ML22354A249

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1_RBS-2022-12 Outlines Draft
ML22354A249
Person / Time
Site: River Bend 
Issue date: 12/14/2022
From: Heather Gepford
NRC/RGN-IV/DORS/OB
To:
Entergy Operations
References
Download: ML22354A249 (1)
v • d • e


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: Riverbend Station Date of Exam: Dec 2022 Tier Group RO K/A Category Points SRO-Only Points K1 K 2

K 3

K4 K 5

K 6

A1 A2 A 3

A 4

G Total A2 G

Total

1.

Emergency and Abnormal Plant Evolutions 1

4 3 4 N/A 3

3 N/A 3

20 2

1 1 1 1

1 1

6 Tier Totals 5

4 5 4

4 4

26

2.

Plant Systems 1

2 2 3 2

3 2 2

3 2 3 2 26 2

2 0 1 1

0 2 1

2 1 0 1 11 Tier Totals 4

2 4 3

3 4 3

5 3 3 3 37

3.

Generic Knowledge and Abilities Categories CO EC RC EM 6

CO EC RC EM 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 K

1 K

2 K

3 A

1 A

2 G

K/A Topic(

s)

IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation X (G2.4.22) Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations (CFR: 41.7 / 41.10 / 43.5 / 45.12) 3.6 38 295003 (APE 3) Partial or Complete Loss of AC Power X

(AK1.08) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the PARTIAL OR COMPLETE LOSS OF AC POWER: Emergency diesel generator load limits (CFR: 41.5 /

41.7 / 45.7 / 45.8) 4.0 39 295004 (APE 4) Partial or Total Loss of DC Power X

(AK3.02) Knowledge of the reasons for the following responses or actions as they apply to PARTIAL OR COMPLETE LOSS OF DC POWER: Ground isolation/fault determination.

(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.3 40 295005 (APE 5) Main Turbine Generator Trip X

(AK1.01) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the MAIN TURBINE GENERATOR TRIP: Reactor pressure control (CFR: 41.5 / 41.7 / 45.7 / 45.8) 4.3 41 295006 (APE 6) Scram X

(AA2.02) Ability to determine or interpret the following as they apply to (APE 6) SCRAM: Control rod position (CFR: 41.10 / 43.5 / 45.13) 4.4 42 295016 (APE 16) Control Room Abandonment X

(AA1.11) Ability to operate or monitor the following as they apply to CONTROL ROOM ABANDONMENT: RCIC (CFR: 41.5 / 41.7 / 45.5 to 45.8) 4.2 43 295018 (APE 18) Partial or Complete Loss of CCW X

(AK3.01) Knowledge of the reasons for the following responses or actions as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER (CCW): Isolation of non-essential heat loads 3.4 44

(CFR: 41.5 / 41.10 / 45.6 / 45.13) 295019 (APE 19) Partial or Complete Loss of Instrument Air X

(AK2.09) Knowledge of the relationship between the PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following systems or components: Primary containment and auxiliaries (CFR: 41.8 / 41.10 / 45.3) 3.4 45 295021 (APE 21) Loss of Shutdown Cooling X

(AK1.03) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the LOSS OF SHUTDOWN COOLING:

Adequate core cooling (CFR: 41.5 / 41.7 / 45.7 / 45.8) 4.4 46 295023 (APE 23) Refueling Accidents X

(AK2.01) Knowledge of the relationship between the REFUELING ACCIDENTS and the following systems or components: Fuel handling equipment (CFR: 41.8 / 41.10 / 45.3) 3.5 47 295024 (EPE 1) High Drywell Pressure X (G2.4.21) Knowledge of the parameters and logic used to assess the status of emergency operating procedures critical safety functions or shutdown critical safety functions (CFR:

41.7 / 43.5 / 45.12) 4 48 295025 (EPE 2) High Reactor Pressure X

(EA1.01) Ability to operate or monitor the following as they apply to HIGH REACTOR PRESSURE: Main and reheat steam (CFR: 41.5 / 41.7 / 45.5 to 45.8) 3.1 49 295026 (EPE 3) Suppression Pool High Water Temperature X

(EK2.04) Knowledge of the relationship between the SUPPRESSION POOL HIGH WATER TEMPERATURE and the following systems or components: Plant process computer/parameter display systems (CFR: 41.7 / 41.10 / 45.3) 2.9 50 295027 (EPE 4) High Containment Temperature (Mark III Containment Only)

X (EK3.03) Knowledge of the reasons for the following responses or actions as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY):

Reactor SCRAM (CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.9 51 295028 (EPE 5) High Drywell

RED = Topic sampled on SRO Exam Temperature (Mark I and Mark II only) 295030 (EPE 7) Low Suppression Pool Water Level X

(EK3.05) Knowledge of the reasons for the following responses or actions as they apply to LOW SUPPRESSION POOL WATER LEVEL:

Suppression pool makeup system(s) operation (CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.6 52 295031 (EPE 8) Reactor Low Water Level X

(EA1.13) Ability to operate or monitor the following as they apply to REACTOR LOW WATER LEVEL: Reactor water level control system (CFR: 41.5 / 41.7 / 45.5 to 45.8) 4.1 53 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown X

(EA2.06) Ability to determine or interpret the following as they apply to (EPE 14) SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor pressure (CFR: 41.10 / 43.5 / 45.13) 4.2 54 295038 (EPE 15) High Offsite Radioactivity Release Rate X

(EA2.03) Ability to determine or interpret the following as they apply to HIGH OFFSITE RADIOACTIVITY RELEASE RATE: Radiation levels (CFR: 41.10 / 43.5 / 45.13) 3.4 55 600000 (APE 24) Plant Fire On Site X (G2.4.26) Knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage (CFR: 41.10 / 43.5 / 45.12) 3.1 56 700000 (APE 25) Generator Voltage and Electric Grid Disturbances X

(AK1.02) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Over-excitation (CFR: 41.5 / 41.7 / 45.7 / 45.8) 3.1 57 K/A Category Totals:

4 3 4

3 3

3 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 K

1 K

2 K

3 A

1 A

2 G

K/A Topic(

s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum 295007 (APE 7) High Reactor Pressure 295008 (APE 8) High Reactor Water Level 295009 (APE 9) Low Reactor Water Level 295010 (APE 10) High Drywell Pressure X

(AK3.02) Knowledge of the reasons for the following responses or actions as they apply to HIGH DRYWELL PRESSURE: Increased drywell cooling (CFR: 41.5 / 41.10 / 45.6 /

45.13) 3.5 58 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

(AK1.11) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the INADVERTENT REACTIVITY ADDITION: Thermal-hydraulic instability (CFR: 41.5 / 41.7 / 45.7 / 45.8) 4.0 59 295015 (APE 15**) Incomplete Scram 295017 (APE 17) High Offsite Release Rate 295020 (APE 20) Inadvertent Containment X 2.1.19 Ability to use available indications to evaluate system or component status 3.9 60

Isolation (CFR: 41.10 / 45.12) 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 X

(EA1.02) Ability to operate or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Leak detection system (CFR: 41.5 / 41.7 / 45.5 to 45.8) 3.8 61 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 X

(EA2.01) Ability to determine or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

Secondary containment pressure.

