ML20276A152

1-CW-2020-08 Draft Outlines

ML20276A152
Person / Time
Site:	Callaway
Issue date:	09/09/2020
From:	Greg Werner Operations Branch IV
To:	Ameren Missouri
References
Download: ML20276A152 (69)

Text

ES-401 PWR Examination Outline - RO Form ES-401-2 Facility: Callaway Plant Date of Exam: August 31, 2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 4 3 4 2 3 2 18 6 Emergency and Abnormal Plant 2 2 2 1 N/A 1 1 N/A 2 9 4 Evolutions Tier Totals 6 5 5 3 4 4 27 10 1 3 1 4 3 1 1 3 3 3 3 3 28 5

2. 2 1 0 2 1 2 1 1 0 0 0 Plant 2 10 3 Systems Tier Totals 5 2 4 5 2 3 4 4 3 3 3 38 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 3 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

• These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

• • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

EA2.02 Ability to determine or interpret the following 000007 (EPE 7; BW E02&E10; CE E02) X as they apply to a reactor trip: 4.3 1 Reactor Trip, Stabilization, Recovery / 1 Proper actions to be taken if the automatic safety functions have not taken place (CFR 43.5 / 45.13)

AK3.05 Knowledge of the reasons for the following 000008 (APE 8) Pressurizer Vapor Space X responses as they apply to the Pressurizer Vapor 4.0 2 Accident / 3 Space Accident:

ECCS termination or throttling criteria.

(CFR 41.5,41.10 / 45.6 / 45.13) 000009 (EPE 9) Small Break LOCA / 3 EA1.05 Ability to operate and monitor the following 000011 (EPE 11) Large Break LOCA / 3 X as they apply to a Large Break LOCA: 4.3 3 Manual and/or automatic transfer of suction of charging pumps to borated source (CFR 41.7 / 45.5 / 45.6).

AK1.02 Knowledge of the operational implications 000015 (APE 15) Reactor Coolant Pump X of the following concepts as they apply to Reactor 3.7 4 Malfunctions / 4 Coolant Pump Malfunctions (Loss of RC Flow):

Consequences of an RCPS failure (CFR 41.8 / 41.10 / 45.3)

AK3.02 Knowledge of the reasons for the following 000022 (APE 22) Loss of Reactor Coolant X responses as they apply to the Loss of Reactor 3.5 5 Makeup / 2 Coolant Makeup:

Actions contained in SOPs and EOPs for RCPs, loss of makeup, loss of charging, and abnormal charging (CFR 41.5, 41.10 / 45.6 / 45.13)

AK2.01 Knowledge of the interrelations between the 000025 (APE 25) Loss of Residual Heat X Loss of Residual Heat Removal System and the 2.9 6 Removal System / 4 following:

RHR heat exchangers.

(CFR 41.7 / 45.7) 000026 (APE 26) Loss of Component X 2.1.23 Ability to perform specific system and 4.3 7 Cooling Water / 8 integrated plant procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6)

AK3.03 Knowledge of the reasons for the following 000027 (APE 27) Pressurizer Pressure X responses as they apply to the Pressurizer 3.7 8 Control System Malfunction / 3 Pressure Control Malfunctions:

Actions contained in EOP for PZR PCS malfunctions (CFR 41.5,41.10 / 45.6 / 45.13)

EA1.14 Ability to operate and monitor the following 000029 (EPE 29) Anticipated Transient X as they apply to an ATWS: 4.2 9 Without Scram / 1 Driving of control rods into the core (CFR 41.7 / 45.5 / 45.6)

ES-401 3 Form ES-401-2 EA2.16 Ability to determine or interpret the following 000038 (EPE 38) Steam Generator Tube X as they apply to a SGTR: 4.2 10 Rupture / 3 Actions to be taken if S/G goes solid and water enters steam line.

(CFR 43.5 / 45.13)

AK1.07 Knowledge of the operational implications 000040 (APE 40; BW E05; CE E05; W E12) X of the following concepts as they apply to Steam 3.4 11 Steam Line RuptureExcessive Heat Line Rupture:

Transfer / 4 Effects of feedwater introduction on dry S/G (CFR 41.8 / 41.10 / 45.3) 000054 (APE 54; CE E06) Loss of Main X 2.4.4 Ability to recognize abnormal indications for 4.5 12 Feedwater /4 system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

(CFR: 41.10 / 43.2 / 45.6)

EK1.01 Knowledge of the operational implications 000055 (EPE 55) Station Blackout / 6 X of the following concepts as they apply to the 3.3 13 Station Blackout:

Effect of battery discharge rates on capacity (CFR 41.8 / 41.10 / 45.3) 000056 (APE 56) Loss of Offsite Power / 6 AA2.19 Ability to determine and interpret the 000057 (APE 57) Loss of Vital AC X following as they apply to the Loss of Vital AC 4.0 14 Instrument Bus / 6 Instrument Bus:

The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus (CFR: 43.5 / 45.13) 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 AK3.03 Knowledge of the reasons for the following 000065 (APE 65) Loss of Instrument Air / 8 X responses as they apply to the Loss of Instrument 2.9 15 Air:

Knowing effects on plant operation of isolating certain equipment from instrument air (CFR 41.5,41.10 / 45.6 / 45.13) 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 EK2.1 Knowledge of the interrelations between the (W E04) LOCA Outside Containment / 3 X (LOCA Outside Containment) and the following: 3.5 16 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(CFR: 41.7 / 45.7)

ES-401 4 Form ES-401-2 EK2.2 Knowledge of the interrelations between the (W E11) Loss of Emergency Coolant X (Loss of Emergency Coolant Recirculation) and the 3.9 17 Recirculation / 4 following:

Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

(CFR: 41.7 / 45.7)

EK1.3 Knowledge of the operational implications of (BW E04; W E05) Inadequate Heat X the following concepts as they apply to the (Loss of 3.9 18 TransferLoss of Secondary Heat Sink / 4 Secondary Heat Sink):

Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Secondary Heat Sink).

