ML22157A119

Administered SRO Written Exam

ML22157A119
Person / Time
Site:	Clinton
Issue date:	01/28/2022
From:	Bruce Bartlett NRC/RGN-III/DRS/OLB
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Bartlett B
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CONFIDENTIAL - Exam Material VISION Report (Test Key) ILT 20-1 SRO Exam Approved CPS ILT NRC/Cert Examinations January 27, 2022 Test ILT 20-1 SRO Exam Approved VISION ID 355971 Status The linked image cannot be displayed. The file may have been moved, renamed, or deleted. Verify that the link points to the correct file and location.

NISP-TR-01 Form 7 Revision 3 Page 1 of 1 EXAMINATION COVER SHEET Exam Title (ID)

ILT 20-1 SRO Exam Approved (355971)

Training Program CPS ILT NRC/Cert Examinations LMS Component ID None Total Points 25.00 Pass Criteria = 80 %

Trainee Name Employee ID Graded By / Date Grade

____ / 25.00 = _______ %

Review and Approval Instructor Date Technical Review Date Training Supv Date Examination Rules

1. References may NOT be used during this exam, unless otherwise stated.

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

3. Conversation with other trainees during the exam 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. Restroom trips are limited and only one examinee at a time may leave.

6. For exams with time limits, you have ____ minutes to complete the exam.

7. The examinee agrees to refrain from discussing the content of the exam until the end of the exam cycle.

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 Exam Rules stated above. Further, I have not given, received, or observed any aid or information regarding this exam prior to or during its administration that could compromise this exam.

Examinee 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 exam questions with the instructor to ensure my understanding.

Examinee Signature ___________________________________________ Date ______________

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 3 of 85 Question 1 ID: 2202355 Points: 1.00 CPS was operating at rated thermal power with the following indications:

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 4 of 85 THEN, an event occurred resulting in the following indications:

The action with the highest priority based on completion time must be taken within __________ to meet procedural / tech spec requirements.

A.

15 minutes B.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Answer B

Answer Explanation B is correct:

The indications provided in the stem point to a failed jet pump in RR Loop B, requiring entry into the following LCO's:

ITS 3.4.1 Recirculation Loops Operating - loop flow mismatch is outside the limit of SR 3.4.1.1.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 5 of 85 Per ITS 3.4.3 Jet Pumps - one jet pump riser has failed. SR 3.4.3.1 criteria a and b are both no longer met.

ITS 3.4.1 RA A.1 requires an RR loop to be shutdown within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if RR Loop flow mismatch is not within limits.

ITS 3.4.3 RA A.1 requires the plant to be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if one or more jet pumps are inoperable.

Incorrect responses:

A is incorrect but plausible. This response is plausible because CPS 3005.01 section 6.4.2 Stability Control Concerns - MELLLA Limit, requires action to be commenced within 15 minutes after plant is stable. However, there has not been a MELLLA limit violation.

C is incorrect but plausible. This response is plausible because if either RR FCV has failed, entry into ITS 3.4.2 Flow Control Valves (FCVs) is required, and RA A.1 to lock up the affected FCV within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is appropriate. However, with jet pump flow in the opposite loop unaffected, there is no indication of a failed FCV.

D is incorrect but plausible. This response is plausible because a failed jet pump does require MODE 3 to be entered within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. However, the completion time for ITS 3.4.1 (loop flow mismatch) is more limiting.

Question Information Topic Failed Jet Pump User ID CL-ILT-N20076 System ID 2202355 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.2 Facility operating limitations in the technical specifications and their bases.

References Provided

• ITS 3.4.1 with LCO, applicability statement, and surveillance requirements redacted ITS 3.4.2 with LCO, applicability statement, and surveillance requirements redacted ITS 3.4.3 LCO, applicability statement, and surveillance requirements redacted CPS 3005.01 Fig. 1 Reactor Recirculation PPDS Graphics (Internal)

K/A Justification Question meets the KA because the examinee must use the indications given in the stem to determine that a jet pump has failed.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 6 of 85 SRO-Only Justification This question is linked to SRO-only task 140109.23 (Apply the administrative requirements for execution of Technical Specifications and Off-Site Dose Calculation Manual Requirements). Also linked to 10CFR55.43(b)(2), Facility operating limitations in the Technical Specifications and their bases.

Additional Information Question is high cog, written at the analysis and comprehension level. The examinee must analyze several conditions in the stem to determine a jet pump has failed and then determine a course of action based on that determination (3-SPR).

NRC Exams Only Question Type Bank (CL-ILT-A15092)

Difficulty N/A Technical Reference and Revision #

• ITS 3.4.1 (3.4-1,2 Amend. 171, 3.4-3 Amend. 192)

ITS 3.4.2 (3.4-6 Amend 192)

ITS 3.4.3 (3.4-8 Amend. 95, 3.4-9 Amend. 192)

CPS 3005.01, Rev. 46a Training Objective DB400801.01.08 Given an abnormal Reactor coolant flow condition, describe specific actions for a jet pump failure.

Previous NRC Exam Use None K/A Reference(s) 295001.AA2.04 Safety Function 1 Tier 1 Group 1 RO Imp: 3.0 SRO Imp: 3.1 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR: 41.10 / 43.5 / 45.13)

Individual jet pump flows: Not-BWR-1&2 Learning Objective(s)

Q1/76 295001 A2.04 (BH)

User (Sys) ID N/A (1554371)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 7 of 85 Question 2 ID: 2202460 Points: 1.00 The plant was operating in Mode 1 when a transient occurred resulting in the sequence of events shown below:

Time Event 10:00 Station experiences a loss of ALL AC power.

10:00 MCR begins performing actions per EOP-1 RPV Control and CPS 4200.01 Loss of AC Power.

11:00 Shift Manager declares an Extended Loss of AC Power (ELAP) condition exists.

11:00 MCR enters CPS 4306.01 Extended Loss of AC Power / Loss of Ultimate Heat Sink.

14:00 RPV level is being maintained using the preferred injection source.

15:30 Upper Containment Pools are dumped IAW CPS 4306.01.

Current plant conditions are as follows:

Parameter Value Trend Suppression Pool Level 21' 0" steady Suppression Pool Temperature 163°F rising RPV Pressure 900 psig cycling between 800-1000 psig Which of the following RPV pressure control actions is required?

A.

Control and maintain RPV pressure band of 800 - 1065 psig.

B.

Rapidly depressurize RPV to a pressure band of 150 - 250 psig.

C.

Perform RPV Blowdown to an RPV pressure of less than 38 psig.

D.

Commence cooldown to an RPV pressure band of 150 - 250 psig using SRVs at less than 100°F/hr.

Answer B

Answer Explanation B is correct.

Per CPS 4402.01, EOP-6 Primary Containment Control Heat Capacity Temperature Limit curve, the HCTL is being exceeded with Suppression Pool Temperature at 163°F, Suppression Pool Level at 21" 0",

and RPV pressure cycling between 800-1000 psig. The SP temperature leg of EOP-6 directs performing a blowdown if suppression pool temperature cannot be held below the HCTL limit.

Per CPS 4401.02 EOP-1B Alternate RPV Level/Pressure Control - RPV Pressure leg, if Reactor Core Isolation Cooling (RCIC) is available AND needed for RPV injection and RPV Blowdown required by any EOP, then rapidly depressurize to the Preferred RCIC Control Band (150-250 psig).

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 8 of 85 The EOP/SAG Technical Basis, Vol II discussion for EOP-1B states:

When a blowdown is required, it is generally desirable to fully depressurize the RPV and cool down the plant to cold shutdown conditions. In EOP-1B, however, maintaining adequate core cooling is of higher immediate priority and takes precedence over blowdown requirements in other EOPs.

Since full RPV depressurization may result in loss of RCIC, RPV blowdown must be coordinated with core cooling strategies. Full depressurization is appropriate only if adequate core cooling will not be sacrificed as a result. Loss of adequate core cooling would compound the plant challenges requiring a blowdown and increase any resulting radioactivity release. Core cooling is thus prioritized over other EOP objectives. If a blowdown is required by EOP-6, EOP-8, or EOP-9, RPV pressure is reduced to mitigate the primary containment, secondary containment, or radioactivity release challenge, but the depressurization is limited to the minimum pressure required for RCIC operation.

With RCIC as the only available injection source during a complete loss of AC event, the RPV should not be fully depressurized to avoid loss of injection flow required for adequate core cooling, and should be maintained between 150 to 250 psig as described above.

Incorrect Responses:

A is incorrect but plausible. This response is plausible because the RPV Pressure Control Block of CPS 4306.01, Extended Loss of AC Power / Loss of Ultimate Heat Sink, directs controlling RPV pressure per EOP-1 RPV Control. The pressure leg of EOP-1 directs stabilizing RPV pressure below 1065 psig. OP-CL-101-111-1001 Strategies For Successful Transient Mitigation section 4.1.3.6.B, EOP-1 Pressure Leg, directs providing an initial RPV pressure band of 800-1065 psig. However, the actions required by the EOP-6 Suppression Pool Temperature Leg override the pressure stabilization actions prescribed in CPS 4306.01.

C is incorrect but plausible. EOP-6 Primary Containment Control Suppression Pool Temperature leg requires a blowdown if SP temp cannot be maintained below HCTL. The EOP/SAG Technical Basis, Vol II discussion for EOP-1 states: The RPV is considered depressurized when the RPV pressure is less than the Decay Heat Removal Pressure (38 psig). The Decay Heat Removal Pressure is used as the basis for the depressurized state since, below this pressure, the rate of energy addition to the containment will be within the capacity of the containment vent path. This choice is plausible if the examinee fails to recognize that RCIC is the only available injection source during a complete loss of AC event and that the RPV should not be fully depressurized to avoid loss of injection flow required for adequate core cooling.

D is incorrect but plausible. This response is plausible because the RPV Pressure Control Block of CPS 4306.01 Extended Loss of AC Power / Loss of Ultimate Heat Sink RPV Pressure Control Actions directs performing a cooldown to 150-250 psig using SRVs at < 100°F/hr after pressure is stabilized below 1065 psig and is being controlled per EOP-1 (answer choice A). However, the actions required by the EOP-6 Suppression Pool Temperature Leg override the pressure stabilization actions prescribed in CPS 4306.01.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 9 of 85 Question Information Topic HCTL User ID CL-ILT-N20077 System ID 2202460 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided HCTL K/A Justification Question meets the KA because the examinee must interpret Suppression Pool Temperature, Reactor Pressure, and Suppression Pool Level to determine the correct answer.

SRO-Only Justification Question is linked to SRO Only task 440201.02 Enter and Execute EOP-6 when required, and to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information Question is high cog, written at the analysis and comprehension level. The examinee must analyze the data provided in the stem and then determine required actions based on knowledge of several procedures (3-SPK).

Note that both this question and Q3/78 utilize the HCTL graph as a reference. In this case, the correct answer is obtained by recognizing that HCTL has been exceeded and that a blowdown is necessary. No information from this question could be used to assist in correctly answering Q3/78.

NRC Exams Only Question Type Bank (CL-ILT-N15076)

Difficulty N/A Technical Reference and Revision #

• CPS 4401.01 Rev. 31 CPS 4402.01 Rev. 31 CPS 4200.01 Rev. 26e CPS 4306.01 Rev. 01 EOP/SAG Technical Bases, Rev. 0 OP-CL-101-111-1001 Rev. 17 Training Objective LP87558.01.17 Given a condition resulting in approaching the Heat Capacity Limit, Fig. P, and a diagram of EOP-6, determine when it would be appropriate to Blowdown.

Previous NRC Exam Use ILT 15-1 NRC Exam

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 10 of 85 K/A Reference(s) 295003.AA2.02 Safety Function 6 Tier 1 Group 1 RO Imp: 4.2*

SRO Imp: 4.3*

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER : (CFR: 41.10 / 43.5 / 45.13)

Reactor power / pressure / and level Learning Objective(s)

Q2/77 295003 A2.02 (BH)

User (Sys) ID N/A (1554372)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 11 of 85 Question 3 ID: 2202397 Points: 1.00 With the plant operating at rated thermal power:

At 0105 a valid scram signal was received.

At 0110 the immediate actions for the scram were completed.

Plant conditions at 0112 are as follows:

RPV pressure is cycling between ~930 and 1090 psig.

Suppression Pool temperature is 147°F and slowly rising.

Suppression Pool level is 18' 11" and slowly rising.

Which of the following describes the appropriate actions required for RPV pressure control?

A.

Restore and maintain RPV pressure to below HCTL and STPL limits but above 350 psig. Ok to exceed 100°F per hour.

B.

Restore and hold RPV pressure 350 psig to 1065 psig per the pressure leg of EOP-1A ATWS RPV Control. Hold cooldown rate below 100°F per hour.

C.

Depressurize the RPV using the pressure leg of EOP-1 RPV Control. Hold cooldown rate below 100°F per hour.

D.

Rapidly depressurize the RPV using the Main Turbine Bypass Valves. Ok to exceed 100°F per hour.

Answer A

Answer Explanation A is correct.

