ML22144A029
| ML22144A029 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/12/2022 |
| From: | Heather Gepford NRC/RGN-IV/DORS/OB |
| To: | Arizona Public Service Co |
| References | |
| Download: ML22144A029 (34) | |
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: 5/12/2022 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)
AA2.03 PZR PORV and/or safety valve position 3.5 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)
EK1.04 Evaluating RCP trip strategy 3.7 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)
AK2.07 RCS 3.2 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)
AK2.06 T-ave/T-ref deviation meter 3.4 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)
A1.12 Alarms and lights 3.3 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)
A3.03 Inverter swap to backup 3.3 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 X
Ability to predict and/or monitor changes in parameters associated with operation of the Control Rod Drive System, including:
(CFR: 41.5 / 45.5)
A1.17 Control bank sequence and overlap 3.7 58 002 (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 X
Knowledge of the effect that a loss or malfunction of the Nuclear Instrumentation System will have on the following systems or system parameters: (CFR: 41.7 / 45.6)
K3.03 FHS 2.9 56 016 (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 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 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)
A4.02 Radiation monitor function 3.4 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: 5/12/2022 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 K1.03 192004 Reactivity Coefficients Describe the effect on the magnitude of the temperature coefficient of reactivity from changes in the following: core age 3.1 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 4.1-1 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) 065 AK2.09 The only valves in the ECCS system which are air-operated are SIT drain valves which are normally closed and fail closed so there is little to no effect of a loss of IA on the ECCS system.
Additionally, it would not be plausible that a SIT drain valve would fail open on a loss of IA (draining the SIT), therefore reselected 065 AK2.07.
1 /1 (Q12) 008 AA2.06 PVNGS does not have PORVs, nor a low-pressure logic control for our PZR Code Safeties, therefore the randomly selected K/A does not apply. Reselected 008 AA2.03 1 /1 (Q13)
CE E05 EK1.12 At PVNGS, RCPs are not secured during an ESD if it is outside Containment (2 are left running and there is no reason we would restart one of the 2 secured), and if the ESD is inside Containment all 4 RCPs are secured due to the loss of cooling water to the RCPs and would therefore not be restarted. As such, reselected CE E05 EK1.04.
1 / 2 (Q20) 001 AK2.07 There is no relationship or interlock between a continuous rod withdrawal and the Boric Acid Makeup Pumps (or pump running lights) at PVNGS. Reselected 001 AK2.06.
2 / 1 (Q29) 010 A1.03 There is no operational way that the PPCS can impact RDT pressure/temp at PVNGS, and even if there was, the only possible impact to Quench Tank pressure/temp would be for it to rise, making any wrong answers implausible. Reselected 010 A1.12 2 / 1 (Q47) 063 A3.02 There is only one DC load that is stripped at PVNGS, and that concept was asked on Q27 (K/A: DC Electrical Distribution -
Operate/monitor: Load shedding), therefore had to replace KA on either Q27 or Q47. Reselected 063 A3.03.
2 / 2 (Q56) 034 K3.03 ROs at PVNGS responsibilities during Fuel Handling Operations is limited to monitoring NI response (audible and visual) as well as radiation levels. Additionally, the Fuel Handling equipment doesnt have any real interaction with Reactor components (other than moving the fuel assemblies). In order to keep the fuel handling aspect of this KA, reselected 015 K3.03 (Nuclear Instrumentation - impact to FHS) 2 / 2 (Q58) 028 A1.03 The only recombiner temperatures with any relevance at PVNGS are the temperature to which containment air is heated to during recombination, which carries virtually zero operational relevance. Additionally, the Hydrogen Recombiner System at PVNGS is currently in the process of being retired in place.
Reselected 001 A1.17 (and verified that 001 - Control Rod Drive System - is not selected as a T2G2 question on either the RO or SRO exam.
2 / 2 (Q61) 072 A4.01 Alarm and interlock setpoint checks and adjustments are solely done by Effluents personnel at PVNGS and is are outside the job function for Operations personnel. Reselected 072 A4.02.
Rev. 12 4
(Q75) 192004 K1.18 Made several attempts to write a question to this KA but couldnt seem to write one that was operationally relevant and had plausible but incorrect distractors. Also searched the NRC GFES exam bank for a question for this KA and found none.
Reselected 192004 K1.03.
