ML21096A024

RB-2021-02 Draft Outlines

ML21096A024
Person / Time
Site:	River Bend
Issue date:	02/19/2021
From:	Greg Werner Operations Branch IV
To:	Entergy Operations
References
Download: ML21096A024 (42)

Text

ES-401 1

Form ES-401-1 Facility: River Bend Station RO Exam Date of Exam: 2-8-2021 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 4

4 N/A 3

3 N/A 3

20 7

2 2

1 1

1 1

1 7

3 Tier Totals 5

5 5

4 4

4 27 10

2.

Plant Systems 1

3 2

3 2

2 2

2 3

2 2

3 26 5

2 1

1 1

1 1

1 1

1 2

1 1

12 3

Tier Totals 4

3 4

3 3

3 3

4 4

3 4

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 3

2 2

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply.) Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following:

(CFR: 41.7 / 45.8)

AK2.06 Reactor Power 3.8 39 295003 (APE 3) Partial or Complete Loss of AC Power / 6 X

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER :

(CFR: 41.10 / 43.5 / 45.13)

AA2.01 Cause of partial or complete loss of A.C. power 3.4 40 295004 (APE 4) Partial or Total Loss of DC Power / 6 X

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :

(CFR: 41.8 to 41.10)

AK1.05 Loss of breaker protection 3.3 41 295005 (APE 5) Main Turbine Generator Trip /

3 X

G2.2.40 Ability to apply Technical Specifications for a system. l (CFR: 41.10 / 43.2 / 43.5 / 45.3) 3.4 42 295006 (APE 6) Scram / 1 X

Knowledge of the reasons for the following responses as they apply to SCRAM:

(CFR: 41.5 / 45.6)

AK3.02 Reactor power response 4.1 43 295016 (APE 16) Control Room Abandonment

/ 7 X

G2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12) 4.0 44 295018 (APE 18) Partial or Complete Loss of CCW / 8 X

Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

(CFR: 41.7 / 45.6)

AA1.02 System loads 3.3 45 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following:

(CFR: 41.7 / 45.8)

AK2.05 Main Steam System 3.4 46 295021 (APE 21) Loss of Shutdown Cooling /

4 X

Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING:

(CFR: 41.7 / 45.6)

AA1.06 Containment/ drywell temperature 2.8 47

ES-401 3

Form ES-401-1 295023 (APE 23) Refueling Accidents / 8 X

Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS:

(CFR: 41.10 / 43.5 / 45.13)

AA2.01 Area radiation levels 3.6 48 295024 High Drywell Pressure / 5 X

Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following:

(CFR: 41.7 / 45.8)

EK2.17 Auxiliary building isolation logic:

Plant-Specific 3.0 49 295025 (EPE 2) High Reactor Pressure / 3 X

Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE:

(CFR: 41.5 / 45.6)

EK3.09 Low-low set initiation 3.9 50 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

(CFR: 41.8 to 41.10)

EK1.02 Steam condensation 3.5 51 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 X

Ability to operate and/or monitor the following as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY) :

(CFR: 41.7 / 45.6)

EA1.03 Emergency Depressurization Mark III 3.5 52 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 295030 (EPE 7) Low Suppression Pool Water Level / 5 X

Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and the following:

(CFR: 41.7 / 45.8)

EK2.07 Downcomer/ horizontal vent submergence 3.5 53 295031 (EPE 8) Reactor Low Water Level / 2 X

Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL:

(CFR: 41.8 to 41.10)

EK1.02 Natural circulation 3.8 54 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTO POWER ABOVE APRM DOWNSCALE OR UNKNOWN :

(CFR: 41.5 / 45.6)

EK3.02 SBLC injection 4.3 55

ES-401 4

Form ES-401-1 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following:

(CFR: 41.7 / 45.8)

EK2.01 Radwaste 3.1 56 600000 (APE 24) Plant Fire On Site / 8 X

G2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

(CFR: 41.10 / 43.2 / 45.6) 4.6 57 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

(CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)

AA2.04 VARs outside capability curve 3.6 58 K/A Category Totals:

3 4

4 3

3 3

Group Point Total:

20

ES-401 5

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum / 3 295007 (APE 7) High Reactor Pressure / 3 X

G2.1.30 Ability to locate and operate components, including local controls.

(CFR: 41.7 / 45.7) 4.4 59 295008 (APE 8) High Reactor Water Level / 2 295009 (APE 9) Low Reactor Water Level / 2 295010 (APE 10) High Drywell Pressure / 5 X

Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE :

(CFR: 41.7 / 45.6)

AA1.02 Drywell floor and equipment drain sumps 3.6 60 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 295012 (APE 12) High Drywell Temperature /

5 X

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE :

(CFR: 41.5 / 45.6)

AK3.01 Increased drywell cooling 3.5 61 295013 (APE 13) High Suppression Pool Temperature. / 5 X

Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL TEMPERATURE:

(CFR: 41.8 to 41.10)

AK1.01 Pool stratification 2.5 62 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) Abnormal Offsite Release Rate / 9 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 295029 (EPE 6) High Suppression Pool Water Level / 5 X

Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL:

(CFR: 41.10 / 43.5 / 45.13)

