ML22214A610
| ML22214A610 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/21/2022 |
| From: | Heather Gepford NRC/RGN-IV/DORS/OB |
| To: | Nebraska Public Power District (NPPD) |
| References | |
| Download: ML22214A610 (72) | |
Text
Form 4.1-BWR RO Boiling-Water Reactor Examination Outline Notes: CO = Conduct of Operations; EC = Equipment Control; RC = Radiation Control; EM = Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan Facility: Cooper Nuclear Station Date of Exam: June 2022 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Total A2 G
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
3 4
3 N/A 4
4 N/A 2
20 7
2 1
2 1
1 1
0 6
3 Tier Totals 4
6 4
5 5
2 26 10
- 2.
Plant Systems 1
3 2
2 5
3 2 2
2 2 1 2 26 5
2 0
1 3
1 0 1 2
1 0 1 1 11 3
Tier Totals 3
3 5
6 3 3 4
3 2 2 3 37 8
- 3.
Generic Knowledge and Abilities Categories CO EC RC EM 6
CO EC RC EM 7
2 2
1 1
- 4. Theory Reactor Theory Thermodynamics 6
3 3
Form 4.1-BWR BWR Examination Outline Page 2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO)
E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)
IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation X
AK1.04 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation:
Thermal-hydraulic instabilities (CFR: 41.8 to 41.10) 4.3 1
295003 (APE 3) Partial or Complete Loss of AC Power X
AK1.02 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Partial or Complete Loss of AC Power:
Load shedding (CFR: 41.8 to 41.10) 3.8 2
295004 (APE 4) Partial or Total Complete Loss of DC Power X
AA1.01 Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of DC Power:
DC electrical distribution (CFR: 41.7 / 45.6) 3.8 3
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 4
295006 (APE 6) Scram X
AK2.03 Knowledge of the relationship between SCRAM and the following systems or components:
CRD hydraulic system (CFR: 41.7 / 45.8) 3.9 5
295016 (APE 16) Control Room Abandonment X 2.4.12 Knowledge of operating crew responsibilities during emergency and abnormal operations (CFR: 41.10 / 45.12) 4.0 6
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 7
295019 (APE 19) Partial or Complete Loss of Instrument Air x
AA2.01 Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Instrument Air:
Instrument air pressure (CFR: 41.10 / 43.5 / 45.13) 4.1 8
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 9
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 10 295024 (EPE 1) High Drywell Pressure X
EK1.01 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to High Drywell Pressure: Drywell integrity (CFR: 41.8 to 41.10) 4.3 11 295025 (EPE 2) High Reactor Pressure X
EK2.01 Knowledge of the relationship between High Reactor Pressure and the following systems or components:
RPS (CFR: 41.7 / 45.8) 4.2 12 295026 (EPE 3) Suppression Pool High Water Temperature X
EA1.08 Ability to operate and/or monitor the following as they apply to Suppression Pool High Water Temperature: LPCS (CFR: 41.7 / 45.6) 3.8 13 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only)
X EK3.03 Knowledge of the reasons for the following responses or actions as they apply to High Drywell Temperature: Drywell spray (CFR: 41.5 / 45.6) 3.8 14 295030 (EPE 7) Low Suppression Pool Water Level X
EA1.04 Ability to operate and/or monitor the following as they 3.5 15
RED = Topic sampled on SRO Exam apply to Low Suppression Pool Water Level:
Suppression pool makeup system(s)
(CFR: 41.7 / 45.6) 295031 (EPE 8) Reactor Low Water Level X
EA2.04 Ability to determine and/or interpret the following as they apply to Reactor Low Water Level: Adequate core cooling (CFR: 41.10 / 43.5 / 45.13) 4.9 16 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown X
EK2.02 Knowledge of the relationship between SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown and the following systems or components:
Redundant reactivity control system (CFR: 41.7 41.8 / 45.8) 4.0 17 295038 (EPE 15) High Offsite Radioactivity Release Rate X
EA2.04 Ability to determine and/or interpret the following as they apply to High Offsite Radioactivity Release Rate:
Source of offsite release (CFR: 41.10 / 43.5 / 45.13) 3.8 18 600000 (APE 24) Plant Fire On Site X 2.4.26 Knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage (CFR: 41.10 / 43.5 / 45.12) 3.1 19 700000 (APE 25) Generator Voltage and Electric Grid Disturbances X
AK3.01 Knowledge of the reasons for the following responses or actions as they apply to Generator Voltage and Electric Grid Disturbances:
Reactor and turbine trip criteria (CFR: 41.4 / 41.5 / 41.7 / 41.10 /
45.8) 3.8 20 K/A Category Totals:
3 4 3 4 4 2 Group Point Total:
20
Form 4.1-BWR BWR Examination Outline Page 3 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)
E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)
IR 295002 (APE 2) Loss of Main Condenser Vacuum X
AA1.12 Ability to operate and/or monitor the following as they apply to Loss of Main Condenser Vacuum:
Condenser air removal system (CFR: 41.7 / 45.6) 3.5 21 295007 (APE 7) High Reactor Pressure 295008 (APE 8) High Reactor Water Level 295009 (APE 9) Low Reactor Water Level 295010 (APE 10) High Drywell Pressure 295011 (APE 11) High Containment Temperature (Mark III Containment only) 295012 (APE 12) High Drywell Temperature 295013 (APE 13) High Suppression Pool Water Temperature/ 5 295014 (APE 14) Inadvertent Reactivity Addition 295015 (APE 15**) Incomplete Scram 295017 (APE 17) High Offsite Release Rate 295020 (APE 20) Inadvertent Containment Isolation 295022 (APE 22) Loss of Control Rod Drive Pumps X
AA2.01 Ability to determine and/or interpret the following as they apply to Loss of Control Rod Drive Pumps:
Accumulator pressure (CFR: 41.10 / 43.5 / 45.13) 4.0 22 295029 (EPE 6) High Suppression Pool Water Level 295032 (EPE 9) High Secondary Containment Area Temperature X
EK3.02 Knowledge of the reasons for the following responses or actions as they apply to High Secondary Containment Area Temperature:
Reactor SCRAM (CFR: 41.5 / 45.6) 4.0 23 295033 (EPE 10) High Secondary
Containment Area Radiation Levels 295034 (EPE 11) Secondary Containment Ventilation High Radiation
/ 9 X
EK2.01 Knowledge of the relationship between Secondary Containment Ventilation High Radiation and the following systems or components:
Process radiation monitoring system (CFR: 41.7 / 45.8) 3.7 24 295035 (EPE 12) Secondary Containment High Differential Pressure 295036 (EPE 13) Secondary Containment High Sump/Area Water Level X
EK1.04 Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Secondary Containment High Sump/Area Water Level:
Maximum safe operating limit (CFR: 41.8 to 41.10) 4.1 25 500000 (EPE 16) High Containment Hydrogen Concentration X
EK2.05 Knowledge of the relationship between High Containment Hydrogen Concentration and the following systems or components:
Hydrogen and oxygen recombiners (CFR: 41.7 / 45.8)
EK2.01 Knowledge of the relationship between High Containment Hydrogen Concentration and the following systems or components:
Containment hydrogen monitors (CFR: 41.7 / 45.8) 3.5 3.8 26 K/A Category Point Totals:
1 2 1 1 1 0 Group Point Total:
6
Form 4.1-BWR BWR Examination Outline Page 4 Plant SystemsTier 2/Group 1 (RO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)
IR 203000 (SF2, SF4 RHR/LPCI)
RHR/LPCI: Injection Mode X
K2.01 Knowledge of electrical power supplies to the following:
Pumps (CFR: 41.7) 4.1 27 203000 (SF2, SF4 RHR/LPCI)
RHR/LPCI: Injection Mode X
K4.01 Knowledge of RHR/LPCI:
Injection Mode design features and/or interlocks that provide for the following:
Automatic system initiation/injection (CFR: 41.7) 4.4 28 205000 (SF4 SCS) Shutdown Cooling System (RHR Shutdown Cooling Mode)
X A2.08 Ability to (a) predict the impacts of the following on the Shutdown Cooling System (RHR Shutdown Cooling Mode) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Loss of heat exchanger cooling (CFR: 41.5 / 45.6) 4.0 29 206000 (SF2, SF4 HPCI)
High-Pressure Coolant Injection X
A3.09 Ability to monitor automatic operation of the High-Pressure Coolant Injection System, including:
System isolation (CFR: 41.7 / 45.7) 4.3 30 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)
Low-Pressure Core Spray X
K6.08 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Low-Pressure Core Spray System: Keep fill system (CFR: 41.7 / 45.7) 3.4 31 209002 (SF2, SF4 HPCS)
High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Control X
K4.04 Knowledge of Standby Liquid Control System design features and/or interlocks that provide for the following:
Indication of a fault in squib valves firing circuits (CFR: 41.7) 3.7 32
212000 (SF7 RPS) Reactor Protection X
K3.06 Knowledge of the effect that a loss or malfunction of the Reactor Protection System will have on the following systems or system parameters:
SCRAM air header solenoid-operated valves (CFR: 41.7 / 45.4) 4.1 33 215003 (SF7 IRM)
Intermediate-Range Monitor X
A1.05 Ability to predict and/or monitor changes in parameters associated with operation of the Intermediate Range Monitor System, including:
SCRAM and rod block trip setpoints (CFR: 41.5 / 45.5 3.9 34 215004 (SF7 SRMS) Source-Range Monitor X
A4.04 Ability to manually operate and/or monitor in the control room:
SRMS drive control switches (CFR: 41.7 / 45.5 to 45.8) 3.5 35 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X
K1.14 Knowledge of the physical connections and/or cause and effect relationships between the Average Power Range Monitor/Local Power Range Monitor System and the following systems:
Reactor vessel and internals (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.0 36 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X 2.2.7 Knowledge of the process for conducting infrequently performed tests or evolutions (CFR: 41.10 / 43.3 / 45.13) 2.9 37 218000 (SF3 ADS) Automatic Depressurization X
K5.01 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Automatic Depressurization System:
ADS logic operation (CFR: 41.5 / 45.3) 4.3 38 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X
K4.08 Knowledge of Primary Containment Isolation System/Nuclear Steam Supply Shutoff design features and/or interlocks that provide for the following:
Manual defeating of selected isolations during specified emergency conditions (CFR: 41.7) 4.0 39
239002 (SF3 SRV) Safety Relief Valves X
A2.02 Ability to (a) predict the impacts of the following on the Safety Relief Valves and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Leaking SRV (CFR: 41.5 / 43.5 / 45.6) 3.7 40 259002 (SF2 RWLCS) Reactor Water Level Control X
K3.02 Knowledge of the effect that a loss or malfunction of the Reactor Water Level Control System will have on the following systems or system parameters:
Feedwater system (CFR: 41.7 / 45.4 to 45.8) 4.0 41 261000 (SF9 SGTS) Standby Gas Treatment X
K1.02 Knowledge of the physical connections and/or cause and effect relationships between the Standby Gas Treatment System and the following systems:
Primary containment system and auxiliaries (CFR: 41.4 to 41.9 / 45.7 to 45.8) 3.7 42 262001 (SF6 AC) AC Electrical Distribution X
K2.02 Knowledge of electrical power supplies to the following:
(CFR: 41.7)
AC breaker control power 3.7 43 262001 (SF6 AC) AC Electrical Distribution X
K4.03 Knowledge of AC Electrical Distribution design features and/or interlocks that provide for the following:
Automatic bus transfer (CFR: 41.7) 3.8 44 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)
X A3.01 Ability to monitor automatic operation of the Uninterruptable Power Supply (AC/DC), including:
Transfer of power sources (CFR: 41.7 / 45.7) 3.4 45 263000 (SF6 DC) DC Electrical Distribution X 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication (CFR: 41.7 / 43.5 / 45.4) 4.3 46 264000 (SF6 EGE) Emergency Generators (Diesel/Jet)
X K4.08 Knowledge of Emergency Generators design features and/or interlocks that provide for the following: Automatic startup (CFR: 41.7) 4.2 47
264000 (SF6 EGE) Emergency Generators (Diesel/Jet)
X A1.10 Ability to predict and/or monitor changes in parameters associated with operation of the Emergency Generators, including: Lights and alarms (CFR: 41.5 / 45.5) 3.5 48 300000 (SF8 IA) Instrument Air X
K5.13 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Instrument Air System: Low instrument air pressure (CFR: 41.5 / 45.3) 3.9 49 400000 (SF8 CCW) Component Cooling Water X
K1.02 Knowledge of the physical connections and/or cause and effect relationships between the Component Cooling Water System and the following systems:
Loads cooled by CCW (CFR: 41.4 to 41.5 / 41.7 to 41.9 /
45.6 to 45.8) 3.8 50 510000 (SF4 SWS*) Service Water X