Per N-CL-OPS-239001 Main Steam System, after the initial high pressure (> 1103 psig) the Low-Low Set (LLS) function arms, causing the five LLS SRVs to stay open longer, reclosing at a lower pressure. The LLS function also lowers the reopening setpoint for two of the LLS SRVs.

Per CPS 5066.06, Annunciator 6C (Low-Low Setpoint Div 2 Sealed In), the relief setpoints of the following LLS valves are changed to:

VALVE OPEN CLOSE 1B21-F051D 1033 psig 926 psig 1B21-F051C 1073 psig 936 psig 1B21-F047F 1113 psig 946 psig 1B21-F051B 1113 psig 946 psig 1B21-F051G (ADS) 1113 psig 946 psig Based on the pressure response provided in the stem (cycling between ~ 930 and 1090 psig) and since a single SRV can pass ~ 6% steam flow, it can be inferred that reactor power is between 6% and 12%

(1B21-F051D remains open and 1B21-F051C is cycling).

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 12 of 85 Based on reactor power, the SRO would enter EOP-1 RPV Control and transition to EOP-1A ATWS RPV Control (due to the implication that shutdown criteria is not met and/or the reactor is still critical). EOP-1A and EOP-6 Primary Containment Control directs lowering RPV pressure to stay below the HCTL and the STPL (if challenged). Additionally, cooldown exceeding 100°F is allowed.

Incorrect Responses:

B is incorrect but plausible. This answer would be correct if HCTL was not being challenged.

C is incorrect but plausible. This answer would be correct if shutdown criteria was met. As explained above, reactor power can be inferred to be between 6% and 12% based on the pressure indications provided in the stem.

D is incorrect but plausible. This answer would be correct if anticipating blowdown was required for the plant conditions that currently exist. Anticipating blowdown is only authorized in EOP-1 RPV Control when you anticipate reaching a blowdown requirement. EOP-1A ATWS RPV Control directs lowering RPV pressure to stay below the HCTL and the STPL.

Question Information Topic High RPV Pressure User ID CL-ILT-N20078 System ID 2202397 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided HCTL Figure K/A Justification Question meets the KA because the examinee must interpret a high reactor pressure condition and determine the proper EOP mitigation strategy to correct it.

SRO-Only Justification Question is linked to SRO Only task 440201.02 Enter and Execute EOP-6 when required, and to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information This is a high cog question, written at the analysis and comprehension level. The examinee must solve a problem using knowledge and a reference to answer the question correctly (3-SPK/SPR).

Note that both this question and Q2/77 utilize the HCTL graph as a reference. In this case, it is used to eliminate distractor C, and no information from this question could be used to assist in correctly answering Q2/77.

NRC Exams Only

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 13 of 85 Question Type Bank (CL-ILT-A17085)

Difficulty N/A Technical Reference and Revision #

• CPS 4402.01 Rev. 31 CPS 4404.01 Rev. 31 CPS 5066.06(6C) Rev. 25e N-CL-OPS-239001 Rev. 009 Training Objective LP87558.01.08 Given a diagram of EOP-6, explain the use and/or function of the following inserts:.08 Figure P, Heat Capacity Limit Previous NRC Exam Use None K/A Reference(s) 295025.EA2.01 Safety Function 3 Tier 1 Group 1 RO Imp: 4.3*

SRO Imp: 4.3*

Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: (CFR: 41.10 / 43.5

/ 45.13)

Reactor pressure Learning Objective(s)

Q3/78 295025 A2.01 (BH)

User (Sys) ID N/A (1554373)

Cross Reference Links Table: TRAINING - QUESTIONS - Track Questions Modified in this Project (CL-OPS-EXAM-ILT)

Tracking link in project CL-OPS-EXAM-ILT to source question 2202397

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 14 of 85 Question 4 ID: 2202351 Points: 1.00 The plant is operating at rated thermal power.

THEN, a transient occurs requiring entry into EOP-6 Primary Containment Control.

The Suppression Pool level trend is shown on the Plant Process Computer (PPC) screen below:

What is the latest time that dumping upper pools is necessary?

A.

11:01 B.

11:05 C.

11:06 D.

11:09 Answer B

Answer Explanation B is correct.

At 11:05, suppression pool level has reached 15.2 feet. Per CPS 4402.01, EOP-6 Primary Containment Control, if cannot hold suppression pool level above 18 feet 11 inches (18.9 feet), before suppression pool drops to 15 feet 1 inch (15.1 feet), dump upper pools.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 15 of 85 Incorrect Responses:

A is incorrect but plausible. At 11:01, suppression pool level has reached 19.2 feet and is approaching 18.9 feet. Per EOP-6, action is required to maintain suppression pool level above 18.9 feet, including use of the Suppression Pool Makeup and Suppression Pool Cleanup and Transfer systems. However, dumping of upper pools is not required until level approaches 15.1 feet.

C is incorrect but plausible. At 11:06 minutes, suppression pool level has reached 14.2 feet. This response is plausible because a blowdown is required if level cannot be maintained over 15.1 feet, which has occurred at this point. However, dumping of upper pools should have been attempted prior to this point.

D is incorrect but plausible. At 11:09, suppression pool level has reached 11.2 feet and is lowering. Per EOP-6, NPSH/vortex limits have been exceeded and dumping of upper pools should have been attempted prior to this point.

Question Information Topic Low Suppression Pool Level User ID CL-ILT-N20079 System ID 2202351 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the examinee must interpret a graph of suppression level over time and determine action required for a low suppression pool level condition.

SRO-Only Justification Question is linked to SRO Only task 440201. Dump the upper pools as directed by EOP-6, and to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information Question is high cog written at the analysis and application level. The candidate must analyze the parameters in the stem, including a graph, and then determine appropriate actions based on that analysis (3-SPK/SPR).

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• CPS 4402.01, Rev. 31 Training Objective N-CL-OPS-DB-LP87558.01.07Given a decreasing Suppression Pool water level, determine when it is

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 16 of 85 appropriate to dump the Upper Pools in terms of approaching a level of 15 ft 1 in.

Previous NRC Exam Use None K/A Reference(s)

B2.1.25 Safety Function 3 Tier 3 Group RO Imp: 3.9 SRO Imp: 4.2 Ability to interpret reference materials, such as graphs, curves, tables, etc.

(CFR: 41.10 / 43.5 / 45.12)

GS.295030 Safety Function 5 Tier 1 Group 1 RO Imp:

SRO Imp:

Low Suppression Pool Water Level Learning Objective(s)

Q4/79 295030 2.1.25 (NH)

User (Sys) ID N/A (1554374)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 17 of 85 Question 5 ID: 2202503 Points: 1.00 An ATWS is in progress.

Plant conditions are as follows:

ADS has been inhibited.

Reactor Power is 55%.

Reactor Water level is -62".

Reactor Pressure is being controlled by cycling SRVs.

Suppression Pool Level is 115°F.

Drywell pressure is 1.79 psig.

HPCS and RCIC are unavailable.

Given the trends in the table below, the final reactor water level band will be -187" to _______.

Time Drywell Pressure (psig)

Reactor Power (%)

Wide Range RPV water level (inches) 0200:00 1.79 55

-62 0200:10 1.77 51

-69 0200:20 1.73 48

-75 0200:30 1.69 41

-80 0200:40 1.67 36

-86 0200:50 1.64 28

-90 0201:00 1.60 26

-96 0201:10 1.54 18

-101 0201:20 1.48 15

-109 0201:30 1.41 9

-120 0201:40 1.32 6

-132 0201:50 1.22 4

-144 0202:00 1.13 3

-155 0202:10 1.02 2

-162 (TAF) 0202:20 0.94 2

-162 (TAF) 0202:30 0.90 1

-162 (TAF) 0202:40 0.88 1

-162 (TAF)

A.

-60" B.

-86" C.

-144" D.

-162" Answer C

Answer Explanation

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 18 of 85 C is correct. Per CPS 4404.01, EOP-1A ATWS RPV CONTROL, RPV Water Level leg, IF Reactor Power is above 5% or unknown, AND RPV water level is between -162" (TAF) and -60", AND suppression pool temperature is above 110°F AND SRV open or drywell pressure above 1.68 psig: Lower RPV water level to reduce reactor power.

The conditions given in the stem meet all of the above, so water level must be lowered until reactor power is below 5%, OR RPV water level drops to -162" (TAF), OR all SRVs stay closed and drywell pressure stays below 1.68 psig. The earliest time one of these conditions is met occurs at 0201:50, when reactor power goes below 5%. The final level band at L-3 would then be -187" to the final level recorded when lowering water level to reduce reactor power is stopped at 5%, or -144".

Incorrect responses:

A is incorrect but plausible. This response is plausible because injecting water to maintain water level between -187" and Level 8 is appropriate if any of the above conditions for lowering level had not been met, e.g., if suppression pool temperature were less than 110°F or no SRVs open with drywell pressure less than 1.68 psig. In that case, the level control band at L-3 would be -187" to -60".

B is incorrect but plausible. This response is plausible because drywell pressure goes below 1.68 psig at this point, and if lowering water level to reduce reactor power was stopped at this point, the level band at L-3 would be -187" to -86". However, the stem states that SRVs are cycling to control pressure. Only with drywell pressure less than 1.68 and all SRVs staying closed is it appropriate to inject to maintain water level.

D is incorrect but plausible. This response is plausible because level reaches TAF at this point, and it would be appropriate to inject to maintain level if reactor power had stayed above 5%. However, injection to maintain level should have been initiated sooner, when reactor power is below 5% at time 0201:50.

Question Information Topic ATWS Level Bands User ID CL-ILT-N20080 System ID 2202503 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided

• CPS 4404.01 EOP-1A ATWS RPV Control (selected excerpts)

K/A Justification Question meets the KA because the examinee must evaluate the trends in reactor power, reactor pressure, and drywell pressure to determine the correct reactor water level band following an ATWS.

SRO-Only Justification Question is linked to SRO Only task 440401.04 Determine when to initiate RPV injection with ATWS systems for EOP-1A, and to 10CFR55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 19 of 85 emergency situations.

Additional Information Question is high cog, written at the analysis and comprehension level. The examinee must evaluate several conditions in the stem and the trends given to determine when re-injecting to the core is appropriate (3-SPK).

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• CPS 4404.01, Rev. 31 Training Objective N-CL-OPS-DB-LP87553.01.09 Explain why injecting too fast with low level during ATWS conditions may damage the core.

Previous NRC Exam Use None K/A Reference(s)

B2.4.47 Safety Function 5 Tier 3 Group RO Imp: 4.2 SRO Imp: 4.2 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 / 43.5 / 45.12)

GS.295037 Safety Function 1 Tier 1 Group 1 RO Imp:

SRO Imp:

SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Learning Objective(s)

Q5/80 295037 2.4.47 (NH)

User (Sys) ID N/A (1554375)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 20 of 85 Question 6 ID: 2202396 Points: 1.00 Following a radioactivity release, the Operations Support Center (OSC) dispatched field monitoring teams to conduct surveys at the following locations:

Field Team 1 - Meteorological Tower Field Team 2 - NTD building perimeter Field Team 3 - Switchyard Which of the field teams are OFF-SITE for the purposes of Monitoring Teams?

A.

1 and 2 ONLY B.

2 and 3 ONLY C.

1 and 3 ONLY D.

1, 2, and 3 Answer D

Answer Explanation D is correct.

Per EP-AA-112-300-F-20, Onsite Monitoring Team Checklist, section 2.1, "Onsite" is defined as all areas inside the Protected Area Fence. Offsite Monitoring Teams will normally survey outside the fence. EP-AA-1003, Exelon Nuclear Radiological Emergency Plan Annex for Clinton Station, Figure 4-1 shows the Protected Area as encompassing the plant itself and the screenhouse. Areas outside of the Protected Area include the switchyard, the parking lots, and the NTD. The Meteorological Tower, NTD building, and switchyard are all "offsite" for Monitoring Team purposes.

Incorrect Responses:

A is incorrect but plausible. Field Teams 1 and 2 are considered "offsite," but Field Team 3 is as well.

This answer may be selected since the Switchyard is in the Owner Controlled Area and is within the Exelon property boundary.

B is incorrect but plausible. Field Teams 2 and 3 are considered "offsite," but Field Team 1 is as well.

This answer may be selected since the Meteorological Tower is in the contractor parking lot and is within the Exelon property boundary.

C is incorrect but plausible. Field Teams 1 and 3 are considered "offsite," but Field Team 2 is as well.

This answer may be selected since the NTD contains the Technical Support Center and is within the Exelon property boundary.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 21 of 85 Question Information Topic Radiological Release User ID CL-ILT-N20081 System ID 2202396 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the examinee must determine which teams are off-site during a radiological release to correctly answer the question.

SRO-Only Justification Question is linked to SRO-only task 997777.03, Emergency Plan Activities performed by an SRO and to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information Question is low cog, written at the memory level. The examinee has to recall facts from a procedure to answer the question (1-F/1-P).

