ML21144A273

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SRO Written Exam
ML21144A273
Person / Time
Site: Clinton Constellation icon.png
Issue date: 04/13/2021
From: Gregory Roach
Operations Branch III
To:
NRC/RGN-III/DRS/OLB
Roach G
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NISP-TR-01 Form 7 Revision 3 Page 1 of 1 EXAMINATION COVER SHEET Exam Title (ID) ILT 19-1 NRC SRO (343246)

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 19-1 NRC SRO Test Key Question 1 ID: 2147008 Points: 1.00 The unit was operating in Mode 2 with NO LCOs in effect when multiple annunciators were received on 1H13-P877-5060. 1H13-P877 indications are as follows:

The following is an excerpt from ITS 3.8.4 DC Sources - Operating:

3.8.4 DC Sources - Operating LCO 3.8.4 The Division 1, Division 2, Division 3, and Division 4 DC electrical power subsystems shall be OPERABLE.

Applicability:

Actions Condition Required Action Completion Time A. One battery charger on A.1 Restore battery terminal 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Division 1 or 2 inoperable. voltage to greater than or equal to the minimum established float voltage.

AND A.2 Verify battery float current 2 Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> amps.

AND Test ID: 343246 12/02/2020 2 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key A.3 Restore battery charger to 7 days OPERABLE status B. One battery on Division 1 or 2 B.1 Restore battery to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> inoperable. OPERABLE status.

For the conditions provided in the stem:

  • ITS LCO Actions(s) associated with Condition(s) _____(1)_____ must be entered.
  • Entry into Mode 1 _____(2)_____ permitted by Technical Specifications.

A. (1) A ONLY (2) is B. (1) A AND B (2) is C. (1) A ONLY (2) is NOT D. (1) A AND B (2) is NOT Answer D Answer Explanation D is correct:

Part 1 The meter indications (DC amps and volts) provided in the stem indicate that DC MCC 1A is de-energized (Battery Charger and Battery outputs are de-energized). If either source was energized (battery or battery charger), the DC voltage and amperage meters would be indicating values > 0 VDC.

With DC MCC 1A de-energized, required actions are to declare Division 1 DC electrical power subsystem inoperable per ITS 3.8.4 and enter Required Actions A (A.1, A.2, and A.3) and B.1, with actions to be completed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The definition of 125VDC electrical power system is that it consists of 4 independent subsystems, each containing a battery, associated battery charger, and all associated control equipment and interconnecting cabling. This information is described in ITS B3.8.4.

Part 2 Per ITS 3.0.4, when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

  • When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;
  • After performance of a risk assessment addressing inoperable systems and Test ID: 343246 12/02/2020 3 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

  • When an allowance is stated in the individual value, parameter, or other Specification.

This specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

Since Conditions A and B both have 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> completion time limits and the LCO does not specify that 3.0.4 does NOT apply, then entering Mode 1 is NOT permitted by Technical Specifications.

Incorrect Responses:

A is incorrect but plausible. This answer is partially correct in that the Div 1 Battery Charger has tripped requiring entry into ITS 3.8.4 Condition A. Part B is plausible because ITS does permit mode changes for certain equipment inoperabilities.

B is incorrect but plausible because ITS does permit mode changes for certain equipment inoperabilities.

C is incorrect but plausible. This answer is partially correct in that the Div 1 Battery Charger has tripped requiring entry into ITS 3.8.4 Condition A.

Question Information The unit was operating in Mode 2 with NO LCOs in effect when multiple annunciators were Topic received on User ID CL-ILT-N19076 System ID 2147008 Status Active Point Value 1.00 Time (min) 1 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I References Provided None Question meets the KA because the examinee must be K/A Justification able to determine the extent of the loss of DC using the indications provided in the stem.

SRO-Only Justification 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 a high cog question written at the analysis and comprehension level. The examinee has to analyze indications in a graphic and then determine required actions based on that analysis (3-SPK).

Test ID: 343246 12/02/2020 4 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key NRC Exams Only Question Type Bank (CL-ILT-N12076) Difficulty NA Technical Reference and Revision #

  • ITS 3.8.4 (3.8-24) Amend. 187
  • ITS 3.0.4 (3.0-1) Amend. 220
  • ITS 3.0.4 (3.0-2) Amend. 213 Training Objective 263000.12 Given BATTERY & DC DISTRIBUTION 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.

.1 AC Electrical Distribution Previous NRC Exam Use ILT 12-1 NRC Exam K/A Reference(s) 295004.AA2.02 Safety Function 6 Tier 1 Group 1 RO Imp: 3.5 SRO Imp: 3.9 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C.

POWER : (CFR: 41.10 / 43.5 / 45.13)

Extent of partial or complete loss of D.C. power Learning Objective(s)

Q1/76 295004 AA2.02 User (Sys) ID N/A (1537896)

Cross Reference Links None Test ID: 343246 12/02/2020 5 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 2 ID: 2153670 Points: 1.00 The plant is in Mode 4 with the Residual Heat Removal (RHR) System in Shutdown Cooling Mode.

THEN, shutdown cooling is lost.

  • Reactor vessel pressure is 12 psig and slowly rising.

When is the earliest off-site notification required?

A. No off-site notification is required.

B. NRC Operations Center within 15 minutes.

C. State and local authorities within 15 minutes.

D. NRC Operations Center within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Answer C Answer Explanation C is correct.

An unplanned RCS pressure rise > 10 psig as a result of temperature increase is an Alert (CA5). Per EP-MW-114-100 Midwest Region Off-Site Notifications, state and local notification must be completed within 15 minutes of the declaration of the emergency. Mode 4 requires the reactor vessel to be depressurized, so a pressure rise to 12 psig means pressure has increased by 12 psig from 0 psig.

Incorrect responses:

A is incorrect but plausible. This response is plausible because the stem conditions given do not meet the threshold for CU5 (no temperature was specified). However, CA5 was exceeded, which requires EAL declaration and notification.

B is incorrect but plausible. NRC Operation Center notification is required, but per LS-AA-1020 Reportability Tables and Decision Trees (pg. 1), the NRC must be notified "immediately by ENS after notification of State and local agencies, but within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of declaration of Emergency Class.

D is incorrect but plausible. LS-AA-1020 (p. 6) requires NRC notification via ENS within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> due to the RHR failure, but this is not the earliest required notification.

Test ID: 343246 12/02/2020 6 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Information The plant is in Mode 4 with the Residual Heat Removal (RHR) System in Shutdown Cooling Mode.

Topic T

User ID CL-ILT-N19077 System ID 2153670 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference OPEN Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided EP-AA-1003 Addendum 3 (pg CL-2-17) Rev. 2 K/A Justification Question meets the KA because the candidate must evaluate loss of shutdown cooling given in the stem and determine reporting requirements to external agencies.

Question is linked to SRO-only Task 996666.11 (Analyze SRO-Only Justification conditions to determine if NRC Notifications are required per 10CFR50.72, 10CFR50.73, and 10CFR20). Also 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 and comprehension level. The candidate must interpret the indications provided and recognize that an Alert declaration is required, and then determine the off-site notification requirements (3-SPK).

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

  • CPS 4301.01 Rev. 16b
  • LS-AA-1020 Rev. 31 Training Objective LP85804.2.4.30Knowledge of events related to system operation / status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Previous NRC Exam Use None Test ID: 343246 12/02/2020 7 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s)

B2.4.30 Safety Function 6 Tier 3 Group RO Imp: 2.7 SRO Imp: 4.1 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

(CFR: 41.10 / 43.5 / 45.11)

GS.295021 Safety Function 4 Tier 1 Group 1 RO Imp: SRO Imp:

Loss of Shutdown Cooling Learning Objective(s)

Q2/77 295021 2.4.30 User (Sys) ID N/A (1537897)

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 2149762 Test ID: 343246 12/02/2020 8 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Test ID: 343246 12/02/2020 9 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 3 ID: 2147011 Points: 1.00 The plant is in Mode 5 with core alterations in progress.

  • Spent fuel is being temporarily stored in the upper containment pool racks.
  • Continuous Containment Purge (CCP) is operating in Unfiltered mode and Fuel Building Ventilation (VF) is operating normally.

At 0100, the following annunciators were received:

  • 5067-3L Hi-Hi Level Drywell Sump Equip/Flr Drn
  • 5067-4L High Flow Drywell Equip Drn
  • 5068-7A Refuel Bellows Leakage At 0107, MCR AR/PR LAN alarms (YELLOW tiles) with values slowly trending up for:
  • Fuel Building Exhaust Radiation Monitors 1RIX-PR006A, B, C, D At 0110, the Refuel SRO reports Upper Containment Pool level is at the top of the Reactor Vessel/Steam Dryer Pools Weir Wall and lowering.

(1) Which action is required to be performed?

(2) Given the following excerpt from EP-AA-1003 Addendum 3 Emergency Action Levels for Clinton Station, what is the correct emergency classification for this event?

RA2 Significant lowering of RU2 Unplanned loss of water water level above, or level above irradiated fuel damage to, irradiated Emergency Action Level (EAL):

fuel. 1. a. UNPLANNED water level Emergency Action Level (EAL): drop in the REFUELING

1. Uncovery of irradiated fuel in PATHWAY as indicated by ANY the REFUELING PATHWAY. of the following:

OR

  • Refueling Cavity water
2. Damage to irradiated fuel level <22 ft. 8 in. above the Table R1 resulting in a release of Reactor Vessel Flange Fuel Handling Incident radioactivity from the fuel as OR Radiation Monitors indicated by ANY Table R1
  • Spent Fuel Pool or Upper
  • Fuel Building Exhaust Radiation Monitoring reading Containment Fuel Storage (1PR006A-D) >1000 mRem/hr. Pool water level < 23 ft.
  • Containment Exhaust 3. Lowering of spent fuel pool
  • Indication or report of a (1PR001A-D) level to 11.00 ft. as indicated on drop in water level in the
  • Containment Fuel xfer 1LI-FC221A(B). REFUELING PATHWAY Plenum (1PR008A-D) AND
b. UNPLANNED Area Radiation Monitor reading rise on one or more radiation monitor in Table R1.

Test ID: 343246 12/02/2020 10 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key A. (1) Establish secondary containment.

(2) RU2 B. (1) Establish secondary containment.

(2) RA2 C. (1) Secure the operating Shutdown Cooling loop.

(2) RU2 D. (1) Secure the operating Shutdown Cooling loop.

(2) RA2 Answer A Answer Explanation A is correct.

Part 1 Explanation:

Failure of the refueling bellows will result in draining the upper containment pools to the Drywell Equipment Drain system and reducing the level of water/shielding above the fuel stored in the upper containment pool racks and the RPV.

Receipt of annunciators 5067-3L and 5068-7A are symptoms of reactor cavity leakage during refueling requiring entry into CPS 4011.01 Reactor Cavity Leakage During Refueling.

If level approaches the top of any irradiated fuel, level cannot be restored, or high radiation levels exist, CPS 4011.01 Reactor Cavity Leakage During Refueling, subsequent action step 4.4 directs the following actions:

  • Enter EOP-1 RPV Control (use ECCS to restore level above the fuel)
  • Prohibit access to CNMT Refueling Floor
  • Establish Secondary CNMT integrity Part 2 Explanation:

Based on plant conditions presented in the stem:

  • Normal Upper Containment Pool level is 827' 3" and was reported at the top of the Reactor Vessel/Steam Dryer Pools Weir Wall (~ 827' 1"), signifying an UNPLANNED drop in water level in the REFUELING PATHWAY
  • CCP Exhaust Rad Monitors (1RIX-PR042A-D) and FB Exhaust Rad Monitors (1RIX-PR06A-D) alarming at the ALERT level (Yellow tiles imply that PR042A-D read ~ 5 mRem/hr and PR006A-D read ~ 2 mRem/hr)) and slowly trending up, signifying an UNPLANNED Area Radiation Monitor reading rise on one or more radiation monitor (in Table R1)

Per EP-AA-1003 Addendum 3 Radiological Emergency Plan Annex for Clinton Station Cold Shutdown/Refueling Matrix for Abnormal Rad Levels / Radiological Effluents, the conditions for UNUSUAL EVENT RU2 Unplanned loss of water level above irradiated fuel has been met.

Incorrect Responses:

B is incorrect but plausible. The first part of this response is correct. The second part of this answer Test ID: 343246 12/02/2020 11 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key would be correct if the irradiated fuel in the REFUELING PATHWAY had been UNCOVERED (NOT reported by Refuel SRO) or the alarming Rad Monitors (in Table R1) were reading >1000 mRem/hr (based on the Yellow tiles) indicating damage to irradiated fuel, leading the candidate to the conclusion that the conditions for ALERT RA2 Significant lowering of water level above, or damage to, irradiated fuel had been met. However, since neither of these conditions were met, UNUSUAL EVENT RU2 Unplanned loss of water level above irradiated fuel has been met, is the appropriate call.

C is incorrect but plausible. The first part of this answer would be correct if the leakage source had not been identified. Then the proper course of action would be CPS 4011.01 step 4.2 which directs securing the operating shutdown cooling RHR pump(s). However, since the source of the leakage is known (failure of the refueling bellows), actions per CPS 4011.01 Reactor Cavity Leakage During Refueling (establish secondary containment integrity) is appropriate. The secons part of this response is correct.

D is incorrect but plausible. This answer would be correct if:

  • the leakage source had not been identified. Then the proper course of action would be CPS 4011.01 step 4.2 which directs securing the operating shutdown cooling RHR pump(s), and
  • the irradiated fuel in the REFUELING PATHWAY had been UNCOVERED (NOT reported by Refuel SRO) or the alarming Rad Monitors (in Table R1) were reading >1000 mRem/hr (based on the Yellow tiles) indicating damage to irradiated fuel, leading the candidate to the conclusion that the conditions for ALERT RA2 Significant lowering of water level above, or damage to, irradiated fuel had been met.

Question Information The plant is in Mode 5 with core alterations in progress.

Topic Spent fuel is being temporarily stor User ID CL-ILT-N19078 System ID 2147011 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 Question meets the KA because given a refuel accident, the examinee must be able to evaluate/interpret the conditions presented in the stem and make operational judgements leading to an emergency classification.

SRO-Only Justification Question linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during emergency situations and 10CFR55.43(b)(7) for SRO Fuel handling facilities and procedures. Also linked to SRO only task 999999.12 Direct refueling activities as refueling supervisor.

