ML23143A208

7-PV-2023-05 Final Outlines

ML23143A208
Person / Time
Site:	Palo Verde
Issue date:	05/11/2023
From:	Heather Gepford Operations Branch IV
To:	Arizona Public Service Co
References
Download: ML23143A208 (1)

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 Note: Systems / E/APEs in RED are sampled on the SRO Exam.

Facility: PVNGS Date of Exam: May 2023 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

1 1

1 1

2 8

4 Tier Totals 5

4 4

4 4

5 26 10

2.

Plant Systems 1

2 2

2 3

3 3

3 3

3 2

2 28 5

2 1

0 1

1 1

1 1

1 1

0 1

9 3

Tier Totals 3

3 4

4 4

4 4

4 3

2 2

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)

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

K/A Topic(s)

IR 000007 (EPE 7) Reactor Trip X

Ability to determine and/or interpret the following as they apply to a Reactor Trip: (CFR:

41.7 / 45.5 / 45.6)

EA2.12 AC distribution availability 3.7 1

(CE E02) Standard Post-Trip Actions and Reactor Trip Recovery / 1 X

Knowledge of the relationship between Standard Post-Trip Actions and Reactor Trip Recovery and the following systems or components: (CFR: 41.7 / 41.8 / 45.4 / 45.7 /

45.8)

EK2.16 SGS 3.9 2

000008 (APE 8) Pressurizer Vapor Space Accident / 3 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)

EA1.07 CCS 3.5 3

000011 (EPE 11) Large Break LOCA / 3 X

2.4.35 Knowledge of nonlicensed operator responsibilities during an emergency (CFR:

41.10 / 43.1/ 43.5 / 45.13) 3.8 4

000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Reactor Coolant Pump Malfunctions: (CFR: 41.8 / 41.10 / 45.3)

AK1.01 Natural Circulation 3.9 5

000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 X

Knowledge of the reasons for the following responses and/or actions as they apply to the Loss of the Residual Heat Removal System:

(CFR: 41.5 / 41.10 / 45.6 / 45.13)

AK3.07 Restoring RHR flow 4.0 6

000026 (APE 26) Loss of Component Cooling Water / 8 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)

AA1.01 PZR heaters, sprays, and PORVs 3.8 7

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

Knowledge of the reasons for the following responses and/or actions as they apply to an Anticipated Transient Without Scram: (CFR:

41.5 / 41.10 / 45.6 / 45.13)

EK3.12 Actions contained in an EOP for ATWS 4.3 8

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 / 41.8 /

45.4 / 45.7 / 45.8)

EK2.14 PZR LCS 3.5 9

000040 (APE 40) Steam Line Rupture / 4 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to a Steamline Rupture:

(CFR: 41.8 / 41.10 / 45.3)

AK1.09 RCS temperature control after the faulted S/G dries out 4.0 10

Rev. 12 (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)

EK1.07 RCS temperature control following blowdown using the least effected S/G 4.0 11 000054 (APE 54) Loss of Main Feedwater /4 X

Ability to operate and/or monitor the following as they apply to Loss of Main Feedwater: (CFR:

41.7 / 45.5 / 45.6)

AA1.05 MFW regulating control valves 3.3 12 (CE E06) Loss of Feedwater /4 X

2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures (CFR: 41.10 / 43.2 / 45.6) 4.5 13 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 X

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

(CFR: 43.5 / 45.13)

AA2.44 Indications of a LOOP 4.1 14 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 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)

AA2.01 Verification that alternate power sources have come online 3.4 15 000062 (APE 62) Loss of Service Water / 4 X

Knowledge of the reasons for the following responses and/or actions as they apply to Loss of Service Water: (CFR: 41.4 / 41.8 / 45.7)

AK3.03 Guidance actions contained in AOPs for loss of service water 3.9 16 000065 (APE 65) Loss of Instrument Air / 8 X

2.4.6 Knowledge of emergency and abnormal operating procedures major action categories (CFR: 41.10 / 43.5 / 45.13) 3.7 17 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X

Knowledge of the relationship between Generator Voltage and Electric Grid Disturbances and the following systems or components: (CFR: 41.4 / 41.5 / 41.7 / 41.10 /

45.8)

AK2.12 AC electrical distribution 3.7 18 K/A Category Totals:

3 3

3 3

3 3

Group Point Total:

18

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 4 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)

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

K/A Topic(s)

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

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Continuous Rod Withdrawal: (CFR: 41.8 / 41.10 / 45.3)

AK1.05 Effects of turbine-reactor power mismatch on rod control 3.7 19 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 X

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

41.10 / 45.3) 4.2 20 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 X

Knowledge of the reasons for the following responses and/or actions as they apply to Fuel Handling Incidents: (CFR: 41.10 / 45.6 / 45.13)

AK3.05 Establishing ventilation alignments 3.6 21 000037 (APE 37) Steam Generator Tube Leak /

3 000051 (APE 51) Loss of Condenser Vacuum / 4 X

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

AK2.13 SDS 3.5 22 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) Control Room Evacuation / 8 X

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Control Room Evacuation: (CFR: 41.8 / 41.10 / 45.3)

AK1.02 Control room habitability 4.1 23 000069 (APE 69) Loss of Containment Integrity /

5 000074 (EPE 74) Inadequate Core Cooling / 4 X

Ability to operate and/or monitor the following as they apply to Inadequate Core Cooling:

(CFR: 41.7 / 45.5 / 45.6)

EA1.21 Secondary inventory control 3.3 24 000076 (APE 76) High Reactor Coolant Activity /

9 X

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

AA2.05 CVCS letdown flow rate 2.8 25 000078 (APE 78) RCS Leak / 3 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery X

2.4.20 Knowledge of the operational implications of emergency and abnormal operating procedures warnings, cautions, and notes (CFR: 41.10 / 43.5 / 45.13) 3.8 26

Rev. 12 (CE E13) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:

2 1

1 1

1 2

Group Point Total:

8

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

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 2.1.27 Knowledge of system purpose and/or function (CFR: 41.7) 3.9 27 004 (SF1; SF2 CVCS) Chemical and Volume Control X

Knowledge of the physical connections and/or cause and effect relationships between the Chemical and Volume Control System and the following systems: (CFR:

41.3 / 41.5 to 41.8 / 41.10)

K1.18 CCWS 3.3 28 005 (SF4P RHR) Residual Heat Removal X

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

K3.06 CSS 3.6 29 006 (SF2; SF3 ECCS) Emergency Core Cooling X

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Emergency Core Cooling System: (CFR:

41.5 / 45.7)

K5.13 Hot-leg injection 3.6 30 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X

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

A1.02 PRT/quench tank pressure 3.1 31 008 (SF8 CCW) Component Cooling Water X

Ability to monitor automatic features of the Component Cooling Water System, including: (CFR: 41.7 / 45.5)

A3.08 Automatic actions associated with the CCWS that occur as a result of an ESFAS signal 4.1 32 010 (SF3 PZR PCS) Pressurizer Pressure Control X

Knowledge of the physical connections and/or cause and effect relationships between the Pressurizer Pressure Control System and the following systems: (CFR:

41.2 to 41.9 / 45.7 / 45.8)

K1.06 CVCS 3.7 33 012 (SF7 RPS) Reactor Protection X

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

K4.03 Protection and control signals 4.1 34 013 (SF2 ESFAS) Engineered Safety Features Actuation X

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Engineered Safety Features Actuation System: (CFR:

41.6 / 41.7 / 41.8 / 45.5 to 45.8)

K6.04 Trip setpoint calculators 3.5 35 022 (SF5 CCS) Containment Cooling X

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

(CFR: 41.7 / 45.5)

A3.01 Initiation of ESFAS mode of operation 4.2 36 026 (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)

A2.08 When to secure CSS 4.0 37

Rev. 12 039 (SF4S MSS) Main and Reheat Steam X

Knowledge of Main and Reheat Steam System design features and/or interlocks that provide for the following: (CFR: 41.7)

K4.06 Prevent reverse steam flow on steamline break 3.5 38 059 (SF4S MFW) Main Feedwater X

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

41.7 / 45.5)

A3.04 MFW turbine 3.4 39 061 (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 40 062 (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)

K5.03 Paralleling between two AC sources 3.5 41 063 (SF6 ED DC) DC Electrical Distribution X

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

A1.04 Battery charger voltage and/or current 3.3 42 064 (SF6 EDG) Emergency Diesel Generator X

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

K2.02 Fuel oil pumps 3.2 43 073 (SF7 PRM) Process Radiation Monitoring X

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

45.9)

A4.02 RMS control panel 3.6 44 076 (SF4S SW) Service Water X

191004 Centrifugal Pumps K1.08 Purpose of starting a pump with discharge valve closed 2.6 45 078 (SF8 IAS) Instrument Air X

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Instrument Air System: (CFR: 41.7 / 45.7)

K6.15 Cooling water 2.8 46 103 (SF5 CNT) Containment X

Ability to (a) predict the impacts of the following on the Containment 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)

A2.10 FHS malfunctions 2.6 47 Multiple-sampled systems start here.

