ML22144A026

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PV-2022-05-Draft Outlines
ML22144A026
Person / Time
Site: Palo Verde  Arizona Public Service icon.png
Issue date: 05/12/2022
From: Heather Gepford
NRC/RGN-IV/DORS/OB
To:
Arizona Public Service Co
References
Download: ML22144A026 (51)
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Text

Rev. 12 Form 4.1-PWR Pressurized-Water Reactor Examination Outline Notes: CO = Conduct of Operations; EC = Equipment Control; RC = Radiation Control; EM = Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan Facility: Palo Verde Date of Exam:

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 3

3 N/A 3

3 N/A 3

18 6

2 2

2 1

1 1

1 8

4 Tier Totals 5

5 4

4 4

4 26 10

2.

Plant Systems 1

3 3

2 2

2 2

2 3

3 3

3 28 5

2 1

1 1

1 1

0 1

0 1

1 1

9 3

Tier Totals 4

4 3

3 3

2 3

3 4

4 4

37 8

3.

Generic Knowledge and Abilities Categories CO EC RC EM 6

CO EC RC EM 7

2 2

1 1

4. Theory Reactor Theory Thermodynamics 6

3 3

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)

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

K/A Topic(s)

IR 000007 (EPE 7) Reactor Trip (CE E02) Standard Post-Trip Actions and Reactor Trip Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 X

Ability to determine and/or interpret the following as they apply to a Pressurizer Vapor Space Accident: (CFR: 43.5 I 45.13)

AA2.06 PORV logic control under low-pressure conditions 3.0 12 000009 (EPE 9) Small Break LOCA / 3 X

Ability to operate and/or monitor the following as they apply to a Small-Break LOCA: (CFR: 41. 7

/ 45.5 / 45.6)

EA1.01 RCS pressure and temperature 4.0 15 000011 (EPE 11) Large Break LOCA / 3 X

2.2.38 Knowledge of conditions and limitations in the facility license (CFR: 41.7 / 41.10 / 43.1 /

45.13) 3.6 14 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X

2.4.21 Knowledge of the parameters and logic used to assess the status of emergency operating procedures critical safety functions or shutdown critical safety functions (CFR: 41.7 /

43.5 / 45.12) 4.0 17 000025 (APE 25) Loss of Residual Heat Removal System / 4 X

Ability to operate and/or monitor the following as they apply to the Loss of the Residual Heat Removal System: (CFR: 41.7 / 45.5 / 45.6)

AA1.03 RHR 4.0 10 000026 (APE 26) Loss of Component Cooling Water / 8 X

Knowledge of the reasons for the following responses and/or actions as they apply to Loss of Component Cooling Water: (CFR: 41.5 /

41.10 / 45.6 / 45.13)

AK3.05 Tripping the reactor 4.1 1

000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X

Ability to operate and/or monitor the following as they apply to a Pressurizer Pressure Control System Malfunction: (CFR: 41. 7 / 45.5 / 45.6)

AA1.04 Pressure recovery using emergency-only heaters 3.5 2

000029 (EPE 29) Anticipated Transient Without Scram / 1 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Anticipated Transient Without Scram: (CFR: 41.8 / 41.10 / 45.3)

EK1.03 Addition of negative reactivity 4.2 7

000038 (EPE 38) Steam Generator Tube Rupture / 3 X

Knowledge of the relationship between a Steam Generator Tube Rupture and the following systems or components: (CFR: 41.7 141.8145.4145.7 145.8)

EK2.09 CVCS 3.3 9

000040 (APE 40) Steam Line Rupture / 4 (CE E05) Excess Steam Demand / 4 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Excess Steam Demand: (CFR: 41.5 / 41.7 / 45.7 / 45.8 / 45.9)

EK1.12 Evaluating RCP restart criteria 3.0 13

Rev. 12 000054 (APE 54) Loss of Main Feedwater /4 X

2.2.39 Knowledge of less than or equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> technical specification action statements (This K/A does not include action statements of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less that follow the expiration of a completion time for a technical specification condition for which an action statement has already been entered.) (CFR: 41.7 / 41.10 / 43.2

/ 45.13) 3.9 4

(CE E06) Loss of Feedwater /4 000055 (EPE 55) Station Blackout / 6 X

Knowledge of the relationship between a Station Blackout and the following systems or components: (CFR: 41.7 / 45.7)

EK1.01 Effect of battery discharge rates on capacity 3.8 3

000056 (APE 56) Loss of Offsite Power / 6 X

Knowledge of the reasons for the following responses and/or actions as they apply to Loss of Offsite Power: (CFR: 41.5,41.10 / 45.6 /

45.13)

AK3.01 Order and time to initiation of power for the load sequencer 3.6 6

000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X

Ability to determine and/or interpret the following as they apply to Loss of Vital AC Electrical Instrument Bus: (CFR: 43.5 / 45.13)

AA2.15 Verification that a loss of AC has occurred 4.1 11 000058 (APE 58) Loss of DC Power / 6 X

Ability to determine and/or interpret the following as they apply to Loss of DC Power:

(CFR: 43.5 / 45.13)

AA2.02 125-V DC bus voltage 3.6 18 000062 (APE 62) Loss of Nuclear Service Water

/ 4 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Loss of Service Water: (CFR: 41.8 / 41.10 / 45.3)

AK1.01 Effect on loads cooled by service water 3.8 8

000065 (APE 65) Loss of Instrument Air / 8 X

Knowledge of the relationship between Loss of Instrument Air and the following systems or components: (CFR: 41.7 / 45.7)

AK2.09 ECCS 3.5 5

000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X

Knowledge of the reasons for the following responses and/or actions as they apply to Generator Voltage and Electric Grid Disturbances: (CFR: 41.4 / 41.5 / 41. 7 / 41.10 /

45.8)

AK3.01 Reactor and turbine trip criteria 3.9 16 K/A Category Totals:

3 3

3 3

3 3

Group Point Total:

18

Form 4.1-PWR PWR Examination Outline Page 4 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

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

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 X

Knowledge of the relationship between Continuous Rod Withdrawal and the following systems or components: (CFR: 41.7 / 45.7)

AK2.07 Boric acid pump running lights 3.0 20 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak /

3 X

2.1.32 Ability to explain and apply system precautions, limitations, notes, or cautions (CFR: 41.10 / 43.2 / 45.12) 3.8 22 000051 (APE 51) Loss of Condenser Vacuum / 4 X

Ability to operate and/or monitor the following as they apply to Loss of Condenser Vacuum:

(CFR: 41. 7 I 45.5 I 45.6)

AA1.09 CWS 3.3 25 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X

Knowledge of the reasons for the following responses and/or actions as they apply to an Accidental Liquid Radwaste Release: (CFR:

41.5 / 41.10 / 45.6 / 45.13)

AK3.04 Guidance contained in procedures 3.6 26 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms / 7 000067 (APE 67) Plant Fire On Site / 8 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Plant Fire on Site:

(CFR: 41.8 / 41.10 / 45.3)

AK1.02 Fire-fighting methods for each type of fire 3.0 24 000068 (APE 68) Control Room Evacuation / 8 000069 (APE 69) Loss of Containment Integrity /

5 000074 (EPE 74) Inadequate Core Cooling / 4 000076 (APE 76) High Reactor Coolant Activity /

9 000078 (APE 78*) RCS Leak / 3 X

Ability to determine and/or interpret the following as they apply to a Reactor Coolant System Leak: (CFR: 43.5 / 45.13)

AA2.05 Letdown isolation valve position indication 3.5 23

Rev. 12 (CE A16) Excess RCS Leakage / 2 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Excess RCS Leakage: (CFR: 41.5141.7145.7145.8145.9)

AK1.07 How RCS leakage isolation can affect other systems 3.3 19 (CE E09) Functional Recovery (CE E13) Loss of Forced Circulation/LOOP/Blackout / 4 X

Knowledge of the relationship between Loss of Forced Circulation and/or LOOP and/or a Blackout and the following systems or components: (CFR: 41.7 / 41.8 / 45.4 / 45.7 /

45.8)

EK2.04 PZR LCS and PCS 3.2 21 K/A Category Point Totals:

2 2

1 1

1 1

Group Point Total:

8

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 6 Plant SystemsTier 2/Group 1 (RO/SRO)

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

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump X

Knowledge of Reactor Coolant Pump System design features and/or interlocks that provide for the following: (CFR: 41.7)

K4.05 Prevention of reverse rotation 3.0 36 004 (SF1; SF2 CVCS) Chemical and Volume Control X

Ability to (a) predict the impacts of the following on the Chemical and Volume Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 to