Additional Information Question is high cog, written at the analysis/

comprehension level. The candidate must analyze the plant conditions provided in the stem, determine the appropriate subsequent actions to be taken as well as an emergency classification based on that analysis (3-SPK/SPR).

Test ID: 343246 12/02/2020 12 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key NRC Exams Only Question Type Bank (CL-ILT-N15081) Difficulty N/A Technical Reference and Revision #

  • CPS 5067.03 (3L) Rev. 32c
  • CPS 5067.04 (4L) Rev. 31
  • CPS 5068.07 (7A) Rev. 24
  • CPS 5140.63 Rev. 1c
  • CPS 5140.65 Rev. 0d
  • CPS 4011.01 Rev. 5b
  • CPS 9000.02D001 Rev. 38a Training Objective PB401101.01.01 Given specific plant conditions, determine if CPS No. 4011.01, REACTOR CAVITY LEAKAGE DURING REFUELING, should be used.

LP87537.01.10Given 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

.06 Loss of Shutdown Systems

.07 Reactor Scram

.08 Electrical Power Failure

.09 Control Room Events

.10 Fire

.11 Security Events

.12 Natural Phenomenon

.13 Contaminated Injury

.14 Other Hazardous Conditions

.15 Loss of Annunciators

.16 Failure to meet Tech Spec Action Statement Previous NRC Exam Use ILT 15-1 NRC Exam K/A Reference(s) 295023.AA2.01 Safety Function 8 Tier 1 Group 1 RO Imp: 3.6 SRO Imp: 4.0 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS : (CFR: 41.10 / 43.5 /

45.13)

Area radiation levels Learning Objective(s)

Q3/78 295023 AA2.01 User (Sys) ID N/A (1537898)

Cross Reference Links None Test ID: 343246 12/02/2020 13 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 4 ID: 2157241 Points: 1.00 Plant conditions:

  • MODE Switch is in SHUTDOWN
  • Multiple rods failed to insert
  • All injection systems except for SLC, RCIC and CRD are terminated and prevented
  • 7 ADS SRVs are opened In accordance with ____(1)____, the CRS should direct the 'A' RO to commence injection into the RPV once reactor pressure first drops below the ____(2)____.

A. (1) EOP-1A (2) Decay Heat Removal Pressure B. (1) EOP-1A (2) Minimum Steam Cooling Pressure C. (1) EOP-3 (2) Decay Heat Removal Pressure D. (1) EOP-3 (2) Minimum Steam Cooling Pressure Answer B Answer Explanation B is correct. Per CPS 4100.01 Reactor Scram, Shutdown Criteria is met when:

  • All rods are inserted to position 00 (Full In) or position 02, OR
  • No more than 1 rod out beyond position 02, with all other rods in at 00, OR
  • A qualified Reactor Engineer has determined that the reactor will remain shutdown under all conditions without boron.

Based on the conditions presented in the stem (multiple rods beyond position 02), shutdown criteria is NOT met. Per EOP-1 RPV Control, the CRS would have transitioned from EOP-1 to EOP-1A ATWS RPV Control.

Again, based on conditions presented in the stem, the CRS transitioned from EOP-1A to EOP-3 Emergency RPV Depressurization (Blow Down) and has initiated a Blow Down (as evidenced by the 7 open SRVs)

Per EOP-3, once the 7 SRVs are open, the CRS should re-enter EOP-1A and wait until RPV pressure lowers below the value in Table J. Per the EOP Technical Bases, as pressure lowers Adequate Core Cooling (ACC) is maintained as long as reactor pressure remains above the Minimum Steam Cooling Pressure (MSCP). At pressures above this value, the steam flow provides sufficient cooling to maintain ACC. Once reactor pressure drops below this value, injection must be restored.

Test ID: 343246 12/02/2020 14 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Incorrect Responses:

A is incorrect but plausible. Once the SRVs are open the CRS must transition back to EOP-1A, but does not depressurize to the Decay Heat Removal Pressure prior to commencing injection. Per the EOP Technical Bases (pg 7-16), when continuing in EOP-3 (non-ATWS condition) AND less than 5 SRVs open, "depressurized" is defined to be an RPV pressure less than 52 psig, the Decay Heat Removal Pressure.

C is incorrect but plausible. This response is plausible because the conditions given in the stem note that a blowdown is in progress per EOP-3. However, re-entry to EOP-1A is required once 7 SRVs are open.

Additionally, injection is required when pressure drops below the Minimum Steam Cooling Pressure.

D is incorrect but plausible. This response is plausible because the conditions given in the stem note that a blowdown is in progress per EOP-3. However, re-entry to EOP-1A is required once 7 SRVs are open.

The second part of the response is correct.

Question Information Plant conditions:

  • MODE Switch is in SHUTDOWN Topic Multiple rods failed to insert
  • All injection sy User ID CL-ILT-N19079 System ID 2157241 Status Active Point Value 1.00 Time (min) 2 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 recognize that based on plant conditions RPV water level is below -162 inches (TAF) and therefore Adequate Core Cooling can no longer be maintained by RPV water level.

Therefore, the candidate must understand how Adequate Core Cooling is determined based on RPV pressure.

SRO-Only Justification Question linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency conditions. Requires knowledge of specific procedure content to perform Emergency RPV Depressurization and Recovery during ATWS conditions. Also linked to SRO only task 440701.04 (Direct actions to Emergency Depressurize per EOP-3 during ATWS).

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

NRC Exams Only Question Type Bank (CL-ILT-15082) Difficulty N/A Test ID: 343246 12/02/2020 15 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Technical Reference and Revision #

  • CPS 4407.01 Rev 30
  • CPS 4404.01 Rev 30
  • EOP Tech Bases Rev 7 Training Objective LP87550.01.01 Recall the definition and bases for the following EOP Variables and Curves:.01 Adequate Core Cooling

.02 Shutdown Criteria

.25 Minimum Steam Cooling Pressure Previous NRC Exam Use ILT 15-1 NRC K/A Reference(s) 295024.EA2.03 Safety Function 5 Tier 1 Group 1 RO Imp: 3.8 SRO Imp: 3.8 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.10 / 43.5

/ 45.13)

Suppression pool level Learning Objective(s)

Q4/79 295031 EA2.03 User (Sys) ID N/A (1537899)

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 2147210 Test ID: 343246 12/02/2020 16 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 5 ID: 2147286 Points: 1.00 The plant is operating at rated thermal power (RTP) with NO testing in progress.

THEN, the following annunciators were received:

  • 5067-8L SRV MONITORING SYSTEM TROUBLE
  • 5004-3F SPDS CSF ALARM (Suppression Pool Temperature alarming 2°F above the setpoint)

Appropriate operator actions have been taken.

If Suppression Pool temperature rises an additional 3°F over the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, entry into CPS

____(1)____ is required.

This action is taken to prevent exceeding the Design Basis maximum allowable value for ____(2)____

temperature.

A. (1) 3006.01 Unit Shutdown (2) Drywell B. (1) 4100.01 Reactor Scram (2) Drywell C. (1) 3006.01 Unit Shutdown (2) Primary Containment D. (1) 4100.01 Reactor Scram (2) Primary Containment Answer C Answer Explanation C is correct.

Part 1:

  • Per ITS 3.6.2.1 Suppression Pool Average Temperature Required Action B.1, if suppression pool average temperature cannot be restored to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> then THERMAL POWER must be reduced 1% RTP within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> requiring entry into CPS 3006.01 Unit Shutdown.

Part 2:

  • Per ITS B3.6.2.1 Suppression Pool Average Temperature, maintaining suppression pool temperature below the LCO limit (95°F) is required to assure that the primary containment conditions assumed for the safety analyses are met. This limitation subsequently ensures that peak primary containment pressures and temperatures do not exceed maximum allowed Test ID: 343246 12/02/2020 17 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key values during a postulated DBA or any transient resulting in heatup of the suppression pool.

Incorrect Responses:

A is incorrect but plausible. This answer is partially correct in that the plant must be shutdown, however this action is taken to prevent exceeding the Design Basis maximum allowable value for Primary Containment temperature. Per ITS 3.6.5.5 Drywell Air Temperature, maintaining drywell air temperature below the LCO limit (150°F) is required to assure that the drywell conditions assumed for the safety analyses are met. This limitation ensures that the peak LOCA drywell temperature does not exceed the maximum allowable temperature of 330°F.

B is incorrect but plausible:

  • Part 1 - ITS requires the plant to be scrammed if suppression pool average temperature is >

110°F but 120°F (ITS 3.6.2.1 D.1). This answer is incorrect because SP temperature has not yet reached this value (currently at 100°F).

  • Part 2 - ITS bases requires maintaining drywell air temperature below the LCO limit (150°F) to assure that the drywell conditions assumed for the safety analyses are met. This answer is incorrect because drywell temperature has not exceeded the LCO limit.

D is incorrect but plausible. ITS requires the plant to be scrammed if suppression pool average temperature is > 110°F but 120°F (ITS 3.6.2.1 D.1). This answer is incorrect because SP temperature has not yet reached this value (currently at 100°F).

Question Information The plant is operating at rated thermal power (RTP) with NO testing in progress.

Topic THEN, the follo User ID CL-ILT-N19080 System ID 2147286 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.2 Facility operating limitations in the technical specifications and 10CFR55 Content their bases.

References Provided None K/A Justification Question meets the KA because the examinee must demonstrate the knowledge of ITS LCO bases with regard to Suppression Pool high water temperature to determine the correct response.

SRO-Only Justification This 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 Question is high cog written at the analysis and comprehension level. The examinee has to analyze the conditions presented in the stem and then determine the appropriate LCO/basis based on ITS (3-SPK).

Test ID: 343246 12/02/2020 18 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key NRC Exams Only Question Type New Difficulty N/A Technical Reference and Revision #

  • ITS 3.6.2.1 (3.6-28/29) Amend. 95
  • ITS B3.6.2.1 (3.6-49) Rev. 20-2 Training Objective 223001.14 Given PRIMARY CONTAINMENT System operability status and a copy of Tech Specs, DISCUSS the bases for the PRIMARY CONTAINMENT System Tech Spec LCO, related safety limits and Limiting Safety System Settings.

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

B2.2.25 Safety Function 5 Tier 3 Group RO Imp: 3.2 SRO Imp: 4.2 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2)

GS.295026 Safety Function 5 Tier 1 Group 1 RO Imp: SRO Imp:

Suppression Pool High Water Temperature Learning Objective(s)

Q5/80 295026 2.2.25 User (Sys) ID N/A (1537900)

Cross Reference Links None Test ID: 343246 12/02/2020 19 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 6 ID: 2147326 Points: 1.00 A transient has occurred requiring entry into EOP-6 Primary Containment Control.

Given the detail below:

In which of the following scenarios is Containment Spray permitted?

RPV CNMT CNMT Pressure Level Temperature Scenario 1 -92" 140°F 1.5 psig Scenario 2 -92" 150°F 3.0 psig Scenario 3 -102" 140°F 1.5 psig Scenario 4 -102" 150°F 3.0 psig A. Scenario 1 ONLY B. Scenario 2 ONLY C. Scenario 1 AND 3 D. Scenario 2 AND 4 Test ID: 343246 12/02/2020 20 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Answer B Answer Explanation B is correct. Per EOP-6 PRIMARY CONTAINMENT CONTROL and the EOP Tech Bases, containment sprays are initiated when containment pressure and temperature is within the "OK TO SPRAY" (unshaded) region of Figure O Containment Spray Initiation Limit. The note states "Do not use RHR pumps you need for core cooling".

Per OP-CL-101-111-1001 Strategies for Successful Transient Mitigation, Alignment of Systems Needed For Adequate Core Cooling (ACC): 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.

With containment pressure and temperature in the "OK TO SPRAY" region of Figure O and RPV level above the trigger point of -100 inches, the RHR pumps are not needed for core cooling and containment spray IS permitted.

Incorrect Responses.

A is incorrect but plausible. This answer would be correct if containment sprays were allowed within the shaded region of Figure O Containment Spray Initiation Limit. However, containment sprays are initiated when containment pressure and temperature is within the "OK TO SPRAY" (unshaded) region of Figure O.

C is incorrect but plausible. This answer would be correct if:

  • RPV level in Scenario 1 was above the trigger point of -100 inches, AND

D is incorrect but plausible. Containment spray is permitted in Scenario 2. This answer would be correct if RPV level in Scenario 4 was above the trigger point of -100 inches. However, with RPV level below the trigger point of -100 inches, RHR pumps ARE needed for core cooling in Scenario 4.

Test ID: 343246 12/02/2020 21 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Information A transient has occurred requiring entry into EOP-6 Primary Containment Control.

Topic Given the detai User ID CL-ILT-N19081 System ID 2147326 Status Active Point Value 1.00 Time (min) 1 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 interpret containment pressure with regard to high containment temperature parameters presented in the stem and determine correct procedural actions to answer the question.

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 candidate must analyze the data provided in the stem and then determine required actions based on knowledge of EOP-6 Primary Containment Control (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-N15075) Difficulty N/A

  • CPS 4402.01 Rev. 30 Technical Reference and Revision #
  • OP-CL-101-111-1001 Rev. 15d LP87558.01.08 Given a diagram of EOP-6, explain the Training Objective use and/or function of the following inserts:

.07 Figure O, Containment Spray Initiation Limit Previous NRC Exam Use ILT 15-1 NRC Exam K/A Reference(s) 295027.EA2.02 Safety Function 5 Tier 1 Group 1 RO Imp: 3.7 SRO Imp: 3.7 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 Test ID: 343246 12/02/2020 22 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Learning Objective(s)

Q6/81 295027 EA2.02 User (Sys) ID N/A (1537901)

Cross Reference Links None Test ID: 343246 12/02/2020 23 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 7 ID: 2147327 Points: 1.00 The plant is operating at rated thermal power. Fuel inspections are in progress in the Fuel Building.

THEN, a spent fuel bundle is severely damaged.

  • The Fuel Building Ventilation (VF) system failed to isolate automatically and manually. The VF supply and exhaust fans are OFF and the Standby Gas Treatment System (SGTS) is in Standby.
  • The radiation release rate from the Fuel Building has exceeded the Emergency Plan UNUSUAL EVENT level.

The following additional conditions are present:

  • All Fuel Building Exhaust Vent Plenum Monitors indicate UPSCALE.
  • Fuel Building Fuel Pool Cooling Pump Room Survey indicates 500 Rem/hr.
  • Fuel Building Fuel Pool Cooling Heat Exchanger Room Survey indicates 700 Rem/hr.
  • Fuel Building General Area Elevation 737' Survey indicates 20 Rem/hr.

