ML22354A255

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7_RBS-2022-12 Outlines Final
ML22354A255
Person / Time
Site: River Bend Entergy icon.png
Issue date: 12/14/2022
From: Heather Gepford
NRC/RGN-IV/DORS/OB
To:
Entergy Operations
References
Download: ML22354A255 (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 (CFR: 41.5 / 41.10 / 45.6 /

45.13) 3.4 44 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 3.5 47

or components: Fuel handling equipment (CFR: 41.8 / 41.10 / 45.3) 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 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 3.6 52

WATER LEVEL:

Suppression pool makeup system(s) operation (CFR: 41.5 / 41.10 / 45.6 /

45.13) 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 3.1 57

RED = Topic sampled on SRO Exam GRID DISTURBANCES:

Over-excitation (CFR: 41.5 / 41.7 / 45.7 /

45.8)

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:

4.0 59

Thermal-hydraulic instability (CFR: 41.5 / 41.7 / 45.7 /

45.8) 295015 (APE 15**)

Incomplete Scram 295017 (APE 17) High Offsite Release Rate 295020 (APE 20) Inadvertent Containment Isolation X 2.1.19 Ability to use available indications to evaluate system or component status (CFR: 41.10 / 45.12) 3.9 60 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 4.0 62

PRESSURE: Secondary containment pressure.

(CFR: 41.10 / 43.5 /

45.13) 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 3.5 4

(CFR: 41.5 / 45.3) 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 indicators for circuit breakers and protective relays (CFR: 41.7) [Tier 4 Generic KA]

2.9 7

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 STEAM SUPPLY SHUTOFF : Primary containment integrity (CFR: 41.5 / 45.3) 4.1 14 239002 (SF3 SRV)

Safety Relief Valves X (A4.04) Ability to manually operate and/or monitor the SAFETY RELIEF VALVES in 4.2 15

the control room: Suppression pool temperature (CFR: 41.7 / 45.5 to 45.8) 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 Emergency Generators and the following systems:

Plant ventilation systems (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.0 23 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 CLEANUP SYSTEM: PCIS/NSSSS (CFR: 41.7 / 45.7) 3.8 31 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 2.7 32

following systems or system parameters:

Fuel pool water chemistry (CFR: 41.7 / 45.6) 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 2.8 35

(CFR: 41.7 / 45.7) 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 Thermodynam ics 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

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

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

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 4.4 78

ACCIDENTS: Emergency plan implementation (CFR: 41.10 / 43.5 / 45.13) 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: Suppression pool water temperature (CFR: 41.10 / 43.5 / 45.13) 4.0 80 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 3.9 84

(CFR: 41.10 / 43.5 /

45.13) 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 X (G2.2.13) EQUIPMENT CONTROL Knowledge of 4.3 88

Monitor/Local Power Range Monitor tagging and clearance procedures (CFR: 41.10 / 43.1 / 45.13) 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 3.9 90

abnormal operations: Loss of CCW pump (CFR: 41.5 / 45.6) 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)

(CFR: 41.7 / 45.7) 3.4 92 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSIVLC) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS)

Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste X (G 2.3.14) Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities, such as analysis and interpretation of radiation and activity readings as they pertain to administrative, normal, 4.3 93

abnormal, and emergency procedures, or analysis and interpretation of coolant activity, including comparison to emergency plan or regulatory limits 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 Subtotal N/

A 1

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 (G2.4.5) Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions 4.3 95 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 3

Tier 3 Point Total 7

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

2/2 268000 (SF9 RW) Radwaste Q93 G 2.3.14 Knowledge of radiation and contamination hazards (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)

K/A was rejected due to difficulty writing question to SRO level linked to Radwaste. Used KA as Tier 3.

3 G2.4.5 Q95 G2.4.5 Knowledge of operating procedures (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)

K/A was rejected to use previously identified KA as a Tier 3 instead of Tier 2 question.