EA2.02 Reactor pressure 3.5 63 295032 (EPE 9) High Secondary Containment Area Temperature / 5 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9

ES-401 6

Form ES-401-1 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 X

Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

(CFR: 41.8 to 41.10)

EK1.01 Secondary containment integrity 3.9 64 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 X

Knowledge of the interrelations between HIGH CONTAINMENT HYDROGEN CONCENTRATIONS the following:

(CFR: 41.7 / 45.8)

EK2.04 Drywell recirculating fan 2.7 65 K/A Category Point Totals:

2 1

1 1

1 1

Group Point Total:

7

ES-401 7

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 1 (RO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode X

Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.03 Valve closures 3.2 1

205000 (SF4 SCS) Shutdown Cooling X

Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE):

(CFR: 41.7 / 45.7)

K6.02 D.C. electrical power 2.7 2

205000 (SF4 SCS) Shutdown Cooling X

Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

A4.06 Reactor water level 3.8 3

206000 (SF2, SF4 HPCIS)

High-Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X G2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

(CFR: 41.10 / 43.2 / 45.6) 4.5 4

209002 (SF2, SF4 HPCS)

High-Pressure Core Spray X

Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE CORE SPRAY SYSTEM (HPCS):

(CFR: 41.5 / 45.3)

K5.02 Heat removal (transfer) mechanism:

BWR-5,6 2.6 5

209002 (SF2, SF4 HPCS)

High-Pressure Core Spray X

Ability to predict and/or monitor changes in parameters associated with operating the HIGH-PRESSURE CORE SPRAY SYSTEM (HPCS) controls including:

(CFR: 41.5 / 45.5)

A1.01 HPCS flow: BWR-5,6 3.6 6

211000 (SF1 SLCS) Standby Liquid Control X

Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY LIQUID CONTROL SYSTEM:

(CFR: 41.7 / 45.7)

K6.03 A.C Power 3.2 7

ES-401 8

Form ES-401-1 211000 (SF1 SLCS) Standby Liquid Control X G2.1.20 Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12) 4.6 8

212000 (SF7 RPS) Reactor Protection X

Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM will have on following:

(CFR: 41.7 / 45.4)

K3.11 Recirculation system 3.0 9

212000 (SF7 RPS) Reactor Protection X

Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including:

(CFR: 41.7 / 45.7)

A3.06 Main turbine trip 4.2 10 215003 (SF7 IRM)

Intermediate-Range Monitor X

Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.07 Reactor vessel 3.0 11 215004 (SF7 SRMS) Source-Range Monitor X

Knowledge of SOURCE RANGE MONITOR (SRM) SYSTEM design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7)

K4.01 Rod withdrawal blocks 3.7 12 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including:

(CFR: 41.7 / 45.7)

A3.04 Annunciator and alarm signals 3.2 13 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following:

(CFR: 41.7 / 45.4)

K3.01 Reactor water level 3.7 14 218000 (SF3 ADS) Automatic Depressurization X

Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including:

(CFR: 41.7 / 45.7)

A3.01 ADS valve operation 4.2 15

ES-401 9

Form ES-401-1 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including:

(CFR: 41.5 / 45.5)

A1.01 System indicating lights and alarms 3.5 16 239002 (SF3 SRV) Safety Relief Valves X

Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.01 Stuck open vacuum breakers 3.0 17 259002 (SF2 RWLCS) Reactor Water Level Control X

Knowledge of the physical connections and/or cause effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.03 Reactor Water Level 3.8 18 261000 (SF9 SGTS) Standby Gas Treatment X

Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.07 A.C. electrical failure 2.7 19 262001 (SF6 AC) AC Electrical Distribution X

Knowledge of the operational implications of the following concepts as they apply to A.C. ELECTRICAL DISTRIBUTION:

(CFR: 41.5 / 45.3)

K5.01 Principle involved with paralleling two A.C. sources 3.1 20 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following:

(CFR: 41.7 / 45.4)

K3.01 Water Level COntrol: Plant-Specific 3.1 21 263000 (SF6 DC) DC Electrical Distribution X

Knowledge of electrical power supplies to the following:

(CFR: 41.7)

K2.01 Major D.C. loads 3.1 22

ES-401 10 Form ES-401-1 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

Knowledge of the physical connections and/or cause effect relationships between EMERGENCY GENERATORS (DIESEL/JET) and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.05 Emergency generator fuel oil supply system 3.2 23 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X G2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) 3.8 24 300000 (SF8 IA) Instrument Air X

Knowledge of electrical power supplies to the following:

(CFR: 41.7)

K2.01 Instrument air compressor 2.8 25 400000 (SF8 CCS) Component Cooling Water X

Knowledge of CCWS design feature(s) and or interlocks which provide for the following:

(CFR: 41.7)

K4.01 Automatic start of standby pump 3.4 26 510000 (SF4 SWS*) Service Water (Normal and Emergency)

K/A Category Point Totals:

3 2

3 2

2 2

2 3

2 2

3 Group Point Total:

26

ES-401 11 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 2 (RO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism X

Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD AND DRIVE MECHANISM:

(CFR: 41.5 / 45.3)