41. 7 / 43.5 / 45.3 / 45.5)

A2.11 Loss of IAS 3.3 43 005 (SF4P RHR) Residual Heat Removal X

2.4.31 Knowledge of annunciator alarms, indications, or response procedures (CFR:

41.10 / 45.3) 4.2 46 006 (SF2; SF3 ECCS) Emergency Core Cooling X

Knowledge of the physical connections and/or cause and effect relationships between the Emergency Core Cooling System and the following systems: (CFR:

41.2 to 41.8 / 45.3 / 45. 7 / 45.8)

K1.11 CCWS 3.7 34 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Pressurizer Relief Tank/Quench Tank System: (CFR:

41.7 / 45.7)

K6.11 Leakage collection 2.8 32 008 (SF8 CCW) Component Cooling Water X

Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.01 CCW Valves 3.0 37 010 (SF3 PZR PCS) Pressurizer Pressure Control X

Ability to (a) predict the impacts of the following on the Pressurizer Pressure Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 /

43.5 I 45.3 I 45.13)

A2.01 Heater failures 3.6 53 012 (SF7 RPS) Reactor Protection X

Knowledge of Reactor Protection System design features and/or interlocks that provide for the following: (CFR: 41.7)

K4.01 Trip logic when one channel is out of service or in test 4.2 41 013 (SF2 ESFAS) Engineered Safety Features Actuation X

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Engineered Safety Features Actuation System: (CFR: 41.3 / 41.4 / 41.5 / 45. 7)

K5.16 ESFAS signal with one train in test 3.8 30 022 (SF5 CCS) Containment Cooling X

Knowledge of the effect that a loss or malfunction of the Containment Cooling System will have on the following systems or system parameters: (CFR: 41. 7 / 45.6)

K3.04 Containment 3.9 51 025 (SF5 ICE) Ice Condenser

Rev. 12 026 (SF5 CSS) Containment Spray X

Ability to monitor automatic features of the Containment Spray System, including:

(CFR: 41. 7 / 45.5)

A3.01 Pump starts and correct valve positioning 4.1 54 039 (SF4S MSS) Main and Reheat Steam X

2.1.28 Knowledge of the purpose and function of major system components and controls (CFR: 41.7) 4.1 39 059 (SF4S MFW) Main Feedwater X

Ability to monitor automatic features of the Main Feedwater System, including: (CFR:

41. 7 / 45.5)

A3.03 Feedwater pump suction flow/pressure 3.6 48 061 (SF4S AFW) Auxiliary/Emergency Feedwater X

Ability to manually operate and monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.02 AFW Flow 4.2 33 062 (SF6 ED AC) AC Electrical Distribution X

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the AC Electrical Distribution System: (CFR: 41.5 /

45.7)

K5.09 Consequence of paralleling out-of-phase/mismatch in volts 3.7 28 063 (SF6 ED DC) DC Electrical Distribution X

Ability to manually operate and/or monitor in the control room: (CFR: 41. 7 / 45.5 to 45.8)

A4.02 Load shedding 3.6 27 064 (SF6 EDG) Emergency Diesel Generator X

Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.02 Fuel oil pumps 3.2 40 073 (SF7 PRM) Process Radiation Monitoring X

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Process Radiation Monitoring System: (CFR: 41.7 /

8-9)

K6.01 PRM component failures 3.2 38 076 (SF4S SW) Service Water X

Knowledge of the effect that a loss or malfunction of the Service Water System will have on the following systems or system parameters: (CFR: 41.7 / 45.6)

K3.16 AFW 3.4 44 078 (SF8 IAS) Instrument Air X

Knowledge of the physical connections and/or cause and effect relationships between the Instrument Air System and the following systems: (CFR: 41.3 to 41.8 /

45.7 / 45.8)

K1.04 Cooling water to compressor 2.9 35 103 (SF5 CNT) Containment X

Ability to predict and/or monitor changes in parameters associated with operation of the Containment System, including: (CFR:

41.5 / 45.5)

A1.01 Containment pressure, temperature, and/or humidity 3.9 42 053 (SF1; SF4P ICS*) Integrated Control 005 (SF4P RHR) Residual Heat Removal X

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.05 Raising or lowering refueling cavity level 3.4 31

Rev. 12 006 (SF2; SF3 ECCS) Emergency Core Cooling X

2.2.42 Ability to recognize system parameters that are entry-level conditions for technical specifications (CFR: 41.7 /

41.10 / 43.2 / 43.3 / 45.3) 3.9 49 010 (SF3 PZR PCS) Pressurizer Pressure Control X

Ability to predict and/or monitor changes in parameters associated with operation of the Pressurizer Pressure Control System, including: (CFR: 41.5 / 45.5)

A1.03 PRT/quench tank pressure and temperature 3.2 29 022 (SF5 CCS) Containment Cooling X

Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.01 CCS fans 3.6 50 026 (SF5 CSS) Containment Spray X

Ability to (a) predict the impacts of the following on the Containment Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 43.5 /

45.3 / 45.13)

A2.03 Failure of ESF 3.9 52 063 (SF6 ED DC) DC Electrical Distribution X

Ability to monitor automatic features of the DC Electrical Distribution System, including: (CFR: 41.7 / 45.5)

A3.02 Battery charger undervoltage stripping 3.1 47 064 (SF6 EDG) Emergency Diesel Generator X

Knowledge of the physical connections and/or cause and effect relationships between the Emergency Diesel Generators and the following systems: (CFR: 41.3 to 41.8 / 45.7 / 45.8)

K1.04 DC distribution system 3.9 45 K/A Category Point Totals:

3 3

2 2

2 2

2 3

3 3

3 Group Point Total:

28

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 9 Plant SystemsTier 2/Group 2 (RO/SRO)

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

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant X

Knowledge of Reactor Coolant System design features and/or interlocks that provide for the following: (CFR: 41.7 / 41.3)

K4.01 Filling and draining the RCS, the refueling cavity, and/or refueling canal 3.2 59 011 (SF2 PZR LCS) Pressurizer Level Control X

2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup (CFR: 41.10 / 45.12) 4.6 60 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation X

Knowledge of the physical connections and/or cause and effect relationships between the Nonnuclear Instrumentation System and the following systems: (CFR:

41.2 to 41.9 / 45.7 / 45.8)

K1.06 AFW system 3.7 55 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X

Ability to predict and/or monitor changes in parameters associated with operation of the Hydrogen Recombiner and Purge Control System, including: (CFR: 41.5 /

45.5)

A1.03 Recombiner temperature 2.8 58 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling X

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Spent Fuel Pool Cooling System: (CFR: 41

.5 / 45. 7)

K5.05 Decay heat 3.8 62 034 (SF8 FHS) Fuel-Handling Equipment X

Knowledge of the effect that a loss or malfunction of the Fuel Handling Equipment System will have on the following systems or system parameters:

(CFR: 41.2 to 41. 7 / 43.6 / 43. 7 / 45.6 to 45.8)

K3.03 Reactor components 2.9 56 035 (SF 4P SG) Steam Generator X

Ability to monitor automatic features of the Steam Generator System, including: (CFR:

41.7 / 45.5)

A3.01 SIG water level control 3.9 57 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X

Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.03 Turbine bypass control loop and valve power 2.9 63 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 10 Plant SystemsTier 2/Group 2 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s)

IR 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring X

Ability to manually operate and/or monitor in the control room: (CFR: 41. 7 / 45.8 /

45.9)

A4.01 Alarm and interlock setpoint checks and adjustments 3.3 61 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:

1 1

1 1

1 0

1 0

1 1

1 Group Point Total:

9

Rev. 12 Form 4.1-COMMON Common Examination Outline Facility: Palo Verde Date of Exam:

Generic Knowledge and AbilitiesTier 3 (RO/SRO)

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.21 Ability to verify that a copy of a controlled procedure is the proper revision (CFR: 41.10 / 45.10 / 45.13) 3.5 65 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication (CFR: 41.7 /

43.5 / 45.4) 4.3 66 Subtotal N/A N/A

2.

Equipment Control 2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels (CFR: 41.6 / 41.7 / 45.2) 4.6 68 2.2.22 Knowledge of limiting conditions for operation and safety limits (CFR: 41.5 / 43.2 / 45.2) 4.0 67 Subtotal N/A N/A

3.

Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms or personnel monitoring equipment (CFR: 41.11 / 41.12 / 43.4 / 45.9) 2.9 69 Subtotal N/A N/A

4.