Form 3.2-1 Administrative Topics Outline Facility: RBS________________

Date of Examination: 12/5/2022____

Examination Level: RO SRO Operating Test Number: __________

Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

A1 Conduct of Operations Determine Decay Heat Removal Injection Rate.

K/A 2.1.7-Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

IMPORTANCE RO 4.4 N

R A2 Conduct of Operations Determination of Primary Containment Water Level.

K/A 2.1.19-Ability to use available indications to evaluate system or component status.

IMPORTANCE RO 3.9 D

R A3 Equipment Control Tagout of FPW-P3 pump.

K/A 2.2.13-Knowledge of tagging and clearance procedures.

IMPORTANCE RO 4.1 M

R A4 Radiation Control Determine expected dose and entry requirements for a high dose evolution.

K/A 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.

IMPORTANCE RO 3.2 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[0], (S)imulator[0], or Class(R)oom [4]

Source and Source Criteria:

(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)[0]

(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes)[1]

(N)ew or Significantly[2] (M)odified from bank (no fewer than one)[1]

  • 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).

Rev 3 Form 3.2-1 Administrative Topics Outline Facility: RBS________________

Date of Examination: 12/5/2022 Examination Level: RO SRO Operating Test Number: _____

Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

A5 Conduct of Operations Verification of Core Parameters following Reactivity Manipulations.

K/A 2.1.7-Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

IMPORTANCE SRO 4.7 N

R A6 Conduct of Operations Determination of Primary Containment Water Level K/A 2.1.19-Ability to use available indications to evaluate system or component status.

IMPORTANCE SRO 3.9 M

R A7 Equipment Control LPRM operability.

K/A 2.2.37-Ability to determine operability or availability of safety-related equipment.

IMPORTANCE SRO 4.6 N

R A8 Radiation Control Determine Dose and entry requirements for a High Dose Evolution K/A 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities, such as analysis and interpretation of radiation and activity readings as they pertain to administrative, normal, abnormal, and emergency procedures, or analysis and interpretation of coolant activity, including comparison to emergency plan or regulatory limits IMPORTANCE SRO 3.8 N

R A9 Emergency Plan Determine EAL classification.

K/A 2.4.44-Knowledge of emergency plan implementing procedures protective action Recommendations.

IMPORTANCE SRO 4.4 M

R

Rev 3 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[0], (S)imulator[0], or Class(R)oom [5]

Source and Source Criteria:

(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)[0]

(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes)[1]

(N)ew or Significantly[2] (M)odified from bank (no fewer than one)[3]

  • 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: 12/5/2022 Operating Test Number: RB-2022-12 Exam Level:

RO SRO-I SRO-U System/JPM Title Type Code Safety Function Control Room Systems S1. Alternating CRD Pumps S1 A-N-S 1

S2. Alternating CCS Pumps A-N-S 8

S3. HPCS CST TO CST With Initiation Signal (SRO-U)

A-EN-N-S 2

S4. Sample Containment Through CMS (SRO-U)

N-EN-S 5

S5. Supplying Fuel Pool Heat Exchanger(s) from NSW N-S 4

S6. Rod Withdrawal Limiter Functional Test A-N-S 7

S7. Parallel Div 1 Diesel From Control Room (RO Only)

D-S 6

S8. Operate FB Vent System Following High-High Radiation L-N-S 9

In-Plant Systems P1. Defeating MSR interlocks (SRO-U)

D-E-L-R 3

P2. Reduce RCIC flow at the RSS panel (SRO-U)

A-E-EN-N-L 4

P3. Align Div 1 Standby Service Water to Fire Protection Water Supply (SRO-U)

N-E 8

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 5

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

2 8

1 4

1 (E)mergency or abnormal in-plant 1

3 1

3 1

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

1 3

1 3

(L)ow power/shutdown 1

2 1

2 1

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

2 9

1 9

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

3 0

2 0

(R)adiologically controlled area 1

1 1

1 1

1 (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

Control Room Systems S1. ALTERNATING CRD PUMPS:

K/A: 201001 Control Rod Drive Hydraulic System A2.14 Low drive header pressure Rating: 3.9 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.