(CFR: 41.8 / 41.10, 45.3)

K/A Category Totals: 4 3 4 2 3 2 Group Point Total: 18

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

000001 (APE 1) Continuous Rod Withdrawal / 1 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 X AK3.03 Knowledge of the reasons 3.5 19 Malfunction / 2 for the following responses as they apply to the Pressurizer Level Control Malfunctions:

False indication of PZR level when PROV or spray valve is open and RCS saturated (CFR 41.5,41.10 / 45.6 / 45.13) 000032 (APE 32) Loss of Source Range Nuclear X AK1.01 Knowledge of the 2.5 20 Instrumentation / 7 operational implications of the following concepts as they apply to Loss of Source Range Nuclear Instrumentation:

Effects of voltage changes on performance (CFR 41.8 / 41.10 / 45.3) 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 X AA2.02 Ability to determine and 3.9 21 interpret the following as they apply to the Loss of Condenser Vacuum:

Conditions requiring reactor and/or turbine trip (CFR: 43.5 / 45.13) 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms X 2.1.30 Ability to locate and 4.4 22

/7 operate components, including local controls.

(CFR: 41.7 / 45.7) 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 AK2.01 Knowledge of the 2.6 23 interrelations between the High Reactor Coolant Activity and the following:

Process radiation monitors (CFR 41.7 / 45.7) 000078 (APE 78*) RCS Leak / 3 NA

ES-401 6 Form ES-401-2 (W E01 & E02) Rediagnosis & SI Termination / 3 X EK1.2 Knowledge of the 3.4 24 operational implications of the following concepts as they apply to the (SI Termination):

Normal, abnormal and emergency operating procedures associated with (SI Termination).

(CFR: 41.8 / 41.10, 45.3)

(W E13) Steam Generator Overpressure / 4 X 2.2.44 Ability to interpret control 4.2 25 room indication to verify the status and operations of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12)

(W E15) Containment Flooding / 5 X EA1.3 Ability to operate and / or 2.8 26 monitor the following as they apply to the (Containment Flooding):

Desired operating results during abnormal and emergency situations.

(CFR: 41.7 / 45.5 / 45.6)

(W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 NA (BW A02 & A03) Loss of NNI-X/Y/7 NA (BW A04) Turbine Trip / 4 NA (BW A05) Emergency Diesel Actuation / 6 NA (BW A07) Flooding / 8 NA (BW E03) Inadequate Subcooling Margin / 4 NA (BW E08; W E03) LOCA CooldownDepressurization / 4 X EK2.2 Knowledge of the 3.7 27 interrelations between the (LOCA Cooldown and Depressurization) and the following:

Facility*s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

(CFR: 41.7 / 45.7)

(BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures NA (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 NA (CE E09) Functional Recovery NA (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 NA K/A Category Point Totals: 2 2 1 1 1 2 Group Point Total: 9

ES-401 7 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO)

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

3 K1.03 Knowledge of the physical connections 003 (SF4P RCP) Reactor Coolant X and/or cause-effect relationships between the 3.3 28 Pump RCPS and the following systems:

ROX2 RCP seal system (CFR: 41.2 to 41.9 / 45.7 to 45.8)

A3.05 Ability to monitor automatic operation of 003 (SF4P RCP) Reactor Coolant X the RCPS, including: 2.7 29 Pump RCP lube oil and bearing lift pumps ROX2 (CFR: 41.7 / 45.5)

K1.19 Knowledge of the physical connections 004 (SF1; SF2 CVCS) Chemical and X and/or cause-effect relationships between the 2.7 30 Volume Control CVCS and the following systems:

Primary grade water supply (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K3.07 Knowledge of the effect that a loss or 005 (SF4P RHR) Residual Heat X malfunction of the RHRS will have on the 3.2 31 Removal following:

ROX2 Refueling operations (CFR: 41.7 / 45.6)

A4.02 Ability to manually operate and/or 005 (SF4P RHR) Residual Heat X monitor in the control room: 3.4 32 Removal Heat exchanger bypass flow control ROX2 (CFR: 41.7 / 45.5 to 45.8)

K4.11 Knowledge of ECCS design feature(s) 006 (SF2; SF3 ECCS) Emergency X and/or interlock(s) which provide for the 3.9 33 Core Cooling following:

Reset of SIS (CFR: 41.7)

G2.1.23 Ability to perform specific system and 007 (SF5 PRTS) Pressurizer X integrated plant procedures during all modes 4.3 34 Relief/Quench Tank of plant operation (CFR: 41.10 / 43.5 / 45.2 / 45.6)

A2.01 Ability to (a) predict the impacts of the 008 (SF8 CCW) Component Cooling X following malfunctions or operations on the 3.3 35 Water CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate ROX2 the consequences of those malfunctions or operations:

Loss of CCW pump (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A1.02 Ability to predict and/or monitor changes 008 (SF8 CCW) Component Cooling X in parameters (to prevent exceeding design 2.9 36 Water limits) associated with operating the CCWS controls including:

ROX2 CCW temperature (CFR: 41.5 / 45.5)

ES-401 8 Form ES-401-2 010 (SF3 PZR PCS) Pressurizer X 2.1.32 Ability to explain and apply system 3.8 37 Pressure Control limits and precautions.