Emergency Procedures/

Plan 2.4.12 Knowledge of operating crew responsibilities during emergency and abnormal operations (CFR: 41.10 /

45.12) 4.0 64 Subtotal N/A N/A Tier 3 Point Total 6

7 TheoryTier 4 (RO)

Category K/A #

Topic RO IR Reactor Theory K1.01 192003 Reactor Kinetics and Neutron Sources Explain the concept of subcritical multiplication 2.8 70 K1.18 192004 Reactivity Coefficients Describe the effect on boron reactivity worth from changes in the following: Explain and describe the effect of power defect and Doppler defect on reactivity 2.9 75 K1.21 192008 Reactor Operational Physics Explain the relationship between steam flow and reactor power given specific conditions 3.8 74 Subtotal N/A Thermodynamics K1.03 193005 Thermodynamic Cycles Describe how changes in system parameters affect thermodynamic efficiency 2.6 72 K1.01 193009 Core Thermal Limits Explain radial peaking factor 2.8 71 K1.05 193010 Brittle Fracture and Vessel Thermal Stress State the effect of fast neutron irradiation on reactor vessel metals 3.0 73 Subtotal N/A Tier 4 Point Total 6

Rev. 12 Form 4.1-PWR Pressurized-Water Reactor Examination Outline Notes: CO = Conduct of Operations; EC = Equipment Control; RC = Radiation Control; EM = Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan Facility: Palo Verde Date of Exam:

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 3

18 3

3 6

2 1

8 2

2 4

Tier Totals 4

26 5

5 10

2.

Plant Systems 1

28 3

2 5

2 9

2 1

3 Tier Totals 37 5

3 8

3.

Generic Knowledge and Abilities Categories CO EC RC EM 6

CO EC RC EM 7

2 2

1 1

2 2

1 2

4. Theory Reactor Theory Thermodynamics 6

3 3

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)

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

K/A Topic(s)

IR 000007 (EPE 7; BW E02&E10; CE E02) Reactor Trip, Stabilization, Recovery / 1 X

2.4.6 Knowledge of emergency and abnormal operating procedures major action categories (CFR: 41.10 / 43.5 / 45.13) 4.7 78 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X

Ability to determine and/or interpret the following as they apply to Reactor Coolant Pump Malfunctions: (CFR: 43.5 I 45.13)

AA2.10 Loss of cooling or seal injection 3.8 80 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 X

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response procedure (CFR: 41.10 / 43.5 / 45.3) 4.0 81 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient Without Scram / 1 000038 (EPE 38) Steam Generator Tube Rupture / 3 X

Ability to determine and/or interpret the following as they apply to a Steam Generator Tube Rupture: (CFR: 43.5 / 45.13)

EA2.25 PRT/quench tank temperature, pressure, and setpoints 3.0 79 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer /

4 X 2.1.20 Ability to interpret and execute procedure steps (CFR: 41.10 / 43.5 / 45.12) 4.6 77 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

Ability to determine and/or interpret the following as they apply to Loss of Feedwater:

(CFR: 43.5 / 45.13)

EA2.05 S/G level and pressure 4.0 76 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water

/ 4 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment / 3 (W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat Transfer Loss of Secondary Heat Sink / 4 K/A Category Totals:

Group Point Total:

6

Form 4.1-PWR PWR Examination Outline Page 3 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

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

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 X

Ability to determine and/or interpret the following as they apply to a Dropped Control Rod: (CFR: 43.5 / 45.13)

AA2.07 In-core NIS 3.4 82 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 X

2.1.23 Ability to perform general or normal operating procedures during any plant condition (CFR: 41.10 / 43.5 / 45.2 / 45.6) 4.4 85 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak /

3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms / 7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling / 4 000076 (APE 76) High Reactor Coolant Activity /

9 X

2.4.31 Knowledge of annunciator alarms, indications, or response procedures (CFR:

41.10 / 45.3) 4.1 83 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA Cooldown Depressurization / 4

Rev. 12 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS Overcooling Pressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery X

Ability to determine and/or interpret the following as they apply to Functional Recovery: (CFR: 41.10 / 43.5 / 45.13)

EA2.05 Charging and letdown flow 3.4 84 (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:

2 2

Group Point Total:

4

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 5 Plant SystemsTier 2/Group 1 (RO/SRO)

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

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW) Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:

Group Point Total:

28/5

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 6 Plant SystemsTier 2/Group 2 (RO/SRO)

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

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:

Group Point Total:

9/3

Rev. 12 Form 4.1-COMMON Common Examination Outline Facility:

Date of Exam:

Generic Knowledge and AbilitiesTier 3 (RO/SRO)

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations Subtotal N/A N/A

2.

Equipment Control Subtotal N/A N/A

3.

Radiation Control Subtotal N/A N/A

4.

Emergency Procedures/

Plan Subtotal N/A N/A Tier 3 Point Total 6

7 TheoryTier 4 (RO)

Category K/A #

Topic RO IR Reactor Theory Subtotal N/A Thermodynamics Subtotal N/A Tier 4 Point Total 6

Form 4.1-1 Record of Rejected Knowledge and Abilities Refer to Examination Standard (ES)-4.2, Developing Written Examinations, Section B.3, for deviations from the approved written examination outline.

Tier/Group Randomly Selected K/A Reason for Rejection NONE. NRC developed outline

Form 3.2-1 Administrative Topics Outline Facility:

PVNGS Date of Examination:

5/2/2022 Examination Level: RO X

SRO Operating Test Number:

2022 Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations (A1)

K/A: 2.1.20 IR: 4.6 JPM

Description:

The applicant will determine the minimum required Arc Flash Boundary as well as the minimum required PPE/EPE that must be worn in order to enter the AFB while racking a class 4kV breaker to the TEST position per 01DP-0IS13, Palo Verde Generating Station Electrical Safe Work Practices.

M, R Conduct of Operations (A2)

K/A: 2.1.25 IR: 3.9 JPM

Description:

The applicant will perform 73ST-9AF01, Auxiliary Feedwater N - Inservice Test. Raw pump performance data will be provided, and the applicant will have to use the raw data to determine pump suction pressure using a conversion table, then calculate pump D/P and determine if the D/P is acceptable by comparing the D/P to the appropriate unit acceptance criteria.

N, R Equipment Control (A3)

K/A: 2.2.12 IR: 3.7 JPM

Description:

The applicant will perform 40ST-9ZZM1, Operations Mode 1 Surveillance Logs, Section 6.1.5, Plant Protection System (PPS) Instrument Channel Checks. The applicant will be provided with all 4 channel indications for 13 PPS transmitters and have to determine if the maximum deviation between channels does or does not meet the acceptance criteria for each parameter.

M, R Radiation Control (A4)

K/A: 2.3.12 IR: 3.2 JPM

Description:

The applicant will be provided a survey map and Radiation Work Permit for the Unit 3 B HPSI Pump room in preparation for an IV on a tag hanging on a valve in the room. The applicant will have to determine the proper task to use on the RWP, the Protective Clothing requirements for the IV (if any),

and whether or not an RP Tech Spec briefing is required for the evolution.

N, R

Instructions for completing Form 3.2-1, Administrative Topics Outline

1. For each license level, determine the number of administrative job performance measures (JPMs) and topic areas as follows:

Topic Number of JPMs

  • Reactor operators (RO) applicants do not need to be evaluated on every topic (i.e., Equipment Control, Radiation Control, or Emergency Plan can be omitted by doubling up on Conduct of Operations), unless the applicant is taking only the administrative topics part of the operating test (with a waiver or excusal of the other portions).

RO*

SRO and RO Retakes Conduct of Operations 1 (or 2) 2 Equipment Control 1 (or 0) 1 Radiation Control 1 (or 0) 1 Emergency Plan 1 (or 0) 1 Total 4

5

2. Enter the associated knowledge and abilities (K/A) statement and summarize the administrative activities for each JPM.
3. For each JPM, specify the type codes for location and source as follows:

Location:

(C)ontrol Room, (S)imulator, or Class(R)oom Source and Source Criteria:

(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams (D)irect from bank (no more than three for ROs, no more than four for senior reactor operators (SROs) and RO retakes)

(N)ew of Significantly (M)odified from bank (no fewer than one)

Form 3.2-1 Administrative Topics Outline Facility:

PVNGS Date of Examination:

5/2/2022 Examination Level: RO SRO X Operating Test Number:

2022 Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations (A5)

K/A: 2.1.25 IR: 4.2 JPM

Description:

The applicant will be directed to determine long-term (> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) power reduction requirements following a slipped CEA and the minimum required CEA withdrawal time when the CEA can be withdrawn using various graphs and tables per 40AO-9ZZ11, CEA Malfunctions.