010 (SF3 PZR PCS) Pressurizer Pressure Control X

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Pressurizer Pressure Control System: (CFR: 41.7 /

45.7)

K6.03 PZR sprays and heaters 3.8 48 022 (SF5 CCS) Containment Cooling X

Ability to (a) predict the impacts of the following on the Containment Cooling 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)

A2.04 Cooling water system malfunction 3.7 49

Rev. 12 026 (SF5 CSS) Containment Spray X

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

K2.01 Containment spray pumps 3.9 50 059 (SF4S MFW) Main Feedwater X

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

K3.02 AFW system 3.8 51 064 (SF6 EDG) Emergency Diesel Generator X

Ability to predict and/or monitor changes in parameters associated with operation of the Emergency Diesel Generators, including: (CFR: 41.5 / 45.5)

A1.11 Fuel oil storage, day tank levels, and/or temperatures 3.3 52 073 (SF7 PRM) Process Radiation Monitoring X

Knowledge of Process Radiation Monitoring System design features and/or interlocks that provide for the following:

(CFR: 41.7)

K4.02 System actuations based on PRM signals 3.7 53 076 (SF4S SW) Service Water X

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Service Water System: (CFR: 41.4 / 41.7 /

45.5)

K5.05 Radiation alarms on SWS 3.1 54 K/A Category Point Totals:

2 2

3 3

3 3

3 3

2 2

2 Group Point Total:

28

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

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

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive X

2.2.3 (Multi-unit license) Knowledge of the design, procedural, and/or operational differences between units (CFR: 41.5 / 41.6

/ 41.7 / 41.10 / 45.12) 3.8 55 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication X

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

K3.02 Plant Computer 3.2 56 015 (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 X

Knowledge of Spent Fuel Pool Cooling System design features and/or interlocks that provide for the following: (CFR: 41.7)

K4.01 Maintaining spent fuel level at specified levels 3.6 57 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Steam Dump System and Turbine Bypass Control: (CFR: 41.5 / 45.7)

K5.04 Basis for plant cooldown rates 3.4 58 045 (SF 4S MTG) Main Turbine Generator X

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

A1.10 Turbine valve indicators (throttle, governor, control, stop, and intercept) 3.2 59 050 (SF 9 CRV) Control Room Ventilation 055 (SF4S CARS) Condenser Air Removal X

Ability to monitor automatic features of the Condenser Air Removal System, including:

(CFR: 41.7 / 45.5)

A3.01 Air removal pump 3.0 60 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste X

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Liquid Radwaste System: (CFR: 41.7 / 45.7 to 45.9)

K6.13 CWS 2.6 61 071 (SF9 WGS) Waste Gas Disposal

Rev. 12 072 (SF7 ARM) Area Radiation Monitoring X

Knowledge of the physical connections and/or cause and effect relationships between the Area Radiation Monitoring System and the following systems: (CFR:

41.7 to 41.9 / 45.8 / 9 / 11)

K1.03 Fuel building isolation 3.6 62 075 (SF8 CW) Circulating Water X

Ability to (a) predict the impacts of the following on the Circulating Water 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)

A2.12 Main condenser tube leak 3.5 63 086 Fire Protection K/A Category Point Totals:

1 0

1 1

1 1

1 1

1 0

1 Group Point Total:

9

Rev. 12 Form 4.1-COMMON Common Examination Outline Facility: PVNGS Date of Exam: May 2023 Generic Knowledge and AbilitiesTier 3 (RO)

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.17 Ability to make accurate, clear, and concise verbal reports (CFR: 41.10 / 45.12 / 45.13) 3.9 64 2.1.39 Knowledge of conservative decision-making practices (CFR: 41.10 / 43.5 / 45.12) 4.3 65 Subtotal N/A N/A

2.

Equipment Control 2.2.13 Knowledge of tagging and clearance procedures (CFR:

41.10 / 43.1 / 45.13) 4.1 66 2.2.43 Knowledge of the process used to track inoperable alarms (CFR: 41.10 / 43.5 / 45.13) 3.0 67 Subtotal N/A N/A

3.

Radiation Control 2.3.12 Knowledge of radiological safety principles and procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, or alignment of filters (CFR: 41.12 / 43.4 / 45.9 / 45.10) 3.2 68 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 69 Subtotal N/A N/A Tier 3 Point Total 6

7 TheoryTier 4 (RO)

Category K/A #

Topic RO IR Reactor Theory 192004 K1.02 Describe the effect on the magnitude of the temperature coefficient of reactivity from changes in the following: moderator temperature 3.1 70 192005 K1.03 Predict direction of change in reactor power for a change in control rod position 3.6 71 192008 K1.14 Describe reactor power and startup rate response prior to reaching the POAH 3.1 72 Subtotal N/A Thermodynamics 193007 K1.04 Describe how the presence of gases or steam can affect heat transfer and fluid flow in heat exchangers 3.0 73 193008 K1.06 Describe critical heat flux 2.9 74 193009 K1.09 Describe the effects of quadrant power tilt (symmetric offset), including long-range effects 3.2 75 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: PVNGS Date of Exam: May 2023 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 18 3

3 6

2 8

2 2

4 Tier Totals 26 5

5 10

2.

Plant Systems 1

28 3

2 5

2 9

1 1

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 13 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)

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

K/A Topic(s)

IR 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 / 45.13)

AA2.14 Subcooling 3.9 76 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X

Ability to determine and/or interpret the following as they apply to Loss of Reactor Coolant Makeup: (CFR: 43.5 / 45.13)

AA2.04 PZR level 3.8 77 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 X

2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls (CFR: 41.10 / 43.2 / 45.6) 4.4 78 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

2.4.29 Knowledge of the emergency plan implementing procedures (CFR: 41.10 / 43.5 /

45.11) 4.4 79 000040 (APE 40) Steam Line Rupture / 4 (CE E05) Excess Steam Demand / 4 000054 (APE 54) Loss of Main Feedwater /4 (CE E06) Loss of Feedwater /4 000055 (EPE 55) Station Blackout / 6 X

Ability to determine and/or interpret the following as they apply to a Station Blackout:

(CFR: 43.5 / 45.13)

EA2.08 In-core thermocouple temperatures 4.0 80 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X

2.1.32 Ability to explain and apply system precautions, limitations, notes, or cautions (CFR: 41.10 / 43.2 / 45.12) 4.0 81 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 K/A Category Totals:

3 3

Group Point Total:

6

Form 4.1-PWR PWR Examination Outline Page 14 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)

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

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 X

2.2.37 Ability to determine operability or availability of safety-related equipment (SRO Only) (CFR: 43.2 / 43.5 / 45.12) 4.7 82 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 X

Ability to determine and/or interpret the following as they apply to Loss of Source Range Nuclear Instrumentation: (CFR: 43.5 /

45.13)

AA2.09 Effect of improper high-voltage setting 3.0 83 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

Ability to determine and/or interpret the following as they apply to a Steam Generator Tube Leak: (CFR: 41.7 / 41.10 / 43.5 / 45.13)

AA2.06 S/G tube failure 4.0 84 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) 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 (CE A16) Excess RCS Leakage / 2 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) 4.6 85 (CE E09) Functional Recovery (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 15 Plant SystemsTier 2/Group 1 (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 X

Ability to (a) predict the impacts of the following on the Residual Heat Removal 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)

A2.05 RHR heat exchanger malfunction 3.5 86 006 (SF2; SF3 ECCS) Emergency Core Cooling X