(CFR: 41.10 / 43.5 / 45.13) 4.0 62 295036 (EPE 13) Secondary Containment High Sump/Area Water Level X

(EK2.03) Knowledge of the relationship between the SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL and the following systems or components:

Radwaste system (CFR: 41.8 / 41.10 / 45.3) 2.9 63 500000 (EPE 16) High Containment Hydrogen Concentration K/A Category Point Totals:

1 1 1 1 1 1 Group Point Total:

6

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 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 1

205000 (SF4 SCS) Shutdown Cooling X

(A3.01) Ability to monitor automatic operation of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) including: Valve operation (CFR: 41.7 / 45.7) 3.7 2

206000 (SF2, SF4 HPCI)

High-Pressure Coolant Injection 207000 (SF4 IC)

Isolation (Emergency)

Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

(K5.04) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the LOW PRESSURE CORE SPRAY SYSTEM : Heat removal (transfer) mechanisms (CFR: 41.5 / 45.3) 3.2 3

209002 (SF2, SF4 HPCS)

High-Pressure Core Spray X

(K5.01) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the HIGH PRESSURE CORE SPRAY SYSTEM : Indications of pump cavitation (CFR: 41.5 / 45.3) 3.5 4

211000 (SF1 SLCS) Standby Liquid Control X (A4.02) Ability to manually operate and/or monitor the STANDBY LIQUID CONTROL SYSTEM in the control room: SLCS control switch (CFR: 41.7 / 45.5 to 45.8) 4.1 5

212000 (SF7 RPS) Reactor Protection X

(A2.09) Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: High Containment/drywell pressure (CFR: 41.5 / 45.6) 4.3 6

212000 (SF7 RPS) Reactor Protection X (291008K1.12) BREAKERS, RELAYS, AND DISCONNECTS Trip 2.9 7

indicators for circuit breakers and protective relays (CFR: 41.7) [Tier 4 Generic KA]

215003 (SF7 IRM)

Intermediate-Range Monitor X

(A2.01) Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Power supply degradation (CFR: 41.5 / 45.6) 3.1 8

215003 (SF7 IRM)

Intermediate-Range Monitor X

(K6.01) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the INTERMEDIATE RANGE MONITOR SYSTEM:

Reactor protection system (power supply)

(CFR: 41.7 / 45.7) 3.7 9

215004 (SF7 SRMS) Source-Range Monitor X

(A1.04) Ability to predict and/or monitor changes in parameters associated with operation of the SOURCE RANGE MONITOR SYSTEM including: Control rod block status (CFR: 41.5 / 45.5) 3.6 10 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

(A3.05) Ability to monitor automatic operation of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR including: Flow converter/comparator signals (CFR: 41.7 / 45.7) 3.2 11 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

(A1.05) Ability to predict and/or monitor changes in parameters associated with operation of the REACTOR CORE ISOLATION COOLING SYSTEM including: RCIC turbine speed (CFR: 41.5 / 45.5) 3.7 12 218000 (SF3 ADS)

Automatic Depressurization X

(K1.02) Knowledge of the physical connections and/or cause and effect relationships between the AUTOMATIC DEPRESSURIZATION SYSTEM and the following systems:

LPCS system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 4.2 13 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

(K5.01) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the PRIMARY CONTAINMENT ISOLATION SYSTEM / NUCLEAR 4.1 14

STEAM SUPPLY SHUTOFF :

Primary containment integrity (CFR: 41.5 / 45.3) 239002 (SF3 SRV) Safety Relief Valves X (A4.04) Ability to manually operate and/or monitor the SAFETY RELIEF VALVES in the control room:

Suppression pool temperature (CFR: 41.7 / 45.5 to 45.8) 4.2 15 259002 (SF2 RWLCS) Reactor Water Level Control X

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

Reactor water level control system (CFR: 41.7) 3.3 16 261000 (SF9 SGTS) Standby Gas Treatment X

(K4.04) Knowledge of STANDBY GAS TREATMENT SYSTEM design features and/or interlocks that provide for the following: Radioactive particulate filtration (CFR: 41.7) 3.4 17 262001 (SF6 AC) AC Electrical Distribution X

(K6.02) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the AC ELECTRICAL DISTRIBUTION:

Offsite power (CFR: 41.7 / 45.7) 4.2 18 262001 (SF6 AC) AC Electrical Distribution x

(K3.01) Knowledge of the effect that a loss or malfunction of the AC Electrical Distribution will have on the following systems or system parameters: Operationally significant AC loads (CFR: 41.7 / 45.4) 4.1 19 262002 (SF6 UPS)

Uninterruptable Power Supply (AC/DC)

X (K3.01) Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (AC/DC) will have on the following systems or system parameters: Reactor water level control system.

(CFR: 41.7 / 45.4) 3.5 20 263000 (SF6 DC) DC Electrical Distribution X (G2.1.27) CONDUCT OF OPERATIONS Knowledge of system purpose and/or function (CFR: 41.7) 3.9 21 264000 (SF6 EGE)

Emergency Generators (Diesel/Jet)

X (A4.03) Ability to manually operate and/or monitor the EMERGENCY GENERATORS (DIESEL/JET) in the control room: Transfer of emergency control between manual and automatic (CFR: 41.7 / 45.5 to 45.8) 3.6 22 264000 (SF6 EGE)

Emergency Generators (Diesel/Jet) x (K1.08) Knowledge of the physical connections and/or cause and effect relationships between the 3.0 23

Emergency Generators and the following systems:

Plant ventilation systems (CFR: 41.2 to 41.9 / 45.7 to 45.8) 300000 (SF8 IA) Instrument Air X

(A2.03) Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Low instrument air pressure (CFR: 41.5 / 45.6) 3.9 24 400000 (SF8 CCW)