(CFR: 41.10 / 43.2 / 45.12)

K4.02 Knowledge of RPS design feature(s) 012 (SF7 RPS) Reactor Protection X and/or interlock(s) which provide for the 3.9 38 following:

ROX2 Automatic reactor trip when RPS Setpoints are exceeded for each RPS function; basis for each (CFR: 41.7)

A2.07 Ability to (a) predict the impacts of the 012 (SF7 RPS) Reactor Protection X following malfunctions or operations on the 3.2 39 RPS; and (b) based on those predictions, use ROX2 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of dc control.

(CFR: 41.5 / 43.5 / 45.3 / 45.5)

K2.01 Knowledge of bus power supplies to the 013 (SF2 ESFAS) Engineered X following: 3.6 40 Safety Features Actuation ESFAS/safeguards equipment control (CFR: 41.7)

K3.01 Knowledge of the effect that a loss or 022 (SF5 CCS) Containment Cooling X malfunction of the CCS will have on the 2.9 41 following:

Containment equipment subject to damage by high or low temperature, humidity, and pressure (CFR: 41.7 / 45.6) 025 (SF5 ICE) Ice Condenser NA 026 (SF5 CSS) Containment Spray X A1.01 Ability to predict and/or monitor changes 3.9 42 in parameters (to prevent exceeding design ROX2 limits) associated with operating the CSS controls including:

Containment Pressure (CFR: 41.5 / 45.5) 026 (SF5 CSS) Containment Spray X K3.01 Knowledge of the effect that a loss or 3.9 43 malfunction of the CSS will have on the ROX2 following:

CCS (CFR: 41.7 / 45.6) 039 (SF4S MSS) Main and Reheat X A1.03 Ability to predict and/or monitor changes 2.6 44 Steam in parameters (to prevent exceeding design limits) associated with operating the MRSS ROX2 controls including:

Primary system temperature indications, and required values, during main steam system warm-up (CFR: 41.5 / 45.5)

ES-401 9 Form ES-401-2 039 (SF4S MSS) Main and Reheat X K5.01 Knowledge of the operational 2.9 45 Steam implications of the following concepts as the apply to the MRSS:

ROX2 Definition and causes of steam/water hammer (CFR: 41.5 / 45.7) 059 (SF4S MFW) Main Feedwater X A3.06 Ability to monitor automatic operation of 3.2 46 the MFW, including:

Feedwater Isolation (CFR: 41.7 / 45.5) 061 (SF4S AFW) X A3.01 Ability to monitor automatic operation of 4.2 47 Auxiliary/Emergency Feedwater the AFW, including:

AFW startup and flows (CFR: 41.7 / 45.5) 062 (SF6 ED AC) AC Electrical X A4.02 Ability to manually operate and/or 2.5 48 Distribution monitor in the control room:

Remote racking in and out of breakers (CFR: 41.7 / 45.5 / to 45.8) 063 (SF6 ED DC) DC Electrical X 2.4.46 Ability to verify that the alarms are 4.2 49 Distribution consistent with the plant conditions.

(CFR: 41.10 / 43.5 / 45.3 / 45.12) 064 (SF6 EDG) Emergency Diesel X K6.07 Knowledge of the effect of a loss or 2.7 50 Generator malfunction of the following will have on the ED/G system:

Air receivers (CFR: 41.7 / 45.7) 073 (SF7 PRM) Process Radiation X K4.01 Knowledge of PRM system design 4.0 51 Monitoring feature(s) and/or interlock(s) which provide for the following:

Release termination when radiation exceeds setpoint (CFR: 41.7) 076 (SF4S SW) Service Water X A2.01 Ability to (a) predict the impacts of the 3.5 52 following malfunctions or operations on the ROX2 SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of SWS (CFR: 41.5 / 43.5 / 45.3 / 45.13) 076 (SF4S SW) Service Water X K1.21 Knowledge of the physical connections 2.7 53 and/or cause- effect relationships between the ROX2 SWS and the following systems:

Auxiliary backup SWS (CFR: 41.2 to 41.9 / 45.7 to 45.8)

ES-401 10 Form ES-401-2 078 (SF8 IAS) Instrument Air X K3.02 Knowledge of the effect that a loss or 3.4 54 malfunction of the IAS will have on the following:

Systems having pneumatic valves and controls (CFR: 41.7 / 45.6) 103 (SF5 CNT) Containment X A4.04 Ability to manually operate and/or 3.5 55 monitor in the control room:

Phase A and phase B resets (CFR: 41.7 / 45.5 to 45.8) 053 (SF1; SF4P ICS*) Integrated NA Control K/A Category Point Totals: 3 1 4 3 1 1 3 3 3 3 3 Group Point Total: 28

ES-401 11 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO)

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

001 (SF1 CRDS) Control Rod Drive X K6.11 Knowledge of the effect of a loss or 2.9 56 malfunction on the following CRDS components:

Location and operation of CRDS fault detection (trouble alarms) and reset system, including rod control annunciator (CFR: 41.7/45.7) 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position X K4.06 Knowledge of RPIS design feature(s) 3.4 57 Indication and/or interlock(s) which provide for the following:

Individual and group misalignment (CFR: 41.5 / 45.7) 015 (SF7 NI) Nuclear X K2.01 Knowledge of bus power supplies to the 3.3 58 Instrumentation following:

NIS channels, components, and interconnections (CFR: 41.7) 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature X K6.01 Knowledge of the effect of a loss or 2.7 59 Monitor malfunction of the following ITM system components:

Sensors and detectors (CFR: 41.7 / 45.7) 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 X A1.01 Ability to predict and/or monitor changes 2.7 60 Cooling in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including:

Spent fuel pool water level (CFR: 41.5 / 45.5) 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator X K5.03 Knowledge of operational implications of 2.8 61 the following concepts as the apply to the S/GS:

Shrink and swell concept (CFR: 41.5 / 45.7) 041 (SF4S SDS) Steam X K4.18 Knowledge of SDS design feature(s) 3.4 62 Dump/Turbine Bypass Control and/or interlock(s) which provide for the following:

Turbine trip (CFR: 41.7) 045 (SF 4S MTG) Main Turbine Generator

ES-401 12 Form ES-401-2 055 (SF4S CARS) Condenser Air X K1.06 Knowledge of the physical connections 2.6 63 Removal and/or cause-effect relationships between the CARS and the following systems:

PRM system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste X K1.07 Knowledge of the physical connections 2.7 64 and/or cause effect relationships between the Liquid Radwaste System and the following systems:

Sources of liquid wastes for LRS (CFR: 41.2 to 41.9 / 45.7 to 45.8) 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water X A2.01 Ability to (a) predict the impacts of the 3.0 65 following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of intake structure (CFR: 41.5 / 43.5 / 45.3 / 45.13) 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room NA Ventilation K/A Category Point Totals: 2 1 0 2 1 2 1 1 0 0 0 Group Point Total: 10

ES-401 Generic Knowledge and Abilities Outline (Tier 3) RO Form ES-401-3 Facility: Date of Exam:

Category K/A # Topic RO SRO-only IR # IR #

2.1.26 Knowledge of industrial safety procedures (such as 3.4 66 rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

(CFR: 41.10 / 45.12) 2.1.37 Knowledge of procedures, guidelines, or limitations 4.3 67 associated with reactivity management.

(CFR: 41.1 / 43.6 / 45.6)

1. Conduct of Operations 2.1.44 Knowledge of RO duties in the control room during fuel 3.9 68 handling, such as l responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

(CFR: 41.10 / 43.7 / 45.12)

Subtotal 3 2.2.14 Knowledge of the process for controlling equipment 3.9 69 configuration or status.

(CFR: 41.10 / 43.3 / 45.13) 2.2.41 Ability to obtain and interpret station electrical and 3.5 70

2. Equipment mechanical drawings.

Control (CFR: 41.10 / 45.12 / 45.13) 2.2.43 Knowledge of the process used to track inoperable 3.0 71 alarms.

(CFR: 41.10 / 43.5 / 45.13)

Subtotal 3 2.3.4 Knowledge of radiation exposure limits under normal or 3.2 72 emergency conditions.

(CFR: 41.12 / 43.4 / 45.10)

3. Radiation 2.3.15 Knowledge of radiation monitoring systems, such as 2.9 73 Control fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.12 / 43.4 / 45.9)

Subtotal 2 2.4.3 Ability to identify post-accident instrumentation. 3.7 74 (CFR: 41.6 / 45.4)

4. Emergency Procedures/Plan 2.4.17 Knowledge of EOP terms and definitions. 3.9 75 (CFR: 41.10 / 45.13)

Subtotal 2 Tier 3 Point Total 10

ES-401 PWR Examination Outline - SRO Form ES-401-2 Facility: Callaway Plant Date of Exam: August 31, 2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 18 3 3 6 Emergency and Abnormal Plant 2 N/A N/A 9 2 2 4 Evolutions Tier Totals 27 5 5 10 1 28 3 2 5 2.

Plant 2 10 2 1 3 Systems Tier Totals 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

• These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

• • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 15 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

000007 (EPE 7; BW E02&E10; CE E02)

Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 X 2.4.30 Knowledge of events related to system 4.1 76 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) 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 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 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 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 X 2.2.25 Knowledge of the bases in Technical 4.2 77 Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 X AA2.03 Ability to determine and interpret the 3.9 78 following as they apply to the Loss of DC Power:

DC loads lost; impact on ability to operate and monitor plant systems (CFR: 43.5 / 45.13) 000062 (APE 62) Loss of Nuclear Service X 2.1.32 Ability to explain and apply system limits and 4.0 80 Water / 4 precautions.

(CFR: 41.10 / 43.2 / 45.12) 000065 (APE 65) Loss of Instrument Air / 8

ES-401 16 Form ES-401-2 000077 (APE 77) Generator Voltage and X AA2.05 Ability to determine and interpret the 3.8 79 Electric Grid Disturbances / 6 following as they apply to Generator Voltage and Electric Grid Disturbances:

Operational status of offsite circuit.

(CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)

(W E04) LOCA Outside Containment / 3 (W E11) Loss of Emergency Coolant X EA2.1 Ability to determine and interpret the 4.2 81 Recirculation / 4 following as they apply to the (Loss of Emergency Coolant Recirculation):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(CFR: 43.5 / 45.13)

(BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals: 3 3 Group Point Total: 6

ES-401 17 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 X 2.2.40 Ability to apply Technical 4.7 82 Specifications for a system.

