ML23018A042
| ML23018A042 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/22/2022 |
| From: | Heather Gepford NRC/RGN-IV/DORS/OB |
| To: | Nebraska Public Power District (NPPD) |
| References | |
| NLS2022015 | |
| Download: ML23018A042 (1) | |
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: 9/19/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-09 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 K/A 2.1.45 (4.3)
(R) (N)
Conduct of Operations A.2 Obtain and Interpret GARDEL Periodic Case K/A 2.1.7 (4.4)
(R) (D)
Equipment Control A.3 Determine Mechanical and Electrical Isolation Boundaries K/A 2.2.13 (4.1)
(R) (N)
Radiation Control A.4 Determine Dose Requirements and Administrative 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: 9/19/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-09 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 K/A 2.1.4 (3.8)
(R) (N)
Conduct of Operations A.6 Review of Jet Pump Operability K/A 2.1.25 (4.2)
(R) (D)
Equipment Control A.7 Determine Equipment Status Control Requirements K/A 2.2.14 (4.3)
(R) (N)
Radiation Control A.8 Authorize Stable Iodine Thyroid Blocking K/A 2.3.14 (3.8) (SRO ONLY)
(R) (D)
Emergency Plan A.9 Determine Emergency Classifications EAL K/A 2.4.41 (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: 9/19/2022 Operating Test Number: CNS-2022-09 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.)
NRC K/A 400000 A4.01 (3.8)
RO, SROI,.
P, D, S, E, 8
S2. Manually Initiate HPCI Pressure Control Mode (Alternate Path)
NRC K/A 206000 A1.02 (4.1)
RO, SROI, D, S, A, L 4
S3. Perform a Quick Restart of RFPT B (Alternate Path)
(Developed for 19-01 Exam.)
NRC K/A 259001 A4.02 (4.0)
RO, SROI, P, S, A 2
S4. Manually Initiate Drywell Sprays (Alternate Path)
NRC K/A: 226001 A1.01 (4.5)
RO, SROI, SROU, N, S, A, L 5
S5. Conduct RR Pump Quick Restart (Mode 3)
(Alternate Path)
(Developed for 19-01 Re-Take Exam. Was uploaded to Adams, but the Exam was never administered.)
NRC K/A 202001 A4.01 (4.0)
RO.
P, D, S, E, L, A
1
S6. Secure SDG from the Control Room NRC K/A 264000 A4.04 (4.1)
RO, SROI.
N, S 6
S7. Manually Insert a Group 6 Isolation (Alternate Path)
NRC K/A 272000 A2.16 (2.9)
RO, SROI, SROU.
N, S, E, 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, E, 3
In-Plant Systems P1. Placing SW-TCV-451A In Service and in AUTO NRC K/A 400000 A1.02 (3.0)
RO, SROI, SROU.
R, N 7
P2. Securing Fire Pump C Locally NRC K/A 286000 A1.05 (3.1)
RO, SROI, SROU.
D 8
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 (5) 4-6 (4) 2-3 (2)
(C)ontrol room (D)irect from bank
< 9 (5)
< 8 (4)
< 4 (3)
(E)mergency or abnormal in-plant
> 1 (4)
> 1 (3) > 1 (2)
(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 (8)
> 2 (8)
> 1 (6)
(P)revious two exams (randomly selected)
< 3 (3)
< 3 (2)
< 2 (0)
(R)adiologically controlled area
> 1 (2) > 1 (2) > 1 (1)
(S)imulator
Form 3.3-1 Scenario Outline Facility:
CNS___________
Scenario #: 2_____________________
Scenario Source:
New__________
Op. Test #:
Examiners:
Applicants/
Operators: _________________________
Initial Conditions: BOL (IC9)
Turnover:
See attached turnover page on last page.
Critical Tasks:
Identified in the event.
Event No.
Malf.
No.
Event Type*
Event Description 1
Raise RX Power ATC raises power per 10.13 using control rods to continue startup.
BOP provides peer checking for control rod movement.
CRS provides the reactivity manager during the change in power.
IRM C INOP ATC Responds to 9-5-1/C-8, IRM RPS A UPSCALE TRIP OR INOP. ATC will bypass IRM C and reset half scram per procedure 2.1.5.
BOP provides peer check for IRM bypass and half scram reset.
CRS determines that LCO 3.3.1.1 is a potential LCO only.
3 ZDIHVSW EFR1A to OFF MC (BOP)
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 ZDIPCISS WS3A to ON C (BOP, CRS)
TS (CRS)
MSIV 80A Fails Closed BOP will respond to closing MSIV and take control switch for MSIV 80A to close. Will ensure MSL drain in service.
CRS will enter 2.4MSIV for MSIV 80A indicating closed.
CRS will determine LCO 3.3.1.1 is not met and enter Condition A.
5 RR20A 5
RR leak in Drywell Crew takes action per 2.4PC BOP vents containment ATC scrams the reactor
6 RR20A 25-30 over 10 min ramp HP11 FW22A FW22B inserted 10 min after recirc leak starts Manual trigger 5 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 prior to RPV level reaching -183 CFZ.