M, R Conduct of Operations (A6)

K/A: 2.1.39 IR: 4.3 JPM

Description:

The applicant will determine the minimum required Arc Flash Boundary as well as the minimum required PPE/EPE that must be worn in order to enter the AFB while racking a class 4kV breaker to the TEST position per 01DP-0IS13, Palo Verde Generating Station Electrical Safe Work Practices.

Following determination of PPE/EPE, the applicant will be given a second cue sheet indicating that a fatality occurred during the evolution, requiring the applicant to determine the reporting requirements of the event per the PVNGS Event Reporting Manual.

M, R Equipment Control (A7)

K/A: 2.2.36 IR: 4.2 JPM

Description:

The applicant will be directed to evaluate the TS impacts of a switchyard breaker which has relayed using PVNGS Technical Specifications.

N, R Radiation Control (A8)

K/A: 2.3.14 IR: 3.8 JPM

Description:

The applicant will be directed to determine if a release through the Plant Vent may continue following a loss of power and what actions are required to continue (or re-initiate) the release per the PVNGS Offsite Dose Calculation Manual.

D, R Emergency Plan (A9)

K/A: 2.4.41 IR: 4.6 JPM

Description:

The applicant will be directed to analyze plant conditions and determine the appropriate EAL classification.

N, R

Instructions for completing Form 3.2-1, Administrative Topics Outline

1. For each license level, determine the number of administrative job performance measures (JPMs) and topic areas as follows:

Topic Number of JPMs

  • Reactor operators (RO) applicants do not need to be evaluated on every topic (i.e., Equipment Control, Radiation Control, or Emergency Plan can be omitted by doubling up on Conduct of Operations), unless the applicant is taking only the administrative topics part of the operating test (with a waiver or excusal of the other portions).

RO*

SRO and RO Retakes Conduct of Operations 1 (or 2) 2 Equipment Control 1 (or 0) 1 Radiation Control 1 (or 0) 1 Emergency Plan 1 (or 0) 1 Total 4

5

2. Enter the associated knowledge and abilities (K/A) statement and summarize the administrative activities for each JPM.
3. For each JPM, specify the type codes for location and source as follows:

Location:

(C)ontrol Room, (S)imulator, or Class(R)oom Source and Source Criteria:

(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams (D)irect from bank (no more than three for ROs, no more than four for senior reactor operators (SROs) and RO retakes)

(N)ew of Significantly (M)odified from bank (no fewer than one)

Form 3.2-2 Control Room-Plant Systems Outline Facility:

PVNGS Date of Examination:

5/2/2022 Operating Test Number:

2022 Exam Level:

X RO X

SRO-I X

SRO-U System / JPM Type Type Code SF Control Room Systems S1 (RO only) 014 A2.04 Reset CEA positions in the Plant Computer and Core Monitoring Computer following a slipped CEA per 40AO-9ZZ11, CEA Malfunctions D, S 1

S2 (U) 013 A4.01 Establish correct equipment lineup and adequate SI flow following a Loss of Coolant Accident per 40EP-9EO03.

A, D, EN, L, S 2

S3 006 A4.02 Raise SIT 1A pressure to clear the low pressure alarm and stop raising pressure prior to bringing in the high pressure alarm per 40OP-9SI03, Safety Injection Tank Operations D,S 3

S4 003 A2.06 Respond to a loss of Nuclear Cooling Water per 40AO-9ZZ03, Loss of Cooling Water.

A, D, S 4P S5 045 A2.17 Respond to a partial Main Turbine load rejection per 40AO-9ZZ08, Load Rejection.

A, N, S 4S S6 (U) 062 A4.08 Transfer 13.8 kV Bus NAN-S01 to NAN-S03 per 40OP-9NA03, 13.8 kV Electrical Systems and recognize and respond to an ATWS.

A, M, S 6

S7 008 A2.01 Place Train A LPSI on SDC per Appendix 23 and establish EW cooling using the NC system per Appendix 243, NC Cross-Tie to EW Train A.

A, L, N, S 8

S8 (U) 050 A4.01 Restore Control Room Ventilation to a normal lineup following a CREFAS actuation per 40OP-9HJ01, Control Building HVAC.

EN, N, S 9

In-Plant Systems P1 (U) 008 A2.01 Perform (simulate) manual valve operations in the field for cross-tying NC to Train A EW per Appendix 243-A, NC Cross-Tie to EW Train A E, N, L, R 8

P2 (U) 045 K6.08 Perform Turbine Building verifications and (simulate) starting Main Lube Oil Pumps which failed to automatically start per 40AO-9ZZ18, Shutdown Outside Control Room, Appendix B, Turbine Building Actions A, D, E 4S P3 064 A4.06 Reset (simulate) the B EDG overspeed trip and manually start the EDG during a Blackout per Appendix 56, Restoring DG B to PBB-S04 E, M 6

SRO-U will perform JPMs S2, S6, S8, P1, and P2 RO Only JPM will be JPM S1

S1: The applicant will be directed to reset CEA positions in the Plant Computer and Core Monitoring Computer following a slipped CEA per 40AO-9ZZ11, CEA Malfunctions. The applicant will have to determine actual CEA position then locate, update and confirm the correct location in the PC and CMC.

This is a bank JPM covering Safety Function 1.

S2: The applicant will be directed to respond to a Loss of Coolant Accident per 40EP-9EO03. The applicant will have to identify multiple components which failed to auto actuate on SIAS (and manually start them), manually open 2 SI injection valves (one will fail to open) to establish minimum required SI flow (minimum flow wont be met), and trip all 4 RCPs due to a loss of NPSH. This is a bank alternate path JPM covering Safety Function 2.

S3: The applicant will be directed to raise SIT 1A pressure to clear the low pressure alarm and stop raising pressure prior to bringing in the high pressure alarm per 40OP-9SI03, Safety Injection Tank Operations. This is a bank JPM covering Safety Function 3.

S4: The applicant will be directed to respond to a loss of Nuclear Cooling Water per 40AO-9ZZ03, Loss of Cooling Water. The loss is due to an NC CIV that has failed closed. When the applicant attempts to reopen the valve it will not reopen requiring a manual Reactor trip, securing of all 4 RCPs, and isolating Seal Bleedoff from each RCP. One of the Seal Bleedoff isolation valves will be seized open, requiring isolation via alternate means. This is a bank alternate path JPM covering Safety Function 4P.

S5: The applicant will be directed to respond to a partial Main Turbine load rejection per 40AO-9ZZ08, Load Rejection. The examinee will have to restore RCS Tave/Tref mismatch using either CEAs. After Tave/Tref mismatch has been restored, the applicant will lower the Main Turbine Load Set Potentiometer to put the Load Limit Potentiometer back in control of Main Turbine load, however the Load Limit Limiting light will fail to illuminate requiring the applicant to use diverse indications to realize the potentiometer is back in control of Main Turbine load (PV OE from Oct 2021). This is a new alternate path JPM covering Safety Function 4S.

S6: The applicant will be directed to transfer 13.8 kV Bus NAN-S01 to NAN-S03 per 40OP-9NA03, 13.8 kV Electrical Systems. When the transfer is made, NAN-S03 will fault resulting in a loss of 2 RCPs. The Reactor will fail to automatically trip, requiring the applicant to recognize the ATWS in progress and manually trip the Reactor. This is a modified alternate path JPM covering Safety Function 6.

S7: The applicant will be directed to place Train A LPSI on SDC per Appendix 23, SDC Initiation. The A EW Pump will trip when started (third step of Appendix 23), requiring the applicant to cross-tie NC to EW using Appendix 243, NC Cross-Tie to EW Train A. This is a new alternate path JPM covering Safety Function 8.

S8: The applicant will be directed to restore Control Room Ventilation to a normal lineup following a CREFAS actuation per 40OP-9HJ01, Control Building HVAC. This is a new JPM covering Safety Function 9.

P1: The applicant will be directed to perform (simulate) manual valve operations in the field for cross-tying NC to Train A EW per Appendix 243-A, NC Cross-Tie to EW Train A. This is a new JPM covering Safety Function 8.

P2: The applicant will be directed to perform (simulate) 40AO-9ZZ18, Shutdown Outside Control Room, Appendix B, Turbine Building Actions. The applicant will verify the status of electrical buses and Main Turbine equipment following a Control Room evacuation. The applicant will identify two lube oil pumps which failed to auto start, and simulate manually starting the pumps. This is a bank alternate path JPM covering Safety Function 4S.