Below is Table U from EOP-8, Secondary Containment Control.

U Area Radiation Limits Area Method Max Normal Max Safe Fuel Pool Clg Heat Exch Rm Survey 100 mR/hr 400 R/hr Fuel Bldg Gen Area EL 712' Survey 2.5 mR/hr 25 R/hr Fuel Bldg Pipe Valve Room Survey 10 mR/hr 400 R/hr Fuel Bldg Fuel Pool Clg Pmp Rm Survey 20 mR/hr 400 R/hr Fuel Bldg Gen Area EL 737' Survey 2.5 mR/hr 25 R/hr Which of the following actions must be performed NEXT?

A. Perform a normal reactor plant shutdown.

B. Scram and perform an Emergency RPV Depressurization.

C. Scram and depressurize the RPV rapidly using Main Turbine Bypass Valves.

D. Verify Turbine Building Ventilation (VT) is in operation OR restart VT if necessary.

Answer A Answer Explanation A is correct:

Per CPS 4406.01 EOP-8 Secondary Containment Control, when two or more areas (same parameter) are Test ID: 343246 12/02/2020 24 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key above max safe values, EOP-8 requires the MCR to shutdown the reactor per 3006.01 Unit Shutdown.

A blowdown is only required if a primary system is discharging into the secondary containment in EOP-8, or if a primary system is discharging outside the primary and secondary containments and off-site release rates are above EP-AA-1003 Emergency Plan Radiological Effluent "Alert" level in EOP-9.

Incorrect Responses:

B is incorrect but plausible. This answer would be correct if:

  • two (2) or more areas of the same parameter above max safe.

However, there is NO primary system discharging to the secondary containment, so with two areas of the same parameter above max safe EOP-8 requires the MCR to shutdown the reactor per CPS 3006.01 Unit Shutdown.

C is incorrect but plausible. This answer would be correct if:

  • one (1) area above max safe and anticipating a second area (same parameter) to rise above max safe.

However, there is NO primary system discharging to the secondary containment with two areas of the same parameter above max safe. EOP-8 requires the MCR to shutdown the reactor per 3006.01 Unit Test ID: 343246 12/02/2020 25 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Shutdown.

D is incorrect but plausible. This answer would be correct if the off-site release rate from the Fuel Building (FB) was above the EP-AA-1003 Emergency Plan Radiological Effluent "Alert" Level.

The first action of EOP-9 is:

  • IF Turbine Building Ventilation (VT) is shutdown
  • THEN Restart Turbine Building Ventilation.

Question Information The plant is operating at rated thermal power. Fuel inspections are in progress in the Fuel Topic Buildin User ID CL-ILT-N19082 System ID 2147327 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the examinee must analyze/verify instrument setpoints are greater than Area Radiation Limits (concurrent with an off-site release) and take action to mitigate those conditions presented in the stem.

SRO-Only Justification Question is linked to SRO Only task 440601.02 Respond to a Secondary Containment Control Emergency per EOP-8, 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 candidate must analyze the data provided in the stem and then determine required actions based on knowledge of EOP-8 Secondary Containment Control (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-A11079) Difficulty N/A Technical Reference and Revision #

  • CPS 4406.01 Rev. 30 N-CL-OPS-DB-LP87559.01.07Given EOP-8 and the Training Objective following increasing Secondary Containment parameters, state when a Reactor Scram is required:

.01 Temperature (Table T).

.02 Radiation Level (Table U)

.03 Water Level (Table W)

Test ID: 343246 12/02/2020 26 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Previous NRC Exam Use None K/A Reference(s)

GS.295038 Safety Function 9 Tier 1 Group 1 RO Imp: SRO Imp:

High Off-Site Release Rate B2.4.50 Safety Function 9 Tier 3 Group RO Imp: 4.2 SRO Imp: 4.0 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

(CFR: 41.10 / 43.5 / 45.3)

Learning Objective(s)

Q7/82 295038 2.4.50 User (Sys) ID N/A (1537902)

Cross Reference Links None Test ID: 343246 12/02/2020 27 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 8 ID: 2148786 Points: 1.00 A reactor scram has occurred.

  • NO other EOPs have been entered.
  • RPV pressure and level stabilization efforts are in progress using ONLY RCIC and SRVs.
  • Opening SRVs has resulted in multiple trips of the RCIC Turbine.

Under these conditions, which of the following describes the maximum RPV level and pressure bands that are permitted to be established?

(1) level band (2) pressure band A. (1) +8.9 to +52 inches (2) 600 - 1065 psig B. (1) +8.9 to +52 inches (2) 800 - 1065 psig C. (1) -30 to +40 inches (2) 800 - 1065 psig D. (1) -30 to +40 inches (2) 600 - 1065 psig Answer D Answer Explanation D is correct.

Per OP-CL-101-111-1001 Strategies For Successful Transient Mitigation EOP 1 RPV Control, direct initial RPV:

  • level band of Level 3 (8.9 in.) to Level 8 (52 in.) IAW CPS 4411.03
  • pressure band of 800-1065 psig IAW CPS 4411.09 Per CPS 4100.01 Reactor Scram section 4.3 Level Control Actions / 4.4 Pressure Control Actions:
  • stabilize RPV level Level 3 (8.9 in.) to Level 8 (52 in.)
  • stabilize RPV pressure < 1065 psig
  • if RPV level can not be stabilized, then expand level band to -30 to +40 in. Wide Range with a target of 0 in. to +10 in. Wide Range. Expanding the pressure band to 600 - 1065 psig will assist with pressure/level coordination efforts.

Although, OP-CL-101-111-1001 Strategies For Successful Transient Mitigation EOP 1 RPV Control points out that use of the expanded pressure band is not required to be used concurrent with using an Test ID: 343246 12/02/2020 28 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key expanded level band; however:

  • expanding the pressure band may assist with level control actions, and
  • the maximum permitted pressure band would be the expanded pressure band Based on plant conditions presented in the stem (RPV can not be stabilized), the expanded level and pressure bands should be utilized.

Incorrect Responses:

A is incorrect but plausible. The first part of this answer would be correct if RPV level could be stabilized per CPS 4100.01 Level Control Actions and OP-CL-101-111-1001 Strategies For Successful Transient Mitigation. However, based on plant conditions presented in the stem (RPV can not be stabilized), the expanded level band should be utilized. The second part of the question is correct.

B is incorrect but plausible. This answer would be correct if RPV level could be stabilized per CPS 4100.01 Level Control Actions and OP-CL-101-111-1001 Strategies For Successful Transient Mitigation.

However, based on plant conditions presented in the stem (RPV can not be stabilized), the expanded level and pressure bands should be utilized.

C is incorrect but plausible. The first part of the answer is correct. The second part of this answer would be correct if RPV level could be stabilized per CPS 4100.01 Level Control Actions and OP-CL-101-111-1001 Strategies For Successful Transient Mitigation. However, based on plant conditions presented in the stem (RPV can not be stabilized), the expanded pressure band should be utilized in conjunction with the expanded level band.

Question Information A reactor scram has occurred.

Topic EOP-1 RPV Control has been entered.

NO other EOPs have been enter User ID CL-ILT-N19083 System ID 2148786 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 10 CFR 55.43 SRO WRITTEN EXAMINATION References Provided None K/A Justification Question meets the KA because the examinee has to analyze the conditions in the stem, and then determine that the RPV level and pressure band must be expanded to assist with RPV pressure and level coordination efforts.

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.

Additional Information Question is low cog written at the memory level. The Test ID: 343246 12/02/2020 29 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key candidate must recall the pressure and level bands prescribed by CPS procedures (1-F).

NRC Exams Only Question Type Bank (CL-ILT-A14085) Difficulty N/A

  • CPS 4401.01 Rev. 30 Technical Reference and Revision #
  • CPS 4100.01 Rev. 23f
  • OP-CL-101-111-1001 Rev. 15d 410001.01) respond to the following Reactor Scram Training Objective conditions:

.2 Turbine/Generator trip without bypass valves.

Previous NRC Exam Use None K/A Reference(s) 295008.AA2.01 Safety Function 2 Tier 1 Group 2 RO Imp: 3.9 SRO Imp: 3.9 Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL : (CFR: 41.10 /

43.5 / 45.13)

Reactor water level Learning Objective(s)

Q8/83 295008 AA2.01 User (Sys) ID N/A (1537903)

Cross Reference Links None Test ID: 343246 12/02/2020 30 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 9 ID: 2148787 Points: 1.00 The plant is operating at 90% power, THEN a transient occurs.

The 'A' Reactor Operator (RO) observes and reports the following:

  • A step change in Reactor Power to 93%.
  • Reactor recirculation flow, flow control line (FCL), and feedwater temperature are unchanged.

The CRS shall direct the 'A' RO to lower reactor power to the original level using ____(1)____ first.

(2) Who is required to be notified?

A. (1) RR flow (2) Reactor Engineer (RE) only.

B. (1) RR flow (2) Reactor Engineer (RE) and Nuclear Station Engineering Department (NSED).

C. (1) rod insertion (2) Reactor Engineer (RE) only.

D. (1) rod insertion (2) Reactor Engineer (RE) and Nuclear Station Engineering Department (NSED).

Answer A Answer Explanation A is correct.

Per CPS 4007.03 Rod Drop Section 1.0 Symptoms, the information provided in the stem is indicative of a rod drop.

Per CPS 4007.03 Section 4.0 Subsequent Actions:

  • Lower reactor power to the original level using first core flow and then rod insertion.
  • Notify the Reactor Engineer (RE) of the event.

Incorrect Responses:

B is incorrect but plausible. The first part of this response is correct. The second part would be correct if the stem conditions were indicative of core shroud cracking above the top guide, which will also cause an abrupt change in reactor power. However, based on the plant conditions presented in the stem, it is appropriate to enter CPS 4007.03 Rod Drop which requires the Reactor Engineer (RE) only to be notified.

C is incorrect but plausible. The first part of this response is plausible because CPS 3005.01 Unit Power Changes requires rod insertion for an inadvertent MELLLA Limit violation or forced entry into the Controlled Entry Region. However, CPS 4007.03 Rod Drop requires reactor power to be lowered first by Test ID: 343246 12/02/2020 31 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key core flow. The second part is correct.

D is incorrect but plausible.

  • The first part of this response is plausible because CPS 3005.01 Unit Power Changes requires rod insertion for an inadvertent MELLLA Limit violation or forced entry into the Controlled Entry Region. However, CPS 4007.03 Rod Drop requires reactor power to be lowered first by core flow.
  • The second part would be correct if the stem conditions were indicative of core shroud cracking above the top guide, which will also cause an abrupt change in reactor power. However, based on the plant conditions presented in the stem, it is appropriate to enter CPS 4007.03 Rod Drop which requires the Reactor Engineer (RE) only to be notified.

Question Information The plant is operating at 90% power, THEN a transient occurs.

Topic The 'A' Reactor Operator (RO) obse User ID CL-ILT-N19084 System ID 2148787 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 10 CFR 55.43 SRO WRITTEN EXAMINATION References Provided None K/A Justification Question meets the KA because the examinee must interpret the conditions in the stem (to include reactor power) following an Inadvertent Reactivity Addition, and then determine the appropriate actions that the CRS must direct and who he/she must notify.

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.

Additional Information Question is high cog written at the analysis level. The candidate must diagnose the condition given in the stem and then determine appropriate actions (3-SPK).

NRC Exams Only Question Type New Difficulty N/A

  • CPS 4007.03 Rev. 8d Technical Reference and Revision #
  • CPS 4007.02 Rev. 13c
  • CPS 3005.01 Rev. 46 PB400703.01 Training Objective Given specific plant conditions, determine if CPS No.

4007.03, ROD DROP, should be used.

Previous NRC Exam Use None Test ID: 343246 12/02/2020 32 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s) 295014.AA2.01 Safety Function 1 Tier 1 Group 2 RO Imp: 4.1* SRO Imp: 4.2*

Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION : (CFR:

41.10 / 43.5 / 45.13)

Reactor power Learning Objective(s)

Q9/84 295014 AA2.01 User (Sys) ID N/A (1537904)

Cross Reference Links None Test ID: 343246 12/02/2020 33 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 10 ID: 2149626 Points: 1.00 The plant was operating at rated thermal power when annunciator 5013-5D HIGH-HIGH LEVEL FLR/EQUIP DRN SUMP - AUX BLDG was received.

The 'A' RO reports:

  • computer point CM-BC813 RCIC Floor Drn Sump Lvl indicates HI HI, and
  • Suppression Pool level is lowering at 1 inch per minute.

CPS 4304.01 Flooding Table 2 - Suppression Pool Level / ECCS Room Equalization Levels ECCS Room Final Pool Level ECCS Room Final Pool Level HPCS ~ 15' 3" RHR A ~ 15' 5" LPCS ~ 14' 4" RHR B ~ 15' 5" RCIC ~ 16' 2" RHR C ~ 14' 10" If Suppression Pool level lowers to the equalization level, what EOP action is required and what is the reason for that action?

A. Scram to reduce the break flow into the secondary containment.

B. Enter EOP-3 and blowdown to prevent exceeding the heat capacity of the suppression pool.

C. Shutdown the reactor due to the widespread and immediate threat to equipment in the secondary containment.

D. Isolate the discharge into the RCIC Pump Room to protect equipment necessary for safe operation of the plant.

Answer D Answer Explanation D is correct:

Per CPS 4406.01 Secondary Containment Control, an EOP-8 entry condition has been exceeded due to floor drain sump level above max normal (Table V). If floor drain sump level cannot be held below max normal, then the required action is to isolate all discharges into affected areas except systems needed for damage control or EOP actions. Per the EOP-TB page 9-1, EOP-8 Secondary Containment Control is used to protect equipment in the secondary containment.

Additional actions in 4406.01 Secondary Containment Control are not required because:

  • The leak from the Suppression Pool does not constitute a primary leak, precluding the need to perform actions at 32, and
  • multiple areas of the secondary containment are not impacted by the leakage (only the RCIC room) requiring the reactor to be shutdown.

Test ID: 343246 12/02/2020 34 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key EOP-6 Primary Containment Control will also be entered due to Suppression Pool Level below 18.9 feet. Actions are required to initiate normal Suppression Pool makeup methods to hold level above 15.1 ft and to scram and enter EOP-3 blowdown if SP level cannot be maintained above 15.1 ft. Since the equalization level for the RCIC Room is 16' 2", a scram and blowdown will not be required.

