ML23018A044

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CNS-2022-11 Final Outlines
ML23018A044
Person / Time
Site: Cooper Entergy icon.png
Issue date: 11/22/2022
From: Heather Gepford
NRC/RGN-IV/DORS/OB
To:
Nebraska Public Power District (NPPD)
References
Download: ML23018A044 (1)
v • d • e


Text

Form 4.1-BWR Boiling-Water Reactor Examination Outline - Rev 12 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: September 19, 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 3

3 N/A 3

4 N/A 4

20 4

3 7

2 1

1 1

1 1

1 6

2 1

3 Tier Totals 4

4 4

4 5

5 26 6

4 10

2.

Plant Systems 1

2 2

2 3

3 2 2

2 2 3 3 26 3

2 5

2 1

1 1

1 1 1 1

1 1 1 1 11 1

1 1

3 Tier Totals 3

3 3

4 4 3 3

3 3 4 4 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 Emergency and Abnormal Plant Evolutions - Tier 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 2.1.20 Ability to interpret and execute procedure steps (CFR: 41.10 / 43.5 / 45.12) 4.6 11 295003 (APE 3) Partial or Complete Loss of AC Power X

AK2.05 Knowledge of the relationship between Partial or Complete Loss of AC Power and the following systems or components: Decay heat removal systems (CFR: 41.7 / 45.8) 4.2 12 295004 (APE 4) Partial or Total Loss of DC Power X

AA1.03 Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of DC Power: AC electrical distribution (CFR: 41.7 / 45.6) 3.5 13 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 14 295006 (APE 6) Scram X

AK1.04 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to SCRAM: Pressure control (CFR: 41.8 to 41.10) 4.1 15 295016 (APE 16) Control Room Abandonment X

AA2.01 Ability to determine and/or interpret the following as they apply to Control Room Abandonment: Reactor Power (CFR: 41.10 / 43.5 / 45.13) 4.3 16 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 17 295019 (APE 19) Partial or Complete Loss of Instrument Air X

2.1.30 Ability to locate and operate components, including local controls (CFR: 41.7 / 45.7) 4.4 18 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 19 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 20 295024 (EPE 1) High Drywell Pressure X

EK3.02 Knowledge of the reasons for the following responses or actions as they apply to High Drywell Pressure: Suppression pool spray (CFR: 41.5 / 45.6) 4.1 21 295025 (EPE 2) High Reactor Pressure X

EK1.05 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to High Reactor Pressure: Exceeding safety limits (CFR: 41.8 to 41.10) 4.6 22 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) 4.1 23 295027 (EPE 4) High Containment Temperature (Mark III Containment Only)

N/A for CNS 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only)

X 2.2.42 Ability to recognize system parameters that are entry-level conditions for technical specifications (CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) 3.9 24

Form 4.1-BWR BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (RO)

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

IR 295030 (EPE 7) Low Suppression Pool Water Level X

EA2.05 Ability to determine and/or interpret the following as they apply to Low Suppression Pool Water Level: ECCS/RCIC pump flow (CFR: 41.10 / 43.5 / 45.13) 4.1 25 295031 (EPE 8) Reactor Low Water Level X

EA1.08 Ability to operate and/or monitor the following as they apply to Reactor Low Water Level: Alternate injection systems (CFR: 41.7 / 45.6) 3.9 26 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown X

EK3.03 Knowledge of the reasons for the following responses or actions as they apply to SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown: Reactor water level control strategies (CFR: 41.5 / 41.7 / 45.6) 4.3 27 295038 (EPE 15) High Offsite Radioactivity Release Rate X

EK2.02 Knowledge of the relationship between High Offsite Radioactivity Release Rate and the following systems or components: Offgas system (CFR: 41.7 / 45.8) 3.8 28 600000 (APE 24) Plant Fire on Site X

AK1.02 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Plant Fire on Site: Firefighting methods for each type of fire (CFR 41.8 / 41.10 / 45.3) 3.4 29 700000 (APE 25) Generator Voltage and Electric Grid Disturbances X

2.4.2 Knowledge of system setpoints, interlocks and automatic actions associated with emergency and abnormal operating procedure entry conditions (CFR: 41.7 / 45.7 / 45.8) 4.5 30 K/A Category Totals:

3 3

3 3

4 4 Group Point Total:

20

Form 4.1-BWR BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 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 X

AA2.03 Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of AC Power: Battery status (CFR: 41.10 / 43.5 / 45.13) 3.9 76 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 295018 (APE 18) Partial or Complete Loss of CCW X

2.2.45 Ability to determine or interpret technical specifications with action statements of greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (SRO Only)

(CFR: 43.2 / 43.5 / 45.3) 4.7 77 295019 (APE 19) Partial or Complete Loss of Instrument Air 295021 (APE 21) Loss of Shutdown Cooling 295023 (APE 23) Refueling Accidents X

AA2.03 Ability to determine and/or interpret the following as they apply to Refueling Accidents:

Airborne contamination levels (CFR: 41.10 / 43.5 / 45.13) 3.2 78 295024 (EPE 1) High Drywell Pressure X

2.4.18 Knowledge of the specific bases for emergency and abnormal operating procedures (CFR: 41.10 / 43.1 / 45.13) 4.0 79 295025 (EPE 2) High Reactor Pressure X

EA2.04 Ability to determine and/or interpret the following as they apply to High Reactor Pressure: Suppression pool level (CFR: 41.10 / 43.5 / 45.13) 3.4 80 295026 (EPE 3) Suppression Pool High Water Temperature 295027 (EPE 4) High Containment Temperature (Mark III Containment Only)

N/A for CNS 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) 295030 (EPE 7) Low Suppression Pool Water Level 295031 (EPE 8) Reactor Low Water Level 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown X

2.4.51 Knowledge of emergency operating procedure exit conditions (e.g., emergency condition no longer exists, or severe accident guideline entry is required)

(CFR: 41.10 / 43.5 / 45.13) 4.0 81 295038 (EPE 15) High Offsite Radioactivity Release Rate 600000 (APE 24) Plant Fire on Site 700000 (APE 25) Generator Voltage and Electric Grid Disturbances X

AA2.06 Ability to determine and/or interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Generator frequency limitations (CFR: 41.5 and 43.5 / 45.5 / 45.7 /

45.8) 3.0 82 K/A Category Totals:

4 3 Group Point Total:

7

Form 4.1-BWR BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 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 295007 (APE 7) High Reactor Pressure 295008 (APE 8) High Reactor Water Level X AK1.03 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to High Reactor Water Level: Feed flow/steam flow mismatch (CFR: 41.8 to 41.10) 3.6 47 295009 (APE 9) Low Reactor Water Level 295010 (APE 10) High Drywell Pressure 295011 (APE 11) High Containment Temperature (Mark III Containment only)

N/A for CNS 295012 (APE 12) High Drywell Temperature X

AK3.02 Knowledge of the reasons for the following responses or actions as they apply to High Drywell Temperature: Venting (CFR: 41.5 / 45.6) 3.9 48 295013 (APE 13) High Suppression Pool Water Temperature/ 5 295014 (APE 14) Inadvertent Reactivity Addition X

AK2.01 Knowledge of the relationship between Inadvertent Reactivity Addition and the following systems or components: RPS (CFR: 41.7 / 45.8) 4.1 49 295015 (APE 15**) Incomplete Scram 295017 (APE 17) High Offsite Release Rate 295020 (APE 20) Inadvertent Containment Isolation X

AA1.04 Ability to operate and/or monitor the following as they apply to Inadvertent Containment Isolation: SGTS / FRVS (CFR: 41.7 / 45.6) 3.6 50 295022 (APE 22) Loss of Control Rod Drive Pumps 295029 (EPE 6) High Suppression Pool Water Level X

2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures (CFR: 41.10 / 43.2 / 45.6) 4.5 51 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 295035 (EPE 12) Secondary Containment High Differential Pressure 295036 (EPE 13) Secondary Containment High Sump/Area Water Level X

EA2.02 Ability to determine and/or interpret the following as they apply to Secondary Containment High Sump/Area Water Level:

Water level in the affected area (CFR: 41.10 / 43.5 / 45.13) 3.5 52 500000 (EPE 16) High Containment Hydrogen Concentration K/A Category Point Totals:

1 1

1 1

1 1 Group Point Total:

6

Form 4.1-BWR BWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 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 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 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 X

2.4.20 Knowledge of the operational implications of emergency and abnormal operating procedures warnings, cautions, and notes (CFR: 41.10 / 43.5 / 45.13) 4.3 83 295034 (EPE 11) Secondary Containment Ventilation High Radiation 295035 (EPE 12) Secondary Containment High Differential Pressure X

EA2.02 Ability to determine and/or interpret the following as they apply to Secondary Containment High Differential Pressure:

Radiation release rate (CFR: 41.8 to 41.10) 3.9 84 295036 (EPE 13) Secondary Containment High Sump/Area Water Level 500000 (EPE 16) High Containment Hydrogen Concentration X

EA2.03 Ability to determine and/or interpret the following as they apply to High Containment Hydrogen Concentration: Hydrogen concentration limits for drywell (CFR 41.10 / 43.5

/ 45.13) 3.8 85 K/A Category Point Totals:

2 1 Group Point Total:

3

Form 4.1-BWR BWR Examination Outline Plant Systems - Tier 2 / Group 1 (RO)

E/APE # /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

2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup (CFR: 41.10 / 45.12) 4.6 1

205000 (SF4 SCS) Shutdown Cooling X

A4.12 Ability to manually operate and/or monitor in the control room: Recirculation loop temperatures (CFR: 41.7 / 45.5 to 45.8) 3.7 2

206000 (SF2, SF4 HPCI)