Component Cooling Water X

(K3.06) Knowledge of the effect that a loss or malfunction of the COMPONENT COOLING WATER SYSTEM will have on the following systems or system parameters:

Recirculation system (CFR: 41.7 / 45.4) 3.8 25 510000 (SF4 SWS*) Service Water X

(K2.01) SERVICE WATER SYSTEM Knowledge of electrical power supplies to the following: Service water system pumps (Class 1E)

(CFR: 41.7) 3.7 26 K/A Category Point Totals:

2 2 3 2 3 2 2 3 2 3 2 Group Point Total:

26

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G K/A Topic(s)

I R #

201001 (SF1 CRDH) CRD Hydraulic X

(K1.10) Knowledge of the physical connections and/or cause and effect relationships between the Control Rod Drive Hydraulic System and the following systems: Control rod drive mechanisms (CFR: 41.1-3 to 41.5-8 / 45.1-6 / 45.8) 3.8 27 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism X

(A2.09) Ability to (a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Low reactor pressure (CFR: 41.5 / 45.6) 4 28 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information X

(K4.02) Knowledge of ROD CONTROL AND INFORMATION SYSTEM design features and/or interlocks that provide for the following: Bank position withdrawal sequence (CFR: 41.7) 3.7 29 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation X

(K6.03) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Recirculation System: AC electrical distribution system (CFR: 41.7 / 45.7) 3.5 30 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup x

(K6.08) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the REACTOR WATER 3.8 31

CLEANUP SYSTEM:

PCIS/NSSSS (CFR: 41.7 / 45.7) 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:

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

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

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

(K3.04) Knowledge of the effect that a loss or malfunction of the Fuel Pool Cooling and Cleanup will have on the following systems or system parameters:

Fuel pool water chemistry (CFR: 41.7 / 45.6) 2.7 32 234000 (SF8 FH) Fuel Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSIVLC) Main Steam Isolation Valve Leakage Control X

(A2.11) Ability to (a) predict the impacts of the following on the MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:

High reactor pressure (CFR: 41.5 / 45.6) 3 33 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating X

(A1.03) Ability to predict and/or monitor changes in parameters associated with operation of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM including: Reactor water level (CFR: 41.5 / 45.5) 3.8 34 245000 (SF4 MTGEN)

Main Turbine

Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas X

(A3.03) Ability to monitor automatic operation of the OFFGAS SYSTEM including:

System temperature control (CFR: 41.7 / 45.7) 2.8 35 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation X (G2.2.13) Knowledge of tagging and clearance procedures (CFR: 41.10 / 43.1 / 45.13) 4.1 36 290002 (SF4 RVI) Reactor Vessel Internals X

(K1.14) Knowledge of the physical connections and/or cause and effect relationships between the REACTOR VESSEL INTERNALS and the following systems:

Reactor water cleanup system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.5 37 510001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

2 0 1 1 0 2 1 2 1 0 1 Group Point Total:

11

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

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.

(G2.1.3) Knowledge of shift or short-term relief turnover practices (CFR: 41.10 / 45.13) 3.2 64 2.1.

(G2.1.4) 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 65 Subtotal 2

N/A

2.

Equipment Control 2.2.

(G2.2.15) Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, lineups or, tagouts (reference potential) (CFR:

41.10 / 43.3 / 45.13) 3.9 66 2.2.

(G2.2.6) 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.

(G2.3.12) 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.

(G2.4.17) Knowledge of emergency and abnormal operating procedures terms and definitions (CFR: 41.10 / 45.13) 3.9 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 292003 Neutrons K1.01 Explain the concept of subcritical multiplication 3.0 71 6.1 292005 Control Rods K1.10 State the purpose of flux shaping 2.9 72 Subtotal N/A

Thermodynamics 6.2 293005 Thermodynamic Cycles K1.03 Describe the steam quality/moisture effects on turbine integrity and efficiency 2.7 73 6.2 293009 Core Thermal Limits K1.13 Define MAPLHGR 3.6 74 6.2 293006 Fluid Statics and Dynamics K1.06 Discuss methods of prevention of fluid/water hammer 3.2 75 Subtotal N/A Tier 4 Point Total 6

Form 4.1-BWR SRO Boiling-Water Reactor Examination Outline Facility: Riverbend Station Date of Exam: Dec 2022 Tier Group RO K/A Category Points SRO-Only Points K1 K 2

K 3

K4 K 5

K 6

A1 A2 A 3

A 4

G Total A2 G

Total

1.

Emergency and Abnormal Plant Evolutions 1

N/A N/A 4

3 7

2 2

1 3

Tier Totals 6

4 10

2.

Plant Systems 1

2 3

5 2

1 1

1 3

Tier Totals 4

4 8

3.

Generic Knowledge and Abilities Categories CO EC RC EM CO EC RC EM 7

2 2

1 2

4. Theory Reactor Theory Thermodynamics

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

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 X

(AA2.05) Ability to determine or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Jet pump operability (CFR: 41.10 / 43.5 / 45.13) 3.5 76 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 X (G2.4.47) Ability to diagnose and recognize trends in an accurate and timely manner using the appropriate control room reference material (reference potential)

(CFR: 41.10 / 43.5 / 45.12) 4.2 77 295006 (APE 6) Scram 295016 (APE 16) Control Room Abandonment

/ 7 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 or interpret the following as they apply to REFUELING ACCIDENTS: Emergency plan implementation (CFR: 41.10 / 43.5 / 45.13) 4.4 78 295024 (EPE 1) High Drywell Pressure 295025 (EPE 2) High Reactor Pressure X (G2.3.6) RADIATION CONTROL Ability to approve liquid or gaseous release permits (CFR:

41.13 / 43.4 / 45.10) 3.8 79 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:

4.0 80

Suppression pool water temperature (CFR: 41.10 / 43.5 / 45.13) 295027 (EPE 4) High Containment Temperature (Mark III Containment Only)

X (G2.4.40) EMERGENCY PROCEDURES/PLAN Knowledge of SRO responsibilities in emergency plan implementing procedures (SRO Only) (CFR: 43.5 / 45.11) 4.5 81 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) 295030 (EPE 7) Low Suppression Pool Water Level 295031 (EPE 8) Reactor Low Water Level 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown X

(EA2.08) Ability to determine or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