K5.07 How control rod movements affect core reactivity 3.3 27 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 X

Knowledge of RECIRCULATION System design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7)

K4.05 Seal cooling 2.9 28 202002 (SF1 RSCTL) Recirculation Flow Control X

Ability to monitor automatic operations of the RECIRCULATION FLOW CONTROL SYSTEM including:

(CFR: 41.7 / 45.7)

A3.01 Flow control valve operation: BWR-5,6 3.6 29 204000 (SF2 RWCU) Reactor Water Cleanup X

Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER CLEANUP SYSTEM:

(CFR: 41.7 / 45.7)

K6.08 PCIS/NSSSS 3.5 30 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

Ability to predict and/or monitor changes in parameters associated with operating the NUCLEAR BOILER INSTRUMENTATION controls including:

(CFR: 41.5 / 45.5)

A1.03 Surveillance testing 2.9 31 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode

ES-401 12 Form ES-401-1 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

Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.01 Interlock failure 3.3 32 239001 (SF3, SF4 MRSS) Main and Reheat Steam X

Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

A4.04 System pressure 3.8 33 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control X

Ability to monitor automatic operations of the MSIV LEAKAGE CONTROL SYSTEM including:

(CFR: 41.7 / 45.7)

A3.09 Reactor building temperature: BWR-4,5,6(P-Spec) 2.6 34 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating X

Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following:

(CFR: 41.7 / 45.4)

K3.07 Main stop/throttle valves 3.3 35 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater X

Knowledge of electrical power supplies to the following:

(CFR: 41.7)

K2.01 Reactor feedwater pump(s): Motor-Driven-Only 3.3 36 268000 (SF9 RW) Radwaste

ES-401 13 Form ES-401-1 271000 (SF9 OG) Offgas X

Knowledge of the physical connections and/or cause effect relationships between OFFGAS SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.04 Condensate system 2.7 37 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 X Ability to (a) predict the impacts of the following on the REACTOR VESSEL INTERNALS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.02 Over pressurization transient 3.6 38 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

1 1

1 1

1 1

1 1

2 1

1 Group Point Total:

12

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: River Bend Station (RO Exam)

Date of Exam: 2/8/2021 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.13 Knowledge of facility requirements for controlling vital/controlled access.

(CFR: 41.10 / 43.5 / 45.9 / 45.10) 2.5 66 2.1.21 Ability to verify the controlled procedure copy.

(CFR: 41.10 / 45.10 / 45.13) 3.5 67 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

(CFR: 41.10 / 45.12) 3.4 68 Subtotal 3

2. Equipment Control 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

(CFR: 41.10 / 43.5 / 45.13) 2.6 69 2.2.43 Knowledge of the process used to track inoperable alarms.

(CFR: 41.10 / 43.5 / 45.13) 3.0 70 2.2.35 Ability to determine Technical Specification Mode of Operation.

(CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.6 71 Subtotal 3

3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10) 3.2 72 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(CFR: 41.12 / 45.10) 3.5 73 2.3.

Subtotal 2

4. Emergency Procedures/Plan 2.4.3 Ability to identify post-accident instrumentation.

(CFR: 41.6 / 45.4) 3.7 74 2.4.1 Knowledge of EOP entry conditions and immediate action steps.

(CFR: 41.10 / 43.5 / 45.13) 4.6 75 2.4.

Subtotal 2

Tier 3 Point Total 10 10 7

ES-401 1

Form ES-401-1 Facility: Riverbend Station (SRO Exam)

Date of Exam: 2/8/2021 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

N/A N/A 20 3

4 7

2 7

2 1

3 Tier Totals 27 5

5 10

2.

Plant Systems 1

26 3

2 5

2 12 1

2 3

Tier Totals 38 4

4 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

2 2

1 2

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply.) Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295003 (APE 3) Partial or Complete Loss of AC Power / 6 295004 (APE 4) Partial or Total Loss of DC Power / 6 295005 (APE 5) Main Turbine Generator Trip /

3 295006 (APE 6) Scram / 1 295016 (APE 16) Control Room Abandonment

/ 7 295018 (APE 18) Partial or Complete Loss of CCW / 8 X

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

(CFR: 41.10 / 43.5 / 45.13)

AA2.04 System flow 2.9 76 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 295021 (APE 21) Loss of Shutdown Cooling /

4 295023 (APE 23) Refueling Accidents / 8 295024 High Drywell Pressure / 5 X

2.1.32 Ability to explain and apply system limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 4.0 77 295025 (EPE 2) High Reactor Pressure / 3 X

Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:

(CFR: 41.10 / 43.5 / 45.13)

EA2.05 Decay heat generation 3.6 78 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

2.2.22 Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2) 4.7 79 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 295030 (EPE 7) Low Suppression Pool Water Level / 5 295031 (EPE 8) Reactor Low Water Level / 2 X

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6) 4.4 80

ES-401 3

Form ES-401-1 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 600000 (APE 24) Plant Fire On Site / 8 X

Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE:

AA2.07 Whether malfunction is due to common-mode electrical failures 3.0 81 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12) 4.3 82 K/A Category Totals:

3 4

Group Point Total:

7

ES-401 4

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum / 3 295007 (APE 7) High Reactor Pressure / 3 295008 (APE 8) High Reactor Water Level / 2 X

Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL:

(CFR: 41.10 / 43.5 / 45.13)

AA2.05 Swell 3.1 83 295009 (APE 9) Low Reactor Water Level / 2 295010 (APE 10) High Drywell Pressure / 5 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 295012 (APE 12) High Drywell Temperature /

5 295013 (APE 13) High Suppression Pool Temperature. / 5 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) Abnormal Offsite Release Rate / 9 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 X

2.1.7 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 84 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 295029 (EPE 6) High Suppression Pool Water Level / 5 295032 (EPE 9) High Secondary Containment Area Temperature / 5 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 X

Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:

(CFR: 41.10 / 43.5 / 45.13)

EA2.01 Ventilation radiation levels 3.2 85 295035 (EPE 12) Secondary Containment High Differential Pressure / 5

ES-401 5

Form ES-401-1 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals:

2 1

Group Point Total:

3

ES-401 6

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 1 (SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode 205000 (SF4 SCS) Shutdown Cooling 206000 (SF2, SF4 HPCIS)

High-Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

Ability to (a) predict the impacts of the following on the LOW-PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.06 Inadequate system flow 3.2 86 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Control 212000 (SF7 RPS) Reactor Protection 215003 (SF7 IRM)

Intermediate-Range Monitor X

Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.05 Faulty or erratic operation of detectors/system 3.5 87 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

(CFR: 41.10 / 43.5 / 45.13) 4.5 88 239002 (SF3 SRV) Safety Relief Valves

ES-401 7

Form ES-401-1 259002 (SF2 RWLCS) Reactor Water Level Control X

Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.02 Loss of any number of reactor feedwater flow inputs 3.4 89 261000 (SF9 SGTS) Standby Gas Treatment X 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) 4.2 90 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG 300000 (SF8 IA) Instrument Air 400000 (SF8 CCS) Component Cooling Water 510000 (SF4 SWS*) Service Water (Normal and Emergency)

K/A Category Point Totals:

3 2

Group Point Total:

5

ES-401 8

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 2 (SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information X 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13) 4.2 91 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries 226001 (SF5 RHR CSS) RHR/LPCI: Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring X 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

l (CFR: 41.10 / 43.5 / 45.13) 4.9 92

ES-401 9

Form ES-401-1 286000 (SF8 FPS) Fire Protection X

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 conditions or operations:

(CFR: 41.5 / 45.6)

A2.01 System logic failure: Plant-Specific 2.9 93 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

1 2

Group Point Total:

3

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: River Bend Station (SRO Exam)

Date of Exam: 2/8/2021 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, operations memos, etc.

(CFR: 41.10 / 45.12) 3.4 94 2.1.3 Knowledge of shift or short-term relief turnover practices. l (CFR: 41.10 / 45.13) 3.9 95 2.1.

Subtotal 2

2. Equipment Control 2.2.11 Knowledge of the process for controlling temporary design changes.

(CFR: 41.10 / 43.3 / 45.13) 3.3 96 2.2.37 2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12) 4.6 97 2.2.

Subtotal 2

3. Radiation Control 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.12 / 43.4 / 45.9) 3.1 98 2.3.

2.3.

Subtotal 1

4. Emergency Procedures/Plan 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

(CFR: 41.10 / 43.5 / 45.11) 4.1 99 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation. l (CFR: 41.10 / 43.5 / 45.11) 4.5 100 2.4.

Subtotal 2

Tier 3 Point Total 7

7

ES-401 Record of Rejected K/As Form ES-401-4 3/10/2020 Rev 2 Tier /

Group Randomly Selected K/A Reason for Rejection SRO 1/1 295026 2.1.31 Chief examiner randomly selected new K/A due to ability to write an SRO only question. New randomly selected K/A 2.2.22 SRO 1/1 600000 AA2.09 Chief examiner randomly selected a new K/A due to ability to write an SRO only question. New randomly selected K/A AA2.07 SRO 2/1 259002 2.2.44 Chief examiner randomly selected a new K/A due to ability to write an SRO only question. New randomly selected K/A 2.2.25 SRO Generic 2.1.18 Chief examiner randomly selected a new K/A due to ability to write an SRO only question. New randomly selected K/A 2.1.15 SRO Generic 2.2.41 Chief examiner randomly selected a new K/A due to ability to write an SRO only question. New randomly selected K/A 2.2.37 RO Generic 2.2.25 Chief examiner randomly selected a new K/A due to overlap with the SRO written exam. New randomly selected K/A 2.2.43 Rev 1 RO 1/1 295030 EK3.02 EK3.02 is not applicable to RBS. Chief randomly selected New K/A to be 295030 EK 2.07 Downcomer/ horizontal vent submergence RO 2/1 211000 K6.05 K6.05 is not applicable to RBS. Chief randomly selected 211000 K6.03 A.C. Power RO 2/1 239002 A2.10 A2.10 words on ES-401-1 do not match NUREG-1123. Cut and paste error Chief randomly selected 239002 A2.01 Stuck open Vacuum Breakers RO 2/1 259002 K1.11 K1.11 words on ES-401-1 do not match NUREG-1123. Cut and paste error corrected to the correct wording. No change in K/A RO 2/1 Missing K/A Chief Randomly selected 205000 (SF4 SCS) Shutdown Cooling, A4.06 Reactor Water Level SRO 2/1 261000 A2.15 A2.15 is not applicable to RBS. Chief randomly selected 261000 A2.12 High fuel pool ventilation radiation: Plant-Specific SRO 2/2 268000 A2.02 Request new K/A due to ability to write an SRO only question. Chief Examiner randomly selected 286000 A2.01 System logic failure: Plant-Specific Rev 2 SRO 2/1 261000 A2.12 A2.12 is not applicable to RBS. There is no direct tie between fuel pool ventilation and standby gas. Chief Examiner replaced the A2.12 K/A with Generic 2.2.25 for the same topic.

