ML18239A242

7-RBS 2018-07 Final Outlines

ML18239A242
Person / Time
Site:	River Bend
Issue date:	08/03/2018
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML18239A242 (47)

Text

ES-401 1

Form ES-401-1 Rev 1 Facility: River Bend Station Date of Exam: July 2018 Tier Group RO K/A Category Points SRO-Only Points K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 4

3 N/A 3

4 N/A 3

20 7

2 2

1 1

1 1

1 7

3 Tier Totals 5

5 4

4 5

4 27 10

2.

Plant Systems 1

2 2

2 4

2 2

3 3

2 3

1 26 5

2 1

1 1

1 1

1 1

1 1

1 2

12 3

Tier Totals 3

3 3

5 3

3 4

4 3

4 3

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 2

2 3

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.

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

RB-2018-07 E/APE # / Name / Safety Function 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 / 1 & 4 X

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

AA2.02 Neutron monitoring 3.1 39 295003 (APE 3) Partial or Complete Loss of AC Power / 6 X

2.1.31 Ability to locate control room switches, controls, and indications, and to l determine that they correctly reflect the desired plant lineup.

4.6 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:

AK1.06 Prevention of inadvertent system(s) actuation upon restoration of D.C. power 3.3 41 295005 (APE 5) Main Turbine Generator Trip / 3 X

Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following:

AK2.03 Recirculation system 3.2 42 295006 (APE 6) Scram / 1 X

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

AK3.04 Reactor water level setpoint setdown:

Plant-Specific 3.1 43 295016 (APE 16) Control Room Abandonment /

7 X

Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT:

AA1.08 Reactor pressure 4.0 44 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:

AA2.03 Cause for partial or complete loss 3.2 45 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

4.2 46 295021 (APE 21) Loss of Shutdown Cooling / 4 X

Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING:

AK1.01 Decay heat 3.6 47 295023 (APE 23) Refueling Accidents / 8 X

Knowledge of the interrelations between REFUELING ACCIDENTS and the following:

AK2.05 Secondary containment ventilation 3.5 48 295024 High Drywell Pressure / 5 X

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

EK3.03 Containment venting: Mark-III 3.6 49 295025 (EPE 2) High Reactor Pressure / 3 X

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

EA1.07 ARI/RPT/ATWS: Plant-Specific 4.1 50 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

4.2 51

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

RB-2018-07 E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 X

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

EA2.03 Reactor pressure: Mark-III 3.3 52 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 Not Selected 295030 (EPE 7) Low Suppression Pool Water Level / 5 X

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

EK2.08 SRV discharge submergence 3.5 53 295031 (EPE 8) Reactor Low Water Level / 2 X

Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL:

EK3.01 Automatic depressurization system actuation 3.9 54 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

EA1.10 Alternate boron injection methods: Plant-Specific 3.7 55 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE:

EA2.03 Radiation levels 3.5 56 600000 (APE 24) Plant Fire On Site / 8 X

Knowledge of the interrelations between PLANT FIRE ON SITE and the following:

AK2.01 Sensors / detectors and valves 2.6 57 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

Knowledge of the operational implications of the following concepts as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

AK1.03 Under-excitation 3.3 58 K/A Category Totals:

3 4 3 3 4 3 Group Point Total:

20

ES-401 4

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

RB-2018-07 E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum / 3 Not Selected 295007 (APE 7) High Reactor Pressure / 3 Not Selected 295008 (APE 8) High Reactor Water Level / 2 X

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

AK1.03 Feed flow/steam flow mismatch 3.2 59 295009 (APE 9) Low Reactor Water Level / 2 X

Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following:

AK2.02 Reactor water level control 3.9 60 295010 (APE 10) High Drywell Pressure / 5 X

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

AK3.04 Leak investigation 3.5 61 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 Not Selected 295012 (APE 12) High Drywell Temperature /

5 Not Selected 295013 (APE 13) High Suppression Pool Temperature. / 5 Not Selected 295014 (APE 14) Inadvertent Reactivity Addition / 1 X

Ability to operate and/or monitor the following as they apply to INADVERTENT REACTIVITY ADDITION:

AA1.06 Reactor/turbine pressure regulating system 3.3 62 295015 (APE 15) Incomplete Scram / 1 Not Selected 295017 (APE 17) Abnormal Offsite Release Rate / 9 Not Selected 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 Not Selected 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 Not Selected 295029 (EPE 6) High Suppression Pool Water Level / 5 Not Selected 295032 (EPE 9) High Secondary Containment Area Temperature / 5 Not Selected 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 X

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

EA2.03 Cause of high area radiation 3.7 63 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 X

2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

4.5 64 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 Not Selected 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 Not Selected

ES-401 5

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

RB-2018-07 E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 500000 (EPE 16) High Containment Hydrogen Concentration / 5 X

Knowledge of the operational implications of the following concepts as they apply to HIGH CONTAINMENT HYDROGEN CONCENTRATIONS:

EK1.01 Containment integrity 3.3 65 K/A Category Point Totals:

2 1

1 1

1 1

Group Point Total:

7

ES-401 6

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

RB-2018-07 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

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:

A2.14 Initiating logic failure 3.8 1

205000 (SF4 SCS) Shutdown Cooling X

Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including:

A1.08 Heat exchanger temperatures 3.1 2

206000 (SF2, SF4 HPCIS) High-Pressure Coolant Injection N/A for RBS 207000 (SF4 IC) Isolation (Emergency)

Condenser N/A for RBS 209001 (SF2, SF4 LPCS) Low-Pressure Core Spray X

X Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.09 Load sequencing Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM:

K6.01 A.C. power 3.3 3.4 3

4 209002 (SF2, SF4 HPCS) High-Pressure Core Spray X

X Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) and the following:

K1.04 HPCS diesel generator: BWR-5,6 Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE CORE SPRAY SYSTEM (HPCS):

K5.04 Adequate core cooling: BWR-5,6 3.8 3.8 5

6 211000 (SF1 SLCS) Standby Liquid Control X

Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.04 Indication of fault in explosive valve firing circuits 3.8 7

ES-401 7

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

RB-2018-07 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 212000 (SF7 RPS) Reactor Protection X

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

K3.05 RPS logic channels Ability to manually operate and/or monitor in the control room:

A4.07 System status lights and alarms 3.7 4.0 8

9 215003 (SF7 IRM) Intermediate-Range Monitor X

Knowledge of electrical power supplies to the following:

K2.01 IRM channels/detectors 2.5 10 215004 (SF7 SRMS) Source-Range Monitor X

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

K1.06 Reactor vessel 2.8 11 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following:

K3.04 Rod control and information system 3.4 12 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

Knowledge of electrical power supplies to the following:

K2.02 RCIC initiation signals (logic) 2.8 13 218000 (SF3 ADS) Automatic Depressurization X

Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM:

K6.04 Air supply to ADS valves: Plant-Specific 3.6 14 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following:

K4.01 Redundancy 3.0 15 239002 (SF3 SRV) Safety Relief Valves X

Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including:

A1.08 Suppression pool water temperature 3.8 16 259002 (SF2 RWLCS) Reactor Water Level Control X

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

K5.03 Water level measurement 3.1 17

ES-401 8

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

RB-2018-07 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 261000 (SF9 SGTS) Standby Gas Treatment X

Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including:

A3.01 System flow 3.2 18 262001 (SF6 AC) AC Electrical Distribution X

X Knowledge of A.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following:

K4.01 Bus lockouts Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.11 Degraded system voltages 3.0 3.2 19 20 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including:

A3.01 Transfer from preferred to alternate source 2.8 21 263000 (SF6 DC) DC Electrical Distribution X

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

4.1 22 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

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

A4.03 Transfer of emergency control between manual and automatic 3.2 23 300000 (SF8 IA) Instrument Air X

Ability to monitor automatic operations of the INSTRUMENT AIR SYSTEM including:

A3.02 Air temperature 2.9 24 400000 (SF8 CCS) Component Cooling Water X

Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

A2.02 High/low surge tank level 2.8 25 510000 (SF4 SWS*) Service Water (Normal and Emergency)

X Ability to predict and/or monitor changes in parameters associated with operating the Service Water System controls including:

A.1.03 Service Water Pressure 2.7 26

ES-401 9

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

RB-2018-07 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 K/A Category Point Totals:

2 2

2 4

2 2

3 3

2 3

1 Group Point Total:

26

ES-401 10 Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO)

RB-2018-07 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 201001 (SF1 CRDH) CRD Hydraulic Not Selected 201002 (SF1 RMCS) Reactor Manual Control N/A for RBS 201003 (SF1 CRDM) Control Rod and Drive Mechanism X

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

4.4 27 201004 (SF7 RSCS) Rod Sequence Control N/A for RBS 201005 (SF1, SF7 RCIS) Rod Control and Information X

Knowledge of the physical connections and/or cause-effect relationships between ROD CONTROL AND INFORMATION SYSTEM (RCIS) and the following:

K1.02 Reactor/turbine pressure control system: BWR-6 3.3 28 201006 (SF7 RWMS) Rod Worth Minimizer N/A for RBS 202001 (SF1, SF4 RS) Recirculation X

Knowledge of electrical power supplies to the following:

K2.02 MG sets: Plant-Specific 3.2 29 202002 (SF1 RSCTL) Recirculation Flow Control X

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

K3.01 Core flow 3.5 30 204000 (SF2 RWCU) Reactor Water Cleanup Not Selected 214000 (SF7 RPIS) Rod Position Information N/A for RBS 215001 (SF7 TIP) Traversing In-Core Probe Not Selected 215002 (SF7 RBMS) Rod Block Monitor N/A for RBS 216000 (SF7 NBI) Nuclear Boiler Instrumentation Not Selected 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode Not Selected 223001 (SF5 PCS) Primary Containment and Auxiliaries Not Selected 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode (Mark IV)

N/A for RBS 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode N/A for RBS 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup X

Knowledge of FUEL POOL COOLING AND CLEAN-UP design feature(s) and/or interlocks which provide for the following:

K4.06 Maintenance of adequate pool level 2.9 31

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

RB-2018-07 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 234000 (SF8 FH) Fuel-Handling Equipment X

Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT:

K5.02 Fuel handling equipment interlocks 3.1 32 239001 (SF3, SF4 MRSS) Main and Reheat Steam X

Knowledge of the effect that a loss or malfunction of the following will have on the MAIN AND REHEAT STEAM SYSTEM:

K6.10 ADS/low low set: Plant-Specific 3.6 33 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control Not Selected 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating X

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including:

A1.14 Pressure setpoint/pressure demand 3.4 34 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary X

Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.05 Generator trip 3.6 35 256000 (SF2 CDS) Condensate X

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

A3.05 Lights and alarms 3.0 36 259001 (SF2 FWS) Feedwater X

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

A4.08 FWRV position 3.3 37 268000 (SF9 RW) Radwaste Not Selected 271000 (SF9 OG) Offgas Not Selected 272000 (SF7, SF9 RMS) Radiation Monitoring Not Selected 286000 (SF8 FPS) Fire Protection Not Selected 288000 (SF9 PVS) Plant Ventilation X

2.4.21 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.

38 290001 (SF5 SC) Secondary Containment Not Selected

ES-401 12 Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO)

RB-2018-07 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 290003 (SF9 CRV) Control Room Ventilation Not Selected 290002 (SF4 RVI) Reactor Vessel Internals Not Selected 51001 (SF8 CWS*) Circulating Water Not Selected K/A Category Point Totals:

1 1

1 1

1 1

1 1

1 1

2 Group Point Total:

12 Last two NRC exams overlap September 2016 June 2015 December 2014 RO Written Examination X

X SRO Written Examination X

X Operating Examination X

X Last examinations:

September 2016 - Full NRC examination 2015 - RO makeup written examination December 2014 - Full NRC examination March 2014 - Full NRC examination

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

Form ES-401-3 Facility: River Bend Station Date of Exam: July 2018 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements.

3.8 66 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

3.3 67 2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

4.1 68 Subtotal

2. Equipment Control 2.2.7 Knowledge of the process for conducting special or infrequent tests.

2.9 69 2.2.23 Ability to track Technical Specification limiting conditions for operations.