SCRAM discharge volume level (CFR: 41.10 / 43.5 / 45.13) 4 82 295038 (EPE 15) High Offsite Radioactivity Release Rate 600000 (APE 24) Plant Fire On Site 700000 (APE 25) Generator Voltage and Electric Grid Disturbances 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 K

1 K

2 K

3 A

1 A

2 G

K/A Topic(

s)

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

(AA2.04) Ability to determine and/or interpret the following as they apply to High Reactor Pressure: 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 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 X

(AA2.03) Ability to determine or interpret the following as they apply to ABNORMAL OFFSITE RELEASE RATE: Radiation levels (CFR: 41.10 / 43.5 / 45.13) 3.9 84 295020 (APE 20) Inadvertent Containment Isolation 295022 (APE 22) Loss of Control Rod Drive Pumps 295029 (EPE 6) High Suppression Pool Water Level 295032 (EPE 9) High Secondary

Containment Area Temperature 295033 (EPE 10) High Secondary Containment Area Radiation Levels X (G2.2.38) Knowledge of conditions and limitations in the facility license (CFR: 41.7 / 41.10 / 43.1 /

45.13) 4.5 85 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 K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode X (G2.2.22) Knowledge of limiting conditions for operation and safety limits (CFR: 41.5 / 43.2 / 45.2) 4.7 86 205000 (SF4 SCS) Shutdown Cooling 206000 (SF2, SF4 HPCI)

High-Pressure Coolant Injection 207000 (SF4 IC)

Isolation (Emergency)

Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray X

(A2.12) Ability to (a) predict the impacts of the following on the HIGH PRESSURE CORE SPRAY SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: High suppression pool level (CFR: 41.5 / 45.6) 3.2 87 211000 (SF1 SLCS) Standby Liquid Control 212000 (SF7 RPS) Reactor Protection 215003 (SF7 IRM)

Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X (G2.2.13) EQUIPMENT CONTROL Knowledge of tagging and clearance procedures (CFR: 41.10 / 43.1 / 45.13) 4.3 88 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS)

Automatic Depressurization 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X (G2.4.25) Knowledge of fire protection procedures (CFR: 41.10 / 43.5 / 45.13) 3.7 89

239002 (SF3 SRV) Safety Relief Valves 259002 (SF2 RWLCS) Reactor Water Level Control 261000 (SF9 SGTS) Standby Gas Treatment 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS)

Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE)

Emergency Generators (Diesel/Jet) 300000 (SF8 IA) Instrument Air 400000 (SF8 CCW)

Component Cooling Water X

(A2.01) Ability to (a) predict the impacts of the following on the COMPONENT COOLING WATER SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of CCW pump (CFR: 41.5 / 45.6) 3.9 90 510000 (SF4 SWS*) Service Water K/A Category Point Totals:

2 3 Group Point Total:

5

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G K/A Topic(s)

I R #

201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation x

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

Detector malfunctions (CFR: 41.5 / 43.5 / 45.6) 3.7 91 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 X

(K6.05) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the FUEL HANDLING: Upper fuel pool water inventory (Mark III) 3.4 92

(CFR: 41.7 / 45.7) 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 X (G2.4.5) Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions (CFR:

41.10 / 43.5 / 45.13) 4.3 93 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:

1 1

1 Group Point Total:

3

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

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.

(G2.1.35) Knowledge of the fuel handling responsibilities of SROs (CFR: 43.7) 3.9 94 2.1.

(G2.1.7) Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation (CFR: 41.5 / 43.5 /

45.12 / 45.13) 4.7 95 Subtotal N/A 2

2.

Equipment Control 2.2.

(G2.2.17) 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.

(G2.2.38) Knowledge of conditions and limitations in the facility license (CFR: 41.7 /

41.10 / 43.1 / 45.13) 4.5 97 Subtotal N/A 2

3.

Radiation Control 2.3.

(G2.3.6) 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.

(G2.4.6) Knowledge of emergency and abnormal operating procedures major action categories (CFR: 41.10 / 43.5 / 45.13) 4.7 99 2.4.

(G2.4.38) Ability to take actions required by the facility emergency plan implementing procedures, including supporting or acting as emergency coordinator (CFR: 41.10 / 43.5 /

45.11) 4.4 100 Subtotal N/A 2

Tier 3 Point Total 7

Form 4.1-1 Record of Rejected Knowledge and Abilities Tier/Group New Randomly Selected K/A Reason for Rejection (Old K/A) 2/1 300000 (SF8 IA)

Instrument Air Q.24 (A2.03)

Low instrument air pressure (A2.02) Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Component cooling water system malfunction.

RBS instrument Air compressors are air cooled, not water cooled.

(CFR: 41.5 / 45.6) 1/1 295027 (EPE 4)

Containment Temperature (Mark III Containment Only)

Q. 51 (EK3.03)

Reactor Scram (EK3.02) Knowledge of the reasons for the following responses or actions as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY):

(Containment spray)

(CFR: 41.5 / 41.10 / 45.6 / 45.13)

RBS does not have Containment Spray.

2/1 400000 (SF8 CCW)

Component Cooling Water Q. 90 (A2.01)

Loss of CCW pump (A2.12) Ability to (a) predict the impacts of the following on the COMPONENT COOLING WATER SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of cooling to reactor recirculation pump variable frequency drive (CFR: 41.5 / 45.6)

Recirc Pumps do not have variable frequency drives.

2/2 241000 (SF3 RTPRS)

Reactor/Turbine Pressure Regulating Q 34 (A1.03)

Reactor Level (A1.24) Ability to predict and/or monitor changes in parameters associated with operation of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM including: Main turbine eccentricity (CFR: 41.5 / 45.5)

K/A was too difficult to write a question.

2/2 290003 (SF9 CRV)

Control Room Ventilation Q 36 (2.2.13)

Knowledge of tagging and clearance procedures (G2.2.12) Knowledge of surveillance procedures (CFR: 41.10 / 43.2 / 45.13)

K/A was too difficult to write a question.

Form 3.2-1 Administrative Topics Outline Facility: ____________RBS________________

Date of Examination: _____________

Examination Level: RO SRO Operating Test Number: __________

Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations Determine RCIC flow rate based on the minimum Decay Heat Removal Injection Rate.

K/A 2.1.7 N

R Conduct of Operations Determine Suppression Pool level utilizing 3.

K/A 2.1.19 D

R Equipment Control Creation of a tagout for pump.

K/A 2.2.13 M

R Radiation Control Determine expected dose and maximum stay time for a locked high rad entry.

