ML23319A404
| ML23319A404 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 09/20/2023 |
| From: | Heather Gepford Operations Branch IV |
| To: | Ameren Union Electric Co |
| References | |
| Download: ML23319A404 (1) | |
Text
Form 4.1-PWR Pressurized-Water Reactor Examination Outline Facility:
Callaway K/A Catalog Rev. 3 Rev.
2 Date of Exam:
09/20/2023 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
2 4
3 3
18 3
3 6
2 1
2 2
1 1
1 8
2 2
4 Tier Totals 4
5 4
5 4
4 26 5
5 10
- 2.
Plant Systems 1
3 2
2 3
2 3
3 2
4 2
2 28 3
2 5
2 1
1 1
1 0
1 1
1 0
1 1
9 1
1 1
3 Tier Totals 4
3 3
4 2
4 4
3 4
3 3
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
Notes: CO
=
EM =
Conduct of Operations; EC = Equipment Control; RC = Radiation Control; 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.
ES-4.1-PWR PWR Examination Outline (Callaway)
Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)
Item E/APE # / Name /
Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR Q#
1 (000007) (EPE 7; BW E02 & E10; CE E02)
Reactor Trip, Stabilization, Recovery X
(000007EK2.14) Knowledge of the relationship between (EPE 7) REACTOR TRIP, STABILIZATION, RECOVERY and the following systems or components: PZR PORVs (CFR: 41.8 / 41.10 / 45.3) 3.3 1
2 (000008) (APE 8)
Pressurizer Vapor Space Accident X
(000008AK3.04) Knowledge of the reasons for the following responses and/or actions as they apply to (APE 8)
PRESSURIZER VAPOR Space Accident: RCP tripping requirements (CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.5 2
3 (000009) (EPE 9)
Small Break LOCA X
(000009EA1.01) Ability to operate and/or monitor the following as they apply to (EPE 9) SMALL-Break LOCA:
RCS pressure and temperature (CFR: 41.5 / 41.7 / 45.5 to 45.8) 4.0 3
4 (000015) (APE 15)
Reactor Coolant Pump Malfunctions X
(000015) (APE 15) Reactor Coolant Pump Malfunctions (G2.1.20) CONDUCT OF OPERATIONS:
Ability to interpret and execute procedure steps (CFR: 41.10 / 43.5 / 45.12) 4.6 4
5 (000025) (APE 25)
Loss of Residual Heat Removal System X
(000025AK2.01) Knowledge of the relationship between (APE 25) LOSS OF RESIDUAL Heat Removal System and the following systems or components: RHR (CFR: 41.8 / 41.10 / 45.3) 4.0 5
6 (000027) (APE 27)
Pressurizer Pressure Control System Malfunction X
(000027AK1.05) Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to a Pressurizer Pressure Control System Malfunction:
TS limits for RCS pressure (CFR: 41.8 / 41.10 / 41.7 / 45.3) 4.1 6
7 (000038) (EPE 38)
Steam Generator Tube Rupture X
(000038EK2.09) Knowledge of the relationship between (EPE 38) STEAM GENERATOR Tube Rupture and the following systems or components: CVCS (CFR: 41.8 / 41.10 / 45.3) 3.3 7
8 (000040) (APE 40; BW E05; CE E05; W E12) Steam Line Rupture - Excessive Heat Transfer X
(WE12EA2.09) Ability to determine and/or interpret the following as they apply to (W E12) UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS:
Containment pressure (CFR: 41.10 / 43.5 / 45.13) 3.6 8
9 (000054) (APE 54; CE E06) Loss of Main Feedwater X
(000054AA2.01) Ability to determine and/or interpret the following as they apply to (APE 54) LOSS OF Main Feedwater: Occurrence of reactor and/or turbine trip (CFR: 41.10 / 43.5 / 45.13) 3.7 9
10 (000055) (EPE 55)
Station Blackout X
(000055EA1.04) Ability to operate and/or monitor the following as they apply to (EPE 55) Station Blackout: Load shedding (CFR: 41.5 / 41.7 / 45.5 to 45.8) 4.0 10 11 (000056) (APE 56)
(000056AA1.30) Ability to operate and/or monitor the following as they apply to (APE 56) Loss of Offsite Power:
AFW flow control valve (CFR: 41.5 / 41.7 / 45.5 to 45.8) 3.8 11 12 (000057) (APE 57)
Loss of Vital AC Instrument Bus X
(000057AA2.14) Ability to determine and/or interpret the following as they apply to (APE 57) LOSS OF VITAL AC ELECTRICAL INSTRUMENT BUS: Verification that substitute power sources have come on line on a loss of initial AC (CFR: 41.10 / 43.5 / 45.13) 3.7 12 13 (000058) (APE 58)
Loss of DC Power X
(000058AA1.03) Ability to operate and/or monitor the following as they apply to (APE 58) LOSS OF DC Power:
Vital bus and battery bus components 3.8 13
14 (000062) (APE 62)
Loss of Nuclear Service Water X
(000062AK1.01) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (APE 62) LOSS OF SERVICE WATER: Effect on loads cooled by service water (CFR: 41.5 / 41.7 / 45.7 / 45.8) 3.8 14 15 (000065) (APE 65)
Loss of Instrument Air X
(000065) (APE 65) Loss of Instrument Air (G2.2.44)
EQUIPMENT CONTROL: Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions (CFR: 41.5 / 43.5 / 45.12) 4.2 15 16 (000077) (APE 77)
Generator Voltage and Electric Grid Disturbances X
(000077) (APE 77) Generator Voltage and Electric Grid Disturbances (G2.1.19) CONDUCT OF OPERATIONS:
Ability to use available indications to evaluate system or component status (CFR: 41.10 / 45.12) 3.9 16 17 (W E04) LOCA Outside Containment X
(WE04EK1.05) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (W E04) LOCA Outside Containment: Leakage accumulation in RHR pump area (CFR: 41.5 / 41.7 / 45.7 / 45.8) 3.6 17 18 (W E11) Loss of Emergency Coolant Recirculation X
(WE11EK3.21) Knowledge of the reasons for the following responses and/or actions as they apply to (W E11) LOSS OF EMERGENCY Coolant Recirculation: Placing RHRS in operation and maintaining RCS heat removal (CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.8 18 19 (000011) (EPE 11)
Large Break LOCA X
(000011EA2.10) Ability to determine and/or interpret the following as they apply to (EPE 11) LARGE-Break LOCA:
Adequate core cooling (CFR: 41.10 / 43.5 / 45.13) 4.2 76 20 (000022) (APE 22)
Loss of Reactor Coolant Makeup X
(000022) (APE 22) Loss of Reactor Coolant Makeup (G2.1.28) CONDUCT OF OPERATIONS: Knowledge of the purpose and function of major system components and controls (CFR: 41.7) 4.1 77 21 (000026) (APE 26)
Loss of Component Cooling Water X
(000026AA2.01) Ability to determine and/or interpret the following as they apply to (APE 26) LOSS OF Component Cooling Water: Location of a leak in the CCWS (CFR: 41.10 / 43.5 / 45.13) 3.5 78 22 (000029) (EPE 29)
Anticipated Transient Without Scram X
(000029) (EPE 29) Anticipated Transient Without Scram (G2.4.35) EMERGENCY PLAN: Knowledge of nonlicensed operator responsibilities during an emergency (CFR: 41.10 / 43.1 / 43.5 / 45.13) 4.0 79 23 (000055) (EPE 55)
Station Blackout X
(000055) (EPE 55) Station Blackout (G2.4.37)
EMERGENCY PLAN: Knowledge of the lines of authority during implementation of the emergency plan implementing procedures (CFR: 41.10 / 45.13) 4.1 80 24 (BW E04; W E05)
Inadequate Heat Transfer - Loss of Secondary Heat Sink X
(WE05EA2.10) Ability to determine and/or interpret the following as they apply to (W E05) Loss of Secondary Heat Sink: High-head SI flow (CFR: 41.10 / 43.5 / 45.13) 3.7 81 K/A Category Totals:
3 3
2 4
3 3
Group Point Total:
24
ES-4.1-PWR PWR Examination Outline (Callaway)
Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
Item E/APE # / Name /
Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR Q#
25 (000003) (APE 3)
Dropped Control Rod X
(000003AK2.08) Knowledge of the relationship between (APE 3) DROPPED Control Rod and the following systems or components: NIS (CFR: 41.8 / 41.10 / 45.3) 3.9 19 26 (000032) (APE 32)
Loss of Source Range Nuclear Instrumentation X
(000032AK2.02) Knowledge of the relationship between (APE 32) LOSS OF SOURCE RANGE Nuclear Instrumentation and the following systems or components:
Intermediate/log power detectors (CFR: 41.8 / 41.10 / 45.3) 3.2 20 27 (000060) (APE 60)
Accidental Gaseous Radwaste Release X
(000060AA1.04) Ability to operate and/or monitor the following as they apply to an Accidental Gaseous Radwaste Release: Gaseous radwaste release isolation valve.