3.1 70 Subtotal

3. Radiation Control 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

3.5 71 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.4 72 Subtotal

4. Emergency Procedures/Plan 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

3.8 73 2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

3.6 74 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

4.2 75 Subtotal Tier 3 Point Total 10

ES-401 1

Form ES-401-1 Rev 1 Facility: River Bend Station Date of Exam: July 2018 Tier Group RO K/A Category Points SRO-Only Points K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

N/A N/A 20 4

3 7

2 7

2 1

3 Tier Totals 27 6

4 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 Rev 2 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)

E/APE # / Name / Safety Function 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 / 1 & 4 295003 (APE 3) Partial or Complete Loss of AC Power / 6 X

AA2 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER:

AA2.04 System lineups 3.7 84 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 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

4.5 85 295021 (APE 21) Loss of Shutdown Cooling / 4 295023 (APE 23) Refueling Accidents / 8 X 2.4.41 Knowledge of the emergency action level thresholds and classifications.

4.6 86 295024 High Drywell Pressure / 5 X

EA2 - Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

EA2.03 Suppression pool level 3.8 87 295025 (EPE 2) High Reactor Pressure / 3 X

2.2.44 Ability to interpret control room indications to verify the status and l operation of a system, and understand how operator actions and directives affect plant and system conditions.

4.4 88 295026 (EPE 3) Suppression Pool High Water Temperature / 5 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 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

EA2 -Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :

EA2.02 Reactor water level 4.2 89 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 600000 (APE 24) Plant Fire On Site / 8

ES-401 3

Form ES-401-1 Rev 2 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

AA2 - Ability to determine and/or interpret the following as l they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

AA2.07 Operational status of engineered safety features 4.0 90 K/A Category Totals:

0 0 0 0 4 3 Group Point Total:

7

ES-401 4

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 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 295009 (APE 9) Low Reactor Water Level /

2 X 2.4.6 Knowledge of EOP mitigation strategies.

4.7 91 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 X

AA2 - Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION :

AA2.01 Reactor power 4.2 92 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) High Offsite Release Rate / 9 X

AA2 -Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE :

AA2.01 Off-site release rate: Plant-Specific 4.2 93 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 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 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals:

0 0

0 0

2 1

Group Point Total:

3

ES-401 5

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 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 205000 (SF4 SCS) Shutdown Cooling 206000 (SF2, SF4 HPCIS) High-Pressure Coolant Injection N/A for RBS 207000 (SF4 IC) Isolation (Emergency)

Condenser N/A for RBS 209001 (SF2, SF4 LPCS) Low-Pressure Core Spray X

A2 - 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:

A2.09 Low suppression pool level 3.3 76 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 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization X

A2 - Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.05 Loss of A.C. or D.C. power to ADS valves 3.6 78 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff 239002 (SF3 SRV) Safety Relief Valves 259002 (SF2 RWLCS) Reactor Water Level Control 261000 (SF9 SGTS) Standby Gas Treatment 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

ES-401 6

Form ES-401-1 Rev 1 263000 (SF6 DC) DC Electrical Distribution X

2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

4.2 79 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

4.3 77 300000 (SF8 IA) Instrument Air 400000 (SF8 CCS) Component Cooling Water 510000 (SF4 SWS*) Service Water (Normal and Emergency)

X A2.01 - Ability to (a) predict the impacts of the following on the SWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

A2.01 Loss of SWS pump 3.4 80 K/A Category Point Totals:

0 0

0 0

0 0

0 3

0 0

2 Group Point Total:

5

ES-401 7

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 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)

IR 201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control N/A for RBS 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control N/A for RBS 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 N/A for RBS 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor N/A for RBS 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries X

A2 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.06 High containment pressure 4.1 81 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode N/A for RBS 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode N/A for RBS 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 X 2.1.32 Ability to explain and apply system limits and precautions.

4.0 82 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater

ES-401 8

Form ES-401-1 Rev 1 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection X

2.1.28 Knowledge of the purpose and function of major system components and controls.

4.1 83 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:

0 0

0 0

0 0

0 1

0 0

2 Group Point Total:

3

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

Form ES-401-3 Rev 1 2.1.39 Facility: River Bend Station Date of Exam: July 2018 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements.

4.2 94 2.1.40 2.1.40 Knowledge of refueling administrative requirements.

3.9 95 Subtotal

2. Equipment Control 2.2.12 Knowledge of surveillance procedures.

4.1 96 2.2.37 Ability to determine operability and/or availability of safety related equipment.

4.6 97 Subtotal

3. Radiation Control 2.3.13 Knowledge of radiological safety 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, aligning filters, etc.

3.8 98 Subtotal

4. Emergency Procedures/Plan 2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

4.4 99 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation.

4.5 100 Subtotal Tier 3 Point Total 10 7

Rev 1 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group (Original)

Randomly Selected K/A (New)

Reason for Rejection RO T1G1 295006 AK3.05 AK3.04 Original K/A: 295006 SCRAM, AK3. Knowledge of the reasons for the following responses as they apply to SCRAM: Direct turbine generator trip: Plant-Specific Reason for Rejection: Original K/A is not applicable to RBS.

Randomly Selected K/A: 295006 SCRAM, AK3. Knowledge of the reasons for the following responses as they apply to SCRAM: AK3.04 Reactor water level setpoint setdown: Plant-Specific Page 1 point totals not affected by this change.

RO T2G1 215005 K3.07 K3.04 Original K/A: 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor, Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following: K3.07 Rod block monitor:

Plant-Specific Reason for Rejection: Original K/A is not applicable to RBS.

Randomly selected K/A: 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor, K3. Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following: K3.04 Rod control and information system.

Page 1 point totals not affected by this change.

RO T2G1 259002 K5.09 K5.03 Original K/A: 259002 Reactor Water Level Control System, K5.

Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM : K5.09 Adequate core cooling: FWCI Reason for Rejection: Difficulty writing to RO level and eliminate overlap with other questions.

Randomly selected K/A: 259002 Reactor Water Level Control System, K5. Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM : K5.03 Water level measurement

Rev 1 RO T2G2 201006 K1.02 201005 K1.02 Original K/A: 201006 Rod Worth Minimizer System (RWM) (Plant Specific), K1. Knowledge of the physical connections and/or cause-effect relationships between ROD WORTH MINIMIZER SYSTEM (RWM)

(PLANT SPECIFIC) and the following: Rod position indication system: P-Spec(Not-BWR6)

Reason for Rejection: Original K/A is not applicable to RBS.