S2. ALTERNATING CCS PUMPS:

K/A: 400000 Component Cooling Water System A4.01 CCW indications and control Rating: 3.8 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.

S3. HPCS MANUAL STARTUP CST TO CST:

K/A: 209002 High-Pressure Core Spray System A2.01 Automatic system initiation Rating:

4.5 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 and 1000 gpm flow. The Applicant will secure the system IAW AOP-34.

S4. SAMPLE CONTAINMENT THROUGH CMS:

K/A: 223001 Primary Containment System and Auxiliaries K3.04 Hydrogen concentration Rating: 3.5 Section 4.2 of SOP-84. The Applicant will start the system and sample the RWCU valve nest room.

S5. SUPPLYING FUEL POOL HEAT EXCHANGER(S) FROM NORMAL SERVICE WATER:

K/A: 233000 Fuel Pool Cooling and Cleanup A4.01 Valve operations Rating: 3.8 Section 5.9 of SOP-16. The Applicant will manipulate a total of 10 valves.

Control Room Systems S6. ROD WITHDRAWAL LIMITER FUNCTIONAL TEST:

K/A: 201005 Rod Control and Information System A2.13 Rod drift Rating: 4.4 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.

S7. PARALLEL DIV 1 DIESEL FROM THE CONTROL ROOM :

K/A: 264000 Emergency Generators A4.04 Voltage/Frequency: 4.1 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.

S8. OPERATE THE FB VENT SYSTEM FOLLOWING HIGH-HIGH RADIATION:

K/A: 288000 Plant Ventilation Systems A2.04 High Radiation Rating: 3.6 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.

In-Plant Systems P1. DEFEATING MSR INTERLOCKS K/A: 239001 MRSS Main and Reheat Steam System AA1.08 Reactor pressure Rating: 4.2 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.

P2. REDUCE RCIC FLOW AT THE RSS PANEL:

K/A: 295016 Control Room Abandonment AA1.11 RCIC Rating: 4.2 2 of AOP-31. The Applicant will attempt to throttle two valves to establish desired flow rate. One of the valves will not open. The applicant will then use the RCIC flow controller to adjust flow.

Note-The JPM takes place in control building outside RCA.

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

P3. ALIGN DIV 1 STANDBY SERVICE WATER TO FIRE PROTECTION WATER SUPPLY :

K/A: 286000 FPS Fire Protection System, A2.03, Ability to (a) predict the impacts of the following on the Fire Protection System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: AC electrical distribution failure Rating: 2.8/3.0 SOP-37, Section 5.3. The applicant will align division 1 SSW to FPW for firefighting in the control building.

Note-The JPM takes place in control building outside RCA.

REV 2 Facility: __RBS_________

Scenario #: ______1___________

Scenario Source: New__________

Op. Test #: RB-2022-12________

Examiners: __________________

Applicants/

Operators: __________________

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

Critical Tasks:

1. Manually Scram the Reactor before RPV water level reaches -43 inches.
2. Restore and Maintain RPV water level above -162 inches using low pressure injection systems within 15 minutes of water level lowering below -162 inches.

Event Event Description

  • Attribute type
    • CRS **ATC
    • BOP 1

Start Standby Cooling Tower #1 Fans IAW STP-256-0203.

N N

2 Degraded CCP pump A, standby pump fails to auto-start.

A C

C, MC 3

Low Feed pump suction pressure from High Condensate Demin D/P.

A C

C 4

Spurious Pressure perturbation causes a RCIC isolation, E51-F063 valve fails to close.

A TS,C C, MC 5

Failed level transmitter results in a low Hotwell level.

A I

I, MC 6

Control Power fuses for SSW Pump A blow.

TS 7

Loss of the Feed pumps from High Condensate Demin D/P. RPS fails to actuate. A bus fault will occur on DIV 3 480 VAC bus resulting in a loss of all high-pressure injection sources.