(CFR: 41.7 / 45.5 / 45.6) 3.7 21 28 (000067) (APE 67)
Plant Fire On Site X
(000067AA2.12) Ability to determine and/or interpret the following as they apply to (APE 67) PLANT Fire On Site:
Location of vital equipment within fire zone (CFR: 41.10 / 43.5 / 45.13) 3.3 22 29 (000076) (APE 76)
High Reactor Coolant Activity X
(000076AK3.04) Knowledge of the reasons for the following responses and/or actions as they apply to (APE 76) HIGH REACTOR COOLANT ACTIVITY: Maximizing demineralizer flow rates (CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.3 23 30 (W E01 & E02)
Rediagnosis & SI Termination X
(WE02EK3.06) Knowledge of the reasons for the following responses and/or actions as they apply to (W E02) SI TERMINATION: Resetting SI and/or containment isolation signal(s)
(CFR: 41.5 / 41.10 / 45.6 / 45.13) 3.9 24 31 (W E13) Steam Generator Overpressure X
(W E13) Steam Generator Overpressure (G2.1.27)
CONDUCT OF OPERATIONS: Knowledge of system purpose and/or function (CFR: 41.7) 3.9 25 32 (CE A11**; W E08)
RCS Overcooling -
Pressurized Thermal Shock X
(WE08EK1.04) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (W E08) Pressurized Thermal Shock: Maintaining steam supply to turbine-driven AFW pump if it is the only available source of feed flow CFR: 41.5 / 41.7 / 45.7 / 45.8) 3.9 26 33 (000001) (APE 1)
Continuous Rod Withdrawal X
(000001) (APE 1) Continuous Rod Withdrawal (G2.1.7)
Conduct of Operations: 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 82 34 (000028) (APE 28)
Pressurizer (PZR)
Level Control Malfunction X
(000028AA2.13) Ability to determine and/or interpret the following as they apply to (APE 28) PRESSURIZER (PZR) Level Control Malfunction: The actual PZR level, given an uncompensated level with an appropriate graph (CFR: 41.10 / 43.5 / 45.13) 3.3 83 35 (000051) (APE 51)
(000051AA2.02) Ability to determine and/or interpret the following as they apply to (APE 51) LOSS OF Condenser Vacuum: Conditions requiring reactor and/or turbine trip (CFR: 41.10 / 43.5 / 45.13) 4.2 84 36 (BW E08; W E03)
LOCA Cooldown -
Depressurization X
(BW E08; W E03) LOCA Cooldown - Depressurization (G2.4.12) EMERGENCY PROCEDURES/PLAN:
Knowledge of operating crew responsibilities during emergency and abnormal operations (CFR: 41.10 /
45.12) 4.3 85 000005 (APE 5)
Inoperable/Stuck Control Rod / 1 000024 (APE 24)
Emergency Boration /
1
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 000059 (APE 59)
Accidental Liquid Radwaste Release /
9 000061 (APE 61)
Area Radiation Monitoring System Alarms / 7 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 000078 (APE 78*)
RCS Leak / 3 (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 E09; CE A13**;
W E09 & E10)
Natural Circulation/4 (BW E13 & E14)
EOP Rules and Enclosures
(CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation /
LOOP / Blackout / 4 K/A Category Totals:
1 2
2 1
1 1
Group Point Total:
12
ES-4.1-PWR PWR Examination Outline (Callaway)
Plant SystemsTier 2/Group 1 (RO/SRO)
Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR Q#
37 (003) (SF4P RCP)
REACTOR COOLANT PUMP SYSTEM X
(003A2.06) Ability to (a) predict the impacts of the following on the (SF4P RCP) REACTOR COOLANT PUMP SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: CCWS malfunction (CFR: 41.5 / 45.6) 3.5 27 38 (004) (SF1; SF2 CVCS) CHEMICAL AND VOLUME CONTROL SYSTEM X
(004) (SF1; SF2 CVCS)
CHEMICAL AND VOLUME CONTROL SYSTEM (G2.1.2)
CONDUCT OF OPERATIONS:
Knowledge of operator responsibilities during any mode of plant operation (CFR:
41.10 / 43.1 / 45.13) 4.1 28 39 (004) (SF1; SF2 CVCS) CHEMICAL AND VOLUME CONTROL SYSTEM X
(004K4.10) Knowledge of (SF1; SF2 CVCS) CHEMICAL AND VOLUME CONTROL SYSTEM design features and/or interlocks that provide for the following: Minimum temperature requirements on borated systems (CFR: 41.7) 3.0 29 40 (005) (SF4P RHR)
RESIDUAL HEAT REMOVAL SYSTEM X
(005A4.06) Ability to manually operate and/or monitor the (SF4P RHR) RESIDUAL HEAT REMOVAL SYSTEM in the control room: RCS and containment isolation valves (CFR: 41.7 / 45.5 to 45.8) 3.8 30 41 (006) (SF2; SF3 ECCS)
EMERGENCY CORE COOLING SYSTEM X
(006A1.10) Ability to predict and/or monitor changes in parameters associated with operation of the (SF2; SF3 ECCS) EMERGENCY CORE COOLING SYSTEM, including:
CVCS letdown flow (CFR: 41.5 / 45.5) 3.0 31 42 (007) (SF5 PRTS)
PRESSURIZER RELIEF/QUENCH TANK SYSTEM X
(007K3.01) Knowledge of the effect that a loss or malfunction of the (SF5 PRTS)
PRESSURIZER RELIEF/QUENCH TANK SYSTEM will have on the following systems or system parameters: Containment (CFR: 41.7 / 45.4) 3.4 32 43 (008) (SF8 CCW)
COMPONENT COOLING WATER SYSTEM X
(008K6.08) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF8 CCW) COMPONENT COOLING WATER SYSTEM:
Power supply to CCWS pumps and/or valves (CFR: 41.7 / 45.7) 3.7 33
44 (010) (SF3 PZR PCS)
PRESSURIZER PRESSURE CONTROL SYSTEM X
(010A3.03) Ability to monitor automatic features of the (SF3 PZR PCS) PRESSURIZER PRESSURE CONTROL SYSTEM, including: PZR heater operation (CFR: 41.7 / 45.7) 3.3 34 45 (012) (SF7 RPS)
(012K3.06) Knowledge of the effect that a loss or malfunction of the (SF7 RPS) REACTOR PROTECTION SYSTEM will have on the following systems or system parameters: ECCS (CFR: 41.7 / 45.4) 3.9 35 46 (012) (SF7 RPS)
(012K4.08) Knowledge of (SF7 RPS) REACTOR PROTECTION SYSTEM design features and/or interlocks that provide for the following: Logic matrix testing (CFR: 41.7) 3.3 36 47 (013) (SF2 ESFAS)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM X
(013) (SF2 ESFAS)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (191008 K1.04)
BREAKERS, RELAYS, AND DISCONNECTS: Operation of various push buttons, switches, and handles and the resulting action on breakers (CFR: 41.7) 3.0 37 48 (022) (SF5 CCS)
CONTAINMENT COOLING SYSTEM X
(022K2.01) Knowledge of electrical power supplies to the following: (SF5 CCS)
CONTAINMENT COOLING SYSTEM CCS fans (CFR: 41.7) 3.6 38 49 (026) (SF5 CSS)
CONTAINMENT SPRAY SYSTEM X
(026K2.01) Knowledge of electrical power supplies to the following: (SF5 CSS)
CONTAINMENT SPRAY SYSTEM Containment spray pumps (CFR: 41.7) 3.9 39 50 (039) (SF4S MSS)
MAIN AND REHEAT STEAM SYSTEM X
(039A2.04) Ability to (a) predict the impacts of the following on the (SF4S MSS) MAIN AND REHEAT STEAM SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations Malfunctioning steam dump (CFR: 41.5 / 45.6):
4.0 40 51 (039) (SF4S MSS)
MAIN AND REHEAT STEAM SYSTEM X
(039A3.03) Ability to monitor automatic features of the (SF4S MSS) MAIN AND REHEAT STEAM SYSTEM, including: Atmospheric relief valves (CFR: 41.7 / 45.7) 3.7 41 52 (059) (SF4S MFW)
MAIN FEEDWATER SYSTEM X
(059A4.13) Ability to manually operate and/or monitor the (SF4S MFW) MAIN FEEDWATER SYSTEM in the control room: S/G water LCS (CFR: 41.7 / 45.5 to 45.8) 3.9 42
53 (061) (SF4S AFW)
AUXILIARY /
(061A3.01) Ability to monitor automatic features of the (SF4S AFW)
AUXILIARY/EMERGENCY FEEDWATER SYSTEM, including: AFW system automatic start (CFR: 41.7 / 45.7) 4.2 43 54 (061) (SF4S AFW)
AUXILIARY /
(061K1.07) Knowledge of the physical connections and/or cause and effect relationships between the (SF4S AFW)
AUXILIARY/EMERGENCY FEEDWATER SYSTEM and the following systems:
Emergency water source.
(CFR: 41.2 to 41.9 / 45.7 to 45.8) 4.2 44 55 (062) (SF6 ED AC)
AC ELECTRICAL DISTRIBUTION SYSTEM X
(062A1.07) Ability to predict and/or monitor changes in parameters associated with operation of the (SF6 ED AC)
AC ELECTRICAL DISTRIBUTION SYSTEM, including: Inverter outputs (CFR: 41.5 / 45.5) 3.2 45 56 (062) (SF6 ED AC)
AC ELECTRICAL DISTRIBUTION SYSTEM X
(062A3.09) Ability to monitor automatic features of the (SF6 ED AC) AC ELECTRICAL DISTRIBUTION SYSTEM, including: Load sequencing (CFR: 41.7 / 45.7) 3.7 46 57 (063) (SF6 ED DC)
DC ELECTRICAL DISTRIBUTION SYSTEM X
(063K5.04) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF6 ED DC)
DC ELECTRICAL DISTRIBUTION SYSTEM:
System ground (CFR: 41.5 / 45.3) 2.9 47 58 (064) (SF6 EDG)
EMERGENCY DIESEL GENERATOR SYSTEM X
(064K1.04) Knowledge of the physical connections and/or cause and effect relationships between the (SF6 EDG)
EMERGENCY DIESEL GENERATOR SYSTEM and the following systems: DC distribution system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.9 48 59 (064) (SF6 EDG)
EMERGENCY DIESEL GENERATOR SYSTEM X
(064K6.13) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF6 EDG)
EMERGENCY DIESEL GENERATOR SYSTEM:
ESFAS (CFR: 41.7 / 45.7) 4.1 49 60 (073) (SF7 PRM)
PROCESS RADIATION MONITORING SYSTEM X
(073A1.02) Ability to predict and/or monitor changes in parameters associated with operation of the (SF7 PRM)
PROCESS RADIATION MONITORING SYSTEM, including: Lights and alarms (CFR: 41.5 / 45.5) 3.2 50
61 (073) (SF7 PRM)
PROCESS RADIATION MONITORING SYSTEM X
(073K6.01) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF7 PRM)
PROCESS RADIATION MONITORING SYSTEM: PRM component failures (CFR: 41.7 / 45.7) 3.2 51 62 (076) (SF4S SW)
SERVICE WATER SYSTEM X
(076K5.03) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF4S SW)
SERVICE WATER SYSTEM:
Pump cavitation (CFR: 41.5 / 45.3) 3.1 52 63 (078) (SF8 IAS)
INSTRUMENT AIR SYSTEM X
(078K4.05) Knowledge of (SF8 IAS) INSTRUMENT AIR SYSTEM design features and/or interlocks that provide for the following: Isolation of instrument air to containment (CFR: 41.7) 3.2 53 64 (103) (SF5 CNT)
CONTAINMENT SYSTEM X
(103K1.07) Knowledge of the physical connections and/or cause and effect relationships between the (SF5 CNT)
CONTAINMENT SYSTEM and the following systems:
Containment vacuum system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.3 54 65 (003) (SF4P RCP)
REACTOR COOLANT PUMP SYSTEM X
(003A2.02) Ability to (a) predict the impacts of the following on the (SF4P RCP)
REACTOR COOLANT PUMP SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Conditions that exist for an abnormal shutdown of an RCP compared to a normal shutdown of an RCP (CFR: 41.5 / 45.6) 3.8 86 66 (005) (SF4P RHR)
RESIDUAL HEAT REMOVAL SYSTEM X
(005A2.03) Ability to (a) predict the impacts of the following on the (SF4P RHR)
RESIDUAL HEAT REMOVAL SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: RHR pump/motor malfunction (CFR: 41.5 / 45.6) 3.9 87
67 (010) (SF3 PZR PCS)
PRESSURIZER PRESSURE CONTROL SYSTEM X
(010) (SF3 PZR PCS)
PRESSURIZER PRESSURE CONTROL SYSTEM (G2.2.7)
EQUIPMENT CONTROL:
Knowledge of the process for conducting infrequently performed tests or evolutions (CFR: 41.10 / 43.3 / 45.13) 3.6 88 68 (026) (SF5 CSS)
CONTAINMENT SPRAY SYSTEM X