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• EP-AA-1003, Rev. 29 EP-AA-112-300-F-20, Rev. A Training Objective 502041 Discuss the Defined Terms associated with the EALs and Basis document Previous NRC Exam Use None K/A Reference(s) 295038.EA2.01 Safety Function 9 Tier 1 Group 1 RO Imp: 3.3*

SRO Imp: 4.3*

Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE : (CFR: 41.10 /

43.5 / 45.13)

Off-site Learning Objective(s)

Q6/81 295038 A2.01 (NL)

User (Sys) ID N/A (1554376)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 22 of 85

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 23 of 85 Question 7 ID: 2202525 Points: 1.00 The unit is at rated thermal power with:

1DG04TA DG 1A Air Receiver 'A' tagged out of service, and 1DG04TB DG 1A Air Receiver 'B' in service and pressurized to 230 psig.

A grid disturbance causes 4160V Bus 1A1 voltage to lower.

A short time later, annunciator 5060-3D AC UNDERVOLTAGE SECOND LEVEL 4160V BUS, is received.

Div 1 DG is...

A.

operable AND supplying 4160V Bus 1A1.

B.

inoperable but available AND supplying 4160V Bus 1A1.

C.

inoperable but available AND NOT supplying 4160V Bus 1A1.

D.

inoperable AND unavailable to restore 4160V Bus 1A1 voltage.

Answer A

Answer Explanation A is correct:

ITS B3.8.3 states that a single air receiver with pressure above 200 psig is designed to provide multiple starts of its associated DG. This is verified in 9080.01 Quick Start section by valving out one of the air start subsystems and verifying that the DG reaches rated voltage and frequency within the required time.

CPS 5060.03(3D) AC UNDERVOLTAGE SECOND LEVEL 4160V BUS is alarmed when 4160V Bus 1A1 experiences a degraded voltage < 4072 VAC (nominal). Under these conditions, the Div 1 DG will start, the Div 1 Reserve (Main) feeder will open and the Div 1 Main (Reserve) feeder will lock-out; the 4.16KV Bus 1A1 will be stripped of its loads; and the Div 1 DG will tie onto the bus.

Incorrect answers:

B is incorrect but plausible because for most systems, operability requires that all support systems be aligned and able to support the safety function of the supported system. However, both air start subsystems are not required for Diesel Generator operability.

C is incorrect but plausible. The first part of this response is plausible because for most systems, operability requires that all support systems be aligned and able to support the safety function of the supported system. However, both air start subsystems are not required for Diesel Generator operability.

Additionally, the second part of this response is plausible because a valid 1st level undervoltage condition would not cause the Div 1 DG to start and tie onto the bus if the reserve source is available; however, stem conditions indicate a 2nd level undervoltage condition, which does cause the DG to automatically energize the bus.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 24 of 85 D is incorrect but plausible because for most systems, operability and availability require that all support systems be aligned and able to support the safety function of the supported system. However, both air start subsystems are not required for Diesel Generator operability and availability.

Question Information Topic Electrical Grid Disturbance User ID CL-ILT-N20082 System ID 2202525 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.2 Facility operating limitations in the technical specifications and their bases.

References Provided None K/A Justification This question meets the KA because the examinee has to determine the operational/available status of the Div 1 DG during an Electric Grid Disturbance event.

SRO-Only Justification This question is linked to 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Additional Information This is a high cog question written at the analysis and comprehension level. The examinee has to analyze the conditions in the stem, and then predict expected outcomes and determine required actions based on that analysis. (3-PEO)

NRC Exams Only Question Type Bank (CL-ILT-A17082)

Difficulty N/A Technical Reference and Revision #

• CPS 5060.03 (3D) Rev. 30b ITS B3.8.3 Rev. 3-2 CPS 9080.01 Rev. 57 Training Objective

• 262001.06 Given an Auxiliary Power System Annunciator, DESCRIBE: a. The condition causing the annunciator b. Any automatic actions c. Any operational implications 264000.14 Given DIESEL GENERATOR/DIESEL FUEL OIL System operability status and a copy of Tech Specs, DISCUSS the bases for the DIESEL GENERATOR/DIESEL FUEL OIL System Tech Spec LCO, related safety limits and Limiting Safety System Settings.

Previous NRC Exam Use None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 25 of 85 K/A Reference(s)

B2.2.37 Safety Function 9 Tier 3 Group RO Imp: 3.6 SRO Imp: 4.6 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12)

GS.700000 Safety Function 9 Tier Group RO Imp:

SRO Imp:

Generator Voltage and Electric Grid Disturbances Learning Objective(s)

Q7/82 700000 2.2.37 (BH)

User (Sys) ID N/A (1554377)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 26 of 85 Question 8 ID: 2202526 Points: 1.00 0000 - Plant conditions are as follows:

97% Reactor Power.

LPCS is out of service for maintenance.

0100 - the plant experiences a LOCA. The following plant conditions exist:

All Control Rods are fully inserted.

Reactor Pressure is 400 psig.

Reactor Water Level is -90 inches on Wide Range and slowly rising.

The Reserve Auxiliary Transformer has locked out.

Div 1 AND Div 2 DGs are running; Div 3 DG failed to start.

RCIC is injecting.

RHR 'A' is aligned for LPCI injection.

RHR 'B' is aligned for Suppression Pool Cooling.

Suppression Pool Temperature is 102°F and slowly lowering.

Containment Pressure is 3.0 psig and slowly rising.

Containment Temperature is 180°F and slowly rising.

0102 - RHR 'A' pump trips.

Which action(s) should be taken NEXT?

A.

Maintain Suppression Pool Cooling.

B.

Initiate Feedwater Leakage Control System (FWLCS).

C.

Shutdown Suppression Pool Cooling and Initiate CNMT Spray.

D.

Shutdown Suppression Pool Cooling and Initiate LPCI Injection.

Answer C

Answer Explanation C is correct:

Based on plant conditions provided in the question stem, actions in EOP-1 RPV CONTROL and multiple legs of EOP-6 PRIMARY CONTAINMENT CONTROL are applicable and should be executed as follows:

EOP-6 Suppression Pool Temperature Leg - with Suppression Pool Temperature > 95°F - EOP-6 requires all available pool cooling to be started. This action is modified by a note that states, "Do not use RHR pumps you need for core cooling or containment spray".

EOP-6 Containment Temperature and Drywell/Containment Pressure legs - with containment pressure vs. containment temperature in the 'OK to SPRAY' region of Figure O - EOP-6 directs starting containment sprays. This action is modified by a note that states, "Do not use RHR pumps you need for core cooling".

EOP-1 RPV Water Level Leg - requires using preferred injection systems and alternate injection systems (if needed) to hold RPV Level above TAF (-162").

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 27 of 85 With Containment Pressure in the "OK to Spray" region of the Containment Spray Initiation Limit (CSIL) figure, RHR 'B' should be diverted from Suppression Pool Cooling to Containment Spray based on the actions directed in the Containment Temperature leg and the note in the Suppression Pool Temperature leg discussed above.

In addition, with RPV pressure at 400 psig (greater than the discharge head of RHR Pump 'A'), and RPV level at -90 inches and rising, it can be determined that RHR 'A' is not needed for core cooling, and could have been realigned to the Containment Spray mode (had it not tripped).

Therefore, the correct actions to be taken NEXT would be to shutdown suppression pool cooling and initiate CNMT spray.

Incorrect Responses:

A is incorrect but plausible. This response is plausible because maintaining suppression pool cooling would be appropriate if containment spray initiation was not required. However, with Suppression Pool Temperature > 95°F - EOP-6 requires all available pool cooling to be started.

B is incorrect but plausible. This response is plausible because starting the Feedwater Leakage Control System would be appropriate if the entry conditions for CPS 4001.02 AUTOMATIC ISOLATION have been met. However, the Feedwater Leakage Control System is required to be initiated when a DBA LOCA signature has been identified (RPV level below Level 1 (-145.5") and High DW pressure (>1.68 psig) with RPV pressure < 330 psig. Since these conditions have not been met, initiating the Feedwater Leakage Control System should not be performed.

D is incorrect but plausible. This response is plausible due to a common misconception that RHR 'B' is required to be used for LPCI injection. However, OP-CL-101-111-1001 STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION, Section 6.C.3 EOP-1 RPV Control - Alignment of Systems Needed For Adequate Core Cooling (ACC) provides additional guidance on the use of RPV injection sources. It states, "A trigger point of -100 inches and lowering is recommended for evaluating the need to re-align injection systems and/or initiate/maintain containment sprays if those systems will be needed to restore and hold ACC." With RPV pressure above the discharge head of RHR Pump 'B' and RPV level at -90" and rising, RHR 'B' is not needed for core cooling.

Question Information Topic High Containment Temperature User ID CL-ILT-N20083 System ID 2202526 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided CISL K/A Justification Question meets the KA because the examinee must interpret containment pressure as it applies to a high containment temperature and determine the need for

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 28 of 85 containment sprays based of the given plant conditions.

SRO-Only Justification Question is linked to SRO-only task 440201.03 Determine when Containment Sprays are required, when executing EOPs, and to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information Question is high cog, written at the analysis and comprehension level. The examinee must analyze the data provided in the stem and then determine required actions based on knowledge of several procedures (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-N14088)

Difficulty N/A Technical Reference and Revision #

• CPS 4001.02 Rev. 17f CPS 4401.01 Rev. 31 CPS 4402.01 Rev. 31 OP-CL-101-111-1001 Rev. 16 Training Objective LP87558.01.08, N-CL-OPS-DB-LP87558.01.08 Given a diagram of EOP-6, explain the use and/or function of the following inserts:

.07 Figure O, Containment Spray Initiation Limit Previous NRC Exam Use ILT14-1 NRC Exam K/A Reference(s) 295011.AA2.02 Safety Function 5 Tier 1 Group 2 RO Imp: 4.0*

SRO Imp: 4.1*

Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY) : (CFR: 41.10 / 43.5 / 45.13)

Containment pressure: Mark-III Learning Objective(s)

Q8/83 295011 A2.02 (BH)

User (Sys) ID N/A (1554378)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 29 of 85 Question 9 ID: 2228987 Points: 1.00 CPS is at rated thermal power.

THEN, an unisolable main steam line leak into the Auxiliary Building Steam Tunnel occurs, resulting in a Loss of the Primary Containment.

From AR/PR:

From EP-AA-1003, Addendum 3:

Radiological Effluents General Emergency RG1 Site Area Emergency RS1 Alert RA1 Unusual Event RU1 HVAC + SGTS Rad Monitors > 8.91 E+03 Ci/sec for 15 minutes HVAC + SGTS Rad Monitors > 8.91 E+02 Ci/sec for 15 minutes HVAC + SGTS Rad Monitors > 8.91 E+01 Ci/sec for 15 minutes HVAC + SGTS Rad Monitors > 1.02 Ci/sec for 60 minutes

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 30 of 85 (1) What is the lowest sum of HVAC and SGTS Radiation Monitor readings which would require the Shift Emergency Director to recommend evacuation of Sub Areas per EP-AA-111-F-07 Clinton PAR Flowchart during the initial PAR call?

(2) Which Sub Areas should be recommended for evacuation?

A.

(1) 8.92 E+02 Ci/sec for > 15 minutes (2) 1, 2, and 3 B.

(1) 8.92 E+02 Ci/sec for > 15 minutes (2) 1, 5, 6, and 7 C.

(1) 8.92 E+03 Ci/sec for > 15 minutes (2) 1, 2, and 3 D.

(1) 8.92 E+03 Ci/sec for > 15 minutes (2) 1, 5, 6, and 7 Answer C

Answer Explanation C is correct.

PARs are recommended at the General Emergency Level.

Per EP-AA-1003, Addendum 3 excerpt given in stem, RG1 is met at a combined HVAC and SGTS Radiation Monitor reading of 8.91 E+03 Ci/sec for > 15 minutes.

The stem states that the Primary Containment barrier is lost.

The above constitutes a Rapidly Progressing Severe Accident per the Clinton PAR Flowchart. Using Table 1 of the Clinton PAR Flowchart, the candidate will determine that with winds from 232 degrees, evacuation of Sub Areas 1, 2, and 3 is required.

Incorrect Responses:

A is incorrect but plausible. This response is plausible because a a combined HVAC and SGTS Radiation Monitor reading of 8.91 E+02 Ci/sec for > 15 minutes meets conditions for a Site Area Emergency. However, PARs are not recommended until the threshold for a General Emergency are met.

The second part of the response is correct.

B is incorrect but plausible. This response is plausible because a a combined HVAC and SGTS Radiation Monitor reading of 8.91 E+02 Ci/sec for > 15 minutes meets conditions for a Site Area Emergency. However, PARs are not recommended until the threshold for a General Emergency are met.Additionally, the second part is plausible if the wind direction is incorrectly interpreted as travelling to 232 degrees (or coming from 52 degrees) instead of the correct coming from 232 degrees. This is a misconception that has occurred in past emergency preparedness drills.

D is incorrect but plausible. The first part of this response is correct. The second part is plausible if the wind direction is incorrectly intepreted as traveling to 232 degrees (or coming from 53 degrees) instead of the correct coming from 232 degrees. This is a misconception that has occurred in past emergency preparedness drills.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 31 of 85 Question Information Topic Met Tower PARS Data User ID CL-ILT-N20084 System ID 2228987 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.4 Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

References Provided

• EP-AA-111-F-07 Clinton PAR Flowchart, page 1 with references to General Emergency redacted.

