ML22214A599

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1_CNS-2022-06 Draft Outlines
ML22214A599
Person / Time
Site: Cooper 
Issue date: 06/21/2022
From: Heather Gepford
NRC/RGN-IV/DORS/OB
To:
Nebraska Public Power District (NPPD)
References
Download: ML22214A599 (60)
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Text

Form 4.1-BWR RO Boiling-Water Reactor Examination Outline Notes: CO = Conduct of Operations; EC = Equipment Control; RC = Radiation Control; EM = Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan Facility: Cooper Nuclear Station Date of Exam: June 2022 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Total A2 G

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 4

3 N/A 4

4 N/A 2

20 7

2 1

2 1

1 1

0 6

3 Tier Totals 4

6 4

5 5

2 26 10

2.

Plant Systems 1

3 2

2 5

3 2 2

2 2 1 2 26 5

2 0

1 3

1 0 1 2

1 0 1 1 11 3

Tier Totals 3

3 5

6 3 3 4

3 2 2 3 37 8

3.

Generic Knowledge and Abilities Categories CO EC RC EM 6

CO EC RC EM 7

2 2

1 1

4. Theory Reactor Theory Thermodynamics 6

3 3

Form 4.1-BWR BWR Examination Outline Page 2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO)

E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)

IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation X

AK1.04 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation:

Thermal-hydraulic instabilities (CFR: 41.8 to 41.10) 4.3 295003 (APE 3) Partial or Complete Loss of AC Power X

AK1.02 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Partial or Complete Loss of AC Power:

Load shedding (CFR: 41.8 to 41.10) 3.8 295004 (APE 4) Partial or Total Loss of DC Power X

AA1.01 Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of DC Power:

DC electrical distribution (CFR: 41.7 / 45.6) 3.8 295005 (APE 5) Main Turbine Generator Trip X

AK3.05 Knowledge of the reasons for the following responses or actions as they apply to Main Turbine Generator Trip:

Extraction steam/moisture separator isolations (CFR: 41.5 / 45.6) 2.8 295006 (APE 6) Scram X

AK2.03 Knowledge of the relationship between SCRAM and the following systems or components:

CRD hydraulic system (CFR: 41.7 / 45.8) 3.9 295016 (APE 16) Control Room Abandonment X 2.4.12 Knowledge of operating crew responsibilities during emergency and abnormal operations (CFR: 41.10 / 45.12) 4.0 295018 (APE 18) Partial or Complete Loss of CCW X

AK2.02 Knowledge of the relationship between Partial or Complete Loss of Component Cooling Water and the following systems or components:

Plant operations (CFR: 41.7 / 45.8) 3.9

295019 (APE 19) Partial or Complete Loss of Instrument Air x

AA2.01 Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Instrument Air:

Instrument air pressure (CFR: 41.10 / 43.5 / 45.13) 4.1 295021 (APE 21) Loss of Shutdown Cooling X

AA2.04 Ability to determine and/or interpret the following as they apply to Loss of Shutdown Cooling: Reactor water temperature (CFR: 41.10 / 43.5 / 45.13) 4.6 295023 (APE 23) Refueling Accidents X

AA1.01 Ability to operate and/or monitor the following as they apply to Refueling Accidents:

Secondary containment ventilation (CFR: 41.7 / 45.6) 3.8 295024 (EPE 1) High Drywell Pressure X

EK1.01 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to High Drywell Pressure: Drywell integrity (CFR: 41.8 to 41.10) 4.3 295025 (EPE 2) High Reactor Pressure X

EK2.01 Knowledge of the relationship between High Reactor Pressure and the following systems or components:

RPS (CFR: 41.7 / 45.8) 4.2 295026 (EPE 3) Suppression Pool High Water Temperature X

EA1.08 Ability to operate and/or monitor the following as they apply to Suppression Pool High Water Temperature: LPCS (CFR: 41.7 / 45.6) 3.8 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only)

X EK3.03 Knowledge of the reasons for the following responses or actions as they apply to High Drywell Temperature: Drywell spray (CFR: 41.5 / 45.6) 3.8 295030 (EPE 7) Low Suppression Pool Water Level X

EA1.04 Ability to operate and/or monitor the following as they 3.5

RED = Topic sampled on SRO Exam apply to Low Suppression Pool Water Level:

Suppression pool makeup system(s)

(CFR: 41.7 / 45.6) 295031 (EPE 8) Reactor Low Water Level X

EA2.04 Ability to determine and/or interpret the following as they apply to Reactor Low Water Level: Adequate core cooling (CFR: 41.10 / 43.5 / 45.13) 4.9 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown X

EK2.02 Knowledge of the relationship between SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown and the following systems or components:

Redundant reactivity control system (CFR: 41.7 41.8 / 45.8) 4.0 295038 (EPE 15) High Offsite Radioactivity Release Rate X

EA2.04 Ability to determine and/or interpret the following as they apply to High Offsite Radioactivity Release Rate:

Source of offsite release (CFR: 41.10 / 43.5 / 45.13) 3.8 600000 (APE 24) Plant Fire On Site X 2.4.26 Knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage (CFR: 41.10 / 43.5 / 45.12) 3.1 700000 (APE 25) Generator Voltage and Electric Grid Disturbances X

AK3.01 Knowledge of the reasons for the following responses or actions as they apply to Generator Voltage and Electric Grid Disturbances:

Reactor and turbine trip criteria (CFR: 41.4 / 41.5 / 41.7 / 41.10 /

45.8) 3.8 K/A Category Totals:

3 4 3 4 4 2 Group Point Total:

20

Form 4.1-BWR BWR Examination Outline Page 3 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)

E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum X

AA1.12 Ability to operate and/or monitor the following as they apply to Loss of Main Condenser Vacuum:

Condenser air removal system (CFR: 41.7 / 45.6) 3.5 295007 (APE 7) High Reactor Pressure 295008 (APE 8) High Reactor Water Level 295009 (APE 9) Low Reactor Water Level 295010 (APE 10) High Drywell Pressure 295011 (APE 11) High Containment Temperature (Mark III Containment only) 295012 (APE 12) High Drywell Temperature 295013 (APE 13) High Suppression Pool Water Temperature/ 5 295014 (APE 14) Inadvertent Reactivity Addition 295015 (APE 15**) Incomplete Scram 295017 (APE 17) High Offsite Release Rate 295020 (APE 20) Inadvertent Containment Isolation 295022 (APE 22) Loss of Control Rod Drive Pumps X

AA2.01 Ability to determine and/or interpret the following as they apply to Loss of Control Rod Drive Pumps:

Accumulator pressure (CFR: 41.10 / 43.5 / 45.13) 4.0 295029 (EPE 6) High Suppression Pool Water Level 295032 (EPE 9) High Secondary Containment Area Temperature X

EK3.02 Knowledge of the reasons for the following responses or actions as they apply to High Secondary Containment Area Temperature:

Reactor SCRAM (CFR: 41.5 / 45.6) 4.0 295033 (EPE 10) High Secondary

Containment Area Radiation Levels 295034 (EPE 11) Secondary Containment Ventilation High Radiation

/ 9 X

EK2.01 Knowledge of the relationship between Secondary Containment Ventilation High Radiation and the following systems or components:

Process radiation monitoring system (CFR: 41.7 / 45.8) 3.7 295035 (EPE 12) Secondary Containment High Differential Pressure 295036 (EPE 13) Secondary Containment High Sump/Area Water Level X

EK1.04 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Secondary Containment High Sump/Area Water Level:

Maximum safe operating limit (CFR: 41.8 to 41.10) 4.1 500000 (EPE 16) High Containment Hydrogen Concentration X

EK2.05 Knowledge of the relationship between High Containment Hydrogen Concentration and the following systems or components:

Hydrogen and oxygen recombiners (CFR: 41.7 / 45.8) 3.5 K/A Category Point Totals:

1 2 1 1 1 0 Group Point Total:

6

Form 4.1-BWR BWR Examination Outline Page 4 Plant SystemsTier 2/Group 1 (RO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode X

K2.01 Knowledge of electrical power supplies to the following:

Pumps (CFR: 41.7) 4.1 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode X

K4.01 Knowledge of RHR/LPCI:

Injection Mode design features and/or interlocks that provide for the following:

Automatic system initiation/injection (CFR: 41.7) 4.4 205000 (SF4 SCS) Shutdown Cooling X

A2.08 Ability to (a) predict the impacts of the following on the Shutdown Cooling System (RHR Shutdown Cooling Mode) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:

Loss of heat exchanger cooling (CFR: 41.5 / 45.6) 4.0 206000 (SF2, SF4 HPCI)

High-Pressure Coolant Injection X

A3.09 Ability to monitor automatic operation of the High-Pressure Coolant Injection System, including:

System isolation (CFR: 41.7 / 45.7) 4.3 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

K6.08 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Low-Pressure Core Spray System: Keep fill system (CFR: 41.7 / 45.7) 3.4 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Control X

K4.04 Knowledge of Standby Liquid Control System design features and/or interlocks that provide for the following:

Indication of a fault in squib valves firing circuits (CFR: 41.7) 3.7

212000 (SF7 RPS) Reactor Protection X

K3.06 Knowledge of the effect that a loss or malfunction of the Reactor Protection System will have on the following systems or system parameters:

SCRAM air header solenoid-operated valves (CFR: 41.7 / 45.4) 4.1 215003 (SF7 IRM)

Intermediate-Range Monitor X

A1.05 Ability to predict and/or monitor changes in parameters associated with operation of the Intermediate Range Monitor System, including:

SCRAM and rod block trip setpoints (CFR: 41.5 / 45.5 3.9 215004 (SF7 SRMS) Source-Range Monitor X

A4.04 Ability to manually operate and/or monitor in the control room:

SRMS drive control switches (CFR: 41.7 / 45.5 to 45.8) 3.5 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

K1.14 Knowledge of the physical connections and/or cause and effect relationships between the Average Power Range Monitor/Local Power Range Monitor System and the following systems:

Reactor vessel and internals (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.0 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X 2.2.7 Knowledge of the process for conducting infrequently performed tests or evolutions (CFR: 41.10 / 43.3 / 45.13) 2.9 218000 (SF3 ADS) Automatic Depressurization X

K5.01 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Automatic Depressurization System:

ADS logic operation (CFR: 41.5 / 45.3) 4.3 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

K4.08 Knowledge of Primary Containment Isolation System/Nuclear Steam Supply Shutoff design features and/or interlocks that provide for the following:

Manual defeating of selected isolations during specified emergency conditions (CFR: 41.7) 4.0

239002 (SF3 SRV) Safety Relief Valves X

A2.02 Ability to (a) predict the impacts of the following on the Safety Relief Valves and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:

Leaking SRV (CFR: 41.5 / 43.5 / 45.6) 3.7 259002 (SF2 RWLCS) Reactor Water Level Control X

K3.02 Knowledge of the effect that a loss or malfunction of the Reactor Water Level Control System will have on the following systems or system parameters:

Feedwater system (CFR: 41.7 / 45.4 to 45.8) 4.0 261000 (SF9 SGTS) Standby Gas Treatment X

K1.02 Knowledge of the physical connections and/or cause and effect relationships between the Standby Gas Treatment System and the following systems:

Primary containment system and auxiliaries (CFR: 41.4 to 41.9 / 45.7 to 45.8) 3.7 262001 (SF6 AC) AC Electrical Distribution X

K2.02 Knowledge of electrical power supplies to the following:

(CFR: 41.7)

AC breaker control power 3.7 262001 (SF6 AC) AC Electrical Distribution X

K4.03 Knowledge of AC Electrical Distribution design features and/or interlocks that provide for the following:

Automatic bus transfer (CFR: 41.7) 3.8 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X A3.01 Ability to monitor automatic operation of the Uninterruptable Power Supply (AC/DC), including:

Transfer of power sources (CFR: 41.7 / 45.7) 3.4 263000 (SF6 DC) DC Electrical Distribution X 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication (CFR: 41.7 / 43.5 / 45.4) 4.3 264000 (SF6 EGE) Emergency Generators (Diesel/Jet)

X K4.08 Knowledge of Emergency Generators design features and/or interlocks that provide for the following: Automatic startup (CFR: 41.7) 4.2

264000 (SF6 EGE) Emergency Generators (Diesel/Jet)

X A1.10 Ability to predict and/or monitor changes in parameters associated with operation of the Emergency Generators, including: Lights and alarms (CFR: 41.5 / 45.5) 3.5 300000 (SF8 IA) Instrument Air X

K5.13 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Instrument Air System: Low instrument air pressure (CFR: 41.5 / 45.3) 3.9 400000 (SF8 CCW) Component Cooling Water X

K1.02 Knowledge of the physical connections and/or cause and effect relationships between the Component Cooling Water System and the following systems:

Loads cooled by CCW (CFR: 41.4 to 41.5 / 41.7 to 41.9 /

45.6 to 45.8) 3.8 510000 (SF4 SWS*) Service Water X

K5.01 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Service Water System: Intake/traveling screen high differential pressure/

differential level (CFR: 41.4, 41.7 / 45.5) 3.3 510000 (SF4 SWS*) Service Water X

K6.07 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Service Water System: Loss of AC electrical distribution (CFR: 41.7 / 45.7) 3.6 K/A Category Point Totals:

3 2 2 5 3 2 2 2 2 1 2 Group Point Total:

26

Form 4.1-BWR BWR Examination Outline Page 5 Plant SystemsTier 2/Group 2 (RO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)

IR #

201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control X

A1.06 Ability to predict and/or monitor changes in parameters associated with operation of the Recirculation Flow Control System, including: Reactor core flow (CFR: 41.5 / 45.5) 4.2 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information X

K4.05 Knowledge of Rod Position Information System design features and/or interlocks that provide for the following:

Detection of an uncoupled control rod (CFR: 41.7) 4.0 215001 (SF7 TIP) Traversing In-Core Probe X

A4.03 Ability to manually operate and/or monitor in the control room:

Isolation valves (CFR: 41.7 / 45.5 to 45.8) 3.6 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation X

K3.25 Knowledge of the effect that a loss or malfunction of the Nuclear Boiler Instrumentation will have on the following systems or system parameters:

Vessel pressure (CFR: 41.7 / 45.4) 3.6 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool

Cooling/Cleanup 234000 (SF8 FH) Fuel Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam X 5.0 Components 291002 Sensors and Detectors K1.13 Pressure-Modes of Failures 3.1 239003 (SF9 MSIVLC) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary X

A1.07 Ability to predict and/or monitor changes in parameters associated with operation of the Main Turbine Generator and Auxiliary Systems, including:

First stage turbine pressure (CFR: 41.5 / 45.5) 3.3 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater X

K6.02 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Feedwater System:

Condensate system (CFR: 41.7 / 45.7) 3.5 268000 (SF9 RW) Radwaste X

K2.01 Knowledge of electrical power supplies to the following:

Radiological release isolation valves (CFR: 41.7) 2.7 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring X

K3.05 Knowledge of the effect that a loss or malfunction of the Radiation Monitoring System will have on the following systems or system parameters:

Offgas system (CFR: 41.5 / 45.3) 3.6 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation X

A2.02 Ability to (a) predict the impacts of the following on 3.5

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

Extreme environmental conditions (fire, toxic gas, smoke, radiation, etc.)