High-Pressure Coolant Injection X

A3.03 Ability to monitor automatic operation of the High-Pressure Coolant Injection System, including:

System initiation (CFR: 41.7 / 45.7) 4.4 3

207000 (SF4 IC) Isolation (Emergency)

Condenser N/A for CNS 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

A2.0 5 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: Core spray line break (CFR: 41.5 / 43.5 / 45.6) 4.0 4

209002 (SF2, SF4 HPCS)

High-Pressure Core Spray N/A for CNS 211000 (SF1 SLCS) Standby Liquid Control X

A1.01 Ability to predict or monitor changes in parameters associated with operation of the Standby Liquid Control System, including: Tank Level (CFR: 41.5 / 45.5) 3.8 5

212000 (SF7 RPS) Reactor Protection X

X K6.11 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Reactor Protection System:

Primary containment and auxiliaries (CFR: 41.7 / 45.7) 2.1.19 Ability to use available indications to evaluate system or component status (CFR: 41.10 / 45.12) 3.5 3.9 6

7 215003 (SF7 IRM)

Intermediate-Range Monitor X

K5.02 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Intermediate Range Monitor System: Gamma discrimination (CFR: 41.5 / 45.3) 2.6 8

215004 (SF7 SRMS) Source-Range Monitor X

K4.01 Knowledge of Source Range Monitor System design features and/or interlocks that provide for the following: Rod withdrawal blocks (CFR: 41.7) 3.9 9

215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

K3.03 Knowledge of the effect that a loss or malfunction of the Average Power Range Monitor/Local Power Range Monitor System will have on the following systems or system parameters: RMCS (BWR 2, 3, 4, 5)

(CFR: 41.7 / 45.4) 3.4 10 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

X K2.02 Knowledge of electrical power supplies to the following: Initiation/isolation logic (CFR: 41.7)

A4.02 Ability to manually operate and/or monitor in the control room: Turbine trip throttle valve reset (CFR: 41.7 / 45.5 to 45.8) 3.7 4.0 31 32 218000 (SF3 ADS) Automatic Depressurization X

K1.07 Knowledge of the physical connections and/or cause and effect relationships between the Automatic Depressurization System and the following systems: Reactor vessel and internals (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.2 33 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

2.2.22 Knowledge of limiting conditions for operation and safety limits (CFR: 41.5 / 43.2 / 45.2) 4.0 34

Form 4.1-BWR BWR Examination Outline Plant Systems - Tier 2 / Group 1 (RO)

E/APE # /Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G

K/A Topic(s)

IR 239002 (SF3 SRV) Safety Relief Valves X

A4.04 Ability to manually operate and/or monitor in the control room: Suppression pool temperature (CFR: 41.7 / 45.5 to 45.8) 4.2 35 259002 (SF2 RWLCS) Reactor Water Level Control X

X K5.02 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Reactor Water Level Control System: Controller operation (CFR: 41.5 / 45.3)

A3.06 Ability to monitor automatic features of the Reactor Water Level Control System, including:

Reactor water level setpoint set-down following a reactor SCRAM (CFR: 41.7 / 45.7) 3.8 3.6 36 37 261000 (SF9 SGTS) Standby Gas Treatment X

A2.07 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: AC electrical distribution failure (CFR: 41.5 / 43.5 /

45.6) 3.3 38 262001 (SF6 AC) AC Electrical Distribution X

X K4.04 Knowledge of AC Electrical Distribution design features and/or interlocks that provide for the following: Protective relaying (CFR: 41.7)

A1.03 Ability to predict and/or monitor changes in parameters associated with operation of the AC Electrical Distribution, including: Bus voltage (CFR: 41.5 / 45.5) 3.5 3.6 39 40 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X K6.03 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Uninterruptable Power Supply (AC/DC): Static switch/inverter (CFR: 41.7 / 45.7) 3.4 41 263000 (SF6 DC) DC Electrical Distribution X

K5.04 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the DC Electrical Distribution: Ground detection (CFR: 41.5 / 45.3) 2.9 42 264000 (SF6 EGE) Emergency Generators (Diesel/Jet)

X K4.10 Knowledge of Emergency Generators design features and/or interlocks that provide for the following: Automatic start logic (CFR: 41.7) 4.2 43 300000 (SF8 IA) Instrument Air X

K3.25 Knowledge of the effect that a loss or malfunction of the Instrument Air System will have on the following systems or system parameters:

Reactor water cleanup system (CFR: 41.7 / 45.6) 3.2 44 400000 (SF8 CCW) Component Cooling Water X

K2.01 Knowledge of electrical power supplies to the following: CCW Pumps (CFR: 41.7) 3.4 45 510000 (SF4 SWS) Service Water X

K1.05 Knowledge of the physical connections and/or cause and effect relationships between the Service Water System and the following systems:

High-pressure coolant injection system (CFR: 41.4 to 41.8 / 45.7 to 45.8) 3.0 46 K/A Category Point Totals:

2 2

2 3

3 2

2 2

2 3

3 Group Point Total:

26

Form 4.1-BWR BWR Examination Outline Plant Systems - Tier 2 / Group 1 (SRO)

E/APE # /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

A2.17 Ability to (a) predict the impacts of the following on the RHR/LPCI: Injection Mode and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Keep fill system failure (CFR: 41.5 / 43.5 / 45.6) 3.1 86 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 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 87 212000 (SF7 RPS) Reactor Protection 215003 (SF7 IRM)

Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor 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 88 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization 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 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution X

2.2.45 Ability to determine or interpret technical specifications with action statements of greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (SRO Only)

(CFR: 43.2 / 43.5 / 45.3) 4.7 89 264000 (SF6 EGE) Emergency Generators (Diesel/Jet)

X A2.08 Ability to (a) predict the impacts of the following on the Emergency Generators and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Initiation of emergency generator room fire protection system (CFR: 41.5 / 43.5 / 45.6) 3.5 90 300000 (SF8 IA) Instrument Air

Form 4.1-BWR BWR Examination Outline Plant Systems - Tier 2 / Group 1 (SRO)

E/APE # /Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G

K/A Topic(s)

IR 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 Plant Systems - Tier 2 / Group 2 (RO)

E/APE # /Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G

K/A Topic(s)

IR 201001 (SF1 CRDH) CRD Hydraulic X

K5.06 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Control Rod Drive Hydraulic System: Differential pressure indication (CFR: 41.5-7 / 41.10 / 45.1-6 / 45.12-13) 3.4 53 201002 (SF1 RMCS) Reactor Manual Control X

K6.03 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Reactor Manual Control System: Rod worth minimizer (CFR: 41.7 / 45.7) 3.4 54 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

A2.06 Ability to (a) predict the impacts of the following on the Recirculation Flow Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Low reactor water level (CFR: 41.5 / 43.5 / 45.6) 4.0 55 204000 (SF2 RWCU) Reactor Water Cleanup X

A3.04 Ability to monitor automatic operation of the Reactor Water Cleanup System, including: System interlocks and trips (CFR: 41.7 / 45.7) 3.8 56 214000 (SF7 RPIS) Rod Position Information 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 57 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation X

A4.04 Ability to manually operate and/or monitor in the control room: Analog trip units (CFR: 41.7 /

45.5 to 45.8) 3.3 58 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 X

A1.06 Ability to predict and/or monitor changes in parameters associated with operation of the RHR/LPCI: Containment Spray System Mode, including: System flow (CFR: 41.5 / 45.5) 3.8 59 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode X

K1.01 Knowledge of the physical connections and/or cause and effect relationships between the RHR/LPCI: Torus/Suppression Pool Spray Mode and the following systems: Primary containment (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.9 60 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup X

K2.01 Knowledge of electrical power supplies to the following: Fuel pool cooling pumps (CFR: 41.7) 3.1 61 234000 (SF8 FH) Fuel Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSIVLC) Main Steam Isolation Valve Leakage Control

Form 4.1-BWR BWR Examination Outline Plant Systems - Tier 2 / Group 2 (RO)

E/APE # /Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G

K/A Topic(s)

IR 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater X

K3.06 Knowledge of the effect that a loss or malfunction of the Feedwater System will have on the following systems or system parameters: Core inlet subcooling (CFR: 41.7 / 45.4) 3.3 62 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 X

K4.01 Knowledge of Secondary Containment design features and/or interlocks that provide for the following: Personnel access without breaching secondary containment (CFR: 41.7) 3.4 63 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals 510001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

1 1

1 1

1 1

1 1

1 1

1 Group Point Total:

11

Form 4.1-BWR BWR Examination Outline Plant Systems - Tier 2 / Group 2 (SRO)

E/APE # /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 216000 (SF7 NBI) Nuclear Boiler Instrumentation 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 X

A2.19 Ability to (a) predict the impacts of the following on the RHR/LPCI: Containment Spray System Mode and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Low (or negative) suppression chamber pressure during system operation (Mark I, II) (CFR: 41.5 / 43.5 /

45.6) 3.8 91 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel Handling Equipment X

K5.02 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Fuel Handling System: FH equipment interlocks (CFR: 41.5 /

45.3) 3.7 92 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

Form 4.1-BWR BWR Examination Outline Plant Systems - Tier 2 / Group 2 (SRO)

E/APE # /Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G

K/A Topic(s)

IR 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 X

2.2.38 Knowledge of conditions and limitations in the facility license (CFR: 41.7 / 41.10 / 43.1 / 45.13) 4.5 93 290002 (SF4 RVI) Reactor Vessel Internals 510001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

1 1

1 Group Point Total:

3

Form 4.1-COMMON Common Examination Outline Facility: Cooper Nuclear Station Date of Exam: September 19, 2022 (Op Test)

Generic Knowledge and AbilitiesTier 3 (RO/SRO)

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements (CFR: 41.10 / 43.10 / 45.13) 3.8 64 2.1.3 Knowledge of shift or short-term relief turnover practices (CFR: 41.10 / 45.13) 3.7 65 2.1.34 Knowledge of RCS or balance of plant chemistry controls, including parameters measured and reasons for the control (CFR: 41.10 / 43.5 / 45.12) 3.5 94 2.1.36 Knowledge of procedures and limitations involved in core alterations (CFR: 41.10 / 43.6 / 45.7) 4.1 95 Subtotal N/A N/A

2.