P3: The applicant will be directed to restore the B EDG during a Blackout per Appendix 56, Restore DG B to PBB-S04. The applicant will determine that the B EDG has tripped due to overspeed and will take

action to simulate resetting the intake air butterfly valve, ensure it is relatched, and locally start the B EDG. This is a modified JPM covering Safety Function 6.

Form 3.2-2 Instructions for Control Room-Plant Systems Outline

1.

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

License Level Control Room In-Plant Total Reactor Operator (RO) 8 3

11 Senior Reactor Operator-Instant (SRO-I) 7 3

10 Senior Reactor Operator-Upgrade (SRO-U) 2 or 3 3 or 2 5

2.

Select safety functions and system for each JPM as follows:

Refer to Section 1.9 pf the applicable knowledge and abilities (K/A) catalog for the plant systems organized by safety function. For pressurized-water reactor operating tests, the primary and secondary systems listed under Safety Function 4, Heat Removal from Reactor Core, in Section 1.9 of NUREG-1122 or NUREG-2103 may be treated as separate safety functions (i.e., two systems, one primary and one secondary, may be selected from Safety Function 4).

From the safety function groupings identified in the K/A catalog, select the appropriate number of plant systems by safety functions to be evaluated based on the applicants license level (see the table in step 1).

The emergency and abnormal plant evolutions listed in Section 1.10 of the applicable K/A catalog may also be used to evaluate the applicable safety function (as specified for each emergency and abnormal plant evolution in the first tier of the written examination outlines in ES-4.1, Preparing Written Examination Outlines).

For RO/SRO-I applicants: Each of the control room systems JPMs and, separately, each of the in-plant systems JPMs must evaluate a different safety function, and the same system or evolution cannot be used to evaluate more than one safety function in each location. One of the control room systems JPMs must be an engineered safety feature.

For the SRO-U applicants: Evaluate SRO-U applicants on five different safety functions. One of the control room systems JPMs must be an engineered safety feature, and the same system or evolution cannot be used to evaluate more than one safety function.

3.

Select a task for each JPM that supports, either directly or indirectly and in a meaningful way, the successful fulfillment of the associated safety function. Select the task from the applicable K/A catalog or the facility licensees site-specific task list.

If this task has an associated K/A, the K/A should have an importance rating of at least 2.5 in the RO column. K/As that have importance ratings of less than 2.5 may be used if justified based on plant priorities; inform the NRC chief examiner if selecting K/As with an importance rating less than 2.5.

The selected tasks must be different from the events and evolutions conducted during the simulator operating test and tasks tested on the written examination. A task that is similar to a simulator scenario event may be acceptable if the actions required to complete the task are significantly different from those required in response to the scenario event.

Apply the following specific task selection criteria:

At least one of the tasks shall be related to a shutdown or low-power condition.

Four to six of the tasks for RO and SRO-I applicants shall require execution of alternative paths within the facility licensees operating procedures. Two to three of the tasks for SRO-U applicants shall require execution of alternative paths within the facility licensees operating procedures.

At least one alternate path JPM must be new or modified from the bank.

At least one of the tasks conducted in the plant shall require the applicant to enter the radiologically controlled area.

This provides an excellent opportunity for the applicant to discuss or demonstrate radiation control administrative subjects.

If it is not possible to develop or locate a suitable task for a selected system, return to step 2 and select a different system

4.

For each JPM, specify the codes for type, source, and location:

Code License Level Criteria RO SRO-I SRO-U (A)lternate path 4-6 (6) 4-6 (6) 2-3 (3)

(C)ontrol room (D)irect from bank 9 (5) 8 (4) 4 (2)

(E)mergency or abnormal in-plant 1 (3) 1 (3) 1 (2)

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

(L)ow power/shutdown 1 (3) 1 (3) 1 (3)

(N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 2 (6 - 3A) 2 (6 - 3A) 1 (3 - 1A)

(P)revious two exams (randomly selected) 3 (0) 3 (0) 2 (0)

(R)adiologically Controlled Area 1 (1) 1 (1) 1 (1)

(S)imulator

Form 3-3.1 Scenario Outline 2022 NRC Scenario #2 Facility:

Palo Verde Scenario: 2 Test:

2022 NRC Exam Examiners:

Operators:

Initial Conditions: 75% power, MOC, B HPSI Pump OOS, B BAMP OOS Turnover: Select CEA Subgroups for RPCB per 40OP-9SF04, Operation of the Reactor Power Cutback System Event Number Event Type*

Event Description CRS OATC BOP 1

N Select CEA Subgroups for RPCB 2

I, TS I

B Safety Channel NI Middle Detector Fails High 3

I I

I TLI #1 Fails High 4

C, TS C

C, MC CEA 43 Slips 50 into the Core 5

C C, MC Loss of NHN-M10 (loss of A BAMP) 6 M

M M, MC ESD from SG #1 Inside Containment (3 min ramp) 7 I

I, MC SG D/P Lockout Failure SG #1

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control Actual Target Quantitative Attributes 1

Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 Entry into a contingency EOP with substantive actions (1 per scenario set) 2 Pre-identified CTs (2 or more)

Form 3-3.1 Scenario Event Summary NRC Exam Scenario #2 2022 NRC Scenario #2 2022 NRC Exam Scenario #2 Overview Event 1 Upon taking the shift, the OATC will select CEA subgroups (for RPCB) per 40OP-9SF04, Operation of the Reactor Power Cutback System.

Event 2 After CEA subgroups have been selected, the Channel B Safety Channel NI will fail high. The CRS will address Technical Specifications and direct the BOP to bypass the associated bistable at the PPS cabinet.

Event 3 After the bistable has been bypassed, TLI #1 will fail high. The CRS will enter 40AO-9ZZ16, RRS Malfunctions, and direct the BOP to select TLI #2 at the RRS cabinet. In response to the failure of TLI #1, the OATC will take CEDMCS out of AUTO, and if directed by the CRS, take SBCS out of REMOTE/AUTO operation.

Event 4 When TLI #2 has been selected and the unit has been stabilized, CEA xxx will slip 50 into the core. The CRS will enter 40AO-9ZZ11, CEA Malfunctions, and direct the crew to commence a Main Turbine load reduction and an RCS boration.

Event 5 When the crew has commenced a boration for the slipped CEA, a loss of 480V MCC, NHN-M10 will occur. The loss of NHN-M10 will result in a loss of power to the A Boric Acid Makeup Pump, leaving no Boric Acid Makeup Pumps available to continue the boration. The crew will either recommence a boration using 40AO-9ZZ01, Emergency Boration, or the CRS will direct a manual Reactor trip.

NOTE - If the Reactor is tripped due to the loss of both BAMPs, the driver will immediately initiate the ESD inside Containment.

Event 6 After (if) the emergency boration is commenced, an ESD from SG #1 will occur inside Containment. The Reactor will either be manually or automatically tripped.

Event 7 The SG D/P lockout on SG #1 will fail to actuate (the lockout would disable AFAS from actuating on SG #1 to prevent feeding a faulted SG), requiring the crew to override AFAS and stop feed flow into the faulted SG.

The crew will monitor for rebound in SG #1 and when rebound occurs, the crew will stabilize Tcold using ADVs on SG #2.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #2 2022 NRC Scenario #2 Critical Task #1: Following an Excess Steam Demand event, stabilize RCS temperature and throttle SI flow as necessary to prevent lifting any Pressurizer Relief Valves Safety Significance: Failure to control RCS pressure below the lift setpoint of a pressurizer relief valve will result in compounding the ESD by unnecessarily creating a loss of coolant event.

Cueing: Rising pressurizer pressure and RCS temperature following termination of the ESD (dryout).

Measurable Performance Indicator: Trending of pressurizer pressure and pressurizer relief valve status to ensure the reliefs did not lift throughout the duration of the event.

Performance Feedback: Red LED pressurizer relief valve position on Board 4 and ERFDADS pressurizer pressure trends.

Critical Task #2: Secure all 4 RCPs within 30 minutes of the loss of cooling water to the RCPs (occurs when CSAS actuates)

Safety Significance: Engineering analysis shows that RCPs can only run for 30 minutes with no cooling water supplied to the seals until seal breakthrough will occur. This would create a new, unisolable LOCA inside Containment.

Cueing: Procedural direction to stop RCPs with no cooling water available is contained in SPTAs.

Measurable Performance Indicator: The CT is met when all 4 RCPs have been stopped as indicated by each pumps red light off and green light on, as well as indication of 0 amps on each pump motor.

Performance Feedback: The crew will have confirmation that the RCPs have stopped by each RCPs red light off and green light on, as well as indication of 0 amps on each pump motor.