Randomly Selected K/A: 201005 Rod Control and Information System (RCIS), K1. Knowledge of the physical connections and/or cause-effect relationships between ROD CONTROL AND INFORMATION SYSTEM (RCIS) and the following: Reactor/turbine pressure control system: BWR-6 Page 1 point totals not affected by this change.

SRO T1/G1 295019 2.4.11 2.4.8 Original K/A: 295019 Partial or Complete Loss of Instrument Air, 2.4.11 Knowledge of abnormal condition procedures.

Reason for Rejection: Original K/A is not listed in NUREG 1021, ES-401 for Tier 1 and 2 selected K/As.

Randomly Selected K/A: 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Page 1 point totals not affected by this change.

SRO T1/G1 295023 2.4.3 2.4.41 Original K/A: 295023 Refueling Accidents, 2.4.3 Ability to identify post-accident instrumentation.

Reason for Rejection: Unable to tie SRO only ability to identify post-accident instrumentation with the parent KA of Refueling Accidents.

Randomly Selected K/A: 2.4.41 Knowledge of the emergency action level thresholds and classifications.

Page 1 point totals not affected by this change.

SRO T2/G1 209002 2.4.20 209002 2.4.20 Original K/A: 209002 (SF2, SF4 HPCS) High-Pressure Core Spray, 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

Reason for Rejection: Replaced K/A due to overlap with RO examination.

Randomly Selected K/A: 264000 Emergency Generators (Diesel/Jet) 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

Page 1 point totals not affected by this change.

Rev 1 SRO T2/G2 226001 A2.16 223001 A2.06 Original K/A: 226001 RHR/LPCI: Containment Spray System Mode, A2.16 Loss of, or inadequate heat exchanger cooling flow Reason for Rejection: Original K/A is not applicable to RBS. RHR does not have a Containment Spray Mode.

Randomly Selected K/A: 223001 Primary Containment System and Auxiliaries, A2.06 High containment pressure: Mark-III Page 1 point totals not affected by this change.

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

River Bend Nuclear Station Date of Examination:

7/23/2018 Examination Level: RO SRO Operating Test Number:

2018-07 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R,N Determine If the Reactor Remains Shutdown under All Conditions - Incomplete SCRAM K/A: 2.1.37 JPM: RJPM-NRC18-A1 Conduct of Operations R,D Perform Surveillances Required Following Entry Into Single Loop Operation K/A: 2.1.7 JPM: RJPM-NRC18-A2 Equipment Control R,N Determine Protected Equipment Posting Requirements for Risk Related Equipment Out of Service K/A: 2.2.14 JPM: RJPM-NRC18-A3 Radiation Control R,M Prepare for RCA Entry K/A: 2.3.7 JPM: RJPM-NRC18-A4 Emergency Plan 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 Revision 1 Facility:

River Bend Nuclear Station Date of Examination:

7/23/2018 Examination Level: RO SRO Operating Test Number:

2018-07 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R,N Determine If the Reactor Remains Shutdown under All Conditions - Incomplete SCRAM K/A: 2.1.37 JPM: RJPM-NRC18-A5 Conduct of Operations R,D Determine If Core Alterations Are Allowed K/A: 2.1.36 JPM: RJPM-NRC18-A6 Equipment Control R,D Determine Plant Safety Level During Shutdown Conditions K/A: 2.2.18 JPM: RJPM-NRC18-A7 Radiation Control R,N Evaluate And Administer Potassium Iodide To Individual Following Airborne Exposure.

K/A: 2.3.4 JPM: RJPM-NRC18-A8 Emergency Plan R,N NRC Notification Requirements K/A: 2.4.30 JPM: RJPM-NRC18-A9 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)

Rev 4 Facility:

River Bend Date of Examination:

7/23/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

2018-07 Control Room Systems:* 8 for RO, 7 for SRO-I, and 3 for SRO-U System/JPM Title Type Code*

Safety Function S1 STP-052-0101, Fully Withdrawn Control Rod Insertion Operability Check K/A: 201003 A2.02 RJPM-NRC18-S1 A D S 1

S2 Start 3rd condensate pump 104-01 K/A: 256000 A4.01 RJPM-NRC18-S2 A D S 2

S3 Main Stop Valve Testing (1 for retest) per OSP-0102 K/A: 241000 A4.07 RJPM-NRC18-S3 N S 3

S4 Align LPCS to maintain RWL K/A: 209001 A1.01 RJPM-NRC18-S4 A E EN L N S 4

S5 Makeup Supp Pool Level with HPCS [02/14 Audit]

K/A: 223001 A1.08 RJPM-NRC18-S5 M E EN 5

S6 Div 1 Manual Scram Pushbuttons IAW STP-508-0201 K/A: 212000 A4.01 RJPM-NRC18-S6 D S 7

S7 Emergency Operation of Containment Coolers with service water

[03/14 NRC]

K/A: 223001 A2.01 & A2.13 RJPM-NRC18-S7 M E EN 8

S8 SPC Pump Rotation K/A: 233000 A4.01 RJPM-NRC18-S8 A N S 9

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 2 for SRO-U P1 Manually Startup RHR B In Suppression Pool Cooling From RSP 12/10 NRC K/A 219000 A1.02 RJPM-NRC18-P1 E D EN 5

P2 Transfer E51-F063 (RCIC STEAM SUPPLY INBD ISOL VALVE) to Alternate Power K/A: 217000 A2.04 RJPM-NRC18-P2 E EN D L R 4

P3 Manual Start HPCS Diesel Generator K/A: 264000 A4.04 RJPM-NRC18-P3 A EN N 6

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions,

Rev 4 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





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RO SRO-I SRO-U Required Actual Required Actual Required Actual A

4-6 5

4-6 5

2-3 3

D



5



4



2 E



5

1 5



2 EN



6



6



3 L



2



2



1 N or M



6



6



3 P



0



0



0 R



1



1



1 Simulator Schedule S1 and S7 S2 and S5 S3 and S8 S4 S6

2018 NRC Scenario 1 1 of 29 Revision 6 Facility: River Bend Nuclear Station Scenario No.: 1 Op-Test No.: 2018-07 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100% Power. RHR A is in suppression pool cooling. Containment Purge is in service.

Inoperable Equipment: None Turnover: Slow roll RCIC per SOP-35 RCIC System for retest after governor oil replacement.