Equipment Control 2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels (CFR: 41.6 / 41.7 / 45.2) 4.6 66 2.2.14 Knowledge of the process for controlling equipment configuration or status (CFR: 41.10 /

43.3 / 45.13) 3.9 67 2.2.5 Knowledge of the process for making design or operating changes to the facility, such as 10 CFR 50.59, Changes, Tests and Experiments, screening and evaluation processes, administrative processes for temporary modifications, disabling annunciators, or installation of temporary equipment (CFR: 41.10 / 43.3 / 45.13) 3.2 96 2.2.20 Knowledge of the process for managing troubleshooting activities (CFR: 41.10 / 43.5 /

45.13) 3.8 97 Subtotal N/A N/A

3.

Radiation Control 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 68 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities, such as analysis and interpretation of radiation and activity readings as they pertain to administrative, normal, abnormal, and emergency procedures, or analysis and interpretation of coolant activity, including comparison to emergency plan or regulatory limits (SRO Only) (CFR: 43.4 / 45.10) 3.8 98 Subtotal N/A N/A

4.

Emergency Procedures/

Plan 2.4.3 Ability to identify post-accident instrumentation (CFR: 41.6 / 45.4) 3.7 69 2.4.25 Knowledge of fire protection procedures (CFR: 41.10 / 43.5 / 45.13) 3.7 99

2.4.28 Knowledge of procedures relating to a security event (ensure that the test item includes no safeguards information) (CFR: 41.10 / 43.5 /

45.13) 4.1 100 Subtotal N/A N/A Tier 3 Point Total 6

7

Facility: Cooper Nuclear Station Date of Exam: September 19, 2022 (Op Test)

TheoryTier 4 (RO)

Category K/A #

Topic RO IR Reactor Theory 6.1 292003 Reactor Kinetics and Neutron Sources K1.01 Explain the concept of subcritical multiplication 3.0 70 6.1 292004 Reactivity Coefficients K1.14Compare the relative magnitudes of the temperature, Doppler, and void coefficients of reactivity 3.3 71 6.1 292008 Reactor Operational Physics K1.03 Describe count rate and instrument response that should be observed for rod withdrawal during the approach to criticality 4.1 72 Thermodynamics 6.2 293003 Steam K1.23 Use saturated and superheated steam tables 3.1 73 6.2 293007 Heat Transfer K1.13 Calculate core thermal power using a simplified heat balance 2.9 74 6.2 293009 Core Thermal Limits K1.27 Explain the purpose of the flow biasing correlation factor, (K), as it relates to MCPR limits 3.3 75 Subtotal N/A Tier 4 Point Total 6

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

ES-3.2, Page 11 of 18 Form 3.2-1 Administrative Topics Outline Facility: Cooper Nuclear Station Date of Examination: 11/15/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-11 Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations A.1 Conduct Average Suppression Pool Temperature Calculation (Examinee will conduct calculation and determine one channel high and use alternate temperature point.)

K/A 2.1.45 (4.3)

(R) (N)

Conduct of Operations A.2 Obtain and Interpret GARDEL Periodic Case (Obtain periodic case and recognize 1 GAF out and one safety limit out)

K/A 2.1.7 (4.4)

(R) (D)

Equipment Control A.3 Determine Mechanical and Electrical Isolation Boundaries (Examinee determines tagging requirements for REC Pump B)

K/A 2.2.13 (4.1)

(R) (D)

Radiation Control A.4 Determine Dose Requirements and Administrative Limits (Will determine work area and requirements and calculate total dose to verify exceeding limits.)

K/A 2.3.12 (3.2)

(R) (D)

Emergency Plan N/A N/A

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: 11/15/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-11 Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations A.5 Determine Shift Staffing Requirements for Mode Change (Examinee reviews staffing and determines current requirements and requirements to go to Mode 3.)

K/A 2.1.4 (3.8)

(R) (N)

Conduct of Operations A.6 Review of Jet Pump Operability (Reviews 6.LOG.601 and verifies it was incorrectly filled out and determines LCO entry is required.)

K/A 2.1.25 (4.2)

(R) (D)

Equipment Control A.7 Determine Equipment Status Control Requirements (Reviews work and determines requirements to have TCV out of normal alignment for 35 days.)

K/A 2.2.14 (4.3)

(R) (N)

Radiation Control A.8 Authorize Stable Iodine Thyroid Blocking (Examinee determines KI is required and determines who will receive it for a medical.)