NOTE: (Per NUREG-1021) 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 Critical Task identified in the post-scenario review

Form 3-3.1 Driver Set-Up Instructions NRC Exam Scenario #2 2022 NRC Scenario #2 Driver Setup Instructions Reset to IC-612 Run scenario file 2022 NRC Scn #2 Ensure OOS tags hanging on B HPSI and BAMP Pumps Ensure A HPSI Pump is protected

Form 3-3.1 Scenario File Description NRC Exam Scenario #2 2022 NRC Scenario #2 Scenario File IRF crB2SI01SIBP02_1 f:OPEN IRF crB2SI01SIBP02_2 f:RACK_OUT IRF crB4CV08CHNP02B_1 f:OPEN IMF mfNI04B k:2 f:200 IMF cmTRMS03MTNPT11A_1 k:3 r:1 i:535.832 f:839 IMF mfRD02G k:4 f:33 IMF cmBKED12NGNL25C3_5 k:5 IMF mfMS01B k:6 r:3:00 f:10 IMF cmBSRP01BSSG2DPHIAT_1 IMF cmBSRP01BSSG2DPHIBT_1 IMF cmBSRP01BSSG2DPHICT_1 IMF cmBSRP01BSSG2DPHIDT_1

Form 3-3.1 Crew Turnover Sheet NRC Exam Scenario #2 2022 NRC Scenario #2 Plant Conditions:

Unit 1 is operating at 75% power, MOC Equipment Out of Service:

B HPSI Pump OOS for corrective maintenance B BAMP OOS for corrective maintenance Planned Shift Activities:

Select CEA Subgroups for RPCB per 40OP-9SF04, Operation of the Reactor Power Cutback System Maintain power constant at 75% power

Form 3-3.1 Scenario Outline 2022 NRC Scenario #3 Facility:

Palo Verde Scenario: 3 Test:

2022 NRC Exam Examiners:

Operators:

Initial Conditions: 100% power, MOC, B HPSI OOS, B BAMP OOS Turnover: Take Channel C RWT LO trip bistable out of bypass per 40OP-9SB01, Plant Protection System Bypass Operations Event Number Event Type*

Event Description CRS OATC BOP 1

N Take C RWT LO Trip Bistable Out of Bypass 2

TS Inadvertent Closure of RU-1 Inlet, HCB-UV-47 (may get a bean for the BOP if the CRS directs reopening UV-47, but not crediting this until validations show what is likely to happen during the scenario) 3 I

I I

Control Channel NI #1 Fails High 4

I, TS I, MC I, MC Inadvertent B CSAS (UV-671 Fails to Close from Control Room) 5 C

C, MC C, MC Loss of Non-Class Instrument Bus NNN-D15 (RPCB) 6 M

M M

Large Break LOCA Inside Containment 7

Train A Class 4kV Bus PBA-S03 supply transformer NBN-X03 fault (Rx Trip + 10 sec) 8 C

C A Spray Pond Pump Trip (Rx Trip + 5 min)

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control Actual Target Quantitative Attributes 2

Malfunctions after EOP entry (1-2) 3 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 1 Entry into a contingency EOP with substantive actions (1 per scenario set) 2 Pre-identified CTs (2 or more)

Form 3-3.1 Scenario Event Summary NRC Exam Scenario #3 2022 NRC Scenario #3 2022 NRC Exam Scenario #3 Overview Event 1 Upon taking the shift, the BOP will remove the C RWT LO PPS bistable from bypass per 40OP-9SB02, Plant Protection System Bypass Operations.

Event 2 When the bistable has been removed from bypass, Containment Area Radiation Monitor, RU-1, Inlet Isolation Valve, HCB-UV-47, will fail closed. The crew will address the Radiation Monitoring ARP and the CRS will address TS.

Event 3 After Technical Specifications have been addressed, Control Channel NI #1 will fail high.

The CRS will enter 40AO-9ZZ16, RRS Malfunctions, and direct the BOP to select the unaffected Control Channel NI at the RRS cabinet. In response to the failure, the OATC will remove CEDMCS from AUTO and place the affected NI in maintenance to remove it as input to the FWCS.

Event 4 When Control Channel NI #2 has been selected, an inadvertent Train B CSAS will occur. On the CSAS, Train B CS Header Isolation Valve, SIB-UV-671,will open but will then be seized open and will not be able to be closed from the Control Room, requiring the crew to isolate the header via alternate means. The CRS will enter 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations, and direct the OATC to ensure CS Header alternate isolation valves SIB-HV-693 and 695 and closed to stop CS flow (693 normally closed, 695 normally open - will be closed), stop the B CS Pump and stop the B LPSI Pump.

Event 5 After the CRS has addressed Technical Specifications for the CSAS, Non-Class Instrument Bus, NNN-D15, will lose power. The CRS will enter 40AO-9ZZ14, Loss of Non-Class Instrument or Control Power, and direct the crew to trip the A MFP. As a result of the A MFP trip, the CRS will also enter 40AO-9ZZ09, Reactor Power Cutback (LOFP). The crew will take action to stabilize the unit after the MFP trip.

Event 6 When the crew has stabilized the unit, a large break LOCA will occur inside Containment. The Reactor will either be manually tripped or will automatically trip.

Event 7 On the Reactor trip, Train A Class 4kV Bus supply transformer will fault and the bus will be re-energized by the A EDG.

When CSAS actuates, the crew will either have to manually unisolate the Train B CS flowpath by opening SIB-HV-693 or re-opening SIB-HV-695 and re-start the B CS Pump, or will establish CS flow via Train A following restoration of power to Train A 4kV power.

Event 8 5 minutes after the Reactor trip, the A Spray Pond Pump will trip, requiring the crew to trip the A EDG. This will leave the crew will no HPSI flow, causing the CRS to transition to the Functional Recovery procedure. The CRS will direct the crew to cross-tie the B EDG to the Train A 4kV bus in order to restore HPSI flow using the A HPSI Pump (and CS flow if not done previously using Train B equipment realignment).

Form 3-3.1 Critical Task Summary NRC Exam Scenario #3 2022 NRC Scenario #3 Critical Task #1: Re-establish SI flow sufficient to meet Appendix 2 limits within 15 minutes of the restoration of power to PBA-S03 Basis for Bounding Criteria: Normally, we allow 30 minutes for a crew to restore a safety function which is not met (15 minutes to recognize, 15 minutes to restore), however in this case SI flow cannot be restored until power is restored to PBA-S03, therefore we allow 15 minutes to restore flow when power is restored as the time during which power is being restored should allow for the crew to diagnose the failed inventory control safety function.

Safety Significance: The lack of SI flow during a break in the RCS could lead to core uncovery, and ultimately, fuel damage.

Cueing: Procedural direction exists in both SPTAs as well as the LOCA EOP to ensure SI flow is adequate when conditions warrant a SIAS actuation and RCS pressure is less than 1600 psia.

Measurable Performance Indicator: The CT can be confirmed by ensuring the A HPSI Pump is started and injection valves are open. Adequate flow is verified by comparing HPSI flowrates to Appendix 2, Figures (which show the minimum required SI flow depending on current RCS pressure).

HPSI flowrates can be observed on the control boards as well as on any ERFDADS terminal.

Performance Feedback: The crew will have positive confirmation of HPSI flow by observing a red start light on the A HPSI Pump (when manually started) and observing HPSI injection valves being open as well as SI flow on B02 (or ERFDADS) meeting the minimum required flowrates of Appendix 2, Figures.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #3 2022 NRC Scenario #3 Critical Task #2: Establish at least one train of CS flow of at least 4350 gpm within 30 minutes of exceeding the CSAS setpoint (8.06 psig containment pressure)

Basis for Bounding Criteria: Although we allow 15 minutes from the time of power restoration for SI flow to be restored, the crew can restore CS flow prior to power being restored by realigning the Train B CS injection path which was isolated during the inadvertent CSAS prior to the trip. As such, if power restoration to Train A takes > 15 minutes, the crew should not be given the additional time to restore CS flow since they have all needed equipment and indication to restore flow prior to the restoration of Train A power.

Safety Significance: Potential degradation of any barrier to fission product release. Failure to maintain containment temperature and pressure control may challenge containment integrity.

Cueing: Procedural direction as well as indication of containment pressure and temperature exceeding the CSAS setpoint with corresponding indications of inadequate containment spray flow.

Measurable Performance Indicator: The crew will either ensure the B CS Pump is running and open EITHER SIB-HV-693 or SIB-HV-695 to establish flow via the B CS header, OR will ensure the A CS Pump is running and header isolation valve is open following the restoration of power to PBA-S03 within 30 minutes of exceeding 8.06 psig inside Containment.