Ability to (a) predict the impacts of the following on the Emergency Core Cooling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 45.5 /

45.3 / 45.4)

A2.14 Gas accumulation 3.2 87 007 (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 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam X

2.4.41 Knowledge of the emergency action level thresholds and classifications (CFR:

43.5 / 45.11) 4.6 88 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW) Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution 063 (SF6 ED DC) DC Electrical Distribution X

Ability to (a) predict the impacts of the following on the DC Electrical Distribution 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)

A2.01 Grounds 2.8 89 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air X

2.2.37 Ability to determine operability or availability of safety-related equipment (SRO only) (CFR: 43.2 / 43.5 / 45.12) 4.6 90 103 (SF5 CNT) Containment K/A Category Point Totals:

3 2 Group Point Total:

5

Rev. 12

Rev. 12 Form 4.1-PWR PWR Examination Outline Page 17 Plant SystemsTier 2/Group 2 (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 X

Ability to (a) predict the impacts of the following on the Nonnuclear Instrumentation 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 / 45.5)

A2.02 Loss of power supply 3.3 91 017 (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 X

Ability to predict and/or monitor changes in parameters associated with operation of the Fuel Handling Equipment System, including: (CFR: 41.5 / 41.6 / 43.7 / 45.5 /

45.8)

A1.01 Fuel handling equipment load limits 3.0 92 035 (SF 4P SG) Steam Generator X

2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation (CFR: 41.5 /

43.5 / 45.12 / 45.13) 4.7 93 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 050 (SF 9 CRV) Control Room Ventilation 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 086 Fire Protection K/A Category Point Totals:

1 1

1 Group Point Total:

3

Rev. 12 Form 4.1-COMMON Common Examination Outline Facility: PVNGS Date of Exam: May 2023 Generic Knowledge and AbilitiesTier 3 (SRO)

Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement or overtime limitations (reference potential) (CFR: 41.10 / 43.5 / 45.12) 3.9 94 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 95 Subtotal N/A N/A

2.

Equipment Control 2.2.14 Ability to explain the variations in control room layouts, systems, instrumentation, or procedural actions between units at a facility (CFR: 41.6 / 41.7 / 41.10 / 45.1 / 45.13) 4.3 96 2.2.20 Knowledge of the process for managing troubleshooting activities (CFR: 41.10 / 43.5 / 45.13) 3.8 97 Subtotal N/A N/A

3.

Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities, such as analysis and interpretation of radiation and activity readings as they pertain to administrative, normal, abnormal, and emergency procedures, or analysis and interpretation of coolant activity, including comparison to emergency plan or regulatory limits (SRO Only) (CFR: 43.4 / 45.10) 3.8 98 Subtotal N/A N/A

4.

Emergency Procedures/

Plan 2.4.29 Knowledge of the emergency plan implementing procedures (CFR: 41.10 / 43.5 / 45.11) 4.4 99 2.4.14 Knowledge of general guidelines for emergency and abnormal operating procedures usage (CFR: 41.10 / 43.1

/ 45.13) 4.5 100 Subtotal N/A N/A Tier 3 Point Total 6

7

Rev. 12 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 1 / 1 (Q5) 015 AK1.06 There is no procedural guidance for an RCP flywheel, and none of the four Tier 1 question criteria can be met in a question about the operational implications of an RCP flywheel during an RCP malfunction event. Reselected 015 AK1.01.

1 / 1 (Q8) 029 EK3.07 During a ATWS, the Main Turbine is never procedurally directed to be tripped locally. If tripping the Main Turbine fails to work from the Control Room, the EHC Pumps are stopped from the Control Room, and if that fails, MSIVs are closed from the Control Room. Reselected 029 EK3.12.

1 / 1 (Q10) 040 AK1.13 There are no operational or procedural differences between an ATWS with or without a concurrent ATWS at PVNGS.

Reselected 040 AK1.09.

1 / 1 (Q19) 001 AK1.19 In a PWR, void coefficient is virtually negligible and you would not expect to get any voiding due to a continuous rod withdrawal. Reselected 001 AK1.05.

1 / 2 (Q21) 036 AK3.06 Could not create plausible distractors for the reason for placing fuel in a safe location following a fuel handling incident.

Reselected 036 KA3.05.

1 / 2 (Q22) 051 AK2.11 SG Blowdown is not operated or adjusted during a loss of condenser vacuum at PVNGS, and it is not even mentioned in the Loss of Condenser Vacuum AOP. Reselected 051 AK2.13.

1 / 2 (Q24) 074 EA1.23 The PZR Safeties at PVNGS do not have block valves.

Reselected 074 EA1.21.

2 / 1 (Q28) 004 K1.25 PVNGS does not have an excess letdown flowpath (or equivalent). Reselected 004 K1.18.

2 / 1 (Q36) 022 K3.04 Unable to make distractors plausible for the effects of a loss of containment cooling on Containment (temp and press will rise).

Reselected 022 A3.01 at CE direction.

2 / 1 (Q44) 073 A4.03 Operations department does not perform source checks on RMs at PVNGS therefore this KA has no operational relevance.

Reselected 073 A4.02.

2 / 1 (Q51) 059 K3.05 Myself and the Operations Representative could not find any tie between a failure of the Main Feedwater System and Extraction Steam and therefore could not write a question for the selected KA. Reselected 059 K3.02.

2 / 2 (Q55) 001 K2.03 KA replaced at Chief Examiners request based on the recent Ovation Rod Control modification (installed only on Unit 3 as of the exam date). New KA is 001 G 2.2.3 2 / 2 (Q56) 028 K3.01 The Hydrogen Recombiner System was retired in place in fall of 2022 and therefore has no operational relevance. Given the modification in Unit 3 (with Units 1 and 2 to follow in the next 18-24 months), reselected 014 K3.02 to add a newly modified system to the outline in lieu of the retired system.

Rev. 12 2 / 2 (Q63) 075 A2.04 Unable to create a plausible distractor for how to adjust turbine load during excessive heat and/or humidity conditions.

Reselected 075 A2.12.

1 / 1 (Q76) 008 AA2.05 PVNGS does not have PORVs, and the Pressurizer Safety Valves do not have block valves. Reselected 008 AA2.14.

1 / 1 (Q79) 038 G 2.2.40 KA is for LCO actions less than or equal to one hour, however all required actions in LCOs 3.4.14 and 3.4.18 (only ones related to SG tube leakage) are greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Reselected 038 G 2.4.29.

2 / 1 (Q88) 039 G 2.2.3 No procedural or operational differences exist between the units for the Main and Reheat Steam System. Reselected 039 G 2.4.41.

2 / 1 (Q90) 078 G 2.1.25 There arent any graphs, curves, or tables that I am aware of pertaining to operation of the IA system, other than the Loss of IA AOP which has a table showing when each piece of equipment will fail (at what pressure they fail). However providing this table would make the question direct lookup, and asking these failure positions from memory would be a knowledge item, not an ability (as the KA says). Reselected 078 G 2.2.37.

3 (Q100)

G 2.4.51 Unable to write a question that both addressed the KA and met the requirements for a Tier 3 question. Reselected G 2.4.14.

Form 3.2-1 Administrative Topics Outline Facility:

PVNGS Date of Examination:

5/1/2023 Examination Level: RO X

SRO Operating Test Number:

2023 Administrative Topic (Step 1)

Activity and Associated K/A (Step 2)

Type Code (Step 3)

Conduct of Operations (A1)

K/A: 2.1.19 IR: 3.9 JPM

Description:

Perform SIT checks per 40ST-9ZZM3, Operations MODE 3 Surveillance Logs, while in MODE 3 with RCS pressure < 1837 psia.

N, R Conduct of Operations (A2)

K/A: 2.1.25 IR:

JPM

Description:

Determine the maximum allowable time until SDC must be placed in service based on CST level following a Reactor trip and the feed rate for decay heat removal per Appendix 4, CST Level vs Time to Shutdown Cooling.

D, R Equipment Control (A3)

K/A: 2.2.12 IR:

JPM

Description:

Determine RCS leakrate using 40ST-9RC05, Manual Calculation of RCS Water Inventory Balance.