EP, CT C

C, MC 8

SRV B21-F051D will fail open after the scram.

E C

C, MC 9

RWCU valves G33-F001/F004 will fail to isolate and the RWCU Pumps will fail trip on any trip signal.

E C

C, MC 10 An unisolable coolant leak occurs on the A Recirc Loop. ED is required for low-pressure ECCS injection.

EC, 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

REV 2 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 Scram the Reactor before RPV water level reaches Level 2 (-43 inches WR).

Restore and Maintain RPV water level above -162 inches using low pressure injection systems within 15 minutes of water level lowering below -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.

RPV water level below -162 inches for 15 minutes would constitute an escalation in emergency classification level to a site area emergency due a potential loss of the fuel clad barrier.

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

When RPV water level has lowered below -

162 inches FZ.

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

Lowering RPV pressure from performing an emergency depressurization and rising RPV water level from a low-pressure ECCS injection source.

Success Path Positioning the Mode Switch to the Shutdown position.

Emergency Depressurization followed by injection from low pressure ECCS systems.

Measurable Performance Standard Prior to RPV water level reaching Level 2 (-43 inches WR).

Within 15 minutes of RPV water level lowering below -162 inches FZ.

If an applicant/operator or the crew significantly deviates from, or fails to follow, procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review

SUMMARY

REV 2 Event 1: Start Standby Cooling Tower #1 Fans IAW STP-256-0203.

The BOP operator starts the Standby Cooling Tower #1 Fans IAW STP-256-0203.

Event 2: Manual Trigger CCP Pump A begins degrading resulting CCP discharge header lowering to 87 psig. The standby CCP Pump (C) will fail to auto start at 95 psig. The BOP operator should take manual action and start the standby CCP pump. The CRS should enter AOP-11, LOSS OF REACTOR PLANT COMPONENT COOLING WATER.

Event 3: Manual Trigger Condensate Demineralizer D/P begins to elevate causing Feed pump suction header pressure to lower to 275 psig. The ATC will reduce Reactor Power to restore Feed pump suction pressure back to normal. The CRS should enter AOP-6, CONDENSATE/

FEEDWATER FAILURES and AOP-24, THERMAL HYDRAULIC STABILITY CONTROLS.

The Feed pump suction pressure alarm will clear when Reactor Power is below 95%.

Event 4: Manual Trigger Spurious Steam Pressure perturbation causes a RCIC System isolation. E51-F063, RCIC STEAM SUPPLY INBD ISOL VALVE, fails too automatically close. The BOP operator will take manual action IAW ARP-601-21 to close the PCIV.

The CRS should reference and enter TS LCO 3.5.3 Condition A for the RCIC being inoperable, and TS LCO 3.6.1.3 Condition A for the PCIV being inoperable.

Event 5: Manual Trigger Failure of CNS-LT103 level transmitter to CNS-LCV103 CNDS NORMAL MAKEUP LEVEL CONTROL VALVE and CNS-LCV104 CNDS EMERGENCY MAKEUP LEVEL CONTROL VALVE causes Condenser Hotwell level to lower. The ATC operator should manually restore Hotwell level by adjusting the tapeset for either makeup valve IAW the ARP.

Event 6: Manual Trigger Control Power fuses for SSW-P2A, SSW Pump A, blow causing a loss of control to the pump.

The CRS should reference and enter TS LCO 3.7.1 Condition E for SSW-P2A being inoperable.

Event 7: Manual Trigger Condensate Demineralizer D/P begins to elevate again causing the Feed pumps to ultimately trip on low suction header pressure. RPS will fail to actuate automatically when RPV level reaches 9.7 inches NR. The ATC will insert a manual scram prior to RPV level reaching -43 WR inches. The CRS will enter EOP-1, RPV Control on Level 3 and direct actions from that procedure. A loss of E22-S002 occurs after the scram resulting in a loss of all high-pressure feed sources.

Event 8:

SRV B21-F051D fails open after Lo-Lo set initiates. The BOP operator will take manual action and close the SRV.