Performance Feedback: After the crew has re-established CS flow on either train, they can verify adequate CS flow on the control boards or on any ERFDADS terminal.

NOTE: (Per NUREG-1021) 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 Critical Task identified in the post-scenario review

Form 3-3.1 Driver Set-Up Instructions NRC Exam Scenario #3 2022 NRC Scenario #3 Driver Setup Instructions Reset to IC-613 Run scenario file NRC Scenario #3 Ensure OOS tags hanging on B HPSI and BAMP Pumps Ensure A HPSI Pump is protected

Form 3-3.1 Scenario File Description NRC Exam Scenario #3 2022 NRC Scenario #3 Scenario File IRF crB2SI01SIBP02_1 f:OPEN IRF crB2SI01SIBP02_2 f:RACK_OUT IRF crB4CV08CHNP02B_1 f:OPEN IOR diCH_ZDHCBHS47 k:2 c:5 f:CLOSE IMF mfNI02A k:3 r:1 f:125 IMF mfRP06H1 k:4 IMF mfRP06H2 k:4 IMF cmMVRH06SIBUV671_6 k:4 d:20 IMF mfED13C k:5 IMF mfTH01A k:6 f:10 IMF mfED10A e:RPSCHC d:10 IMF mfCC05A e:RPSCHC d:5:00 IRF rfEG04 k:31 f:STOP

Form 3-3.1 Crew Turnover Sheet NRC Exam Scenario #3 2022 NRC Scenario #3 Plant Conditions:

Unit 1 is operating at 100% power, MOC Equipment Out of Service:

B HPSI Pump OOS for corrective maintenance B BAMP OOS for corrective maintenance Planned Shift Activities:

Take Channel C RWT LO trip bistable out of bypass per 40OP-9SB01, Plant Protection System Bypass Operations Maintain power constant at 100% power

Form 3-3.1 Scenario Outline 2022 NRC Scenario #4 Facility:

Palo Verde Scenario: 4 Test:

2022 NRC Exam Examiners:

Operators:

Initial Conditions: 3% power, BOC, B MFP in-service Turnover: Raise Pressurizer Level to 40% per 40OP-9CH01, CVCS Normal Operations Event Number Event Type*

Event Description CRS OATC BOP 1

N Raise Pressurizer Level to 40%

2 I

I, MC DFWCS Feed Flow Transmitter FT-1122Y Fails High 3

C, TS C

C Loss of PBB-S04, B EDG O/P Breaker FTAC (may lose letdown) 4 I, TS I, MC I, MC A SIAS with B CEDM Fans FTAS 5

M M

M B MFP High Vibrations (Trip Initiator) 6 C

C Auto/Manual Reactor Trip Fails at B05 (ATWS) 7 C

C, MC AFA-P01 Overspeed Trip, AFN-P01 Shaft Shear 8

C C

C LOOP with B EDG Trip (Rx Trip +5 min)

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control Actual Target Quantitative Attributes 2

Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 1 Entry into a contingency EOP with substantive actions (1 per scenario set) 2 Pre-identified CTs (2 or more)

Form 3-3.1 Scenario Event Summary NRC Exam Scenario #4 2022 NRC Scenario #4 2022 NRC Exam Scenario #4 Overview Event 1 Upon taking the shift, the OATC will place RCN-LIC-110, Pressurizer Level Control, in LOCAL/AUTO and raise Pressurizer level from 33% to 40% using the local setpoint thumbwheel.

Event 2 When Pressurizer level has been raised, DFWCS SG #2 Feed Flow transmitter FT-1122Y will fail high. The BOP will address the B06 ARP and take action to restore feed flow to SG #2 and place the faulty transmitter in maintenance.

Event 3 After the faulty transmitter has been placed in maintenance, a loss of Class 4kV Bus PBB-S04 will occur (supply transformer will relay) and the B EDG Output Breaker will fail to auto close to re-energize the bus. The CRS will enter 40AO-9ZZ12, Degraded Electrical Power, and direct the BOP to attempt to manually close the EDG Output Breaker to re-energize the bus. The OATC may manually start the E Charging Pump to maintain letdown in service. If the E Charging Pump is not started within ~ 2 minutes, letdown will isolate and the CRS will also enter 40AO-9ZZ05, Loss of Charging or Letdown. If letdown isolates, the CRS will likely direct the OATC to restore letdown using 40AO-9ZZ05, Loss of Charging or Letdown.

Event 4 After PBB-S04 has been re-energized and (if applicable) letdown has been restored, an inadvertent Train A SIAS Leg 2-4 will occur. The CRS will enter 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations, and direct the crew override and stop SI pumps which started as a result of the signal. Additionally, the standby CEDM fans will fail to auto start (running fans trip on the A SIAS), requiring the BOP to restart CEDM fans.

The CRS will also address Technical Specifications due to the inadvertent SIAS.

Event 5 When the plant has been stabilized, the B MFP will start vibrating abnormally. The BOP will address the ARP and dispatch AOs to investigate locally. The MFP will eventually reach trip criteria and the crew will attempt to manually trip the Reactor.

Following the Reactor trip, the BOP will trip the B MFP and transition to an alternate feed source.

Event 6 The Reactor will fail to trip from B05 (either manually or automatically) and the crew will trip the Reactor from B01 by opening the L03 and L10 breakers to de-energize the CEDMCS MG Sets.

Event 7 During SPTAs, the BOP will attempt to restore feed flow to at least one SG, however if AFA-P01 is used, the pump will trip on overspeed, and if AFN-P01 is used, the pump will not produce any flow due to a sheared shaft. The crew will be successful in restoring feed when AFB-P01 is used, however AFB-P01 will not remain in service due to the impending loss of PBB-S04 (will be restored later in the scenario).

Event 8 5 minutes after the Reactor trip, a loss of power will occur and 30 seconds later, the B EDG will trip, resulting in a complete loss of feed flow. The CRS will transition to the Functional Recovery procedure and direct the crew to re-energize PBB-S04 by cross-tying the A EDG to PBB-S04 in order to start (or re-start) AFB-P01 to restore feed to the SGs.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #4 2022 NRC Scenario #4 Critical Task # 1: Following a failure of the Reactor to automatically and/or manually trip using the Reactor Trip Pushbuttons, trip the Reactor by de-energizing load centers L03 and L10 at Board 1 in the Control Room prior to performing step 2 of SPTAs, Vital Auxiliaries verification Safety Significance: Failure to ensure the reactor is tripped following an automatic reactor trip signal during an excess steam demand event, could result in exceeding power limits and fuel temperatures.

Cueing: Failure of the reactor to trip using the reactor trip pushbuttons and/or RPS reactor trip setpoints being exceeded as indicated by red RPS alarms on Board 5 in the Control Room.

Measurable Performance Indicator: Open the feeder breakers for L03 and L10 on Board 1 in the control room.

Performance Feedback: All CEAs inserted as indicated by rod bottom lights on Board 4 in the Control Room, lowering reactor power and a negative startup rate.

Critical Task #2: Restore feed to at least one SG prior to lifting a Primary Safety Valve Safety Significance: If operator action is not taken to restore the loss of the RCS Heat Removal safety function, the RCS will heat up and pressure will rise until a primary safety valve lifts, essentially creating a self-imposed LOCA and challenging RCS inventory control to compound the loss of heat removal.

Cueing: The crew will have indication of the loss of feed on ERFDADS (indicating 0 gpm feed flow) due to the overspeed trip light for AFA-P01, the low amps and no discharge pressure on AFN-P01, the trip of the B MFP, and the inability to use either the A MFP or Condensate Pumps due to the loss of offsite power. Procedure direction to restore feed exists in SPTAs, as well as the Functional Recovery procedure which also contains steps to restore power in order to restore feed using AFB-P01.

Measurable Performance Indicator: The crew will have to close breakers to connect the Train A and Train B 4kV buses, start AFB-P01, and open flow control valves to commence feeding at least one SG.

Performance Feedback: When the crew has restored power to AFB-P01, started AFB-P01 and aligned a feed path to at least one SG, the crew will have indication of feed flow to at least one SG as well as a rising trend on SG level(s), and depending on feed flow rate, a lowering trend on RCS temperature.