ES-401 Record of Rejected K/As Form ES-401-4 3/10/2020 Rev 2 SRO 2/1 259002 2.2.25 Request new KA due to not having any reactor water level control system ties to RBS tech specs. Chief Examiner replaced the Generic 2.2.25 K/A with A2.02 of the same topic.

Rev 3 RO 1/1 295001 AK2.05 Request new K/A due to LPCI logic interrelationship is N/A to RBS.

Chief Examiner replaced AK2.05 with AK2.06 of the same topic.

RO 1/1 295019 AK2.11 Request new K/A due to difficulty of writing question on specific interrelationship. Chief Examiner replaced AK2.11 with AK2.05 of the same topic RO 1/1 295027 EA1.01 Request new K/A due to containment spray topic is N/A to RBS. Chief Examiner replaced Request EA1.01 with EA1.03 of the same topic.

RO 2/1 259002 K1.11 Request new K/A due to drywell pressure: FWCI/HPCI topic is N/A to RBS. Chief Examiner replaced Request K1.11 with K1.03 of the same topic.

RO 2/1 262002 K3.03 Request new K/A due to RFPT topic is N/A to RBS. Chief Examiner replaced Request K3.03 with K3.01 of the same topic.

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

RBS Date of Examination:

2/8/2021 Examination Level: RO SRO Operating Test Number:

2021-02 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R,N Loss of Feedwater Heating event. Applicant must utilize AOP-7 and AOP-24 to determine placement on the Power to Flow Map.

K/A G2.1.43 Conduct of Operations R,N Determine drain time using a PID and OSP-33.

K/A G2.1.25 Equipment Control R,N Using an RPS print, determine the effect of the pulling two fuses on the system and lights.

K/A 2.2.15 Radiation Control R,N The operator will evaluate a condition involving radiological conditions and determine actions required to administratively control the dose received by determining total stay time with current exposure. Determines approval requirements for exceeding the limit.

K/A 2.3.4 Emergency Plan N/A NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

RBS Date of Examination:

2/8/2021 Examination Level: RO SRO Operating Test Number:

2021-02 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R,N Loss of Feedwater Heating event. Applicant must utilize AOP-7 and AOP-24 to determine placement on the Power to Flow Map. CRS must actions taken IAW AOP-24.

K/A G2.1.43 Conduct of Operations R,N Determine drain time using a PID and OSP-33.

CRS must determine LCO status IAW Tech Specs.

K/A G2.1.25 Equipment Control R,N Using an RPS print, determine the effect of the pulling two fuses on the system and lights.

Determine reportability of a full RPS actuation at power.

K/A 2.2.15 Radiation Control R,N Evaluate the impact of a failed Radiation Monitor IAW Tech Specs.

K/A 2.3.13 Emergency Plan R,N Make an EAL call based given Plant Conditions.

K/A 2.4.29 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

RBS Date of Examination:

2/8/2021 Exam Level: RO SRO-I SRO-U Operating Test Number:

2021-02 Control Room Systems:

• 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function

a. RB-2021-02-S1 Transferring Recirc Pumps from Fast Speed to Slow Speed N,S 1

b. RB-2021-02-S2 RCIC Slow Roll Startup A,EN,N,S 2

c. RB-2021-02-S8 Alternating RPCCW Pumps (SRO-U)

A,E,N,S 8

d. RB-2021-02-S4 Alternate Decay Heat Removal Shutdown-Configuration 1 (SRO-U)

A,L,N,S 3

e. RB-2021-02-S5 Rod Withdrawal Limiter Functional Test M,S 7

f. RB-2021-02-S6 Transfer NPS-SWG1A from normal to preferred source A.L,N,S 6

g. RB-2021-02-S7 Adjusting Reactor Pressure using Pressure Regulator Set A,N,S 4

h. RB-2021-02-S3 Start Containment Low Volume Purge (RO ONLY)

D,S 9

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. RB-2021-02-P1 Transferring BYS-INV04 from Normal Operation to Maintenance Bypass (SRO-U)

N 6

j. RB-2021-02-P2 Startup of the Hydrogen Purge System (SRO-U)

D,E,R 5

k. RB-2021-02-P3 Injection into RPV with Condensate via RHR C IAW Encl. 6 (SRO-U)