K5.01 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Service Water System: Intake/traveling screen high differential pressure/
differential level (CFR: 41.4, 41.7 / 45.5) 3.3 51 510000 (SF4 SWS*) Service Water X
K6.07 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Service Water System: Loss of AC electrical distribution (CFR: 41.7 / 45.7) 3.6 52 K/A Category Point Totals:
3 2 2 5 3 2 2 2 2 1 2 Group Point Total:
26
Form 4.1-BWR BWR Examination Outline Page 5 Plant SystemsTier 2/Group 2 (RO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)
IR #
201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control X
A1.06 Ability to predict and/or monitor changes in parameters associated with operation of the Recirculation Flow Control System, including: Reactor core flow (CFR: 41.5 / 45.5) 4.2 53 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information X
K4.05 Knowledge of Rod Position Information System design features and/or interlocks that provide for the following:
Detection of an uncoupled control rod (CFR: 41.7) 4.0 54 215001 (SF7 TIP) Traversing In-Core Probe X
A4.03 Ability to manually operate and/or monitor in the control room:
Isolation valves (CFR: 41.7 / 45.5 to 45.8) 3.6 55 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation X
K3.25 Knowledge of the effect that a loss or malfunction of the Nuclear Boiler Instrumentation will have on the following systems or system parameters:
Vessel pressure (CFR: 41.7 / 45.4) 3.6 56 219000 (SF5 RHR SPC) RHR/LPCI:
Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries 226001 (SF5 RHR CSS) RHR/LPCI:
Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:
Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool
Cooling/Cleanup 234000 (SF8 FH) Fuel Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam X 5.0 Components 291002 Sensors and Detectors K1.13 Pressure-Modes of Failures 3.1 57 239003 (SF9 MSIVLC) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary X
A1.07 Ability to predict and/or monitor changes in parameters associated with operation of the Main Turbine Generator and Auxiliary Systems, including:
First stage turbine pressure (CFR: 41.5 / 45.5) 3.3 58 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater X
K6.02 Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Feedwater System:
Condensate system (CFR: 41.7 / 45.7) 3.5 59 268000 (SF9 RW) Radwaste X
K2.01 Knowledge of electrical power supplies to the following:
Radiological release isolation valves (CFR: 41.7) 2.7 60 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring X
K3.05 Knowledge of the effect that a loss or malfunction of the Radiation Monitoring System will have on the following systems or system parameters:
Offgas system (CFR: 41.5 / 45.3) 3.6 61 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation X
A2.02 Ability to (a) predict the impacts of the following on 3.5 62
the Control Room Ventilation and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Extreme environmental conditions (fire, toxic gas, smoke, radiation, etc.)
(CFR: 41.5 / 43.5 / 45.6) 290002 (SF4 RVI) Reactor Vessel Internals 510001 (SF8 CWS*) Circulating Water X
K3.01 Knowledge of the effect that a loss or malfunction of the Circulating Water System will have on the following systems or system parameters:
Main turbine generator and auxiliary systems (CFR: 41.7 / 45.4) 3.5 63 K/A Category Point Totals:
0 1
3 1
0 1
2 1
0 1
1 Group Point Total:
11
Form 4.1-COMMON RO Common Examination Outline Facility: Cooper Nuclear Station Date of Exam: June 2022 Generic Knowledge and AbilitiesTier 3 (RO)
Category K/A #
Topic RO SRO-Only IR IR
- 1.
Conduct of Operations 2.1.
2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10 CFR Part 55 (CFR: 41.10 / 43.2) 3.3 64 2.1.
2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, or switches (CFR: 41.10 / 45.1 / 45.12) 4.1 65 Subtotal 2
N/A
- 2.
Equipment Control 2.2.
2.2.13 Knowledge of tagging and clearance procedures (CFR: 41.10 / 43.1 / 45.13) 4.1 66 2.2.
2.2.6 Knowledge of the process for making changes to procedures (CFR: 41.10 / 43.3 / 45.13) 3.0 67 Subtotal 2
N/A
- 3.
Radiation Control 2.3.
2.3.12 Knowledge of radiological safety principles and procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, or alignment of filters (CFR: 41.12 / 43.4 / 45.9 / 45.10) 3.2 68 Subtotal 1
N/A
- 4.
Emergency Procedures/
Plan 2.4.
2.4.34 Knowledge of RO responsibilities outside the main control room during an emergency (CFR: 41.10 / 43.5 / 45.13) 4.2 69 Subtotal 1
N/A Tier 3 Point Total 6
TheoryTier 4 (RO)
Category K/A #
Topic RO IR Reactor Theory 6.1 292008 Reactor Operational Physics K1.01 List parameters that should be monitored and controlled during the approach to criticality 3.9 70 6.1 292001 Neutrons K1.02 Define prompt and delayed neutrons 3.1 71 6.1 292005 Control Rods K1.07 Define control rod worth, differential CRW, and integral control rod worth 2.6 72 Subtotal N/A
Thermodynamics 6.2 293007 Heat Transfer K1.01 Describe three mechanisms of heat transfer 3.2 73 6.2 293009 Core Thermal Limits K1.10 Define APLHGR 3.7 74 6.2 293010 Brittle Fracture and Vessel Thermal Stress K1.04 State how the possibility of brittle fracture is minimized by operating limitations 3.2 75 Subtotal N/A Tier 4 Point Total 6
Form 4.1-BWR SRO Boiling-Water Reactor Examination Outline Notes: CO = Conduct of Operations; EC = Equipment Control; RC = Radiation Control; EM = Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan Facility: Cooper Nuclear Station Date of Exam: June 2022 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Total A2 G
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
N/A N/A 20 4
3 7
2 6
2 1
3 Tier Totals 26 6
4 10
- 2.
Plant Systems 1
26 3
2 5
2 11 2
1 3
Tier Totals 37 5
3 8
- 3.
Generic Knowledge and Abilities Categories CO EC RC EM 6
CO EC RC EM 7
2 2
1 1
2 2
1 2
- 4. Theory Reactor Theory Thermodynamics 6
3 3
Form 4.1-BWR BWR Examination Outline Page 2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)
E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)
IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation 295003 (APE 3) Partial or Complete Loss of AC Power 295004 (APE 4) Partial or Total Loss of DC Power 295005 (APE 5) Main Turbine Generator Trip 295006 (APE 6) Scram 295016 (APE 16) Control Room Abandonment
/ 7 X
AA2.04 Ability to determine and/or interpret the following as they apply to Control Room Abandonment:
Suppression Pool Temperature (CFR: 41.10 / 43.5 / 45.13) 3.8 76 295018 (APE 18) Partial or Complete Loss of CCW 295019 (APE 19) Partial or Complete Loss of Instrument Air 295021 (APE 21) Loss of Shutdown Cooling 295023 (APE 23) Refueling Accidents X
AA2.05 Ability to determine and/or interpret the following as they apply to Refueling Accidents:
Emergency plan implementation (CFR: 41.10 / 43.5 / 45.13) 4.4 77 295024 (EPE 1) High Drywell Pressure 295025 (EPE 2) High Reactor Pressure 295026 (EPE 3) Suppression Pool High Water Temperature X
EA2.01 Ability to determine and/or interpret the following as they apply to Suppression Pool High Water Temperature:
Suppression pool water temperature (CFR: 41.10 / 43.5 / 45.13) 4.0 78 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only)
X EA2.01 Ability to determine and/or interpret the following as they apply to High Drywell Temperature:
Drywell temperature (CFR: 41.10 / 43.5 / 45.13) 4.2 79
295030 (EPE 7) Low Suppression Pool Water Level X 2.2.25 Knowledge of the bases in technical specifications for limiting conditions for operation and safety limits (SRO Only)
(CFR: 43.2) 4.2 80 295031 (EPE 8) Reactor Low Water Level 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown 295038 (EPE 15) High Offsite Radioactivity Release Rate X 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator (CFR: 41.10 / 43.5 / 45.11) 4.1 81 600000 (APE 24) Plant Fire On Site 700000 (APE 25) Generator Voltage and Electric Grid Disturbances X 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator (CFR: 41.10 / 43.5 / 45.13) 3.8 82 K/A Category Totals:
4 3 Group Point Total:
7
Form 4.1-BWR BWR Examination Outline Page 3 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)
E/APE # / Name K1 K2 K3 A1 A2 G K/A Topic(s)
IR 295002 (APE 2) Loss of Main Condenser Vacuum 295007 (APE 7) High Reactor Pressure X
Ability to determine and/or interpret the following as they apply to High Reactor Pressure: AA2.04 Bypass valve capacity (CFR: 41.10 / 43.5 / 45.13) 4.0 83 295008 (APE 8) High Reactor Water Level 295009 (APE 9) Low Reactor Water Level X
AA2.02 Ability to determine and/or interpret the following as they apply to Low Reactor Water Level:
Steam flow/feed flow mismatch (CFR: 41.10 / 43.5 / 45.13) 3.8 84 295010 (APE 10) High Drywell Pressure 295011 (APE 11) High Containment Temperature (Mark III Containment only) 295012 (APE 12) High Drywell Temperature 295013 (APE 13) High Suppression Pool Water Temperature/ 5 295014 (APE 14) Inadvertent Reactivity Addition X 2.2.38 Knowledge of conditions and limitations in the facility license (CFR: 41.7 / 41.10 / 43.1 /
45.13) 4.5 85 295015 (APE 15**) Incomplete Scram 295017 (APE 17) High Offsite Release Rate 295020 (APE 20) Inadvertent Containment Isolation 295022 (APE 22) Loss of Control Rod Drive Pumps 295029 (EPE 6) High Suppression Pool Water Level 295032 (EPE 9) High Secondary Containment Area Temperature 295033 (EPE 10) High Secondary Containment Area Radiation Levels 295034 (EPE 11) Secondary Containment Ventilation High Radiation
/ 9 295035 (EPE 12) Secondary
Containment High Differential Pressure 295036 (EPE 13) Secondary Containment High Sump/Area Water Level 500000 (EPE 16) High Containment Hydrogen Concentration K/A Category Point Totals:
2 1 Group Point Total:
3
Form 4.1-BWR BWR Examination Outline Page 4 Plant SystemsTier 2/Group 1 (SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)
IR 203000 (SF2, SF4 RHR/LPCI)
RHR/LPCI: Injection Mode 205000 (SF4 SCS) Shutdown Cooling 206000 (SF2, SF4 HPCI)
High-Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)
Low-Pressure Core Spray X
A2.08 Ability to (a) predict the impacts of the following on the Low-Pressure Core Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Valve openings due to malfunction(s)
(CFR: 41.5 / 43.5 / 45.6) 3.5 86 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 X 2.2.22 Knowledge of limiting conditions for operation and safety limits (CFR: 41.5 / 43.2 / 45.2) 4.7 88 215003 (SF7 IRM)
Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization X 2.2.12 Knowledge of surveillance procedures (CFR: 41.10 / 43.2 / 45.13) 4.1 89 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff 239002 (SF3 SRV) Safety Relief Valves
259002 (SF2 RWLCS) Reactor Water Level Control 261000 (SF9 SGTS) Standby Gas Treatment X
A2.03 Ability to (a) predict the impacts of the following on the Standby Gas Treatment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: High train temperature (CFR: 41.5 / 43.5 /
45.6) 3.3 90 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) 300000 (SF8 IA) Instrument Air 400000 (SF8 CCW) Component Cooling Water 510000 (SF4 SWS*) Service Water K/A Category Point Totals:
3 2 Group Point Total:
5
Form 4.1-BWR BWR Examination Outline Page 5 Plant SystemsTier 2/Group 2 (SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic(s)
IR #
201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor X 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management (CFR: 41.1 / 41.5 / 41.10 /
43.6 / 45.6) 4.6 91 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:
Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries X
A2.09 Ability to (a) predict the impacts of the following on the Primary Containment System and Auxiliaries and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Vacuum breaker malfunction (CFR: 41.5 / 45.6) 3.7 92 226001 (SF5 RHR CSS) RHR/LPCI:
Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:
Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup X
A2.01 Ability to (a) predict the impacts of the following on the Fuel Pool Cooling and Cleanup and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Abnormal fuel pool level (CFR: 41.5 / 43.5 / 45.6) 3.8 93 234000 (SF8 FH) Fuel Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSIVLC) Main Steam Isolation
Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals 510001 (SF8 CWS*) Circulating Water K/A Category Point Totals:
2 1
Group Point Total:
3
Form 4.1-COMMON SRO Common Examination Outline Facility: Cooper Nuclear Station Date of Exam: June 2022 Generic Knowledge and AbilitiesTier 3 (SRO)
Category K/A #
Topic RO SRO-Only IR IR
- 1.