K/A Justification Question meets the KA because the examinee must interpret the meteorological data given to determine which areas require evacuation during an off-site radioactivity release.

SRO-Only Justification This question is linked to SRO-only task 997777.02 Given a postulated E-Plan condition, determine and recommend Offsite Protective actions (PARs), IAW corporate EP, and station specific EP procedures and to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information Question is written at the analysis/comprehension level (high cog). The examinee has to analyze conditions in the stem and solve a problem using knowledge to answer the question (3-SPK).

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• EP-AA-1003, Rev. 29 EP-AA-111-F-07, Rev. H Training Objective LP87537.01.10 Given section 3 of EP-AA-1003, Radiological Emergency Plan Annex For Clinton Station, and plant parameters indicative of one or more of the following events, properly classify the emergency.

.01 Fission Product Boundary Failure

.02 Fuel Damage/Degraded Core

.03 Radiological Emergency

.04 Abnormal Reactor Coolant Leaks, Temperatures and/or Pressures

.05 Steam Line Breaks/Safety Relief Valve Failure Previous NRC Exam Use None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 32 of 85 K/A Reference(s) 295017.AA2.05 Safety Function 9 Tier 1 Group 2 RO Imp: 2.5 SRO Imp: 3.8 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE : (CFR: 41.10 /

43.5 / 45.13)

Meteorological data Learning Objective(s)

Q9/84 295017 A2.05 (NH)

User (Sys) ID N/A (1554379)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 33 of 85 Question 10 ID: 2203811 Points: 1.00 The plant is operating at rated power when a transient results in a steam leak into Secondary Containment.

Plant conditions are currently as follows:

Reactor power is 65%, stable.

Point 8 on recorder 1TR-CM326 indicates 120°F, rising slowly.

Secondary Containment Differential Pressure is +0.00 inches and rising slowly.

Rising Fuel Building radiation levels are hampering efforts to isolate the steam leak.

Fuel Building Exhaust radiation monitors, 1RIX-PR006A, B, C, AND D are reading 11 mr/hr and rising slowly.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 34 of 85

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 35 of 85 What is the NEXT required action?

A.

Scram the Reactor and enter EOP-1, RPV CONTROL.

B.

Defeat VF Interlocks (CPS 4410.00C011) and restart VF (CPS 3404.01)

C.

Shutdown the Reactor in accordance with CPS 3006.01, UNIT SHUTDOWN.

D.

Isolate all discharges into affected areas EXCEPT systems needed for firefighting or EOP actions.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 36 of 85 Answer B

Answer Explanation B is correct.

Per EOP-8 Secondary Containment Control:

Entry is required based on Fuel Building Ventilation (VF) radiation levels being above 10 mr/hr.

VF will isolate and Standby Gas Treatment (VG) will start.

IF VG cannot restore and hold Secondary Containment Differential Pressure below 0 in., AND radiation levels may interfere with operation of systems needed for damage control or EOP actions, THEN restart VF. (OK to defeat VF interlocks - 4410.00C011)

Per EOP/SAG Technical Bases, Vol. II:

Under beyond-design basis conditions, such as extended loss of spent fuel pool cooling resulting in addition of large amounts of water vapor into the secondary containment atmosphere, SGTS may be incapable of maintaining the desired differential pressure. If a negative secondary containment pressure cannot be maintained, gaseous radioactivity may leak into the surrounding environment since the secondary containment is not an airtight structure.

This release could restrict access to the secondary containment or surrounding areas, interfering with damage control activities or the performance of other steps in the EOPs. Under such conditions, Fuel Building Ventilation is restarted to preserve secondary containment accessibility and facilitate damage control activities, thereby reducing risks to personnel, plant structures, and essential equipment.

Incorrect Responses:

A is incorrect but plausible. Scramming the reactor is appropriate if an unisolable primary system leak into the secondary containment exists. However, the RCIC Room Ambient Temp is below the Group 5 &

6 isolation setpoint on RCIC Room Ambient Temp (192°F). Therefore, an isolation has not yet been attempted. EOP-8 only requires a scram if a primary system is discharging into the secondary containment AND is unisolable, before temperatures reach max safe values.

C is incorrect but plausible because EOP-8 requires a reactor shutdown if 2 or more areas exceed max safe temperatures. However, the conditions in the stem do not indicate 2 or more areas above max safe of the same parameter.

D is incorrect but plausible. This response is plausible because the temperature leg of EOP-8 Secondary Containment Control directs isolating all discharges into affected areas except for systems needed for firefighting and EOP actions if any Table T temperature is above max normal. Although Point 8 is above the alarm setpoint, it is not yet above max normal. Additionally, VF should be restarted to preserve secondary containment accessibility and facilitate damage control activities.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 37 of 85 Question Information Topic EOP-8 User ID CL-ILT-N20085 System ID 2203811 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided CPS 4406.01 - EOP-8 Tables T/U (Internal)

K/A Justification Question meets the KA because the examinee must demonstrate knowledge of VF/VG system setpoints, interlocks and automatic actions associated with EOP-8.

Additionally, the candidate must recognize how a Secondary Containment high differential pressure is mitigated.

SRO-Only Justification Question is linked to SRO-only task 440601.02 Respond to a Secondary Containment Control Emergency per EOP-8 and 10CFR55.43(b)(5), Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information Question is high cog, written at the analysis and application level. The candidate must analyze the conditions in the stem and then determine appropriate actions based on that analysis (3-SPK).

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• CPS 4001.02C001 Rev. 16d CPS 4406.01 Rev. 31 CPS 5140.63 Rev. 1d EOP/SAG Technical Bases, Vol II, Rev. 0 Training Objective 290001.16EVALUATE the following FUEL BUILDING HVAC AND SECONDARY CONTAINMENT indications/responses and DETERMINE if the indication/

response is expected and normal.

.1 VF System Isolation and VG Actuation..2 Secondary Containment Diff Press Control.

Previous NRC Exam Use None K/A Reference(s)

B2.4.02 Safety Function 9 Tier 3 Group RO Imp: 4.5 SRO Imp: 4.6 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

(CFR: 41.7 / 45.7 / 45.8)

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 38 of 85 GS.295035 Safety Function 5 Tier 1 Group 2 RO Imp:

SRO Imp:

Secondary Containment High Differential Pressure Learning Objective(s)

Q10/85 295035 2.4.02 (NH)

User (Sys) ID N/A (1554380)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 39 of 85 Question 11 ID: 2202535 Points: 1.00 CPS is at rated thermal power.

THEN, an inadvertent Group 1 isolation occurred.

Very little rod motion occurred on the subsequent scram.

Neither manual scram nor ARI were successful.

The crew initiated SLC at 12:05, but one pump tripped immediately upon starting.

Which of the following is the FIRST instance in which RPV depressurization below 350 psig is permitted?

A.

IRMs < range 7 at 12:20.

B.

At time 12:28.

C.

At time 12:50.

D.

All rods in at time 12:55.

Answer C

Answer Explanation C is correct. Per the Clinton Power Station EOP/SAG Technical Bases for EOP-1A:

Under ATWS conditions, RPV pressure is preferentially maintained above the Minimum ATWS RPV Pressure (350 psig) until it is determined that the reactor will remain shutdown, even if a blowdown is required.

When the reactor is shutdown with no boron injection or injection of Cold Shutdown Boron is completed the risk of unstable control is eliminated and maintaining an elevated pressure is no longer warranted. The blowdown sequence then continues at #16 where full depressurization is performed using SRVs to mitigate the plant condition that required a blowdown.

The CPS 4404.01 EOP-1A ATWS RPV Control hold point in the Pressure Leg requires that prior to commencing a cooldown either cold shutdown boron weight be injected or the reactor is subcritical with no boron injected. Per Detail X, Cold Shutdown Boron Weight has been injected when one SLC pump has been running for 45 minutes.

Incorrect Answers:

A is incorrect but plausible. This response is plausible because it would be correct if no boron had been injected. With one SLC pump running, this criteria cannot be used.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 40 of 85 B is incorrect but plausible. This response is plausible because if both SLC pumps were running, Cold Shutdown Boron Weight would have been injected at 12:28 (23 minutes following starting SLC pumps) per Detail X. However, the stem indicates that one SLC pump tripped immediately.

D is incorrect but plausible. This response is plausible because all rods in meets the hold point for further depressurization; however, this is not the FIRST opportunity the crew must depressurize the reactor.

Question Information Topic SLC Pump Trip User ID CL-ILT-N20086 System ID 2202535 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the examinee must determine the effect of an SLC pump trip on the ability to execute EOP-1A and determine when to depressurize to

< 350 psig following the pump trip.

SRO-Only Justification Although it is RO knowledge to know that SLC pumps are run procedurally for 23 minutes (2 pumps) or 45 minutes (1 pump) to ensure adequate boron, it is SRO knowledge of the EOP-1A hold point (procedure content), subsequent assessment of plant conditions and determination of when to proceed with depressurization to < 350 psig that makes this question SRO only. Linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information Question is high cog, written at the analysis/comprehension level. The examinee must analyze the plant conditions presented in the stem and determine the correct actions to take based on a selection of abnormal/support procedures to be performed in conjunction with the EOP. (3-SPK)

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• CPS 4404.01, Rev. 31 EOP/SAG Technical Bases, Rev. 0 Training Objective N-CP-OPS-DB-LP87553.01.04 State the boron injection methods and flowpaths which can be used to respond to an ATWS.

Previous NRC Exam Use None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 41 of 85 K/A Reference(s) 211000.A2.01 Safety Function 1 Tier 2 Group 1 RO Imp: 3.5 SRO Imp: 3.8*

Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

Pump trip Learning Objective(s)

Q11/86 211000 A2.01 (NH)

User (Sys) ID N/A (1554381)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 42 of 85 Question 12 ID: 2202555 Points: 1.00 The plant is operating at 100% rated thermal power.

Which of the following scenarios would require entry into ITS 3.3.1.1 RPS Instrumentation Condition A (one or more Functions with one channel inoperable)?

Scenario Event 1

16 of the 33 LPRMs on APRM 'A' are failed downscale.

2 It is discovered that Reactor Recirculation Pump 'A' breakers 3A and 4A are incapable of opening.

3 During performance of a Turbine Stop Valve Closure Channel Functional Surveillance, one of the instrument channels tripped outside the allowable value listed in ITS.

4 During performance of Scram Discharge Volume Vent/Drain valve stroke time testing, 1C11-F010 Scram Discharge Volume Vent Valve closed, but exceeded the section 9.1 Acceptance Criteria time limit listed in the surveillance.

Scenario...

A.

1 B.

2 C.

3 D.

4 Answer C

Answer Explanation C is correct.

Per ITS B3.3.1.1 RPS Instrumentation, The OPERABILITY of the RPS is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.1.1-1.

Each Function must have four OPERABLE channels, with their setpoints within the specified Allowable Value, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.

Allowable Values are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 43 of 85 Incorrect Responses A is incorrect but plausible due to the large number of LPRMs that are failed. Per ITS B 3.3.1.1 Functions 2.a and 2.b, the APRM System is composed of four channels, each providing an input to each of the four RPS trip logic divisions. All four Average Power Range Monitor Neutron Flux-High, Setdown and Average Power Range Monitor Flow Biased Simulated, Thermal Power-High channels are required to be OPERABLE to ensure that no single failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 16 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located. The stem states that 16 of the 33 LPRMs on APRM 'A' are failed, which leaves 17 operable, which is greater than the number of LPRMs required to consider APRM 'A' OPERABLE. This distracter would be correct if 2 additional LPRMs were failed.

B is incorrect but plausible. RR Pump 'A' 3A and 4A breakers will trip if an EOC/RPT signal is received.

The EOC/RPT signal is generated from a Turbine Stop Valve or Turbine Control Valve closure signal, which will generate an RPS trip signal (ITS 3.3.1.1 RPS Instrumentation Functions 9 and 10). The trip of the RR Pumps, however, is not an RPS function, so failure of the RR 3A and 4A breakers would not require entry into ITS 3.3.1.1. The failure would require entry into ITS 3.3.4.1 EOC-RPT Instrumentation.

D is incorrect but plausible. The Scram Discharge Volume Vent and Drain Valves automatically close on an RPS initiation and are stroke time tested per CPS 9031.15 Reactor Mode Switch Functional Test. Per CPS 9031.15, the operability requirements for the SDV V&D Valves are tied to ITS SR 3.1.8.3 Scram Discharge Volume Vent and Drain Valves and is based on limiting the amount of reactor coolant discharged during a scram. The valves are required to close to satisfy the requirements of ITS 3.3.1.1, but the stroke time requirement only applies to ITS 3.1.8.

Question Information Topic RPS Instrumentation User ID CL-ILT-N20087 System ID 2202555 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.2 Facility operating limitations in the technical specifications and their bases.

References Provided None K/A Justification This question meets the KA because the examinee must demonstrate knowledge of conditions and limitations in the facility license for the RPS system to answer the question correctly.