Two hours into LCO 3.5.3. condition A for RCIC, HPCS is protected.

Event No.

Malf. No.

Event Type Event Description 1

RCIC009 I (BOP/CRS)

TS (CRS)

A (ALL)

RCIC started/H13-P601/21A/C01DIV I RCIC ISOL MN STM SPLY LINE DIFF PRESS HIGH alarms / Division 1 isolation with failed MOV, TS 3.3.6.1 Condition A, TS 3.6.1.3 Condition A entered.

2 LPRMUP0615D I (ATC/CRS)

A (ALL)LPRM 06-15D fails upscale / half scram 3

CNM004B C (ATC/CRS)

A (ALL)

CNM-P1B trips / power reduction 4

ED003A C (BOP/CRS)

A(CREW)

TS (CRS)

NNS-SWGR1C LOCKOUT Division 3 bus fault /Division 3 DG starts/secure unloaded DG, TS 3.8.9 Condition E, TS 3.5.1 Condition B, TS 3.7.1 Condition E, and TS 3.5.3 Condition B 5

CNM001A-K M (ALL)

Loss of all condensate/scram 6

RCS007 M (ALL)

Reactor coolant leak / loss of RPV water level CT-1 7

RHR009A LPCS005 C (BOP/CRS)

Division 1 ECCS fails to auto initiate/ manually initiated.

CT-2 8

RHR001B RHR001C C (ATC/CRS)

RHR B and C injection valves fail to auto open, but can be manually opened.

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)

2018 NRC Scenario 1 2 of 29 Revision 6 Quantitative Attributes Table Attribute E3-301-4 Target Actual Description Malfunctions after EOP entry 1-2 2

x Div 1 ECCS Fails to auto initiate manual action works x RHR B and C injection valves fail closed. Injection valves can be manually opened.

Abnormal Events 2-4 4

x RCIC started/H13-P601/21A/C01DIV I RCIC ISOL MN STM SPLY LINE DIFF PRESS HIGH alarms / Division 1 isolation with failed MOV, TS 3.3.6.1 Condition A x

LPRM 06-15D fails upscale / half scram x

CNM-P1B trips / power reduction x

NNS-SWGR1C LOCKOUT Division 3 bus fault Major Transients 1-2 2

x All condensate demineralizers and filters clog resulting in a complete loss of feed. Crew will manually scram.

(EOP-1, AOP-1) x Reactor coolant leak in the drywell causes a loss of RPV water level. (EOP-1)

EOP entries requiring substantive action 1-2 1

x Reactor coolant leak in the drywell causes a loss of RPV water level. (EOP-1)

Entry into a contingency EOP with substantive actions 1

1 x Reactor coolant leak in the drywell causes a loss of RPV water level. (EOP-1) Crew must perform Emergency Depressurization due to low reactor water level.

Preidentified critical tasks 2-3 2

x CT-1: Emergency Depressurize with at least one injection source lined up for injection when level cannot be restored and maintained above -187, within 15 minutes. (15 minutes start when RPV level reaches -187 inches).

x CT-2: Restore reactor water level prior to exiting EOP-1, RPV Control, and entering the SAPs.

2018 NRC Scenario 1 3 of 29 Revision 6 SCENARIO ACTIVITIES:

Initial Conditions Initial Conditions: 100% Power. RHR A is in suppression pool cooling. Containment Purge is in service.

Inoperable Equipment: None Turnover: Slow roll RCIC per SOP-35, RCIC System, for retest after governor oil replacement.

These are done:

4.3.2. Place RHR into Suppression Pool Cooling mode per SOP-0031, Residual Heat Removal.

4.3.3. Place Containment Purge in service per SOP-0059, Containment HVAC System.

Event 1 - (Initial Setup - Automatic)

RCIC is manually started for retest after oil in governor replaced. When RCIC is up to normal speed (>2500 rpm), H13-P601/21A/C01DIV I RCIC ISOL MN STM SPLY LINE DIFF PRESS HIGH, alarms requiring Division 1 RCIC system isolation. Per AOP-3, Automatic Isolations, E51-F064, RCIC STEAM SUPPLY OUTBD ISOL VALVE, fails to auto close and must be manually closed.

Tech Spec 3.3.6.1. Primary Containment and Drywell Isolation Instrumentation, Condition A entered.

Tech Spec 3.6.1.3 Primary Containment Isolation Valves (PCIVs), Condition A entered.

Event 2 - (Triggered by Lead Examiner)

A failure of LPRM 06-15D upscale will cause a half scram on RPS B. Crew should implement ARP-680-06A-A03 and C03. Bypass APRM F and reset RPS half scram.

Review Tech Specs 3.3.1.1 and TR 3.3.2.1 for APRM operability (no LCO entered).

BypassLPRM 06-15D and restore APRM F to service.

Event 3 - (Triggered by Lead Examiner)

One of the three operating condensate pumps (CNM-P1B) trips resulting in AOP-0006, CONDENSATE/FEEDWATER FAILURES, entry and required power reduction to 90%

with recirculation flow.

2018 NRC Scenario 1 4 of 29 Revision 6 Event 4 - (Triggered by Lead Examiner)

NNS-SWGR1C LOCKOUT Division 3 bus fault. Alarm H13-P601/16A/B03, DIV III 4KV BUS AUTO TRIP, indicates bus fault. Division 3 Diesel Generator starts and does not pick up the bus due to fault. Crew must secure the diesel generator.

Tech Spec 3.8.9 Distribution SystemsíOperating condition E entered.

Tech Spec 3.5.1, ECCS -Operating, condition B entered.

Tech Spec 3.7.1, Standby Service Water (SSW) System and Ultimate Heat Sink (UHS),

Condition E entered.

Tech Spec 3.5.3 RCIC System, Condition B entered.

Event 5 - (Triggered by Lead Examiner)

All condensate demineralizers and filters are clogged resulting in a complete loss of condensate and feed. Crew will manually scram. (This will prevent all feed and condensate availability for the rest of the scenario.) (EOP-1, AOP-1)

Event 6 - (Initial Setup - Automatic)

Reactor coolant leak in the drywell with no high pressure injection sources available will cause a loss of RPV water level. (EOP-1) RPV level will slowly lower due to loss of all high pressure injection systems. The crew will emergency depressurization to restore RPV water level with Low Pressure ECCS.