7 RH08A RH08C ZDIRHRS WS3C to STOP C (ATC, CRS)
MC (ATC)
RHR Pump A and C Failure to Auto Start/ RHR Pump C failure to start ATC will recognize failure of RHR Pump A to auto start on low RPV water level/high drywell pressure and take action to manually start RHR Pump A.
MC (ATC)
Failure of CS-MO-12A ATC will recognize failure of CS-MO-12A to auto open and can be manually opened.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Quantitative Attributes Table Attribute ES-3.3-1 Target Actual Description Events after EOP entry 1-2 2
- 1. RHR Pump A failure to auto start
- 2. CS-MO-12A failure to auto open Abnormal Events 2-4 3
- 1. MSIV 80A Closure (2.4MSIV)
- 2. High Drywell Temperature and Pressure (2.4PC)
- 3. R-2/E-4, REACTOR BLDG EXHAUST FAN FAILURE.
Major Transients 1-2 1
- 1. EOP-3A
2 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 prior to RPV level reaching -183 (CFZ).
Normal Events N/A 0
Reactivity Manipulations N/A 1
- 1. Raise RX power with control rods to continue startup.
(ATC)
Manual Control of Automatic Function 1
3
- 1. Failure of Reactor building exhaust fan to auto start. (BOP)
- 3. CS-MO-12A failure to auto open (ATC)
Instrument/
Component Failures N/A 5
- 2. Reactor building exhaust fan failure (BOP)
- 5. CS-MO-12A failure to auto open (ATC)
Total Malfunctions N/A 10
- 1. IRM C INOP failure
- 2. Reactor building exhaust fan failure
- 3. MSIV 80a fails closed
- 4. HPCI auxiliary oil pump failure
- 5. RHR Pump A and C failure to auto start and C wont start.
- 6. CS-MO-12A failure to auto open
- 7. RHR-MO-25B loses power.
- 8. RF Startup FCVs fail closed
- 9. RF-MO-29/30 fail closed
TS Evaluation 2
2
- 1. LCO 3.3.1.1 Condition A
- 2. LCO 3.6.4.1 Condition A
Scenario Record Crew Critical Tasks 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).
This critical task is identified as critical because without operator action, with lowering water level due to a leak, not lining up injection systems prior to -158(CFZ) could complicate level recovery during ED and cause damage to fuel.
SAT / UNSAT 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 prior to RPV level reaching -183 CFZ.
This critical task is identified as critical because without operator action, localized fuel failure could occur.
SAT / UNSAT
Shift Turnover INITIAL CONDITIONS A.
Plant Status
- 1.
The plant is starting up after a refueling outage at ~7.5% power MOL.
B.
Technical Specifications LCOs in effect
- 1.
Day 1 of LCO 3.5.1 Condition A for Core Spray B tagged out.
C.
Significant problems/abnormalities
- 1.
Core Spray B is tagged out for maintenance.
D.
Overall Risk Status
- 1.
Green E.
Evolutions/maintenance for the on-coming shift
- 1.
Raise Rx power with control rods to continue the plant startup.
Form 3.3-1 Scenario Outline Facility:
CNS___________
Scenario #: 4_____________________
Scenario Source:
New___________
Op. Test #:
Examiners:
Applicants/
Operators: _________________________
Initial Conditions: MOL (IC19)
Turnover:
See Attached Turnover page on last page.
Critical Tasks:
Identified in the Event.
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 ZAORFCL I94B to 0 TS (CRS)
Failure of one channel of RFP high level trip Crew recognizes RFC-LI-94B failed downscale CRS will determine LCO 3.3.2.2 Condition A not met and enter Condition A.
3 Trigger off CS pump Start switch C ( BOP, CRS)
TS (CRS)
Perform 6.1CS.101 Surveillance/CS-MO-5A failure BOP continues surveillance at step 3.18.
CRS will enter LCO 3.5.1 Condition A for CS Pump A inop..
CS-MO-5A will fail closed 10 seconds after pump start. This causes a failure of acceptance criteria.
4 AD06D at 100. Will be removed when reset buttons pushed.
TS (CRS)
SRV-71D Fails Open/2.4SRV BOP Responds to 9-3-1/A-2 Relief Valve Open and 9-3-1/C-1 Safety Relief valve leaking due to SRV-71D failing open.
ATC will rapidly lower power to less than 90%. (ATC)
CRS will enter and direct actions of 2.4SRV.
SRV-71D will close when LLS Logic reset buttons are pushed.
(BOP)
CRS will determine that LCO 3.6.1.6 not met and enter Condition A.
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 bulk torus temperature reaching 110°F.
Loss of Vacuum/Scram BOP will notice megawatts electric lowering and vacuum degrading. Announce 2.4VAC and 2.4OFFGAS entry conditions.