K/A 2.3.12 N

R Emergency Plan N/A

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:

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

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)

  • 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).

Form 3.2-1 Administrative Topics Outline Facility: ____________RBS________________

Date of Examination: _____________

Examination Level: RO SRO Operating Test Number: __________

Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations Determine if a predictor case is in the acceptable range of actual plant data.

K/A 2.1.7 N

R Conduct of Operations Determine Suppression Pool level utilizing 3 and direct EOP actions based on data.

K/A 2.1.19 M

R Equipment Control Verify APRM/LPRM operability from a test procedure.

K/A 2.2.37 M

R Radiation Control Determine expected dose and maximum stay time for a locked high rad entry and fill out appropriate forms based on the data.

K/A 2.3.12 N

R Emergency Plan Determine EAL classification and PAR recommendations.

K/A 2.4.44 M

R

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:

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

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)

  • 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).

Form 3.2-2 Control Room/In-Plant Systems Outline Facility: ________RBS _______________ Date of Examination: _____________

Operating Test Number: _2022_____

Exam Level:

RO-only SRO-I SRO-U System/JPM Title Type Code Safety Function Control Room Systems

a. Alternating CRD Pumps, RJPM-OPS-S1 A-N-S 1
b. Alternating CCS Pumps, RJPM-OPS-S2 A-N-S 8
c. HPCS CST to CST flow test, RJPM-OPS-S3 A-EN-N-S 2
d. CMS Hydrogen Analyzer Ctmt sample, RJPM-OPS-S4 N-EN-S 5
e. Supplying Fuel Pool Heat Exchanger(s) from NSW, RJPM-OPS-S5 N-E-S 4
f. Control Rod Withdrawal Limiter test, RJPM-OPS-S6 A-N-S 7
g. Paralleling the DIV 1 Diesel Generator RJPM-OPS-S7 D-S 6
h. Overriding high rad signal for Fuel Building Ventilation RJPM-OPS-S8 L-N-S 9

In-Plant Systems

i. Enclosure 5, RJPM-OPS-P1 D-E-L-R 3
j. Reduce RCIC flow, RJPM-OPS-P2 A-E-N-L 4
k. Restoration of Fuel Building Dampers, RJPM-OPS-P3 N-E-L-R 9

Code License Level Criteria RO SRO-I SRO-U Req Actual Req Actual Req Actual (A)lternate path 4-6 5

4-6 5

2-3 2

(C)ontrol room (D)irect from bank 9

2 8

1 4

1 (E)mergency or abnormal in-plant 1

4 1

4 1

3 (EN)gineered safety feature (for control room system) 1 2

1 2

1 2

(L)ow power/shutdown 1

4 1

4 1

3 (N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 2 9

2 9

1 4

(P)revious two exams (randomly selected) 3 0

3 0

2 0

(R)adiologically controlled area 1

2 1

2 1

2 (S)imulator

1. Determine the number of control room system and in-plant system job performance measures (JPMs) to develop using the following table:

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

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:

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:

Code License Level Criteria RO SRO-I SRO-U (A)lternate path 4-6 4-6 2-3 (C)ontrol room (D)irect from bank 9

8 4

(E)mergency or abnormal in-plant 1

1 1

(EN)gineered safety feature (for control room system) 1 1

1 (L)ow power/shutdown 1

1 1

(N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 2 2

1 (P)revious two exams (randomly selected) 3 3

2 (R)adiologically controlled area 1

1 1

(S)imulator

ALTERNATING CRD PUMPS:

Section 6.1 of SOP-2. The Applicant will start the standby pump and secure the other pump.

Alternate Path-The procedure prompts the operator to check drive water D/P is 250 PSID and if not adjust it per section 6.16. D/P will be 230 PSID.

ALTERNATING CCS PUMPS:

Section 5.1 of SOP-12. The Applicant will start the standby pump.

Alternate Path-The procedure prompts the operator to check motor amps. The motor overload alarm will come in. The Applicant will stop the pump.

HPCS MANUAL STARTUP CST TO CST:

Section 4.5 of SOP-30. The Applicant will start the pump.

Alternate Path-An auto initiation signal will come in after the pumps is started. The Applicant will secure the system IAW AOP-34.

MANUAL CMS SYSTEM STARTUP AND OPERATION:

Section 4.2 of SOP-84. The Applicant will start the system and sample the RWCU valve nest room.

SUPPLYING FUEL POOL HEAT EXCHANGER(S) FROM SERVICE WATER:

Section 5.9 of SOP-16. The Applicant will manipulate a total of 10 valves.

ROD WITHDRAWAL LIMITER FUNCTIONAL TEST:

STP-500-0704. The Applicant will start on step 8.

Alternate Path-The control rod will drift out when the applicant tests the rod block signal.

The applicant should fully insert the control rod.

PARALLELING THE DIV 1 DIESEL GENERATOR :

Section 4.5 of SOP-53. The Applicant will start on step 13. The Applicant will adjust voltage and frequency and then close the diesel output breaker.

OVERRIDING HIGH RAD SIGNAL FOR FUEL BUILDING VENTILATION:

Section 5.4 of SOP-62. The Applicant will start on step 4. The Applicant will manipulate two override switches, open 2 dampers, start a supply fan, and close two dampers.

ENCLOSURE 5:

of EOP-5. The applicant will simulate performing actions to defeat the Main Turbine cross-around pressure closure signals for MSR steam supply valves.

Note-The JPM takes place in the RCA/Turbine Bldg.

REDUCE RCIC FLOW IAW AOP-31:

2 of AOP-31. The Applicant will attempt to throttle two valves at the remote shutdown panel to establish desired flow rate. One of the valves will not open.

Alternate Path-The Applicant will have to make flow adjustments with the flow controller instead.

RESTORATION OF FUEL BUILDING DAMPERS :

.2 of AOP-31. The Applicant will simulate manipulating two manual valves.

Note-The JPM takes place in the RCA/Fuel Bldg.

Facility: __RBS____________

Scenario #: ______1___________

Scenario Source: __________________

Op. Test #: __________________

Examiners: __________________

Applicants/

Operators: __________________

Initial Conditions: Mode 1, 100% Power Turnover: STP-256-0203, Division I Standby Cooling Tower Fans Operability Test.

Critical Tasks:

1. Manually Scram the Reactor within 5 minutes of RPV water level reaching Level 3.
2. Emergency Depressurize and inject with low pressure ECCS systems to restore RPV water level within 15 minutes of RPV water level reaching -162 inches.