Event 9:

The RWCU System will fail to isolate on a Level 2 (-43 inches WR) signal. The ATC operator will manually trip the RWCU pumps.

Event 10:

5 minutes after the Reactor Scram, an unisolable coolant leak on Recirc Loop A causing a rise in Drywell D/P and loss of RPV inventory. The CRS will direct the BOP operator to perform an emergency depressurization when RPV lowers below -162 inches FZ to allow low pressure injection sources to recover RPV level.

REV 2 Facility: __RBS____________

Scenario #: ______2___________

Scenario Source: __________________

Op. Test #: __________________

Examiners: __________________

Applicants/

Operators: __________________

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

Critical Tasks:

1. Manually inject with Standby Liquid Control prior to 110° F Suppression Pool Temperature
2. Restart a Reactor Feed Pump and restore RPV water level prior to -162 inches WR Event Event Description
  • Attribute type
    • CRS
    • ATC
    • BOP 1

Place RHR A in Suppression Pool Cooling mode.

N N

2 Main Condenser vacuum degradation.

A C

C 3

Blown Fuse on SLC Pump A TS 4

RHR A experiences a sheared shaft A

TS/C C

5 Main Turbine Bearing Header pressure degradation.

A I

I,MC 6

Main Condenser vacuum begins to degrade more requiring a Reactor Scram.

A, EP C

C C

7 ATWS - 10% power. Injection with SLC Pump B is required to lower power.

E,EP,CT M

M M

8 Heater string isolation & and trip of the feedpumps. RCIC trip and HPCS injection valve will not open.

E,CT C

C 9

MSL A outboard MSIV fail to auto close E

C 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 SPARE

REV 2 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 inject with Standby Liquid Control prior to 110° F Suppression Pool Temperature Restore feed pump to restore RPV water level prior to water level reaching -162 inches.

CT Criteria This action would prevent a challenge to plant safety by preventing Level / Power Conditions requiring lowering RWL to TAF This action that would prevent a challenge to plant safety by preventing a condition that warrants the initiation of emergency depressurization.

Safety Significance A combination of high reactor power

(>5%), high suppression pool temperature (110°F), and an open SRV or high drywell pressure indicates that heat is being added to the suppression pool faster than it is being removed. Resulting suppression pool heatup could result in loss of NPSH for ECCS pumps, primary containment overpressurization, and loss of primary containment integrity. Loss of primary containment integrity, in turn, could lead to a loss of adequate core cooling and uncontrolled release of radioactivity to the environment. ATWS strategies coordinate control of SP temperature, RPV water level, RPV pressure, and power to maintain margin to HCTL. Actions include injecting with boron.

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 Reactor Power greater than 5%

following manual scram actions taken Lowering RPV water level.

REV 2 Performance Feedback SLC pump injection pressure above reactor pressure SLC tank level lowering Rising RPV water level.

Success Path Inject with SLC Pump A or B prior to 110° F Suppression Pool Temperature Manually restore heater string flow path and starting a feed pump.

Measurable Performance Standard Prior to 110° F Suppression Pool Temperature Prior to RPV water level reaching -162 inches (TS Safety Limit).

If an applicant/operator or the crew significantly deviates from, or fails to follow, procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review

EVENT

SUMMARY

Event 1:

Place RHR A in Suppression Pool Cooling mode IAW SOP-31, Residual Heat Removal.

Event 2:

Main Condenser in leakage begins to degrade vacuum. The CRS will direct the ATC to lower Reactor Power to stabilize vacuum and assign AOP-5, Loss of Condenser Vacuum to the ATC operator.

Event 3:

Fuse C41-F3 for Standby Liquid Control Pump A blows causing a loss of continuity for the squib valve. The CRS should reference and enter TS LCO 3.1.7 Condition B for SLC A being inoperable. This will required SLC B for CT.

Event 4:

RHR Pump A experiences sheared shaft while operating in Suppression Pool Cooling Mode.