Event 7 - (Initial Setup - Automatic)

Division 1 ECCS fails to auto initiate, but can be manually initiated. Division 1 ECCS should initiate on low RPV level or high drywell pressure. The pump should start prior to Emergency Depressurization. Crew should recognize and manually initiate/align Division 1 ECCS systems.

Event 8 - (Initial Setup - Automatic)

RHR B and C injection valves fail to auto open, but can be manually opened. The injection valves should open on low RPV pressure to allow injection of low pressure system. The crew should recognize this after emergency depressurization performed and manually open the injection valves as needed to restore RPV level.

2018 NRC Scenario 1 5 of 29 Revision 6 CT-1 CT-2 Critical Task

• • Emergency Depressurize with at least one injection source lined up for injection when level cannot be restored and maintained above -187, within 15 minutes. (15 minutes start when RPV level reaches -187 inches)

Restore reactor water level above TAF

(-162 inches) prior to exiting EOP-1, RPV Control, and entering the SAPs.

EVENT 6

7&8 Safety Significance Per EOP-1, ALC-7, 8, and 11, Emergency RPV depressurization (signaled by Step ALC-11) permits injection from low head systems, maximizes the total injection flow, and minimizes the flow through any primary system break.

If an injection source is available, emergency depressurization should be delayed at least until RPV level reaches the top of the active fuel, but may be performed anytime RPV level is between the top of the active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL).

If it is believed that available injection sources are capable of restoring and maintaining RPV water level above the Minimum Steam Cooling RPV Water Level following RPV depressurization, the blowdown may be performed as soon as RPV level reaches the top of the active fuel. (For example, all low pressure ECCS are running but cannot inject until RPV pressure decreases below their shutoff heads.)

Per EOP-1, ALC-12, the contingent of Step ALC-12 provides one last opportunity for restoring adequate core cooling before the requirement for containment flooding occurs.

Adequate core cooling is ensured following emergency depressurization as long as one of two conditions exists:

• Steam Cooling with Injection: RPV water level can be restored and maintained above the Minimum Steam Cooling RPV Water Level (MSCRWL). The core is then cooled by a combination of submergence and steam cooling even with no core spray flow.

• Spray Cooling: Design core spray flow requirements are satisfied and RPV level can be restored and maintained at or above the elevation of the jet pump suctions.

Cueing The ADS/SRVs will indicate open using red lights, steam flow, and acoustic monitoring. Reactor pressure will lower.

RPV level will rise and system flow will be indicated after emergency depressurization, when a low pressure ECCS system is aligned for injection.

• 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 a significant plant degradation or significantly alters a mitigation strategy

• • Per EPSTG*0002, EOP-1, Step ED-6: Five open SRVs is the Minimum Number of SRVs Required for Emergency Depressurization (MNSRED) and is the least number of SRVs which corresponds to a Minimum Steam Cooling Pressure (MSCP) sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding MSCP. The MNSRED is utilized to assure the RPV will depressurize and remain depressurized when emergency depressurization is required. Refer to Appendix A for a detailed discussion of the MNSRED and the MSCP.

2018 NRC Scenario 1 6 of 29 Revision 6 Top 10 systems and operator actions important to risk that are tested:

Scenario 1 Event 3 Loss of FW/Cond - #7 Internal Events Event 4 EJS 480VAC Power - #5 Risk Significant System NPS-13.2kV - #10 Risk Significant System Event 5 Loss of FW/Cond - #7 Internal Events Event 5 Reactor Trip/Turbine Trip -#6 Internal Events Event 6 SRV Depressurization - #3 Risk Significant System Manual depressurization of reactor vessel - #1 Operator Actions Start maximized CRD injection - #10 Operator Actions Simulator Setup:

IC: 116 RHR placed into Suppression Pool Cooling mode per SOP-0031, Residual Heat Removal.

Containment Purge in service per SOP-0059, Containment HVAC System.

Protect HPCS.

2018 NRC Scenario 2 1 of 28 Revision 6 Facility: River Bend Nuclear Station Scenario No.: 2 Op-Test No.: 2018-07 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100% Power. Normal operating conditions.

Event No.

Malf. No.

Event Type Event Description 1

P601_16a:h-5 LO_E22-D2-A I (BOP/CRS)

TS (CRS)

A (BOP/CRS)

HPCS CST instrument fails low / manual swap to suppression pool, TS 3.3.5.1 Condition A, D 2

RMS013A, 1.0 E-1 C (BOP/CRS)

A (ALL)

TS (CRS)

RMS-RE13A fails upscale, automatic start of control room filter train/ HVC-AOD51A fails to isolate, manually isolated,

Tech Spec 3.3.7.1 Control Room Fresh Air (CRFA) System Instrumentation and Tech Spec 3.7.2 Control Room Fresh Air (CRFA) System 3

CNM001 (8)

C(ATC/CRS)

A (ALL)

Main Condenser vacuum degrades/ reduce power to maintain vacuum 4

B21001A I (ATC/CRS)

A (ALL)

RPV level transmitter A fails high / swap feedwater level control signals 5

CNM001 C (ALL)

A (ALL)

Vacuum continues to degrades / manual scram 6

RPS001A M (ALL)

ATWS / lower RPV level / insert rods EOP-1A CT-1 7

CNM001 C (ALL)

MSIVs FAIL to automatically close / manually close MSIVs.

CT-3 8

DI_C41-C001A DI_C41-C001B C (ALL)

C41-S1A(B), SLC PUMP A(B) fails to inject / start second SLC pump.

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)

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

2018 NRC Scenario 2 2 of 28 Revision 6 Quantitative Attributes Table Attribute E3-301-4 Target Actual Description Malfunctions after EOP entry 1-2 2

x MSIVs fail to close on low vacuum x 1st SLC pump start fails - key switch broken Abnormal Events 2-4 4

x HPCS CST instrument fails low x Main Condenser vacuum degrades x RMS-RE13A fails upscale x RPV level transmitter A fails Major Transients 1-2 1

x ATWS (fail to deenergize all RPS relays). EOP-1A, RPV Control - ATWS EOP entries requiring substantive action 1-2 1

x EOP-1, RPV Control Entry into a contingency EOP with substantive actions 1

1 x EOP-1A, ATWS Level/Power Control Preidentified critical tasks 2-3 3

x Terminate and prevent all injection into the RPV except boron injection, CRD, and RCIC to lower RPV level to <-56 inches prior to meeting Level/Power Conditions.

x Inject SLC to shutdown reactor, prior to exceeding Suppression Pool temperature (110ºF).

x Close MSIVs within 15 minutes of reaching 8.5 inches vacuum in the main condenser (auto isolation signal).