M, R Emergency Plan (A4)

K/A: 2.4.47 IR:

JPM

Description:

Determine time until Minimum Required Power Reduction curve would be violated if power reduction was paused following a slipped CEA.

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)

A1: The applicant will be directed to perform SIT checks per 40ST-9ZZM3, Operations MODE 3 Surveillance Logs, while in MODE 3 with RCS pressure < 1837 psia. The applicant will be provided with pictures the relevant meters and indications then have to determine which operability requirements apply given the conditions in the cue and the pictures.

A2: The applicant will be directed to calculate the rate at which CST level is lowering and compare that to a graph comparing CST level and time after shutdown to determine the maximum amount of time until SDC must be in service per Appendix 4, CST Level vs Time to Shutdown Cooling.

A3: The applicant will be directed to determine RCS leakrate using 40ST-9RC05, Manual Calculation of RCS Water Inventory Balance. This JPM was selected due to the OE from July 2022 in which RC05 was improperly performed.

This JPM was selected due to the OE from July 2022 in which RC05 was improperly performed.

(CR # 22-07824)

A4: The applicant will be directed to determine how long the power reduction can be paused for (following a dropped CEA) to allow for Pressurizer level to lower (as it approaches the high TS limit) based on power reduction requirements and initial power levels per 40AO-9ZZ11, CEA Malfunctions.

Form 3.2-1 Administrative Topics Outline Facility:

PVNGS Date of Examination:

5/1/2023 Examination Level: RO SRO X

Operating Test Number:

2023 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:

Determine stay time inside Containment per 01DP-0IS17, Heat Stress Prevention Program.

M, R Conduct of Operations (A6)

K/A: 2.1.4 IR: 3.8 JPM

Description:

Given a timeline of events at minimum manning, determine when an on-shift SM who failed a random FFD test must be replaced and determine the reporting requirements for the event.

N, R Equipment Control (A7)

K/A: 2.2.23 IR: 4.6 JPM

Description:

Given a timeline of events, determine LCO entry and exit times.

D, R Radiation Control (A8)

K/A: 2.3.14 IR: 3.8 JPM

Description:

Given the need for emergency exposure for equipment repairs, determine who should perform the repair, authorization required, and reportability for the exposure.

N, R Emergency Plan (A9)

K/A: 2.4.41 IR: 4.6 JPM

Description:

Classify an event using the EAL charts.

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)

A5: The applicant will be directed to determine the maximum stay time for a worker entering Containment to perform work, as well as the minimum required rest/recovery time following the work per 01DP-0IS17, Heat Stress Prevention Program. This JPM was selected due to the OE from October 2022 in which three workers experienced heat related illnesses following work inside Containment.

This JPM was selected due to the OE from October 2022 in which three workers experienced heat related illnesses following work inside Containment.

(CR # 22-10526)

A6: The applicant will be directed to determine the impact to site staffing as well as the event reporting requirements following an on-shift SM failing a for-cause FFD evaluation during the work shift.

This JPM was selected following OE from August 2022 in which a licensed operator failed a random FFD test.

A7: The applicant will be directed to determine LCO entry and exit times based on a timeline of events regarding the status of ECCS equipment and an EDG.

A8: The applicant will be directed to determine which worker should be chosen to receive an emergency exposure while performing necessary repairs during an emergency, whose authorization is required, and the reportability requirements (to the NRC) as a result of the emergency exposure.

A9: The applicant will be directed to classify an event using the EAL charts and PVNGS Emergency Plan.

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

PVNGS Date of Examination:

5/1/2023 Operating Test Number:

2023 Exam Level:

X RO X

SRO-I X

SRO-U System / JPM Type Type Code SF Control Room Systems S1 001 A4.17 Restore CEA Group overlap using Ovation Rod Control per 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump)

N, S 1

S2 013 A2.06 Respond to an inadvertent A CSAS per 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations A, D, EN, S 2

S3 010 A2.02 Place the Pressurizer in Boron Equalization A, D, S 3

S4 005 A3.01 Respond to a RAS per 40EP-9EO03, Loss of Coolant Accident A, D, EN, L, S 4P S5 045 A2.02 Swap from B Stator Cooling Water Pump to A Stator Cooling Water Pump.

A, N, S 4S S6 APE 008 AK3.03 Vent the RCS to the RDT to collapse voids per Appendix 15, RCS Void Control A, D, L, S 5

S7 058 AA1.05 Respond to a loss of class 125 VDC control power during A EDG Load Run A, D, S 6

S8 RO Only 015 A4.02 Place Boron Dilution Alarm System in service following a Reactor Trip D, L, S 7

In-Plant Systems P1 EPE 011 EA1.05 Appendix 208-A - Aligning the A Charging Pump for Hot Leg Injection E, EN, L, N, R 2

P2 068 AA1.01 Establish RCS Heat Removal at the Remote Shutdown Panel per 40AO-9ZZ19, Control Room Fire (Performed in RSD Panel Simulator)

E, N 4S P3 055 EA1.06 Appendix 109-A - Energize NHN-M03 Loads from the SBOG E, L, N 6

SRO-U will perform S1, S4, S5, P1, and P3 (Shaded in grey)

RO Only JPM will be S8

S1: The applicant will be directed to restore CEA overlap following a Reactor Power Cutback per 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump). The applicant will determine how far Group 4 CEAs must be withdrawn to re-establish proper CEA group overlap, then use the new Ovation Rod Control System to realign the CEA groups. This is a new JPM covering Safety Function 1.

DUE TO THE OVATION MODIFICATION ONLY BEING INSTALLED IN SIMULATOR A, THIS JPM WILL ONLY BE RUN IN SIMULATOR A.

S2: The applicant will be directed to respond to an inadvertent Train A CSAS per 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations. The applicant will stop the A CS Pump and attempt to override and isolate the A CS Header. The A CS Header Isolation Valve will be seized in the open position, requiring the operator to isolate the header using an alternate isolation flowpath. This is a bank alternate path JPM covering Safety Function 2 S3: The applicant will be directed to place the Pressurizer in Boron Equalization per 40OP-9CH01, CVCS Normal Operations. When the first Pressurizer Backup Heater is energized, the output on Pressure Master Control, RCN-PIC-100, will fail high, resulting in the Main Spray Valves both opening and Pressurizer Pressure lowering. The applicant will address the ARP and take actions to diagnose the failure and regain control of Pressurizer pressure prior to an automatic Reactor trip. This is a bank alternate path JPM covering Safety Function 3.

S4: The applicant will be directed verify a RAS actuation during a large break LOCA per 40EP-9EO03, Loss of Coolant Accident. One of the Train B SDC suction valves from containment will be seized closed requiring the applicant to trip the B HPSI and B CS Pumps. The applicant will then manually close the RWT outlet valves (time critical action to close within 5 minutes of the RAS actuation), then ensure all of the HPSI, LPSI, and CS Pump recirc valves are closed (one will fail to auto close and must be manually closed by the applicant). This is a bank alternate path time-critical JPM covering Safety Function 4P.

S5: The applicant will be directed to swap from B Stator Cooling Water Pump to A Stator Cooling Water Pump per 40OP-9CE01, Stator Cooling System. During the pump swap, the standby pump is started using the auto start test feature prior to securing the running pump. The A Pump (standby, initially) will have a sheared shaft when started. This will not show indications until the B Pump is stopped. When the B Pump is stopped, 2-3 Stator Cooling Water trouble alarms will annunciate and temperatures will begin to rise on Stator Cooling Water cooled equipment. The B Pump will have a fail to auto start malfunction installed so the applicant will have to recognize that the A Pump is not providing sufficient cooling, and manually re-start the B Pump within 70 seconds of stopping the B Pump or the Main Turbine will trip. This is a new alternate path JPM covering Safety Function 4S.

S6: The applicant will be directed to vent the Reactor Vessel Head to the RDT to collapse voids during a LOCA per Appendix 15, RCS Void Control. During the vent alignment, one of the head vent valves will fail to open requiring the applicant to align the vent from RCS to Containment atmosphere instead. This is a bank alternate path JPM covering Safety Function 5.

S7: The applicant will be directed to secure from an EDG load run. During the load reduction, Train A 125 VDC Control Power Bus, PKA-M41, will de-energize resulting in an automatic trip of the A EDG.