E,EN, L,N,R 2

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 8/ 4 1/ 1/ 1 1/ 1/ 1 (control room system) 1/ 1/ 1 2/ 2/ 1 3/ 3/ 2 (randomly selected) 1/ 1/ 1

ES-301 Transient and Event Checklist Form ES-301-5 Facility: RBS Date of Exam:2/8/2021 Operating Test No.: RB-2021-02 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 0

0 0

1 1 0 NOR 0

0 0

1 1 1 I/C 6

3 3

4 4 2 MAJ 1

1 1

2 2 1 TS 2

0 0

0 2 2 RO SRO-I SRO-U RX 0

0 0

1 1 0 NOR 0

0 0

1 1 1 I/C 5

3 3

4 4 2 MAJ 1

1 1

2 2 1 TS 2

0 0

0 2 2 RO SRO-I SRO-U RX 0

0 0

1 1 0 NOR 0

0 0

1 1 1 I/C 6

3 3

4 4 2 MAJ 1

1 1

2 2 1 TS 3

0 0

0 2 2 RO SRO-I SRO-U RX 0

0 0

1 1 0 NOR 0

0 0

1 1 1 I/C 6

3 3

4 4 2 MAJ 1

1 1

2 2 1 TS 3

0 0

0 2 2 Instructions:

1.

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one-for-one basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

Facility: River Bend Nuclear Station Scenario No.: 1 Op-Test No.: RB-2021-02 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100% Power.

Turnover: N/A Event No.

Malf. No.

Event Type Event Description 1

FWS001A RCS015A I (ATC/CRS)

TS (CRS)

A (ALL)

Trip of Feedwater Pump A, and FCV A fails to runback. ATC should take manual control of FCV A and match Recirc loop flows.

Reference/entry into AOP-6 and AOP-24.

CRS enters TS 3.4.1 2

CRD001A C(BOP/CRS)

A (BOP/CRS)

CRD Pump A will experience high amps and trip on overcurrent. The BOP will start the standby CRD Pump IAW SOP-2.

CRDM3617 (DRIFT)

C (ALL)

TS (CRS)

A(ALL)

Control Rod 36-17 will drift out after the start of the CRD pump. ATC will provide a continuous insert signal to maintain the control rod full in. The BOP will lower drive water pressure IAW ARP H1-P680-07 to maintain the Control Rod full in.

Reference/entry into AOP-61 CRS enters TS 3.1.3 3

ED002B M (ALL)

Loss of NPS-SWG1B. This results in a loss of ALL Feedwater Pumps. Reactor water level will lower causing a Reactor Scram.

EOP-1, RPV Control AOP-1 and AOP-2 4

HPCS003 HPCS001 RCIC001 C (BOP)

HPCS will fail to start and RCIC will start but trip on Overspeed. The BOP must coordinate with the Reactor Building operator and reset the trip IAW OSP-53 attachment 40 to restore RPV water level.

CT-1 5

MGEN003 C (ATC/CRS)

The Main Turbine Generator will not trip on reverse power.

Must be tripped manually IAW AOP-2 within 20 minutes.

CT-2 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Reason for Revision:

Quantitative Attributes Table Attribute E3-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2

RCIC trip on Overspeed Main Generator fails to trip on reverse power Abnormal Events 2-4 3

Trip of Feedwater Pump A, and FCV A fails to runback.

CRD Pump A will experience high amps and trip on overcurrent.

Control Rod 36-17 will begin drifting out Major Transients 1-2 1

Loss of NPS-SWG1B requires entry into EOP-1 when level reaches Level 3.

EOP entries requiring substantive action 1-2 1

EOP-1, RPV control EOP contingencies requiring substantive action 1

1 EOP-1, Alternate Level Control Preidentified critical tasks 2-3 2

RPV level must be restored prior to reaching -162 inches.

Main Generator must be tripped within 20 minutes of turbine trip.

SCENARIO ACTIVITIES:

Initial Conditions 100% Power. Normal operating conditions.

Inoperable Equipment: None Turnover: N/A Event 1 - (Triggered by Lead Examiner)

Trip of Feedwater Pump A, and FCV A fails to runback. ATC should take manual control of FCV A and match Recirc loop flows. Reference/entry into AOP-6, Condensate/Feedwater Failures and AOP-24, Thermal Hydraulics Stability Controls. CRS should reference TS 3.4.1, Recirculation Loops Operating.

Event 2 - (Triggered by Lead Examiner)

CRD Pump A will experience high amps and trip on overcurrent. The BOP will start the standby CRD Pump IAW SOP-2. Control Rod 36-17 will drift out after the start of the CRD pump. ATC will provide a continuous insert signal to maintain the control rod full in. The BOP will lower drive water pressure IAW ARP H1-P680-07 to maintain the Control Rod full in.

Reference/entry into AOP-61, Control Rod(s) Mispositioned/Malfunction.

CRS should reference TS 3.1.3, Control Rod OPERABILITY.

Event 3 - (Triggered by Lead Examiner)

Loss of NPS-SWG1B. This results in a loss of ALL Feedwater Pumps. Reactor water level will lower causing a Reactor Scram.