CRS will enter and direct actions for 2.4VAC and 2.4OFFGAS with 2.4VAC as the priority.
ATC will perform a rapid power reduction with Reactor Recirc per
2.1.10 Hard Card to 40 mlbm/hr. (ATC)
CRS will direct a reactor scram when vacuum reaches 23Hg ATC will insert a manual scram. (ATC)
BOP will manually trip the main turbine on the scram. (BOP) 6 MS01A MS07A MS07B MS07E MS07F PC02B M (ATC, BOP, CRS)
MSL A and B Leak Inside Primary Containment Main Steam Lines A and B rupture inside primary containment on the scram and cannot be isolated.
CRS will enter 2.4PC and EOP-3A on rising drywell pressure Torus/Drywell vacuum breaker will fail open causing torus to drywell pressure to track together.
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#2 Given a condition where torus pressure exceeds the pressure suppression pressure, the crew will emergency depressurize the RPV prior to exceeding PCPL-A limit.
7 CS06B MC (ATC)
Failure of Core Spray Pump B auto start ATC will recognize the failure of Core Spray Pump B to auto start on drywell pressure and manually start it.
8 ZDIRHRS WS17A in NAR C (BOP, CRS)
Have PC spray failures go active after the scram.
RHR Loop A Spray Valve Control Failure/ RHR Loop B Failure of Containment Sprays BOP will attempt to spray the torus and drywell and will not be able to.
CRS will determine that ED will be required when torus pressure cannot be maintained below PSP.
Crew will not be able to anticipate due to a Group 1 isolation on low vacuum.
9 RH04B RH08A RH08C RH04A P1134 override to off CS06B C, (ATC, CRS)
MC (ATC)
RHR-MO-25B Loss of Power/RHR Pumps A & C Failure to Auto Start ATC will recognize failure of RHR Pumps A and C to auto start and take action to manually start.
ATC will recognize failure of Core Spray Pump B to auto start and take action to manually start.
ATC will recognize RHR-MO-25B loses power 1 second after switch is taken to open and uses RHR Loop A and Core Spray B to maintain RPV level during ED.
BOP will assist with level control and monitor for ED to be complete.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Quantitative Attributes Table Attribute ES-3.3-1 Target Actual Description Events after EOP entry 1-2 3
- 1. Core Spray Pump B failure to auto start.
- 2. Loss of Spray Valve Control
- 3. RHR-MO-25B failure to open.
Abnormal Events 2-4 2
- 1. Inadvertent SRV opening (2.4SRV)
- 2. Lowering condenser vacuum (2.4VAC/2.4OFFGAS)
Major Transients 1-2 1
- 1. Main Steam Line Break inside Primary Containment EOP entries requiring substantive action 1-2 2
- 1. EOP-3A
2 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 bulk torus temperature reaching 110°F.
CT#2 Given a condition where torus pressure exceeds the pressure suppression pressure, the crew will emergency depressurize the RPV prior to exceeding PCPL-A limit.
Normal Events N/A 1
- 1. Shift RFLO pumps per 2.2.28.1 Reactivity Manipulations N/A 1
- 1. Power reduction to <90%.
Manual Control of Automatic Function 1
3
- 1. Failure of Core Spray Pump B to auto start.
- 2. Failure of RHR Pumps A and C to auto start Instrument/
Component Failures N/A 5
- 1. CS-MO-5A failure (BOP)
- 4. Core Spray Pump B failure to auto start (ATC)
Total Malfunctions N/A 8
- 1. CS-MO-5A failure
- 2. SRV-71D fails open
- 3. Core Spray Pump B failure to auto start
- 4. RHR Loop A Spray Valve Control failure
- 5. RHR-MO-25B failure to open
- 6. RHR-MO-26B and RHR-MO-39B failure to open
- 7. Torus/DW vacuum breaker failure
- 8. MSL A break inside primary containment
TS Evaluation 2
2
- 1. LCO 3.6.1.6 Condition A
- 2. LCO 3.3.2.2 Condition A.
Scenario Record Crew Critical Tasks 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 bulk torus temperature reaching 110°F.
This critical task is identified as critical because without operator action, failing to close the valve or shutdown the reactor would violate BIIT(Boron Injection Initiation Temperature) and challenge the ability of primary containment to absorb the energy in the reactor and violate HCTL.
SAT / UNSAT CT-2 Given a condition where torus pressure exceeds the pressure suppression pressure, the crew will emergency depressurize the RPV prior to exceeding PCPL-A limit.
This critical task is identified as critical because without operator action, exceeding the limit may challenge vent valve operability, SRV operability, and the structural integrity of the primary containment.
SAT / UNSAT
Shift Turnover INITIAL CONDITIONS A.
Plant Status
- 1.
The plant is at 100% power MOL.
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.
Shift RFPT oil pumps for both Reactor Feed Pumps per 2.2.28.1, Section 13.
- 2.
Continue with surveillance 6.1CS.101, Section 3, starting at step 3.18.