The loss of PKA will also result in the loss of control power to the A EDG Output Breaker, resulting in the A EDG being motorized by the grid. The applicant will address either the ARP or AOP (or use guidance in the precautions section of the ST) and direct an AO to locally open the A EDG Output Breaker to prevent damage to the EDG. This is a bank, time critical, alternate path JPM covering Safety Function 6.

S8: The applicant will be directed to place the Boron Dilution Alarm System in service per Appendix 8, Boron Dilution Alarm Check. The applicant will select the Startup Channel NIs in the Control Room, then manipulate switches at the Excore cabinet outside the Control Room and evaluate voltage differences between the two BDAS channels to ensure proper indication. This is a bank JPM covering Safety Function 7.

P1: The applicant will be directed to perform (simulate) the field actions of Appendix 208, LM - Charging Pump Hot Leg Injection Train A HPSI, to establish a valve lineup for hot leg injection. This is a new RCA JPM covering Safety Function 2.

P2: The applicant will be directed to establish a control source of steaming from the SGs and establish Auxiliary Feed flow to the SGs from the Remote Shutdown Panel following control room evacuation due to a fire. This JPM will be performed in the RSD Panel simulator following the applicants identifying the location of the RSD during in-plant JPMs. This is a new JPM covering Safety Function 4S.

P3: The applicant will be directed to establish conditions to re-energize communications equipment following a blackout per Appendix 109-A, Energize NHN-M03 Loads From the SBOG. The applicant will locate panels in the Turbine and Control Buildings and simulate aligning breakers as directed by Appendix 109-A. This is a new 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 (2)

(C)ontrol room (D)irect from bank 9 (6) 8 (5) 4 (1)

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

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

(L)ow power/shutdown 1 (5) 1 (4) 1 (2)

(N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 2 (5 - 1A) 2 (5 - 1A) 1 (4 - 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 2023 NRC Scenario #1 Facility:

PVNGS Scenario: 1 Test:

2023 NRC Exam Examiners:

Operators:

Initial Conditions: 100% power, B CS Pump OOS, B NC Pump OOS.

Turnover: Perform 20 gallon shiftly dilution per 40OP-9CH01, CVCS Normal Operations. Maintain power stable at 100%.

Event Number Event Type*

Event Description CRS OATC BOP 1

R R

Perform a 20 gallon shiftly dilution per 40OP-9CH01, CVCS Normal Operations 2

I, TS I

Containment Pressure (WR) transmitter, HCC-PI-352C, fails high 3

I I, MC Charging Line to RC Loop 2A DP Controller, CHN-PDIC-240, output fails high in Auto 4

C C

Degraded Main Condenser Vacuum 5

C, TS C

C, MC Loss of Train B 4kV Bus PBB-S04 / A & C Containment Normal ACUs Fail to Auto Start 6

C C

C A MFP Trip (RPCB) 7 M

M M

Loss of Offsite Power / A EDG Trip - Loss of Lube Oil (Blackout) 8 I

I, MC AFAS Fails to Auto Actuate

• (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) 5 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)

SPARE

Form 3-3.1 Scenario Event Summary NRC Exam Scenario #1 2023 NRC Scenario #1 2023 NRC Exam Scenario #1 Overview Event 1 Upon taking the shift, the OATC will perform a 20 gallon shiftly dilution per 40OP-9CH01, CVCS Normal Operations.

Event 2 Following the dilution, Containment WR Pressure transmitter, HCC-PI-352C, will fail high. The BOP will address the ARP, the CRS will address Technical Specifications, and direct the BOP to bypass the affected bistable at the PPS cabinet.

Event 3 After the affected bistable has been bypassed, Charging Line to RC Loop 2A DP Controller, CHN-PDIC-240, output will fail high in Auto. The OATC will address the ARP and take manual control of CHN-PDIC-240 and adjust the output to restore DP to the normal band and restore Seal Injection to the normal band.

Event 4 When PDIC-240 has been restored to normal output, Main Condenser vacuum will begin to degrade. The CRS will enter 40AO-9ZZ07, Loss of Condenser Vacuum, and direct the crew to search for the source of the loss of vacuum and take load off the Main Turbine to stabilize vacuum. When the crew starts to reduce turbine load, an AO will report that they discovered an empty loop seal and have commenced filling it.

Event 5 After the crew has lowered Main Turbine load and directed filling the loop seal, a loss of Train B 4kV Bus PBB-S04 will occur. The CRS will enter 40AO-9ZZ12, Degraded Electrical Power. The OATC will either start the E Charging Pump to prevent a loss of letdown (may occur using Prompt and Prudent action), or letdown will isolate on high letdown temperature. If letdown isolates, the CRS will enter 40AO-9ZZ05, Loss of Charging or Letdown, and establish conditions for extended operation with letdown isolated. Additionally, on the loss of PBB-S04, the CEDM and Containment Normal ACUs will lose power. The standby Containment Normal ACUs will fail to auto start, requiring a manual start from the BOP or OATC. The CRS will also address Technical Specifications as a result of the loss of the class bus.

Event 6 After Technical Specifications have been addressed, the A MFP will trip. The CRS will enter 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump), and direct the crew to take RPCB out of service, and remove CEDMCS from automatic.

Event 7 After the crew has stabilized the plant, a loss of offsite power will occur, resulting in a Reactor trip. The A EDG will trip 60 seconds after starting due to a loss of lube oil, putting the unit in a blackout condition. Following SPTAs, the CRS will transition to 40EP-9EO08, Blackout, and direct the crew to start and align an SBOG to PBA-S03 to restore class 4kV power to the unit.

Event 8 AFAS will fail to auto actuate, requiring the BOP to manually start turbine-driven AFW Pump, AFA-P01 to restore feed flow to the unit.

When power has been restored to a class 4kV bus and feed has been restored to at least one SG, the scenario may be terminated.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #1 2023 NRC Scenario #1 Critical Task # 1: Re-establish steaming and/or feeding to at least one SG to prevent lifting a Primary Safety Valve Safety Significance: Failure to reestablish a steaming/feed source to at least one SG will result in the continued heatup of the RCS, potentially to the point at which an otherwise preventable lifting of a primary safety valve, increasing the risk of a stuck-open valve and unisolable steam space leak inside Containment.

Cueing: Procedural direction in SPTAs as well as the Blackout EOP to feed at least one SG and use the SGs to maintain RCS temperature in the post-trip bands, which will directly aid in maintaining RCS pressure in the post-trip bands, preventing the lifting of a primary safety valve.

Measurable Performance Indicator: The status of RCS pressure is able to be read on several control board indications as well as any ERFDADS terminal. The status of the pressurizer safety valves is available on B04. The status of feedwater to the SGs is also available on multiple control board indicators as well on every ERFDADS terminal.

To restore feed:

To start AFA-P01:

o Open EITHER (or both) SGA-HS-134A or SGA-HS-138A to admit steam to AFA-P01 and start the pump To feed SG 1:

o Open AFA-HS-37A and throttle open AFC-HS-33A To feed SG 2:

o Open AFC-HS-36A and throttle open AFA-HS-32A To establish steaming:

For SG 1 Line 1 ADV: Place SGA-HS-184A and SGC-HS-184B to OPEN PERM, then raise output on SGA-HIC-184A until SG 1 Line 1 ADV opens For SG 1 Line 2 ADV: Place SGB-HS-178A and SGD-HS-178B to OPEN PERM, then raise output on SGB-HIC-178A until SG 1 Line 2 ADV opens For SG 2 Line 1 ADV: Place SGB-HS-185A and SGD-HS-185B to OPEN PERM, then raise output on SGB-HIC-185A until SG 2 Line 1 ADV opens For SG 2 Line 2 ADV: Place SGA-HS-179A and SGC-HS-179B to OPEN PERM, then raise output on SGA-HIC-179A until SG 2 Line 2 ADV opens Performance Feedback: The crew can confirm that feed has been restored using the control board indications and/or ERFDADS terminals. The status of the pressurizer safety valves may be monitored using the safety valve indicators on B04.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #1 2023 NRC Scenario #1 Critical Task # 2: Align an SBOG to a Class 4kV Bus within one hour of the blackout condition Safety Significance: Restoration of power to at least one class 4kV bus is required to ensure the class batteries do not drop below minimum voltage values, which would result in an escalation from a site area emergency to a general emergency condition. Failure to align an SBOG within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> also requires entry into 79IS-9ZZ07, PVNGS Extended Loss of All Site AC Guideline MODES 1-4, which represents an escalation in response, increased personnel dispatch, and alignment of FLEX equipment which is otherwise avoidable with minimally competent crew response.