EOP-1, RPV Control AOP-1, Reactor Scram and AOP-2, Main Turbine and Generator Trips.

Event 4 - (Initial Setup - Automatic)

HPCS will fail to start and RCIC will start but trip on Overspeed. The BOP must coordinate with the Reactor Building operator and reset the trip IAW OSP-53, Emergency and Transient Response Support Procedure attachment 40 to restore RPV water level.

CT-1 Event 5 - (Initial Setup - Automatic)

The Main Turbine Generator will not trip on reverse power. Must be tripped manually IAW AOP-2, Main Turbine and Generator Trips within 20 minutes.

CT-2

CT-1 CT-2 Critical Task Main Generator must be tripped within 20 minutes of turbine trip IAW AOP-2.

Reset RCIC trip and restore RPV level before reaching -162 inches.

EVENT 4

5 Safety Significanc e

Excessive motoring will cause major turbine blade damage from overheating.

If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes less than two-thirds of the core height.

Cueing Main Generator output breakers still closed after the turbine trp.

RCIC trip after system start.

• If an 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 (NUREG 1021, Appendix D). An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in significant plant degradation or significantly alters a mitigation strategy.

Facility: River Bend Nuclear Station Scenario No.: 3 Op-Test No.: RB-2021-02 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100% Power.

Turnover: N/A Event No.

Malf. No.

Event Type Event Description 1

I (BOP/CRS)

TS (CRS)

A (ALL)

Spurious initiation of RCIC due to failed Level trip units.

The BOP will trip RCIC IAW AOP-34.

CRS enters TS 3.3.5.3 and 3.5.3 2

C (BOP)

TS (CRS)

A(ALL)

RCIC steam leak occurs in the RCIC room. E51-MOVF063 closes but E51-MOVF064 fails to close. The BOP will close the E51-MOVF064 valve.

CRS enters TS 3.3.6.1 3

I (ATC/CRS)

TS (CRS)

A (ALL)

B21-N004A fails UPSCALE causing actual RPV to lower due the increase in Feedwater flow. The ATC should place the master controller in manual IAW AOP-6. The ATC should swap from A to B channel for RPV level signal.

CRS enters TRM 3.3.7.3 4

C (ATC/CRS)

A (ALL)

Loss of extraction steam to A first point will cause a reduction of Feedwater heating. The ATC will lower power to 90% IAW AOP-7.

5 M (ALL)

Drywell steam leak will occur. CRS will direct scramming the Reactor when Drywell D/P reaches 0.8 psid. After the scram E22-S004 will have a bus fault and the Feedwater Pumps will trip due to loss of suction pressure.

6 C (BOP)

The MSIVs will fail to close when RPV level reaches Level

1. Must manually isolated before cooldown exceeds 100°F/hr.

CT-1 7

C(ATC)

Manually open low pressure ECCS injection valves prior to entering the SAPs.

CT-2 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Reason for Revision:

Quantitative Attributes Table Attribute E3-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2

MSIVs fail to close on isolation signal Low pressure ECCS systems injection valves fail to open automatically.

Abnormal Events 2-4 4

RCIC instrumentation failure RCIC isolation valve fails to close RPV level transmitter fails Loss of Extraction steam Major Transients 1-2 1

Steam leak in the Drywell EOP entries requiring substantive action 1-2 1

EOP-1, RPV control EOP contingencies requiring substantive action 1

1 Emergency RPV depressurization Preidentified critical tasks 2-3 2

Manually isolate MSIVs before cooldown exceeds 100°F/hr.

Emergency depressurize RPV and manually open low pressure ECCS injection valves to restore RPV level prior to entering the SAPs.

SCENARIO ACTIVITIES:

Initial Conditions 100% Power. Normal operating conditions.

Inoperable Equipment: None Turnover: N/A Event 1 - (Triggered by Lead Examiner)

Spurious initiation of RCIC due failed Level trip units. The BOP will trip RCIC IAW AOP-34.

CRS enters TS 3.3.5.3 and 3.5.3 Event 2 - (Triggered by Lead Examiner)

RCIC steam leak occurs in the RCIC room. E51-MOVF063 closes but E51-MOVF064 fails to close. The BOP will close the E51-MOVF064 valve.

CRS enters TS 3.3.6.1 Event 3 - (Triggered by Lead Examiner)

B21-N004A fails UPSCALE causing actual RPV to lower due the increase in Feedwater flow. The ATC should place the master controller in manual IAW AOP-6. The ATC should swap from A to B channel for RPV level signal.

CRS enters TRM 3.3.7.3 Event 4 - (Initial Setup - Automatic)

Loss of extraction steam to A first point will cause a reduction of Feedwater heating. The ATC will lower power to 90% IAW AOP-7, Loss of Feedwater Heating Event.

Event 5,6,7 - (Initial Setup - Automatic)

Drywell steam leak will occur. CRS will direct scramming the Reactor when Drywell D/P reaches 0.8 psid. After the scram E22-S004 will have a bus fault, the Feedwater Pumps will trip due to loss of suction pressure, and all low pressure ECCS systems injection valves will not auto open. The MSIVs will fail to close will RPV level reaches Level 1. Must manually isolated before cooldown exceeds 100°F/hr. Restore RPV water level prior to reaching -162 inches.