SRO-Only Justification Question is linked to SRO only task 140109.23 Apply The Administrative Requirements For Execution Of Technical Specifications Or Off-Site Dose Calculation Manual Requirements and 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases. This question also requires knowledge of TS bases to analyze TS required actions and terminology.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 44 of 85 Additional Information This is a high cog question written at the analysis and application level. The examinee has to recall several pieces of information and then compare that with knowledge of the bases for RPS instrumentation technical specifications to answer the question (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-N17086)

Difficulty N/A Technical Reference and Revision #

• CPS 5005.01 (1H) Rev. 28e CPS 9031.15 Rev. 27c ITS B3.3.1.1 (page B3.3-3) Revision No. 10-6 ITS B3.3.1.1 (page B3.3-7) Revision No. 7-5 ITS B3.3.1.1 (page B3.3-8) Revision No. 0 ITS B3.3.4.1 (page B3.3-66) Revision No. 7-5 Training Objective 212000.14 Given Reactor Protection System (RPS) and Alternate Rod Insertion (ARI) System operability status and a copy of Tech Specs, DISCUSS the bases for the Reactor Protection System (RPS) and Alternate Rod Insertion (ARI) System Tech Spec LCO, related safety limits and Limiting Safety System Settings.

Previous NRC Exam Use ILT 17-1 NRC Exam K/A Reference(s)

GS.212000 Safety Function 7 Tier 2 Group 1 RO Imp:

SRO Imp:

Reactor Protection System B2.2.38 Safety Function 7 Tier 3 Group RO Imp: 3.6 SRO Imp: 4.5 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13)

Learning Objective(s)

Q12/87 212000 2.2.38 (BH)

User (Sys) ID N/A (1554382)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 45 of 85 Question 13 ID: 2202504 Points: 1.00 CPS is operating at rated thermal power. Numerous LPRM inputs are bypassed (marked with an asterisk and shaded) due to detector faults. All other LPRMs are working properly.

Div 1 APRM Channel A LPRM Detectors06-23A 06-15B 14-15C 14-07D 06-39A 06-31B 14-31C

• 14-23D 22-07A 22-15B

• 14-47C 14-39D 22-23A 22-31B 30-15C 30-07D 22-39A 22-47B 30-31C 30-23D 38-07A 38-15B

• 30-47C 30-39D

• 38-23A 38-31B 46-15C

• 46-23D 38-39A 38-47B 46-31C 46-39D

• 47-47C

• Div 2 APRM Channel B LPRM Detectors14-07A

• 06-23B

• 06-15C 14-15D

• 14-23A

• 06-39B 06-31C 14-31D 14-39A 22-07B 22-15C 14-47D

• 30-07A 22-23B

• 22-31C 30-15D 30-23A

• 22-39B 22-47C 30-31D

• 30-39A 38-07B

• 38-15C

• 30-47D 46-23A

• 38-23B

• 38-31C 46-15D 46-39A 38-39B 38-47C

• 46-31D

• 46-47D Div 3 APRM Channel C LPRM Detectors14-15A 14-07B 06-23C 06-15D

• 14-31A 14-23B 06-39C 06-31D

• 14-47A 14-39B

• 22-07C 22-15D

• 30-15A 30-07B 22-23C 22-31D

• 30-31A

• 30-23B 22-39C 22-47D

• 30-47A

• 30-39B 38-07C 38-15D

• 46-15A 46-23B 38-23C 38-31D 46-31A 46-39B 38-39C 38-47D

• 46-47A Div 4 APRM Channel D LPRM Detectors06-15A

• 14-15B

• 14-07C

• 06-23D 06-31A

• 14-31B

• 14-23C 06-39D 22-15A 14-47B 14-39C

• 22-07D 22-31A

• 30-15B 30-07C 22-23D

• 22-47A

• 30-31B 30-23C 22-39D 38-15A 30-47B

• 30-39C 38-07D 38-31A

• 46-15B 46-23C 38-23D 38-47A

• 46-31B

• 46-39C 38-39D 46-47B

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 46 of 85 THEN, the channel 'A' APRM power supply fails low.

No operator actions have been taken.

(1) How many APRM channels must be declared inoperable?

(2) What actions are necessary to mitigate the plant impact?

A.

(1) 1 (2) Perform immediate actions of CPS 4100.01, Reactor Scram.

B.

(1) 2 (2) Perform immediate actions of CPS 4100.01, Reactor Scram.

C.

(1) 1 (2) Place the Division 1 Sensor Bypass Switch to the BYPASS position ONLY per CPS 3305.01, Reactor Protective System.

D.

(1) 2 (2) Place the Division 1 Sensor Bypass Switch to the BYPASS position ONLY per CPS 3305.01, Reactor Protective System.

Answer D

Answer Explanation D is correct. Per CPS 5004.01 ALARM PANEL ANNUNCIATORS - ROW 1, APRM A UPSC TRIP OR INOP, when the APRM "A" fails low it causes an INOP annunciator for that channel and will generate a rod withdrawal block. Although APRM channel 'C' is inoperable per ITS, having less than two operable LPRM inputs per level will not activate an APRM INOP trip per CPS 3308.01 LOCAL/AVG POWER RANGE MONITORS (L/APRM), section 6.3. The required action to mitigate the rod withdrawal block is to bypass the 'A' APRM by placing the Division 1 Sensor Bypass control switch to BYPASS per CPS 3305.01 REACTOR PROTECTIVE SYSTEM.

Incorrect responses:

A is incorrect but plausible. This response may be selected based on a misconception that the low input voltage to the 'A' APRM channel will not cause it to become inoperable and/or generate a rod block or failure to recognize that APRM 'C' is also inoperable based on having less than two operable LPRM inputs per level. Additionally, per CPS 5004.01(H), one APRM INOP signal will cause a rod withdrawal block - scramming the reactor is a plausible response based on confusion of actions required per the ARP.

B is incorrect but plausible. The first part of the response is correct. However, a reactor scram is not required. This answer may be selected because two APRM INOP signals will cause a reactor scram but having less than two operable LPRM inputs per level will not activate an APRM INOP trip.

C is incorrect but plausible. This response may be selected based on a misconception that the low input voltage to the 'A' APRM channel will not cause it to become inoperable and/or generate a rod block or failure to recognize that APRM 'C' is also inoperable based on having less than two operable LPRM inputs per level. The second part of the response is correct.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 47 of 85 Question Information Topic Loss of APRM Power User ID CL-ILT-N20088 System ID 2202504 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the examinee must understand the impact of a loss of power to the APRM and then determine mitigating actions required as a result of the loss of power.

SRO-Only Justification Question is linked to SRO only task 140109.23 Apply The Administrative Requirements For Execution Of Technical Specifications Or Off-Site Dose Calculation Manual Requirements and 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Additional Information Question is high cog, written at the analysis and comprehension level. The examinee must determine based on the conditions given in the stem that one APRM channel has an INOP trip and then determine how to mitigate the condition (3-SPK).

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• ITS B3.3.1.1 (B3.3-7 Rev. 7-5)

CPS 5004.01, Rev. 28e CPS 3308.01, Rev. 11f Training Objective 215005.14 Given AVERAGE POWER RANGE MONITOR System operability status and a copy of Tech Specs, DISCUSS the bases for the AVERAGE POWER RANGE MONITOR System Tech Spec LCO, related safety limits and Limiting Safety System Settings.

Previous NRC Exam Use None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 48 of 85 K/A Reference(s) 215005.A2.01 Safety Function 7 Tier 2 Group 1 RO Imp: 2.7 SRO Imp: 3.1 Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

Power supply degraded Learning Objective(s)

Q13/88 215005 A2.01 (NH)

User (Sys) ID N/A (1554383)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 49 of 85 Question 14 ID: 2202616 Points: 1.00 The plant is operating at rated thermal power. CPS 9381.01 MOV Thermal Overload Bypass Verification is in progress on the 1E22-F004 HPCS To CNMT Outboard Isolation Valve following maintenance.

Currently:

HPCS OUT OF SERVICE switch is in NORMAL.

HPCS MOV TEST PREP switch is in NORMAL.

ORM OR 2.5.2 MOV Thermal Overload Protection has been entered for the conduct of the surveillance.

Then, the following indications are received UNEXPECTEDLY:

Annunciator 5062-8E HPCS OUT OF SERVICE HPCS PWR LOSS OR OVLD ANY VLV status light Position indication lights for 1E22-F004 are extinguished (1) Is this condition reportable in accordance with LS-AA-1110 Reportability Manual AND (2) In addition to restoring the HPCS system to an OPERABLE status, what are the MINIMUM actions REQUIRED in accordance with Tech Spec 3.5.1 ECCS Operating?

A.

(1) No (2) Perform CPS 9054.01 RCIC System Operability Check.

B.

(1) No (2) Examine logs or other information to determine RCIC system operability.

C.

(1) Yes (2) Perform CPS 9054.01 RCIC System Operability Check.

D.

(1) Yes (2) Examine logs or other information to determine RCIC system operability.

Answer D

Answer Explanation D is correct. Given the conditions presented in the stem and per CPS 5062.03 Alarm Panel 5062 Annunciators - Row 8 annunciator 5062-8E HPCS OUT OF SERVICE, a blown control power fuse of 1E22-F004 HPCS To CNMT Outbd Isln Valve is the cause. This means the 1E22-F004 is electrically disabled due to a loss of control power.

Per ITS B3.5.1 ECCS - Operating, the HPCS System consists of a single motor driven pump, a spray sparger above the core, and piping and valves to transfer water from the suction source to the sparger.

With the 1E22-F004 electrically disabled, the HPCS System is INOPERABLE.

Per ITS LCO 3.5.1 ECCS - Operating

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 50 of 85 B.

High Pressure Core Spray (HPCS)

System inoperable B.1 AN D

B.2 Verify by administrative means RCIC System is OPERABLE when RCIC is required to be OPERABLE.

Restore HPCS System to OPERABLE STATUS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 14 days Per ITS B3.5.1, OPERABILITY of RCIC is therefore verified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when the HPCS System is inoperable. This may be performed as an administrative check, by examining logs or other information, to determine if the RCIC is out of service for maintenance or other reasons. Verification does not require performing the Surveillances needed to demonstrate the OPERABILITY of the RCIC System.

Per LS-AA-1110 Reportability Manual 50.73(a)(2)(v): The licensee shall report...any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed for:

(A) Shutdown the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

Incorrect Responses:

A is incorrect but plausible. A plausible misconception that the condition is not reportable due to the HPCS System being a "redundant" ECCS system and therefore IAW LS-AA-1110 50.73(a)(2)(vi), not reportable because equipment in the same system (i.e., ECCS) was operable and available to perform the required safety function. The second part of the response is plausible because performing a surveillance is the normal method of verifying operability. However, examining logs or other information to determine RCIC system operability is acceptable per ITS 3.5.1.

B is incorrect but plausible. A plausible misconception that the condition is not reportable due to the HPCS System being a "redundant" ECCS system and therefore IAW LS-AA-1110 50.73(a)(2)(vi), not reportable because equipment in the same system (i.e., ECCS) was operable and available to perform the required safety function. The second part of the response is correct.

C is incorrect but plausible. The first part of this response is correct. The second part of the response is plausible because performing a surveillance is the normal method of verifying operability. However, examining logs or other information to determine RCIC system operability is acceptable per ITS 3.5.1.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 51 of 85 Question Information Topic RCIC ITS User ID CL-ILT-N20089 System ID 2202616 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I 10CFR55 Content CFR: 43.2 Facility operating limitations in the technical specifications and their bases.

References Provided LS-AA-1110 pages 30-34 K/A Justification Question meets the KA because the examinee must determine the effect on the RCIC system of a blown fuse (degraded power supply) which occurred while performing maintenance.

SRO-Only Justification The administrative actions required to verify RCIC system operability is contained in the Tech Spec Bases, which is SRO only knowledge. Linked to 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Additional Information Question is low cog, written at the memory level. The examinee must recall facts and information contained in a procedure (1-F).

NRC Exams Only Question Type Bank (CL-ILT-A17088)

Difficulty N/A Technical Reference and Revision #

• LS-AA-1110 Rev. 31 ITS LCO 3.5.1 (3.5-1, Amend. 231)

ITS B3.5.1 (B3.5-3, Rev. 20-2)

CPS 5062.08(8E), Rev. 26c Training Objective 217000.09 DISCUSS the effect:

.1. A total loss or malfunction of the REACTOR CORE ISOLATION COOLING (RI) System has on the plant.

.2. A total loss or malfunction of various plant systems has on the REACTOR CORE ISOLATION COOLING (RI) System.

Previous NRC Exam Use None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 52 of 85 K/A Reference(s)

GS.217000 Safety Function 2 Tier 2 Group 1 RO Imp:

SRO Imp:

Reactor Core Isolation Cooling System (RCIC)

B2.2.36 Safety Function 2 Tier 3 Group RO Imp: 3.1 SRO Imp: 4.2 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13)

Learning Objective(s)

Q14/89 217000 2.2.36 (BL)

User (Sys) ID N/A (1554384)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 53 of 85 Question 15 ID: 2202403 Points: 1.00 A loss of Service Air/Instrument Air (SA/IA) pressure with the reactor at rated thermal power, resulted in the following conditions:

Reactor scrammed, all rods in.

Main Steam Equalizing header pressure at 0 psig.

SA/IA pressure is now recovering and is currently 40 psig and rising at 1 psig per minute.