(026A2.04) Ability to (a) predict the impacts of the following on the (SF5 CSS)
CONTAINMENT SPRAY SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Failure of spray pump (CFR: 41.5 / 45.6) 4.0 89 69 (063) (SF6 ED DC)
DC ELECTRICAL DISTRIBUTION SYSTEM X
(063) (SF6 ED DC) DC ELECTRICAL DISTRIBUTION SYSTEM (G2.4.41)
EMERGENCY PROCEDURES/PLAN:
Knowledge of the emergency action level thresholds and classifications (SRO Only)
(CFR: 43.5 / 45.11) 4.6 90 025 (SF5 ICE) ICE CONDENSER SYSTEM 053 (SF1; SF4P ICS*) INTEGRATED CONTROL SYSTEM K/A Category Totals:
3 2
2 3
2 3
3 2
4 2
2 Group Point Total:
33
ES-4.1-PWR PWR Examination Outline (Callaway)
Plant SystemsTier 2/Group 2 (RO/SRO)
Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR Q#
70 (002) (SF2; SF4P RCS) REACTOR COOLANT SYSTEM X
(002K4.01) Knowledge of (SF2; SF4P RCS) REACTOR COOLANT SYSTEM design features and/or interlocks that provide for the following: Filling and draining the RCS, the refueling cavity, and/or refueling canal (CFR: 41.7) 3.2 55 71 (011) (SF2 PZR LCS)
PRESSURIZER LEVEL CONTROL SYSTEM X
(011A1.06) Ability to predict and/or monitor changes in parameters associated with operation of the (SF2 PZR LCS) PRESSURIZER LEVEL CONTROL SYSTEM, including:
PZR temperature (CFR: 41.5 / 45.5) 3.3 56 72 (015) (SF7 NI)
NUCLEAR INSTRUMENTATION SYSTEM X
(015A4.02) Ability to manually operate and/or monitor the (SF7 NI) NUCLEAR INSTRUMENTATION SYSTEM in the control room: NIS indicators (CFR: 41.7 / 45.5 to 45.8) 3.6 57 73 (017) (SF7 ITM) IN CORE TEMPERATURE MONITOR SYSTEM X
(017A2.02) Ability to (a) predict the impacts of the following on the (SF7 ITM) IN CORE TEMPERATURE MONITOR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Elevated in-core temperatures that can cause or have caused core damage (CFR: 41.5 / 45.6) 4.1 58 74 (027) (SF5 CIRS)
CONTAINMENT IODINE REMOVAL SYSTEM X
(027) (SF5 CIRS)
CONTAINMENT IODINE REMOVAL SYSTEM (G2.1.27)
CONDUCT OF OPERATIONS:
Knowledge of system purpose and / or function (CFR: 41.7) 3.9 59 75 (033) (SF8 SFPCS)
SPENT FUEL POOL COOLING SYSTEM X
(033K1.07) Knowledge of the physical connections and/or cause and effect relationships between the Spent Fuel Pool Cooling System and the following systems: Emergency makeup water systems (CFR: 41.2 to 41.9 / 45.7 /
45.8))
3.6 60 76 (035) (SF4P SG)
STEAM GENERATOR SYSTEM X
(035K3.01) Knowledge of the effect that a loss or malfunction of the (SF4P SG) STEAM GENERATOR SYSTEM will have on the following systems or system parameters: RCS (CFR: 41.7 / 45.4) 4.3 61
77 (041) (SF4S SDS)
STEAM DUMP/TURBINE BYPASS CONTROL SYSTEM X
(041K2.03) Knowledge of electrical power supplies to the following: (SF4S SDS) STEAM DUMP/TURBINE BYPASS CONTROL SYSTEM Turbine bypass control loop and valve power (CFR: 41.7) 2.9 62 78 (071) (SF9 WGS)
WASTE GAS DISPOSAL SYSTEM X
(071K6.07) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF9 WGS) WASTE GAS DISPOSAL SYSTEM: Waste gas compressors (CFR: 41.7 / 45.7) 2.6 63 79 (029) (SF8 CPS)
CONTAINMENT PURGE SYSTEM X
(029) (SF8 CPS)
CONTAINMENT PURGE SYSTEM (G2.2.38)
EQUIPMENT CONTROL:
Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 /
43.1 / 45.13) 4.5 91 80 (034) (SF8 FHS)
FUEL HANDLING EQUIPMENT SYSTEM X
(034K5.02) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF8 FHS) FUEL HANDLING EQUIPMENT SYSTEM: Load limitations (CFR: 41.4 / 41.5 / 43.7 /45.7) 3.0 92 81 (075) (SF8 CW)
(075A2.04) Ability to (a) predict the impacts of the following on the (SF8 CW)
CIRCULATING WATER SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Effects of extremes in ambient temperature on cooling tower operation (CFR: 41.5 / 45.6) 2.8 93 001 (SF1 CRDS)
CONTROL ROD DRIVE SYSTEM 014 (SF1 RPI) ROD POSITION INDICATION SYSTEM 016 (SF7 NNI)
NONNUCLEAR INSTRUMENTATION SYSTEM 028 (SF5 HRPS)
HYDROGEN RECOMBINER AND PURGE CONTROL SYSTEM
045 (SF4S MTG)
MAIN TURBINE GENERATOR SYSTEM 050 (SF9 CRV*)
CONTROL ROOM VENTILATION 055 (SF4S CARS)
CONDENSER AIR REMOVAL SYSTEM 056 (SF4S CDS)
CONDENSATE SYSTEM 068 (SF9 LRS)
LIQUID RADWASTE SYSTEM 072 (SF7 ARM)
AREA RADIATION MONITORING SYSTEM 079 (SF8 SAS**)
STATION AIR SYSTEM 086 (SF8 FPS) FIRE PROTECTION SYSTEM K/A Category Totals:
1 1
1 1
0 1
1 1
0 1
1 Group Point Total:
12
Form 4.1-COMMON Common Examination Outline ES-4.1-COMMON COMMON Examination Outline (Callaway)
Facility:
Callaway Date of Exam:
09/11/2023 Generic Knowledge and Abilities Outline (Tier 3) (RO/SRO)
Category K/A #
Topic RO SRO-Only Item #
IR Q#
IR Q#
- 1.
Conduct of Operations G2.1.29 (G2.1.29) CONDUCT OF OPERATIONS: Knowledge of how to conduct system lineups, such as valves, breakers, or switches (CFR: 41.10 / 45.1 / 45.12) 82 4.1 64 G2.1.47 (G2.1.47) CONDUCT OF OPERATIONS: Ability to direct nonlicensed personnel activities inside the control room (CFR:
41.10 / 43.5 / 45.5 / 45.12 / 45.13) 83 3.2 65 G2.1.18 (G2.1.18) CONDUCT OF OPERATIONS: Ability to make accurate, clear, and concise logs, records, status boards, and reports (CFR: 41.10 / 45.12 / 45.13) 84 3.8 94 G2.1.40 (G2.1.40) CONDUCT OF OPERATIONS: Knowledge of refueling administrative requirements (CFR: 41.10 / 43.5/ 43.6) 85 3.9 95 Subtotal N/A 2
N/A 2
- 2.
Equipment Control G2.2.44 (G2.2.44) EQUIPMENT CONTROL: Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions (CFR: 41.5 / 43.5 / 45.12) 86 4.2 66 G2.2.13 (G2.2.13) Knowledge of tagging and clearance procedures (CFR: 41.10 / 43.1 / 45.13) 87 4.1 67 G2.2.21 (G2.2.21) EQUIPMENT CONTROL: Knowledge of pre-and post-maintenance operability requirements (CFR: 41.10 /
43.2) 88 4.1 96 G2.2.17 (G2.2.17) EQUIPMENT CONTROL: 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) 89 3.8 97 Subtotal N/A 2
N/A 2
- 3.
Radiation Control G2.3.5 (G2.3.5) RADIATION CONTROL: Ability to use RMSs, such as fixed radiation monitors and alarms or personnel monitoring equipment (CFR: 41.11 / 41.12 / 43.4 / 45.9) 90 2.9 68 G2.3.12 (G2.3.12) RADIATION CONTROL: 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) 91 3.7 98 Subtotal N/A 1
N/A 1
- 4.
Emergency Procedures /
Plan G2.4.51 (G2.4.51) EMERGENCY PROCEDURES/PLAN: Knowledge of emergency operating procedure exit conditions (e.g., emergency condition no longer exists or severe accident guideline entry is required) (CFR: 41.10 / 43.5 /45.13) 92 3
69 G2.4.40 (G2.4.40) EMERGENCY PROCEDURES/PLAN: Knowledge of SRO responsibilities in emergency plan implementing procedures (SRO Only) (CFR: 43.5 / 45.11) 93 4.5 99 G2.4.42 (G2.4.42) EMERGENCY PROCEDURES/PLAN: Knowledge of emergency response facilities (CFR: 41.10 / 45.11) 94 3.8 100 Subtotal N/A 1
N/A 2
Tier 3 Point Total N/A 6
N/A 7
Form 4.1-COMMON Common Examination Outline
ES-4.1-COMMON COMMON Examination Outline (Callaway)
Facility:
Callaway Date of Exam:
09/11/2023 Theory (Tier 4) (RO)
Category K/A #
Topic RO Item #
IR Q#
Reactor Theory 192006 (192006K1.06) FISSION PRODUCT POISONS: Describe the following processes and state their effect on reactor operations: --
transient xenon (CFR: 41.1) 95 3.4 70 192007 (192007K1.01) FUEL DEPLETION AND BURNABLE POISONS:
Define burnable poison and state its use in the reactor (CFR:
41.1) 96 2.5 71 192008 (192008K1.09) REACTOR OPERATIONAL PHYSICS:
(CRITICALITY) Define criticality as related to a reactor startup (CFR: 41.1) 97 3.3 72 Subtotal N/A 3
Thermodynamics 193003 (193003K1.25) STEAM: Explain and use saturated and superheated steam tables (CFR: 41.14) 98 3.4 73 193004 (193004K1.11) THERMODYNAMIC PROCESS: (CONDENSERS)
Describe the process of condensate depression (subcooling) and its effect on plant operation (CFR: 41.14) 99 2.5 74 193009 (193009K1.02) CORE THERMAL LIMITS: Explain axial peaking factor (CFR: 41.14) 100 2.8 75 Subtotal N/A 3
Tier 4 Point Total N/A 6
Form 4.1-1 Record of Rejected Knowledge and Abilities Refer to Examination Standard (ES)-4.2, Developing Written Examinations, Section B.3, for deviations from the approved written examination outline.
Tier/Group Randomly Selected K/A Reason for Rejection 1/1 APE 27 K1.05 (4.1)
Q#6 - randomly selected a new K/A from the same APE due to overlap with operating test and other written exam questions (Q1, Q34) (APE27AK3.07(4.0)). New K/A's category was determined by observing that Tier 1 K1 category was lower than the rest while K3 category was higher, hence to maintain an more even distribution new K/A was randomly selected from APE27 AK1.
1/2 000060 (APE 60)
AA1.04 (3.7)
Q#21 - randomly selected new K/A from category not used in Tier /
Group as the original K/A (000060AK1.04 (2.8)) is outside the job scope of Callaway's reactor operators. The task is performed by the emergency plan responders.
2/2 002K4.01 (3.2)
Q#55 - selected the first K/A in the K/A Category due to the inability to write a question with plausible distractors with the original K/A (002K4.06 (2.7)).
2/2 027G2.1.27 (3.9)
Q#59 - randomly selected new Generic K/A due to the inability to write a question with plausible distractor of the original K/A (027G2.4.50 (4.2)). This was due to plant design as alarm response procedure only containing information about motor overloads.
2/2 033K1.07 (3.6)
Q#60 - After a discussion with the Chief Examiner, swapped original question #92 and #60 as Question #92 was a RO level topic.
Updated K/A value to maintain the Tier 2 / Group 2 distribution of 1 or 0 per K/A Category. Original K/A of (033A2.04 (3.4)).
4 193009K1.02 (2.8)
Q#75 - randomly selected new Theory K/A due to the inability to write a question pertinent to reactor operator position for the original K/A (193009K1.03(2.7)). For Callaway, Axial flux Difference (AFD) and Quadrant Power Tilt Ratio (QPTR) are applicable to the RO job scope via surveillances, Nuclear Instrument panel indication, etc.
SRO 1/1 000026 (APE26)
AA2.01 (3.5)
Q#78 - randomly selected new K/A in the K/A Category due to overlap concerns with Q#27 and the original K/A (000026AA2.06 (3.3)).
SRO 1/1 EPE 000029 G2.4.35 (4.0)
Q#79 - new K/A provided by the Chief Examiner after a discussion about the topic and references. Original K/A (EPE 000029 G2.1.47 (3.2)).
SRO 1/2 000001 (APE1)
G2.1.7 (4.7)
Q#82 - randomly selected new Generic K/A due to the inability to write a question to the topic and to minimize topic recurrence e.g.
K/A present in Q#89. Original K/A (000001 G2.2.23 (4.6)).
SRO 2/2 029G2.2.38 (4.5)
Q#91 - randomly selected new Generic K/A due to overlap of SRO Administrative JPM topics and the original K/A (029G2.3.6 (3.6)).
SRO 2/2 034K5.02 (3.0)
Q#92 - After a discussion with the Chief Examiner, swapped K/As with original question #92 and #60 as Question #60 was a SRO topic. Updated K/A value.