Conduct of Operations 2.1.
2.1.35 Knowledge of the fuel handling responsibilities of SROs (CFR: 43.7) 3.9 94 2.1.
2.1.15 Knowledge of administrative requirements for temporary management direction, such as standing orders, night orders, or operations memoranda (CFR: 41.10 / 45.12) 3.4 95 Subtotal N/A 2
- 2.
Equipment Control 2.2.
2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator (CFR: 41.10 / 43.5 / 45.13) 3.8 96 2.2.
2.2.43 Knowledge of the process used to track inoperable alarms (CFR: 41.10 / 43.5 / 45.13) 3.3 97 Subtotal N/A 2
- 3.
Radiation Control 2.3.
2.3.6 Ability to approve liquid or gaseous release permits (CFR: 41.13 / 43.4 / 45.10) 3.8 98 Subtotal N/A 1
- 4.
Emergency Procedures/
Plan 2.4.
2.4.29 Knowledge of the emergency plan implementing procedures (CFR: 41.10 / 43.5 / 45.11) 4.4 99 2.4.
2.4.52 Knowledge of the lines of authority during implementation of the emergency plan, emergency plan implementing procedures, emergency operating procedures, or severe accident guidelines (CFR: 41.10 / 45.13) 4.0 100 Subtotal N/A 2
Tier 3 Point Total 7
TheoryTier 4 (RO)
Category K/A #
Topic RO IR Reactor Theory 6
6 6
Subtotal N/A Thermodynamics 6
6 6
Subtotal N/A Tier 4 Point Total 6
Form 4.1-1 Record of Rejected Knowledge and Abilities Tier/Group Randomly Selected K/A Reason for Rejection RO T2/G2 239001 Main and Reheat Steam Systems, 6.2, 293005, K1.05 (IR -2.8) 5.0 Components 239001 Main and Reheat Steam Systems, 291002 Sensors and Detectors K1.13 Pressure-Modes of Failures (IR-3.1)
Original question was from section 6 THEORY of 1123 Rev 3 and the new question is from Section 5 COMPONENTS of NUREG 1123 Rev 3. Change is IAW with NUREG 1021 REV
- 12. This Change was discussed with the CE.
SRO T1/G1 295016 (APE
- 16) Control Room Abandonment
/ 7 AA2.07 Ability to determine and/or interpret the following as they apply to Control Room Abandonment:
Suppression chamber pressure IR 3.4 295016 (APE 16)
Control Room Abandonment
/ 7 AA2.04 Ability to determine and/or interpret the following as they apply to Control Room Abandonment:
Suppression Pool temperature IR 3.8 Because CNS procedures do not contain instructions for observing or controlling Suppression Chamber Pressure during control room abandonment, 295016 AA2.07 (Suppression chamber pressure) was replaced with 295016 AA2.04 (Suppression Pool temperature). Page 1 point totals not affected by this change. (Rev 1).
500000 (EPE
- 16) High Containment Hydrogen Concentration EK2.05 Knowledge of the relationship between High Containment Hydrogen Concentration and the following systems or components:
Hydrogen and oxygen recombiners (CFR: 41.7 /
45.8)
IR 3.5 500000 (EPE 16)
High Containment Hydrogen Concentration EK2.01 Knowledge of the relationship between High Containment Hydrogen Concentration and the following systems or components:
Containment hydrogen monitors (CFR: 41.7 / 45.8)
IR 3.8 K/A was replaced since CNS does not have a hydrogen and oxygen recombiner in our PC. Discussed with CE and replaced with random K/A. Page 1 point totals not affected by this change. (Rev 2). K/A stayed as 500000 (EPE 16) High Containment Hydrogen Concentration but changed to EK2.01 (Containment hydrogen monitors) from EK2.05 (Hydrogen and oxygen recombiners)
ES-3.2, Page 11 of 18 Form 3.2-1 Administrative Topics Outline Facility: Cooper Nuclear Station Date of Examination: 6/13/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-06 Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
Conduct of Operations A1, Perform a DG Fuel Oil Availability K/A 2.1.23 (4.3)
(R) (D)
Conduct of Operations A2, Determine Peer checking requirements for Control Room equipment (Multiple)
K/A 2.1.1 (3.8)
(R) (N)
Equipment Control A3, Determine Proper Sequence for Hanging a Clearance Order (MODE 5 CRD-P-B)
K/A 2.2.13 (4.1)
(R) (N)
Radiation Control N/A N/A Emergency Plan A4, Identify Available and Reliable Instrumentation During a Fire K/A 2.4.3 (3.7) 2.4.25 (3.3)
(R) (N)
ES-3.2, Page 12 of 18 Facility: Cooper Nuclear Station Date of Examination: 6/13/2022 Examination Level: RO SRO Operating Test Number: CNS-2022-06 Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
Conduct of Operations A5, Determination of Work Hour Limitations for Watch Standing K/A G2.1.5 (3.9)
(R) (N)
Conduct of Operations A6, Perform SRO duties for protecting FPC following a Refueling Outage K/A G2.1.25 (4.2)
(R) (N)
Equipment Control A7, Determine LCO and Required Actions (Tech Spec 3.3.2.1 (RWM))
K/A G2.2.35 (4.5)
(R) (N)
Radiation Control A8, Determine Emergency Exposure Requirements Used on the 2020-04 exam.
K/A G2.3.14 (3.8) (SRO ONLY)
(R) (D)
Emergency Plan A9, Determine Emergency Classifications EAL (SA6.1)
K/A G2.4.41 (4.6) (SRO ONLY)
(R) (M)
ES-3.2, Page 13 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 3.2-2 Control Room/In-Plant Systems Outline Facility: Cooper Nuclear Station Date of Examination: 6/14/2022 Operating Test Number: CNS-2022-06 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. Perform Rapid Power Reduction (Alternate Path)
NRC K/A 202002 A2.05 (3.3/3.7)
RO, SROI, SROU.
M, S, A 1
S2. Perform RFPT thrust bearing wear and failure alarm test (Alternate Path-High Vibrations)
NRC K/A 259001 A3.10 (3.9)
RO, SROI.
N, S, A 2
S3. Lowering DEH Pressure Set point in TARGET Mode (developed for 2021-06 Retake exam, entered into ADAMS, but test never administered to examinees)
NRC K/A 241000 A4.06 (4.2)
RO, SROI.
P, S, L 3
S4. Adjust Generator Voltage Regulator (Auto &
Manual) (Alternate Path)
NRC K/A: 245000 A4.12 (3.3)
RO, SROI.
N, S, A 4
S5. Vent the PC for early venting per 5.8.18 NRC K/A: 223001 K4.08 (3.8), 223001 A4.17 (3.8)
RO.
D, S, L 5
S6. Manually Bypass RWM (Restoration)
NRC K/A 201006 A4.01 (3.4)
RO, SROI, SROU.
N, S 7
S7. Respond to Loss of Condenser Vacuum due to Ice buildup NRC K/A 295002 AA1.07 (3.4), 510001 A4.01 (3.3)
RO, SROI.
N, S, 8
S8. Restoration of a Group 6 Primary Containment Isolation (Alternate Path)
NRC K/A: 288000 A4.01 (3.2)
RO, SROI, SROU.
N, S, EN, A 9
In-Plant Systems P1. Conduct Alternate Pressure Control (Failure-to-Scram) using Steam Jet Air Ejectors (Not used from 2012 to present)
(developed for 2021-06 Retake exam, entered into ADAMS, but test never administered to examinees)
NRC K/A 295025 EK1.07 (4.2), EK2.12 (3.1), 239001 K1.07 (3.2)
RO, SROI, SROU.
P, R, L, E 3
P2. Respond To No Break Power Panel Failure (Alternate Path)
NRC K/A 262002 K4.01 (3.5)
Last Used on 2012 ILT Class NRC exam RO, SROI, SROU.
D, E, A 6
P3. Backwash a Circulating Water Pump NRC K/A 295002 AK2.08 (3.5), 510001 A1.06 (3.9)
RO, SROI.
D, E 8
Form 3.2-2 Instructions for Control Room/In-Plant Systems Outline
- 1. Determine the number of control room system and in-plant system job performance measures (JPMs) to develop using the following table:
- 2. Select safety functions and systems for each JPM as follows:
Refer to Section 1.9 of the applicable knowledge and abilities (K/A) catalog for the plant systems organized by safety function. For pressurized-water reactor operating tests, the primary and secondary systems listed under Safety Function 4, Heat Removal from Reactor Core, in Section 1.9 of the applicable K/A catalog, may be treated as separate safety functions (i.e., two systems, one primary and one secondary, may be selected from Safety Function 4). From the safety function groupings identified in the K/A catalog, select the appropriate number of plant systems by safety functions to be evaluated based on the applicants license level (see the table in step 1).