K/A 2.3.14 (3.8) (SRO ONLY)

(R) (D)

Emergency Plan A.9 Determine Emergency Classifications EAL (Examinee will review conditions and determine EAL HU2.1 is required to be declared)

K/A 2.4.41 (4.6) (SRO ONLY)

(R) (N)

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: 11/15/2022 Operating Test Number: CNS-2022-11 Exam Level:

RO SRO-I SRO-U System/JPM Title Type Code Safety Function Control Room Systems C, D, E, S, A, N, M 1-9 S1. Operate Service Water Backups for Critical Loop Cooling (Developed for 19-01 Re-Take Exam - never administered the exam)

NRC K/A 400000 A4.01 (3.8)

RO, SROI, P, D, S, L 8

S2. Manually Initiate HPCI Pressure Control Mode (Alternate Path)

(HPCI speed will stay below min speed and will need to be secured.)

NRC K/A 206000 A1.02 (4.1)

RO, SROI, SROU D, S, A, L 4

S3. Manually Inject into the RPV with Core Spray NRC K/A 209001 A4.05 (4.2)

RO D, S, L 2

S4. Manually Initiate Drywell Sprays (Alternate Path)

(Will need to switch loops to commence drywell sprays)

NRC K/A: 226001 A1.01 (4.5)

RO, SROI, N, S, A, L, EN 5

S5. Withdraw Control Rod from Position 00 (Alternate Path)

(Will try multiple paths to withdraw CR from 0. Will withdraw when Drive water pressure is raised)

D, S, L, A 1

NRC K/A 201003 A2.01 (4.0)

RO, SROI S6. Secure SDG from the Control Room NRC K/A 262001 A2.12 (4.6/4.5)

RO, SROI.

N, S, L 6

S7. Manually Insert a Group 6 Isolation (Alternate Path)

(Need to do both steps to manually initiate the Group 6 Isolation and identify the SGT A didnt start)

NRC K/A 272000 A2.17 (3.4/3.3)

RO, SROI, SROU.

N, S, EN, A 9

S8. Control Reactor Pressure Using Gland Sealing Steam System NRC K/A: 241000 A1.01 (4.4)

RO, SROI, SROU.

N, S, L 3

In-Plant Systems P1. Placing SW-TCV-451A In Service and in AUTO NRC K/A 400000 A1.02 (3.4)

RO, SROI, SROU.

N, R 8

P2. RCIC Overspeed Trip Linkage Reset NRC K/A 217000 A2.02 (4.2/4.0)

RO, SROI, SROU.

E, N, R 2

P3. Shift CRD Drive Filters NRC K/A 201001 A1.03 (3.5)

RO, SROI.

N, R 1

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 (2)

(C)ontrol room (D)irect from bank

< 9 (4)

< 8 (3)

< 4 (1)

(E)mergency or abnormal in-plant

> 1 (1)

> 1 (1) > 1 (1)

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

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

(L)ow power/shutdown

> 1 (7) > 1 (6) > 1 (2)

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

> 2 (7)

> 2 (7)

> 1 (4)

(P)revious two exams (randomly selected)

< 3 (1)

< 3 (1)

< 2 (0)

(R)adiologically controlled area

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

(S)imulator

Op-Test No.: CNS 2022-11 Scenario No.: 2 Page 1 of 2 NUREG 1021 FORM 3.3-1 Scenario Outline Facility: Cooper Nuclear Station Scenario No.: 2_

Scenario Source: IC 10 Operating-Test No.: CNS-2022-11 Examiners: ____________________________ Applicants/Operators: _____________________________

Turnover:

The plant is at 7.5% power in Mode 2 following a refueling outage. CS-P-A is tagged due to an oil leak identified during the startup.

1. Raise Reactor power using control rods per 10.13 and 4.3 to establish 25% bypass valve position.

Critical Tasks:

Identified in the Event Description 6 Event No.

Malf. No.

Event Type*

Event Description 1

R (ATC, CRS)

Raise Reactor Power ATC raises power per 10.13 using control rods to continue startup.

2 Trigger NM13C I (ATC, CRS)

IRM C INOP ATC Responds to 9-5-1/D-7, IRM RPS A UPSCALE TRIP OR INOP. ATC will bypass IRM C and reset half scram per procedure 2.1.5.

3 Trigger ZDIHVSWEFR1A to OFF MC (BOP)

C (BOP, CRS)

TS (CRS)

Trip of Reactor Building Exhaust Fan BOP will respond to R-2/E-4, REACTOR BLDG EXHAUST FAN FAILURE. Recognize failure of standby exhaust fan to start and manually start it.

BOP will inform CRS that secondary containment went above

-0.25 and reference LCO 3.6.4.1.

CRS will determine that LCO 3.6.4.1 was not met and enter Condition A.