Form 3.2-1 Administrative Topics Outline Facility: _____Callaway ______________________
Date of Examination: __9/11/2023___
Examination Level:
RO SRO Operating Test Number: ___2023-1__
Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
A1 - Conduct of Operations G2.1.25 (3.9) Ability to interpret reference materials, such as graphs, curves, tables etc.
JPM: Calculate Boron Addition for blocking P-11 with 0 and 1 untrippable control rod.
R, M, P1 A2 - Conduct of Operations G2.1.19 (3.9) Ability to use available indications to evaluate system or component status.
JPM: Determine RPV venting time per EOP Addendum 33.
R, M A3 - Equipment Control G2.2.14 (3.9) Knowledge of the process for controlling equipment configuration or status JPM: Determine what equipment, currently not in service, could be started after NG02 and NG04 load centers are cross connected.
R, M A4 - Radiation Control G2.3.12 (3.2) Knowledge or 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.
JPM: Estimate Total Dose for placing WPA.
R, N Topic Number of JPMs Source and Source Criteria Conduct of Operations 1
Requirement Actual Equipment Control 1
(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams) 1 Radiation Control 1
(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes) 0 Emergency Plan 1
(N)ew or Significantly (M)odified from bank (no fewer than one) 4 Total 4
Note 1: The JPM from the 2020 exam were randomly selected by placing 2 slips of paper labeled R through S in a container. IF "R" was selected then 4 slips of paper (labeled 1-4) were placed in a container for random selection. IF "S" was selected then 5 slips of paper (labeled 1-5) were placed in a container for random selection. No JPMs from the 2022 NRC exam were available for random selection as those JPMs will be used as a part of 2023 Audit Exam.
A1 This is a MODIFIED BANK JPM. The original JPM was used on the 2020 ILT NRC exam and required the applicant to calculate the volume of borated water addition of 3245 to 3560 gallons and 5893 to 5787 gallons for the situations of 0 and 1 untrippable rods respectively (RCS Boron concentration was 870 ppm and Cycle Burnup was 14000 MWD/MTU). In this modified JPM, RCS Boron concentration is 1100 ppm and Cycle Burnup was 5000 MWD/MTU. With these values, the applicant calculated the volume of borated water addition of 6546 to 6597 gallons and 9665 to 9985 gallons for the situations of 0 and 1 untrippable rods respectively.
A2 This is a MODIFIED BANK JPM. The original JPM was used on the 2016 ILT NRC Exam and required the applicant to calculate a RV Venting time between 1.72 and 1.84 minutes (using the initial conditions of CTMT pressure 3.8 pisg, CTMT temperature 167F, CTMT H2 2.3%, RCS Pressure 1925 psig). In this modified JPM, all 4 initial parameters where changed (initial conditions of CTMT pressure 9.5 pisg, CTMT temperature 225F, CTMT H2 1.4%, RCS Pressure 500 psig). This results in a maximum venting time of 17.20 minutes (17.16 to 17.28 minutes is acceptable).
A3 This is a MODIFIED BANK JPM. The original JPM was used on the 2016 ILT NRC Exam and required the applicant to determine what loads could be started after NG01 and NG03 are cross connected in support of maintenance. The JPM is modified such that the initial amps loading are different with different equipment not in service. With these conditions, the applicant determined that only PJ31, CHGR 250 VDC NO 2, or SGK04B, CTRL RM A/C Unit B, can be started without exceeding the 1200 amp limit. If the applicant selects any other idle 'B' Train component, the JPM is unsatisfactory.
A4 This is a NEW JPM. The applicant is to estimate the total dose that will be received by ALL operations personnel (both the tag placer and independent verifier) when placing the WPA when provided with 2 survey maps and time estimates for each WPA step. Furthermore, the applicant is directed to use the dose rate at 30 cm, use the average of dose rates values when a range is present on the survey, and include an additional 10 mrem of dose for an individual to independently verify tag placement. Upon completion of this JPM, the applicant determined the doses from individual tag steps are 6 mrem, 7.5 mrem, 3 mrem, and 5 mrem with a total dose estimate of 31.5 mrem.
Form 3.2-1 Administrative Topics Outline Facility: _____ Callaway ___________________
Date of Examination: ___9/11/2023___
Examination Level:
RO SRO Operating Test Number: __2023-1___
Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
A5 - Conduct of Operations G2.1.40 (3.9) Knowledge of refueling administrative requirements.
JPM: Review OSP-SF-00003, Pre-Core Alteration Verifications, to determine if core alterations can begin.
R, D A6 - Conduct of Operations G2.1.25 (4.2) Ability to interpret reference materials, such as graphs, curves, and tables.
JPM: Determine required actions to isolate the flowpath for inoperable Containment Isolation Valves.
R, N A7 - Equipment Control G2.2.44 (4.4) Ability to interpret control room indication to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.
JPM: Determine RCS Leakage Action level and surveillance frequency.
R, N A8 - Radiation Control G2.3.11 (4.3) Ability to control radiation releases.
JPM: Determine the Administrative requirements after terminating a Liquid Release due to high dose rates.
R, N A9 - Emergency Plan G2.4.44 (4.4) Knowledge of emergency plan protective action recommendations.
JPM: Complete CA 2843, PAR Flowchart.
R, M
Topic Number of JPMs Source and Source Criteria Conduct of Operations 2
Requirement Actual Equipment Control 1
(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams) 0 Radiation Control 1
(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes) 1 Emergency Plan 1
(N)ew or Significantly (M)odified from bank (no fewer than one) 4 Total 5
A5 This is a Bank JPM that was used on the 2019 NRC Exam. The applicant is directed to review OSP-SF-00003, Pre-Core Alteration Verifications, Attachment 2 and 3 to determine if core alterations can begin. The applicant will determine that core alterations can not begin due to the configuration of CTMT Purge Gas Detectors, GTRE0022 and GTRE0033. Additionally, direct communications have not been established supporting core alterations.
A6 This is a NEW JPM. The applicant is provided 2 inoperable Containment Isolation Valves (BLHV8047, RX M/U WTR OUTER CTMT HV ISO and EMHV8823, SI/ACC INJ TEST LINE ISO HV) due to exceeding their individual leakage limits. Using ODP-ZZ-00036, Technical Specification Application for CTMT Isolation Valves, Attachment 1, the applicant determines that Figures 5a and Figure 15 are applicable. From these figures, the applicant determines that for, BLHV8047, BL8046 (check valve) must be used to isolate the flowpath AND flow through BL8046 must be secured. Furthermore, the applicant determined that, for EMHV8823, EMHV8835 AND EMV0067 must be closed to isolate the flowpath. Neither inoperable valve, BLHV8047 and EMHV8823, can be used to isolate its flowpath.
A7 This is a NEW JPM. When provided a series of daily RCS Unidentified Leakage and Primary to Secondary Leakage data, the applicant must determine the current RCS Leakage action level and the frequency of the Primary to Secondary leak rate surveillance. The applicant determined the Highest Tier Action Level for Unidentified RCS leak rate is Tier Three. Additionally, the applicant determined that CTP-ZZ-02590, Primary to Secondary leak rate determination, should be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
A8 This is a NEW JPM. With a liquid release in progress, the applicant will be given sample results indicating unexpected high dose rates. Due to the release rate and associated dose at the site boundary for 60 minutes of exposure, FSAR
16.11.1.2 Condition A and B limits are not met and the Shift Manager is required to notify 8 individuals of the event.
A9 This is a MODIFIED TIME CRITICAL JPM. (The note below this summary explains the JPM modification and original JPM usage.) At the completion of this JPM, the applicant determined that the Affect Sectors are A, B, C. These affected sectors should be Sheltered in Place 10 miles downwind. All sectors within a 2 mile radius of the plant should be Sheltered in Place. Additionally, the applicant completed CA2843 sections: map outline, method, evacuate, reason for type of PAR, and impediments considered correctly (per the included KEY) in less than or equal to () 15 minutes of the start of the JPM.
Note: while completing a PAR was a part of the 2022 and 2020 ILT Exams, this JPM is different from the previous JPMs in multiple ways including: impediments exist (for a different reason) for a non rapidly progressing event, different affects zones, and a different radius (5 miles vs 10 miles) are also present. Therefore, previous knowledge of either 2022 or 2020 JPM would not provide an unfair advantage for the SRO applicant as this JPM is significantly modified.
Form 3.2-2 Control Room/In-Plant Systems Outline Facility: ___Callaway _______________ Date of Examination: ____9/11/2023____
Operating Test Number: __2023 -1____
Exam Level:
RO SRO-I SRO-U System/JPM Title Type Code Safety Function Control Room Systems S1 - 004A2.14 (4.0/4.2) 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: Emergency boration JPM: Emergency Actions / CVCS: Respond to Loss of Shutdown Margin / Emergency Borate. - RO and SRO-I Applicants ONLY D, A, L, S 1
S2 - 006A4.02 (4.2) Ability to manually operate and/or monitor in the control room: ECCS valves JPM: Safety Injection System: Open EMHV8802A ('A' SI Pump Discharge to Hot Leg Inj) in Mode 1,2 or 3. - RO and SRO-I Applicants ONLY EN, D, P1, S
2 S3 - 010A4.05 (3.2) Ability to manually operate and/or monitor in the control room: PZR auxiliary spray valves JPM: Pressurizer Pressure Control System: Initiate Aux Spray in Mode 3 per OTN-BB-00005. - ALL Applicants L, N, S 3
S4 - 005A3.01 (4.1) Ability to monitor automatic features of the Residual Heat Removal System, including: Automatic RHR suction swap-over JPM: RHR System / Perform RHR Suction Valve Automatic Actuation Test on EJHV8701A, RHR Pump A Suction Isolation. - RO Applicants ONLY EN, L, N, S 4P S5 - 103A1.01 (3.9) Ability to predict and/or monitor changes in parameters associated with operation of the Containment System, including: Containment pressure, temperature, and/or humidity JPM: Containment System / Lower Containment Temperature. - RO and SRO-I Applicants ONLY N, S 5
S6 - 062A4.01 (3.5) Ability to manually operate and/or monitor in the control room: All breakers JPM: Reenergized a dead safety related 4160 VAC bus. -
ALL Applicants N, A, L, S, 6
S7 - 016 A2.01 (3.3/3.4) 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: Detector/transmitter failure JPM: Initiate Cold Overpressure Mitigation System then respond to a wide range pressure channel failure. - ALL Applicants EN, M, A, S, L, 7
S8 - 050A4.01 (3.8) Ability to manually operate and/or monitor in the control room: Initiate/reset system JPM: Control Room Ventilation / Actuate B Train Control Room Ventilation Isolation with failures requiring contingency actions. - RO and SRO-I Applicants ONLY EN, A, N, S 9
In-Plant Systems P1 - 061A1.04 (3.8) Ability to predict and/or monitor changes in parameters associated with operation of the Auxiliary/Emergency Feedwater System, including: AFW source tank level JPM: Auxiliary Feedwater System / Emergency Makeup Water to CST per EOP Addendum 23. - RO and SRO-I Applicants ONLY D, E, P1 4S P2 - 006K5.13 (3.6) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Emergency Core Cooling System: Hot-leg injection JPM: Emergency Core Cooling System / Perform ES-1.4, Transfer to Hot Leg Recirculation Step 3. - ALL Applicants E, A, N, R, L
2 P3 - 078K4.06 (2.9) Knowledge of the Instrument Air System design features and/or interlocks that provide for the following: Maintaining dry air JPM: Instrument Air System / Place Train 'A' Air Dryers in service. - ALL Applicants D
8
Note 1: The JPMs from the 2020 exam were randomly selected by placing 11 slips of paper labeled A through K in a container. No JPMs from the 2022 NRC exam were available for random selection as those JPMs will be used as a part of 2023 Audit Exam.