Cueing: Procedural guidance in the Functional Recovery procedure will direct the crew to restore power (the SBOG is the preferred available method to do so in this case).

Measurable Performance Indicator: The CT is met when the crew reenergized at least one class 4kV bus, which will be indicated by voltage indication on the bus (which can also be read on any ERFDADS terminal)

Performance Feedback: The crew will have positive confirmation that power has been restored based on breaker positions and bus voltage indications.

• This Critical Task and the 60 minute time requirement meets Operations Management expectations for an Operating Crew 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 #1 2023 NRC Scenario #1 Driver Setup Instructions Reset to IC-611 Run scenario file 2023 NRC Scn #1 Ensure OOS tags are hung on the B CS Pump and B NC Pump handswitches Ensure PROTECTED EQUIPMENT cover placed over the A CS Pump and A NC Pump handswitches ENSURE CHN-FQIS-210X IS SET TO A TOTAL OF 100 GALLONS BEFORE EACH RUN OF THE SCENARIO

Form 3-3.1 Scenario File Description NRC Exam Scenario #1 2023 NRC Scenario #1 Scenario File IMF cmBSRP01BSSG2LVLLDT_1 IMF cmBSRP01BSSG2LVLLCT_1 IMF cmBSRP01BSSG2LVLLBT_1 IMF cmBSRP01BSSG2LVLLAT_1 IMF cmBSRP01BSSG1LVLLDT_1 IMF cmBSRP01BSSG1LVLLCT_1 IMF cmBSRP01BSSG1LVLLBT_1 IMF cmBSRP01BSSG1LVLLAT_1 IMF cmCPCH02HCNA01C_5 IMF cmCPCH02HCNA01A_5 IMF cmTRCH05HCCPT352C_1 k:2 r:1 i:0.58524 f:85 IMF cmCNCV01CHEPDIC240_2 k:3 r:1 f:100.0 IMF mfMC01B k:4 f:1.2 IMF mfED11C k:5 IRF rfEG21 k:21 f:STOP IMF mfFW17A k:6 IMF mfED02 k:7 IRF rfEG17 k:7 f:CLOSE d:60 IRF rfED86 k:31 f:ENERGIZED IRF rfED85 k:32 f:CLOSE IRF rfED78 k:33 f:CLOSE

Form 3-3.1 Crew Turnover Sheet NRC Exam Scenario #1 2023 NRC Scenario #1 Plant Conditions:

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

B CS Pump (A CS Pump is protected) o LCO 3.6.6 Condition A was entered 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago. Expected return to service in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B NC Pump (A NC Pump is protected)

Planned Shift Activities:

Perform a 20 gallon shiftly dilution per 40OP-9CH01, CVCS Normal Operations, Section 6.46, Makeup - Dilution to the VCT for Normal Power Operations The dilution may be performed using the existing reactor makeup water flowrate

Form 3.3-1 Scenario Outline 2023 NRC Scenario #2 Facility:

PVNGS Scenario: 2 Test:

2023 NRC Exam Examiners:

Operators:

Initial Conditions: 75% power, MOC, B CS Pump OOS, B NC Pump OOS Turnover: Pump the Reactor Drain Tank to the Hold Up Tank from 60% to 57% per 40OP-9CH06, CVCS Miscellaneous Operations. Maintain power stable at 75%.

Event Number Event Type*

Event Description CRS OATC BOP 1

N N

Pump the RDT to the HUT (60% to 57%)

2 I

I, MC I

RRS Tc transmitter RCN-TT-121Y Fails High 3

TS Containment RM RU-1 Inlet CIV, HCB-UV-47, Fails Closed 4

C, TS C

C 10 gpm SGTL SG #1 5

C C, MC Loss of Non-Class Instrument Bus, NNN-D11 6

C C, MC RCP 2A Trip / ATWS (B05) 7 M

M M

ESD Outside Containment 8

C C, MC MSIS Fails to Auto Actuate

• (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) 3 Pre-identified CTs (2 or more)

Form 3-3.1 Scenario Event Summary NRC Exam Scenario #2 2023 NRC Scenario #2 2023 NRC Exam Scenario #2 Overview Event 1 Upon taking the shift, the OATC will pump the Reactor Drain Tank from 60% to 57% per 40OP-9CH06, CVCS Miscellaneous Operations.

Event 2 After the RDT has been pumped, RCS Tcold transmitter, RCN-TT-121Y, will fail high.

The CRS will enter 40AO-9ZZ16, RRS Malfunctions, and direct the crew to take manual control of PLCS, select the unaffected Loop Tave instrument at the RRS cabinet, and restore automatic control of Pressurizer level..

Event 3 When PLCS has been returned to automatic control, HCB-UV-47, Containment Radiation Monitor Containment Isolation Inlet Valve, will spuriously close. The CRS will address Technical Specifications for the failure of the CIV and isolation of the Rad Monitor.

Event 4 After addressing Technical Specifications, a 10 gpm SGTL will occur in SG #1. The CRS will enter 40AO-9ZZ02, Excessive RCS Leakrate, and direct the crew to determine the leakrate and take actions to minimize release to the environment. The CRS will also address Technical Specifications due to the leakage.

Event 5 After the crew has determined the leakrate and the CRS has addressed Technical Specifications, a loss of Non-Class Instrument Bus, NNN-D11, will occur. The CRS will enter 40AO-9ZZ14, Loss of Non-Class Instrument or Control Power, and direct the crew to ensure PLCS and PPCS are selected to unaffected input transmitters and take CEDMCS out of Auto Sequential.

NOTE: Letdown will isolate as a result of this event, however performance of 40AO-9ZZ05, Loss of Charging or Letdown is not needed to be observed since letdown will be re-isolated shortly after the Reactor trip in the next event.

Event 6 When the plant has been stabilized following the loss of NNN-D11, RCP 2A will trip but the Reactor will fail to auto trip, resulting in an ATWS. The crew will recognize the ATWS and take action to manually trip the Reactor at B05.

Event 7 On the Reactor trip, one MSSV on SG #1 will lift and fail to reseat, resulting in a SGTL and unisolable ESD to the environment. Based on this, the CRS will transition to 40EP-9EO09, Functional Recovery, following SPTAs, and direct the crew to commence feeding SG #1 at a rate of 1360-1600 gpm in an effort to cover the top of the U-tubes to minimize the release to the environment.

Event 8 When the ESD causes SG pressures to lower to the MSIS setpoint, MSIS will fail to auto actuate. The crew will manually actuate MSIS to ensure that SG #2 inventory is not removed via the failed MSSV on SG #1.

When MSIS has been actuated and SG #1 is being fed at 1360-1600 gpm, the scenario may be terminated.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #2 2023 NRC Scenario #2 Critical Task # 1: Following a failure of the Reactor to automatically trip, manually trip the Reactor 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 with valid trip signals locked in will result in lowering the margin to safety limits.

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

Measurable Performance Indicator: The Reactor should be tripped by manually depressing the RTCB Pushbuttons on B05.

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.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #2 2023 NRC Scenario #2 Critical Task # 2: When the Main Steam Isolation setpoints are exceeded, ensure Main Steam Isolation has actuated prior to automatic AFAS-2 actuation Safety Significance: Prior to the MSIS, both SGs are affected by the failed open MSSV on SG #1.

As such, an AFAS-2 occurring prior to the initiation of MSIS will result in an avoidable loss of CST inventory (AFAS-1 is not preventable due to the lifted MSSV on SG #1).

Cueing: The crew should recognize the failure of MSIS to actuate when SG pressure lowers to less than 960 psia.

Measurable Performance Indicator: The crew will have to manually actuate MSIS by taking the four handswitches for each ESFAS channel actuation (on B05) to the actuate position. This can be confirmed by the red MSIS lights on the vertical section of B05 as well as the actuation logic lights for each actuation extinguishing on the horizontal section of B05. The AFAS-2 actuation will occur at 25.8% wide range in SG #2 and will be indicated by the red AFAS-2 alarm on the vertical section of B05.