(CFR: 41.5 / 43.5 / 45.6) 290002 (SF4 RVI) Reactor Vessel Internals 510001 (SF8 CWS*) Circulating Water X

K3.01 Knowledge of the effect that a loss or malfunction of the Circulating Water System will have on the following systems or system parameters:

Main turbine generator and auxiliary systems (CFR: 41.7 / 45.4) 3.5 K/A Category Point Totals:

0 1

3 1

0 1

2 1

0 1

1 Group Point Total:

11

Form 4.1-COMMON RO Common Examination Outline Facility: Cooper Nuclear Station Date of Exam: June 2022 Generic Knowledge and AbilitiesTier 3 (RO)

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.

2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10 CFR Part 55 (CFR: 41.10 / 43.2) 3.3 2.1.

2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, or switches (CFR: 41.10 / 45.1 / 45.12) 4.1 Subtotal 2

N/A

2.

Equipment Control 2.2.

2.2.13 Knowledge of tagging and clearance procedures (CFR: 41.10 / 43.1 / 45.13) 4.1 2.2.

2.2.6 Knowledge of the process for making changes to procedures (CFR: 41.10 / 43.3 / 45.13) 3.0 Subtotal 2

N/A

3.

Radiation Control 2.3.

2.3.12 Knowledge of radiological safety principles and procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, or alignment of filters (CFR: 41.12 / 43.4 / 45.9 / 45.10) 3.2 Subtotal 1

N/A

4.

Emergency Procedures/

Plan 2.4.

2.4.34 Knowledge of RO responsibilities outside the main control room during an emergency (CFR: 41.10 / 43.5 / 45.13) 4.2 Subtotal 1

N/A Tier 3 Point Total 6

TheoryTier 4 (RO)

Category K/A #

Topic RO IR Reactor Theory 6.1 292008 Reactor Operational Physics K1.01 List parameters that should be monitored and controlled during the approach to criticality 3.9 6.1 292001 Neutrons K1.02 Define prompt and delayed neutrons 3.1 6.1 292005 Control Rods K1.07 Define control rod worth, differential CRW, and integral control rod worth 2.6 Subtotal N/A

Thermodynamics 6.2 293007 Heat Transfer K1.01 Describe three mechanisms of heat transfer 3.2 6.2 293009 Core Thermal Limits K1.10 Define APLHGR 3.7 6.2 293010 Brittle Fracture and Vessel Thermal Stress K1.04 State how the possibility of brittle fracture is minimized by operating limitations 3.2 Subtotal N/A Tier 4 Point Total 6

Form 4.1-BWR SRO Boiling-Water Reactor Examination Outline Notes: CO = Conduct of Operations; EC = Equipment Control; RC = Radiation Control; EM = Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan Facility: Cooper Nuclear Station Date of Exam: June 2022 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Total A2 G

Total

1.

Emergency and Abnormal Plant Evolutions 1

N/A N/A 20 4

3 7

2 6

2 1

3 Tier Totals 26 6

4 10

2.

Plant Systems 1

26 3

2 5

2 11 2

1 3

Tier Totals 37 5

3 8

3.

Generic Knowledge and Abilities Categories CO EC RC EM 6

CO EC RC EM 7

2 2

1 1

2 2

1 2

4. Theory Reactor Theory Thermodynamics 6

3 3

Form 4.1-BWR BWR Examination Outline Page 2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)

E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)

IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation 295003 (APE 3) Partial or Complete Loss of AC Power 295004 (APE 4) Partial or Total Loss of DC Power 295005 (APE 5) Main Turbine Generator Trip 295006 (APE 6) Scram 295016 (APE 16) Control Room Abandonment

/ 7 X

AA2.04 Ability to determine and/or interpret the following as they apply to Control Room Abandonment:

Suppression Pool Temperature (CFR: 41.10 / 43.5 / 45.13) 3.8 295018 (APE 18) Partial or Complete Loss of CCW 295019 (APE 19) Partial or Complete Loss of Instrument Air 295021 (APE 21) Loss of Shutdown Cooling 295023 (APE 23) Refueling Accidents X

AA2.05 Ability to determine and/or interpret the following as they apply to Refueling Accidents:

Emergency plan implementation (CFR: 41.10 / 43.5 / 45.13) 4.4 295024 (EPE 1) High Drywell Pressure 295025 (EPE 2) High Reactor Pressure 295026 (EPE 3) Suppression Pool High Water Temperature X

EA2.01 Ability to determine and/or interpret the following as they apply to Suppression Pool High Water Temperature:

Suppression pool water temperature (CFR: 41.10 / 43.5 / 45.13) 4.0 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only)

X EA2.01 Ability to determine and/or interpret the following as they apply to High Drywell Temperature:

Drywell temperature (CFR: 41.10 / 43.5 / 45.13) 4.2

295030 (EPE 7) Low Suppression Pool Water Level X 2.2.25 Knowledge of the bases in technical specifications for limiting conditions for operation and safety limits (SRO Only)

(CFR: 43.2) 4.2 295031 (EPE 8) Reactor Low Water Level 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown 295038 (EPE 15) High Offsite Radioactivity Release Rate X 2.4.30 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) 4.1 600000 (APE 24) Plant Fire On Site 700000 (APE 25) Generator Voltage and Electric Grid Disturbances X 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator (CFR: 41.10 / 43.5 / 45.13) 3.8 K/A Category Totals:

4 3 Group Point Total:

7

Form 4.1-BWR BWR Examination Outline Page 3 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)

E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum 295007 (APE 7) High Reactor Pressure X

Ability to determine and/or interpret the following as they apply to High Reactor Pressure: AA2.04 Bypass valve capacity (CFR: 41.10 / 43.5 / 45.13) 4.0 295008 (APE 8) High Reactor Water Level 295009 (APE 9) Low Reactor Water Level X

AA2.02 Ability to determine and/or interpret the following as they apply to Low Reactor Water Level:

Steam flow/feed flow mismatch (CFR: 41.10 / 43.5 / 45.13) 3.8 295010 (APE 10) High Drywell Pressure 295011 (APE 11) High Containment Temperature (Mark III Containment only) 295012 (APE 12) High Drywell Temperature 295013 (APE 13) High Suppression Pool Water Temperature/ 5 295014 (APE 14) Inadvertent Reactivity Addition X 2.2.38 Knowledge of conditions and limitations in the facility license (CFR: 41.7 / 41.10 / 43.1 /

45.13) 4.5 295015 (APE 15**) Incomplete Scram 295017 (APE 17) High Offsite Release Rate 295020 (APE 20) Inadvertent Containment Isolation 295022 (APE 22) Loss of Control Rod Drive Pumps 295029 (EPE 6) High Suppression Pool Water Level 295032 (EPE 9) High Secondary Containment Area Temperature 295033 (EPE 10) High Secondary Containment Area Radiation Levels 295034 (EPE 11) Secondary Containment Ventilation High Radiation

/ 9 295035 (EPE 12) Secondary

Containment High Differential Pressure 295036 (EPE 13) Secondary Containment High Sump/Area Water Level 500000 (EPE 16) High Containment Hydrogen Concentration K/A Category Point Totals:

2 1 Group Point Total:

3

Form 4.1-BWR BWR Examination Outline Page 4 Plant SystemsTier 2/Group 1 (SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode 205000 (SF4 SCS) Shutdown Cooling 206000 (SF2, SF4 HPCI)

High-Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

A2.08 Ability to (a) predict the impacts of the following on the Low-Pressure Core Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Valve openings due to malfunction(s)

(CFR: 41.5 / 43.5 / 45.6) 3.5 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Control X

A2.04 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 operations:

Inadequate SLCS system flow (CFR: 41.5 / 43.5 / 45.6) 3.8 212000 (SF7 RPS) Reactor Protection X 2.2.22 Knowledge of limiting conditions for operation and safety limits (CFR: 41.5 / 43.2 / 45.2) 4.7 215003 (SF7 IRM)

Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization X 2.2.12 Knowledge of surveillance procedures (CFR: 41.10 / 43.2 / 45.13) 4.1 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff 239002 (SF3 SRV) Safety Relief Valves

259002 (SF2 RWLCS) Reactor Water Level Control 261000 (SF9 SGTS) Standby Gas Treatment X

A2.03 Ability to (a) predict the impacts of the following on the Standby Gas Treatment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: High train temperature (CFR: 41.5 / 43.5 /

45.6) 3.3 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) 300000 (SF8 IA) Instrument Air 400000 (SF8 CCW) Component Cooling Water 510000 (SF4 SWS*) Service Water K/A Category Point Totals:

3 2 Group Point Total:

5

Form 4.1-BWR BWR Examination Outline Page 5 Plant SystemsTier 2/Group 2 (SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)

IR #

201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor X 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management (CFR: 41.1 / 41.5 / 41.10 /

43.6 / 45.6) 4.6 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries X

A2.09 Ability to (a) predict the impacts of the following on the Primary Containment System and Auxiliaries and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:

Vacuum breaker malfunction (CFR: 41.5 / 45.6) 3.7 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup X

A2.01 Ability to (a) predict the impacts of the following on the Fuel Pool Cooling and Cleanup and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:

Abnormal fuel pool level (CFR: 41.5 / 43.5 / 45.6) 3.8 234000 (SF8 FH) Fuel Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSIVLC) Main Steam Isolation

Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals 510001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

2 1

Group Point Total:

3

Form 4.1-COMMON SRO Common Examination Outline Facility: Cooper Nuclear Station Date of Exam: June 2022 Generic Knowledge and AbilitiesTier 3 (SRO)

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.

2.1.35 Knowledge of the fuel handling responsibilities of SROs (CFR: 43.7) 3.9 2.1.

2.1.15 Knowledge of administrative requirements for temporary management direction, such as standing orders, night orders, or operations memoranda (CFR: 41.10 / 45.12) 3.4 Subtotal N/A 2

2.

Equipment Control 2.2.

2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator (CFR: 41.10 / 43.5 / 45.13) 3.8 2.2.

2.2.43 Knowledge of the process used to track inoperable alarms (CFR: 41.10 / 43.5 / 45.13) 3.3 Subtotal N/A 2

3.

Radiation Control 2.3.

2.3.6 Ability to approve liquid or gaseous release permits (CFR: 41.13 / 43.4 / 45.10) 3.8 Subtotal N/A 1

4.

Emergency Procedures/

Plan 2.4.

2.4.29 Knowledge of the emergency plan implementing procedures (CFR: 41.10 / 43.5 / 45.11) 4.4 2.4.

2.4.52 Knowledge of the lines of authority during implementation of the emergency plan, emergency plan implementing procedures, emergency operating procedures, or severe accident guidelines (CFR: 41.10 / 45.13) 4.0 Subtotal N/A 2

Tier 3 Point Total 7

TheoryTier 4 (RO)

Category K/A #

Topic RO IR Reactor Theory 6

6 6

Subtotal N/A Thermodynamics 6

6 6

Subtotal N/A Tier 4 Point Total 6

Form 4.1-1 Record of Rejected Knowledge and Abilities Tier/Group Randomly Selected K/A Reason for Rejection

Form 4.1-1 Record of Rejected Knowledge and Abilities Refer to Examination Standard (ES)-4.2, Developing Written Examinations, Section B.3, for deviations from the approved written examination outline.

Tier/Group Randomly Selected K/A Reason for Rejection RO T2/G2 239001 Main and Reheat Steam Systems, 6.2, 293005, K1.05 (IR -2.8) 5.0 Components 239001 Main and Reheat Steam Systems, 291002 Sensors and Detectors K1.13 Pressure-Modes of Failures (IR-3.1)

Original question was from Section 6 THEORY of NUREG 1123 Rev 3 and the new question is from Section 5 COMPONENTS of NUREG 1123 Rev 3. Change is IAW with NUREG 1021 REV 12. This change was discussed with the CE.

SRO T1/G1 295016 (APE 16)

Control Room Abandonment

/ 7 AA2.07 Ability to determine and/or interpret the following as they apply to Control Room Abandonment:

Suppression chamber pressure IR 3.4 295016 (APE 16) Control Room Abandonment

/ 7 AA2.04 Ability to determine and/or interpret the following as they apply to Control Room Abandonment:

Suppression Pool temperature IR 3.8 Because CNS procedures do not contain instructions for observing or controlling Suppression Chamber Pressure during control room abandonment, 295016 AA2.07 (Suppression chamber pressure) was replaced with 295016 AA2.04 (Suppression Pool temperature). Page 1 point totals not affected by this change. (Rev 1).

ES-4.1, Page 28 of 28

ES-3.2, Page 11 of 18 Form 3.2-1 Administrative Topics Outline Facility: Cooper Nuclear Station Date of Examination: 6/13/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-06 Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations A.1, Determine TCAs and actions for failure to meet the Time (100% ATWS with MSIVs Closed)

K/A 2.1.2 (4.1)

(R) (N)

Conduct of Operations A.2, Determine Peer checking requirements for Control Room equipment (Multiple)

K/A 2.1.1 (3.8)

(R) (N)

Equipment Control A.3, Determine Proper Sequence For Hanging A Clearance Order (MODE 5 CRD-P-B)

K/A 2.2.13 (4.1)

(R) (N)

Radiation Control N/A N/A Emergency Plan A.4, Identify PAM instrumentation and criteria to qualify it as NUREG 1.97 (Black Diamond Instrumentation)

K/A 2.4.3 (3.7)

(R) (N)

ES-3.2, Page 12 of 18 Instructions for completing Form 3.2-1, Administrative Topics Outline

1. For each license level, determine the number of administrative job performance measures (JPMs) and topic areas as follows:
  • Reactor operator (RO) applicants do not need to be evaluated on every topic (i.e., Equipment Control, Radiation Control, or Emergency Plan can be omitted by doubling up on Conduct of Operations), unless the applicant is taking only the administrative topics portion of the operating test (with a waiver or excusal of the other portions).
2. Enter the associated knowledge and abilities (K/A) statement and summarize the administrative activities for each JPM.
3. For each JPM, specify the type codes for location and source as follows:

Location:

(C)ontrol room, (S)imulator, or Class(R)oom Source and Source Criteria:

(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)

(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes)

(N)ew or Significantly (M)odified from bank (no fewer than one)

Topic Number of JPMs RO*

SRO and RO Retakes Conduct of Operations 1 (or 2) 2 Equipment Control 1 (or 0) 1 Radiation Control 1 (or 0) 1 Emergency Plan 1 (or 0) 1 Total 4

5

Form 3.2-1 Administrative Topics Outline Facility: Cooper Nuclear Station Date of Examination: 6/13/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-06 Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations A5, Determine Watchstanding Requirements (Newly Licensed RO)

K/A G2.1.4 (3.8)

(R) (N)

Conduct of Operations A6, Perform SRO duties for protecting FPC following a Refueling Outage K/A G2.1.25 (4.2)

(R) (N)

Equipment Control A7, Determine LCO and Required Actions (Tech Spec 3.3.2.1 (RWM))

K/A G2.2.35 (4.5)

(R) (N)

Radiation Control A8, Determine Protective Action Recommendations K/A G2.3.14 (3.8) (SRO ONLY)

(R) (D)

Emergency Plan A9, Determine Emergency Classifications EAL (SA6.1)