Control Room System JPMs S1 This is an Alternate Path Bank JPM that has not been used on a ILT NRC exam since 2014. This JPM is for RO and SRO-I Applicants ONLY. The applicant will perform actions of OTO-ZZ-00003, Loss of Shutdown Margin / Dilution Event to emergency borate with the plant in Mode 3. The applicant will have starting at least one Boric Acid Transfer Pumps (BG HIS-5A and/or 6A) and then opening both BG HIS-110A and 110B to establish a flowpath from the Boric Acid Transfer Pumps to the VCT (due to a failure with the original flowpath to the charging pump suction).
S2 This is a Bank JPM that was on the 2020 ILT NRC Exam. This JPM is for RO and SRO-I Applicants ONLY. The applicant will perform actions of OTN-EM-00001, Safety Injection System, Section 5.11 to open EMHV8802A, SI Pump A Disch To Hot Leg Inj without affecting RWST or RHUT inventory, starting the 'A' Train SI pump nor affecting the lineup (and therefore operability) of the opposite train which would place the plant in a LCO 3.0.3 condition.
S3 This is a New JPM. This JPM is for ALL Applicants. The plant is in Mode 3 with the 'D' RCP Off, normal letdown and charging in service. The applicant will Code License Level Criteria RO SRO-I SRO-U (A)lternate path 5
5 3
(C)ontrol room (D)irect from bank 4
4 1
(E)mergency or abnormal in-plant 2
2 1
(EN)gineered safety feature (for control room system) 4 3
1 (L)ow power/shutdown 6
5 4
(N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 7 6
4 (P)revious two exams (randomly selected) 2 2
0 (R)adiologically controlled area 1
1 1
(S)imulator
perform Section 5.8 of OTN-BB-00005 to place Auxiliary Spray in service. This will consist of opening BGHV8145 and closing BGHV8147. Next the applicant will energized one or more of the PZR Backup Heater Groups A&B by placing BB HIS-51A and / or BB HIS-52A in close. Lastly, the applicant placed the PZR Master Controller, BB PK-455A, at 50% in Auto by pushing the MAN pushbutton, then pushing the output up arrow until the indicator read 50% and then depressed the Auto pushbutton.
S4 This is a New JPM. This JPM is for RO Applicants ONLY. The plant is in Mode 4 with RCS pressure less than 425 psig. The applicant will perform Section 6.2 of OSP-EJ-00001, Testing EJHV8701A, RHR Pump A SUCT ISO, by placing the
'A' RHR pump in Pull to Lock using EJ HIS-1, then closing EJHV8804A, RHR to CHG Pumps, by pushing the close pushbutton on EJ HIS8804A, determined that the RCS >390 psig interlock test is SAT (and recorded it as so) and lastly opened EJHV8701A by pushing the open pushbutton on EJ HIS-8701A when simulated RCS pressure is lowered to less than 360 psig.
S5 This is a New JPM. This JPM is for RO and SRO-I Applicants ONLY. The plant is at 100% power during an extended heat wave. Containment Air Temperature is reaching its maximum allowed by Technical Specifications. The applicant is directed to lower Containment Air Temperature using the 'B' Essential Service Water Train. By the end of the JPM, the applicant will have started the 'B' ESW pump (EF HIS-56A to RUN), opened EF HIS-38 and EF HIS-50 using their individual pushbuttons and closed EF HIS-24,26,40 and 42 using their individual pushbuttons in order to lower Containment Temperature at the fastest rate possible.
S6 This is a New Alternate Path JPM. This JPM is for ALL Applicants. The plant is in Mode 3 with NB01 deenergized due to a loss of offsite power and a failure of the 'A' EDG, NE01. The crew has entered OTO-NB-00001, Loss of Power to NB01, and is performing actions when normal offsite power becomes available.
Using OTN-NB-0001A, power is momentarily restored to NB01 from the switchyard, but is lost again when breaker, 52-3, SAFEGUARD XFMR 13.8 KV PCB, trips open. In order to reenergize NB01, the applicant must determine the only viable source is from the Startup Transformer via XNB02 (B Train) and implement OTN-NB-0001A, Addendum 5, Section 3.2, Re-Energizing NB01 From Alternate Source XNB02. The applicant will then place NB HS-10 in Alt FDR BRKR, place NB HS-7 to ON, and close NB0109 by placing NB HIS-3 to CLOSE in order to successfully reenergize NB01.
S7 This is a Time Critical Alternate Path Modified Bank JPM with the bank JPM on the 2017 ILT NRC Exam as JPM S3. This JPM is for ALL Applicants. The
applicant will be directed to place the Cold Overpressure Mitigation System in service (aka "arm COMS") per OTN-BB-00005, Section 5.6. After COMS is placed in service (both BB HS-8000A and BB HS-8000B in ARM), WR pressure instrument, BB PT-403, will fail to above the COMS pressure initiation setpoint resulting in BB PCV-456A, B PZR PORV, failing opening. Per Annunciator 64C, the applicant will be required to close BB HIS-8000B, B PORV block valve, by returning BBHS8000B to block and depressed BB HIS-8000B's CLS pushbutton because the B PORV will not close using its handswitch. The JPM is complete when the applicant has closed BB HIS-8000B, B PORV block valve, using its handswitch within 7 minutes of the 'B' PORV failing open.
S8 This is a New Alternate Path JPM. This JPM is for RO and SRO-I Applicants ONLY. The plant is at full power with maintenance complete on some 'B' Train CRVIS logic and components. The applicant will perform Section 5.5 of OTN-GK-00001, Control Building HVAC System, to actuate 'B' Train CRVIS and place GK HIS-8 to the OFF position after actuation. Then applicant should determine
'B' Train CRVIS will not fully actuate and the applicant will have to implement Section 5.9, Contingency Actions for the Failure of CRVIS to Fully Actuate, and place 3 hand switches (GK HIS-83, GK HIS-30, and GK HIS-40) in Pull to Lock.
In Plant System JPMs P1 This is a BANK JPM that was on the 2020 ILT NRC Exam. This JPM is for RO and SRO-I Applicants ONLY. The applicant will perform actions of EOP Addendum 23, Local CST Emergency Fill, connecting a fire hose to the CST emergency fill connector on APV0043 and then unlock and open APV0043 and then open a fire hydrant thereby establishing fire water emergency fill to the CST.
P2 This is a New Alternate Path JPM. This JPM is for ALL Applicants. The applicant will perform the inplant actions of ES-1.4, Transfer to Hot Leg Recirculation, Step #3. Due to a failure of EMHV8802B, B Discharge To Hot Leg Injection Valve, to open locally, the applicant must apply the RNO to expand the isolation boundary for the transfer. By the completion of the JPM, the applicant manually closed EMHV8835, SI Pumps Discharge to Cold Leg Injection Isolation Valve, and opened EMHV8802A, SI Pump A Discharge to Hot Leg Injection Valve P3 This is a Bank JPM that has not been used on a ILT NRC exam since 2014. This JPM is for ALL Applicants. The applicant will perform actions on OTN-GK-00001, Addendum 2, Section 5.1 to place the 'A' Train Air Dryers in service and put the 'B' Train Air Dryers in standby. The applicant will be required to open 2 manual valves (KAV0755/0766), throttle 1 valve (KAV0759), and then place
KAHS0358 in the "DRYER A" position in order to successfully place A Train in service.
Page 1 of 38 Form 3.3-1 Scenario Outline Facility:
Callaway Scenario #: 1 Scenario Source:
Site Op. Test #: 2023-1 Examiners:
Applicants/
Operators:
Initial Conditions:
100% MOC Turnover:
NCP is OOS for emergent bearing maintenance. 'B' Train is protected.
Testing is required on CIV#2, ACFCV0054. Perform a downpower to 97% then perform OSP-AC-00006, Section 6.3, to test ACFCV0054.
Event No.
Malf. No.
Event Type*
Event Description 1
N/A CRS (R)
ATC (R)
BOP (R)
Perform a down power from 100% to 97% per OTG-ZZ-00004, Addendum 3.
2 N/A CRS (N)
BOP (N)
Perform OSP-AC-00006, Section 6.3, to test Combined Intermediate Valve #2, ACFCV0054.
3 BB /
BBPT0455 CRS (I, TS)
ATC (I, MC)
PZR Pressure instrument Fails Low, OTO-BB-00006, Pressurizer Pressure Control Malfunction (Technical Specifications 3.3.1, 3.3.2, and 3.3.4) 4 PBG01B_1 PEG01A&
C_2 CRS (C, TS)
ATC (C, MC)
'B' CCP pump trip, OTO-BG-00001 Pressurizer Level Control Malfunction. Failure of either 'A' CCW train pump to autostart when the 'A' CCP is placed in service, manual control available.
(Technical Specification 3.5.2, 16.1.2.2, and 16.1.2.4) 5 CD /
PCD01 = 0 CRS (C)
BOP (C, MC)
Trip of the Main Seal Oil Pump, OTO-MA-00002 MG Hydrogen Leakage / Seal Oil System Malfunction. Emergency Seal Oil Pump Fails to autostart, manual control available.
6 BB /
BB001_C CRS (M)
ATC (M)
BOP (M)
RCS LOCA. RX Trip E-0, Reactor Trip or Safety Injection, then transition to E-1, Loss of Primary or Secondary Coolant.
7 AL /
PAL01A &
PAL02 = 1 CRS (C)
BOP (C, MC)
During E-0, the TDAFP pump will trip and the 'A' MDAFP will fail to autostart but can be started from the main control boards, manual control available.
8 SA /
SIS_A_Blo ck_Auto SIS_B_Blo ck_Auto CRS (I)
ATC (I, MC)
Safety Injection fails to Automatically Actuate.
CT-1, Manually actuate at least one train of SIS-actuated safeguards equipment before the completion of E-0 Immediate Actions (E-0 Steps 1-4).
9 EN /
SB029B &
SB032B CRS (I)
ATC (I, MC)
Containment Spray fails to Automatically actuate.
CT-2, Manually actuate at least one train of Containment Spray before the completion of E-0 Attachment A.
10 EJ / PEJ01 xMI179S=1 CRS (C)
BOP (C)
Loss of both RHR pumps during the performance of E-1, transition to ECA-1.1, Loss of Emergency Coolant Recirculation.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Callaway 2023-1 Scenario #1 Event Description Page 2 of 38 The Plant is 100%, MOC with the NCP tagged OOS for emergent bearing maintenance.
After the crew takes the watch, the crew will down power the unit from 100% to 97% per OTG-ZZ-0004 Addendum 3 in order to perform the CIV surveillance.
The crew will perform OSP-AC-00006, Combined Intermediate Valve Stroke Test, Section 6.3, to test CIV#2 (ACFCV0054).
After CIV#2 is tested, the upper selected PZR Pressure Channel BB PT-455 fails low.
Annunciator 83C, RX PARTIAL TRIP, will alarm to alert the operator of the malfunction. The CRS should enter OTO-BB-00006, Pressurizer Pressure Control Malfunction, and transfer pressure control to remove the failed channel. This failure will result in Technical Specification 3.3.1, 3.3.2, and 3.3.4 not being met.
Once Technical Specifications are determined, the 'B' CCP trips on overcurrent. The crew should enter OTO-BG-00001, PZR Level Control, and place the 'A' CCP in service. When the
'A' CCP is placed in service, neither 'A' CCW train pump will autostart. Manual control of both
'A' CCW and 'C' CCW pumps is available and either pump can be manually started from CR panel RL020. This 'B' CCP failure will result in Technical Specification 3.5.2, 16.1.2.2, and 16.1.2.4 not being met.
After the Technical Specification determination, the Main Generator Seal Oil Pump will trip. The crew will respond with OTO-MA-00002, MG Hydrogen Leakage / Seal Oil System Malfunction and manual start the Emergency Seal Oil Pump as it failed to autostart.