Incorrect Responses:

A is incorrect but plausible. Per EOP-TB page 9-11 and 9-12, break flow into the secondary containment is reduced when a primary system is discharging into the secondary containment by inserting a scram.

Since the leak in the stem is not a primary system, inserting a scram to reduce the break flow is not the action or the reason for performing required EOP actions.

B is incorrect but plausible. EOP-6 requires a blowdown to be performed if Suppression Pool level cannot be held above 15'1" to prevent exceeding the heat capacity of the Suppression Pool. Since the equalization level is above 15'1", performing a blowdown to prevent exceeding the heat capacity limit of the suppression pool is not the action or the reason for performing required EOP actions.

C is incorrect but plausible. This answer would be correct if the SP leakage was impacting another area in the secondary containment (cross flooding). EOP-8 requires a reactor shutdown to be performed if 2 or more areas in the secondary containment are above max safe values for level due to non-primary system leakage. EOP-TB page 9-10 states the bases for this action is due to the indications of a wide-spread problem posing a direct and immediate threat to the secondary containment. Since only 1 area is impacted (the RCIC Pump Room), a reactor shutdown is not required by EOP-8.

Question Information The plant was operating at rated thermal power when annunciator 5013-5D HIGH-HIGH LEVEL Topic FLR/EQUIP D User ID CL-ILT-N19085 System ID 2149626 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the examinee must demonstrate the ability to explain and apply system limits and precautions with regard to Secondary Containment High Sump/Area Water Level 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, Test ID: 343246 12/02/2020 35 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key abnormal, and emergency conditions.

Additional Information Question is high cog written at the analysis and comprehension level. The candidate must analyze the conditions provided in the stem and then determine the effect/consequences/mitigation strategies based on that analysis (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-N17085) Difficulty N/A

  • CPS 4304.01 Rev. 6c Technical Reference and Revision #
  • CPS 4402.01 Rev. 30
  • CPS 4406.01 Rev. 30
  • EOP-TB Rev. 7 Training Objective N-CL-OPS-DB-LP87559.01.10 Describe the normal means of control for a slow increase in Secondary Containment Water Levels.

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

GS.295036 Safety Function 5 Tier 1 Group 2 RO Imp: SRO Imp:

Secondary Containment High Sump/Area Water Level B2.4.18 Safety Function 5 Tier 3 Group RO Imp: 3.3 SRO Imp: 4.0 Knowledge of the specific bases for EOPs.

(CFR: 41.10 / 43.1 / 45.13)

Learning Objective(s)

Q10/85 295036 2.4.18 (17-1N)

User (Sys) ID N/A (1537905)

Cross Reference Links None Test ID: 343246 12/02/2020 36 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 11 ID: 2149636 Points: 1.00 The plant is in Mode 2 performing a startup.

  • Heatup and pressurization is complete.
  • CPS 9433.20A ECCS LPCS Pump Discharge Pressure ADS E21-N052 Channel Calibration is in progress.

During the performance of CPS 9433.20A Section 8.1, Functional Test, Analog Trip Module (ATM) ADS LCPS Pump Discharge Pressure E21-N652 failed to trip.

  • ATM E21-N652 receives an input from transmitter E21-N052.
1) Which of the following actions is/are required?
2) Can the plant be transitioned to Mode 1?

A. 1) Restore instrument channel to operable status in 8 days ONLY.

2) Can enter Mode 1.

B. 1) Restore instrument channel to operable status in 8 days ONLY.

2) Can NOT enter Mode 1.

C. 1) Declare ADS valves inoperable in one hour and restore instrument channel to operable status in 8 days.

2) Can enter Mode 1 D. 1) Declare ADS valves inoperable in one hour and restore instrument channel to operable status in 8 days.
2) Can NOT enter Mode 1 Answer B Answer Explanation B is correct.

Per ITS 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation, with function 4.e, LPCS Pump Discharge Pressure - High INOPERABLE, condition A.1 and Condition G must be entered. Required action G.1 is not required to be taken since initiation capability has not been lost in either ADS trip system (Div 1 or 2). Required Action G.2 requires the inoperable channel to be restored within 8 days.

Since the inoperability does not meet the waiver requirements of ITS LCO 3.0.4, transition to Mode 1 is not permitted.

Incorrect Responses:

A is incorrect but plausible. While the first part of the answer is correct, this answer would be correct if Test ID: 343246 12/02/2020 37 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key performing the required actions of ITS 3.3.5.1 condition G met the requirements of ITS 3.0.4. Since the LCO requires the inoperable instrumentation channel to be restored to operable status in 8 days (not an unlimited period of time), LCO 3.0.4 does not permit entry into Mode 1.

C is incorrect but plausible.

  • The first part of this response is plausible because the LPCS Discharge Press instrument is an input to ADS logic. However, a trip of either of the two discharge pressure transmitter ATMs (LPCS or LPCI 'A' for Div 1; LPCI 'B' or 'C' for Div 2) will satisfy the ADS logic. Therefore, ADS valves are NOT rendered inoperable by a single instrument failure per ITS B3.3.5.1.
  • The second part of the answer is also incorrect. The plausible misconception is that the inoperability meets the waiver requirements of ITS LCO 3.0.4 and transition to Mode 1 is permitted, but it does not.

D is incorrect but plausible. The first part of this response is plausible because the LPCS Discharge Press instrument is an input to ADS logic. However, a trip of either of the two discharge pressure transmitter ATMs (LPCS or LPCI 'A' for Div 1; LPCI 'B' or 'C' for Div 2) will satisfy the ADS logic.

Therefore, ADS valves are NOT rendered inoperable by a single instrument failure per ITS B3.3.5.1.

The second part of the response is correct.

Question Information he plant is in Mode 2 performing a startup.

Topic Heatup and pressurization is complete.

CPS 9433.2 User ID CL-ILT-N19086 System ID 2149636 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.2 Facility operating limitations in the technical specifications and 10CFR55 Content their bases.

References Provided

  • ITS 3.3.5.1 K/A Justification Question meets the K/A because it requires the candidate to recognize the implications of the failed LPCS instrument in the event of an accident at this low power level and take action per Technical Specifications to mitigate it.

SRO-Only Justification Question is 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 (instrumentation channel failure) to the reference (technical specifications) to determine the required actions and ability to enter a Mode change (3-SPR).

NRC Exams Only Test ID: 343246 12/02/2020 38 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Type Bank (CL-ILT-A14088) Difficulty N/A Technical Reference and Revision #

  • ITS 3.3.5.1 (3.3-31) Amendment No 95
  • ITS 3.3.5.1 (3.3-37) Amendment No 95
  • ITS 3.3.5.1 (3.3-42) Amendment No 216
  • ITS 3.0.4 (3.0-1) Amendment No 220
  • ITS 3.0.4 (3.0-2) Amendment No 213
  • ITS B3.3.5.1 (3.3-93/94) Rev. 4-8
  • CPS 9433.20A Rev. 0a Training Objective 209001.12 Given LPCS 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.

Previous NRC Exam Use None.

K/A Reference(s)

B2.4.09 Safety Function 5 Tier 3 Group RO Imp: 3.8 SRO Imp: 4.2 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13)

GS.209001 Safety Function 2 Tier 2 Group 1 RO Imp: SRO Imp:

Low Pressure Core Spray System Learning Objective(s)

Q11/86 209001 2.4.9 User (Sys) ID N/A (1537907)

Cross Reference Links None Test ID: 343246 12/02/2020 39 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 12 ID: 2147168 Points: 1.00 The plant is operating at rated thermal power (RTP).

At 0100, the 'D' Area Equipment Operator reports that the circuit breaker on AB MCC 1H for the Standby Liquid Control (SLC) Storage Tank Operating Heater (C41-D002) was discovered in the OFF position.

At 0600, Electrical Maintenance completes repairs and restores power to the SLC Storage Tank Operating Heater (C41-D002).

The following temperatures are recorded:

Time SLC Storage SLC Pump Tank Suction Solution Piping Temperature Temperature 0100 75°F 75°F 0200 74°F 73°F 0300 73°F 73°F 0400 71°F 70°F 0500 70°F 70°F 0600 69°F 70°F 0700 71°F 70°F 0800 73°F 72°F 0900 75°F 75°F Both SLC subsystems are first INOPERABLE at ____(1)____.

Surveillance(s) ____(2)____ must be completed in order to restore both SLC subsystems to an OPERABLE status.

A. (1) 0200 (2) CPS 9915.01 Standby Liquid Control Chemical Sampling ONLY B. (1) 0200 (2) CPS 9915.01 Standby Liquid Control Chemical Sampling AND CPS 9015.02 Standby Liquid Control Injection Operability C. (1) 0600 (2) CPS 9915.01 Standby Liquid Control Chemical Sampling ONLY D. (1) 0600 (2) CPS 9915.01 Standby Liquid Control Chemical Sampling AND CPS 9015.02 Standby Liquid Control Injection Operability Test ID: 343246 12/02/2020 40 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Answer C Answer Explanation C is correct.

Per ITS 3.1.7 Standby Liquid Control (SLC) System Surveillance Requirements:

  • SR 3.1.7.3 Verify temperature of pump suction piping is 70°F
  • SR 3.1.7.5 Verify the concentration of boron in solution is within the limits of Figure 3.1.7 once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored to 70°F Per CPS 9915.01 Standby Liquid Control Chemical Sampling satisfies ITS surveillance requirement (SR) 3.1.7.5 Per CPS 9015.02 Standby Liquid Control Injection Operability once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after pump suction piping temperature is restored to 70°F, perform section 8.2 Flow Path from Storage Tank to Test Tank Verification Test.

Incorrect Responses:

A is incorrect but plausible. The first part of this response would be correct if the minimum temperature specified in ITS 3.1.7 corresponded to the low end of the SLC Storage Tank temperature controller 1C41-TIC-R002 which controls between 75°F and 85°F. However, ITS 3.1.7 specifies a minimum temperature of 70°F. The second part of the response is correct.

B is incorrect but plausible. This response would be correct if:

  • the minimum temperature specified in ITS 3.1.7 corresponded to the low end of the SLC Storage Tank temperature controller 1C41-TIC-R002 which controls between 75°F and 85°F, and
  • SLC pump suction piping temperature had lowered to < 70°F in addition to SLC Storage Tank temperature.

D is incorrect but plausible, The first part of this response is correct. The second part would be correct if SLC pump suction piping temperature had lowered to < 70°F in addition to SLC Storage Tank temperature. However, since SLC pump suction piping temperature did not go below 70°F, the CPS 9015.02 Standby Liquid Control Injection Operability surveillance is not required to be performed.

Test ID: 343246 12/02/2020 41 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Information The plant is operating at rated thermal power (RTP).

Topic At 0100, the 'D' Area Equipment Operator re User ID CL-ILT-N19087 System ID 2147168 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the KA because the candidate must predict/determine at what point the SLC subsystems become inoperable and then what action(s) must be taken to correct/restore the SLC subsystems to an operable status.

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.

Additional Information Question is high cog because the examinee must analyze the data provided in the stem and then determine when a subsystem became inoperable and the actions necessary to restore operability (3-SPK).

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

  • ITS 3.1.7 (3.1-20/21) Amend. 192
  • CPS 3314.01 Rev. 12b
  • CPS 9015.01 Rev. 43a
  • CPS 9015.02 Rev. 39c Training Objective 211000.12 Given STANDBY LIQUID CONTROL 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.

Previous NRC Exam Use None Test ID: 343246 12/02/2020 42 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s) 211000.A2.05 Safety Function 1 Tier 2 Group 1 RO Imp: 3.1 SRO Imp: 3.4 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)

Loss of SBLC tank heaters Learning Objective(s)

Q12/87 211000 A2.05 User (Sys) ID N/A (1537906)

Cross Reference Links None Test ID: 343246 12/02/2020 43 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 13 ID: 2148909 Points: 1.00 The plant is operating at rated thermal power (RTP), THEN a LOCA occurs.

The 'A' Reactor Operator (RO) observes and reports the following:

  • The reactor scrammed due to lowering RPV level.
  • Drywell pressure is 1.50 psig and rising 0.01 psig/minute.

The 'B' Reactor Operator (RO) observes and reports the following:

  • RCIC turbine is tripped.

(1) Which of the following valves should have isolated?

(2) If those valves do NOT automatically isolate, which procedure directs isolation actions?

A. (1) 1E51-F031 and 1E51-F076, RCIC Suppr Pool Suction Valve & Stm Supply Warm Up Isol Valve (2) CPS 3310.01 Reactor Core Isolation Cooling B. (1) 1E51-F077 and 1E51-F078, Outboard and Inboard RCIC Exh Vac Bkr Valves.

(2) CPS 3310.01 Reactor Core Isolation Cooling C. (1) 1E51-F031 and 1E51-F076, RCIC Suppr Pool Suction Valve & Stm Supply Warm Up Isol Valve (2) EOP-1 RPV Pressure Control D. (1) 1E51-F077 and 1E51-F078, Outboard and Inboard RCIC Exh Vac Bkr Valves.

(2) EOP-1 RPV Pressure Control Answer C Answer Explanation C is correct.

Per CPS 4001.02C001 Automatic Isolation Checklist, RCIC Suppr Pool Suction Valve (1E51-F031) and RHR & RCIC Stm Supp Warm Up Isol Valve (1E51-F076) isolate on a low RCIC steam supply pressure, indicated by the 5063-1A and 5063-2A annunciators.

Per CPS 4401.01 EOP-1 RPV Control, the CRS will direct "Verify needed auto actions."

Per CPS EOP Technical Basis (EOP-TB) for EOP-1, Verify means to confirm that necessary responses occur Test ID: 343246 12/02/2020 44 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key and that appropriate system states exist. Actions that should have occurred but did not should be manually performed.

Incorrect Responses:

A is incorrect but plausible. The first part of the distractor is correct. The second part is plausible since the RCIC procedure is normally used to manipulate these valves; however, direction to isolate comes from EOP-1.

B is incorrect but plausible. 1E51-F077 and F078 isolate on the low RCIC steam supply pressure given in the stem, but only with a concurrent high drywell pressure signal. Since drywell pressure remained below the setpoint, these valves would remain open. The second part is also incorrect; the RCIC procedure is normally used to manipulate these valves; however, direction to isolate comes from EOP-1.

D is incorrect but plausible. 1E51-F077 and F078 isolate on the low RCIC steam supply pressure given in the stem, but only with a concurrent high drywell pressure signal. Since drywell pressure remained below the setpoint, these valves would remain open. The second part of the distractor is correct.