The BOP operator will close the test return valve restoring injection line pressure. The CRS should reference and enter TS LCO 3.5.1 Condition A and 3.6.2.3 Condition A for RHR A being inoperable.

Event 5:

Main Turbine Lube Pump degrades causing bearing header to lower to 14 psig and the TGOP will fail to auto start. The ATC will manually start the TGOP to restore bearing header pressure.

Event 6:

Main Condenser inleakage begins to increase degrading vacuum. The ATC will report the lowering vacuum trend to the CRS. The CRS will direct the ATC operator to place the Mode Switch in shutdown prior to Main Condenser Vacuum reaching 23.0 inches vacuum. The MSIVs will auto close at 8.5 inches vacuum resulting in all of the energy going to the suppression pool via SRVs.

Event 7:

When the reactor is scrammed, a hydraulic ATWS will result with approximately 10% reactor power. The CRS will enter EOP-1, RPV Control on Level 3 and direct actions from that procedure. The crew will inject SLC to lower power (CT).

Event 8 A heater string isolation will result in tripping all feed pumps. The ATC must restore the heater string and a feed pump to restore reactor water level. (CT) RCIC trips on overspeed.

HPCS injection valve will fail closed when overridden closed during terminate and prevent to lower level.

Event 9:

B21-AOVF028A, MSL A OUTBD MSIV will fail to auto close. The BOP will manually close the valve.

REV 2 Facility: __RBS____________

Scenario #: ______3___________

Scenario Source: NEW______________

Op. Test #: RB-2022-12________

Examiners: __________________

Applicants/

Operators: __________________

Initial Conditions: Mode 1, 100% Power, MOL Turnover: The System Dispatcher wants Reactor Power to be reduced to 90%.

Critical Tasks:

1. Manually Scram the Reactor prior to radiation levels reaching 9500 mr/hr in the RCIC room.
2. Emergency Depressurization before Max Safe radiation levels are reached in the third Secondary Containment area (RHR C).

Event Event Description

  • Attribute type
    • CRS
    • ATC
    • BOP 1

Lower Reactor Power to 90% with Recirc Flow.

R R

2 Recirc FCV A will slowly start to drift close.

A TS,C C

3 Trip of an Isophase Bus duct cooling fan.

A C

C 4

CNS makeup valve for the Seal Steam Generator will fail closed.

A C

C,MC 5

APRM F will fail upscale.

I I

6 Spurious initiation of the RCIC system.

A I,TS I

7 Unisolable steam leak will develop in the RCIC room. A Manual Reactor Scram is required prior to radiation levels reaching Max Safe levels in the RCIC room.

EP,CT M

M M

8 RHS-AOV63 will fail to isolate on Level 3.

E C

C,MC 9

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

REV 2 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 reaching Max Safe level.

Emergency Depressurization before Max Safe radiation levels are reached in the third Secondary Containment area (RHR C).

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.

Per EOP-3 bases for steps SC-13, 14, and 15, If radiation levels in any one of the areas listed in Table SC-2 approach their maximum safe operating value, adequate core cooling, containment integrity, safety of personnel, or continued operability of equipment required to perform EOP actions can no longer be assured. A reactor scram is initiated through entry of EOP-1 to reduce the primary system discharge into secondary containment and in anticipation of possible RPV depressurization in Step SC-17 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.

With a continued spread of radiation levels to additional areas would indicate a wide-spread problem which may pose a direct and immediate threat to secondary containment or MSL tunnel integrity, equipment located in the secondary containment or MSL tunnel, and continued safe operation of the plant eventually leading to immediate threat to the health and safety of the public.

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.

REV 2 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.

Prior to a third Secondary Containment Area exceeding Max Safe Radiation levels.

If an applicant/operator or the crew significantly deviates from, or fails to follow, procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review

REV 2 Event 1: Lower power to 90%

The Load Dispatcher has requested RBS station to lower power to 90%.

NOTE: When power reaches 95%, it will fire the next trigger.