2018 NRC Scenario 2 3 of 28 Revision 6 SCENARIO ACTIVITIES:

Initial Conditions 100% Power. Normal operating conditions.

Inoperable Equipment: None Turnover: Continue to operate at normal rated power.

Event 1 - (Triggered by Lead Examiner)

HPCS CST instrument E22-LISN654C CST level transmitter fails low and suction sources should swap automatically but do not require manual swap per ARP-601-16A-H05. Enter Tech Spec 3.3.5.1 Condition A1, Immediately enter Condition D. Enter Tech Spec 3.3.5.1 Condition D.1, Declare HPCS INOP within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Enter Tech Spec 3.3.5.1 Condition D.2.2, Align HPCS pump suction to the suppression pool within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Event 2 - (Triggered by Lead Examiner)

RMS-RE13A, Control Room local air intake rad monitor, fails upscale resulting in an automatic start of the Division 1 control room charcoal filter train. HVC-AOD51A, CR TOILET DN STREAM ISOL, fails to isolate but can be manually isolated per AOP-3, Automatic Isolations. CRS will enter Tech Spec 3.3.7.1 Control Room Fresh Air (CRFA)

System Instrumentation and Tech Spec 3.7.2 Control Room Fresh Air (CRFA) System.

Event 3 - (Triggered by Lead Examiner)

Main Condenser vacuum degrades due to condenser air in leakage. Crew must reduce power to maintain vacuum. AOP-5, Loss of Main Condenser Vacuum, Trip of Circulating Water Pump entered.

Event 4 - (Triggered by Lead Examiner)

RPV level transmitter A fails upscale (slowly ramps up to give time to respond). Crew should recognize transmitter failure per H13-P680 / 03A / C08, RX FW LEVEL CONTROL SIGNAL FAILURE, and alternate feedwater level control signals per SOP-9, Feedwater System, Section 5.1.

Event 5 - (Triggered by Lead Examiner)

Main Condenser vacuum continues to degrade. Crew must manually scram per AOP-5, Loss of Main Condenser Vacuum, Trip of Circulating Water Pump.

2018 NRC Scenario 2 4 of 28 Revision 6 Event 6 - (Initial Setup - Automatic)

ATWS (fail to deenergize: all RPS relays will need to be deenergized to insert control rods by pulling fuses). EOP-1A, RPV Control - ATWS. The rods will be inserted per 0, DE-ENERGIZING SCRAM SOLENOIDS. RPS control power fuses will be removed to de-energize the scram solenoids and drive the control rods in.

Event 7 - (Initial Setup - Automatic)

MSIVs FAIL to close on low vacuum. Crew must manually close MSIVs (8.5 inches Hg vacuum).

Event 8 - (Initial Setup - Automatic) 1st C41-S1A(B), SLC PUMP A(B) attempted fails to start. Crew must start second SLC pump. The failure malfunction is initially on both pumps to ensure either pump that is started first fails. When the first pump is attempted the malfunction is deleted from the second pump.

2018 NRC Scenario 2 5 of 28 Revision 6 Critical Task CT-1 CT-2 CT-3 Terminate and prevent all injection into the RPV except boron injection, CRD, and RCIC to lower RPV level to <-56 inches prior to meeting Level/Power Conditions.

Level/Power Conditions exist when ALL of the following are met:

• Reactor power above 5%

or cannot be determined

• SP temperature above 110°F

• Any SRV open or DW pressure above 1.68 psid

• RPV level above -100 in.

Inject SLC to shutdown reactor, prior to exceeding Suppression Pool temperature (110ºF).

NOTE: Crew may take action to inject SLC but due to fault SLC may not be injecting before 110ºF.

Close MSIVs within 15 minutes of reaching 8.5 inches vacuum in the main condenser (auto isolation signal).

EVENT 5

8 7

Safety Significance Per EOP-1A, RLA-13, With RPV injection terminated and prevented, RPV water level and reactor power decrease at the maximum possible rate allowed by boil off.

Failure to completely stop RPV injection flow (with the exception of CRD, RCIC, and SLC) would delay the reduction in core inlet sub-cooling, thus increasing the potential for flux oscillations.

Per EOP-1A, per RQA-4&5, A scram failure event with reactor power above 5% coupled with an MSIV isolation, however, may result in rapid heat up of the suppression pool due to the steam discharged from the RPV via SRVs. The challenge to containment thus becomes the limiting factor which defines the second of the two possible conditions requiring initiation of boron injection.

Per EIP-2-001, NOUE is declared within 15 minutes if there is any loss or any potential loss of containment. A loss of primary containment per PC3, Primary containment isolation failure or bypass, due to a failure of both valves in any one line to close AND direct downstream pathway to the environment exists after PC isolation.

Cueing Feed, condensate, and HPCS injection is terminated. RPV level is lowering.

The first SLC pump will fail to start, but the second pump will inject. SLC flow will rise and reactor power will begin to slowly lower.

Vacuum will degrade to less than 8.5 inches main condenser vacuum. All of the MSIVs will fail to isolate, but he drain lines will automatically close.

• 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 a significant plant degradation or significantly alters a mitigation strategy

2018 NRC Scenario 2 6 of 28 Revision 6 Top 10 systems and operator actions important to risk that are tested:

Event 5 Reactor Trip/Turbine Trip -#6 Internal Events Event 6 Reactor Protection System - #4 Risk Significant systems Event 7 MSIV Closure - #9 Internal Events Simulator Setup:

IC: 101

2018 NRC Scenario 3 Revision 5 1 of 23 Facility: River Bend Nuclear Station Scenario No.: 3 Op-Test No.: 2018-07 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 86% Power. Normal operating conditions.

Event No.

Malf. No.

Event Type Event Description 1

R (ATC/CRS)

Raise power with rods per reactivity management plan.