(CFR: 41.10/43.2/43.5/45.3) 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 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X 2.2.25 Knowledge of the bases in 4.2 83 Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) 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 AA2.13 Ability to determine and 4.4 84 interpret the following as they apply to the Plant Fire on Site:

Need for emergency plant shutdown (CFR: 43.5 / 45.13) 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 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 NA (BW A02 & A03) Loss of NNI-X/Y/7 NA (BW A04) Turbine Trip / 4 NA (BW A05) Emergency Diesel Actuation / 6 NA (BW A07) Flooding / 8 NA (BW E03) Inadequate Subcooling Margin / 4 NA (BW E08; W E03) LOCA CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures NA

ES-401 18 Form ES-401-2 (CE A11**; W E08) RCS OvercoolingPressurized Thermal X EA2.2 Ability to determine and 4.1 85 Shock / 4 interpret the following as they apply to the (Pressurized Thermal Shock):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments (CFR: 43.5 / 45.13)

(CE A16) Excess RCS Leakage / 2 NA (CE E09) Functional Recovery NA (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 NA K/A Category Point Totals: 2 2 Group Point Total: 4

ES-401 19 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (SRO)

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

003 (SF4P RCP) Reactor Coolant ROX2 Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control 005 (SF4P RHR) Residual Heat ROX2 Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling ROX2 Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection ROX2 013 (SF2 ESFAS) Engineered X A2.04 Ability to (a) predict the impacts of the 4.2 86 Safety Features Actuation following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Loss of instrument bus (CFR: 41.5 / 43.5 / 45.3 / 45.13) 022 (SF5 CCS) Containment Cooling X 2.4.21 Knowledge of the parameters and logic 4.6 87 used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12) 025 (SF5 ICE) Ice Condenser NA 026 (SF5 CSS) Containment Spray ROX2 039 (SF4S MSS) Main and Reheat ROX2 Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)

Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical X 2.2.37 Ability to determine operability and/or 4.6 88 Distribution availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12) 063 (SF6 ED DC) DC Electrical Distribution

ES-401 20 Form ES-401-2 064 (SF6 EDG) Emergency Diesel X A2.06 Ability to (a) predict the impacts of the 3.3 89 Generator following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Operating unloaded, lightly loaded, and highly loaded time limit.

(CFR: 41.5 / 43.5 / 45.3 / 45.13) 073 (SF7 PRM) Process Radiation X A2.01 Ability to (a) predict the impacts of the 2.9 90 Monitoring following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Erratic or failed power supply (CFR: 41.5 / 43.5 / 45.3 / 45.13) 076 (SF4S SW) Service Water ROX2 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated NA Control K/A Category Point Totals: 3 2 Group Point Total: 5

ES-401 21 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (SRO)

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

001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling X A2.03 Ability to (a) predict the impacts of the 4.0 91 Equipment following malfunctions or operations on the Fuel Handling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Mispositioned fuel element (CFR: 41.5 / 43.5 / 45.3 / 45.13) 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation X A2.02 Ability to (a) predict the impacts of the 2.9 92 Monitoring following malfunctions or operations on the ARM system- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Detector failure (CFR: 41.5 / 43.5 / 43.3 / 45.13) 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection X 2.4.41 Knowledge of the emergency action 4.6 93 level thresholds and classifications.

(CFR: 41.10 / 43.5 / 45.11)

ES-401 22 Form ES-401-2 050 (SF 9 CRV*) Control Room NA Ventilation K/A Category Point Totals: 2 1 Group Point Total: 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) SRO Form ES-401-3 Facility: Callaway Plant Date of Exam: August 31, 2020 Category K/A # Topic RO SRO-only IR # IR #

2.1.34 Knowledge of primary and secondary plant chemistry 3.5 94 limits.

1. Conduct of (CFR: 41.10 / 43.5 / 45.12)

Operations 2.1.41 Knowledge of the refueling process. 3.7 95 (CFR: 41.2 / 41.10 / 43.6 / 45.13)

Subtotal 2 2.2.5 Knowledge of the process for making design or 3.2 96 operating changes to the facility.

(CFR: 41.10 / 43.3 / 45.13)

2. Equipment 2.2.17 Knowledge of the process for managing maintenance 3.8 97 Control activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

(CFR: 41.10 / 43.5 / 45.13)

Subtotal 2 2.3.6 Ability to approve release permits. 3.8 98

3. Radiation Control (CFR: 41.13 / 43.4 / 45.10)

Subtotal 1 2.4.29 Knowledge of the emergency plan. 4.4 99 (CFR: 41.10 / 43.5 / 45.11)

4. Emergency 2.4.23 Knowledge of the bases for prioritizing emergency 4.4 100 Procedures/Plan procedure implementation l during emergency operations.

(CFR: 41.10 / 43.5 / 45.13)

Subtotal 2 Tier 3 Point Total 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Selected Reason for Rejection Group K/A 1/1 K/A 027 AK3.01 Q#8 00027 AK3.04 - Pressurizer Pressure Control System AK3.03 - Pressurizer Malfunction: Knowledge of the reasons for the following Pressure Control responses as they apply to the Pressurizer Pressure System Malfunction: Control Malfunctions: Why, if PZR level is lost and then Actions contained in restored, that pressure recovers much more slowly.

EOP for PZR PCS Randomly replaced due to the inability to write a question malfunctions.

with plausible distractors.

AK3.01 Isolation of PZR spray following loss of PZR heaters was reselected as the plant design did not contain the function / isolation supported by the K/A.

1/1 K/A 040 AK1.07 - Q#11 00040 AK1.04 - Steam Line Rupture: Knowledge Steam Line Rupture: of the operational implications of the following concepts Knowledge of the as they apply to Steam Line Rupture: Nil ductility operational temperature.

implications of the Randomly replaced due to the inability to write a question following concepts as with plausible distractors.

they apply to Steam Line Rupture: Effects of feedwater introduction on dry S/G.