For RO/SRO-I applicants: Each of the control room system JPMs and, separately, each of the in-plant system JPMs must evaluate a different safety function, and the same system or evolution cannot be used to evaluate more than one safety function in each location. One of the control room system JPMs must be an engineered safety feature.
For SRO-U applicants: Evaluate SRO-U applicants on five different safety functions.
One of the control room system JPMs must be an engineered safety feature, and the same system or evolution cannot be used to evaluate more than one safety function.
- 3. Select a task for each JPM that supports, either directly or indirectly and in a meaningful way, the successful fulfillment of the associated safety function. Select the task from the applicable K/A catalog (K/As for plant systems or emergency and abnormal plant evolutions) or the facility licensees site-specific task list. If this task has an associated K/A, the K/A should have an importance rating of at least 2.5 in the RO column. K/As that have importance ratings of less than 2.5 may be used if justified based on plant priorities; inform the NRC chief examiner if selecting K/As with an importance rating less than 2.5.
The selected tasks must be different from the events and evolutions conducted during the simulator operating test and tasks tested on the written examination. A task that is similar to a simulator scenario event may be acceptable if the actions required to complete the task are significantly different from those required in response to the scenario event.
Apply the following specific task selection criteria:
License Level Control Room In-Plant Total Reactor Operator (RO) 8 3
11 Senior Reactor Operator-Instant (SRO-I) 7 3
10 Senior Reactor Operator-Upgrade (SRO-U) 2 or 3 3 or 2 5
Form 3.2-2 Instructions for Control Room/In-Plant Systems Outline (continued)
At least one of the tasks shall be related to a shutdown or low-power condition.
Four to six of the tasks for RO and SRO-I applicants shall require execution of alternative paths within the facility licensees operating procedures. Two to three of the tasks for SRO-U applicants shall require execution of alternative paths within the facility licensees operating procedures.
At least one alternate path JPM must be new or modified from the bank.
At least one of the tasks conducted in the plant shall evaluate the applicants ability to implement actions required during an emergency or abnormal condition.
At least one of the tasks conducted in the plant shall require the applicant to enter the radiologically controlled area. This provides an excellent opportunity for the applicant to discuss or demonstrate radiation control administrative subjects.
If it is not possible to develop or locate a suitable task for a selected system, return to step 2 and select a different system.
- 4. For each JPM, specify the codes for type, source, and location: (ACTUAL)
Code License Level Criteria RO SRO-I SRO-U (A)lternate path 4-6 (5) 4-6 (5) 2-3 (3)
(C)ontrol room (D)irect from bank
< 9 (3)
< 8 (2)
< 4 (1)
(E)mergency or abnormal in-plant
> 1 (3)
> 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 (6)
> 2 (6)
> 1 (3)
(P)revious two exams (randomly selected)
< 3 (2)
< 3 (2)
< 2 (1)
(R)adiologically controlled area
> 1 (1) > 1 (1) > 1 (1)
(S)imulator
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 1 of 48 Table of Contents NUREG 1021 FORM 3.3-1................................................................................................................................................. 2 Scenario Summary............................................................................................................................................................. 15 Event 1 - RR Power Reduction.............................................................................................................................................. 17 Event 2 - CRD ACC 26-27 Low Pressure................................................................................................................................ 18 Event 3 - DEH Pump B High Filter D/P................................................................................................................................. 20 Event 4 - B RR Vigh Vibrations............................................................................................................................................ 21 Event 5 - RR Piping Leak into PC........................................................................................................................................... 25 Event 6 - Electrical ATWS...................................................................................................................................................... 29 Event 7 - Fuel Failure............................................................................................................................................................ 39 Scenario Record................................................................................................................................................................. 47 Shift Turnover..................................................................................................................................................................... 48 Revision Summary Rev.
Description Author 00 New Exam scenario.
C. Edgington 01 Comments from Chief Examiner on the Draft Outline submittal.
C. Edgington 02 Comments from Chief Examiner on the Draft Operating Test submittal.
C. Edgington 03 Comments from Chief Examiner during the NRC Validation Week C. Edgington
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 2 of 48 NUREG 1021 FORM 3.3-1 A.
Scenario Outline Facility: Cooper Nuclear Station Scenario No.: 1__
Scenario Source: IC 19 Operating-Test No.: CNS-2022-06 Examiners: ____________________________ Applicants/Operators:
Turnover:
The plant is at 100% power MOL.
1.
TLCO 3.3.5 Condition A for one LEFM instrument INOPERABLE for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> has been entered 2.
Lower Reactor power to < 2381 MWth using Reactor Recirculation to comply with TLCO 3.3.5 Condition B.
Critical Tasks:
Identified in the Event Description on Events 6 and 9 Event No.
Malf.
No.
Event Type*
Event Description 1
Lower RX Power ATC Lowers power per 2.1.10 using RR to 2381MWth BOP provides peer checking for lowering RR Speed CRS provides the reactivity manager during the change in power.
2 RD11 TS (CRS) Low Accumulator Pressure 26-27 ATC addresses alarm card 9-5-2/G-6 CRS enters TS 3.1.5 Condition A.
DEH Pump High filter D/P alarm BOP Shifts DEH pumps per the alarm card B-1/E-7 4
TS (CRS)
B RR Pump rising vibrations ATC will take Alarm card actions for 9-4-3/C-7 which direct tripping B RR Pump (2 dangers)
BOP will take actions per 2.4RR for a tripped RR pump CRS will determine that LCO 3.4.1 is not met and enters Condition B.
5 RR20B 0 to 6 (Ramp over 6 minutes)
RR Piping leak will cause Drywell pressure to rise BOP will Vent PC per 2.4PC CRS will direct entry into 2.4PC and EOP-3A. (EOP-3A entry)
CRS will direct a reactor scram ATC will scram the reactor 6
RD26 RD27 RP01A RP01B MC (BOP)
ATWS conditions ATC reports a failure to scram CRS enters EOP-1A transitions to EOP-6A and EOP-7A ATC and BOP take EOP-7A Table 41 actions.
Spare Scenario
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 3 of 48 RP01C RP01D RP11A RP11B RP11C RP11D RP11E RP11F RP11G RP11H o ATC will inject SLC and lower RX water level to -60 inches WR o BOP will install Group 1 low level jumpers, inhibit ADS and insert rods per 5.8.3.
o BOP will insert all rods by taking manual action to trip RPS via the TEST trip switches IAW 5.8.3.
CT#1 (BOP)
Given a condition where an ATWS has occurred, the crew inhibits ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cool down rate limit during a failure to scram.
Given a condition where an ATWS has occurred and reactor power is above 3%, the crew stops and prevents injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -60 (or LL, as applicable), prior to neutronic oscillations exceeding 25% peak-to-peak indicated on any APRM.
CT#3 (BOP)
Given a condition where an ATWS has occurred and reactor power is above 3%, the crew inserts all control rods to at least position 02 by performing procedure 5.8.3 Attachment 1 ALTERNATE ROD INSERTION METHODS FLOWCHART Legs B and C prior to the FIRST of either:
A) Transitioning to Leg D or F of procedure 5.8.3 (longer actions of manually venting the scram air header or CRDMs); or B) Exceeding the Boron Injection Initiation Temperature (BIIT) curve.
Fuel failure will occur from a broken piece of the B RR pump CRS will direct action per 5.1RAD, 2.4OG, and 5.2FUEL. (EOP-5A entry)
BOP will take actions per 5.1RAD, 2.4OG, and 5.2FUEL.
ATC will perform Attachment 2 of 5.2FUEL 8
RP15 MC (ATC)
Group 6 failure CRS will direct ATC to verify group isolations ATC will determine that a failure of Group 6 exists o ATC will insert a Group 6 isolation per 2.1.22.
CT#4 (ATC)
Given a condition where a failure of an Auto Group 6 isolation has occurred while the drywell is lined up for venting, and all three fission product boundaries have been lost, the crew will insert a Manual Group 6 isolation IAW procedure 2.1.22 to minimize the release to the public prior to Drywell Radiation Monitor (RMA-RM-40A/B) 3.6E+3 REM/Hr.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 4 of 48
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 5 of 48 B.
NUREG 1021 Rev 12 Criteria for Evaluation Scenarios, Table 3.4-1 and Table 3.4-2 Quantitative Attributes Table Attribute ES-3.4-1 Target Actual Description Events after EOP entry 1-2 2
- 1. Fuel Failure
- 2. Failure of Group 6 Isolation Abnormal events 2-4 3
- 1. DEH pump high filter D/P(B-1/E-7)
- 2. Trip of RR pump (2.4RR)
- 3. RR loop leak (2.4PC)
Major transients 1-2 1
- 1. EOP-3A
- 2. EOP-5A
- 3. EOP-6A
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 6 of 48 Pre-identified Critical Tasks 2
3
- 1. (CT#1) Given a condition where an ATWS has occurred, the crew inhibits ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to Scram.
- 2. (CT#2) Given a condition where an ATWS has occurred and reactor power is above 3%, the crew stops and prevents injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -
60 (or LL, as applicable), prior to neutronic oscillations exceeding 25% peak-to-peak indicated on any APRM.
- 3.
(CT#3) Given a condition where an ATWS has occurred and reactor power is above 3%, the crew inserts all control rods to at least position 02 by performing procedure 5.8.3 ALTERNATE ROD INSERTION METHODS FLOWCHART Legs B and C prior to the FIRST of either:
A) Transitioning to Leg D or F of procedure 5.8.3 (longer actions of manually venting the scram air header or CRDMs); or B) Exceeding the Boron Injection Initiation Temperature (BIIT) curve.
- 4.
(CT#4) Given a condition where a failure of an Auto Group 6 isolation has occurred while the drywell is lined up for venting, and all three fission product boundaries have been lost, the crew will insert a Manual Group 6 isolation IAW procedure 2.1.22 to minimize the release to the public prior to Drywell Radiation Monitor (RMA-RM-40A/B) 3.6E+3 REM/Hr.
Reactivity Manipulation N/A 1
- 1. Lower power 2419 MWth to 2381 MWth Manual Control of Automatic Function 1
2
- 1. Failure of AUTO Group 6 isolation (ATC)
- 2. Failure of Automatic Scram (BOP)
Normal Evolution N/A 0
Instrument/
Component (I/C)
Failure N/A 5
- 4. Failure of AUTO Group 6 isolation (ATC)
TS Evaluation 2
2
- 1. LCO 3.4.1 Condition B (Core Flow Loop Mismatch)
- 2. LCO 3.1.5 Condition A (CRD 26-27 Accumulator INOP)
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 7 of 48 Total Malfunctions N/A 6
- 1. Low CRD 26-27 Accumulator pressure
- 2. DEH Pump filter high D/P
- 3. B RR pump rising vibes
- 4. RR discharge piping leak
- 5. Failure of RPS/ARI
- 6. Failure of Group 6 isolation
Op-Test No.: CNS 2022-06 Scenario No.:
1 Page 8 of 48 C.
NUREG 1021 Rev 12 Criteria for Evaluation Scenarios, Table 3.4-1 and Table 3.4-2 Critical Task #1 Given a condition where an ATWS has occurred, the crew inhibits ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cool down rate limit during a failure to Scram.