Question Information The plant is operating at rated thermal power (RTP), THEN a LOCA occurs.

Topic The 'A' Reactor Operato User ID CL-ILT-N19088 System ID 2148909 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None.

K/A Justification Question meets the K/A because the candidate has to demonstrate the ability to verify that the alarms are consistent with the plant conditions (incomplete RCIC isolation) to answer the question.

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 Question is high cog written at the analysis and comprehension level. The candidate must analyze the conditions in the stem and then determine required actions based on that analysis (3-SPK/SPR).

NRC Exams Only Question Type New Difficulty N/A Test ID: 343246 12/02/2020 45 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Technical Reference and Revision #

  • CPS 4001.02C001 Rev 16c
  • CPS 5063.01 (1A) Rev 29c
  • CPS 5063.02 (2A) Rev 31e Training Objective LP85804.2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

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

B2.4.46 Safety Function 1 Tier 3 Group RO Imp: 4.2 SRO Imp: 4.2 Ability to verify that the alarms are consistent with the plant conditions.

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

Learning Objective(s)

Q13/88 223002 2.4.46 User (Sys) ID N/A (1537908)

Cross Reference Links None Test ID: 343246 12/02/2020 46 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 14 ID: 2147146 Points: 1.00 The plant is operating at rated thermal power, THEN a transient occurs.

The 'B' Reactor Operator (RO) observes and reports the following:

Immediate actions of CPS 4009.01 INADVERTENT OPENING/SAFETY RELIEF VALVE have been completed.

Subsequent actions of CPS 4009.01 up to and including all methods of closing 1B21-F051G have been completed. All attempts to close 1B21-F051G have failed.

A ____(1)____ confirms that the SRV is OPEN and not leaking by.

NEXT, the CRS will direct a ____(2)____ per CPS 3005.01 Unit Power Changes.

A. 1) SRV tailpipe temperature of 260°F

2) Power reduction to < 98%

B. 1) Decrease in generator MWe

2) Power reduction to < 98%

C. 1) SRV tailpipe temperature of 260°F

2) Rapid plant shutdown D. 1) Decrease in generator MWe
2) Rapid plant shutdown Answer D Answer Explanation D is correct:

Per CPS 4009.01 Inadvertent Opening Safety/Relief Valve 1.0 Symptoms, a decrease in Generator MWe output is indicative of an open Safety/Relief Valve (SRV).

Per CPS 4009.01 4.0 Subsequent actions, a leaking SRV tailpipe temperature is typically 230-270°F and is confirmed 'not open' by no acoustic monitor noise and no MWe decrease.

The stem of the question states that immediate actions of CPS 4009.01 have been completed and attempts to close the SRV have failed. CPS 4009.01 4.0 Subsequent actions, step 4.9, requires a Rapid Plant Shutdown per 3005.01, UNIT POWER CHANGES if the SRV remains open.

Test ID: 343246 12/02/2020 47 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Incorrect Responses:

A is incorrect but plausible. The SRV tailpipe temperature of 260°F is significantly higher than normal, but within the range stated in CPS 4009.01 for a leaking SRV. The second part of the distractor is also incorrect; lowering power to < 98% is required but would have already been completed as the first subsequent action of 4009.01.

B is incorrect but plausible. The first part of the distractor is correct. The second is not; lowering power to < 98% is required but would have already been completed as the first subsequent action of 4009.01.

C is incorrect but plausible. The SRV tailpipe temperature of 260°F is significantly higher than normal, but within the range stated in CPS 4009.01 for a leaking SRV. The second part of the distractor is correct.

Question Information The plant is operating at rated thermal power, THEN a transient occurs.

Topic The 'B' Reactor Operator User ID CL-ILT-N19089 System ID 2147146 Status Active Point Value 1.00 Time (min) 3 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the (b) portion of the K/A because the candidate must be able to understand the indications of a stuck open SRV and know which procedure to enter and actions to take to mitigate the problem.

As permitted by ES-401 D.2.a, the (a) portion of the K/A (the low cog portion) is not tested.

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 Question is high cog written at the analysis level. The candidate must determine from the conditions in the stem that an SRV is open and then select the appropriate response (3-SPK).

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

  • CPS 4009.01 Rev. 13b Test ID: 343246 12/02/2020 48 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key

  • CPS 5066.05 (5B) Rev. 28a
  • CPS 5067.08 (8L) Rev. 31a 239001.11 Training Objective EVALUATE given key MAIN STEAM System parameters, if needed DETERMINE a course of action to correct or mitigate the following abnormal condition(s):

.1 Trip Signal

.2 Isolation Signal

.3 Auto Start Signal

.4 Condition/Lineup Previous NRC Exam Use None K/A Reference(s) 239002.A2.03 Safety Function 3 Tier 2 Group 1 RO Imp: 4.1 SRO Imp: 4.2*

Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; 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)

Stuck open SRV Learning Objective(s)

Q14/89 239002 A2.03 User (Sys) ID N/A (1537909)

Cross Reference Links None Test ID: 343246 12/02/2020 49 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 15 ID: 2149651 Points: 1.00 A transient has occurred requiring entry into EOP-2 RPV Flooding.

Shutdown criteria are met.

Which of the following represents a condition where injection flow can be slowed or stopped?

A. Inject with Reactor Core Isolation Cooling (RCIC) until RPV pressure starts to rise.

B. Inject with Condensate / Condensate Booster (CD/CB) Pumps until SRV tailpipe temperatures start to rise.

C. Inject with High Pressure Core Spray (HPCS) until 1E22-F004 HPCS To CNMT Outbd Isln Valve closes at Level 8.

D. Defeat Motor Driven Reactor Feed Pump (MDRFP) Level 8 interlocks and inject with the MDRFP until SRV acoustic monitor alarms and Main Steam Line Flow indications are received.

Answer D Answer Explanation D is correct:

Per the EOP Technical Bases (pg 6-1), EOP-2, RPV flooding, is used to cool the core when RPV water level cannot be determined. The specified actions first depressurize the RPV, then control injection to establish and maintain one of the following conditions:

  • The RPV flooded to the elevation of the main steam lines. The core will then be cooled by full submergence. This condition may ultimately be achieved under either shutdown or failure-to-scram conditions.
  • RPV pressure above the Minimum Steam Cooling Pressure. The core will then be cooled by submergence or steam cooling. Since reactor power must be at least 6%-10% to generate the amount of steam required to sustain the Minimum Steam Cooling Pressure, this condition is applicable only under ATWS conditions.

CPS 1005.09M002 STA/IA Guide Sheet provides methodology for determining when level is at the Main Steam Lines and includes the following:

Assess using multiple and diverse items:

  • Acoustic Monitors (may clear when < MSL)
  • SRV/ADS Tail Pipe Temp (will lower when > MSL; will slowly raise up if < MSL)
  • RPV pressure - ~ 10 psig rise indicates at MSL's; sudden large increase indicates solid.
  • MSL - Flow / Hi Flow dP
  • SP Level increase after initial drop?
  • Drastic step changes on level?

Test ID: 343246 12/02/2020 50 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key

  • Fill Rate vs. Leak/Usage Rate?
  • Not expected to reach MSL when > 7% RTP (SRV cap).

Incorrect Responses:

A is incorrect but plausible. Per CPS 1005.09M002 EOP/Off-Normal Performance Aid Matrix, one indication of water level at the main steam lines is RPV pressure increasing; however, RCIC will not be available following the blowdown with reactor pressure below RCIC turbine stall pressure. This answer is plausible because RCIC is normally a preferred injection system.

B is incorrect but plausible. EOP-2 directs RPV flooding via numerous systems, including Condensate/Condensate Boost; however, per CPS 1005.09M002 EOP/Off-Normal Performance Aid Matrix, rising SRV tailpipe temperatures is indicative that RPV level is below the MSLs. This answer is plausible because lowering tailpipe temperatures would indicate water level at the MSLs.

C is incorrect but plausible. HPCS is a preferred system for RPV flooding per EOP-2; however, conditions in the stem indicate that RPV level is unknown. The Clinton EOP Technical Bases (p. 6-38) states that if RPV water level is "unknown," the inputs to the Level 8 trip logic must be considered invalid.

Question Information A transient has occurred requiring entry into EOP-2 RPV Flooding.

Topic Shutdown criteria are met.

User ID CL-ILT-N19090 System ID 2149651 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None.

K/A Justification Question meets the KA because it requires the candidate to interpret and execute steps of EOP-2 to determine conditions under which injection flow must be throttled to control reactor water level.

SRO-Only Justification Question requires knowledge of specific procedure content to perform RPV Flooding. It is linked to SRO only task 100509.16d (Demonstrate knowledge of symptom based EOP mitigation strategies). Also linked to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency conditions.

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

Test ID: 343246 12/02/2020 51 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key NRC Exams Only Question Type Bank (CL-ILT-N12088) Difficulty N/A

  • CPS 1005.09M002 Rev 9 Technical Reference and Revision #
  • CPS 4403.01 Rev 30
  • CPS EOP-TB Rev 7 Training Objective N-CL-OPS-DB-LP87554.01.03 Given plant conditions determine if RPV Flooding is met, as defined in EOP-2.

Previous NRC Exam Use ILT 12-1 NRC K/A Reference(s)

B2.1.20 Safety Function 3 Tier 3 Group RO Imp: 4.6 SRO Imp: 4.6 Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12)

GS.259002 Safety Function 2 Tier 2 Group 1 RO Imp: SRO Imp:

Reactor Water Level Control System Learning Objective(s)

Q15/90 259002 2.1.20 User (Sys) ID N/A (1537910)

Cross Reference Links None Test ID: 343246 12/02/2020 52 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 16 ID: 2148874 Points: 1.00 The plant was operating at rated thermal power (RTP).

At 1500 a transient occurred that resulted in a SCRAM and loss of the Reserve Auxiliary Transformer (RAT).

At 1502 the immediate actions for the scram were completed and CPS 4200.01 Loss of AC Power was entered.

Which of the following subsequent actions will be given the highest priority?

A. Bypass the condensate polishers.

B. Restore RAT supply to the 4.16 KV Buses 1A1, 1B1, and 1C1.

C. Initiate Reactor Recirc Pump Auxiliary Seal Injection Pump operation.

D. Verify all rods inserted at the Rod Action Control System (RACS) panels.

Answer C Answer Explanation C is correct:

Per CPS 4200.01 Loss of AC Power, SBO: Technical Bases Summary, states that system leakage for a SBO, including RR pump seals (~ 38 gpm) and inventory loss from SRVs, it is assumed to be ~ 100 gpm.

The commonality between a SBO and loss of the non-ECCS buses is that both events will result in undervoltage trips of the CRD Pumps, a loss of the RR Pumps, and a loss of the CCW pumps.

Per EP-AA-1003 Addendum 3 Emergency Action Levels For Clinton Station basis for MU6 (CL 2-92), a stuck open SRV or SRV leakage is not considered either identified or unidentified leakage by Technical Specification and, therefore, is not applicable to the MU6 EAL.

Therefore, leakage past the RR Pump seals will result in unidentified leakage in the drywell of >

10 gpm for 15 minutes.

Incorrect Responses:

A is incorrect but plausible. The conditions in the stem (scram and loss of non-vital power) will result in a loss of vacuum with no ability for mitigation with no power available to the Condenser Vacuum Pumps.

CPS 4004.02 Loss of Vacuum directs bypassing the Condensate Polishers to prevent overheating the resin during degrading vacuum conditions. This action would not be prioritized however, because of the loss of non-vital power to 1CD016 Condensate Polisher Bypass Valve and to the Test ID: 343246 12/02/2020 53 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Condensate/Condensate Booster System Pumps.

B is incorrect but plausible. Emergency Action Level (EAL) MA1 is exceeded if AC power capability to the emergency busses is reduced to a single power source for 15 minutes or longer such that an additional single failure would result in a station blackout. Under the conditions provided in the stem, however, the ERAT and DGs are available to the safety busses, so MA1 is not imminently threatened from a loss of the RAT transformers and restoring the RAT supply to the safety busses would be prioritized lower than initiating RR Aux Seal Injection pump operation.

D is incorrect but plausible. Failure of RPS to shutdown the reactor can result in exceeding the threshold for EAL MS3, MA3 or MU3 depending on the severity of the ATWS. The threshold values for these EALs, however, are based on reactor power and not on the ability to determine if shutdown criteria is met.

Furthermore, since the immediate actions for the scram are complete (rods have been verified inserted) and the loss of non-vital AC will not preclude the operator from determining rod positions at P680 -

verifying rod positions at the RACS panels is not required.

Question Information The plant was operating at rated thermal power (RTP).

Topic At 1500 a transient occurred that resulted User ID CL-ILT-N19091 System ID 2148874 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

Test ID: 343246 12/02/2020 54 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key References Provided: None K/A Justification Statement: Question meets the (b) portion of the KA because the candidate has to determine the impact of the loss of BOP power and resultant loss of CRD cooling water flow to the RR Pump seals and then determine that the MU6 threshold values will be exceeded requiring actions to be taken per the Emergency Plan.

As permitted by ES-401 D.2.a, the (a) portion of the K/A (the low cog portion) is not tested.

SRO Only Justification Statement: Question is linked to SRO only task 997777.03 Emergency Plan Activities performed by an SRO 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 comprehension level. The candidate must analyze the conditions in the stem and then determine required actions based on that analysis (3-SPK/SPR).

NRC Exams Only (as applicable)

Question Type: Bank (CL-ILT-N17093) Difficulty: N/A

  • CPS 4200.01 Rev. 26c Technical Reference and Revision #:
  • EP-AA-1003 Addendum 3 (CL 2-4, 2-5, 2-6, 2-91 and 2-92)

Rev. 2 Training Objective: 201001.09 DISCUSS the effect:

a. A total loss or malfunction of the Control Rod Drive Hydraulic System has on the plant.
b. A total loss or malfunction of various plant systems has on the Control Rod Drive Hydraulic System.