K/A 028 AK3.03 1/2 Q#19 00028 AK3.04 - Pressurizer (PZR) Level Control Pressurizer (PZR)

Malfunction: Knowledge of the reasons for the following Level Control responses as they apply to the Pressurizer Level Control Malfunction:

Malfunctions: Change in PZR level with power change, Knowledge of the even though RCS T-ave. constant, due to loop size reasons for the difference.

following responses as they apply to the Plant design does not support this K/A as the loop sizes Pressurizer Level are effectively the same (27.5" ID vs 29").

Control Malfunctions:

False indication of PZR level when PORV or spray valve is open and RCS saturated.

ES-401 Record of Rejected K/As Form ES-401-4 1/2 K/A 061 G2.1.19 Q#22 00061 G2.1.31 - Area Radiation Monitoring G2.1.30 Area (ARM) System Alarms: Ability to use plant computers to Radiation Monitoring evaluate system or component status. Ability to locate (ARM) System control room switches, controls, and indications, and to Alarms: Ability to determine that they correctly reflect the desired plant locate and operate lineup.

components, including While there are some indications on the backpanel that local controls.

may be used during source checks and alarms, plant lineups of ARMs are performed by the rad protection department. Randomly replaced due to the inability to write a question with plausible distractors applicable to control room operators.

Reselected G2.1.19 (Ability to use plant computers to evaluate system or component status.) due to inability to write question with plausible distractors.

K/A WE E13 G2.2.44 1/2 Q#25 WE E13 G2.2.12 - Steam Generator Overpressure:

Steam Generator Knowledge of surveillance procedures. Steam Generator Overpressure: Ability overpressure is a functional restoration yellow path to interpret control procedure that does not have surveillance procedures room indications to associated with it.

verify the status and operation of a system, Randomly replaced K/A.

and understand how operator actions and directives affect plant and system conditions.

K/A 007 G2.1.23:

2/1 Q#34 007 K5.02 - Pressurizer Relief Tank/Quench Tank Pressurizer Relief System: Knowledge of the operational implications of the Tank/Quench Tank following concepts as the apply to PRTS: Method of System: Ability to forming a steam bubble in the PZR.

perform specific system and Unable to write a question with plausible distractors for integrated plant the associated system as it does not apply to the K/A procedures during all topic. The method of drawing a steam bubble in the PZR modes of plant is independent of the PRT operation.

operation.

ES-401 Record of Rejected K/As Form ES-401-4 2/1 K/A 012 A2.07 - Q#39 012 A2.01 - Reactor Protection System: Ability to Reactor Protection (a) predict the impacts of the following malfunctions or System: Ability to (a) operations on the RPS; and (b) based on those predict the impacts of predictions, use procedures to correct, control, or mitigate the following the consequences of those malfunctions or operations:

malfunctions or Faulty bistable operation.

operations on the RPS; Randomly replaced due to the inability to write a question and (b) based on those with plausible distractors.

predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of dc control.

2/1 K/A 026 A1.01 - Q#42 026 A1.06 - Containment Spray System (CSS):

Containment Spray Ability to predict and/or monitor changes in parameters System (CSS): Ability (to prevent exceeding design limits) associated with to predict and/or operating the CSS controls including: Containment spray monitor changes in pump cooling.

parameters (to prevent Randomly replaced as plant design did not support this exceeding design K/A. Callaway's CTMT spray pumps are air cooled.

limits) associated with operating the CSS controls including:

Containment Pressure.

2/1 K/A 026 K3.01 - Q#43 026 K3.02 - Containment Spray System (CSS):

Containment Spray Knowledge of the effect that a loss or malfunction of the System (CSS): CSS will have on the following: Recirculation spray Knowledge of the system.

effect that a loss or Randomly replaced due to the inability to write a question malfunction of the with plausible distractors.

CSS will have on the following: CCS 2/1 K/A 059 A3.06 - Main Q#46 059 A3.02 - Main Feedwater (MFW) System:

Feedwater (MFW) Ability to monitor automatic operation of the MFW, System: Ability to including: Programmed levels of the S/G.

monitor automatic Randomly replaced due to the inability to write a question operation of the MFW, with plausible distractors as programmed SG level is including: Feedwater independent of power level i.e. SG level is not a function isolation.

of reactor power.

ES-401 Record of Rejected K/As Form ES-401-4 2/1 K/A 061 A3.01 - Q#47 061 A3.04 - Auxiliary / Emergency Feedwater Auxiliary / Emergency (AFW) System: Ability to monitor automatic operation of Feedwater (AFW) the AFW, including: Automatic AFW isolation.

System: Ability to Randomly replaced as plant design did not support this monitor automatic K/A. Callaway's AFW system does not have an operation of the AFW, automatic isolation feature.

including: AFW startup and flows.

2/1 K/A 103 A4.04 - Q#55 103 A4.09 - Containment System: Ability to Containment System: manually operate and/or monitor in the control room:

Ability to manually Containment vacuum system.

operate and/or monitor Randomly replaced due to the inability to write a question in the control room:

to this K/A in this system. Containment Purge would be Phase A and phase B the only applicable ventilation system which is covered resets.

under its own system topic.

2/2 K/A 014 K4.06 - Rod Q#57 014 K4.05 - Rod Position Indication System Position Indication (RPIS): Knowledge of RPIS design feature(s) and/or System (RPIS): interlock(s) which provide for the following: Zone Knowledge of RPIS reference lights.

design feature(s)

Plant design does not support this K/A as there are no and/or interlock(s) specific rod hold interlocks, there is rod overlap and rod which provide for the movement rates and insertions due to runback etc but no following: Individual hold interlocks.

and group misalignment.

SRO 1 / 2 K/A 059 G2.4.41 Q#83 - 059 G2.4.35 Accidental Liquid Radwaste G2.2.25 - Accidental Release: Knowledge of local auxiliary operator tasks Liquid Radwaste during an emergency and the resultant operational effects.