K/A G2.4.41 (4.6) (SRO ONLY)

(R) (M)

Instructions for completing Form 3.2-1, Administrative Topics Outline

1. For each license level, determine the number of administrative job performance measures (JPMs) and topic areas as follows:
  • Reactor operator (RO) applicants do not need to be evaluated on every topic (i.e., Equipment Control, Radiation Control, or Emergency Plan can be omitted by doubling up on Conduct of Operations), unless the applicant is taking only the administrative topics portion of the operating test (with a waiver or excusal of the other portions).
2. Enter the associated knowledge and abilities (K/A) statement and summarize the administrative activities for each JPM.
3. For each JPM, specify the type codes for location and source as follows:

Location:

(C)ontrol room, (S)imulator, or Class(R)oom Source and Source Criteria:

(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)

(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes)

(N)ew or Significantly (M)odified from bank (no fewer than one)

Topic Number of JPMs RO*

SRO and RO Retakes Conduct of Operations 1 (or 2) 2 Equipment Control 1 (or 0) 1 Radiation Control 1 (or 0) 1 Emergency Plan 1 (or 0) 1 Total 4

5

Form 3.2-2 Control Room/In-Plant Systems Outline Facility: Cooper Nuclear Station Date of Examination: 6/13/2022 Operating Test Number: CNS-2022-06 System/JPM Title Type Code Safety Function Control Room Systems C, D, E, S, A, N, M 1-9 S1. Perform Rapid Power Reduction (Alternate Path)

NRC K/A 202002 A2.05 (3.3/3.7)

RO, SROI, SROU.

M, S, E, A 1

S2. Perform RFPT thrust bearing wear and failure alarm test (Alternate Path-High Vibrations)

NRC K/A 259001 A3.10 (3.9)

RO, SROI.

N, S, A 2

S3. Lowering DEH Pressure Set point in TARGET Mode (If ever used for ILT NRC, not used from 2012 to present)

(developed for 2021-06 Retake exam, entered into ADAMS, but test never administered to examinees)

NRC K/A 241000 A4.06 (4.2)

RO, SROI.

P, S, L 3

S4. Adjust Generator Voltage Regulator (Auto &

Manual) (Alternate Path)

NRC K/A: 245000 A4.12 (3.3)

RO, SROI, SROU.

N, S, A 4

S5. Vent the PC for early venting per 5.8.18 NRC K/A: 223001 K4.08 (3.8), 223001 A4.17 (3.8)

RO.

D, S, E, L 5

S6. Manually Bypass RWM (Restoration)

NRC K/A 201006 A4.01 (3.4)

RO, SROI.

N, S 7

S7. Respond to Loss of Condenser Vacuum due to Ice buildup NRC K/A 295002 AA1.07 (3.4), 510001 A4.01 (3.3)

RO, SROI.

N, S, E 8

S8. Restoration of a Group 6 Primary Containment Isolation (Alternate Path)

NRC K/A: 288000 A4.01 (3.2)

RO, SROI, SROU.

N, S, EN, E, A

9 In-Plant Systems P1. Conduct Alternate Pressure Control (Failure-to-Scram) using Steam Jet Air Ejectors (Not used from 2012 to present)

(developed for 2021-06 Retake exam, entered into ADAMS, but test never administered to examinees)

NRC K/A 295025 EK1.07 (4.2), EK2.12 (3.1), 239001 K1.07 (3.2)

RO, SROI, SROU.

P, R, L, E 3

P2. 5.1ASD Control Building Actions; De-energize RPS Last Used on 17-01 ILT Class (2018-09 Exam)

NRC 295016 AA1.01 (3.8), 212000 K2.01 (3.8) and K2.03 (3.9)

RO, SROI, SROU.

D, E 7

P3. Backwash a Circulating Water Pump NRC K/A 295002 AK2.08 (3.5), 510001 A1.06 (3.9)

RO, SROI.

D, E 8

Form 3.2-2 Instructions for Control Room/In-Plant Systems Outline

1. Determine the number of control room system and in-plant system job performance measures (JPMs) to develop using the following table:
2. Select safety functions and systems for each JPM as follows:

Refer to Section 1.9 of the applicable knowledge and abilities (K/A) catalog for the plant systems organized by safety function. For pressurized-water reactor operating tests, the primary and secondary systems listed under Safety Function 4, Heat Removal from Reactor Core, in Section 1.9 of the applicable K/A catalog, may be treated as separate safety functions (i.e., two systems, one primary and one secondary, may be selected from Safety Function 4). From the safety function groupings identified in the K/A catalog, select the appropriate number of plant systems by safety functions to be evaluated based on the applicants license level (see the table in step 1).

For RO/SRO-I applicants: Each of the control room system JPMs and, separately, each of the in-plant system JPMs must evaluate a different safety function, and the same system or evolution cannot be used to evaluate more than one safety function in each location. One of the control room system JPMs must be an engineered safety feature.

For SRO-U applicants: Evaluate SRO-U applicants on five different safety functions.

One of the control room system JPMs must be an engineered safety feature, and the same system or evolution cannot be used to evaluate more than one safety function.

3. Select a task for each JPM that supports, either directly or indirectly and in a meaningful way, the successful fulfillment of the associated safety function. Select the task from the applicable K/A catalog (K/As for plant systems or emergency and abnormal plant evolutions) or the facility licensees site-specific task list. If this task has an associated K/A, the K/A should have an importance rating of at least 2.5 in the RO column. K/As that have importance ratings of less than 2.5 may be used if justified based on plant priorities; inform the NRC chief examiner if selecting K/As with an importance rating less than 2.5.

The selected tasks must be different from the events and evolutions conducted during the simulator operating test and tasks tested on the written examination. A task that is similar to a simulator scenario event may be acceptable if the actions required to complete the task are significantly different from those required in response to the scenario event.

Apply the following specific task selection criteria:

License Level Control Room In-Plant Total Reactor Operator (RO) 8 3

11 Senior Reactor Operator-Instant (SRO-I) 7 3

10 Senior Reactor Operator-Upgrade (SRO-U) 2 or 3 3 or 2 5

Form 3.2-2 Instructions for Control Room/In-Plant Systems Outline (continued)

At least one of the tasks shall be related to a shutdown or low-power condition.

Four to six of the tasks for RO and SRO-I applicants shall require execution of alternative paths within the facility licensees operating procedures. Two to three of the tasks for SRO-U applicants shall require execution of alternative paths within the facility licensees operating procedures.

At least one alternate path JPM must be new or modified from the bank.

At least one of the tasks conducted in the plant shall evaluate the applicants ability to implement actions required during an emergency or abnormal condition.

At least one of the tasks conducted in the plant shall require the applicant to enter the radiologically controlled area. This provides an excellent opportunity for the applicant to discuss or demonstrate radiation control administrative subjects.

If it is not possible to develop or locate a suitable task for a selected system, return to step 2 and select a different system.

4. For each JPM, specify the codes for type, source, and location: (ACTUAL)

Code License Level Criteria RO SRO-I SRO-U (A)lternate path 4-6 (4) 4-6 (4) 2-3 (3)

(C)ontrol room (D)irect from bank

< 9 (3)

< 8 (2)

< 4 (1)

(E)mergency or abnormal in-plant

> 1 (7)

> 1 (6) > 1 (4)

(EN)gineered safety feature (for control room system)

> 1 (1) > 1 (1) > 1 (1)

(L)ow power/shutdown

> 1 (3) > 1 (2) > 1 (1)

(N)ew or (M)odified from bank (must apply to at least one alternate path JPM)

> 2 (6)

> 2 (6)

> 1 (3)

(P)revious two exams (randomly selected)

< 3 (2)

< 3 (2)

< 2 (1)

(R)adiologically controlled area

> 1 (1) > 1 (1) > 1 (1)

(S)imulator

Form 3.3-1 Scenario Outline Facility:

CNS___________

Scenario #: 1_____________________

Scenario Source:

Op. Test #:

Examiners:

Applicants/

Operators: _________________________

Initial Conditions: MOL (IC19)

Turnover:

Critical Tasks:

Event No.