What actions are required to be performed NEXT?

A.

Place all SRV control switches to OFF.

B.

Open IA ring header isolation valves to the Control Building.

C.

Open IA ring header isolation valves to the Radwaste Building.

D.

Place all Main Steam Isolation Valve (MSIV) control switches to CLOSE.

Answer D

Answer Explanation D is correct. Per CPS 4004.01 Instrument Air Loss Subsequent Action 4.4, the control switches for Main Steam Isolation Valves (MSIVs) are required to be placed to CLOSE to prevent an inadvertent reopening when air pressure is restored.

Incorrect Responses:

A is incorrect but plausible. EOP-1 detail P1 directs placing all SRV control switches to off if the IA supply to the SRVs is lost; however, the SRV IA supply from the back up air bottles has not been impacted by the conditions presented in the stem. Therefore, placing SRV control switches to off is unnecessary.

B is incorrect but plausible. Although this action is not addressed in CPS 4004.01, plausible because the recovery of the Control Building IA ring header is addressed in the normal operating procedure CPS 3214.01, Plant Air (IA & SA) section 8.2.1.3 Control Building Instrument Air Header Auto Isolation. This is also incorrect because recovery of the CB IA Header cannot be accomplished until air header pressure is

> 70 psig. Placing the MSIV control switches in CLOSE is prioritized higher to prevent an uncontrolled cooldown of the RPV.

C is incorrect but plausible. Although this action is not addressed in CPS 4004.01, plausible because the recovery of the Radwaste Building IA ring header is addressed in the normal operating procedure CPS 3214.01, Plant Air (IA & SA) section 8.2.1.2 Radwaste Building Instrument Air Header Auto Isolation.

This is also incorrect because recovery of the RW IA Header cannot be accomplished until air header pressure is > 70 psig. Placing the MSIV control switches in CLOSE is prioritized higher to prevent an uncontrolled cooldown of the RPV.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 54 of 85 Question Information Topic Loss of Service Air/Instrument Air User ID CL-ILT-N20090 System ID 2202403 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the examinee must understand the status of the instrument air system and determine what operator action must be directed for the conditions given to answer the question correctly.

SRO-Only Justification Question is linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency conditions.

Additional Information This is a high cog question, written at the analysis and comprehension level. The examinee must analyze the conditions provided in the stem and then determine which action is prioritized the highest based on the analysis (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-N11082)

Difficulty N/A Technical Reference and Revision #

• CPS 4004.01 Rev. 10c CPS 3214.01 Rev. 27c EOP/SAG Technical Bases, Rev. 0 Training Objective 4004.01 (400401.01) From the MCR, complete Control Room actions to respond to a Loss of Instrument Air IAW CPS 4004.01.

Previous NRC Exam Use ILT 10-1 NRC exam K/A Reference(s)

GS.300000 Safety Function 8 Tier 2 Group 1 RO Imp:

SRO Imp:

Instrument Air System (IAS)

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 55 of 85 B2.2.44 Safety Function 8 Tier 3 Group RO Imp: 4.2 SRO Imp: 4.4 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)

Learning Objective(s)

Q15/90 300000 2.2.44 (BH)

User (Sys) ID N/A (1554385)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 56 of 85 Question 16 ID: 2202303 Points: 1.00 Plant conditions are as follows:

• The reactor is at 30% power during a startup following refueling.

• Reactor pressure is 950 psig.

• Individual control rod scram time testing surveillance has just started.

Scram times for the first two rods to be tested are listed below:

Control Rod 28-29 Control rod 32-29 Time to notch 43 0.31 seconds 0.29 seconds Time to notch 29 0.79 seconds 0.75 seconds Time to notch 13 1.44 seconds 7.35 seconds (1) What is the status of these rods?

(2) Is a plant shutdown required?

A.

(1) One rod is considered "slow."

(2) Commence plant shutdown.

B.

(1) One rod is considered "slow."

(2) Plant shutdown is not necessary.

C.

(1) Both rods are considered "slow."

(2) Commence plant shutdown.

D.

(1) Both rods are considered "slow."

(2) Plant shutdown is not necessary.

Answer B

Answer Explanation B is correct. Per CPS ITS 3.1.4, Table 3.1.4-1, control rod 28-29 is considered "slow." Table 3.1.4-1 also states that any rod with a scram time >7 seconds to notch position 13 is not considered "slow," and is inoperable. Per CPS ITS 3.1.3, if 9 or more rods are inoperable, the plant must be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Per CPS ITS 3.1.4, if 10 or more operable rods are considered "slow," the plant must be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Incorrect responses:

A is incorrect. The first part of the response is correct. However, a plant shutdown is only required if more than 10 control rods are "slow" or if 2 "slow" rods are adjacent to one another. This response is plausible since the 2 rods in question occupy adjacent locations.

C is incorrect. Rod 32-29 is not considered "slow;" it is inoperable per ITS 3.1.4., and a plant shutdown is only required if more than 10 control rods are "slow" or if 2 "slow" rods are adjacent to one another. This

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 57 of 85 response is plausible because of the long scram time associated with rod 32-29, and since the 2 rods in question occupy adjacent locations.

D is incorrect. Rod 32-29 is not considered "slow;" it is inoperable per ITS 3.1.4. This response is plausible because of the long scram time associated with rod 32-29. The second part of the response is correct.

Question Information Topic Slow Control Rods User ID CL-ILT-N20091 System ID 2202303 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided CPS ITS 3.1.4 K/A Justification This question meets the KA because the examinee must predict the impact of a long scram time for two rods and determine the action required per Technical Specifications.

SRO-Only Justification This question is linked to SRO-only task 140109.23 (Apply the administrative requirements for execution of Technical Specifications and Off-Site Dose Calculation Manual Requirements). Also linked to 10CFR55.43(b)(2), Facility operating limitations in the Technical Specifications and their bases.

Additional Information Question is high cog written at the application level. The candidate must apply the conditions presented in the stem (control rod scram times) to the reference (technical specifications) to determine the required actions (3-SPR).

NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

• CPS ITS 3.1.3 (3.1-7 Amend. 188, 3.1-8 Amend.

149, 3.1-9, 10 Amend. 192)

CPS ITS 3.1.4 (3.1-11 Amend. 95, 3.1-12 Amend.

192, 3.1.13 Amend. 95, 3.1-14 Amend. 95)

Training Objective 201003.12 Given Control Rod and Control Rod Drive Mechanism System operability status OR key parameter indications, plant conditions, and a copy of Tech Specs, DETERMINE if Tech Spec Limiting Condition for Operations have been met, and required actions if any.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 58 of 85 Previous NRC Exam Use None K/A Reference(s) 201003.A2.10 Safety Function 1 Tier 2 Group 2 RO Imp: 3.0 SRO Imp: 3.4 Ability to (a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

Excessive SCRAM time for a given drive mechanism Learning Objective(s)

Q16/91 201003 A2.10 (NH)

User (Sys) ID N/A (1554386)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 59 of 85 Question 17 ID: 2202597 Points: 1.00 The plant was operating at power with parameters as follows:

2697 MWth 90% Rod Line 66.5 Mlb/hr core flow THEN, RR FCV B drifted closed (no apparent cause). Immediate operator actions were taken to stop the transient and stabilize the plant.

Current plant parameters are as follows:

Reactor power at 2415 MWTh and steady Total core flow at 55.8 Mlb/hr and steady

'A' RR loop flow is 33.4 Mlb/hr

'B' RR loop flow is 22.6 Mlb/hr Reactor operation is confirmed to be below the MELLLA limit and outside the Controlled Entry Region.

(1) What is the impact of operating the Reactor Coolant Recirculation System under these conditions?

(2) To mitigate this condition, the CRS should direct the 'A' RO to SHUT the 'A' Reactor Recirc Pump FCV until the 'A' recirculation loop jet pump flow is at a MAXIMUM of _________.

A.

(1) Unacceptable flow coastdown and core response during a design basis LOCA.

(2) 26.8 Mlb/hr B.

(1) Unacceptable flow coastdown and core response during a design basis LOCA.

(2) 31.0 Mlb/hr C.

(1) Non-conservative APRM flow biased simulated thermal power setpoints.

(2) 26.8 Mlb/hr D.

(1) Non-conservative APRM flow biased simulated thermal power setpoints.

(2) 31.0 Mlb/hr Answer B

Answer Explanation B is correct. In accordance with ITS 3.4.1 Recirculation Loops Operating and its associated bases, the operation of the Reactor Coolant Recirculation System is an initial condition assumed in the design basis loss of coolant accident (LOCA). The analyses assume that both loops are operating at the same flow prior to the accident. The LOCA analysis was reviewed for the case with a flow mismatch between the two loops with the pipe break assumed to be in the loop with the higher flow. A small mismatch was determined to be acceptable based on engineering judgment. However, outside that limit, the flow coastdown and core response are potentially more severe.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 60 of 85 The SRO would have 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore recirculation loop jet pump flow mismatch within the limits of 10% of rated core flow (8.45 mlb/hr) when operating at < 70% rated core flow (59.15 mlb/hr) or one recirculation loop must be shut down. Since the 'B' FCV is locked out, 'A' FCV must be SHUT until the 'A' recirculation loop jet pump flow mismatch is at a MAXIMUM of 31.0 Mlb/hr.

Incorrect Responses:

A is incorrect but plausible. The first part of the response is correct. The second part is plausible because 26.8 Mlb/hr corresponds to a 5% mismatch between the two loops. A 5% mismatch is the minimum acceptable flow mismatch for 70% of rated core flow; however, flow given in the stem is less than 70%.

C is incorrect but plausible. The first part of the response is plausible because with only one recirculation loop in service, the APRM flow biased simulated thermal power setpoints are considered non-conservative based on flow mismatch per 3.3.1.1 RPS Instrumentation bases. However, the conditions indicated in the stem have both loops in operation. Additionally, the second part is plausible because 26.8 Mlb/hr corresponds to a 5% mismatch between the two loops. A 5% mismatch is the minimum acceptable flow mismatch for 70% of rated core flow; however, flow given in the stem is less than 70%.

D is incorrect but plausible. The first part of the response is plausible because with only one recirculation loop in service, the APRM flow biased simulated thermal power setpoints are considered non-conservative based on flow mismatch per 3.3.1.1 RPS Instrumentation bases. However, the conditions indicated in the stem have both loops in operation. The second part of the response is correct.

Question Information Topic RR FCV Malfunction User ID CL-ILT-N20092 System ID 2202597 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.2 Facility operating limitations in the technical specifications and their bases.

References Provided None K/A Justification This question meets the KA because the examinee must demonstrate knowledge of the impact of a Recirculation flow Control System malfunction (SF2) and how to mitigate that impact.

SRO-Only Justification Question is linked to SRO only task 999999.07 Apply Technical Specification requirements and 10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Additional Information Question is high cog, written at the analysis level. The examinee has to analyze plant condition before and after an event then use knowledge to predict an impact based on the event and determine the correct actions to mitigate the event. (3-SPK)

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 61 of 85 NRC Exams Only Question Type Bank (CL-ILT-N14093)

Difficulty N/A Technical Reference and Revision #

• ITS 3.4.1 (pg. 3.4-1) Amend. 171 ITS 3.4.1 (pg. 3.4-3) Amend. 192 ITS B 3.4.1 (pg. B 3.4-2) Rev. No. 1-1 ITS B 3.4.1 (pg. B 3.4-3) Rev. No. 11-1 ITS B 3.3.1.1 (pg. B 3.3-8) Rev. No. 0 Training Objective 202001.14 Given Reactor Recirculation System operability status and a copy of Tech Specs, DISCUSS the bases for the Reactor Recirculation System Tech Spec LCO, related safety limits and Limiting Safety System Settings.

Previous NRC Exam Use ILT 14-1 NRC Exam K/A Reference(s)

B2.4.21 Safety Function 1 Tier 3 Group RO Imp: 4.0 SRO Imp: 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.

(CFR: 41.7 / 43.5 / 45.12)

GS.202002 Safety Function 1 Tier 2 Group 2 RO Imp:

SRO Imp:

Recirculation Flow Control System Learning Objective(s)

Q17/92 202002 2.4.08 (BH)

User (Sys) ID N/A (1554387)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 62 of 85 Question 18 ID: 2228838 Points: 1.00 The plant was operating with normal system alignments at 25% power when annunciator 5007-1B TURB TRIP EHC SYS was received.

(1) Which of the following graphics (graphic 1 or graphic 2) represents the expected status of GCB 4506, GCB 4510, and MOD 4508 five minutes after receipt of annunciator 5007-1B?

Required actions are to _____(2)_____.

(1)

(2)

A.

(1) Graphic 1 (2) insert control rods until power is < 21.6%

B.

(1) Graphic 2 (2) insert control rods until power is < 21.6%

C.

(1) Graphic 1 (2) secure one of the three operating Circulating Water Pumps

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 63 of 85 D.

(1) Graphic 2 (2) secure one of the three operating Circulating Water Pumps Answer A

Answer Explanation A is correct:

Per CPS 3004.01 Turbine Startup and Generator Synchronization, the Main Generator is synchronized to the grid between 15% and 18% reactor power, closing Gas Circuit Breakers (GCB) 4510 and 4506.