Performance Feedback: The crew will have indication of successful actuations by observing the red MSIS lights on the vertical section of B05, the actuation logic lights for the MSIS actuation extinguishing on the horizontal section of B05, as well as by observing the actuated equipment for the MSIS (MSIVs, FWIVs, etc.)

Form 3-3.1 Critical Task Summary NRC Exam Scenario #2 2023 NRC Scenario #2 Critical Task # 3: Establish a feedrate of 1360-1600 gpm to SG #1 prior to exiting HR-2, RCS and Core Heat Removal, SG with SI Safety Significance: An event in which a SG has a tube leak or rupture concurrently with an unisolable steam leak to atmosphere will result in a radioactive release to the atmosphere. A feedrate of 1360-1600 gpm to the affected SG is performed in order to expeditiously establish sufficient inventory in the affected SG to ensure the U-tubes are covered (~ 45% NR), thus minimizing the release to the environment.

Cueing: The crew will have indication of SG tube leakage on SG #1 prior to the reactor trip (leak occurs in Event 4) from ERFDADS indicating a rising leakrate and SG #1 level rising. Once entering an EOP, the crew can also get confirmation of the SGTL from chemistry. The ESD outside of containment will be indicated steam flow on SG #1 rising and pressure on SG #1 lowering.

Measurable Performance Indicator: The crew will align 2 AFW pumps to supply feedwater to SG #1 for a total of 1360-1600 gpm, per step 15 of 40EP-9EO09, Functional Recovery, HR-2, SG with SI.

Performance Feedback: Total feed flow to the affected SG will be available using any ERFDADS computer 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 #2 2023 NRC Scenario #2 Driver Setup Instructions Reset to IC-612 Run scenario file 2023 NRC Scn #2 Ensure OOS tags are hung on the B CS Pump and B NC Pump handswitches Ensure PROTECTED EQUIPMENT cover placed over the A CS Pump and A NC Pump handswitches

Form 3-3.1 Scenario File Description NRC Exam Scenario #2 2023 NRC Scenario #2 Scenario File IMF cmSRRP01ARRPSK1_2 IMF cmSRRP01ARRPSK2_2 IMF cmSRRP01ARRPSK3_2 IMF cmSRRP01ARRPSK4_2 IMF cmBSRP01BSCNTPRHIAT_1 IMF cmBSRP01BSCNTPRHIBT_1 IMF cmBSRP01BSCNTPRHICT_1 IMF cmBSRP01BSCNTPRHIDT_1 imf cmBSRP01BSSG1PRLOAT_1 imf cmBSRP01BSSG1PRLOBT_1 imf cmBSRP01BSSG1PRLOCT_1 imf cmBSRP01BSSG1PRLODT_1 imf cmBSRP01BSSG2PRLOAT_1 imf cmBSRP01BSSG2PRLOBT_1 imf cmBSRP01BSSG2PRLOCT_1 imf cmBSRP01BSSG2PRLODT_1 IMF cmTRRX05RCNTT121Y_1 k:2 r:1 i:559.621 f:650 IOR diCH_ZDHCBHS47 k:3 f:CLOSE MOR diCH_ZDHCBHS47 k:3 c:10 f:CLOSE IMF mfTH06A k:4 f:1 IRF rfFW13 f:0 k:31 IMF mfED13A k:5 IMF mfRC01C k:6 IMF cmRVMS01SGEPSV575_2 k:6 d:5 r:5 f:100.0

Form 3-3.1 Crew Turnover Sheet NRC Exam Scenario #2 2023 NRC Scenario #2 Plant Conditions:

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

B CS Pump (A CS Pump is protected) o LCO 3.6.6 Condition A was entered 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago. Expected return to service in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B NC Pump (A NC Pump is protected)

Planned Shift Activities:

Pump the RDT to the HUT per 40OP-9CH06, CVCS Miscellaneous Operations, Section 6.8, Pumping the Reactor Drain Tank to the Holdup Tank Bypassing the Gas Stripper, until RDT level is 57%, and restore from pumping the RDT per Section 6.9, Restoration from Pumping the Reactor Drain Tank to the Holdup Tank Bypassing the Gas Stripper The AO has been briefed and is on station with a copy of the associated procedure sections

Form 3.3-1 Scenario Outline 2023 NRC Scenario #3 Facility:

PVNGS Scenario: 3 Test:

2023 NRC Exam Examiners:

Operators:

Initial Conditions: 50% power, MOC, B CS Pump OOS, B NC Pump OOS Turnover: Change the order to running Charging Pumps from 1-2-3 to 3-1-2 per 40OP-9CH01, CVCS Normal Operations. Maintain power stable at 50%.

Event Number Event Type*

Event Description CRS OATC BOP 1

N N

Change the order of running Charging Pumps from 1-2-3 to 3-1-2 2

I I

SG #2 Level transmitter SGN-LT-1121 fails low 3

I, TS I

Pressurizer Level transmitter RCA-LT-110Y fails to 50%

4 C, TS C, MC Inadvertent Train B AFAS-2 5

C, TS C

C Loss of NC - Cross-Tie NC and EW 6

C C, MC A Turbine Cooling Water Pump Shaft Shear / Standby Pump Fails to Auto Start 7

M M

M Feed Line Break on SG #2 Inside Containment 8

C C, MC Train A CS Pump Fails to Auto Start

• (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) 5 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 #3 2023 NRC Scenario #3 2023 NRC Exam Scenario #3 Overview Event 1 Upon taking the shift, the OATC will change the order of running Charging Pumps from 1-2-3 to 3-1-2 per 40OP-9CH01, CVCS Normal Operations.

Event 2 When the order of running Charging Pumps has been changed, SG #2 level transmitter, SGN-LT-1121, will fail low. The BOP will address the ARP and placed the failed transmitter in maintenance.

Event 3 When LT-1121 has been placed in maintenance, Pressurizer level transmitter, RCA-LT-110Y, will fail to 50%. The OATC will address the ARP and place the unaffected level transmitter in service for PLCS, restore automatic control of Pressurizer level, and ensure Pressurizer heaters are returned to normal configuration. The CRS will also address Technical Specifications for the failed transmitter.

Event 4 When Technical Specifications have been addressed, an inadvertent Train B AFAS-2 will occur. The CRS will enter 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations, and direct the BOP to override and stop feeding SG #2. The CRS will also address Technical Specifications following the inadvertent actuation.

Event 5 When Technical Specifications have been addressed, the running Nuclear Cooling Water Pump will trip. The CRS will enter 40AO-9ZZ03, Loss of Cooling Water, and direct the BOP to cross-tie Train A EW with NC to restore cooling to the RCPs.

Letdown will be also be lost as a result of the loss of NC so the CRS will also enter 40AO-9ZZ05, Loss of Charging or Letdown, and direct the OATC to establish conditions for extended operation with letdown isolated. The CRS will also have to address Technical Specifications following the cross-tie of NC with EW.

Event 6 When NC and EW have been cross-tied and Technical Specifications have been addressed, the running Turbine Cooling Water Pump will experience a sheared shaft and the standby pump will fail to auto start. The BOP will address the ARP and start the standby TC Pump.

Event 7 When the standby TC Pump is started, a feed line break will occur on SG #2 inside Containment. The crew will trip the Reactor and perform SPTAs. When SIAS actuates, the RCPs will lose cooling flow and the crew will have to recognize this and trip all 4 RCPs. When SG #2 reaches dryout, the BOP will take action to stabilize RCS temperature to prevent lifting a Pressurizer Safety Valve.

Event 8 When Containment pressure exceeds 8.06 psig, CSAS will occur. However the B CS Pump is OOS and the A CSAS will fail to auto start on the CSAS, requiring the crew to manually start the A CS Pump to establish CS flow.

When the crew has stabilized RCS temperature and established adequate CS flow, the scenario may be terminated.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #3 2023 NRC Scenario #3 Critical Task # 1: Secure all 4 RCPs within 30 minutes of the unrecoverable loss of cooling water 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. The SIAS will close the NC-EW cross-tie valves, resulting in a loss of cooling water the RCPs.

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.