Malf.

No.

Event Type*

Event Description 1

FW38A R (ATC, CRS) Lower RX Power ATCO Lowers power per 2.1.10 using RR to </=

2381MWth BOP provides peer checking for lowering RR Speed CRS provides the reactivity manager during the change in power.

2 RD11 TS (CRS)

Low Accumulator Pressure 26-27 ATCO addresses alarm card 9-5-2/G-6 CRS enters TS 3.1.5 Condition A.

3a RR50B C (ATC, CRS)

TS (CRS)

B RR Pump will get rising vibrations ATCO will take Alarm card actions which direct tripping B RR (2 dangers)

BOP will take actions per 2.4RR for a tripped RR pump CRS will determine that LCO 3.4.1 is not met and enters Condition B.

3b RR20B C (BOP, CRS) RR Piping leak will cause Drywell pressure to rise ATC will scram the reactor BOP will Vent PC per 2.4PC CRS will direct a reactor scram and 2.4PC entry. (EOP 1A and 3A entry) 3c CR01 M (ATC, BOP, CRS)

Fuel failure will occur from a broken piece of the B RR pump CRS will direct action per 5.1RAD, 2.4OG, and 5.2FUEL.

(EOP 5A entry)

BOP will take actions per 5.1RAD, 2.4OG, and 5.2FUEL.

ATCO will perform Attachment 2 of 5.2FUEL

Form 3.3-1 Scenario Outline 3d RH29A/B C (BOP, CRS) RHR MO-39A or B failure CRS directs the BOP to spray the PC.

BOP will identify that the first loop of RHR will fail when trying to get spray valve control and will need to transition to the other RHR loop which will allow sprays.

4a RD26 RD27 RP01A RP01B RP01C RP01D RP11A RP11B RP11C RP11D RP11E RP11F RP11G RP11H M (ATC, BOP, CRS)

C (BOP, CRS)

MC (BOP)

ATWS conditions CRS enters EOP 1A transitions to EOP 6A and EOP 7A ATCO and BOP take EOP 7A table 41 actions.

o ATCO will inject SLC and lower RX water level to -

60 inches o BOP will install group 1 low level jumpers, inhibit ADS and insert rods per 5.8.3.

o BOP will insert all rods by taking manual action to trip RPS via the TEST trip switches IAW 5.8.3.

CT#1 (BOP)

Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to Scram.

CT #2 (ATCO)

Given a condition where an ATWS has occurred and reactor power is above 3%, the crew stops and prevents injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -60 (or LL, as applicable), prior to neutronic oscillations exceeding 25% peak-to-peak indicated on any APRM.

4b RP15 MC (ATC)

C (ATC, CRS)

Group 6 failure CRS will direct BOP to verify group isolations BOP will determine that a failure of group 6 exists o BOP will insert a group 6 isolation per 2.1.22.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control

Form 3.3-1 Scenario Outline Quantitative Attributes Table Attribute ES-3.3-1 Target Actual Description Events after EOP entry 1-2 3

1. ATWS
2. Failure of AUTO Group 6 isolation
3. Failure of RHR Spray valve control Abnormal Events 2-4 4
1. Low Accumulator pressure (9-5-2/G-6)
2. Rising Vibrations on B RR pump (2.4RR)
3. RR leak (2.4PC)
4. Fuel Failure (5.1RAD, 5.2FUEL, and 2.4OG)

Major Transients 1-2 2

1. Fuel Failure
2. ATWS EOP entries requiring substantive action 1-2 4
1. EOP-6A
2. EOP-7A
3. EOP-3A
4. EOP-5A EOP contingencies requiring substantive action 1 per s

e t

1

1. EOP-6A Contingency #5 - Level/Power Control Pre-identified Critical Tasks 2

2

1. (CT#1) Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to Scram.
2. (CT#2) Given a condition where an ATWS has occurred and reactor power is above 3%, the crew stops and prevents injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -60 (or LL, as applicable), prior to neutronic oscillations exceeding 25% peak-to-peak indicated on any APRM.

Normal Events N/A 0

Reactivity Manipulations N/A 1

1. Lower RX power with RR to less than 2381 MWth (ATC)

Manual Control of Automatic Function 1

2

1. Failure of AUTO Group 6 isolation (ATC)
2. Failure of Automatic Scram (BOP)

Form 3.3-1 Scenario Outline Instrument/

Component Failures N/A 5

1. Rising Vibrations on B RR pump (ATC)
2. Failure of AUTO Group 6 isolation (ATC)
3. Failure of RHR-MO-39A(B) (BOP)
4. Failure of RPS/ARI (BOP)
5. RR Piping Leak (BOP)

Total Malfunctions N/A 6

1. Rising Vibrations on B RR pump
2. Failure of AUTO Group 6 isolation
3. Failure of RHR Spray valve control
4. Fuel Failure
5. Failure of RPS/ARI
6. RR Leak TS Evaluation 2

2

1. LCO 3.1.5 Condition A
2. LCO 3.4.1 Condition B

Form 3.3-1 Scenario Outline Scenario Record Crew Critical Tasks CCT-1 Given a condition where an ATWS has occurred, the crew inhibits ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cool down rate limit during a failure to scram.

This critical task is identified as critical because without operator action, while lowering water level for the ATWS, ADS could initiate and lower reactor pressure below the low pressure ECCS injection pressure and an uncontrolled cold water injection could occur.

SAT / UNSAT CCT-2 Given a condition where an ATWS has occurred and reactor power is above 3%, the crew stops and prevents injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -60 (or LL, as applicable), prior to neutronic oscillations exceeding 25% peak-to-peak indicated on any APRM.

This critical task is identified as critical because without operator action, localized fuel failure could occur.

SAT / UNSAT

Form 3.3-1 Scenario Outline Shift Turnover INITIAL CONDITIONS A.

Plant Status

1.

The plant is at 100% power MOL.

B.

Technical Specifications LCOs in effect

1.

TLCO 3.3.5 Condition A for one LEFM instrument INOPERABLE for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C.

Significant problems/abnormalities

1.

None D.

Overall Risk Status

1.

Green E.

Evolutions/maintenance for the on-coming shift

1.

Lower Rx power to < 2381 MWth using Reactor Recirculation to comply with TLCO 3.3.5 Condition B.

Form 3.3-1 Scenario Outline Facility:

CNS Scenario #: 2 Scenario Source:

Op. Test #:

Examiners:

Applicants/

Operators:

Initial Conditions: EOL (IC25)

Turnover:

Critical Tasks:

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N (BOP, CRS) 6.DWLD.301 BOP Performs surveillance 2a N/A R (BOP, CRS)

Raise RX Power ATCO Monitors 9-5 parameters and Reactor Power on PMIS BOP changes pressure set on DEH from 926 psig to 945 psig to raise power IAW 2.2.77.1 and 2.1.10 for EOL coastdown.

CRS provides the reactivity manager during the change in power.

2b NM14A I (ATC, CRS) Respond to A APRM failure ATCO addresses the alarm card and bypasses APRM A CRS reviews Tech Specs and determines a potential LCO only 3a HV02A TS (CRS) Respond to a minor Earthquake BOP addresses the Drywell vent valve closure CRS determines that LCO 3.4.5 Condition B is not met 3b CU01B RP12 C, MC (BOP)

C (CRS)

Respond to a RWCU leak BOP addresses the rising RWCU pump room temperatures BOP isolates RWCU prior to 2 areas exceeding max safe CRS Enters EOP 5A CT#1 When a primary system is discharging into the secondary containment through a break and automatic isolation fails, the crew takes manual action from the control room which effects isolation of the system prior to reaching Maximum Safe values in two areas for the same parameter.