Per CPS 5007-1B Turb Trip EHC Sys, a turbine trip will cause the generator to trip on reverse power, causing GCBs 4506 and 4510 to open (Graphic 1).

Per CPS 5004.03 (3D), the RPS trip on TSV/TCV closure is bypassed when < 33.3% RTP. At 25% power, the turbine will trip, but the reactor will NOT scram.

Per CPS 4005.01 Loss of Feedwater Heating step 4.5, if the main turbine trips without a SCRAM, then insert control rods to reduce power to < 21.6%.

Incorrect Responses:

B is incorrect but plausible. Opening Motor-Operated Disconnect (MOD) 4508 would disconnect the Main Generator from the grid, similar to how the RAT transformers are disconnected on a RAT trip, where the RAT Circuit Switcher 4538 opens. However, in the case of the Main Turbine, GCBs 4510 and 4506 are tripped and MOD 4508 remains closed. The MOD is not operated under load. The second part of the response is correct.

C is incorrect but plausible. The first part of the response is correct. CPS 4100.01 Scram, section 4.1 Electric Concerns Post Scram, step 4.1.2 directs securing 1 of the 3 CW pumps if operating with two Reactor Recirculating (RR) pumps, three Circulating Water (CW) pumps and the Motor-Driven Reactor Feed Pump (MDRFP) on RAT 'A'. Since the stem states that reactor power is at 25%, both RR Pumps are operating in slow speed. In slow speed, both RR pumps are powered from 4160V Buses 1A and 1B via RAT 'B' following the turbine trip, so the requirement to secure a Circulating Water Pump does not exist.

D is incorrect but plausible. Opening Motor-Operated Disconnect (MOD) 4508 would disconnect the Main Generator from the grid, similar to how the RAT transformers are disconnected on a RAT trip, where the RAT Circuit Switcher 4538 opens. However, in the case of the Main Turbine, GCBs 4510 and 4506 are tripped and MOD 4508 remains closed. The MOD is not operated under load. Additionally, CPS 4100.01 Scram, section 4.1 Electric Concerns Post Scram, step 4.1.2 directs securing 1 of the 3 CW pumps if operating with two Reactor Recirculating (RR) pumps, three Circulating Water (CW) pumps and the Motor-Driven Reactor Feed Pump (MDRFP) on RAT 'A'. Since the stem states that reactor power is at 25%, both RR Pumps are operating in slow speed. In slow speed, both RR pumps are powered from 4160V Buses 1A and 1B via RAT 'B' following the turbine trip, so the requirement to secure a Circulating Water Pump does not exist.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 64 of 85 Question Information Topic Main Turbine Trip User ID CL-ILT-N20093 System ID 2228838 Status Active Point Value 1.00 Time (min) 2 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I References Provided:

None K/A Justification Statement:

Question meets the KA because the candidate must demonstrate the ability to predict the impact of a Turbine trip on the Main Turbine Generator and Auxiliary System controls by choosing a graphic that shows the predicted alignment following a turbine trip to answer the question.

SRO Only Justification Statement:

This question is linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Additional Information:

Question is written at the comprehension level (high cog). The examinee must recognize the interaction between systems and determine consequences/implications. (2-RI)

NRC Exams Only (as applicable)

Question Type:

Bank (CL-ILT-N17091)

Difficulty:

N/A Technical Reference and Revision #:

CPS 3004.01 Rev. 34e CPS 5007.01 (1B) Rev. 30f CPS 5004.03 (3D) Rev. 28c CPS 4005.01 Rev. 19b CPS 4100.01 Rev. 24 Training Objective:

(410001.01) From the MCR, complete Control Room Actions to respond to a Reactor Scram, IAW CPS 4100.01 Reactor Scram.

Previous NRC Exam Use:

ILT 17-1 NRC Exam

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 65 of 85 K/A Reference(s) 245000.A2.01 Safety Function 4 Tier 2 Group 2 RO Imp: 3.7 SRO Imp: 3.9 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

(CFR: 41.5 / 45.6)

Turbine trip Learning Objective(s)

Q18/93 214000 A2.01 (NH)

User (Sys) ID N/A (1554388)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 66 of 85 Question 19 ID: 2202302 Points: 1.00 Core Alterations were stopped by the Refuel SRO due to plant conditions.

Who has the authority to grant permission to resume fuel movement?

A.

Refuel SRO B.

Shift Manager C.

Lead Refuel SRO D.

Control Room Supervisor Answer B

Answer Explanation B is correct.

Per CPS 3703.01 Core Alterations Section 4.29, the Shift Manager's permission is required to resume fuel movement.

Incorrect Responses:

A is incorrect but plausible. The Refuel SRO has a number of responsibilities during core alterations but does not possess the authority to grant permission to resume fuel movement.

C is incorrect but plausible. The Lead Refuel SRO has responsibilities during core alterations but does not possess the authority to grant permission to resume fuel movement.

D is incorrect but plausible. The Control Room Supervisor has responsibilities during core alterations but does not possess the authority to grant permission to resume fuel movement.

Question Information Topic Refueling SRO Responsibilities User ID CL-ILT-N20094 System ID 2202302 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I 10CFR55 Content CFR: 43.7 Fuel handling facilities and procedures.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 67 of 85 References Provided None K/A Justification This question meets the KA because the examinee has to identify the actions required of the Refueling SRO in the event that core alts are impaired due to plant conditions.

SRO-Only Justification This question is linked to 10CFR55.43(b)(6) -

Procedures and limitations involved in core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. This question also tests knowledge of SRO-only responsibilities during refueling operations.

Additional Information This is a low cog question written at the memory level.

The examinee has to recall facts from a procedure to answer the question (1-F/1-P).

NRC Exams Only Question Type Bank (CL-ILT-N11093, CL-ILT-N17098)

Difficulty N/A Technical Reference and Revision #

• CPS 3703.01, Rev. 29b Training Objective LP85802.2.2.26 Knowledge of refueling administrative requirements.

Previous NRC Exam Use ILT 10-1 NRC Exam, ILT 17-1 NRC Exam K/A Reference(s)

B2.1.35 Safety Function 4 Tier 3 Group RO Imp: 2.2 SRO Imp: 3.9 Knowledge of the fuel-handling responsibilities of SROs.

(CFR: 41.10 / 43.7)

Learning Objective(s)

Q19/94 2.1.35 (BL)

User (Sys) ID N/A (1554389)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 68 of 85 Question 20 ID: 2202261 Points: 1.00 A fuel shuffle is in progress. The fuel movement sheet indicates that a fuel assembly will be inserted into core location 15-38.

Before lowering the fuel bundle into its new core location, which direction should the channel fastener be pointing?

A.

SW B.

NW C.

NE D.

SE Answer D

Answer Explanation X-AXIS Y-AXIS

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 69 of 85 D is correct. Per N-CL-OPS-290004, proper orientation of fuel assemblies is assured by verification procedures during core loading. Five separate visual indications of proper fuel assembly orientation exist:

The channel fastener assemblies are to the center of the control rod blade.

The handle orientation boss points toward the control blade (toward the center of the cell).

The channel spacer buttons are toward the control rod blade.

The alphanumeric serial numbers on the handles can be read from the center of the cell.

There is cell-to-cell symmetry.

Since the fuel assembly is in the bottom right of the fuel cell (NW), the channel fastener should point toward the center of the cell (SE).

Incorrect Responses:

A is incorrect but plausible if the examinee fails to recall the location of the channel fastener in relation to the control rod blade center on the fuel bundle or makes an error determining the correct core location, etc.

B is incorrect but plausible if the examinee fails to recall the location of the channel fastener in relation to the control rod blade center on the fuel bundle or makes an error determining the correct core location, etc.

C is incorrect but plausible if the examinee fails to recall the location of the channel fastener in relation to the control rod blade center on the fuel bundle or makes an error determining the correct core location, etc.

Question Information Topic Fuel Bundle Orientation User ID CL-ILT-N20095 System ID 2202261 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I References Provided None K/A Justification Question meets the KA because the examinee must recall the administrative requirements for fuel bundle orientation and apply that knowledge to determine correct fuel bundle orientation during core alterations, an operationally valid, safety significant operation.

SRO-Only Justification This question is an SRO-only question because the refueling SRO is responsible for ensuring correct orientation of the fuel bundle before releasing the grapple (CPS 3703.01 step 4.22). Also linked to 10CFR55.43(b)(7) Fuel handling facilities and procedures.

Additional Information This question is high cog, written at the analysis/

comprehension level. The examinee must solve a problem (correct fuel bundle orientation) using

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 70 of 85 knowledge (fuel bundle construction, relationship of the fuel bundle to the fuel cell, etc.) (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-N15094)

Difficulty N/A Technical Reference and Revision #

• CPS 3703.01, Rev. 29b N-CL-OPS-290004, Rev. 7 Training Objective 290004.18 Describe five indications used to verify proper Fuel Bundle orientation in the Reactor.

LP85801.2.1.40 Knowledge of refueling administrative requirements.

LP86610.01.09Describe four (4) methods used to verify proper fuel assembly orientation.

Previous NRC Exam Use ILT 15-1 NRC Exam K/A Reference(s)

B2.1.40 Safety Function 4 Tier 3 Group RO Imp: 2.8 SRO Imp: 3.9 Knowledge of refueling administrative requirements.

(CFR: 41.10 / 43.5 / 45.13)

Learning Objective(s)

Q20/95 2.1.40 (BH)

User (Sys) ID N/A (1554390)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 71 of 85 Question 21 ID: 2202260 Points: 1.00 Which of the following activities is a controlled exclusion to a temporary configuration change per CC-AA-112, Temporary Configuration Changes?

A.

Disabling an alarm in accordance with OP-AA-103-102, Watchstanding Practices.

B.

Changing a Control Room alarm setpoint in accordance with an approved troubleshooting plan.

C.

Placement of plastic sheeting over plant ventilation ducts in accordance with an approved painting plan.

D.

Installation of a catch containment under a leaking component in the SX 'A' pump room in accordance with an approved Radiation Protection procedure.

Answer D

Answer Explanation D is correct. Per CC-AA-112 Temporary Configuration Changes:

4.2.11 Catch Basins controlled under procedure RP-AA-502 are Controlled Exclusions from this procedure. Other similar housekeeping type temporary configurations changes controlled by approved procedures are Controlled Exclusions from this procedure.

Incorrect Responses:

A is incorrect but plausible. Disabling alarms is performed occasionally for nuisance alarms in the Main Control Room, and a common misconception is that doing so is not a temporary configuration change.

However, it is listed in CC-AA-112 section 4.2.12 as one of the activities typically not considered a Controlled Exclusion.

B is incorrect but plausible. Per CC-AA-112, section 4.2.12, temporary setpoint changes are typically not considered Controlled Exclusions and must be tagged and controlled per CC-AA-112 Temporary Configuration Changes. This answer may be selected because it is in accordance with an approved procedure; however, MA-AA-716-004, Conduct of Troubleshooting, section 4.1.10.4 requires use of CC-AA-112 Temporary Configuration Changes, if as-left configuration will be different from plant design.

C is incorrect but plausible. Plastic sheeting is a temporary configuration change if it has the potential to significantly affect plant design (ventilation, heat removal, access to critical equipment, fire suppression, etc.). Normally barriers and enclosures used by maintenance do not significantly affect plant design.

Some examples of barriers or enclosures for consideration as temporary barriers include plastic sheeting, canvas, Herculite, Sil temp, or tarps used for contamination control, FME, abatement enclosures, welding shields, etc. Plastic sheeting over ventilation ducts would significantly affect plant design, and would not be a Controlled Exclusion.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 72 of 85 Question Information Topic Temporary Configuration Change User ID CL-ILT-N20096 System ID 2202260 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I References Provided None K/A Justification Question meets the KA because the examinee must have knowledge of the Temporary Configuration Change process to answer the question.

SRO-Only Justification Question is linked to SRO only task 999999.10 CC-AA-112 - Authorize installation of a Temporary Modification.

Question is linked to 10CFR55.43(b)(3) Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Additional Information This question is low cog, written at the memory level.

The examinee has to recall facts from a procedure to answer the question (1-F/1-P).

NRC Exams Only Question Type Modified (CL-ILT-N18097)

Difficulty N/A Technical Reference and Revision #

• CC-AA-112, Rev. 29 MA-AA-716-004, Rev. 18 Training Objective LP85802.2.2.5 Knowledge of the process for making design or operating changes to the facility.

Previous NRC Exam Use None K/A Reference(s)

B2.2.11 Safety Function 4 Tier 3 Group RO Imp: 2.3 SRO Imp: 3.3 Knowledge of the process for controlling temporary design changes.

(CFR: 41.10 / 43.3 / 45.13)

Learning Objective(s)

Q21/96 2.2.11 (ML)

User (Sys) ID N/A (1554391)

Cross Reference Links Table: TRAINING - QUESTIONS - Track Questions Modified in this Project (CL-OPS-EXAM-ILT)

Tracking link in project CL-OPS-EXAM-ILT to source question 2202259

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 73 of 85

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 74 of 85 Question 22 ID: 2202454 Points: 1.00 Per CC-AA-103-100 Configuration Change Control for Permanent Physical Plant Changes, Operations is responsible for A.

performing an independent detailed design verification of Safety Related Configuration Changes.