Previous NRC Exam Use: ILT 17-1 NRC K/A Reference(s) 201001.A2.04 Safety Function 1 Tier 2 Group 2 RO Imp: 3.8 SRO Imp: 3.9*

Ability to (a) predict the impacts of the following on the CONTROL ROD DRIVE HYDRAULIC 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)

Scram conditions Learning Objective(s)

Q16/91 201001 A2.04 (17-1N)

User (Sys) ID N/A (1537911)

Cross Reference Links None Test ID: 343246 12/02/2020 55 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Test ID: 343246 12/02/2020 56 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 17 ID: 2148829 Points: 1.00 A transient has occurred, the reactor has scrammed and plant conditions are as follows:

  • Suppression Pool Temperature 140°F and rising at 0.1°F per minute.
  • Suppression Pool Level is at 15 feet 4 inches and lowering at 0.5 inches per minute.
  • The suppression pool level decrease CANNOT be stopped.
  • Drywell Pressure is at 1.5 psig and steady.
  • Containment Temperature is at 110°F and steady.
  • Containment pressure is at 1.4 psig and steady.
  • RPV pressure is at 600 psig and steady.
  • Suppression pool dump valves, 1SM001A, 1B, 2A and 2B, are shut.
  • Hydrogen Mixing Compressors 1HG02CA(B) are in operation.

See EOP-6 Primary Containment Control, Figures O and P below:

Test ID: 343246 12/02/2020 57 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Which of the following describes the NEXT required operator action?

A. Dump upper pools.

B. Start containment sprays.

C. Stop Hydrogen Mixing Compressors.

D. Lower RPV pressure to below Heat Capacity Limit Curve.

Answer A Answer Explanation A is correct.

Per CPS 4402.01 Primary Containment Control (EOP-6) , SUPPRESSION POOL LEVEL leg, Low Level -

Below 18 ft 11 in.:

  • IF "Cannot hold pool level above 15 ft 1 in."
  • THEN "Before suppression pool level drops to 15 ft 1 in., dump upper pools Incorrect responses:

B is incorrect but plausible. This response would be correct if Drywell pressure exceeded its EOP-6 entry condition since Containment pressure is above the Figure O Containment Spray Initiation Limit.

However, since Drywell and Containment pressures are < 1.68 psig, containment sprays are not yet directed.

C is incorrect. This response is plausible because Hydrogen Mixing Compressors must be secured below a Suppression Pool level of 13 ft 1 in.

Test ID: 343246 12/02/2020 58 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key D is incorrect. This response would be correct if Suppression Pool temperature was rising at a rate that threatened Figure P Heat Capacity Limit before suppression pool level lowered to the point that dumping the upper pools was required. However, at an RPV pressure of 600 psig, Suppression Pool temperature would reach the Figure P limit at ~ 146°F ( or ~ 60 minutes later). A Suppression Pool level of 15 feet 1 inch will be reached in about 6 minutes.

Question Information A transient has occurred, the reactor has scrammed and plant conditions are as follows:

Topic

  • All co User ID CL-ILT-N19092 System ID 2148829 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the K/A because the examinee must demonstrate the ability to perform an integrated plant procedure following failure of the primary containment system.

SRO-Only Justification Question is linked to SRO-only task 100509.07 Execute EOP Decision Symbols. Also 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 application and analysis level. The candidate must evaluate a number of parameters provided in the stem, determine the correct path in EOP-6 and select the correct response (3-SPK).

NRC Exams Only Question Type Bank (CL-ILT-N11079) Difficulty N/A Technical Reference and Revision #

  • CPS 4402.01 (EOP-6) Rev. 30 LP87558.01.07 Given a decreasing Suppression Pool Training Objective water level, determine when it is appropriate to dump the Upper Pools in terms of approaching a level of 15 ft 1 in.

Previous NRC Exam Use ILT 10-1 NRC Test ID: 343246 12/02/2020 59 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s)

GS.223001 Safety Function 5 Tier 2 Group 2 RO Imp: SRO Imp:

Primary Containment System and Auxiliaries B2.1.23 Safety Function 5 Tier 3 Group RO Imp: 4.3 SRO Imp: 4.4 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6)

Learning Objective(s)

Q17/92 223001 2.1.23 User (Sys) ID N/A (1537912)

Cross Reference Links None Test ID: 343246 12/02/2020 60 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 18 ID: 2149283 Points: 1.00 The plant is operating at rated thermal power with VC Train A in operation.

THEN, the following annunciators are received:

  • 5050-3L TROUBLE CONTROL ROOM HEATING COIL A
  • 5050-8J HIGH HUMIDITY CONTROL ROOM TRAIN A (1) Which annunciator requires a Technical Specification LCO entry?

(2) Which LCO is required to be entered?

A. (1) 5050-3L (2) 3.7.3 Control Room Ventilation System B. (1) 5050-3L (2) 3.7.4 Control Room Air Conditioning System C. (1) 5050-8J (2) 3.7.3 Control Room Ventilation System D. (1) 5050-8J (2) 3.7.4 Control Room Air Conditioning System Answer A Answer Explanation A is correct.

CPS 5050.03 Alarm Panel 5050 Annunciators - Row 3, annunciator procedure for 5050-3L TROUBLE CONTROL ROOM HEATING COIL A directs the following:

  • Shift to Control Room HVAC Train B per CPS 3402.01 (VC).
  • If VC A Train is INOP, refer to ITS LCO 3.7.3 Per B3.7.3 Control Room Ventilation System, Each Control Room Ventilation subsystem is considered OPERABLE when the individual components necessary to limit CRE occupant exposure are OPERABLE.

A subsystem is considered OPERABLE when its associated:

  • HEPA filter and charcoal adsorber are not excessively restricting flow and are capable of performing their filtration functions; and
  • Heater, demister, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

Since the Air Handling Unit Heating Coil (0VC01AA) is not OPERABLE, then LCO 3.7.3 Control Room Ventilation System must be entered.

Test ID: 343246 12/02/2020 61 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Incorrect Responses:

B is incorrect but plausible. The first part of the answer is correct. The second part of the answer is plausible because there are also heating coils associated with the Control Room Air Conditioning (AC)

System subsystems. However, per B3.7.4 Control Room AC System, The heating coils and humidification equipment are not required for Control Room AC System OPERABILITY.

C is incorrect but plausible. This answer would be correct if the humidification equipment was associated with the Control Room Ventilation System and the 0VC01AA Heating Coil was associated with the Control Room AC System. In this case, annunciator 5050-8J would be responsible for an entry into LCO 3.7.3 Control Room Ventilation. However, since the 0VC01AA Heating Coil is associated with the Control Room Ventilation System, annunciator 5050-3L is responsible for entry into LCO 3.7.3 Control Room Ventilation. The second part of the answer is correct.

D is incorrect but plausible. This answer would be correct if the humidification equipment was associated with the Control Room Ventilation System and the 0VC01AA Heating Coil was associated with the Control Room AC System. In this case, annunciator 5050-8J would be responsible for an entry into LCO 3.7.3 Control Room Ventilation. The second part of the answer is plausible because there are also heating coils associated with the Control Room Air Conditioning (AC) System subsystems. However, per B3.7.4 Control Room AC System, The heating coils and humidification equipment are not required for Control Room AC System OPERABILITY.

Question Information The plant is operating at rated thermal power with VC Train A in operation.

Topic THEN, the followin User ID CL-ILT-N19093 System ID 2149283 Status Active Point Value 1.00 Time (min) 2 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I CFR: 43.2 Facility operating limitations in the technical specifications and 10CFR55 Content their bases.

References Provided None K/A Justification Question meets the K/A because the candidate must interpret Control Room Ventilation annunciators and based on their significance, choose the appropriate LCO to enter in order to determine the correct response.

SRO-Only Justification 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 low cog written at the memory level. The examinee must recall facts pertaining to and contained in Test ID: 343246 12/02/2020 62 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key a procedures and ITS (1-F).

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

  • CPS 5050.03 (5050-3L) Rev. 30c
  • CPS 5050.08 (5050-8J) Rev. 31d
  • ITS B3.7.3 (B3.7-13) Rev. 14-2
  • ITS B3.7.4 (B3.7-23) Rev. 20-2 290003.06 Given a CONTROL ROOM HVAC Training Objective System Annunciator, DESCRIBE:
a. The condition causing the annunciator
b. Any automatic actions
c. Any operational implications

.1 AUTO START CONT ROOM HVAC M-U AIR FAN A (B)

.2 SMOKE VC SYSTEM

.3 LOW TEMP CONT ROOM HVAC MIXED AIR DIV 1 (2)

.4 HI RADIATION CONT RM HVAC SYS DIVISION 1 (2)

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

B2.4.45 Safety Function 5 Tier 3 Group RO Imp: 4.1 SRO Imp: 4.3 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

GS.290003 Safety Function 9 Tier 2 Group 2 RO Imp: SRO Imp:

Control Room HVAC Learning Objective(s)

Q18/93 290003 2.4.45 User (Sys) ID N/A (1537913)

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 2147108 Test ID: 343246 12/02/2020 63 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Test ID: 343246 12/02/2020 64 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 19 ID: 2148946 Points: 1.00 A transient has occurred resulting in exceeding entry condition E1 on the generic EOP flowchart shown below.

Conditions have been met for A1, but A1 actions have NOT yet been taken.

If the conditions in C1 are exceeded, the CRS will direct the Reactor Operators to perform action

_____(1)_____ NEXT.

Given these conditions, if entry condition E3 is exceeded, re-entry into the EOP _____(2)_____ required.

Test ID: 343246 12/02/2020 65 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key A. (1) A1 (2) is B. (1) A2 (2) is C. (1) A1 (2) is NOT D. (1) A2 (2) is NOT Answer A Answer Explanation A is correct:

Per OP-CL-101-111-1001 Strategies For Successful Transient Mitigation, 4.1.3 EOP Execution, step 4:

All steps must be executed in their specified order when executing control legs of the EOPs. This is to ensure all available mitigating systems are utilized and their effectiveness assessed, even if a blowdown parameter is currently exceeded. As an example, the Containment Pressure leg must be executed, including establishing Containment Sprays, if available, prior to determining if a blowdown is required due to exceeding the Pressure Suppression Pressure (PSP) limit. After Containment Sprays are initiated (valves fully open), then check PSP. If Containment Pressure is below the PSP limit, then blow down is not required.

Per CPS 1005.09 Emergency Operating Procedure (EOP) and Severe Accident Guideline (SAG)

Program, section 8.12.3.3, an EOP shall be reentered upon each receipt of an entry condition.

Incorrect Responses:

B is incorrect but plausible. Although a condition parameter is exceeded, making the distractor plausible, preceding EOP actions must be performed in their specified order. The second part of the distractor is correct.

C is incorrect but plausible. The first part of the distractor is correct. The second part is plausible because most off-normal and and emergency procedures are designed to be followed sequentially without re-entry; however, re-entry to the EOPs are required if any additional entry conditions are received.

D is incorrect but plausible. Although a condition parameter is exceeded, making the distractor plausible, preceding EOP actions must be performed in their specified order. The second part of the distractor is also incorrect. It is plausible because most off-normal and and emergency procedures are designed to be followed sequentially without re-entry; however, re-entry to the EOPs are required if any additional entry conditions are received.

Test ID: 343246 12/02/2020 66 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Information A transient has occurred resulting in exceeding entry condition E1 on the generic EOP flowchart Topic sho User ID CL-ILT-N19094 System ID 2148946 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question meets the K/A because the candidate has to demonstrate the ability to manage the control room crew by selecting the actions they are required to perform during plant transients to answer the question.

SRO-Only Justification Question is linked to SRO-only task 100509.07 Execute EOP Decision Symbols.

Additional Information Question is low cog written at the memory level. The candidate must recall the procedural actions required to execute the EOPs (1-P).

NRC Exams Only Question Type Bank (CL-ILT-N14098) Difficulty N/A Technical Reference and Revision #

  • CPS 1005.09 Rev. 11
  • OP-CL-101-111-1001 Rev. 15d Training Objective LP85801 2.1.6 Ability to manage the control room crew during plant transients.

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

B2.1.06 Safety Function 9 Tier 3 Group RO Imp: 3.8* SRO Imp: 4.8 Ability to manage the control room crew during plant transients.

(CFR: 41.10 / 43.5 / 45.12 / 45.13)

Learning Objective(s)

Q19/94 2.1.6 User (Sys) ID N/A (1537914)

Cross Reference Links None Test ID: 343246 12/02/2020 67 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 20 ID: 2161620 Points: 1.00 Which of the following personnel can manipulate reactivity controls at Clinton Power Station (CPS) when supervised by a licensed reactor operator?

  • Mary, who is licensed as a Reactor Operator at Quad Cities Generating Station (QDC).
  • Sally, who is a Senior Reactor Operator trainee in the current licensed operator training class at CPS.
  • Ted, who has an inactive Reactor Operator license at CPS and is reactivating his license.

A. Sally and Ted ONLY B. Mary and Ted ONLY C. Sally and Mary ONLY D. Sally, Ted, and Mary Answer A Answer Explanation A is correct:

Per OP-AA-300 Reactivity Management section 4.8.6, the Control Room Supervisor (CRS) / Senior Reactor Operator (SRO) ENSURES trainees manipulating reactivity controls are enrolled in an approved training program and directly supervised by a licensed individual.

  • Sally is enrolled in a training program at CPS.

Per OP-AA-105-102 NRC Active License Maintenance section 4.2.1, Reactivate an RO or SRO license to an active status by performing 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift functions in the presence and under the sole direct supervision of an active RO or SRO, as appropriate and in the position to which the individual will be assigned.

  • Ted previously had an active CPS RO license.

Incorrect Responses:

B is incorrect but plausible. The first part of the response is plausible because Mary currently holds an active operating license at QDC. Per OP-AA-101-111-1001 Operations Standards And Expectations 0 List Of Actions That Travelers Can Perform, travelers are typically authorized to manipulate equipment and perform procedures. However, OP-AA-101-111-1001 Attachment 10 does not include supervised reactivity manipulations. The second part of this response is correct.

C is incorrect but plausible. The first part of the response is correct. The second part of the response is plausible because Mary currently holds an active operating license at QDC. Per OP-AA-101-111-1001 Operations Standards And Expectations Attachment 10 List Of Actions That Travelers Can Perform, Test ID: 343246 12/02/2020 68 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key travelers are typically authorized to manipulate equipment and perform procedures. However, OP-AA-101-111-1001 Attachment 10 does not include supervised reactivity manipulations.

D is incorrect but plausible. Sally and Ted may perform reactivity manipulations under direct supervision because they are in approved training programs at CPS; however, Mary's license at QDC does not allow for these reactivity manipulations. Per OP-AA-101-111-1001 Operations Standards And Expectations 0 List Of Actions That Travelers Can Perform, travelers are typically authorized to manipulate equipment and perform procedures. However, OP-AA-101-111-1001 Attachment 10 does not include supervised reactivity manipulations.