After the MG Seal Oil system is stabilized, a RCS LOCA occurs. The crew will enter E-0, Reactor Trip or Safety Injection, and at Step #4 manually initiate Safety Injection as it failed to automatically occur. The crew will proceed through E-0 and perform E-0 Attachment A and transition to E-1, Loss of Reactor or Secondary Coolant, at E-0 Step #16.
During the performance of E-0, the TDAFP pump will trip 45 seconds after it starts (and cannot be reset / restarted) and the 'A' MDAFP will fail to autostart. The 'A' MDAFP can be started from the main control boards.
As containment pressure exceeds 27 psig, Containment Spray fails to automatically actuate.
The crew will manually actuate at least one train of Containment Spray at E-0 Attachment A step #A9.
During the performance of E-1, the 'A' RHR trips on overcurrent and the 'B' RHR won't start from the control room or locally. At E-1, step #12, 'Initiate Evaluation of Plant Status', the crew will determine that RHR is not available and perform the RNO to transition to ECA-1.1, Loss of Emergency Coolant Recirculation.
The scenario is complete when the crew has transitioned ECA-1.1 and performed Step #7 to secure the correct number of running CTMT spray pumps as a function of RWST level and CTMT pressure.
Callaway 2023-1 Scenario #1 Critical Tasks Page 3 of 38 CT-1 CT-2 Critical Tasks Manually actuate at least one train of SIS-actuated safeguards equipment before the completion of E-0 Immediate Actions (E-0 Steps 1-4).
Manually actuate at least one train of Containment Spray before the completion of E-0 Attachment A.
EVENT #
6 7
Safety significance Failure to manually actuate SI under the postulated conditions constitutes mis-operation or incorrect crew performance in which the crew does not prevent degraded emergency core cooling system (ECCS)capacity.
In this case, SI can be manually actuated from the control room. Therefore, failure to manually actuate SI also represents a failure by the crew to demonstrate the following abilities:
Effectively direct or manipulate engineered safety feature (ESF) controls that would prevent (degraded emergency core cooling system (ECCS)capacity)
Recognize a failure or an incorrect automatic actuation of an ESF system or component Take one or more actions that would prevent a challenge to plant safety Failure to manually actuate at least one train of Containment Spray under the postulated conditions demonstrates the inability of the crew to recognize a failure or an incorrect automatic actuation of an ESF system or component. In this case, at least one train of Containment Spray can be manually actuated from the control room. Therefore, failure to manually actuate at least one train of Containment Spray also represents a failure by the crew to demonstrate the ability to effectively direct or manipulate engineered safety feature (ESF) controls" that would prevent or challenge a radiological barrier.
Cueing Indication and/or annunciation that that Sl is required PRZR pressure less than SI actuation setpoint, 1849 psig Containment pressure greater than SI actuation setpoint, 3.5 psig Subcooled margin less than the foldout page criterion for SI actuation in ES-0.1 PRZR water level less than the foldout page criterion for SI actuation in ES-0.1 No indication or annunciation that SI is actuated Indication and/or annunciation that that CSAS (Containment Spray Actuation Signal) is required Containment Pressure greater than 27 psig Annunciator 59A CSAS Annunciator 59B CISB No indication or annunciation that Containment Spray is actuated Performance indicator Manipulation of controls as required to actuate at least one train of Sl SB HS-27 or SB HS-28 Manipulation of controls as required to actuate at least one train of Containment Spray by starting the pump and opening its discharge valve:
EN HIS-3 and EN HIS-6 (A Train)
EN HIS-9 and EN HIS-12 (B Train)
Performance feedback Indication that at least one train of SI - Actuated LOCA Sequencer annunciator 30A - Lit LOCA Sequencer annunciator 31B - Lit SB069 SI Actuate Red Light - Lit SOLID (NOT blinking)
Indication that at least one train of Containment Spray - Actuated CTMT Pressure lowering EFSAS Status panels CSAS Sections White Lights LIT EFSAS Status panels CISB Sections White Lights LIT Justification for the chosen performance limit The crew has had ample opportunity to recognize the need for Sl and the fact that Sl has not automatically actuated. Given the postulated plant conditions, transition from E-0 to ES-0.1 constitutes an error in diagnosing plant conditions and using the E-0 procedure.
The performance of this critical task by the completion of E-0 Attachment A is justified on the basis of minimizing the differential pressure between the containment internal atmosphere and the external atmosphere. Continuous Action Step E-0 Attachment A, Step A9 directs actuating Containment Spray if CTMT pressure is or has exceeded 27 psig, and failure to initiate when directed by the Emergency Procedures represents a failure to protect the health and Safety of the public. CTMT Pressure > 27 psig with no CTMT Spray pumps running is a Orange Path and requires a transition to FR-Z.1. Containment cooling is required to remove heat from the containment internal atmosphere and thereby prevent containment internal pressure from exceeding the design values. Furthermore, manual actuation of at least one train of containment spray for iodine removal is an assumption made in the FSAR for onsite and offsite dose projections. Failure to "scrub" the internal containment atmosphere represents elevated dose rates and a failure to protect the health and safety of the public.
PWR Owners Group Appendix CT-2, Manually actuate SI CT-3, Manually Actuate Containment Cooling 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.
Page 4 of 38 References OSP-AC-00006, Combined Intermediate Valve Stroke Test, rev 8 OTO-BB-00006, Pressurizer Pressure Control Malfunction, rev 20 OTO-BG-00001, Pressurizer Level Control Malfunction, rev 25 OTO-MA-00002, MG Hydrogen Leakage / Seal Oil System Malfunction, rev 14 E-0, Reactor Trip or Safety Injection, rev 28 E-1, Loss of Reactor or Secondary Coolant, rev 24 ECA-1.1, Loss of Emergency Coolant Recirculation, rev 15 Technical Specification 3.5.2, ECCS - Operating Technical Specification 3.3.1, Reactor Trip System Instrumentation Technical Specification 3.3.2, ESFAS Instrumentation ODP-ZZ-00025, EOP/OTO User's Guide OTG-ZZ-00004, Addendum 3, Planned Power Changes from Full Power, rev 12 PRA Systems, Events or Operator Actions (from PRA update 9.01):
An intermediate LOCA represents a 8% contribution to CDF by Accident type.
The RHR system is the 8th highest risk important system.
Target Quantitative Attributes per Scenario/Scenario Set (NUREG 1021, Table 3.4-1)
Quantitative Attribute Target per Scenario Actual Attributes Events after EOP entry 1-2 4
Abnormal events 2-4 3
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 1
Preidentified CTs 2 or more 2
The following events qualify as manual control of an automatic function per ES-3.4.2(4):
Event#3, PZR Pressure instrument fails low requiring manual control to stabilize PZR pressure (ATC)
Event#4, CCW failure to autostart; can be manually started from CR panel (ATC)
Event#5 Main Seal Oil Pump trip with failure of the Emergency Seal Oil Pump to autostart; can be manually started from CR panel (BOP)
Event#7, Failure of the "a" MDAFP pump to autostart post trip; can be manually started from CR panel (BOP)
Event#8, SI fails to auto actuate; can be manually actuated from CR panel (ATC)
Event#9, CSAS fails to auto actuate; can be manually actuated from CR panel (ATC)
Scenario #1 is a LBLOCA in which Containment pressure exceeds 27 psig with the critical task of actuating at least one train of Containment Spray. The leak size is much larger than the amount in 2022 scenario #4 and 2022 scenario #1 requiring different actions / Critical Tasks.
For example, the Critical Task of tripping RCPs is not applicable due to the LOCA size but was critical in 2022 scenario#4. In 2020 scenario #1, RCS pressure did drop below 1425 psig.
Callaway 2023-1 Scenario#1 Simulator Lesson Plan Page 5 of 38 Scenario Setup Guide:
IC-47 (password 047)
=======SCENARIO PRELOADS / SETUP ITEMS==========
Pre-Loads:
o JRM11EXAMSAFE = TRUE(1) o Ensure BB PT455 is selected for PZR Pressure Control o Ensure the NCP is tagged OOS BG47PB0301_BKRTA_BKPOS = 3 (rackout) switch in PTL w/tag o Ensure the 'B' CCP is in service along with the B CCW train.
For specific event malfunctions, see the individual event page and/or Attachment 2, Scenario File Manager o PEG01A_2=1 (failure of the CCW pump to Auto start when CCP in train started) o PEG01C_2=1 (failure of the CCW pump to Auto start when CCP in train started) o SIS_A_Block_Auto = 0 (SI auto actuation blocked) o SIS_B_Block_Auto = 0 (SI auto actuation blocked) o SB029B_CSAS = 0 (CTMT Spray Auto start blocked) o SB032B_CSAS = 0 (CTMT Spray Auto start blocked) o X17I179S 1 delay=0 ramp=0 on=0 off=0 o PEJ01A 1 cd='X01I165A EQ 1' delay=20 ramp=0 on=0 off=0 o PAL01A_3 1 delay=0 ramp=0 on=0 off=0 o PAL02_1 1 cd='rec0009 le 1' delay=45 ramp=0 on=0 off=0 o CD03RL_2ATVSP 0 delay=0 ramp=0 on=0 off=0 o CD03RL_1ATVSP 0 delay=0 ramp=0 on=0 off=0
Page 1 of 35 Form 3.3-1 Scenario Outline Facility:
Callaway Scenario #: 2 Scenario Source:
Site Op. Test #: 2023-1 Examiners:
Applicants/
Operators:
Initial Conditions:
75%, holding per power supply supervisor Turnover:
NCP is OOS for emergent bearing maintenance. 'B' Train is protected. Maintenance on transformer XPB117 is complete with WPA tags removed. Perform OTN-PB-00001 Addendum 2, Section 5.5 to restore non safety 4.16kV busses PB117 and PB218 to their normal lineup.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A CRS (N)
BOP (N)
Return Bus PB117 To Transformer XPB117 From Transformer XPB218 per OTN-PB-00001, Addendum 2, Section 5.5.
2 ABPV00 02A_1 CRS (C)
BOP (C, MC)
Atmospheric Steam Dump 'B' fails open with manual control available. OTO-AB-00001, Steam Dump Malfunction.
3 BG /
x01A104 P = 9 CRS (I)
ATC (I, MC)
BG TK-130, LTDN HX Outlet Temp CTRL, failure in automatic, manual control available.
= 27.5 CRS (TS)
AL PI-38, fails midscale. OTO-AL-00001, Aux Feed Pump LO Suction Pressure Channel Failure, (Technical Specification 3.3.2) 5 SE /
SEN0043 SRO (I, TS)
ATC (I)
BOP (I)
Power Range Channel N43 fails high. OTO-SE-00001, Nuclear Instrument Malfunction, (Technical Specification 3.3.1) 6 AC / AC YE0017 SRO (C)
ATC (C)
BOP (C)
Main Turbine bearing failure resulting in High Vibrations requiring load reduction. OTO-AC-00002, Turbine Vibration 7
SF /
SF006 =
Both Modes SRO (M)
ATC (M)
BOP (M, MC)
Reactor Failure to trip, E-0 Reactor Trip or Safety injection CT#1, Insert negative reactivity into the core by opening both PG 19 and PG 20 feeder breakers (PG HIS-16 and PG HIS-18) within 5 minutes of the failure of the Manual Reactor Trip.
8 AB /
AB002_B
= 1300 CRS (C)
BOP (C)
'B' Faulted SG, Transition from E-0 to E-2.
CT#2, Isolate the faulted B SG before transition out of E-2 9
SA /
SAS9xx_
2=1 CRS (C)
BOP (C, MC)
'B' MSIV, AB HIS-17, fails to automatically close. May be manually closed from control room panel.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control SPARE SCENARIO
Callaway 2023-1 Scenario #2 Event Description Page 2 of 35 The Plant is at ~75% holding per power supply supervisor.