Event 2: AUTOMATIC TRIGGER Recirc FCV A will slowly start to drift close. The ATC will take manual action to lock up the FCV. The CRS should reference enter TS LCO 3.4.1 Condition A for Recirc Loop Flow mismatch.

Event 3: MANUAL TRIGGER Isophase Bus duct cooling fan 1, GML-FN1 DUCT FAN 1, trips causing a rising bus duct temperature. The BOP operator will reference ARP-870-54 and start GML-FN1 DUCT FAN 2 IAW SOP-67, Isolated Phase Bus Duct Cooling System.

Event 4: MANUAL TRIGGER CNA-MOVS9 fails closed causing a lowering level in the Steam Seal Evaporator. The BOP operator will manually open the bypass valve to restore level IAW ARP-870-52.

Event 5: MANUAL TRIGGER APRM F fails upscale causing a Rod Block and a Half Scram. The ATC will bypass APRM F and reset the Half Scram signal.

Event 6: MANUAL TRIGGER Spurious initiation of the RCIC system. The BOP operator will manually trip RCIC IAW AOP-34, INADVERTENT INITIATION OF AN ECCS OR RCIC. The CRS should reference and enter TS LCO 3.5.3 Condition A and LCO 3.3.5.3 Condition A then B.

Event 7: MANUAL TRIGGER An unisolable leak develops in the RCIC room causing rising room temperature and radiation levels. The CRS will enter EOP-3, EMERGENCY OPERATING PROCEDURE -

SECONDARY CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL. The CRS will direct the ATC to place the Mode Switch in shutdown prior to exceeding any RCIC Max Safe Operating Value.

Event 8: AUTOMATIC TRIGGER RHS-AOV63, SPC SUCTION VALVE, fails to isolate on Level 3 signal.

Event 9: AUTOMATIC TRIGGER Radiation Levels in the RHR A room will begin approaching Max Safe Levels. The CRS will direct an Emergency Depressurization IAW EOP-3 when the RCIC room and RHR A room have exceeded Max Safe Rad Levels. The CRS will direct the BOP to open 7 ADS/SRVs IAW EOP-1.

REV 3 Facility: __RBS____________

Scenario #: ______4___________

Scenario Source: __________________

Op. Test #: __________________

Examiners: __________________

Applicants/

Operators: __________________

Initial Conditions: Mode 2, 5% Power, BOL Turnover: Plant Startup IAW GOP-1 Critical Tasks:

1. Manually raise frequency on DIV 1 diesel to close the output breaker within 15 minutes of the station blackout.
2. Restore RCIC prior to to STEAM COOLING required per EOP-1C1.

Event Event Description

  • Attribute type
    • CRS
    • ATC
    • BOP 1

ATC Operator will withdraw one control rod IAW the RMP.

R 2

The second Control Rod selected will begin to drift out when selected.

A C,TS C

C 3

Trip of DIV 1 battery room fan and failure of the standby fan to auto start.

A C

C,MC 4

Startup Feedwater level controller fails low.

A C

C,MC 5

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

A C,TS C

6 The Main Turbine Bypass Valves will fail closed.

A C

C 7

A loss of Offsite Power/Station Blackout occurs. DIV 2 and 3 Diesel Generators trip when they auto start and cannot be manually started. A small Drywell Steam leak occurs after the scram.

EP M

M M

8 DIV 1 Diesel Generator output breaker fails to close requiring operator action to power bus.

Div 1 Diesel Generator trips on overspeed approximately 1 minute after the closure of the output breaker (CT-1)

E,EC C

C,MC 9

RCIC trips on low suction pressure. Crew must stabilize pressure and restore RCIC (CT-2)

E,CT C

C 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

REV 3 Attribute Target Actual Malfunctions after EOP entry 1-2 2

Abnormal Events 2-4 5

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 raise frequency on DIV 1 diesel to close the output breaker within 15 minutes of the station blackout.

Restore RCIC prior to STEAM COOLING required per EOP-1C1.