Safety Significance In order to effect a reduction in reactor power, actions in Contingency #5 may deliberately lower RPV water level to a level below the automatic initiation setpoint of ADS. Actuation of this system imposes a severe thermal transient on the RPV and complicates the efforts to maintain RPV water level within the ranges specified in Contingency #5. Further, rapid and uncontrolled injection of large amounts of relatively cold, unborated water from low pressure injection systems may occur as RPV pressure decreases to and below the shutoff heads of these pumps. Such an occurrence would quickly dilute in-core boron concentration and reduce reactor coolant temperature. When the reactor is not shutdown, or when the shutdown margin is small, sufficient positive reactivity might be added in this way to cause a reactor power excursion large enough to severely damage the core. Therefore, ADS initiation is purposely prevented as the first action of the level/power control procedure. When required, explicit direction to depressurize the RPV is provided in the PSTG, thereby negating any requirement to maintain the automatic initiation capability of ADS.
Cues ADS Timers Actuated alarm 9-3-1/A-1.
Wide Range and Fuel Zone/CFZ RPV level indications approaching or exceeding Level 1 (-113).
ADS valve control switch red and amber indicating lights on Panel 9-3 ON.
Measurable Performance Indicators Manipulation of ADS A and ADS B Inhibit switches on Panel 9-3 vertical section prior to Reactor Pressure lowering to 425 psig.
This accounts for the highest starting pressure of getting reactor pressure up to 1070 psig.
554ºF - 99ºF = 455ºF.
455ºF has a pressure of 426 psig - 429 psig.
Performance Feedback Inhibit switches click into the vertical, inhibit position on Panel 9-3 prior to breaking the tech spec required cooldown rate.
9-3-1/D-1 ADS INHIBITED alarm comes in Applicability ATWS with power >3% following mitigating Scram Actions of procedure 2.1.5 REACTOR SCRAM.
Justification for the chosen performance limit Inhibiting ADS before injection from high volume, cold water systems occur ensures a related power excursion will not be experienced that could challenge to the fuel barrier. Inhibiting ADS before the Tech Spec cooldown limit is exceeded ensures the RPV fission product barrier is not challenged by a significant thermal transient.
BWR Owners Group Appendix App. B, Step RC/Q-6
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1 Page 9 of 48 Scenario Guide Requirements The scenario must be designed to make the crew lower RPV level per EOP-7A Table 41 actions (i.e. ATWS with power >3% and Feedwater or HPCI maintaining level above -60 WR).
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1 Page 10 of 48 Critical Task # 2 Given a condition where an ATWS has occurred and reactor power is above 3%, the crew stops and prevents injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -60 (or LL, as applicable), prior to neutronic oscillations exceeding 25% peak-to-peak indicated on any APRM.
Safety Significance Lowering level below -60 WR, sufficiently below the elevation of the feedwater sparger nozzles prevents or mitigates the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities.
This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.
24" below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that the capability to bypass the low RPV water level MSIV isolation should be able to control RPV water level with feedwater pumps to preclude the isolation.
Cues Manual scram is initiated and RPS fails to de-energize and reactor power remains >3% on Panel 9-5 indications and SPDS and RPV level is > -60 WR on SPDS.
Measurable Performance Indicators Operator manipulates Feedwater HMIs on Panel 9-5 or Panel A as necessary to stop FW injection until RPV level goes below -60 WR.
Operator manipulates HPCI controls on panel 9-3 to stop HPCI injection until RPV level is below -60 WR.
Performance Feedback Feedwater flow indication on panel 9-5 indicate zero.
HPCI flow indication on panel 9-3 indicates zero and/or HPCI injection MOV indicates closed.
Applicability EOP-7A conditions where power remains above 3% following completion of migrating tasks of procedure 2.1.5 REACTOR SCRAM. If, due to scenario design and dynamics, level lowers to below -60 WR without crew action, this should not be selected as a critical task.
Justification for the chosen performance limit Applicability for this CT is during EOP-7A conditions where it is necessary to lower reactor water level to - 60 WR before 25% peak-to-peak neutron flux oscillations potentially could occur with localized fuel power peaking. This value was chosen because it establishes margin to conditions where fuel damaging power oscillations may theoretically occur per the PSTGs.
BWR Owners Group Appendix App. B, Contingency #5
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1 Page 11 of 48 Scenario Guide Requirements Initial conditions, ATWS with greater than 3% following 2.1.5 mitigating actions for a rector scram. The scenario should be designed such that level will remain above -60 WR, apart from crew action.
Critical Task #3 Given a condition where an ATWS has occurred and reactor power is above 3%, the crew inserts all control rods to at least position 02 by performing procedure 5.8.3 Attachment 1 ALTERNATE ROD INSERTION METHODS FLOWCHART Legs B and C prior to the FIRST of either:
A) Transitioning to Leg D or F of procedure 5.8.3 (longer actions of manually venting the scram air header or CRDMs); or B) exceeding the Boron Injection Initiation Temperature (BIIT) curve.
Safety Significance RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits to preserve the integrity of the fuel cladding and the reactor coolant pressure boundary (RCPB) and minimize the energy that must be absorbed following a loss of coolant accident (LOCA). Failure to effect shutdown of the reactor when a RPS setting has been exceeded, even at low power, would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. Action to shut down the reactor is required when RPS and ARI systems fail. With the injection of SLC, the makeup of rod position is in an unanalyzed area and the longer time that this condition occurs means a greater chance of fuel damage due to localized power peaking.
The postulated DBA against which the primary containment performance is evaluated is the entire spectrum of postulated pipe breaks within the primary containment. Inputs to the safety analyses include initial suppression pool water volume and suppression pool temperature. An initial pool temperature of 95°F is assumed for the safety analyses. Reactor shutdown at a pool temperature of 110°F and vessel depressurization at a pool temperature of 120°F are assumed for the safety analyses. The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut down. The design temperature for this condition is BIIT.
Cues Annunciators 9-5-2/A-1 (A-2) RX SCRAM CHANNEL A (B) in alarm with RPS remaining energized.
Completion of 2.1.5 Reactor Scram mitigating actions with reactor power at 100%.
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1 Page 12 of 48 Measurable Performance Indicators Operator takes the RPS TEST TRIP switches to TRIP per 5.8.3.
Performance Feedback RPS Group lights de-energized on panel 9-5.
Control Rod full -in indication on panel 9-5.
Reactor power trend on nuclear instrumentation on panel 9-5.
Applicability Any time a parameter exceeds a scram setting and RPS fails to trip and reactor power is greater than 30% and SLC is being injected into the reactor vessel.
The conditional CT for if the main turbine is tripped prior to performing 5.8.3 actions energy is discharging into primary containment via SRVs or a primary system leak.
This task is only critical if a manual scram and ARI actuation does not cause fully inserting control rods, and performing C leg of the 5.8.3 flowchart would be successful in fully inserting control rods.
Justification for the chosen performance limit With the initial conditions, by having a 100% ATWS with SLC injecting causes the reactor to enter an unanalyzed area until reactor power is lowered to 3%. If the operator does not follow the procedure properly he causes a longer time to take actions in the field.
This is conditional if the crew decides to take action to trip the Main Turbine prior to addressing the ATWS. If this action is taken then heat would be added to containment and the limiting time would be prior to the Suppression Pool exceeding BIIT, which would indicate if an ED was required that there is a chance that the ED would exceed HCTL. This causes a potential loss of Primary containment.
BWR Owners Group Appendix App. B, Part 2 (EPGs-HOT)
Scenario Guide Requirements Electrical Failure to Scram event with ARI failure.
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1 Page 13 of 48
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1 Page 14 of 48 Critical Task # 4 Given a condition where a failure of an Auto Group 6 isolation has occurred while the drywell is lined up for venting, and all three fission product boundaries have been lost, the crew will insert a Manual Group 6 isolation IAW procedure 2.1.22 to minimize the release to the public prior to Drywell Radiation Monitor (RMA-RM-40A/B) 3.6E+3 REM/Hr.
Safety Significance If the crew does not take action to isolate the PC from venting to the environment, escalation from an ALERT initially to a SAE, if PC venting not completed prior Drywell Rad monitors rising to a value of 3.6E+3 REM/Hr. The EAL would elevate from a SAE to a GE. Evacuation of the public would be required. This is very safety significant. Failing to isolate PC venting would result in an unnecessary offsite release and endanger plant personnel.
Cues Indication of rising or Maximum Operating values of Radiation levels in an area of a system which is connected to the RCS.
Rising Drywell Radiation Levels on PMIS Drywell Particulate High Activity Alarm (Q-1/A-1)
Measurable Performance Indicators Will stop the PC venting and prevent the escalation of the EAL by ensuring that (PC-AO-246 or PC-MO-306) and (PC-AO-245 or PC-MO-305) are closed.
Performance Feedback PC Ventilation valves close Applicability A fuel failure has occurred and a failure of Group 6 occurs while PC venting is in progress.
Justification for the chosen performance limit With the initial conditions, the upgrade to a SAE would be within 15 minutes but it is possible that with the on-going conditions, the identification of the EAL since it is normally the SM that evaluates EALs, but the more severe EAL of upgrading to a GE would take ~ 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to get to and the safety significant of the release to the public should be identified and within the given time frame should be able to be met.
BWR Owners Group Appendix App. B, Part 2 (EPGs-HOT)
Scenario Guide Requirements A failure of an Auto Group 6 isolation has occurred while the drywell is lined up for venting and Fuel damage has occurred.
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1 Page 15 of 48 Scenario Summary The plant is operating at 100% power at MOL in the operating cycle when the crew takes the watch. A LEFM input is bad. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> have elapsed since A LEFM input has gone bad and the crew is directed to lower power to less than the license limit of 2381 MWth.
Event 1 begins when crew commences lowering power with Reactor Recirculation to 2381 MWth (ATC ACTION) (RX Manipulation) per 2.1.10. Event ends when Rx power is lowered to less than or equal to 2381MWth.
Event 2 begins when an Annunciator 9-5-2/G-6 CRD ACCUM LOW PRESS OR HIGH LEVEL is received. ATC addresses the alarm card and receives the report that CRD 26-27 accumulator N2 pressure is reading 935 psig. ATC directs the NLO to charge CRD accumulator per procedure 2.2.8 CONTROL ROD DRIVE HYDRAULIC SYSTEM. CRS addresses Tech Specs and determines that LCO 3.1.5 is not met and enters the Required ACTION of condition A and declares CRD 26-27 INOPERABLE or SLOW. (CRS TS ACTION) (Slow most probable).
Event 3 begins when Annunciator B-1/E-7 TURB EH FLUID SUPPLY FILTER B HIGH D/P is received.
The BOP will address the alarm card and start A DEH pump and stop B DEH pump. (BOP ACTION)
Event 4 begins when a B RR pump has rapid rising vibes and when levels rise to 2 danger levels, then the ATC trips B RR pump (ATC ACTION) per 9-4-3/C-7. The CRS enters 2.4RR and directs the BOP to perform subsequent operator actions for a tripped RR pump. The BOP performs Attachment 1 of 2.4RR. (BOP ACTION). The CRS addresses Tech Specs and determines that LCO 3.4.1 is not met and enters the required actions on Condition B. (CRS TS ACTION)
Event 5 begins when Drywell pressure commences to rise due to RCS boundary leakage. The crew enters 2.4PC due to rising PC pressure and temperatures. The BOP will vent containment per 2.2.60 hardcard.