Question Information Which of the following personnel can manipulate reactivity controls at Clinton Power Station Topic (CPS)

User ID CL-ILT-N19095 System ID 2161620 Status Active Point Value 1.00 Time (min) 2 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I CFR: 43.6 Procedures and limitations involved in initial core loading, 10CFR55 Content alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

References Provided None K/A Justification Question meets the K/A because it requires the candidate to demonstrate knowledge of who is permitted to conduct reactivity manipulations, which is a CRS/SRO task designated in the reactivity management procedure.

SRO-Only Justification Question is linked to SRO-only task 999999.60, Administer the Reactivity Management Program 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 a low cog question written at the memory level. The candidate must recall facts in CPS procedures (1-F).

NRC Exams Only Question Type New Difficulty N/A

  • OP-AA-300 Rev. 14 Technical Reference and Revision #
  • OP-AA-101-111-1001 Rev. 25 LP85801.2.1.37 Training Objective Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Previous NRC Exam Use None Test ID: 343246 12/02/2020 69 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s)

B2.1.37 Safety Function 9 Tier 3 Group RO Imp: 4.3 SRO Imp: 4.6 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

(CFR: 41.1 / 43.6 / 45.6)

Learning Objective(s)

Q20/95 2.1.37 User (Sys) ID N/A (1537915)

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 2146458 Test ID: 343246 12/02/2020 70 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 21 ID: 2147086 Points: 1.00 The plant is operating at rated thermal power.

At 1200 on May 1st, a common equipment failure is identified to the Control Room Supervisor (CRS) which makes multiple pieces of Improved Technical Specification (ITS) required equipment INOPERABLE.

  • Corrective measures will take 4 to 6 days to restore compliance with ITS.
  • The affected LCOs are applicable in Modes 1, 2, and 3.
  • The CRS immediately enters Technical Specification LCO 3.0.3 Which one of the following identifies the LATEST time on May 3rd by which the plant must be in Mode 4?

A. 0100 B. 0200 C. 2100 D. 2200 Answer A Answer Explanation A is correct.

Per Improved Technical Specification (ITS) Limiting Condition of Operation (LCO) 3.0.3:

Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:

  • Mode 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;
  • Mode 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and
  • Mode 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Therefore, the plant must be in Mode 4 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> after 1200 on May 1st, which is 0100 on May 3rd.

Incorrect responses:

B is incorrect but plausible. This answer would be correct if the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowed to initiate plant action was in addition to the 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> allowed to place the plant in Mode 4. However, the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowed to prepare for an orderly shutdown before initiating a change in unit operation is part of the 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />, not in addition to the 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

C is incorrect but plausible. This answer would be correct if the time allowed to Mode 4 was a sum of the times to reach Mode 2 (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />), Mode 3 (13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />) and Mode 4 (37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />) since during an orderly Test ID: 343246 12/02/2020 71 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key shutdown you would transition from Mode 1 Mode 2 Mode 3 Mode 4. However, since the unit shall be placed in a Mode in which the LCO is not applicable (Mode 4), 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> is the allowable time to reach Mode 4.

D is incorrect but plausible. This answer would be correct if:

  • the time allowed to Mode 4 was a sum of the times to reach Mode 2 (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />), Mode 3 (13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />) and Mode 4 (37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />) since during an orderly shutdown you would transition from Mode 1 Mode 2 Mode 3 Mode 4, and
  • the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowed to initiate plant action was in addition to the time allowed to place the plant in Mode 4.

Question Information The plant is operating at rated thermal power.

Topic At 1200 on May 1st, a common equipment failure is User ID CL-ILT-N19096 System ID 2147086 Status Active Point Value 1.00 Time (min) 2 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I 10CFR55 Content CFR: 43.1 Conditions and limitations in the facility license.

References Provided None K/A Justification Question meets the K/A because the candidate must demonstrate knowledge of Limiting Conditions for Operation (LCO) applicability and their associated time limits for plant conditions presented in the stem.

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

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

The candidate must recall facts in CPS Improved Technical Specifications (ITS) (1-F).

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

  • ITS 3.0.3 Amend. 220 Training Objective LP85802.2.2.22Knowledge of limiting conditions for operations and safety limits.

Previous NRC Exam Use None Test ID: 343246 12/02/2020 72 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s)

B2.2.22 Safety Function 9 Tier 3 Group RO Imp: 4.0 SRO Imp: 4.7 Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2)

Learning Objective(s)

Q21/96 2.2.22 User (Sys) ID N/A (1537916)

Cross Reference Links None Test ID: 343246 12/02/2020 73 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 22 ID: 2149126 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.

Test ID: 343246 12/02/2020 74 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Information Per CC-AA-103-100 Configuration Change Control for Permanent Physical Plant Changes, Topic Operations is User ID CL-ILT-N19097 System ID 2149126 Status Active Point Value 1.00 Time (min) 2 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I CFR: 43.3 Facility licensee procedures required to obtain authority for 10CFR55 Content design and operating changes in the facility.

References Provided: None K/A Justification Statement: 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 Statement: 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 (as applicable)

Question Type: New Difficulty: N/A Technical Reference and Revision #:

Knowledge of the process for making design or operating changes to the facility.

Previous NRC Exam Use: None Test ID: 343246 12/02/2020 75 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s)

B2.2.05 Safety Function 9 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 User (Sys) ID N/A (1537917)

Cross Reference Links None Test ID: 343246 12/02/2020 76 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 23 ID: 2148794 Points: 1.00 An event has occurred and the Technical Support Center (TSC) has NOT yet been activated or briefed.

Containment Venting that will exceed radioactive release rate limits of CPS 4412.00C002 Sampling Containment Atmosphere Prior To Venting is required.

The CRS shall...

A. proceed with Containment venting with NO additional authorization.

B. proceed with Containment venting ONLY after authorization from the Shift Manager.

C. proceed with Containment venting ONLY after authorization from the Plant Manager.

D. delay Containment venting until the TSC has Command and Control and the venting is authorized by the Station Emergency Director.

Answer B Answer Explanation B is correct.

Per EOP-6 Primary Containment Control, venting the containment to stay below the Primary Containment Pressure Limit which requires exceeding the radioactive release rate limits of CPS 4412.00C002 Sampling Containment Atmosphere Prior To Venting may only be performed when authorized by the Emergency Director.

Per EP-AA-112 Emergency Response Organization (ERO) / Emergency Response Facility (ERF)

Activation And Operation:

  • the Shift Manager shall assume the responsibilities of the Shift Emergency Director following event classification. The Shift Manager retains this responsibility until command and control is transferred to the Station Emergency Director (SED).
  • The SED may assume command and control when the TSC is activated or if in the SED's judgement the following criteria are met:
  • adequate staff levels are present in support of non-delegable responsibilities;
  • staff has been fully briefed; and
  • a turnover has been completed.

Incorrect Responses:

A is incorrect but plausible. This response would be correct if containment venting was not expected to exceed the radioactive release rate limits of CPS 4412.00C002. However, since the question stem specifically states that containment venting will exceed the radioactive release rates of CPS 4412.00C002, authorization of the Emergency Director is required.

Test ID: 343246 12/02/2020 77 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key C is incorrect but plausible. This response would be correct if venting the containment was controlled like other major plant evolutions such as conducting a reactor plant startup, which requires Plant Manager authorization per CPS 3001.01 Preparation For Startup & Approach To Critical. However, EOP-6 specifically requires Emergency Director authorization to perform the evolution presented in the question stem.

D is incorrect but plausible. This response would be correct if the Shift Manager was not designated by EP-AA-112 to assume responsibilities of the Shift Emergency Director prior to the Station Emergency Director being stationed. However, since the Shift Manager initially assumes the responsibilities of the Shift Emergency Director, the Shift Manager can authorize the radioactive release per EOP-6. There is no need to delay.

Question Information An event has occurred and the Technical Support Center (TSC) has NOT yet been activated or Topic briefed.

User ID CL-ILT-N19098 System ID 2148794 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided None K/A Justification Question requires assessing plant conditions and then selecting an appropriate strategy to control the release of radiation.

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 Question is low cog written at the memory level. The candidate must recall procedure steps (1-B).

NRC Exams Only Question Type Bank (CL-ILT-N11096) Difficulty N/A Technical Reference and Revision #

  • CPS 4402.01 Rev 30
  • EP-AA-112 Rev. 22 Training Objective LP85803. 2.3.11Ability to control radiation releases.

Previous NRC Exam Use ILT 10-1 NRC Test ID: 343246 12/02/2020 78 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key K/A Reference(s)

B2.3.11 Safety Function 9 Tier 3 Group RO Imp: 3.8 SRO Imp: 4.3 Ability to control radiation releases.

(CFR: 41.11 / 43.4 / 45.10)

Learning Objective(s)

Q23/98 2.3.11 User (Sys) ID N/A (1537918)

Cross Reference Links None Test ID: 343246 12/02/2020 79 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 24 ID: 2148793 Points: 1.00 The reactor is operating at 98% power.

At 0830, the 'A' RO reports:

  • Drywell pressure is 2.2 psig and rising rapidly
  • Mode switch in SHUTDOWN; Power is 98%
  • Some inward control rod motion was observed At 0837, conditions are as follows:
  • Reactor power is stable at 55%
  • Reactor water level is -5 inches Wide Range, lowering slowly
  • Suppression pool temperature is 150°F, rising slowly
  • Suppression Pool level is 19.8 feet, rising slowly
  • Drywell radiation monitor reading 60 R/hr, rising slowly
  • See EOP-6, Figure P below:

What is the proper Emergency Action Level (EAL) classification at 0837?

Test ID: 343246 12/02/2020 80 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key A. FA1 B. MA3 C. FG1 D. MS3 Answer D Answer Explanation D is correct.

Per EP-AA-1003 Addendum 3 Exelon Nuclear Emergency Action Levels for Clinton Station, a Site Area Emergency threshold has been exceeded based on a MS3 classification - Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.

Emergency Action Level (EAL):

  • Auto scram did not shutdown the reactor as indicated by Reactor power > 5%; AND
  • All manual / ARI actions to shutdown the reactor have been unsuccessful indicated by Reactor power > 5%; AND
  • EOP-6 Fig P Heat Capacity limit exceeded (SRVs operating in LLS - equated to RPV pressure above 900 psig with 150°F suppression pool temperature).

Incorrect Responses:

A is incorrect but plausible. The FA1 threshold has been exceeded due to loss of the RC. This answer would be correct if only the FA1 threshold had been exceeded, but since MS3 has also been exceeded, classification at the higher EAL level is required.

B is incorrect but plausible. The MA3 threshold has been exceeded due to the ATWS condition. This answer would be correct if only the MA3 threshold had been exceeded, but since MS3 has also been exceeded, classification at the higher EAL level is required.

C is incorrect but plausible. The FG1 classification would be correct if conditions presented in the stem exceeded the thresholds causing a loss of two barriers and a potential loss of a third barrier. Since Drywell radiation levels did not exceed the threshold for loss of the Fuel Clad barrier and potential loss of the Containment barrier, classification at the MS3 level is correct.

Test ID: 343246 12/02/2020 81 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Information The reactor is operating at 98% power.

Topic At 0830, the 'A' RO reports:

Drywell pressure is 2.2 psi User ID CL-ILT-N19099 System ID 2148793 Status Active Point Value 1.00 Time (min) 0 Open or Closed Reference OPEN Operator Type_Cognitive Level SRO-HIGH Operator Discipline LO-I CFR: 43.5 Assessment of facility conditions and selection of appropriate 10CFR55 Content procedures during normal, abnormal, and emergency situations.

References Provided EP-AA-1003 Addendum 3, Rev 2 (pages CL 2.3 and CL 2.5)

K/A Justification Question meets the KA because it requires the examinee to interpret the indications provided in the stem and then determine the emergency action level threshold and classification.

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. In addition, classification of emergencies is an SRO only function at CPS.

Additional Information Question is high cog written at the analysis level. The candidate must analyze the conditions provided in the stem and use references to determine which emergency classification is appropriate based on that analysis (3-SPK/SPR).

NRC Exams Only Question Type Bank (CL-ILT-A11084) Difficulty N/A Technical Reference and Revision #

  • CPS 4402.01 Rev. 30 Training Objective LP85804.

2.4.41 Knowledge of the emergency action level thresholds and classifications.

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

B2.4.41 Safety Function 9 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)

Test ID: 343246 12/02/2020 82 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Learning Objective(s)

Q24/99 2.4.41 User (Sys) ID N/A (1537919)

Cross Reference Links None Test ID: 343246 12/02/2020 83 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question 25 ID: 2146438 Points: 1.00 Which of the following on-shift staffing positions may concurrently be assigned as a Fire Brigade Member?

A. Incident Assessor (IA)

B. Safe Shutdown Qualified Operator (SSQ)

C. Emergency Response Organization (ERO) Communicator (EROC)

D. Emergency Response Organization (ERO) Non-Licensed Operator (ENLO)

Answer D Answer Explanation D is correct:

Per OP-CL-101-102-1001, CPS Minimum On-Shift Staffing Functions (Table 1), in Modes 1/2/3, CPS is required to have:

  • One (1) Incident Assessor (IA) that may not have other concurrent ERO or Fire Brigade duties.
  • One (1) Safe Shutdown Qualified Operator (SSQ) that cannot be the 'A' or 'B' RO, Fire Brigade Member or EROC. The SSQ must be C area qualified.
  • One (1) Fire Brigade Leader
  • Four (4) Fire Brigade Members, which may also be ENLO or ERO Access.
  • One (1) designated EROC, which cannot be the SM, STA, IA, A RO, B RO, ENLO, A Fire Brigade position, or SSQ.

Incorrect Responses:

A is incorrect but plausible. The Incident Assessor (IA) position is required to meet the minimum on-shift staffing functions, but may not fill any other concurrent Fire Brigade duties.

B is incorrect but plausible. The Safe Shutdown Qualified Operator (SSQ) position is required to meet the minimum on-shift staffing functions but may not be a member of the Fire Brigade.

C is incorrect but plausible. The Emergency Response Organization (ERO) Communicator (EROC) is required to meet the minimum on-shift staffing functions, but may not be a member of the Fire Brigade.

Test ID: 343246 12/02/2020 84 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key Question Information Which of the following on-shift staffing positions may concurrently be assigned as a Fire Brigade Topic M

User ID CL-ILT-N19100 System ID 2146438 Status Active Point Value 1.00 Time (min) 3 Open or Closed Reference CLOSED Operator Type_Cognitive Level SRO-MEMORY Operator Discipline LO-I 10CFR55 Content 10 CFR 55.43 SRO WRITTEN EXAMINATION References Provided None K/A Justification Question meets the KA because the examinee must be knowledgeable of the facility protection requirements, including fire brigade manning.