4 Trigger RC05 MC (BOP)

C (BOP, CRS)

TS (CRS)

RCIC Inadvertent Initiation BOP will respond to 9-4-1/A-1, RCIC Logic Actuated and take alarm card and/or 2.4CSCS actions to shutdown RCIC.

CRS will enter 2.4CSCS and direct actions.

CRS will determine LCO 3.5.3 is not met and enter Condition A.

5 Trigger RR20A 5

C (ATC, BOP, CRS)

RR leak in Drywell Crew takes action per 2.4PC.

Op-Test No.: CNS 2022-11 Scenario No.: 2 Page 2 of 2 BOP vents containment.

ATC scrams the reactor at ~ 1.84 psig drywell pressure.

6 inserted on the scram RR20A 5-30 over 10 min ramp ZDIRFSW29M V to CLOSE 60 sec TD ZDIRFSW30M V to CLOSE 60 sec TD inserted on the scram HP11 FW22A FW22B M (ATC, BOP, CRS)

LOCA/ED/ HPCI Failure to Operate/SU valves fail closed/RF-MO-29/30 Fail Closed.

CRS will direct inhibiting ADS prior to -113 low RPV level.

BOP will inhibit ADS.

ATC will prepare to line up systems for injection.

ED at -158 RPV water level.

CT#1 Given a condition where a LOCA is occurring and RPV level is lowering, the crew will line up the two available injection subsystems for injection prior to RPV level going below -158 (CFZ).

CT#2 Given a condition where a LOCA is occurring and RPV level is lowering, the crew will enter EOP-2A when RPV level reaches

-158 CFZ and commence emergency depressurization by opening at least 4 SRVs prior to RPV level reaching -183 CFZ.

7 RH08A RH08B RH08C RH08D CS06B RH01C C (ATC, CRS)

MC (ATC)

Low pressure ECCS pumps failure to Auto Start/ RHR Pump C Trip/RHR-MO-25B Loss of Power ATC will recognize failure of all RHR and CS pumps to auto start on low RPV water level/high drywell pressure and take action to manually start RHR Pump A.

ATC will recognize loss of power to RHR-MO-25B and recognize only RHR Loop A and Core Spray Pump B are only injection sources.

8 AD06E &

AD06H to 0 C (BOP, CRS)

Failure of SRV E and H to Open

  • BOP will recognize failure of SRV-71E and SRV-71H to open and open two additional SRVs.

9 CS02B C (ATC, CRS)

MC (ATC)

Failure of CS-MO-12B to Auto Open ATC will recognize failure of CS-MO-12B to auto open and can be manually opened.

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

NUREG 1021 FORM 3.3-1 Scenario Outline Facility: Cooper Nuclear Station Scenario No.: 3__(Spare)

Scenario Source: IC 16 Operating-Test No.: CNS-2022-11 Examiners: ____________________________ Applicants/Operators: _____________________________

Turnover:

The plant is at 75% power BOL.

1. Shift TEC pumps per 2.2.76 Section 7.
2. Raise Reactor power to 78% using Reactor Recirculation IAW 2.1.10, to commence power increase following quarterly down power.

Critical Tasks:

Identified in the Event Description 7.

Event No.

Malf.

No.

Event Type*

Event Description 1

N (BOP, CRS)

Shift TEC pumps per 2.2.76 Section 7.

BOP Shifts TEC pumps.

2 R (ATC, CRS)

Raise Reactor Power ATC Raises power per 2.1.10 using RR to ~78% power for power ascension following downpower for control rod adjustment.

3 Trigger NM09C to 0

I (ATC, CRS)

APRM C Fails Downscale ATC will respond to 9-5-1/C-8, APRM downscale and take alarm card actions to bypass APRM C.

4 Trigger RP03C C (BOP, ATC, CRS)

TS (CRS)

RPS B Loss of Power BOP transfers RPS B to the alternate source.

BOP will verify and reset group isolations IAW 2.1.22.

ATC resets the B half scram IAW 2.1.5 CRS determines that LCO 3.4.5 is not met and enters Condition B.

5 Trigger HP05 MC (BOP)

C (BOP, CRS)

TS (CRS)

Spurious HPCI Initiation BOP will take immediate operator actions to secure HPCI.

CRS will determine LCO 3.5.1 is not met and enter Condition C.

6 Trigger ED04 ZDIDGSW C (ATC, BOP, CRS)

LOOP/Failure of DG1 to start/ Failure of DG2 to auto start BOP closes breaker EG2 when EG2 failed to AUTO close when

Op-Test No.: CNS 2022-09 Scenario No.: 3 CSDG1 to NASP MC (BOP)

DG2 got to rated power.

ATC takes scram actions and places the Reactor Mode switch to Refuel/Shutdown.

7 Trigger HP06 HP09 ZDIHPCIS WS1 to open ZDIHPCIS WS2 to open ZDIHPCIS WS32 to OFF M (ATC, BOP, CRS)

HPCI Steam Line Break/Failure to isolate HPCI line/ED BOP/ATC attempts to isolate HPCI by closing HPCI-MO-15 and HPCI-MO-16. Reports that all actions from the control room failed to isolate HPCI steam line.