(BOP ACTION). The CRS will determine that venting will not be able to maintain PC pressure less than 1.84 psig and will enter 2.1.5 and take the plant offline.
Event 6 begins when the ATC scrams the RX. An electrical ATWS occurs. CRS enters EOP-1A and then transitions to EOP-6A and EOP-7A. The CRS orders EOP-7A Table 41 actions:
Inject SLC Lower level to -60 Reactor Water Level (CT#2)
Inhibit ADS (CT#1)
Install Group 1 low level jumpers.
Perform 5.8.3 actions insert all control rods using Test Keylock switches. (CT#3) (BOP MC)
Event 7 begins 4 minutes after Event 5 started. This will cause rising radiation levels due to broken fuel.
2.4OG, 5.2FUEL, 5.1RAD, and EOP-5A are entered. When drywell pressure gets greater than 1.84 psig, a failure of a Group 6 isolation should be identified. For the fuel failure and rising radiation, the crew will take the following actions:
Insert a Group 6 Isolation per 2.1.22 (ATC ACTION) (ATC MC) (CT#4)
Perform Attachment 2 of 5.2FUEL if In-containment radiation monitors are 250R/HR (ATC ACTION)
(Attachment 2 contains action to isolate RX building sumps to prevent pumping water to RW)
Close MSIVs if reactor shutdown with MAIN STM LINE HI-HI Rad alarm is received (BOP ACTION)
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1 Page 16 of 48 NRC Event 8 is a passive Event and is incorporated into Summary Event 7. A failure of the Group 6 Isolation will occur. This can be identified with the Group 6 isolation signal at 1.84 psig drywell pressure. PC Venting is in-service IAW 2.4PC, the crew will need to insert a Group 6 Isolation IAW procedure 2.1.22 by placing the Reactor Building Radiation Monitor test switches to TRIP TEST.
The scenario can be terminated when all control rods have been inserted, Group 6 isolation is manually inserted, 5.2FUEL Attachment 2 is complete, Drywell pressure, RPV water level, and RPV pressure bands are ordered, and are being recovered towards their prescribed bands.
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2 Table of Contents NUREG 1021 FORM 3.3-1................................................................................................................................................. 2 Scenario Summary............................................................................................................................................................... 9 Event 1 - Shift CRD Pumps.................................................................................................................................................... 10 Event 2 - Raise Pressure Setpoint to Raise RX power........................................................................................................... 12 Event 3 - Respond to A APRM INOP Alarm......................................................................................................................... 14 Event 4 - Earthquake, Drywell Vent Monitor valves fail closed............................................................................................ 15 Event 5 - RWCU Steam Leak, Group 3 failure....................................................................................................................... 18 Event 6 - RWCU-MO-18 Failure to Close, TS......................................................................................................................... 20 Event 7 - A RR Seal Failure, 2.4PC, Scram........................................................................................................................... 21 Event 8 - A SGT failure to start during Group 6.................................................................................................................... 27 Event 9 - MSL Leak, ED, 2 SRVs fail to open.......................................................................................................................... 28 Scenario Record................................................................................................................................................................. 35 Shift Turnover..................................................................................................................................................................... 36 Revision Summary Rev.
Description Author 00 New Exam scenario.
C. Edgington 01 Comments from Chief Examiner on the Draft Outline submittal.
C. Edgington 02 Comments from Chief Examiner on the Draft Operating Test submittal.
C. Edgington 03 Comments from Chief Examiner during the NRC Validation Week C. Edgington
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2 Page 2 of 36 NUREG 1021 FORM 3.3-1 A.
Scenario Outline Facility: Cooper Nuclear Station Scenario No.: 2__
Scenario Source: IC 25 Operating-Test No.: CNS-2022-06 Examiners: ____________________________ Applicants/Operators:
Turnover:
The plant is at 78% power EOL. In Coastdown
- 1.
Shift CRD pumps to allow maintenance to get vibrations on CRD Pump A. (Start CRD Pump A and secure CRD Pump B.)
- 2.
Raise Rx power by raising DEH pressure setpoint from 926 to 945 psig at 7 psig per minute IAW 2.2.77.1 and 2.1.10.
Critical Tasks:
Identified in the Event Description on Events 5 and 8 Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N (ATC, CRS) Shift CRD pumps per 2.2.8 ATC shifts CRD pumps IAW procedure 2.2.8 for maintenance.
2 N/A N (BOP, CRS) Raise RX Power ATCO Monitors 9-5 parameters and Reactor Power on PMIS BOP changes pressure set on DEH from 926 psig to 945 psig to raise power IAW 2.2.77.1 and 2.1.10 for EOL coastdown.
CRS is the reactivity manager during the change in power.
3 NM14A I (ATC, CRS) Respond to APRM A failure ATCO addresses the alarm card and bypasses APRM A CRS reviews Tech Specs and determines a potential LCO only 4
HV02A TS (CRS)
Respond to a minor Earthquake BOP addresses the Drywell Vent Monitor valves closure CRS determines that LCO 3.4.5 Condition B is not met 5
CU01B RP12 MC (ATC)
Respond to a RWCU leak CRS Enters EOP-5A BOP places RX Quad FCUs in-service per EOP-5A ATC isolates RWCU prior to 2 areas exceeding max safe (CT#1) When a primary system is discharging into the secondary containment through a break and automatic isolation fails, the crew takes manual action from the control room to isolate the system prior to reaching Maximum Safe values in two areas for the same parameter.
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2 Page 3 of 36 6
Overrride ZDIPCIS SWS16 to NORM TS (CRS)
Respond to a RWCU-MO-18 failure to close CRS determines that LCO 3.6.1.3 is applicable 7
Respond to a #1 and #2 seal failure on A RR pump (RR-MO-43A will not isolate)
CRS enters 2.4RR BOP addresses 2.4RR ATCO trips the RR pump BOP vents PC per 2.4PC.
CRS enters EOP-1A and EOP-3A on high drywell pressure ATCO scrams on high drywell pressure 8
PC18A MC (BOP)
Respond to failure of SGT A on group 6 isolation BOP verifies Group 6 BOP identifies A SGT failed to start and starts it.
9
Respond to an Earthquake aftershock and a MSL leak BOP addresses 5.1QUAKE ATC addresses the rising steam tunnel temperatures ATC attempts to isolate MSL (2 MSLs fail to isolate)
CRS re-enters EOP-5A CRS determines that ED is required CRS enters EOP-2A (CT#2) When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 4 SRVs when maximum safe operating values are exceeded in two areas for the same parameter prior to the fourth area exceeding Max Safe Operating Temperature. (The two areas SE and SW Quad Areas exceed Max Safe Operating Temperature almost simultaneously; they are treated as a common area. CT is ED before a 4th area (NE Quad exceeds the Max Safe Operating Temperature between 4-6 minutes following the common area) other than those 2 exceeding Max Safe Operating Temperature) 10 TC07A TC07B TC07C C (ATC, CRS) Respond to a failure of SRVs C and H to OPEN Bypass valves fail closed (anticipate not available)
ATC identifies that 2 SRV fail to open and places all 8 SRV switches to open
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
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2 Page 4 of 36 B.
NUREG 1021 Rev 12 Criteria for Evaluation Scenarios, Table 3.4-1 and Table 3.4-2 Quantitative Attributes Table Attribute ES-3.3-1 Target Actual Description Events after EOP entry 1-2 3
- 1. Failure of SGT train to start
- 3. Failure of 2 SRVs to open Abnormal Events 2-4 3
- 1. Seal Failure on B RWCU pump (9-3-1/E-10)
- 2. APRM A INOP (alarm Card 9-5-1)
- 3. Failure of RR A seals (2.4RR)
Major Transients 1-2 1
- 1. MSL LEAK in Secondary Containment EOP entries requiring substantive action 1-2 3
- 1. EOP-2A
- 2. EOP-3A
- 1. EOP-2A Contingency #2 - Emergency Depressurization Pre-identified Critical Tasks 2
2
- 1. (CT#1) When a primary system is discharging into the secondary containment through a break and automatic isolation fails, the crew takes manual action from the control room to isolate the system prior to reaching Maximum Safe values in two areas for the same parameter.
- 2. (CT#2) When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 4 SRVs when maximum safe operating values are exceeded in two areas for the same parameter prior to the fourth area exceeding Max Safe Operating Temperature. (The two areas SE and SW Quad Areas exceed Max Safe Operating Temperature almost simultaneously, they are treated as a common area. CT is ED before a 4th area (NE Quad exceeds the Max Safe Operating Temperature between 4-6 minutes following the common area) other than those 2 exceeding Max Safe Operating Temperature)
Normal Events N/A 1
- 2. Raise RX power with pressure setpoint (BOP)
Reactivity Manipulations N/A 1
- 1. N/A
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2 Page 5 of 36 Manual Control of Automatic Function 1
2
- 2. Failure of the Group 3 (AUTO) (BOP)
Instrument/
Component Failures N/A 5
Total Malfunctions N/A 6
- 1. Failure of APRM A
- 2. Failure of SGT train to start
- 3. Failure of RWCU-MO-18 to Close
- 4. Failure of RR A seals
- 5. MSL Break in SC
- 6. Failure of 2 SRVs to open TS Evaluation 2
2
- 1. LCO 3.4.5 Condition B
- 2. LCO 3.6.1.3 Condition A
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2 Page 6 of 36 C.
NUREG 1021 Rev 12 Criteria for Evaluation Scenarios, Table 3.4-1 and Table 3.4-2 Critical Task # 1 When a primary system is discharging into the secondary containment through a break and automatic isolation fails, the crew takes manual action from the control room to isolate the system prior to reaching Maximum Safe values in two areas for the same parameter.
Safety Significance EOP-5A directs isolating primary system leaks into secondary containment when a maximum normal operating value is exceeded. Failing to do so can result in an unnecessary offsite release and endanger plant personnel.
Isolating the leak terminates the RCS discharge into secondary containment.
Cues Indication of rising or Maximum Operating values in an area of a system which is connected to the RCS, combined with abnormal system parameters (e.g. such as levels, pressures, and flow rates).
Field reports of visible/audible leaks into secondary containment.
Measurable Performance Indicators Crew places the control switch for the applicable isolation valve(s) to CLOSE.
Performance Feedback Indication for applicable isolation valve(s) Green light illuminates and Red light extinguishes.
Secondary Containment parameter(s) eventually stabilizes and lowers.
RPV and/or associated system parameters indicate leak has been isolated.
Applicability EOP-5A conditions where a system (primary or non-primary) is discharging into the secondary containment and manual isolation capability from the control room is possible. This includes manipulation of valve control switches and valve power supply control switches, as applicable. If the leaking system is required for adequate core cooling this task is not applicable.
Technical Bases EOP-5A directs that this action be taken when a maximum normal operating value is exceeded. Failing to do so can significantly change the mitigation strategy as an unnecessary release will result and possibly endangering plant personnel.
Justification for the chosen performance limit Before reaching two Maximum Safe values in two areas for the same parameter was chosen because that is the next EOP-5A significant action threshold, when Emergency Depressurization is required. Isolating the leak before reaching this level will avert the significant thermal transient on the RPV caused by Emergency Depressurization.