Release: Knowledge Randomly replaced due to the inability to write a SRO of the bases in Level question with plausible distractors to this topic as Technical there are few, if any, specific aux operator tasks for this Specifications for event. A spill response team (which may or may not have limiting conditions for Aux operators on it) would address the spill.

operations and safety limits. Reselected G2.4.41 as there was already one EAL question in order to avoid oversampling EALs for SROs.

ES-401 Record of Rejected K/As Form ES-401-4 SRO 2 / 1 K/A 073 A2.01 - Q#90 073 A2.03 - Process Radiation Monitoring (PRM)

Process Radiation System: Ability to (a) predict the impacts of the following Monitoring (PRM) malfunctions or operations on the PRM system; and (b)

System: Ability to (a) based on those predictions, use procedures to correct, predict the impacts of control, or mitigate the consequences of those the following malfunctions or operations: Calibration drift.

malfunctions or Randomly replaced due to the inability to write a SRO operations on the PRM Level question with plausible distractors. If calibration system; and (b) based drift is suspected the procedure selection to mitigate on those predictions, would be made by a different department after being use procedures to notified by the shift.

correct, control, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Callaway Date of Examination: 8/31/2020 Examination Level: RO SRO Operating Test Number: 2020-1 Administrative Topic (see Note) Type Describe activity to be performed Code*

G2.1.25 (3.9) Ability to interpret reference C, S, R materials, such as graphs, curves, tables etc.

Conduct of Operations JPM: Calculate Boron Addition for blocking RO Admin1 N P-11 with 0 and 1 untrippable control rod.

G2.1.4 (3.3) Knowledge of individual licensed C, S, R operator responsibilities related to shift staffing, Conduct of Operations such as medical requirements, no-solo N operation, maintenance of active license status, RO Admin2 10CFR55, etc.

JPM: Select available overtime shifts without violating shift staff limits in APA-ZZ-00905 G2.2.43 (3.0) Knowledge of the process used to S, D track inoperable alarms.

Equipment Control JPM: Complete CA2557, Computer Point RO Admin3 Status Log and delete PPC point from alarm G2.3.14 (3.4) Knowledge of radiation or C, S, R contamination hazards that may arise during Radiation Control normal, abnormal, or emergency conditions or M activities.

RO Admin4 JPM: Determine if a respirator should be worn NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 RO Administrative JPMs:

RO Admin #1 This is a NEW JPM. At the completion of this JPM, the applicant calculated the volume of borated water addition as 3245 lbm (+/- 1lbm) and 5893 lbm (+/- 1lbm) for the situations of 0 and 1 untrippable rods respectively.

RO Admin #2 This is a NEW JPM. At the completion of this JPM, the applicant determined they are available to fill the Thursday and Friday night shifts without violating the work hour restrictions of APA-ZZ-00905.

RO Admin #3 This is a BANK JPM that has not been used on at least the last 3 initial license exams. At the completion of this JPM, the applicant deleted PPC point AFQ0601 from alarm and correctly completed out a CA2557 column 1 and 5.

RO Admin #4 This is a MODIFIED BANK JPM. The original JPM (URO-ADM A004J) has not been used on at least the last 3 NRC ILT Exams.

This JPM was modified by adjusting the dose rate, internal dose without a respirator, the times to complete the job with and without a respirator, and the overall determination to wear a respirator.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Callaway Date of Examination: 8/31/2020 Examination Level: RO SRO Operating Test Number: 2020-1 Administrative Topic (see Note) Type Describe activity to be performed Code*

G2.1.2 (4.4) Knowledge of operator C, S, R responsibilities during all modes of plant Conduct of Operations operations.

SRO Admin1 N JPM: Determine risk management actions for NG07 Switchgear OOS.

G2.1.1 (4.2) Knowledge of conduct of operatins C, S, R requirements.

Conduct of Operations JPM: Determine Shift Manager Communication SRO Admin2 N Requirements.

G2.2.12 (4.1) Knowledge of surveillance C, S, R procedures.

Equipment Control JPM: Review OSP-EG-P01AC, determine SRO Admin3 N errors and document Operability concerns.

G2.3.14 (3.8) Knowledge of radiation or C, S, R contamination hazards that may arise during Radiation Control normal, abnormal, or emergency conditions or M activities.

SRO Admin4 JPM: Select individual to exceed dose limit for accident mitigation.

G2.4.44 (4.4) Knowledge of emergency plan C, S, R protective action recommendations Emergency Plan JPM: Complete CA 2843, PAR Flowchart.

SRO Admin5 N NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 SRO Administrative JPMs:

SRO Admin1 This is a NEW JPM. At the completion of this JPM, the applicant determined that 'B' ESW Train and the TDAFP shall be maintained operable. No work will be allowed on AEPS, Security Diesel, or within 50 feet of the startup transformer. The applicant also determined that 3 areas (NB02, TB 2065' general area, and the condensate polisher area) require a RMA walkdown.

SRO Admin2 This is a NEW JPM. At the completion of this JPM, the applicant listed the 6 individuals/positions that should be notified due to a contaminated, injured person that will be transported via ambulance offsite for medical assistance.

SRO Admin3 This is a NEW JPM. At the completion of this JPM, the applicant determined that the pump differential pressure was incorrectly calculated, Vibration data at position F is in the Alert Range and Vibration data at position J is in the Required Action Range which results in the 'A' CCW pump being declared Inoperable.