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: 5/12/2022 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 2
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
2.2.45 Ability to determine and/or interpret TS with action statements of greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (SRO Only)
(CFR: 43.2 / 43.5 / 45.3) 4.7 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)
EA2.19 S/G level 4.0 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)
EA2.11 S/G level and pressure 3.6 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 X
Ability to (a) predict the impacts of the following on the Reactor Protection System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)
A2.01 Faulty bistable operation 3.6 87 013 (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 X
Ability to (a) predict the impacts of the following on the Main Feedwater 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 I 45.13}
A2.01 Actuation of AFW system 4.1 86 061 (SF4S AFW) Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution X
2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator (CFR: 41.10 / 43.5 / 45.11) 4.1 88 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring X
Ability to (a) predict the impacts of the following on the Process Radiation Monitoring 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 / 8-9)
A2.01 PRM component failures 3.1 90 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air
Rev. 12 103 (SF5 CNT) Containment X
2.2.45 Ability to determine or interpret technical specifications with action statements of greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (SRO Only) (CFR: 43.2 / 43.5 / 45.3) 4.7 89 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:
Group Point Total:
5
Rev. 12 Form 4.1-PWR PWR Examination Outline Page 7 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 X
2.2.25 Knowledge of the bases in technical specifications for limiting conditions for operation and safety limits (SRO Only)
(CFR: 43.2) 4.2 93 027 (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 X
Ability to (a) predict the impacts of the following on the Liquid Radwaste 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 / 45.8 /
45.9 I 45.13)
A2.04 Failure of automatic isolation 3.8 92 071 (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 X
Ability to (a) predict the impacts of the following on the Control Room Ventilation and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 43.5 / 45.6)
A2.03 Initiation/reconfiguration failure 3.8 91 K/A Category Point Totals:
Group Point Total:
3
Rev. 12 Form 4.1-COMMON Common Examination Outline Facility:
Date of Exam: 5/12/2022 Generic Knowledge and AbilitiesTier 3 (RO/SRO)
Category K/A #
Topic RO SRO-Only IR IR
- 1.
Conduct of Operations 2.1.2 Knowledge of operator responsibilities during any mode of plant operation (CFR: 41.10 / 43.1 / 45.13) 4.4 99 2.1.40 Knowledge of refueling administrative requirements (CFR: 41.10 / 43.5 / 43.6 / 45.13) 3.9 95 Subtotal 2
- 2.
Equipment Control 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator (CFR: 41.10 /
43.5 / 45.13) 3.8 96 2.2.20 Knowledge of the process for managing troubleshooting activities (CFR: 41.10 / 43.5 / 45.13) 3.8 100 Subtotal 2
- 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 94 Subtotal 1
- 4.
Emergency Procedures/
Plan 2.4.26 Knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage (CFR: 41.10 / 43.5 / 45.12) 3.6 98 2.4.41 Knowledge of the emergency action level thresholds and classifications (SRO Only) (CFR: 43.5 / 45.11) 4.6 97 Subtotal 2
Tier 3 Point Total 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 (Q79) 038 EA2.25 The PRT/Quench Tank at PVNGS only collects RCP bleedoff and PZR relief valve discharge and is thus unaffected during a SGTR. Reselected 038 EA2.19 1 / 1 (Q81) 026 G 2.4.50 Verification of alarm setpoints and ARP actions didnt lend themselves to any area which could be tied to a CFR55.43 topic. The most logical direction to go was the Technical Specification impacts of a loss of Component Cooling Water, therefore reselected 026 G 2.2.45 (determine/interpret TS actions greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> - SRO only) 1 / 2 (Q84)
CE E09 EA2.05 Since Charging and Letdown are read on meters at PVNGS, there is nothing to determine and/or interpret during the Functional Recovery procedure regarding charging and letdown.
Reselected CE E09 EA2.11.
2 / 2 (Q92) 068 A2.02 Since PVNGS is a zero release plant, could not generate a question which is a match to the randomly selected KA.
Reselected 068 A2.04.
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 cycling an Aux Feed Pump breaker which is racked 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 (different for each unit).
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: 2.2.37 IR: 4.6 JPM
Description:
The applicant will evaluate the operability of MSIVs in each unit, given the accumulator pressures for each MSIV.
M, 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 D, 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 (5) 4-6 (5) 2-3 (3)
(C)ontrol room (D)irect from bank 9 (5) 8 (4) 4 (3)
(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 (2)
(N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 2 (7 - 3A) 2 (6 - 3A) 1 (2 - 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 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, MC C, MC CEA 43 Slips 50 into the Core 5
C C, MC Loss of NHN-M13 (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) 3 Pre-identified CTs (2 or more)
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 BAMP OOS Turnover: Take Channel C RWT LO trip bistable out of bypass per 40OP-9SB01, Plant Protection System Bypass Operations, and continue the B HPSI Pump IST per 73ST-9SI10, HPSI Pump Miniflow - Inservice Test Event Number Event Type*
Event Description CRS OATC BOP 1
N Take C RWT LO Trip Bistable Out of Bypass 2
TS N
B HPSI Pump In-Service Test (pump trips during test) 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 + 1.5 min) - if A EDG not tripped within 5 minutes, A EDG will trip 9
C C
A HPSI Pump Fails to Auto Start (following restoration of power to PBA-S03)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Tech Spec, (MC) Manual Control Actual Target Quantitative Attributes 3
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 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 3
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)