BWR Owners Group Appendix App. B, steps SC/T-3, SC/R-1, SC/L-1 Scenario Guide Requirements The scenario must be able to drive at least one secondary containment parameter to its Max Safe value in two plant areas if the crew does not take action to isolate the leak. The crew scramming and reducing RPV pressure to reduce the driving head of the leak should not prevent reaching the Max Safe value for a parameter in two plant areas.
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2 Page 7 of 36 Critical Task # 2 When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 4 SRVs when maximum safe operating values are exceeded in two areas for the same parameter prior to the fourth area exceeding Max Safe Operating Temperature. (The two areas SE and SW Quad Areas exceed Max Safe Operating Temperature almost simultaneously, they are treated as a common area. CT is ED before a 4th area (NE Quad exceeds the Max Safe Operating Temperature between 4-6 minutes following the common area) other than those 2 exceeding Max Safe Operating Temperature)
Safety Significance Should secondary containment parameters exceed their maximum safe operating values in more than one area, the RPV must be depressurized to preclude further degradation. RPV depressurization places the primary system in its lowest possible energy state, rejects heat to the suppression pool in preference to outside the containment, and reduces the driving head and flow of primary systems that are unisolated and discharging into the secondary containment.
The criteria of "two or more areas" specified identifies the rise in secondary containment parameters as a wide-spread problem which may pose a direct and immediate threat to secondary containment integrity, equipment located in the secondary containment, and continued safe operation of the plant.
Cues SPDS indication for secondary containment parameters indicate area radiation, area temperature, or area water level has exceeded its maximum safe operating value in two areas.
Measurable Performance Indicators Manipulation of any four SRV controls on Panel 9-3:
SRV-71A, SRV-71B, SRV-71E, SRV-71G, SRV-71D, SRV-71F.
Performance Feedback Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.
Applicability EOP-5A conditions, RCS leaks into secondary containment with the RPV pressurized.
Justification for the chosen performance limit Emergency Depressurization is required due to effects of a break spreading into and potentially affecting safety equipment and operations in more than one area; however, emergency depressurization is not allowed until the second area exceeds its Max Safe limit. Before the Max Safe limit is exceeded in a fourth area gives reasonable time for the crew to perform emergency depressurization before the leak hampers equipment or operations in an even more widespread area.
Opening 4 SRVs was chosen to comply with the PSTGs Required Minimum Number of SRVs required to be opened to perform an Emergency
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2 Page 8 of 36 Depressurization. This is also identified in step RC/P-12 of EOP-2A, at which to continue Emergency Depressurization with less than 4 SRVs open it requires going to TABLE 2 RPV DEPRESSURIZATION SYSTEMS to complete the ED. (The expectations is that the operators follow procedure guidance and open 6 SRVs but to satisfactory complete the critical task they need to comply with the PSTGs and are required to open the minimum number of SRVs to perform an Emergency Depressurization.)
BWR Owners Group Appendix App. B, steps SC/T-4.2, SC/R-2.2, SC/L-2.2.
Scenario Guide Requirements The scenario must be able to drive the selected parameter to its Max Safe value in four plant areas. Also, ensure the leak severity itself, or subsequent cold water injection, does not deplete RPV pressure (driving head) so low that Max Safe in a fourth area cannot be reached. The crew should be driven to ED, versus just reducing pressure, to provide a consistent, measurable performance indicator. The CT listed in the scenario should list which areas will exceed their Max Safety Operating Temperature limit first, second, third, and fourth.
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2 Page 9 of 36 Scenario Summary The plant is operating at 78% power at EOL in coast down. When the crew takes the watch, they are directed to shift CRD pumps for maintenance to get vibration data and to adjust pressure setpoint to 945 psig using DEH pressure setpoint to raise power IAW Step 2.5 of 2.2.77.1 DEH Control System.
(Pressure setpoint had been lowered to 926 psig to perform a surveillance the previous shift.)
Event 1 begins when crew takes the watch and the crew shifts CRD pumps for maintenance. (ATC ACTION) (NORMAL).
Event 2 begins when the crew commences adjusting pressure setpoint on DEH to 945 psig. The BOP (BOP ACTION) (RX Manipulation) will make this adjustment while the ATCO monitors 9-5 parameters for power and Reactor pressure to ensure they stay in band as directed per 2.1.10 and 2.2.77.1.
Event 3 begins following the power adjustment; APRM A will fail downscale. The ATCO will address this failure and bypass (ATC ACTION) as directed per the alarm card to clear the rod block. The CRS will determine that this is only a potential LCO 3.3.1.1 only.
Event 4 begins when a minor earthquake occurs causing a RWCU seal failure and the drywell vent monitor valves to close. The crew will enter 5.1QUAKE and the CRS will address Tech Specs for LCO 3.4.5 (RCS Leakage Detection Instrumentation) for the Drywell Vent Monitor out of service. (CRS TS ACTION)
Event 5 begins when it is identified that there are rising temperatures in the RWCU Pump B room.
This will either be identified from PMIS on SPDS 16 or the area high temperature alarm will be received. The CRS will enter EOP-5A and direct the RX Quad FCUS to be placed in-service (BOP ACTION) and direct the ATC to take actions to isolate RWCU (ATC ACTION). The alarm card will direct actions to isolate the leak as well. There will be a failure of Group 3 to auto isolate. RWCU-MO-15 will manually close (BOP ACTION, MC) (CT#1), and RWCU-MO-18 will not close.
Event 6 begins when the failure of RWCU-MO-18 to close is reported to the CRS. The CRS will address Tech Specs for LCO 3.6.1.3 (PCIV) Condition A for RWCU-MO-18. (CRS TS ACTION) If the crew does not isolate prior to the RWCU high area temperature isolation, then the CRS should identify that the instrumentation failed to isolate as well and the CRS would also enter LCO 3.3.6.1 (Primary Containment Isolation Instrumentation) Condition A, B, C, and F. (CRS TS ACTION)
Event 7 begins when RR Pump A has a failure of the #1 seal. The crew will enter 2.4RR to address the seal. Three minutes after the #1 seal failure, the #2 seal will fail. This will require entry into 2.4PC and rapid isolation of RR Pump A (ATC ACTION). RR Pump A will trip, but RR-MO-43A will fail open. The BOP will take action to vent the drywell per 2.4PC. (BOP ACTION). The crew will scram the plant. The CRS will enter EOP-1A and EOP-3A on rising drywell pressure and will order verifying group isolations and ECCS initiation. When verifying the Group 6 isolation, they will discover that SGT System A has failed to auto start and will start the SGT-EF-1E manually. (BOP ACTION).
Event 8 begins when the crew is directed to verify Group Isolations per EOP-1A. While verifying Group 6 Isolation it will be identified that SGT A failed to start. The crew will verify that A SGT will start by taking it to RUN. (BOP ACTION).
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3 Page 1 of 50 Table of Contents NUREG 1021 FORM 3.3-1................................................................................................................................................. 2 Scenario Summary............................................................................................................................................................... 9 Event 1 - B RFP to Idle Speed.............................................................................................................................................. 11 Event 2 - Raise Reactor Power with Control Rods................................................................................................................ 16 Event 3 - IRM D Downscale................................................................................................................................................. 19 Event 4 - TEC Pump C trip................................................................................................................................................... 20 Event 5 - CRD Flow Controller Failure................................................................................................................................... 21 Event 6 - SLC A Pump Breaker Trip..................................................................................................................................... 23 Event 7 - Grid Oscillations..................................................................................................................................................... 24 Event 8 - Torus Leak............................................................................................................................................................. 26 Event 9 - LOOP, DG1 fail to start, DG2 fail to auto start....................................................................................................... 29 Event 10 - ED on Torus level................................................................................................................................................. 33 Scenario Record................................................................................................................................................................. 37 Shift Turnover..................................................................................................................................................................... 38 Revision Summary Rev.
Description Author 00 New Exam scenario.
C. Edgington 01 Comments from Chief Examiner on the Draft Outline submittal.
C. Edgington 02 Comments from Chief Examiner on the Draft Operating Test submittal.
C. Edgington 03 Comments from Chief Examiner during the NRC Validation Week C. Edgington
Op-Test No.: CNS 2022-06 Scenario No.:
3 Page 2 of 50 NUREG 1021 FORM 3.3-1 A.
Scenario Outline Facility: Cooper Nuclear Station Scenario No.: 3__
Scenario Source: IC 8 Operating-Test No.: CNS-2022-06 Examiners: ____________________________ Applicants/Operators:
Turnover:
The plant is ~ 3% power BOL in Mode
- 1.
Place B RFP in-service to idle speed per 2.2.28 FEEDWATER SYSTEM STARTUP AND SHUTDOWN, Section 13, continuing at Step 13.3.
- a.
2.2.28 FEEDWATER SYSTEM STARTUP AND SHUTDOWN, Section 8, has been completed satisfactorily.
- 2.
Withdraw control rods to obtain 25% bypass valve position IAW 2.1.1 and 10.13.
Critical Tasks:
Identified in the Event Description on Events 8 and 9 Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N (BOP, CRS) Place RFP B In-service per 2.2.28 Section 13 (Idle Speed)
BOP performs Section 13 of 2.2.28 and gets RFP B up to Idle Speed 2
N/A R (ATC, CRS) Raise Reactor Power by withdrawing control rods to get to 25%
bypass ATC withdraws control rods (last rod moved 14-31 at 8)
BOP provides peer checking CRS provides the reactivity manager during the change in power.
3 NM05D C (ATC, CRS) Respond to IRM D downscale failure ATC addresses the alarm card and bypasses IRM D 4
SW07C C (BOP, CRS) Respond to TEC C pump trip BOP starts B TEC pump IAW alarm card 5
ZAICRD FC301(2) to 0 MC (ATC)
CRD Flow controller fail to zero CRS enters 2.4CRD ATC places CRD flow controller to manual 6
SL06 TS (CRS)
Respond to SLC Pump A breaker trip CRS determines LCO 3.1.7 is not met and enters Condition A
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3 Page 3 of 50 7
TS (CRS)
Respond to Grid Instabilities BOP addresses the alarm card for Grid oscillations and places 1FS and 1GS to P-T-L, this causes a loss of power to the ESST.
CRS enters 5.3GRID BOP addresses 5.3GRID actions CRS determines that LCO 3.8.1 is not met for loss of ESST 8
Respond to a TORUS Leak CRS enters EOP-3A for lowering Torus level BOP lines up makeup systems to fill the Torus BOP places HPCI to P-T-L prior to lowering below 9.6 feet (CT#1)
CRS enters EOP-2A BOP opens 6 SRVs (CT#3)
CT#1 When torus water level cannot be maintained above 11' and HPCI is not required for adequate core cooling, prevent HPCI operation prior to torus water level lowering below 9.6.
CT#3 When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 4 SRVs prior to torus water level falling below 6.