After the reactivity brief is complete, the crew will return non safety related bus PB117 To transformer XPB117 From transformer XPB218 per OTN-PB-00001, Addendum 2, Section 5.5.
Once busses PB117 and PB218 are returned their normal lineup, the 'B' Atmospheric Steam Dump fails open. The BOP operator should close the ABPV0002 using manual control. The crew should enter OTO-AB-00001, Steam Dump Malfunction.
Once Technical Specification 3.7.4 is reviewed, BG TK-130 will fail in automatic causing Letdown (LTDN) temperatures to rise. Annunciator 39B will alarm at a letdown temperature of 120°F and LTDN will divert to the VCT if temperature reaches 137°F. The crew will respond by performing Annunciator 39B actions which includes taking manual control of BG TK-130.
Once LTDN temperatures have been restored to normal, AL PI-38 will fail midscale. The crew will enter OTO-AL-00001, Aux Feed Pump Lo Suction Pressure Channel Failure and initiate actions to repair the instrument. Technical Specification 3.3.2 is not met.
After Technical Specification are addressed, Power Range Nuclear Instrument Channel N43 fails high. The crew will enter OTO-SE-00001, Nuclear Instrument Malfunction, to bypass channel N43 and restore control rods to desired position. Technical Specification 3.3.1 is not met.
Once Technical Specifications are addressed, a Main Turbine bearing will begin to fail resulting in High Vibrations will develop. High Vibrations will rise and lower above the alarm setpoint based on simulator booth driver actions. The crew will enter OTO-AC-00002, Turbine Vibration, and perform a load reduction while monitoring vibration levels. Eventually, turbine vibrations continue to rise and the crew will be required to manually trip the reactor per OTO-AC-00002 Step #10 RNO.
The manual reactor trip will fail and the crew will enter E-0 and perform E-0 Step #1 RNO b. to shutdown the reactor. Opening both PG-19 and PG-20 feeder breakers (PG HIS-16 & 18) will successfully shutdown the reactor.
When the reactor and main turbine shutdown, a fault develops on 'B' SG. The crew will perform steps of E-0 and at Step #14 transition to E-2, Faulted SG Isolation. The crew will fast close all MSIVs using AB HS-79/80 and determine that 'B' MSIV fails to close. The crew should close the 'B' MSIV, AB HIS-17, from control room panel RL026. The crew will complete 'B' SG isolation per E-2 step #4.
The scenario is complete when the crew has isolated 'B' SG per E-2 or transitioned from E-2 without 'B' SG isolated.
Callaway 2023-1 Scenario #2 Critical Tasks Page 3 of 35 CT-1 CT-2 Critical Tasks Insert negative reactivity into the core by opening both PG 19 and PG 20 feeder breakers (PG HIS-16 and PG HIS-18) within 5 minutes of the failure of the Manual Reactor Trip.
Isolate the faulted B SG before transition out of E-2 EVENT #
7 8
Safety significance In the scenario, failure to insert negative reactivity by one of the methods listed previously can result in the needless continuation of an extreme or a severe challenge to the subcriticality CSF. Although the challenge was not initiated by the crew (was not initiated by operator error), continuation of the challenge is a result of the crew's failure to insert negative reactivity and shutdown the reactor when procedurally directed.
Failure to isolate a faulted SG that can be isolated causes challenges to CSFs beyond those irreparably introduced by the postulated conditions.
Failure to isolate a faulted SG can result in challenges to the following CSFs:
Integrity Subcriticality Containment (if the break is inside containment)
Cueing Indication of ATWS. The reactor is not tripped (i.e. rod bottom lights not LIT and neutron flux is NOT lowering) are indication that the reactor is not shutdown and that a manual reactor trip is not effective)
Both of the following:
Steam pressure and flow rate indications that make it possible to identify B SG as faulted AND Valve position and flow rate indication that AFW continues to be delivered to the faulted B SG Performance indicator Manipulation of controls in the control room as required to initiate the insertion of negative reactivity into the core (at least one of the following)
Open supply breakers to PG19 and PG20.
o PG HIS-16 and PG HIS-18 to Trip position ISOLATE AFW flow to faulted SG(s):
CLOSE associated MD AFP Flow Control Valve(s):
CLOSE associated TD AFP Flow Control Valve(s):
Identify that ABV0085 (SG B) should be closed CLOSE 'B' MSIV:
AB HIS17 Performance feedback Crew will observe the following:
Indication of a negative SUR on the intermediate range of the excore NIS Indication of less than 5% power on the power range of the excore NIS Green Light Lit, Red Light OFF on PG HIS-16 and PG HIS-18 Crew will observe the following:
Any depressurization of intact SGs stops AFW flow rate indication to faulted SG of zero Justification for the chosen performance limit With rising main turbine vibrations, within 5 minutes is an acceptable amount of time for the crew to shutdown the reactor in order to be able to secure the main turbine such that large amounts of physical damage to plant structure does not occur. This potential damage could result in the loss of secondary system such as a heat sink, and /or methods to feed the SGs to maintain a heatsink along with potential damage to onsite AC distribution systems.
before transition out of E-2 is in accordance with the PWR Owners Group Emergency Response Guidelines. It allows enough time for the crew to take the correct action while at the same time preventing avoidable adverse consequences.
PWR Owners Group Appendix CT-52, Insert negative reactivity into the core CT-17 Isolate faulted SG 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.
Callaway 2023-1 Scenario #2 References Page 4 of 35 References OTN-PB-00001 Addendum 2, Energizing and Cross-tying Buses PB117 and PB218, rev 7 OTO-AB-00001, Steam Dump Malfunction, rev 19 OTO-AL-00001, Aux Feed Pump LO Suction Pressure Channel Failure, rev 8 OTA-RK-00018, Addendum 39B, LTDN HX DISCH TEMP HI, rev 1 OTA-RK-00018, Addendum 39A, LTDN HX TEMP HI DIVERT, rev 0 OTO-SE-00001, Nuclear Instrument Malfunction, rev 33 OTO-AC-00002, Turbine Vibration, rev 29 E-0, Reactor Trip or Safety Injection, rev 28 E-2, Faulted SG Isolation, rev 12 ES-0.1, Reactor Trip Response, rev 25 Technical Specification 3.3.1, Reactor Trip System Instrumentation Technical Specification 3.7.4, Atmospheric Steam Dump Valves ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions:
An ATWS represents a 2% contribution to CDF A Main Steam Line Break (Faulted SG) represents a 24% contribution to CDF Target Quantitative Attributes per Scenario/Scenario Set (NUREG 1021, Table 3.4-1)
Quantitative Attribute Target per Scenario Actual Attributes Events after EOP entry 1-2 2
Abnormal events 2-4 4
Major transients 1-2 1
EOPs entered/requiring substantive actions 1-2 2
Entry into a contingency EOP with substantive actions 1 per scenario set 0
Preidentified CTs 2 or more 2
The following events qualify as manual control of an automatic function per ES-3.4.2(4):
Event#2, 'B' ASD Fails open, can be manually closed from CR panel (BOP)
Event#3, BG TK-130 controller failure, Letdown temperatures can be manually controlled (using BG TK-130 in manual) from CR panel (ATC)
Event #7, The ATWS is mitigated by the BOP operator taking manual control of the feeder breakers of PG19 and PG20 i.e. the MG sets to deenergize the buses and shutdown the reactor.
Event #9, AB HIS-17, fails to automatically close but can be manually closed from control room panel (BOP).
Callaway 2023-1 Scenario #2 Page 5 of 35 Scenario #2 consists of an ATWS in which opening both PG-19 and PG-20 feeder breakers (PG HIS-16 & 18) will successfully shutdown the reactor per E-0 Step #1 RNO b. There is no transition to FR-S.1 required. After the reactor is shutdown, a fault develops on 'B' SG. The crew will fast close all MSIVs using AB HS-79/80 and determine that 'B' MSIV fails to auto close.
The crew should close the 'B' MSIV, AB HIS-17, from control room panel RL026. This scenario is different from the two previous ILT exams (2020 and 2022) scenarios in that 2020 the 'A' SG was faulted but after isolation the scenario evolved into a FR-H.1 Loss of Secondary Heat Sink that required RCS bleed and Feed. While 2022 scenario #3 does have the same ATWS malfunction in that opening both PG-19 and PG-20 feeder breakers (PG HIS-16 & 18) is successful that scenario evolves into a FR-H.1 scenario not a SG fault. In summary, 2023 scenario #2 is different from the previous 2 exams due to the combination of events (ATWS then 'B' SG fault) that requires different chain of responses (E-0 to E-2).
Callaway 2023-1 Scenario#2 Simulator Lesson Plan Page 6 of 35 Scenario Setup Guide:
IC-48 (password 048)
=======SCENARIO PRELOADS / SETUP ITEMS==========
Pre-Loads:
o JRM11EXAMSAFE = TRUE(1) o Ensure the NCP is tagged OOS BG47PB0301_BKRTA_BKPOS = 3 (rackout) switch in PTL w/tag o Ensure the 'B' CCP is in service along with the B CCW train o SF006=2 Fail RX to Trip both Auto & Manual o SAS9xx_2=1 B MSIV fail to close o SAS9XX_6 1 delay=0 ramp=0 on=0 off=0 o delia SAS9XX_6 2 delay=0 cd='X25I130C EQ 1' o SIFT 20220040_workaround.sce For specific event malfunctions, see the individual event page and/or Attachment 2, Scenario File Manager
Page 1 of 40 Form 3.3-1 Scenario Outline Facility:
Callaway Scenario #: 3 Scenario Source:
Site Op. Test #: 2023-1 Examiners:
Applicants/
Operators:
Initial Conditions:
48% stable, startup on hold while an issue with the non running MFP is investigated Turnover:
NCP is OOS for emergent bearing maintenance. 'B' Train is protected.
Perform OTN-EA-00001 section 5.3 to swap service water pumps from B & C initially in service to the final configuration of A & B in service. An equipment operator is briefed and standing by.
Event No.
Malf. No.
Event Type*
Event Description 1
N/A CRS (N)
BOP (N)
Perform OTN-EA-00001, Section 5.3 to start the 'A' service water pump and section 5.4 to secure the 'C' service water pump.
2 AE /
AELT0539
= 40 CRS (I, TS)
BOP (I, MC)
'C' SG Level Instrument Fails to 25%. OTO-AE-00002, Steam Generator Water Level Control Instrument Malfunctions.
(Technical Specifications 3.3.1 and 3.3.2) 3 PG /
PG2401 CRS (I)
ATC (I)
PZR pressure control lost due to failure of PZR variable heater, Annunciator 33E PZR Heater Group Lockout.
4 SM04_B=
160 CRS (C, TS)
ATC (C)
BOP (C)
D Control Bank rod, M-4, rod insertion. OTO-SF-00001, Rod Control Malfunctions. OTO-MA-00008, Rapid Load Reduction, to lower power less than 40%. (Technical Specification 3.1.4) 5 PA / PA02 SF /
SF006 CRS (C)
ATC (C, MC)
Loss of PA02 resulting in the loss of 2 RCPs with Reactor failure to automatically trip (ATWS), E-0 Reactor Trip or Safety Injection.
CT#1, Insert negative reactivity into the core by inserting a Manual Rx trip using SB HS-1 or SB HS-42 within 5 minutes of the PA02 bus fault.
6 DGBlock_
2 NE01EXC CRS (M)
ATC (M)
BOP (M)
LOOP, 'A' EDG's excitation fails during start, 'B' EDG will not start locally or remote. Transition to ECA-0.0, Loss of All AC Power (Station Blackout).
7 AL /
ALPA02 CRS (C)
BOP (C, MC)
Failure of the Turbine Driven AFP to automatically start during SBO, can be started from CR panel.
CT#2, Establish greater than 270,000 lbm/hr AFW flow rate to the SGs prior to any SG reaching dryout conditions.