Event Event Description

  • Attribute type
    • CRS
    • ATC
    • BOP 1

Start Standby Cooling Tower #1 Fans IAW STP-256-0203, Division I Standby Cooling Tower Fans Operability Test.

N 2

A degraded CCP pump (Low Discharge Pressure) and the standby pump failures to auto-start on low pressure. Manual start of the standby pump and entry into AOP-11 are required.

A C,MC 3

Plugging of the Condensate Demineralizers causing a low suction pressure for the Feedwater Pumps. The ATC will lower Reactor Power to restore suction pressure IAW AOP-6.

A R

4 Spurious Pressure perturbation causes a RCIC isolation on differential pressure. The DIV 2 RCIC steam isolation valve fails to close requiring manual isolation. Entry into TS LCO 3.5.3 Condition A and TS LCO 3.6.1.3 Condition A. are required A

TS C,MC 5

Failed Condensate Makeup Valves resulting in a low hotwell level. The ATC should take manual control of the makeup valve controller to restore level IAW ARP.

A I,MC 6

Control Power fuses for SSW Pump A blow. Entry into TS LCO 3.7.1 Condition A is required.

TS

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task

Event Event Description

  • Attribute type
  • CRS
  • ATC
  • BOP 7

Plugging of the Condensate Demineralizers continues causing a complete loss of suction pressure for the Feedwater pumps. RPS fails to actuate on Level 3. ATC operator must insert a manual Scram IAW AOP-1.

EOP-1 entry is required.

EP,CT M

M M

8 B21-F051D will fail open after the scram. The BOP Operator will close the SRV with handswitch.

E C,MC 9

A coolant leak occurs on the A Recirc Loop. A bus fault occurs on the DIV 3 bus resulting in a loss of HPCS.

Emergency Depressurization and low pressure ECCS injection is required when RPV lowers below -162 inches.

EC,CT M

M M

10 RWCU Pumps will fail trip on Level 2 signal. The ATC Operator will trip the pumps manually.

E C,MC

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task Attribute Target Actual Malfunctions after EOP entry 1-2 2

Abnormal Events 2-4 4

Major Transients 1-2 2

EOP entries requiring substantive action 1-2 1

EOP contingencies requiring substantive action 1 per set 1

Preidentified critical tasks 2 or more 2

CT-1 CT-2 Critical Task Manually Scram within 5 minutes of RPV water level reaching Level 3.

Emergency Depressurize and inject with low pressure ECCS systems to restore RPV water level within 15 minutes of RPV water level reaching -162 inches.

CT Criteria EOP-directed action that is essential to an events overall mitigative strategy.

EOP-directed action that is essential to an events overall mitigative strategy.

Safety Significance Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission.

Submergence is the preferred method for cooling the core. The core is adequately cooled by submergence when it can be determined that RPV water level is at or above the top of the active fuel (-162 inches).

All fuel nodes are then assumed to be covered with water and heat is removed by boiling heat transfer.

Initiating Cue Lowering RPV water level indication resulting from a loss of the Feedwater System.

When RPV water level indicates -162 inches.

Performance Feedback All Control rods inserted and Reactor Power indicating 0%.

Lowering RPV pressure and rising RPV water level.

Success Path Positioning the Mode Switch to the Shutdown position.

Opening 7 ADS/SRVs Measurable Performance Standard Within 5 minutes of RPV water level reaching Level 3. This is a reasonable amount of time, as agreed upon by Operations Department.

Within 15 minutes of RPV water level reaching -162 inches. This is a reasonable amount of time, as agreed upon by Operations Department.

Facility: __RBS____________

Scenario #: ______2___________

Scenario Source: __________________

Op. Test #: __________________

Examiners: __________________

Applicants/

Operators: __________________

Initial Conditions: Mode 1, 100% Power Turnover: Place RHR A in Suppression Pool Cooling mode IAW SOP-31.

Critical Tasks:

1. Manually Scram the Reactor before vacuum reaches 22.3 Hg Vac.
2. Manually initiate HCPS to restore RPV water level prior to water level reaching -162 inches.

Event Event Description

  • Attribute type
    • CRS
    • ATC
    • BOP 1

Place RHR A in Suppression Pool Cooling mode.

N 2

Main Condenser vacuum begins to degrade slowly. The ATC will lower Reactor power to maintain vacuum IAW AOP-5.

A R

3 Blown Fuse on SLC Pump A causes a loss of continuity. Entry into TS LCO 3.1.7 Condition A is required.

TS 4

RHR A experiences a sheared shaft causing low pressure injection pressure for RHR A and LPCS. The BOP operator will close the test return valve and trip the RHR A pump. Entry into TS LCO 3.5.1 Condition C and TS LCO 3.6.2.3 Condition A are required.

A TS C

5 Main Turbine Bearing pressure will degrade to 14 psig and the TGOP will fail to auto-start. The ATC operator will manually start the TGOP IAW ARP.

A I,MC 6

Main Condenser vacuum begins to degrade more. A manual scram is required before vacuum reaches 22.3 Hg Vac. Main Turbine will trip but the BPV will fail closed. RPS will fail to actuate automatically. EOP-1 entry required.

EP,CT M

M M

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task

Event Event Description

  • Attribute type
  • CRS
  • ATC
  • BOP 7

Motoring of the generator will occur after the Scram/Turbine trip. The ATC operator will force a reverse power trip by lowering VARS IAW AOP-2.

E I,MC 8

NPS buses will fail to fast transfer resulting in a loss of ALL Feed/Condensate pumps. RCIC will trip on Overspeed and will not be restored. The only high pressure Feed source is HPCS. HPCS Pump must be manually initiated for RPV water level control prior to RPV water level reaching -162 inches.

E,CT C,MC

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task Attribute Target Actual Malfunctions after EOP entry 1-2 2

Abnormal Events 2-4 3

Major Transients 1-2 1

EOP entries requiring substantive action 1-2 1

EOP contingencies requiring substantive action 1 per set 0

Preidentified critical tasks 2 or more 2

CT-1 CT-2 Critical Task Manually Scram the Reactor prior to Main Condenser Vacuum reaching 22.3 Hg Vac.(Main Turbine trip setpoint)

Manually initiate HCPS to restore RPV water level prior to water level reaching -162 inches CT Criteria This action would prevent a challenge to plant safety.

This action that would prevent a challenge to plant safety by preventing a condition that warrants the initiation of emergency depressurization.

Safety Significance A reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded during Main Turbine trip transient.