3c Overrride ZDIPCIS SWS16 to NORM TS (CRS) Respond to a RWCU-MO-18 failure to close CRS determines that LCO 3.6.1.3 is applicable

Form 3.3-1 Scenario Outline 4a RR10A RR11A C (ATC, CRS)

Respond to a #1 and # 2 seal failure on A RR pump (RR-MO-43 will not isolate)

CRS enters 2.4RR BOP addresses 2.4RR ATCO trips the RR pump BOP vents PC per 2.4PC.

CRS enters EOP 1A and EOP 3A on high drywell pressure ATCO scrams on high drywell pressure 4b PC18A C, MC (BOP)

C (CRS)

TS (CRS)

Respond to failure of SGT B on group 6 isolation BOP verifies Group 6 BOP identifies B SGT failed to start and started it.

CRS LCO can address the LCO 3.6.4.3 Condition A following the scenario 5a

HV02A, MS03A M (ATC, BOP, CRS)

Respond to an Earthquake and a MSL leak BOP addresses 5.1QUAKE BOP addresses the rising steam tunnel temperatures BOP attempts to isolate MSL (2 MSLs fail to isolate)

CRS re-enters EOP 5A 5b MS03A RP04 C (BOP, CRS)

Respond to a MSL A and B failure to isolate CRS determines that ED is required BOP identifies that 2 SRV fail to open and place all 8 SRV switches to open CT#2 When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs when maximum safe operating values are exceeded in two areas for the same parameter prior to the fourth area exceeding maximum safe value. (The second and third areas exceed 195F degrees at ~ the same time.)

5c TC07A TC07B TC07C C (ATC, CRS)

Respond to a failure of SRVs C and H to OPEN Bypass valves fail closed (anticipate not available)

ATC opens SRVs until 6 are open

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control

Form 3.3-1 Scenario Outline Quantitative Attributes Table Attribute ES-3.3-1 Target Actual Description Events after EOP entry 1-2 5

1. Failure of SGT train to start
2. Failure of RWCU-MO-18 to Close
3. Failure of RR A seals
4. Failure of MSL A and MSL B to close
5. Failure of 2 SRVs to open Abnormal Events 2-4 3
1. Minor Earthquake (5.1QUAKE)
2. Seal Failure on A RR pump (2.4RR, 2.4PC)
3. APRM A INOP (alarm Card 9-5-1)

Major Transients 1-2 1

1. MSL LEAK EOP entries requiring substantive action 1-2 3
1. EOP-2A
2. EOP-3A
3. EOP-5A EOP contingencies requiring substantive action 1 per set 1
1. EOP-2A Contingency #2 - Emergency Depressurization Pre-identified Critical Tasks 2

2

1. (CT#1) When a primary system is discharging into the secondary containment through a break and automatic isolation fails, the crew takes manual action from the control room which effects isolation of the system prior to reaching Maximum Safe values in two areas for the same parameter.
2. (CT#2) When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs when maximum safe operating values are exceeded in two areas for the same parameter prior to the fourth area exceeding maximum safe value. (The second and third areas exceed 195F degrees at ~

the same time.)

Normal Events N/A 1

1. 6.DWLD.301 (BOP)

Reactivity Manipulations N/A 1

1. Raise RX power with pressure setpoint (BOP)

Manual Control of Automatic Function 1

2

1. Failure of SGT train to AUTO start during a GROUP 6 isolation (BOP)
2. Failure of the Group 3 (AUTO) (BOP)

Form 3.3-1 Scenario Outline Instrument/

Component Failures N/A 6

1. Failure of APRM A (ATC)
2. Failure of SGT train to start (BOP)
3. Failure of RWCU-MO-18 to Close (BOP)
4. Failure of RR A seals (ATC)
5. Failure of MSL A and MSL B to close (BOP)
6. Failure of 2 SRVs to open (ATC)

Total Malfunctions N/A 6

6. Failure of APRM A
7. Failure of SGT train to start
8. Failure of RWCU-MO-18 to Close
9. Failure of RR A seals
10. Failure of MSL A and MSL B to close
11. Failure of 2 SRVs to open TS Evaluation 2

3

1. LCO 3.4.5 Condition B
2. LCO 3.6.1.3 Condition A
3. LCO 3.6.4.3 Condition A

Scenario Record Crew Critical Tasks CCT-1 When a primary system is discharging into the secondary containment through a break and automatic isolation fails, the crew takes manual action from the control room which effects isolation of the system prior to reaching Maximum Safe values in two areas for the same parameter.

This critical task is identified as critical because without operator action, secondary area temperatures could rise high enough to damage plant equipment or be harmful to personnel trying to enter the area to mitigate the event.

SAT / UNSAT CCT-2 When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs when maximum safe operating values are exceeded in two areas for the same parameter prior to the fourth area exceeding maximum safe value. (The second and third areas exceed 195F degrees at ~ the same time.)

This critical task is identified as critical because without operator action, secondary area temperatures could rise high enough to damage plant equipment or be harmful to personnel trying to enter the area to mitigate the event.

SAT / UNSAT

Shift Turnover INITIAL CONDITIONS A.

Plant Status

1.

The plant is at 78% power EOL. In coastdown B.

Technical Specifications LCOs in effect

1.

None C.

Significant problems/abnormalities

1.

None D.

Overall Risk Status

1.

Green E.

Evolutions/maintenance for the on-coming shift

1.

Perform 6.DWLD.301 for quarterly surveillance

2.

Raise Rx power by raising DEH pressure setpoint from 926 to 945 psig IAW 2.2.77.1 and 2.1.10.

Form 3.3-1 Scenario Outline Facility:

CNS Scenario #: 3 Scenario Source:

Op. Test #:

Examiners:

Applicants/

Operators:

Initial Conditions: IC10 Turnover:

Critical Tasks:

Event No.

Malf.

No.

Event Type*

Event Description 1a NM05D C (ATC, CRS)

Respond to IRM D downscale failure (prior to mode switch to RUN)

ATC addresses the alarm card and bypasses IRM D 1b N/A N (ATC, CRS)

Place Rx mode switch to RUN ATC places RX mode switch to run 1c N/A R (ATC, CRS)

Raise RX Power SEQ step 18, Rod group 6 12 rods to get to 25%

bypass ATC withdraws control rods BOP provides peer checking CRS provides the reactivity manager during the change in power.

2 SL06 TS (CRS) Respond to SLC Squib valve power failure CRS determines LCO 3.1.7 is not met and enters Condition A 3a ED20 ED21 C (BOP, CRS)

Respond to Grid Instabilities BOP addresses the alarm card for Grid oscillations and places 1FS and 1GS to P-T-L CRS enters 5.3GRID BOP addresses 5.3GRID actions 3b ED06 TS (CRS) Respond to Loss of ESST When BOP addresses the alarm card for Grid oscillations and places 1FS and 1GS to P-T-L the ESST will lose power CRS determines that LCO 3.8.1 is not met for loss of ESST 4a ED04 M (ATC, BOP, CRS)

Respond to a LOOP Crew Responds to a LOOP

Form 3.3-1 Scenario Outline 4b DG06A DG06B C, MC (BOP)

C (CRS)

Respond to a failure DG1 and DG2 to Auto start BOP places both DG1 and DG2 control switches to start CT#1 When a loss of offsite power occurs with a failure of both DGs to Auto start the crew will place the control switches for either DG prior to an ED being required.