8 PA050101
_1 CRS (C)
ATC (C, MC)
BOP (C)
Failure of AEPS EDGS to automatically start when COOP power fails causing PB05 UV, can be started from CR panel. EOP Addendum 39 to restore either NB01 or NB02 using AEPS EDGs.
CT#3, Energize either NB01 or NB02 AC Emergency Bus using AEPS EDGs within 30 minutes of the loss of PA02.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Callaway 2023-1 Scenario #3 Event Description Page 2 of 40 The Plant is at 48% (just below P-9) and stable. Startup is on hold while an issue with the non running MFP is investigated.
After the reactivity brief is complete, the crew will perform OTN-EA-00001 section 5.3 to swap service water pumps from B & C initially in service to the final configuration of A & B in service.
Once the service water pump swap is complete, C SG controlling level channel slowly fails to 25%. The crew will respond using OTO-AE-00002, Steam Generator Water Level Control Instrument Malfunctions, to control SG level. Technical Specification 3.3.1 and 3.3.2 are not met.
Once 'C' SG NR level is restored to the 45-55% band and the MFRV, AE FK-530 is return to automatic control, the PZR variable heaters will fail resulting in a loss of pressure control.
Annunciator 33E will alarm and the crew will restore PZR pressure control by placing the PZR variable heaters in PTL and energizing one set of backup heaters.
Once 1 set of PZR backup heaters is energized, a D control bank control rod, M4, inserts midcore. The crew will enter OTO-SF-00001, Rod Control Malfunctions, and mitigate the event.
Technical Specification 3.1.4 is not met. The crew will adjust turbine load per OTO-MA-00008, Rapid Load Reduction, to lower power to less than 40% to support a normal rod recovery.
After a measurable load reduction, 13.8KV non safety related bus, PA02, will develop a bus fault. This will result in the loss of 2 RCPs but the Reactor not trip despite being above P-7.
The crew will enter E-0, Reactor Trip or Safety Injection, and either manual Rx trip switch (SB HS-1 or SB HS-42) will shutdown the reactor. The crew has a maximum of 5 minutes to Insert negative reactivity and shutdown the reactor beginning at the PA02 fault.
After the reactor is shutdown, a Loss Of Offsite Power (LOOP) occurs. Additionally, the 'A' EDG, NE01, starts but the field will fail to flash, thereby prevent any voltage to develop and supplying NB01. 'B' EDG, NE02, will not start from the control room panel nor locally. This results in a loss of all AC Power and the crew should transition to ECA-0.0 from E-0 Step #3 RNO.
The TDAFP (PAL02) fails to autostart on the NB01 or NB02 UV signal or low SG level. The crew should start the TDAFP per ECA-0.0 Step #3 RNO. After the TDAFP is in service, and attempts to restore the 'A' & 'B' EDGs are unsuccessful, the crew should proceed in ECA-0.0 and restore power to either NB01 or NB02 using the AEPS EDGs per EOP Addendum 39, AEPS. Note: The COOP grid is unstable and normal COOP power is unavailable. The AEPS EDGS did not autostart due to a failure of PA050101 to open on the loss of COOP power but PA050101 can be opened from Control Room HMI PBXY0001 to generate the EDG start signal.
The scenario is complete when the crew has restored power to either NB01 or NB02 with the AEPS EDGs per EOP Addendum 39.
Callaway 2023-1 Scenario #3 Critical Tasks Page 3 of 40 CT-1 CT-2 Critical Tasks Insert negative reactivity into the core by inserting a Manual Rx trip using SB HS-1 or SB-HS-42 within 5 minutes of the PA02 bus fault.
Establish greater than 270,000 lbm/hr AFW flow rate to the SGs prior to any SG reaching dryout conditions (SG WR level less than 10% [25%]).
EVENT #
6 7
Safety significance In the scenario, failure to insert negative reactivity by one of the methods listed previously can result in the needless continuation of an extreme or a severe challenge to the subcriticality CSF. Although the challenge was not initiated by the crew (was not initiated by operator error), continuation of the challenge is a result of the crew's failure to insert negative reactivity and shutdown the reactor when procedurally directed.
Failure to establish minimum AFW flow in this scenario is a violation of the basic objective of ECA-0.0 and of the assumptions of the analyses upon which ECA-0.0 is based. Without AFW flow, the SGs could not support any significant plant cooldown. Thus, the crew would lose the ability to delay the adverse consequences of core uncovery.
Cueing Indication of ATWS. The reactor is not tripped (i.e. rod bottom lights not LIT and neutron flux is NOT lowering) are indication that the reactor is not shutdown and that a manual reactor trip is not effective)
Indication of SBO (neither NB01 or NB02 energized) with no AFW flow indication present Performance indicator Manipulation of controls in the control room as required to initiate the insertion of negative reactivity into the core (at least one of the following)
Manually Rx Trip Switches o
SB HS-1 OR SB HS-42 Manipulation of the:
TDAFP Mechanical Trip/Throttle valve:
o FC HIS-312A Performance feedback Crew will observe the following:
Indication of a negative SUR on the intermediate range of the excore NIS Indication of less than 5% power on the power range of the excore NIS Green Light Lit, Red Light OFF on PG HIS-16 and PG HIS-18 Crew will observe the following:
Greater than 270,000 lbm/hr combined AFW flow to the SGs.
Justification for the chosen performance limit With the PA02 bus fault resulting in the loss of 2 RCPs and no reactor trip, the plant is in an unanalyzed condition. With reactor power greater than the P-7 permissive, the remaining RCPs may not remove reactor heat effectively resulting in rising fuel temperatures which may challenge the fuel clad integrity / fission product barrier. Within 5 minutes is an acceptable amount of time for the crew to shutdown the reactor before any significant degradation of the fission product barrier occurs.
Without AFW flow, decay heat would open the SG safety valves and would rapidly deplete the SG inventory, leading to a loss of secondary heat sink, or SG dryout. Decay heat would then increase RCS temperature and pressure until the pressurizer PORVs open, imposing a larger LOCA than RCP seal leakage.
PWR Owners Group Appendix CT-52, Insert negative reactivity into the core CT - 23, Establish AFW flow during SBO 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.
Callaway 2023-1 Scenario #3 Critical Tasks Page 4 of 40 CT-3 Critical Tasks Energize either NB01 or NB02 AC Emergency Bus using AEPS EDGs within 30 minutes of the loss of PA02 EVENT #
8 Safety significance In the scenario, failure to energize at least one ac emergency bus results in the needless continuation of a situation in which the pumped ECCS capacity and the emergency power capacity are both in a completely degraded status, as are all other active safeguards requiring electrical power. Although the completely degraded status is not due to the crew's action (was not initiated by operator error), continuation in the completely degraded status is a result of the crew's failure to energize at least one ac emergency bus.
Cueing Indication and/or annunciation that all ac emergency buses are de-energized Bus energized lamps extinguished Circuit Breaker Position Bus Voltage EDG status Performance indicator Manipulation of controls as required to energize at least one ac emergency bus from the AEPS:
Using PBXY0001, Open AEPS FDR BKR TO PA501 VIA Central Electric Reform Feeder Breaker To XFMR o
PA50101 Then perform one of the following to restore power to NB01 or NB02:
NB01:
Using PBXY0001 CLOSE AEPS FDR BKR PB0503 TO NB0114 PB0503 CLOSE NB01 AEPS SPLY BKR NB0114 NB HIS-67 NB02 Using PBXY0001 CLOSE AEPS FDR BKR PB0502 TO NB0214 PB0502 Close NB02 AEPS Supply BKR NB0214 NB HIS-68 Performance feedback Indication that NB01 or NB02 is energized:
NB01 or NB02 Bus energized light NB01 or NB02 bus voltage Justification for the chosen performance limit Failure to perform the critical task within 30 minutes results in needless degradation of RCS barrier (and to fission product release, specifically of the RCS barrier at the point of the RCP seals). Failure to perform the critical task means that RCS inventory lost through the RCP seals cannot be replaced. It also means that the RCP seals remain without cooling and gradually deteriorate. As the seals deteriorate the rate of RCS inventory loss increases.
PWR Owners Group Appendix CT - 24, Energize at least one ac emergency bus 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.
Callaway 2023-1 Scenario #3 References Page 5 of 40 References OTN-EA-00001, Service Water System, rev 41 OTO-AE-00002, Steam Generator Water Level Control Instrument Malfunctions, rev 14 OTA-RK-00018, Addendum 33E, PZR Heater Group Lockout, rev 0 OTA-RK-00016, Addendum 17D, PG Transformer Undervoltage, rev 0 OTA-RK-00022, Addendum 86B, Low Flow and P7 Reactor Trip, rev 0 OTO-SF-00001, Rod Control Malfunctions, rev 23 OTO-MA-00008, Rapid Load Reduction, rev 42 E-0, Reactor Trip or Safety Injection, rev 28 ECA-0.0, Loss of All AC Power, rev 33 EOP Addendum 39, Alternate Emergency Power Supply, rev 15 Technical Specification 3.3.1, Reactor Trip System Instrumentation Technical Specification 3.3.2, ESFAS Instrumentation Technical Specification 3.1.4, Rod Group Alignment Limits ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions (from PRA update 9.01):
A Station Blackout represents a 3% contribution to CDF by Accident type.
13.8KVAC (PA system) is the highest risk important System for Callaway.
Target Quantitative Attributes per Scenario/Scenario Set (NUREG 1021, Table 3.4-1)
Quantitative Attribute Target per Scenario Actual Attributes Events after EOP entry 1-2 2
Abnormal events 2-4 2
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 1
Preidentified CTs 2 or more 3
The following events qualify as manual control of an automatic function per ES-3.4.2(4):
Event #2, 'C' SG level instrument fails which requires manual control of the 'C' Main Fed Reg Valve to stabilize 'C' SG level. (BOP)
Event #5, The reactor fails to automatically trip and manually tripping the reactor using SB HIS 1 (ATC) or 42 (BOP) is successful. As SB HIS 1 is likely the first switch manipulated, the ATC would get credit for manual control of a failed automatic action.
Event #7, The TDAFP fails to automatically start, can be manually started from CR panel (BOP)
Event #8, The AEPS EDGs fail to autostart on PB05 UV, can be started from CR panel (ATC)
Callaway 2023-1 Scenario #3 Page 6 of 40 Scenario #3 consists of an ATWS that either manual Rx trip switch (SB HS-1 or SB HS-42) will shutdown the reactor. This is different from the ATWS scenarios in the previous 2 exams as there was no ATWS in 2020 and 2022's ATWS had different success methods (supply breakers or infield action per FR-S.1). After the ATWS is mitigated, a loss of all AC Power occurs. The scenario is complete when the crew has restored power to either NB01 or NB02 with the AEPS EDGs. In the previous 2 exams, power was either restored with COOP power or offsite power, making the response and success path different for this scenario.
Callaway 2023-1 Scenario#3 Simulator Lesson Plan Page 7 of 40 Scenario Setup Guide:
- IC-49 (password 049)
=======SCENARIO PRELOADS / SETUP ITEMS==========
- Pre-Loads:
o JRM11EXAMSAFE = TRUE(1) o Ensure the NCP is tagged OOS BG47PB0301_BKRTA_BKPOS = 3 (rackout) switch in PTL w/tag o Ensure the 'B' CCP is in service along with the B CCW train.
o SF006 = 0 (Failure of the RX to Auto Trip) o PA050101_1= TRUE(1) (Failure of PA050101 to open automatically on Undervoltage (UV))
o PAL02_3 = 1 (Failure of TDAFP to auto start) o DGBlock_2 = 4 (Failure of 'B' EDG to start in auto or manual) o NE01EXC = TRUE(1) ('A' EDG excitation fails)
For specific event malfunctions, see the individual event page and/or Attachment 3, Scenario File Manager