CRS will enter EOP-2A when 2nd area reaches maximum safe operating temperature and direct emergency depressurization.

CT#1 Given a condition when high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to loss of AC power, crew manually closes DG2 output breaker or directs the NLO to close EG2 locally to energize LP ECCS systems prior to RPV water level lowering below -158 CFZ (TAF).

CT#2 Given a condition where two areas exceed their maximum safe operating temperature, the crew will enter EOP-2A and commence emergency depressurization of the RPV by opening at least 4 SRVs prior to a third area reaching its maximum safe operating temperature. (MSOT). (The first area to exceed MSOT is the 903,931 with TS-126A & C. The second area to exceed the MSOT is the SW Quad with TS-99G/TS-105B. The third area to exceed the MSOT is the NW Quad with TS-99C.)

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

Op-Test No.: CNS 2022-11 Scenario No.: 4 NUREG 1021 FORM 3.3-1 Scenario Outline Facility: Cooper Nuclear Station Scenario No.: 4__

Scenario Source: IC 19 Operating-Test No.: CNS-2022-11 Examiners: ____________________________ Applicants/Operators:

Turnover:

The plant is at 100% power MOL.

1. Shift RFLO pumps for both Reactor Feed Pumps IAW 2.2.28.1, Section 13.

Critical Tasks:

Identified in the Event Descriptions 3, 7 and 10.

Event No.

Malf.

No.

Event Type*

Event Description 1

N (BOP, CRS) Shift RFLO pumps on both RFPT per Section 13 of 2.2.28.1.

2 Trigger RR27B to 0

TS (CRS)

Failure of One Channel of RFP High Level Trip ATC addresses RVLC SYSTEM TROUBLE annunciator. (I)

CRS will determine LCO 3.3.2.2 not met and enter Condition A.

(TS) 3 Trigger AD06D at 100. Will be removed when reset buttons pushed.

AD05 to out C (BOP, CRS)

R (ATC, CRS)

TS (CRS)

SRV-71D Fails Open/2.4SRV ATC will rapidly lower power to less than 90%. (R)

BOP takes actions per 2.4SRV and when LLS Logic reset buttons are pushed SRV-D closes. (C)

CRS will determine that LCO 3.6.1.6 not met and enter Condition A. (TS)

CT#1 Given a condition where an SRV fails open, the crew will take action to close the valve or insert a manual scram prior to average torus water temperature reaching 110°F.

Op-Test No.: CNS 2022-11 Scenario No.: 4 4

Trigger ZDIHVS WEFT1B to OFF ZDIHVS WEFT1d to OFF C (BOP, CRS)

Loss of 2 Turbine Building Exhaust Fans BOP will take the Turbine building exhaust fan in OFF to AUTO/RUN per alarm card action. (R-1/A-1, R-1/A-2, R-1/A-3, R-1/B-3, R-1/C-3) 5 Trigger FW12A ramp to 10 in 60 seconds C (ATC, CRS)

FW Flow Transmitter Oscillations CRS will enter 2.4RXLVL.

ATCO will place the level control select to 1 Element Cont (Single Element Control).

6 Trigger MC01 C (ATC, BOP CRS)

Loss of Vacuum/Scram ATC will perform a rapid power reduction with Reactor Recirc per 2.1.10 Hard Card to 40 mlbm/hr.(C)

ATC will insert a manual scram. (C)

BOP will manually trip the main turbine on the scram. (C) 7 Off scram RR20A PC12 M (ATC, BOP, CRS)

RR Leak Inside Primary Containment/ED on PSP CRS will direct torus sprays and drywell sprays IAW EOP-3A.

CRS will enter EOP-2A once PSP has been exceeded.

BOP will open 6 SRVs for emergency depressurization.

ATC will maintain RPV water level.

CT#3 Given a condition where torus pressure exceeds the pressure suppression pressure, the crew will emergency depressurize the RPV prior to exceeding PCPL-A limit.

8 CS06B MC (ATC)

C (ATC, CRS)

Failure of Core Spray Pump B Auto Start ATC will recognize the failure of Core Spray Pump B to auto start on high drywell pressure and manually start it. (MC, C) 9 ZDIRHRS WS9A and 9B to Close.

Array to force first loop of torus sprays to fail.

C (BOP, CRS)

Loss of Drywell Sprays/One Loop of Torus Sprays BOP will attempt to spray the torus. The first RHR Loop will fail and cause BOP to use other loop. Drywell sprays will not be available (C) 10 HP01 MC(ATC)

C(ATC, CRS)

Failure of HPCI to Auto Start ATC will manually start HPCI per the Hardcard for injection CT#2 Given a condition where the operating injection systems cannot maintain RPV level and HPCI fails to automatically align for injection, the crew manually aligns HPCI for injection prior to RPV water level lowering below -158 CFZ (TAF).

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