Form 3-3.1 Critical Task Summary NRC Exam Scenario #3 2023 NRC Scenario #3 Critical Task # 2: When the Containment Spray Actuation setpoint is exceeded, ensure adequate Containment Spray flow to meet the CTPC safety function prior to exiting Standard Post Trip Actions Safety Significance: Potential degradation of any barrier to fission product release. Failure to maintain containment temperature and pressure control may challenge containment integrity.

Restoring this safety function prior to the completion of SPTAs will prevent transition to the Functional Recovery procedure, which would slow the mitigation of the event in progress.

Cueing: In addition to the procedural cue, the crew may use indications of Containment pressure, Containment temperature, Containment fan coolers, Containment Spray pumps, and Containment Spray flow to provide cue to perform elements of this task.

Measurable Performance Indicator: The task is identified by at least one member of the crew manipulating the controls (starting the A CS Pump) to establish Containment Spray flow. If Containment pressure is > 8.5 psig, the crew should ensure a CSAS is actuated and at least one CS header is delivering > 4350 gpm on at least one header.

Performance Feedback: The task provides feedback by observing > 4350 gpm on B02 and ERFDADS flow indicators and Containment pressure lowering.

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 2023 NRC Scenario #3 Driver Setup Instructions Reset to IC-613 Run scenario file 2023 NRC Scn #3 Ensure OOS tags are hung on the B CS Pump and B NC Pump handswitches Ensure PROTECTED EQUIPMENT cover placed over the A CS Pump and A NC Pump handswitches ENSURE 6A06A ALARM IS RESET AND DFWCS ALARMS ARE ACKNOWLEDGED AND RESET PRIOR TO STARTING SCENARIO

Form 3-3.1 Scenario File Description NRC Exam Scenario #3 2023 NRC Scenario #3 Scenario File IMF cmCPRH02SIAP03_5 IMF cmTRFW04SGNLT1121_1 k:2 r:5:00 i:49.9145 f:0 IMF cmTRCV19RCBLT110Y_1 k:3 r:1 i:40.2065 f:50 IMF mfRP06N1 k:4 IMF mfRP06N2 k:4 IMF mfCC01A k:5 IRF rfCC34 k:31 f:90 IMF cmCPTP04TCNP01B_5 k:6 IMF cmCPTP04TCNP01A_1 k:6 d:5 IMF mfFW12B k:7 r:5:00 f:10

Form 3-3.1 Crew Turnover Sheet NRC Exam Scenario #3 2023 NRC Scenario #3 Plant Conditions:

Unit 1 is operating at 50% power, MOC B MFP running, A MFP in standby CEDMCS is in Auto Sequential Equipment Out of Service:

B CS Pump (A CS Pump is protected) o LCO 3.6.6 Condition A was entered 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago. Expected return to service in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B NC Pump (A NC Pump is protected)

Planned Shift Activities:

Shift the order of running Charging Pumps from 1-2-3 to 3-1-2 per 40OP-9CH01, CVCS Normal Operations in preparation for preventative maintenance to be performed next shift AO is on station and pre-start checks have been performed

Facility:

2023 E

A V

M P

E I

M P

N T

N A

L T

O I

R I

T M

G C

T A

U I

A Y

S A

B S

A B

S A

B S

A B

L M

N 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 0

U1 NOR 1

1 2

1 1

1 1

I/C 2,4,5,6

,8 2,3,4,5

,6,8 11 4

4 2

9 MAJ 7

7 2

2 2

1 1

Man. Ctrl.

0 1

1 0

0 13 TS 3,4 3,4,5 5

0 2

2 3

RX 0

1 1

0

-1 R1 NOR 1

1 1

1 1

0 I/C 2,4,5,6 2,4,5,6 8

4 4

2 4

MAJ 7

7 2

2 2

1 0

Man. Ctrl.

2,5,6 4,6 5

1 1

0 4

9 TS 0

0 2

2 0

RX 0

1 1

0

-1 R2 NOR 1

1 1

1 1

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

4 4

2 2

MAJ 7

7 2

2 2

1 0

Man. Ctrl.

8 8

2 1

1 0

1 7

TS 0

0 2

2 0

RX+NOR

+I/C RX+NOR

+I/C RX+NOR

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

2 3

4 POSITION POSITION POSITION POSITION PVNGS 5/1/2023 Operating Test Number:

Form 3.4-1 Events and Evolutions Checklist - CREW 'A' Date of Exam:

Facility:

2023 E

A V

M P

E I

M P

N T

N A

L T

O I

R I

T M

G C

T A

U I

A Y

S A

B S

A B

S A

B S

A B

L M

N 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 0

U2 NOR 1

1 2

1 1

1 1

I/C 2,4,5,6

,8 2,3,4,5

,6,8 11 4

4 2

9 MAJ 7

7 2

2 2

1 1

Man. Ctrl.

0 1

1 0

0 13 TS 3,4 3,4,5 5

0 2

2 3

RX 0

1 1

0

-1 R3 NOR 1

1 1

1 1

0 I/C 2,4,5,6 2,4,5,6 8

4 4

2 4

MAJ 7

7 2

2 2

1 0

Man. Ctrl.

2,5,6 4,6 5

1 1

0 4

9 TS 0

0 2

2 0

RX 0

1 1

0

-1 R4 NOR 1

1 1

1 1

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

4 4

2 2

MAJ 7

7 2

2 2

1 0

Man. Ctrl.

8 8

2 1

1 0

1 7

TS 0

0 2

2 0

RX+NOR

+I/C SRO-U RX+NOR

+I/C RX+NOR

+I/C POSITION POSITION POSITION POSITION RO SRO-I PVNGS Date of Exam: 5/1/2023 Operating Test Number:

Scenarios 1

2 3

4 Form 3.4-1 Events and Evolutions Checklist - CREW 'B'

Facility:

2023 E

A V

M P

E I

M P

N T

N A

L T

O I

R I

T M

G C

T A

U I

A Y

S A

B S

A B

S A

B S

A B

L M

N 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 0

U3 NOR 1

1 2

1 1

1 1

I/C 2,4,5,6

,8 2,3,4,5

,6,8 11 4

4 2

9 MAJ 7

7 2

2 2

1 1

Man. Ctrl.

0 1

1 0

0 13 TS 3,4 3,4,5 5

0 2

2 3

RX 0

1 1

0

-1 R5 NOR 1

1 1

1 1

0 I/C 2,4,5,6 2,4,5,6 8

4 4

2 4

MAJ 7

7 2

2 2

1 0

Man. Ctrl.

2,5,6 4,6 5

1 1

0 4

9 TS 0

0 2

2 0

RX 0

1 1

0

-1 R6 NOR 1

1 1

1 1

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

4 4

2 2

MAJ 7

7 2

2 2

1 0

Man. Ctrl.

8 8

2 1

1 0

1 7

TS 0

0 2

2 0

RX+NOR

+I/C SRO-U RX+NOR

+I/C RX+NOR

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

2 3

4 Form 3.4-1 Events and Evolutions Checklist - CREW 'C' PVNGS Date of Exam: 5/1/2023 Operating Test Number:

Scenarios

Facility:

2023 E

A V

M P

E I

M P

N T

N A

L T

O I

R I

T M

G C

T A

U I

A Y

S A

B S

A B

S A

B S

A B

L M

N 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 0

U4 NOR 1

1 1

1 1

0 I/C 2,4,5,6

,8 2,4,5,6 9

4 4

2 7

MAJ 7

7 2

2 2

1 1

Man. Ctrl.

4,6 2

1 1

0 2

10 TS 3,4 2

0 2

2 0

RX 0

1 1

0

-1 I1 NOR 1

1 2

1 1

1 1

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

,6,8 10 4

4 2

6 MAJ 7

7 2

2 2

1 0

Man. Ctrl.

2,5,6 3

1 1

0 2

12 TS 3,4,5 3

0 2

2 1

RX 0

1 1

0

-1 R7 NOR 1

1 1

1 1

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

4 4

2 2

MAJ 7

7 2

2 2

1 0

Man. Ctrl.

8 8

2 1

1 0

1 7

TS 0

0 2

2 0

RX+NOR

+I/C SRO-U RX+NOR

+I/C RX+NOR

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

2 3

4 Form 3.4-1 Events and Evolutions Checklist - CREW 'D' PVNGS Date of Exam: 5/1/2023 Operating Test Number:

Scenarios