NOTE: (Per NUREG-1021) 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 Critical Task identified in the post-scenario review

Form 3-3.1 Driver Set-Up Instructions NRC Exam Scenario #3 2022 NRC Scenario #4 Driver Setup Instructions Reset to IC-614 Run scenario file 2022 NRC Scn #4 Ensure DFWCS alarms are acknowledged

Form 3-3.1 Scenario File Description NRC Exam Scenario #4 2022 NRC Scenario #4 Scenario File IMF mfRP04A IMF mfRP04C IMF mfRD12A IMF cmTRFW04SGNFT1122Y_1 k:2 r:5:00 i:0.296053 f:10 IMF mfED10B k:3 IMF cmBKEG03PBBS04B_2 k:3 IMF mfRP06C2 k:4 d:5 IMF cmCPCH03HCNA02B_5 k:4 IMF cmCPCH03HCNA02D_5 k:4 IMF mfFW15B k:5 r:5:00 f:100 IMF cmCPFW07AFNP01_1 k:6 IMF mfFW22 k:7 IMF mfED02 e:RPSCHC d:5:00 IMF cmBSEG03DGBPSL8_2 e:RPSCHC d:5:30 IMF cmBSEG03DGBPSL6_2 e:RPSCHC d:5:30 IMF cmBSEG03DGBPSL4_2 e:RPSCHC d:5:30 IMF cmBSEG03DGBPSL10_2 e:RPSCHC d:5:30

Form 3-3.1 Crew Turnover Sheet NRC Exam Scenario #4 2022 NRC Scenario #4 Plant Conditions:

Unit 1 is operating at 3% power The B MFP is in service with DFWCS in AUTO in 1-element control Equipment Out of Service:

None Planned Shift Activities:

Raise Pressurizer Level to 40% per 40OP-9CH01, CVCS Normal Operations Maintain power at 3% power

Facility:

2022 E

A V

M P

E I

P N

T N

L T

O I

I T

M C

T A

U A

Y S

A B

S A

B S

A B

S A

B L

M N

P R

T O

R T

O R

T O

R T

O T

E O

C P

O C

P O

C P

O C

P RX 0

1 1

0 R1 NOR 1

1 2

1 1

1 I/C 3,4,5 3,4,6,8 7

4 4

2 MAJ 6

5 2

2 2

1 Man. Ctrl.

4,5 4

3 1

1 0

9 TS 0

0 2

2 RX 0

1 1

0 R2 NOR 2

1 1

1 1

I/C 3,4,5,8 2,3,4,7

,8 9

4 4

2 MAJ 6

5 2

2 2

1 Man. Ctrl.

4,5 2,4,7 5

1 1

0 10 TS 0

0 2

2 RX 0

1 1

0 R3 NOR 1

1 2

1 1

1 I/C 3,4,7 3,4,6,8 7

4 4

2 MAJ 5

5 2

2 2

1 Man. Ctrl.

3,7 4

3 1

1 0

9 TS 0

0 2

2 RX 0

1 1

0 R4 NOR 0

1 1

1 I/C 2,3,4,6 2,3,4,7

,8 9

4 4

2 MAJ 5

5 2

2 2

1 Man. Ctrl.

2,3 2,4,7 5

1 1

0 9

TS 0

0 2

2 RX 0

1 1

0 R5 NOR 1

1 2

1 1

1 I/C 3,4,5 3,4,6,8 7

4 4

2 MAJ 6

5 2

2 2

1 Man. Ctrl.

4,5 4

3 1

1 0

9 TS 0

0 2

2 RX+NOR

+I/C RO SRO-I SRO-U 1

2 3

4 POSITION POSITION POSITION POSITION RX+NOR

+I/C RX+NOR

+I/C RX+NOR

+I/C RX+NOR

+I/C Scenarios PVNGS 5/2/2022 Operating Test Number:

Form 3.4-1 Events and Evolutions Checklist (Based on running Scenarios 4 / 3 / 1 on Mon, Tues, Wed)

Date of Exam:

Facility:

2022 E

A V

M P

E I

P N

T N

L T

O I

I T

M C

T A

U A

Y S

A B

S A

B S

A B

S A

B L

M N

P R

T O

R T

O R

T O

R T

O T

E O

C P

O C

P O

C P

O C

P RX 0

1 1

0 R6 NOR 2

1 1

1 1

I/C 3,4,5,8 2,3,4,7

,8 9

4 4

2 MAJ 6

5 2

2 2

1 Man. Ctrl.

4,5 2,4,7 5

1 1

0 10 TS 0

0 2

2 RX 0

1 1

0 R7 NOR 1

2 2

1 1

1 I/C 3,4,7 3,4,5,8 7

4 4

2 MAJ 5

6 2

2 2

1 Man. Ctrl.

3,7 4,5 4

1 1

0 9

TS 0

0 2

2 RX 0

1 1

0 R8 NOR 1

1 1

1 1

I/C 2,3,4,6 3,4,5 7

4 4

2 MAJ 5

6 2

2 2

1 Man. Ctrl.

2,3 4,5 4

1 1

0 8

TS 0

0 2

2 RX 0

1 1

0 R9 NOR 2

1 1

1 1

I/C 3,4,5,8 2,3,4,7

,8 9

4 4

2 MAJ 6

5 2

2 2

1 Man. Ctrl.

4,5 2,4,7 5

1 1

0 10 TS 0

0 2

2 RX 0

1 1

0 R10 NOR 1

1 1

1 1

I/C 2,3,4,6 3,4,5 7

4 4

2 MAJ 5

6 2

2 2

1 Man. Ctrl.

2,3 4,5 4

1 1

0 8

TS 0

0 2

2 RX+NOR

+I/C RX+NOR

+I/C SRO-U RX+NOR

+I/C RX+NOR

+I/C RO SRO-I 1

2 3

4 RX+NOR

+I/C POSITION POSITION POSITION POSITION PVNGS Date of Exam: 5/2/2022 Operating Test Number:

Scenarios Form 3.4-1 Events and Evolutions Checklist

Facility:

2022 E

A V

M P

E I

P N

T N

L T

O I

I T

M C

T A

U A

Y S

A B

S A

B S

A B

S A

B L

M N

P R

T O

R T

O R

T O

R T

O T

E O

C P

O C

P O

C P

O C

P RX 0

1 1

0 I1 NOR 1

1 1

1 1

I/C 2,3,4,6

,7 3,4,5,8 3,4,6,8 13 4

4 2

MAJ 5

6 5

3 2

2 1

Man. Ctrl.

4 1

1 1

0 14 TS 2,4 2,4 4

0 2

2 SRO-U RX+NOR

+I/C POSITION POSITION POSITION POSITION RO SRO-I Scenarios 1

2 3

4 Form 3.4-1 Events and Evolutions Checklist PVNGS Date of Exam: 5/2/2022 Operating Test Number:

Facility:

2022 E

A V

M P

E I

P N

T N

L T

O I

I T

M C

T A

U A

Y S

A B

S A

B S

A B

S A

B L

M N

P R

T O

R T

O R

T O

R T

O T

E O

C P

O C

P O

C P

O C

P RX 0

1 1

0 U1 NOR 0

1 1

1 I/C 3,4,5,8 2,3,4,6

,7,8 10 4

4 2

MAJ 6

5 2

2 2

1 Man. Ctrl.

0 1

1 0

10 TS 2,4 3,4 4

0 2

2 RX 0

1 1

0 U2 NOR 0

1 1

1 I/C 2,3,4,6

,7 2,3,4,6

,7,8 11 4

4 2

MAJ 5

5 2

2 2

1 Man. Ctrl.

0 1

1 0

11 TS 2,4 3,4 4

0 2

2 RX 0

1 1

0 U3 NOR 0

1 1

1 I/C 3,4,5,8 2,3,4,6

,7,8 10 4

4 2

MAJ 6

5 2

2 2

1 Man. Ctrl.

0 1

1 0

10 TS 2,4 3,4 4

0 2

2 RX 0

1 1

0 U4 NOR 0

1 1

1 I/C 2,3,4,6

,7 3,4,5,8 9

4 4

2 MAJ 5

6 2

2 2

1 Man. Ctrl.

0 1

1 0

9 TS 2,4 2,4 4

0 2

2 RX 0

1 1

0 U5 NOR 1

1 1

1 1

I/C 3,4,7 2,3,4,6

,7,8 9

4 4

2 MAJ 5

5 2

2 2

1 Man. Ctrl.

3,7 2

1 1

0 10 TS 3,4 2

0 2

2 RX+NOR

+I/C RX+NOR

+I/C RX+NOR

+I/C SRO-U RX+NOR

+I/C RX+NOR

+I/C RO SRO-I PVNGS Date of Exam: 5/2/2022 Operating Test Number:

Scenarios 1

2 3

4 Form 3.4-1 Events and Evolutions Checklist POSITION POSITION POSITION POSITION

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