2 CRDM1637 C (ATC/CRS)

TS (CRS)

A (ALL)

Control Rod 16-37 drift out / hydraulically isolated / Enter Tech Spec 3.1.3.C 3

p870_54a:g_5 C (BOP/CRS)

A (ALL)

Rotate to standby packing exhauster due to high temperature(swap from A to B) 4 TS (CRS)

LPCS Pump Breaker trip/Enter TS 3.5.1 Condition A 5

CRD016 C (BOP/CRS)

A (ALL)

CRD Pump A trip / start CRD-P1B 6

FWS004B C (ATC/CRS)

A (ALL)

FWS Master Controller output fails high / controller to manual 7

WCS004 WCS005 WCS006 M (ALL)

RWCU leak / reactor scram EOP-1/EOP-3 CT-1 8

D3mod124c1v final = 0 I (ALL)

Bypass Valves fail closed / pressure control to SRVs (and Bypass Valves if manually opened).

9 WCS006 C (ALL)

RWCU leak spreads to RCIC Room, Emergency depressurization per EOP-3.

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)

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

2018 NRC Scenario 3 Revision 5 2 of 23 Quantitative Attributes Table Attribute E3-301-4 Target Actual Description Malfunctions after EOP entry 1-2 2

x EHC pressure transmitter failure x

RWCU leak spreads to RCIC Room.

Abnormal Events 2-4 4

x Rod drifts out AOP-61, Control Rod Mispositioned/Malfunction x

Steam Packing Exhauster high temperature condition per ARP H13-P870-54A-G05.

x CRD pump trip, ARP-H13-P601-22A-A01 x

FRV malfunction, AOP-6, Condensate/Feedwater Failures Major Transients 1-2 1

x RWCU leak requires reactor scram per EOP-3, Secondary Containment and Radioactive Release Control EOP entries requiring substantive action 1-2 2

x Post scram RPV level and pressure control per EOP-1, RPV Control x

RWCU leak requires reactor scram per EOP-3, Secondary Containment and Radioactive Release Control Entry into a contingency EOP with substantive actions 1

1 x

RWCU leak spreads to RCIC Room, requires emergency depressurization per EOP-3.

Preidentified critical tasks 2-3 2

x Scram reactor prior to reaching Max Safe Operating Values in any area (SC-2).

x Emergency Depressurize RPV within 15 minutes of any Sec CTMT parameter exceeding its Table SC-2 Max Safe Operating Value in 2 or more areas.

2018 NRC Scenario 3 Revision 5 3 of 23 SCENARIO ACTIVITIES:

Initial Conditions 86% Power. Normal operating conditions.

Event 1 -

Raise power with control rods per reactivity management plan.

Event 2 - (Initial Setup - Automatic)

When withdrawing Control Rod 16-37 it continues to drift out, requiring it to be driven fully in and hydraulically isolated per AOP-61, Control Rod Mispositioned/ Malfunction.

CRS will enter Tech Spec 3.1.3.C when the control rod is isolated.

Event 3 - (Triggered by Lead Examiner)

Steam Packing Exhauster high temperature condition requiring crew to rotate to the standby packing exhauster.

Event 4 - (Triggered by Lead Examiner)

LPCS pump will trip and building operator will report acrid odor (no smoke or fire). CRS should enter Tech Spec 3.5.1 condition A.

Event 5 - (Triggered by Lead Examiner)

CRD Pump A trip. BOP starts the B lube oil pump and then CRD pump B to restore CRD flow.

Event 6 - (Triggered by Lead Examiner)

FWS Master Controller output fails high. ATC must take manual control of the controller per AOP-6, Condensate/Feedwater Failures.

Event 7 - (Triggered by Lead Examiner)

RWCU leak requires reactor scram per EOP-3, Secondary Containment and Radioactive Release Control.

Event 8 - (Initial Setup - Automatic)

Two minutes after manual scram, the RPV pressure controller pressure transmitter will fail. This will cause the Main Turbine Control Valves to close and prevent the Bypass Valves from automatically opening (they may still be manually opened). This results in a shift of RPV pressure control to the SRVs (and Bypass Valves if manually opened).

2018 NRC Scenario 3 Revision 5 4 of 23 Event 9 - (Initial Setup - Automatic, 5 minutes after the mode switch is taken to shutdown, the leak will spread)

RWCU leak spreads to RCIC Room, requires emergency depressurization per EOP-3.

2018 NRC Scenario 3 Revision 5 5 of 23 Critical Task CT-1 CT-2 Scram reactor prior to reaching Max Safe Operating Values in any area (SC-2).

• • Emergency Depressurize RPV within 15 minutes following any Sec CTMT parameter exceeding its Table SC-2 Max Safe Operating Value in 2 or more areas.

EVENT 7

9 Safety Significance Per EOP-3 bases for steps SC-13, 14, and 15, If temperatures, radiation levels or water 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.

Per EIP-2-001, when either RWCU pump room or RCIC room is above Table F2 Max Normal Operating Temperatures and above Table F1 Max Safe Operating Temperatures a Site Area Emergency should be declared per FS1. This is based on a loss of Reactor Coolant System Barrier (RC3) and Primary Containment Barrier (PC3). The crew has 15 minutes to make this upgrade declaration.

Cueing RWCU Pump Room area temperature will rise.

The ADS/SRVs will indicate open using red lights, steam flow, and acoustic monitoring.

Reactor pressure will lower.

• 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 a significant plant degradation or significantly alters a mitigation strategy

• • Per EPSTG*0002, EOP-1, Step ED-6: Five open SRVs is the Minimum Number of SRVs Required for Emergency Depressurization (MNSRED) and is the least number of SRVs which corresponds to a Minimum Steam Cooling Pressure (MSCP) sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding MSCP. The MNSRED is utilized to assure the RPV will depressurize and remain depressurized when emergency depressurization is required. Refer to Appendix A for a detailed discussion of the MNSRED and the MSCP.

2018 NRC Scenario 3 Revision 5 6 of 23 Top 10 systems and operator actions important to risk that are tested:

Event 7 Reactor Trip/Turbine Trip -#6 Internal Events Event 9 SRV Depressurization - #3 Risk Significant System Manual depressurization of reactor vessel - #1 Operator Actions Simulator Setup:

IC: 117 Verify CRD pump A running B in standby.

Verify A Steam Packing Exhauster is running and B in standby.