4c RC04 MC (ATC)

C (ATC, CRS)

Respond to a failure of the RCIC controller in AUTO ATC takes manual control of RCIC controller when it fails to control in AUTO.

5 PC08 M (ATC, BOP, CRS)

C (BOP, CRS)

Respond to a TORUS Leak CRS enters EOP 3A for lowering Torus level BOP lines up makeup systems to fill the Torus BOP places HPCI to P-T-L prior to lowering below 11 feet CRS enters EOP 2A BOP opens 6 SRVs CT#2 When the suppression pool level is lowering below 9.6 feet through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs prior torus water level lowering below 6 feet, which would prevent opening SRVs for ED.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control

Form 3.3-1 Scenario Outline Quantitative Attributes Table Attribute ES-3.3-1 Target Actual Description Events after EOP entry 1-2 2

1. DG1 and DG2 fail to Auto Start
2. RCIC controller fail (AUTO)

Abnormal Events 2-4 3

1. IRM D fail downscale (2.3_9-5-1 alarm card)
2. Oscillating GRID Voltage (5.3GRID)
3. LOOP (5.3 EMPOWER)

Major Transients 1-2 2

1. LOOP
2. PC Leak EOP entries requiring substantive action 1-2 3
1. EOP-3A
2. EOP-2A
3. EOP-5A EOP contingencies requiring substantive action 1 per set 1
1. EOP-2A Contingency #2 - Emergency Depressurization Pre-identified Critical Tasks 2

2

1. (CT#1) When a loss of offsite power occurs with a failure of both DGs to Auto start the crew will place the control switches for either DG prior to an ED being required.
2. (CT#2) When the suppression pool level is lowering below 9.6 feet through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs prior torus water level lowering below 6 feet, which would prevent opening SRVs for ED.

Normal Events N/A 1

1. Transfer Reactor Mode Switch to RUN (ATC)

Reactivity Manipulations N/A 1

1. Raise reactor power by withdrawing control rods (ATC)

Manual Control of Automatic Function 1

2

1. DG1 and DG2 fail to Auto Start (BOP)
2. RCIC controller fail (AUTO) (ATC)

Form 3.3-1 Scenario Outline Instrument/

Component Failures N/A 5

1. IRM D fail downscale (ATC)
2. Oscillating GRID Voltage (BOP)
3. RCIC controller fail (ATC)
4. DG1 and DG2 fail to Auto Start (BOP)
5. Torus Leak (BOP)

Total Malfunctions N/A 6

1. IRM D fail downscale
2. Oscillating GRID Voltage
3. RCIC controller fail
4. DG1 and DG2 fail to Auto Start
5. SLC Squib Valve failure
6. Torus Leak TS Evaluation 2

2

1. LCO 3.1.7 Condition A
2. LCO 3.8.1 Condition A

Scenario Record Crew Critical Tasks CCT-1 CT#1 When a loss of offsite power occurs with a failure of both DGs to Auto start the crew will place the control switches for either DG prior to an ED being required.

This critical task is identified as critical because without operator action, following emergency depressurization then there would not be any system able to inject into the vessel and restore/maintain RX water level above of TAF.

SAT / UNSAT CCT-2 CT#2 When the suppression pool level is lowering below 9.6 feet through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs prior torus water level lowering below 6 feet, which would prevent opening SRVs for ED.

This critical task is identified as critical because without operator action SRV availability would be lost prior to being able to get reactor pressure below the decay heat removal pressure.

SAT / UNSAT

Shift Turnover INITIAL CONDITIONS A.

Plant Status

1.

The plant is ~ 5% power BOL in Mode 2 B.

Technical Specifications LCOs in effect

1.

None C.

Significant problems/abnormalities

1.

None D.

Overall Risk Status

1.

Green E.

Evolutions/maintenance for the on-coming shift

1.

Place Reactor Mode switch to RUN

ES-3.4, Page 9 of 10 Form 3.4-1 Events and Evolutions Checklist E

A V

P E

P N

T L

T O

I T

C T

A A

Y S

A B

S A

B S

A B

S A

B L

N P

R T

O R

T O

R T

O R

T O

T E

O C

P O

C P

O C

P O

C P

RO I

U RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

1 0

1 RX 1

0 1

RX 1

1 0

RX RX 1

0 1

0 1

1 RX U

M*

POSITION POSITION POSITION POSITION Facility:

Date of Exam:

Operating Test No.:

Scenarios 1

2 3

4 M

I N

I M

CNS 6/13/2022 1

1 5

2 0

0 2

1 1

6 1

3 2

1 1

5 2

0 3

2 16 5

0 7

X 1

0 2

1 0

0 0

3 2

1 0

1 1

3 1

2 0

0 0

3 1

0 0

0 0

3 2

1 0

1 1

2 2

1 0

1 1

8 5

2 0

X X

2 1

8 5

4 0

0 2

ES-3.4, Page 10 of 10 Form 3.4-1 Instructions for the Events and Evolutions Checklist

1.

Mark the applicant license level for each simulator operating test number.

2.

For the set of scenario columns, fill in the associated event number from Form 3.3-1, Scenario Outline, to show the specific event types being used for the applicant while in the assigned crew position for that scenario.

  • Minimums are subject to the instructions in Section C.2, License Level Criteria.

KEY: RX = Reactivity Manipulation; NOR = Normal Evolution; I/C = Instrument/Component Failure; MAJ = Major Transient; Man. Ctrl = Manual Control of Automatic Function; TS = Technical Specification Evaluation; RO = Reactor Operator; SRO-I or I = Instant Senior Reactor Operator; SRO-U or U = Upgrade Senior Reactor Operator; SRO = Senior Reactor Operator; ATC = At the Controls; and BOP = Balance of Plan

ES-3.4, Page 9 of 10 Form 3.4-1 Events and Evolutions Checklist E

A V

P E

P N

T L

T O

I T

C T

A A

Y S

A B

S A

B S

A B

S A

B L

N P

R T

O R

T O

R T

O R

T O

T E

O C

P O

C P

O C

P O

C P

RO I

U RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

RO SRO-I NOR 1

1 1

I/C 4

4 2

SRO-U MAJ 2

2 1

Man. Ctrl 1

1 0

TS 0

2 2

1 0

1 RX 1

0 1

RX 1

1 0

RX RX 1

0 1

0 1

1 RX U

M*

POSITION POSITION POSITION POSITION Facility:

Date of Exam:

Operating Test No.:

Scenarios 1

2 3

4 M

I N

I M

CNS 6/13/2022 1

1 5

2 0

0 2

3 2

13 5

2 4

1 0

2 2

1 0

0 0

3 2

1 0

0 0

3 1

0 0

2 1

0 1

1 2

2 1

0 1

1 8

5 2

0 X

2 1

11 5

2 3

X X

1 1

1 2

0 1

1 6

1 0

3 3

0 0

1 1

5 2

0 2

3

ES-3.4, Page 10 of 10 Form 3.4-1 Instructions for the Events and Evolutions Checklist

1.

Mark the applicant license level for each simulator operating test number.

2.

For the set of scenario columns, fill in the associated event number from Form 3.3-1, Scenario Outline, to show the specific event types being used for the applicant while in the assigned crew position for that scenario.

  • Minimums are subject to the instructions in Section C.2, License Level Criteria.

KEY: RX = Reactivity Manipulation; NOR = Normal Evolution; I/C = Instrument/Component Failure; MAJ = Major Transient; Man. Ctrl = Manual Control of Automatic Function; TS = Technical Specification Evaluation; RO = Reactor Operator; SRO-I or I = Instant Senior Reactor Operator; SRO-U or U = Upgrade Senior Reactor Operator; SRO = Senior Reactor Operator; ATC = At the Controls; and BOP = Balance of Plan

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