B.

approving a Configuration Change impacting Operations at the Plant Operations Review Committee (PORC).

C.

identifying the need for and ensuring the revision of specific operating procedures that are affected by a Configuration Change.

D.

determining the need for an Operational Briefing BEFORE the Work Orders that implement the Configuration Change are submitted to Operations.

Answer C

Answer Explanation C is correct.

Per CC-AA-103-100 Configuration Change Control for Permanent Physical Plant Changes, Section 3.9 states that Operations is responsible for identifying and ensuring revision of specific operating procedures and Operator training are updated for the Configuration Change.

Incorrect Responses:

A is incorrect but plausible. A plausible misconception is that Operations personnel are responsible for performing independent detailed design verification of configuration changes. However, the Design Verifier is responsible for this per CC-AA-103-100 section 3.7.

B is incorrect but plausible. This is plausible because the Plant Operations Review Committee (PORC) requires an Operations Representative for configuration changes having to do with Operations. However, the responsibility for approving those changes rests with the Plant Manager at PORC per CC-AA-103-100 section 3.11.

D is incorrect but plausible. This response is plausible because Operations personnel must review any changes and determine if formal training is required; however, the Operational Briefing is separate from training. The Design Engineering Manager (DEM) is responsible for determining if an Operational Briefing is necessary in conjunction with a configuration change per CC-AA-103-100 section 3.6.2.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 75 of 85 Question Information Topic Design Change Procedures User ID CL-ILT-N20097 System ID 2202454 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I 10CFR55 Content CFR: 43.3 Facility licensee procedures required to obtain authority for design and operating changes in the facility.

References Provided None K/A Justification Question meets the KA because the candidate must demonstrate knowledge of the process for making design changes to the facility by determining Operations Department responsibilities for a proposed design change.

SRO-Only Justification Question is linked to 10CFR55.43(b)(3) Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Additional Information Question is low cog written at the memory level. The candidate must recall facts and apply them to the specific situation. (1-F)

NRC Exams Only Question Type Bank (CL-ILT-N19097)

Difficulty N/A Technical Reference and Revision #

• CC-AA-103-100, Rev 2 Training Objective LP85802.2.2.5 Knowledge of the process for making design or operating changes to the facility.

Previous NRC Exam Use ILT 19-1 NRC Exam K/A Reference(s)

B2.2.05 Safety Function 4 Tier 3 Group RO Imp: 2.2 SRO Imp: 3.2 Knowledge of the process for making design or operating changes to the facility.

(CFR: 41.10 / 43.3 / 45.13)

Learning Objective(s)

Q22/97 2.2.5 (PL)

User (Sys) ID N/A (1554392)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 76 of 85

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 77 of 85 Question 23 ID: 2202258 Points: 1.00 A fuel damaging LOCA has occurred.

Containment radiation levels are currently at 43 Rem/hr.

The TSC has NOT been activated.

The MCR has determined that RPV level restoration will be attempted using the Standby Liquid Control (SLC) Storage Tank IAW CPS 4411.03 Injection Flooding Sources Appendix D: SLC (test or storage tank).

The exposure limit (TEDE) for each person performing local operations on the SLC system is

_____(1)_____.

The Shift Manager _____(2)_____ authorize the use of Potassium Iodide (KI) to the personnel assigned to perform local operations on the SLC system.

A.

(1) 10 Rem (2) may B.

(1) 25 Rem (2) may C.

(1) 10 Rem (2) may NOT D.

(1) 25 Rem (2) may NOT Answer A

Answer Explanation A is correct:

Per EP-AA-113 Personnel Protective Actions, 3. Responsibilities, the Shift Manager (Shift Emergency Director) shall perform the responsibilities of the Station Emergency Director until relieved. The Station Emergency Director is responsible for the following protective actions:

Authorization for emergency exposure greater than 5 Rem Authorization for issuance of KI to Exelon Nuclear emergency workers and/or onsite personnel Direction of Assembly, Accountability and Evacuation of personnel.

Per EP-AA-113 Attachment 1, the dose limit for protecting valuable property is 10 Rem TEDE when a lower dose is not practical.

Per EP-AA-113 section 4.4 KI Assessment, step 4.4.1.B, if workers will be entering an unknown radiological atmosphere that is suspected to have a high iodine concentration (i.e. loss of Fuel Clad barrier) the Shift Manager should recommend the issuance of one (1) 130 mg KI tablet to each emergency worker affected per day for 10 consecutive days. Per EP-AA-1003 Addendum 3

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 78 of 85 page CL 2-3 Hot Matrix (barrier criteria), a containment radiation monitoring reading > 41.3 R/hr represents a loss of the Fuel Clad.

Incorrect Responses:

B is incorrect but plausible. 25 rem is the emergency exposure limit for lifesaving operations. The second part of the question is correct.

C is incorrect but plausible. The first part of the question is correct. The Shift Emergency Director is responsible for issuing KI to emergency workers and/or onsite personnel. This answer would be correct if the Shift Manager had been relieved as the Station Emergency Director or if a loss of fuel clad barrier had not occurred.

D is incorrect but plausible - 25 rem is the emergency exposure limit for lifesaving operations, and the Shift Emergency Director is responsible for issuing KI to emergency workers and/or onsite personnel.

This answer would be correct if the Shift Manager had been relieved as the Station Emergency Director or if a loss of fuel clad barrier had not occurred.

Question Information Topic Emergency Exposure User ID CL-ILT-N20098 System ID 2202258 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I References Provided None K/A Justification Question meets the KA because the examinee must have knowledge of the emergency exposure hazards to answer the question.

SRO-Only Justification Question is linked to SRO-only task 997777.03 Emergency Plan Activities performed by an SRO. The Station Emergency Director position is filled by the Shift Manager prior to transferring command and control to the Station Emergency Director.

Additional Information This question is low cog, written at the memory level.

Requires recall of procedure steps (1-B).

NRC Exams Only Question Type Bank (CL-ILT-N18098)

Difficulty N/A Technical Reference and Revision #

• EP-AA-113 Rev. 15 EP-AA-1003 Addendum 3 Rev. 5 Training Objective LP85803.2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Previous NRC Exam Use ILT 18-1 NRC Exam

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 79 of 85 K/A Reference(s)

B2.3.14 Safety Function 4 Tier 3 Group RO Imp: 3.4 SRO Imp: 3.8 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(CFR: 41.12 / 43.4 / 45.10)

Learning Objective(s)

Q23/98 2.3.14 (PL)

User (Sys) ID N/A (1554393)

Cross Reference Links None

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 80 of 85 Question 24 ID: 2202275 Points: 1.00 The upper end of the EOP hierarchy represents guidance which is...

A.

function or intent driven (e.g., inject water, isolate a leak).

B.

overridden by lower tier procedures should a conflict in operating direction exist.

C.

written with a lower level of conservative decision making as they are only concerned with placing the plant in a safe condition.

D.

required to be performed as "direct referral" to the source procedure (i.e. actions cannot be performed from memory) regardless of event severity.

Answer A

Answer Explanation A is correct.

Per CPS 1005.09, CONTROL AND USAGE OF EMERGENCY OPERATING PROCEDURES (EOP) AND SEVERE ACCIDENT GUIDELINES (SAG), section 8.8.1.2, the upper end of the hierarchy represents function or intent driven guidance procedures (e.g., inject water, isolate a leak) which utilize applicable operating details from the lower tier documents.

Incorrect Responses:

B is incorrect but plausible. This response is plausible due to the lower tier procedures utilizing specific operating direction for many plant components and the possible misconception that lower tier procedures take precedence should a conflict in operating direction exist when in fact, the upper tier procedure in use shall take precedence.

C is incorrect but plausible. This response is plausible due to the common misconception that in order to get the plant to a safe condition; upper tier documents are less conservative. Each higher level of emergency response procedure is written with a higher level of conservative decision making designed to bring the plant into the safest condition possible.

D is incorrect but plausible. This response is plausible because actions are not performed from memory except in rare circumstances. However, per CPS 1005.09, section 8.8.2.2, consistent with the severity of the emergency event, these actions may be performed from memory without direct referral to the source document.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 81 of 85 Question Information Topic EOP Hierarchy User ID CL-ILT-N20099 System ID 2202275 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level RO-MEMORY Operator Discipline LO-I References Provided None K/A Justification Question meets the KA because the examinee must demonstrate knowledge of the priority and hierarchy of emergency response procedures.

SRO-Only Justification Question is linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. It requires knowledge of administrative procedures that specify hierarchy of plant normal, abnormal and emergency procedures.

Additional Information Question is low cog, written at the memory level. The candidate must recall facts and apply them to the specific situation. (1-F)

NRC Exams Only Question Type Bank (CL-ILT-N11097)

Difficulty N/A Technical Reference and Revision #

• CPS 1005.09, Rev. 11a Training Objective LP87551.01.02 Explain the relative position of the EOPs in the hierarchy of plant procedures.

Previous NRC Exam Use ILT 11-1 NRC Exam K/A Reference(s)

B2.4.23 Safety Function 4 Tier 3 Group RO Imp: 3.4 SRO Imp: 4.4 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

(CFR: 41.10 / 43.5 / 45.13)

Learning Objective(s)

Q24/99 2.4.23 (BL)

User (Sys) ID N/A (1554394)

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 82 of 85 Cross Reference Links Table: TRAINING - QUESTIONS - Track Questions Modified in this Project (CL-OPS-EXAM-ILT)

Tracking link in project CL-OPS-EXAM-ILT to source question 2202255

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 83 of 85 Question 25 ID: 2202235 Points: 1.00 The plant was operating at rated thermal power.

At 11:00, Security reports that a group of anti-nuclear protestors have assembled on the intersection of Highway 54 and the CPS access road to demonstrate.

At 12:00, A situation requiring evacuation of the Main Control Room (MCR) is ordered due to heavy smoke in the MCR; CPS 4003.01 Remote Shutdown (RS) is entered.

At 12:05, All MCR actions were taken prior to evacuation and all systems responded as expected; the MCR Operators begin their transit to the Remote Shutdown Panel.

At 12:16, MCR Operators arrive at the Remote Shutdown Panel, and assume plant control.

What is the highest Emergency Classification for this event?

A.

HU1 B.

HU3 C.

HA2 D.

HS2 Answer C

Answer Explanation C is correct.

Per EP-AA-1003 Addendum 3 Emergency Action Levels for Clinton Station, the Emergency Action Level (EAL) HA2 Control Room evacuation resulting in transfer of plant control to alternate locations is the highest Emergency Classification for this event.

CPS 4003.01 Remote Shutdown (RS) was entered at 12:00 to evacuate the Main Control Room (MCR).

This meets the requirements for the HA2 declaration.

At time 12:05, the MCR is evacuated.

At time 12:16, Remote Shutdown Panel (RSP) control was established (11 minutes elapsed).

Incorrect Responses:

A is incorrect but plausible. This response is plausible due to the demonstration given in the stem.

However, a demonstration does not rise to the level of a security condition and would not meet the threshold for HU1. Security conditions are defined as a security event that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant.

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 84 of 85 B is incorrect but plausible. This response is plausible due to the heavy smoke given in the stem.

However, the only indication is heavy smoke with no fire alarms; thus entry into HU3 is not required.

D is incorrect but plausible. This response is plausible because the conditions presented in the stem may have constituted the inability to control a key safety function from outside the control room (HS2).

However, the stem stated that control of Table H1 safety functions from the RSP was established within 15 minutes of the last person leaving the MCR; thus entry into HS2 is not required.

Question Information Topic Hazard EAL User ID CL-ILT-N20100 System ID 2202235 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I 10CFR55 Content CFR: 43.5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References Provided EP-AA-1003 Addendum 3, pages CL 2-8 thru CL 2-12.

K/A Justification Question meets the KA because the examinee must demonstrate knowledge of EAL action level thresholds and classifications with respect to a Control Room Abandonment event.

SRO-Only Justification Question is linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. It is also linked to SRO only task 997777.01 From the MCR, classify Emergency Action Level IAW EP-AA-1003, Radiological Emergency Plan Annex For Clinton.

Additional Information This is a high cog question, written at the analysis and application level. The examinee has to evaluate the conditions in the stem to determine the correct answer (3-SPR/SPK).

NRC Exams Only Question Type Bank (CL-ILT-N17077)

Difficulty N/A Technical Reference and Revision # EP-AA-1003 Addendum 3, Rev. 5 Training Objective LP85804.2.4.29 Knowledge of the emergency plan.

Previous NRC Exam Use ILT 17-1 NRC Exam

CONFIDENTIAL - Exam Material ILT 20-1 SRO Exam Approved Test Key Test ID: 355971 01/27/2022 85 of 85 K/A Reference(s)

B2.4.41 Safety Function 4 Tier 3 Group RO Imp: 2.9 SRO Imp: 4.6 Knowledge of the emergency action level thresholds and classifications.

(CFR: 41.10 / 43.5 / 45.11)

Learning Objective(s)

Q25/100 2.4.41 (BH)

User (Sys) ID N/A (1554395)

Cross Reference Links None