Submergence is the preferred method for cooling the core. The core is adequately cooled by submergence when it can be determined that RPV water level is at or above the top of the active fuel(-162 inches).

All fuel nodes are then assumed to be covered with water and heat is removed by boiling heat transfer.

Initiating Cue Lowering Main Condenser Vacuum.

Lowering RPV water level.

Performance Feedback All Control rods inserted and Reactor Power indicating 0%.

Rising RPV water level.

Success Path Positioning the Mode Switch to the Shutdown position.

Manual initiation of the HPCS System.

Measurable Performance Standard Prior to Main Condenser Vacuum reaching 22.3 Hg Vac (Main Turbine trip setpoint).

Prior to RPV water level reaching -162 inches (TS Safety Limit).

Op. Test No.: _____ Scenario No.: _____ Event No.: _____ Page ___ of ___

Event

Description:

Symptoms/Cues:

Time Position Applicants Actions or Behavior

Facility: __RBS____________

Scenario #: ______3___________

Scenario Source: __________________

Op. Test #: __________________

Examiners: __________________

Applicants/

Operators: __________________

Initial Conditions: Mode 1, 100% Power Turnover:

Critical Tasks:

1. Manually Scram the Reactor prior to radiation levels reaching Max Safe levels in the RCIC room.
2. Within 5 minutes, Emergency Depressurize the RPV when an unisolable steam leak causes radiation levels to exceed their Max Safe value in two separate areas of Secondary Containment.

Event Event Description

  • Attribute type
    • CRS
    • ATC
    • BOP 1

Recirc FCV A will slowly start to drift close. ATC Operator will arm and depress the HPU A shutdown pushbutton. Recirc flows will be mismatched after failure and require lowering flow on FCV B for balance.

TS LCO 3.4.1 Condition A entry is required.

A TS R

2 DIV 1 and DIV 2 fuses blow for ADS SRV B21-F047C. Requires entry into TS LCO 3.5.1 Condition E.

TS 3

Trip of an Isophase Bus duct cooling fan. The BOP Operator will start the standby fan IAW ARP.

A C

4 CNS makeup valve for the Seal Steam Generator will fail closed causing a low level. The BOP Operator will open the bypass valve to restore level IAW ARP.

A C,MC 5

APRM E will fail downscale. The ATC Operator will bypass the APRM.

I 6

Spurious initiation of the RCIC system. The BOP Operator will trip the RCIC turbine IAW AOP-34.

A I

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task Event Event Description
  • Attribute type
  • CRS
  • ATC
  • BOP

7 Directly after the RCIC trip, a unisolable steam leak will develop in the RCIC room causing a rise in room temperatures and radiation levels. Entry into EOP-3 required.

EP M

M M

8 Manually Scram the Reactor prior to radiation levels reaching Max Safe levels in the RCIC room. EOP-1 entry on the level 3.

CT 9

Recirc Pump B will fail to downshift on Level 3 signal. ATC Operator will downshift or trip the pump.

E I,MC 10 Emergency Depressurization is required when the RCIC room and RHR A room exceeds Max Safe area radiation levels.

EC.CT

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task Attribute Target Actual Malfunctions after EOP entry 1-2 1

Abnormal Events 2-4 4

Major Transients 1-2 1

EOP entries requiring substantive action 1-2 2

EOP contingencies requiring substantive action 1 per set 1

Preidentified critical tasks 2 or more 2

CT-1 CT-2 Critical Task Manually Scram the Reactor prior to Radiation Levels in the RCIC room reach Max Safe level.

Within 5 minutes, Emergency Depressurize the RPV when an unisolable steam leak causes radiation levels to exceed their Max Safe value in two separate areas of Secondary Containment.

CT Criteria EOP-directed action that is essential to an events overall mitigative strategy.

EOP-directed action that is essential to an events overall mitigative strategy.

Safety Significance If a discharge from a primary system is the source of radioactivity, a Reactor Scram should be adequate to terminate any further increase in secondary containment radiation levels.

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 or MSL tunnel.

Initiating Cue An unisolable steam leak causing radiation levels in the RCIC room to rise above Max Safe levels.

An unisolable steam leak causing radiation levels to exceed their Max Safe value in two separate areas of Secondary Containment.

Performance Feedback All Control rods inserted and Reactor Power indicating 0%.

Lowering RPV pressure and lowering Secondary Containment radiation levels.

Success Path Positioning the Mode Switch to the Shutdown position.

Opening 7 ADS/SRVs Measurable Performance Standard Prior to RCIC room radiation levels exceeding 9505 mR/Hr.

Within 5 minutes of Radiation Levels exceeding 9505 mr/Hr in the RCIC and RHR A rooms.

Op. Test No.: _____ Scenario No.: _____ Event No.: _____ Page ___ of ___

Event

Description:

Symptoms/Cues:

Time Position Applicants Actions or Behavior

Facility: __RBS____________

Scenario #: ______4___________

Scenario Source: __________________

Op. Test #: __________________

Examiners: __________________

Applicants/

Operators: __________________

Initial Conditions: Mode 2, 4% Power Turnover: Plant Startup IAW GOP-1 Critical Tasks:

1. Manually control RPV pressure with drains prior to RPV pressure reaching 1094.7 psig.
2. Manually initiate and control RCIC prior to RPV water level reaching -237 inches.

Event Event Description

  • Attribute type
    • CRS
    • ATC
    • BOP 1

ATC Operator will withdraw one control rod IAW the RMP.

R 2

Next Control Rod selected will begin to drift out and will stick when the ATC begins to insert the Control Rod. CRD Drive Water pressure should be raised to insert the stuck rod IAW AOP-61. Entry into TS LCO 3.1.3 is required.

A TS C

3 Steam Jet Air Ejector pressure control valve fails closed. The BOP operator must open the bypass valve to restore steam pressure and maintain condenser vacuum IAW ARP.

A C,MC 4

Bus fault occurs on EJS-SWG2A. The BOP will need to start Containment Unit Cooler C. Entry into AOP-

13. TS LCO 3.8.9 Condition A is required.

A TS C

5 The Main Turbine Bypass Valves will fail closed. The BOP operator should open Main Steam Line drain valves to maintain reactor pressure below RPS trip setpoint. Entry into AOP-17.

A,CT C,MC 6

A loss of Offsite Power/Station Blackout. Entry into AOP-50 and EOP-1. Transition to EOP-1C1-Alternate Level Control.