SRO-Only Justification Question pertains to verification of minimum shift staffing which is an SRO-only function at CPS and is also linked to 10CFR55.43(b)(1) Conditions and limitations in the facility license.

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

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

  • OP-CL-101-102-1001 Rev 7d Training Objective LP85804 2.4.26 Knowledge of facility protection requirements including fire brigade and portable fire fighting equipment usage.

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

B2.4.26 Safety Function 9 Tier 3 Group RO Imp: 3.1 SRO Imp: 3.6 Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage.

(CFR: 41.10 / 43.5 / 45.12)

Learning Objective(s)

Q25/100 2.4.26 User (Sys) ID N/A (1537920)

Test ID: 343246 12/02/2020 85 of 86

CONFIDENTIAL - Exam Material ILT 19-1 NRC SRO Test Key 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 2103888 Test ID: 343246 12/02/2020 86 of 86

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCO 3.0.2 LCO 3.0.3 LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; (continued)

CLINTON 3.0-1 Amendment No. 220

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 b. After performance of a risk assessment addressing (continued) inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate (exceptions to this Specification are stated in the individual Specifications); or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

LCO 3.0.5 LCO 3.0.6 (continued)

CLINTON 3.0-2 Amendment No. 95

ECCS Instrumentation 3.3.5.1 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.5.1-1.

ACTIONS

NOTE--------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.1-1 for the channel.

(continued)

CLINTON 3.3-31 Amendment No. 95

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. As required by B.1 --------NOTE---------

Required Action A.1 Only applicable for and referenced in Functions 1.a, 1.b, Table 3.3.5.1-1. 2.a, and 2.b.

Declare supported 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability is capability for inoperable. feature(s) in both divisions AND B.2 --------NOTE---------

Only applicable for Functions 3.a and 3.b.

Declare High Pressure 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from Core Spray (HPCS) discovery of System inoperable. loss of HPCS initiation capability AND B.3 Place channel in trip. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (continued)

CLINTON 3.3-32 Amendment No. 216

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 ---------NOTE--------

Required Action A.1 Only applicable for and referenced in Functions 1.c, 1.d, Table 3.3.5.1-1. 2.c, and 2.d.

Declare supported 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND C.2 Restore channel to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

(continued)

CLINTON 3.3-33 Amendment No. 216

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 --------NOTE--------

Required Action A.1 Only applicable if and referenced in HPCS pump suction is Table 3.3.5.1-1. not aligned to the suppression pool.

Declare HPCS System 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from inoperable. discovery of loss of HPCS initiation capability AND D.2.1 Place channel in 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> trip.

OR D.2.2 Align the HPCS pump 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> suction to the suppression pool.

(continued)

CLINTON 3.3-34 Amendment No. 193

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 --------NOTE---------

Required Action A.1 Only applicable for and referenced in Functions 1.e, 1.f, Table 3.3.5.1-1. and 2.e.

Declare supported 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND E.2 Restore channel to 7 days OPERABLE status.

(continued)

CLINTON 3.3-35 Amendment No. 216

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. As required by F.1 Declare Automatic 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from Required Action A.1 Depressurization discovery of and referenced in System (ADS) valves loss of ADS Table 3.3.5.1-1. inoperable. initiation capability in both trip systems AND F.2 Place channel in 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from trip. discovery of inoperable channel concurrent with HPCS or reactor core isolation cooling (RCIC) inoperable AND 8 days (continued)

CLINTON 3.3-36 Amendment No. 95

ECCS Instrumentation 3.3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. As required by G.1 --------NOTES--------

Required Action A.1 Only applicable for and referenced in Functions 4.c, 4.e, Table 3.3.5.1-1. 4.f, 4.g, 5.c, 5.e, and 5.f.

Declare ADS valves 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from inoperable. discovery of loss of ADS initiation capability in both trip systems AND G.2 Restore channel to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from OPERABLE status. discovery of inoperable channel concurrent with HPCS or RCIC inoperable AND 8 days H. Required Action and H.1 Declare associated Immediately associated Completion supported feature(s)

Time of Condition B, inoperable.

C, D, E, F, or G not met.

CLINTON 3.3-37 Amendment No. 95

ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS

NOTES-------------------------------------

1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for Functions 3.c, 3.f, 3.g, and 3.h; and (b) for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for Functions other than 3.c, 3.f, 3.g, and 3.h, provided the associated Function or the redundant Function maintains ECCS initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.3 Calibrate the analog trip module. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.6 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program CLINTON 3.3-38 Amendment No. 192

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 1 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS PER ACTION A.1 REQUIREMENTS VALUE FUNCTION

1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS)

Subsystems

a. Reactor Vessel Water 1,2,3 2(a) B SR 3.3.5.1.1 -148.1 Level-Low Low Low, SR 3.3.5.1.2 inches Level 1 SR 3.3.5.1.3(c)

SR 3.3.5.1.5 SR 3.3.5.1.6(c)

b. Drywell Pressure-High 1,2,3 2(a) B SR 3.3.5.1.1 1.88 psig SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
c. LPCI Pump A 1,2,3 1 C SR 3.3.5.1.2 4.5 seconds Start-Time Delay SR 3.3.5.1.4 and Logic Card SR 3.3.5.1.5 5.5 seconds
d. Reactor Vessel 1,2,3 4 C SR 3.3.5.1.1 454 psig Pressure-Low SR 3.3.5.1.2 and (Injection Permissive) SR 3.3.5.1.3(c) 494 psig SR 3.3.5.1.4(c)

SR 3.3.5.1.5

e. LPCS Pump Discharge 1,2,3 1 E SR 3.3.5.1.1 750 gpm Flow-Low (Bypass) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
f. LPCI Pump A Discharge 1,2,3 1 E SR 3.3.5.1.1 900 gpm Flow-Low (Bypass) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
g. Manual Initiation 1,2,3 1 C SR 3.3.5.1.5 NA (continued)

(a) Also required to initiate the associated diesel generator.

(c) 1. If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined As-Found Tolerance band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value within the As-Left Tolerance of the Actual Trip Setpoint; otherwise, the channel shall be declared inoperable.
3. The Nominal Trip Setpoint and the methodology used to determine the Nominal Trip Setpoint, the predefined As-Found Tolerance and As-Left Tolerance bands shall be specified in the ORM.

CLINTON 3.3-39 Amendment No. 216

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI B and LPCI C Subsystems
a. Reactor Vessel 1,2,3 2(a) B SR 3.3.5.1.1 -148.1 Water Level-Low Low SR 3.3.5.1.2 inches Low, Level 1 SR 3.3.5.1.3(c)

SR 3.3.5.1.5 SR 3.3.5.1.6(c)

b. Drywell Pressure- 1,2,3 2(a) B SR 3.3.5.1.1 1.88 psig High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
c. LPCI Pump B 1,2,3 1 C SR 3.3.5.1.2 4.5 seconds Start-Time Delay SR 3.3.5.1.4 and Logic Card SR 3.3.5.1.5 5.5 seconds
d. Reactor Vessel 1,2,3 4 C SR 3.3.5.1.1 454 psig Pressure-Low SR 3.3.5.1.2 and (Injection SR 3.3.5.1.3(c) 494 psig Permissive) SR 3.3.5.1.4(c)

SR 3.3.5.1.5

e. LPCI Pump B and 1,2,3 1 per pump E SR 3.3.5.1.1 900 gpm LPCI Pump C SR 3.3.5.1.2 Discharge Flow-Low SR 3.3.5.1.3 (Bypass) SR 3.3.5.1.4 SR 3.3.5.1.5
f. Manual Initiation 1,2,3 1 C SR 3.3.5.1.5 NA (continued)

(a) Also required to initiate the associated diesel generator.

(c) 1. If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined As-Found Tolerance band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value within the As-Left Tolerance of the Actual Trip Setpoint; otherwise, the channel shall be declared inoperable.
3. The Nominal Trip Setpoint and the methodology used to determine the Nominal Trip Setpoint, the predefined As-Found Tolerance and As-Left Tolerance bands shall be specified in the ORM.

CLINTON 3.3-40 Amendment No. 216

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

3. High Pressure Core Spray (HPCS) System
a. Reactor Vessel Water 1,2,3 4(a) B SR 3.3.5.1.1 -48.1 inches Level-Low Low, SR 3.3.5.1.2 Level 2 SR 3.3.5.1.3(c)

SR 3.3.5.1.5 SR 3.3.5.1.6(c)

b. Drywell 1,2,3 4(a) B SR 3.3.5.1.1 1.88 psig Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
c. Reactor Vessel Water 1,2,3 2 C SR 3.3.5.1.1 54.6 inches Level-High, SR 3.3.5.1.2 Level 8 SR 3.3.5.1.3(c)

SR 3.3.5.1.5 SR 3.3.5.1.6(c)

d. RCIC Storage Tank 1,2,3 2 D SR 3.3.5.1.1 3.0 inches Level-Low SR 3.3.5.1.2 SR 3.3.5.1.3(c)

SR 3.3.5.1.4(c)

SR 3.3.5.1.5

e. Suppression Pool 1,2,3 2 D SR 3.3.5.1.1 11 inches Water Level-High SR 3.3.5.1.2 SR 3.3.5.1.3(c)

SR 3.3.5.1.4(c)

SR 3.3.5.1.5

f. HPCS Pump 1,2,3 1 E SR 3.3.5.1.1 120 psig Discharge SR 3.3.5.1.2 Pressure-High SR 3.3.5.1.3 (Bypass) SR 3.3.5.1.4 SR 3.3.5.1.5
g. HPCS System Flow 1,2,3 1 E SR 3.3.5.1.1 500 gpm Rate-Low (Bypass) SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
h. Manual Initiation 1,2,3 1 C SR 3.3.5.1.5 NA (continued)

(a) Also required to initiate the associated diesel generator.

(c) 1. If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined As-Found Tolerance band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value within the As-Left Tolerance of the Actual Trip Setpoint; otherwise, the channel shall be declared inoperable.
3. The Nominal Trip Setpoint and the methodology used to determine the Nominal Trip Setpoint, the predefined As-Found Tolerance and As-Left Tolerance bands shall be specified in the ORM.

CLINTON 3.3-41 Amendment No. 216

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 4 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

4. Automatic Depressurization System (ADS) Trip System 1 (Logic A and E)
a. Reactor Vessel Water 1,2(b),3(b) 2 F SR 3.3.5.1.1 -148.1 inches Level-Low Low Low, SR 3.3.5.1.2 Level 1 SR 3.3.5.1.3(c)

SR 3.3.5.1.5 SR 3.3.5.1.6(c)

b. Drywell 1,2(b),3(b) 2 F SR 3.3.5.1.1 1.88 psig Pressure-High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
c. ADS Initiation Timer 1,2(b),3(b) 1 G SR 3.3.5.1.2 117 seconds SR 3.3.5.1.4 SR 3.3.5.1.5
d. Reactor Vessel Water 1,2(b),3(b) 1 F SR 3.3.5.1.1 8.3 inches Level-Low, Level 3 SR 3.3.5.1.2 (Confirmatory) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
e. LPCS Pump Discharge 1,2(b),3(b) 2 G SR 3.3.5.1.1 125 psig and Pressure-High SR 3.3.5.1.2 < 176.3 psig SR 3.3.5.1.3(c)

SR 3.3.5.1.4(c)

SR 3.3.5.1.5

f. LPCI Pump A 1,2(b),3(b) 2 G SR 3.3.5.1.1 115 psig and Discharge Pressure- SR 3.3.5.1.2 < 135 psig High SR 3.3.5.1.3(c)

SR 3.3.5.1.4(c)

SR 3.3.5.1.5

g. ADS Drywell Pressure 1,2(b),3(b) 2 G SR 3.3.5.1.2 6.5 minutes Bypass Timer SR 3.3.5.1.4 SR 3.3.5.1.5
h. Manual Initiation 1,2(b),3(b) 2 G SR 3.3.5.1.5 NA (continued)

(b) With reactor steam dome pressure > 150 psig.

(c) 1. If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined As-Found Tolerance band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value within the As-Left Tolerance of the Actual Trip Setpoint; otherwise, the channel shall be declared inoperable.
3. The Nominal Trip Setpoint and the methodology used to determine the Nominal Trip Setpoint, the predefined As-Found Tolerance and As-Left Tolerance bands shall be specified in the ORM.

CLINTON 3.3-42 Amendment No. 216

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 5 of 5)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES OR REFERENCED OTHER REQUIRED FROM SPECIFIED CHANNELS PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

5. ADS Trip System 2 (Logic B and F)
a. Reactor Vessel Water 1,2(b),3(b) 2 F SR 3.3.5.1.1 -148.1 Level-Low Low Low, SR 3.3.5.1.2 inches Level 1 SR 3.3.5.1.3(c)

SR 3.3.5.1.5 SR 3.3.5.1.6(c)

b. Drywell Pressure-High 1,2(b),3(b) 2 F SR 3.3.5.1.1 1.88 psig SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
c. ADS Initiation Timer 1,2(b),3(b) 1 G SR 3.3.5.1.2 117 seconds SR 3.3.5.1.4 SR 3.3.5.1.5
d. Reactor Vessel Water 1,2(b),3(b) 1 F SR 3.3.5.1.1 8.3 inches Level-Low, Level 3 SR 3.3.5.1.2 (Confirmatory) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
e. LPCI Pumps B & C 1,2(b),3(b) 2 per pump G SR 3.3.5.1.1 115 psig Discharge SR 3.3.5.1.2 and 135 Pressure-High SR 3.3.5.1.3(c) psig SR 3.3.5.1.4(c)

SR 3.3.5.1.5

f. ADS Drywell Pressure 1,2(b),3(b) 2 G SR 3.3.5.1.2 6.5 minutes Bypass Timer SR 3.3.5.1.4 SR 3.3.5.1.5
g. Manual Initiation 1,2(b),3(b) 2 G SR 3.3.5.1.5 NA (b) With reactor steam dome pressure > 150 psig.

(c) 1. If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined As-Found Tolerance band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value within the As-Left Tolerance of the Actual Trip Setpoint; otherwise, the channel shall be declared inoperable.
3. The Nominal Trip Setpoint and the methodology used to determine the Nominal Trip Setpoint, the predefined As-Found Tolerance and As-Left Tolerance bands shall be specified in the ORM.

CLINTON 3.3-43 Amendment No. 216

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