9 DG06A DG06B MC (BOP)
Loss of Offsite power and respond to a failure DG1 and DG2 to Auto start BOP places DG1 control switch to start o DG1 Will not start BOP places DG2 control switches to start (CT#2)
CT#2 When a loss of offsite power occurs with a failure of both DGs to Auto start, the crew will place the control switch for DG#2 to start and restore power within 15 minutes.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
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3 Page 4 of 50 NUREG 1021 Rev 12 Criteria for Evaluation Scenarios, Table 3.4-1 and Table 3.4-2 Quantitative Attributes Table Attribute ES-3.3-1 Target Actual Description Events after EOP entry 1-2 2
- 1. DG1 and DG2 fail to Auto Start
- 2. LOOP (5.3EMPWR)
Abnormal Events 2-4 4
- 1. IRM D fail downscale (2.3_9-5-1 alarm card)
- 2. TEC Pump C Trip (M-2/C-5 alarm card)
- 3. CRD Flow Controller failure (9-5-2/E-6)
- 4. Oscillating GRID Voltage (5.3GRID)
Major Transients 1-2 1
- 1. PC Leak EOP entries requiring substantive action 1-2 3
- 1. EOP-2A
- 2. EOP-3A
- 1. EOP-2A Contingency #2 - Emergency Depressurization Pre-identified Critical Tasks 2
3
- 3. (CT#1) When torus water level cannot be maintained above 11' and HPCI is not required for adequate core cooling, prevent HPCI operation prior to torus water level lowering below 9.6.
- 4. (CT#2) When a loss of offsite power occurs with a failure of both DGs to Auto start, the crew will place the control switch for DG#2 to start and restore power within 15 minutes.
- 5. (CT#3) When torus water level cannot be maintained above 9.6',
crew Emergency Depressurizes by opening at least 4 SRVs prior to torus water level falling below 6.
Normal Events N/A 1
Reactivity Manipulations N/A 1
- 1. Raise reactor power by withdrawing control rods (ATC)
Manual Control of Automatic Function 1
2
- 2. DG2 fail to Auto Start (BOP)
Instrument/
Component Failures N/A 5
- 2. TEC Pump C trip (BOP)
- 4. Oscillating GRID Voltage (BOP)
- 5. DG2 fail to Auto Start (BOP)
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3 Page 5 of 50 Total Malfunctions N/A 7
- 1. IRM D fail downscale
- 2. TEC C Pump Trip
- 3. CRD Flow Controller Failure
- 4. SLC Pump A Breaker Trip
- 5. Oscillating Grid Voltage
- 6. DG1 and DG2 fail to Auto Start
- 7. Torus Leak TS Evaluation 2
2
- 1. LCO 3.1.7 Condition A
- 2. LCO 3.8.1 Condition A
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3 Page 6 of 50 B.
NUREG 1021 Rev 12 Criteria for Evaluation Scenarios, Table 3.4-1 and Table 3.4-2 Critical Task #1 When torus water level cannot be maintained above 11' and HPCI is not required for adequate core cooling, prevent HPCI operation prior to torus water level lowering below 9.6.
Safety Significance Operation of the HPCI System with its exhaust discharge device not submerged will directly pressurize the torus. HPCI operation is therefore secured, as required, to preclude the occurrence of this condition. The consequences of not doing so may extend to failure of the primary containment from over-pressurization.
Cues Lowering Torus water level, at 11, as indicated on SPDS.
Measurable Performance Indicators Crew stops and prevents HPCI by one of the following on Panel 9-3:
- Depressing and holding the HPCI trip pushbutton, and placing HPCI Aux Oil Pump control switch in PTL
- Placing HPCI-MO-16, STM SUPP OUTBD ISOL VLV, control switch in CLOSE Depressing HPCI MANUAL ISOLATION pushbutton, if initiation signal present Performance Feedback HPCI speed lowers to zero on HPCI-SI-2792 on Panel 9-3 HPCI flow lowers to zero on HPCI-FIC-108 on Panel 9-3 Steam supply isolation valve HPCI-MO-15 and/or HPCI-MO-16 control switch green light illuminated and red light extinguished on Panel 9-3.
Applicability EOP-3A conditions with torus water level lowering.
Justification for the chosen performance limit When torus water level cannot be maintained above 11 feet is the EOP-3A, step SP/L-11 criteria for preventing HPCI operation to ensure HPCI exhaust does not directly impinge on the torus air space, thus presenting a challenge to the containment fission product barrier. Per the PSTGs, it states that the task should be done immediately. The next action point to scram is prior to 9.6 feet.
BWR Owners Group Appendix App. B, step SP/L-2.1 Scenario Guide Requirements The scenario should be designed such that it will be apparent to the crew within 10 minutes that they will be unable to stop the drop in suppression pool level (e.g. torus leak). This includes any necessary reports from the field. The initial report from the field should include the elevation of the leak in the torus (e.g. for this CT, at least below 11) and should state the leak is unisolable. The rate of fall in suppression pool level should give the crew at least 20 minutes from the start of the major event before reaching 9.6 feet. This is because emergency depressurization will be required at 9.6 feet. Other injection systems must be available to assure adequate core cooling.
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3 Page 7 of 50 Critical Task # 2 When a loss of offsite power occurs with a failure of both DGs to Auto start, the crew will place the control switch for DG#2 to start and restore power within 15 minutes.
Safety Significance Failure to recognize the auto start not occurring and energizing of the safety bus, and failure to take manual action per Procedure 5.3EMPWR will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.
Cues Indication and/or annunciation that all ac emergency buses are de-energized Bus energized lamps extinguished Circuit breaker position Bus voltage EDG status Control room lighting dimmed Measurable Performance Indicators Manipulation of controls as required to energize Div 2 AC emergency bus from Panel C:
Operator places DIESEL GEN 2 C/S to START on Panel C Performance Feedback Crew will observe light indication for equipment powered by Division 2 AC illuminate on Panel 9-3 and bus voltage ~4200V on Panel C.
Applicability Loss of off-site power events when all sources of off-site power are lost and a diesel generator fails to auto start or energize its bus. This is only applicable if manual action from the Control Room would be effective in energizing the bus.
Justification for the chosen performance limit 15 minutes is assigned because this would cause an upgrade from an ALERT emergency classification to a Site Area Emergency.
BWR Owners Group Appendix App. B, Contingency #1 Scenario Guide Requirements LOOP with a failure to AUTO start the DGs, without manual action the plant would remain in a SBO.
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3 Page 8 of 38 Critical Task # 3 When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening at least 4 SRVs prior to torus water level falling below
- 6.
Safety Significance The RPV is not permitted to remain at pressure if suppression of steam discharged from the RPV into the drywell cannot be assured. When the downcomer vent openings are not adequately submerged, any steam discharged from the RPV into the drywell may not condense in the suppression pool before torus pressure reaches unacceptable levels. RPV depressurization is required at or before the point at which this low water level condition occurs.
This reduces the amount of energy that may be discharged directly to the torus air space to as low as possible.
Cues Lowering Torus water level to 9.6, as indicated on SPDS.
Measurable Performance Indicators Manipulation of any four SRV controls on Panel 9-3:
SRV-71A, SRV-71B, SRV-71C, SRV-71E, SRV-71G, SRV-71H, SRV-71D, SRV-71F.
Performance Feedback Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS and Panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.
Applicability EOP-3A conditions with torus water level lowering.
Justification for the chosen performance limit Inability to maintain torus water level above 9.6 is the EOP-3A, Step SP/L-14 criteria for transitioning to emergency depressurization. Failure to emergency depressurize before the SRV tailpipes become uncovered would prevent a rapid depressurization and require a delay as alternate emergency depressurization would be required.
Opening 4 SRVs was chosen to comply with the PSTGs Required Minimum Number of SRVs required to be opened to perform an Emergency Depressurization. This is also identified in step RC/P-12 of EOP-2A, which to continue Emergency Depressurization with less than 4 SRVs open it requires going to TABLE 2 RPV DEPRESSURIZATION SYSTEMS to complete the ED.
(The expectations is that the operators follow procedure guidance and open 6 SRVs but to satisfactory complete the critical task they need to comply with the PSTGs and are required to open the minimum number of SRVs to perform an Emergency Depressurization.)
BWR Owners Group Appendix App. B, step SP/L-2.2 Scenario Guide Requirements The scenario should be designed such that it will be apparent to the crew within 10 minutes that they will be unable to stop the drop in suppression pool level (e.g. torus leak). This includes any necessary reports from the field. The initial report from the field should include the elevation of the leak in the torus (e.g. for this CT, at least below 11) and should state the leak is unisolable. The rate of fall in suppression pool level should give the crew at least 20 minutes from the start of the major event before reaching 9.6 feet. This is because preventing HPCI operation and scramming will also be required before 9.6 feet.
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3 Page 9 of 50 Scenario Summary The plant is operating at ~ 3% power at BOL in startup when the crew takes the watch. Turnover to the crew is to place RFP B in-service and up to idle speed IAW procedure 2.2.28, Section 13 and then continue withdrawing control rods IAW 10.13 pull sheet to obtain ~ 25% bypass valve position. Current bypass valve position is ~ 16% open.
Event 1 begins when the crew takes the watch and performs Section 13 of 2.2.28 to bring RFP B up to idle speed in preparation for placing it in-service. (BOP NORMAL EVENT)
Event 2 begins when the crew commences withdrawing control rods to get bypass valve position to 25%. (ATC RX-MANIPULATION)
Event 3 begins when IRM D fails downscale causing a Rod Block. IAW alarm card 9-5-1/E-8 IRM DOWNSCALE, the crew will bypass IRM D. (ATC ACTION)
Event 4 begins when Annunciator M-2/C-5 TEC PUMP C TRIP is received. The BOP will address the alarm card and start TEC Pump B. (BOP ACTION)
Event 5 begins when Annunciator 9-5-2/E-6 CRD CHARGING HEADER HIGH PRESSURE is received. The crew will take action per the alarm card and enter 2.4CRD and place the flow controller to Manual IAW 2.4CRD. (ATC ACTION)
Event 6 begins when Annunciator 9-5-2/G-7 LOSS OF CONT TO SQUIB VLVS is received. The CRS addresses Tech Specs and enters TS LCO 3.1.7 Condition A for a loss of one SLC Subsystem. (CRS TS ACTION)
Event 7 begins when multiple alarms are received on Board C due to grid oscillations. The BOP will take action per the alarm cards and place 1FS and 1GS to P-T-L (BOP ACTION) at which time the ESST will lose power and become INOPERABLE. The CRS will address Tech Specs and enter TS LCO 3.8.1 Condition A for a loss of an offsite power source. (CRS TS ACTION)
Event 8 begins when torus level starts lowering. The crew addresses Annunciator 9-3-2/G-5 and performs suppression pool makeup IAW 2.2.69.3. At - 2 inches torus level the CRS will enter EOP-3A and EOP-5A and direct suppression pool makeup per 5.8.14. At 11 feet torus level, the CRS will direct that HPCI be placed in P-T-L (CT#1), IAW EOP-3A and then enter EOP-1A and insert a manual scram.
Event 9 begins when the crew scrams; a loss of offsite power occurs, and both DGs will fail to start.
The crew will take the following actions:
BOP places DG1 C/S to start in an attempt to start DG1 (Fails to start),
BOP places DG2 C/S to start, DG2 will start (CT#2), (BOP ACTION) (BOP MC)
CRS enters 5.3EMPWR Event 10 begins when the torus level cannot be maintained above 9.6 feet. (MAJOR) The CRS will order Emergency Depressurization when Torus level lowers to 9.6 feet. The crew will open at least 4 SRVs. (CT#3)
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3 Page 10 of 50 The scenario can be terminated when Emergency Depressurization is complete and when Post ED RPV water level band has been ordered and is being recovered towards its ordered band.