CT-1 CT-2 Critical Task Manually isolate MSIVs before cooldown exceeds 100°F/hr.

Emergency depressurize RPV and manually open low pressure ECCS injection valves to restore RPV level prior to entering the SAPs.

EVENT 5

5 Safety Significance The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a non-isolable leak or loss of coolant accident If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes less than two-thirds of the core height.

Cueing Step RP-4 in EOP-1 Step ALC-12 in EOP-1

• If an 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 (NUREG 1021, Appendix D). An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in significant plant degradation or significantly alters a mitigation strategy.

Facility: River Bend Nuclear Station Scenario No.: 4 Op-Test No.: 2021-NRC Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 4% Power.

Turnover:

Event No.

Malf. No.

Event Type Event Description 1

I (ATC/CRS)

A (ALL)

IRM A fails downscale causing a rod block. ATC will bypass the IRM.

2 I (ATC/CRS)

TS (CRS)

A(ALL)

IRM C fails upscale a rod block/half scram. ATC will unbypass IRM A and bypass IRM C. Reset the half scram.

CRS enters TS 3.3.1.1 3

C (BOP/CRS)

TS (CRS)

A (ALL)

Suppression Pool leak into the crescent area via RHR A suction line. BOP must isolate the leak by closing RHR A suction valve. Restore Suppression pool level with HPCS IAW Encl.30 CRS enters TS 3.6.2.2 AOP-13 EOP-2, Primary Containment Control CT-1 4

C (BOP/CRS)

TS (CRS)

A (ALL)

MAIN STEAM LINE DIV 1 HI-HI RAD OR INOP and MAIN STEAM LINE DIV 3 HI-HI RAD OR INOP alarms come in at the H13-P601 panel and B33-F019 will fail to isolate requiring manual action from the BOP.

CRS enters TS 3.6.1.3 5

M (ALL)

Turbine bypass will fail closed and steam line drains will not open. RPV pressure will rise requiring a manual scram per AOP-17.

EOP-1, RPV Control CT-2 6

C (ATC)

RPS will fail to actuate automatically.

CT-1 7

C (BOP)

A spurious DIV 3 initiation will occur after the scram. The BOP will need to terminate and prevent the HPCS pump IAW AOP-34.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Reason for Revision:

Quantitative Attributes Table Attribute E3-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2

RPS signals fail to insert control rods Spurious DIV 3 initiation Abnormal Events 2-4 4

IRM failures Suppression Pool leak High Steam Line Radiation BPV failed closed Major Transients 1-2 1

RPV pressure transient with no BPV/drains EOP entries requiring substantive action 1-2 2

EOP-1, RPV control EOP-2, Containment Control EOP contingencies requiring substantive action 1

1 N/A Preidentified critical tasks 2-3 2

Insert a manual Reactor scram before RPV pressure exceeds 1100 psig.

Close RHR A suction valve before Suppression Pool level reaches 15ft, 5in.

SCENARIO ACTIVITIES:

Initial Conditions 4% Power. Startup Inoperable Equipment: None Turnover: N/A Event 1 - (Triggered by Lead Examiner)

IRM A fails downscale causing a rod block. ATC will bypass the IRM.

Event 2 - (Triggered by Lead Examiner)

IRM C fails upscale a rod block/half scram. ATC will unbypass IRM A and bypass IRM C.

Reset the half scram.

CRS enters TS 3.3.1.1 Event 3 - (Triggered by Lead Examiner)

Suppression Pool leak into the crescent area via RHR A suction line. BOP must isolate the leak by closing RHR A suction valve. Restore Suppression pool level with HPCS IAW Encl.30 CRS reference 3.6.2.2 AOP-13, Primary Containment Control EOP-2, Primary Containment Control Event 4 - (Triggered by Lead Examiner)

MAIN STEAM LINE DIV 1 HI-HI RAD OR INOP and MAIN STEAM LINE DIV 3 HI-HI RAD OR INOP alarms come in at the H13-P601 panel and B33-F019 will fail to isolate requiring manual action from the BOP.

CRS enters TS 3.6.1.3 Event 5 (Triggered by Lead Examiner)

Turbine bypass will fail closed and steam line drains will not open. RPV pressure will rise requiring a manual scram per AOP-17. RPS will fail to actuate automatically.

EOP-1, RPV Control

CT-1 CT-2 Critical Task Close RHR A suction valve before Suppression Pool level reaches 15ft, 5in.

Insert a manual Reactor scram before RPV pressure exceeds 1100 psig.

EVENT 3

5 Safety Significanc e

When suppression pool level decreases to two feet above the top of the Mark III horizontal vents, any further drop in water level could result in direct exposure of the drywell atmosphere to the containment airspace thus compromising the pressure suppression function of the containment.

The capability of inserting the control rods provide assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated.

Cueing Step SPL-2 in EOP-2 Step RC-2 in EOP-1

• If an 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 (NUREG 1021, Appendix D). An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in significant plant degradation or significantly alters a mitigation strategy.