EC,EP M

M M

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task Event Event Description
  • Attribute type
  • CRS
  • ATC
  • BOP

7 RCIC fails to start, and the flow controller fails low. Manually initiate RCIC and manually control injection E,CT C,MC

    • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
  • (E)vents after EOP entry, (A)bnormal events, (EP)EOP entry, (EC)Contingency EOP entry, (CT)Critical Task Attribute Target Actual Malfunctions after EOP entry 1-2 2

Abnormal Events 2-4 4

Major Transients 1-2 1

EOP entries requiring substantive action 1-2 1

EOP contingencies requiring substantive action 1 per set 1

Preidentified critical tasks 2 or more 2

CT-1 CT-2 Critical Task Manually control RPV pressure with drains prior to RPV pressure reaching 1094.7 psig. (RPS setpoint)

Manually initiate RCIC to restore RPV water level prior to water level reaching -237 inches CT Criteria This action would prevent a challenge to plant safety.

This action that would prevent a challenge to plant safety by preventing a condition that warrants the initiation of emergency depressurization.

Safety Significance An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB.

If RPV water level drops below the Minimum Zero-Injection RPV Water Level (MZIRWL),

steam cooling may no longer be sufficient to preclude the peak clad temperature from exceeding 1800°F.

Initiating Cue Rising RPV pressure with bypass capability.

RCIC failed to start and control flow automatically.

Performance Feedback Lowering RPV pressure.

RCIC operating with flow above 600 gpm.

Success Path Positioning the Mode Switch to the Shutdown position.

Manual initiation of the RCIC System.

Measurable Performance Standard Prior to RPV pressure reaching 1094.7 psig.(RPS setpoint)

Prior to RPV water level reaching -237 inches.

Op. Test No.: _____ Scenario No.: _____ Event No.: _____ Page ___ of ___

Event

Description:

Symptoms/Cues:

Time Position Applicants Actions or Behavior

Crew A Facility:RBS Date of Exam:12/5/22 Operating Test No.:2022-12 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO1 SRO-I SRO-U RX 1

0 1

1 1

0 NOR 0

0 0

1 1

1 I/C 2

3 5

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 2

1 3

1 1

0 TS 0

0 0

0 2

2 RO2 SRO-I SRO-U RX 0

1 1

1 1

0 NOR 1

0 1

1 1

1 I/C 3

2 5

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 3

2 5

1 1

0 TS 0

0 0

0 2

2 RO9 SRO-I SRO-U RX 0

1 1

1 1

0 NOR 1

0 1

1 1

1 I/C 2

2 4

4 4

2 MAJ 1

1 2

2 2

1 Man. Ctrl 2

1 3

1 1

0 TS 0

0 0

0 2

2 RO SRO-I1 SRO-U RX 0

0 1

1 1

1 0

NOR 0

0 0

0 1

1 1

I/C 5

4 2

11 4

4 2

MAJ 2

1 1

4 2

2 1

Man. Ctrl 0

0 1

1 1

1 0

TS 2

2 0

4 0

2 2

Crew B Facility:RBS Date of Exam:12/5/22 Operating Test No.:2022-12 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO3 SRO-I SRO-U RX 1

0 1

1 1

0 NOR 0

1 1

1 1

1 I/C 2

2 4

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 2

2 4

1 1

0 TS 0

0 0

0 2

2 RO4 SRO-I SRO-U RX 0

1 1

1 1

0 NOR 1

0 1

1 1

1 I/C 3

2 5

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 3

2 5

1 1

0 TS 0

0 0

0 2

2 RO SRO-I2 SRO-U RX 0

0 0

0 1

1 0

NOR 0

0 0

0 1

1 1

I/C 5

4 3

12 4

4 2

MAJ 2

1 1

4 2

2 1

Man. Ctrl 0

0 1

5 1

1 0

TS 2

2 0

4 0

2 2

Crew C Facility:RBS Date of Exam:12/5/22 Operating Test No.:2022-12 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO5 SRO-I SRO-U RX 1

0 1

2 1

1 0

NOR 0

1 0

1 1

1 1

I/C 2

2 2

6 4

4 2

MAJ 2

1 1

4 2

2 1

Man. Ctrl 2

2 1

5 1

1 0

TS 0

0 0

0 0

2 2

RO6 SRO-I SRO-U RX 0

1 0

1 1

1 0

NOR 1

0 0

1 1

1 1

I/C 3

2 3

8 4

4 2

MAJ 2

1 1

4 2

2 1

Man. Ctrl 3

2 1

6 1

1 0

TS 0

0 0

0 0

2 2

RO SRO-I SRO-U1 RX 0

0 0

1 1

0 NOR 0

0 0

1 1

1 I/C 5

4 9

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 0

0 0

1 1

0 TS 2

2 4

0 2

2

Crew D Facility:RBS Date of Exam:12/5/22 Operating Test No.:2022-12 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO7 SRO-I SRO-U RX 1

0 1

1 1

0 NOR 0

1 1

1 1

1 I/C 2

2 4

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 2

2 4

1 1

0 TS 0

0 0

0 2

2 RO8 SRO-I SRO-U RX 0

1 1

1 1

0 NOR 1

0 1

1 1

1 I/C 3

2 5

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 3

2 5

1 1

0 TS 0

0 0

0 2

2 RO SRO-I SRO-U2 RX 0

0 0

1 1

0 NOR 0

0 0

1 1

1 I/C 5

4 9

4 4

2 MAJ 2

1 3

2 2

1 Man. Ctrl 0

0 0

1 1

0 TS 2

2 4

0 2

2

Facility:RBS Date of Exam:12/5/22 Operating Test No.:2022-12 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO SRO-I SRO-U RX 1

0 1

1 1

0 NOR 0

0 0

1 1

1 I/C 2

4 6

4 4

2 MAJ 1

1 1

2 2

1 Man. Ctrl 0

3 3

1 1

0 TS 0

0 0

0 2

2 RO SRO-I SRO-U RX 0

1 0

1 1

1 0

NOR 0

0 0

0 1

1 1

I/C 6

2 4

6 4

4 2

MAJ 1

1 1

1 2

2 1

Man. Ctrl 0

0 3

3 1

1 0

TS 2

0 0

2 0

2 2

RO SRO-I SRO-U RX 0

0 1

1 0

NOR 0

0 1

1 1

I/C 6

6 4

4 2

MAJ 1

1 2

2 1

Man. Ctrl 0

0 1

1 0

TS 2

2 0

2 2

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