CT Criteria This action directly leads to the restoration of multiple safety functions.

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

Safety Significance A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Fifteen minutes was selected as a threshold due to upgrade in Emergency Classification from ALERT to SAE.

Restoring RCIC prior to entering STEAM COOLING leg of EOP-1C1 ensures core cooling is by submergence which is preferrable over steam cooling as steam cooling allows RPV water level to decrease through boil-off until it drops to the Minimum Zero-Injection RPV Water Level (MZIRWL).

The MZIRWL is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude clad temperature in the uncovered portion of the core from exceeding 1800°F.

Initiating Cue Loss of power to all AC sources. DIV 1 DG running but the output breaker has failed to close.

RCIC tripped with RWL inventory lowering Performance Feedback Power is restored to the DIV 1 bus.

RCIC operating with flow at or above 600 gpm.

Success Path Emergency start and raise frequency on the DIV 1 diesel generator.

Reset RCIC trip and restore RCIC injection Measurable Performance Standard Within 15 minutes of the Loss of Offsite RCIC restored injecting into Reactor Vessel with no entry into STEAM COOLING leg of EOP-1C1

REV 3 If an applicant/operator or the crew significantly deviates from, or fails to follow, procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review

SUMMARY

Event 1:

ATC Operator will withdraw one control rod 2437 IAW the RMP.

Event 2: AUTOMATIC TRIGGER The ATC will select the first Control Rod 4029 IAW RMP. When the second Control Rod is selected, another Control Rod will begin to drift out and it will stick when the ATC begins to insert the Control Rod IAW ARP-680-07/AOP-61, Control Rod(s) Mispositioned/

Malfunctioned. The CRS should reference and enter TS LCO 3.1.3 Condition C for the Control Rod being inoperable.

NOTE: The Control Rod cannot be inserted past position 12.

Event 3: MANUAL TRIGGER HVC-FN3A, BATT RM 1A EXH FAN, for the Div 1 Battery Fan will trip and HVC-FN3D, BATT RM 1A EXH FAN will fail to start automatically. The BOP operator will manually start the standby fan IAW ARP-863-74.

Event 4: MANUAL TRIGGER Level Transmitter B21-N004A fails upscale causes a rise in RPV water level. The ATC operator will take immediate action IAW AOP-6 Condensate/Feedwater Failures and place the Feedwater Master Controller in manual.

Event 5: MANUAL TRIGGER A Bus fault occurs on EJS-SWG2A, Div 1 480 Vac Load Center tripping EJS-SWG2A/ACB38, EJS-SWGR2A SPLY BRKR. The BOP will need to start Containment Unit Cooler C and a Drywell Unit Cooler IAW AOP-13, PRIMARY CONTAINMENT CONTROL. The CRS should reference and enter TS LCO 3.8.9 Condition A for the being EJS-SWG2A inoperable.

Event 6: MANUAL TRIGGER The Operating Bypass EHC Pump will trip, and the standby pump will fail to start automatically. The Bypass Valves will fail closed causing Reactor Pressure to rise. The BOP operator will manually open MSL drains to stabilize RPV pressure.

Event 7: MANUAL TRIGGER A loss of Offsite Power/Station Blackout occurs and DIV 2/3 Diesel Generators trip when they auto start and cannot be manually started. A small Drywell Steam leak occurs after the scram resulting in Drywell D/P reaching above 1.68 psid.

Event 8: MANUAL TRIGGER DIV 1 Diesel Generator output breaker fails to close requiring operator action to power bus.

Div 1 Diesel Generator trips on overspeed approximately 1 minute after the closure of the output breaker.

NOTE: The DG will be tripped approximately 1 minute after breaker closure.

Coordinate with the lead evaluator to manually insert Division 1 DG trip after the output breaker is closed.

Event 9: MANUAL TRIGGER RCIC trips on overspeed when initially started. An operator will have to be dispatched to reset the trip and RCIC will have to be manually started and controlled.

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