SRO Admin4 This is a MODIFIED BANK JPM. The bank JPM was used on the 2016 ILT NRC Exam. That JPM has been modified by changing the reason for the dose in excess of the federal limit along with the list of volunteers. Using HDP-ZZ-01450, the applicant will select 1 individual out of 8 choices (different from the 2016 choice) that meets the guidelines and complete section 1 of CA0276, Authorization To Exceed Federal Occupational Dose Limits, correctly per the included key.

SRO Admin5 This is a NEW TIME CRITICAL JPM. At the completion of this JPM, the applicant identified that sector D is also affected and the Evacuation distance changed from Sectors E, F, G 5 miles to D, E, F, and G 10 miles. The applicant also completed 3 sections (outline, method, reason for type of PA) of CA 2843, PAR Flowchart correctly.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Callaway Date of Examination: 8/31/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 2020-1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function Sim1. 024AA2.01 Emergency Boration / Loss of Shutdown Margin in N, S, L, A 1 Mode 4 - Emergency Borate from the RWST using 'B' SI Pump SRO-U Sim2. 006A4.02 Emergency Core Cooling System / Open EN, N, S 2 EMHV8802A ('A' SI Pump Discharge to Hot Leg Inj) in Mode 1

,2 or 3 SRO-U Sim3. 010A4.01 Pressurizer Pressure Control System (BB) / D, S, A 3 Respond to a Master Pressure Controller Failure Sim4. 002A2.03 Reactor Coolant System / Perform ES-0.2, Natural N, S, L, A 4P Circulation Cooldown, Steps #7 and 8 to initiate a RCS cooldown SRO-U Sim5. 103A4.04 Containment / Perform OTO-SA-00001, Attachment EN, N, S 5 C, Containment Isolation Phase A Recovery Sim6. 062A4.01 AC Electrical Distribution / Transfer 4160 VAC N, S, L 6 PB122 to transformer XPB123 (PA02 supplied) from XPB122 (PA01 supplied)

RO only Sim7. 073A4.03 Process Radiation Monitoring System / Source M, S 7 Check GH-RE-10B per OSP-SP-00001 Sim8. 008A4.01 CCW System / Alternate CCW pumps in a single N, S, A 8 train then respond to a loss of both pumps and swap CCW service loop to the other CCW train within 10 minutes.

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 P1. 055EK3.02 Loss of Offsite and Onsite Power / Manually load A, D, E, P1 6 equipment on to an AC Bus SRO-U P2. 061K1.07 Auxiliary Feedwater System / Emergency Makeup D, E 4S Water to CST per EOP Addendum 23 P3. 033A2.03 Spent Fuel Pool Cooling System / Perform Local N, R 8 actions to fill the SFP with Reactor Makeup Water Per OTN-EC-00001 Addendum 6 SRO-U

• All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1/ 1/ 1 (control room system)

(L)ow-Power/Shutdown 1/ 1/ 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 (P)revious 2 exams 3/ 3/ 2 (randomly selected)

(R)CA 1/ 1/ 1 (S)imulator Note 1: The JPMs from the 2017 exam were randomly selected by placing 11 slips of paper labeled A through K in a container. No JPMs from the 2019 NRC exam were available for random selection as those JPMs will be used as a part of 2020 Audit Exam.

Simulator JPMs S1 This is a NEW ALTERNATE PATH JPM. The applicant will perform actions of OTO-ZZ-00003, Loss of Shutdown Margin, and borate the RCS using the B SI Pump and associated piping. Both CCPs and the flowpath to the A SI pump are not available and/or successful in establishing emergency boration thereby requiring the use of the B SI Pump and associated flowpath.

S2 This is a NEW JPM. The applicant will perform actions of OTN-EM-00001, 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.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S3 This is an ALTERNATE PATH, BANK JPM. The JPM (URO-SBB-04-C166J(A))

was used on the 2016 ILT NRC Exam. The applicant will be directed to equalize RCS and Pressurizer Boron Concentration using OTG-ZZ-00004, Addendum 03.

When the master pressure controller is taken to AUTO the PZR spray valves fail open requiring the applicant to manually close the spray valves. Upon completion of this JPM, the master pressure controller failure has been addressed prior to a Reactor Trip being generated on low pressurizer pressure.

S4 This is a NEW ALTERNATE PATH JPM. The applicant will perform actions of ES-0.2, Natural Circulation Cooldown, steps 7 & 8. Conditions are such that the applicant will have to perform the RNO actions of both steps, starting 4 CRDM fans and using the more than one SG ASDs (as the applicant must diagnosis that the condenser is unavailable) to establish a cooldown rate without generating a Steam Line SI nor exceed the cooldown rate.

S5 This is a NEW JPM. The applicant will perform actions of OTO-SA-00001, Attachment C, CTMT Isolation Phase A Recovery. The applicant will have to reset the Phase A signal, establish RCP seal water return, restore sample flow to CTMT monitors, restore reactor makeup water, and provide a CTMT sump discharge flowpath.

S6 This is a NEW JPM. The applicant will perform actions of OTN-PB-00001, Addendum 3, Energizing and Cross-tying Buses PB121, PB122, and PB123, to transfer 4160 VAC bus PB122 from its normal supply (PA01 via XPB122) to its alternate supply (PA02 via XPB123) without deenergizing the bus.

S7 This is a MODIFIED BANK JPM. This JPM was developed and submitted for use on the 2017 ILT NRC exam but not used due to the final class makeup (No ROs in 2017). It was modified as the original JPM source checked a different monitor, BM-RE-52. The applicant will perform actions of OSP-SP-00001 and have successfully source checked all three channels of GH-RE-10B (105, 108, 109) by selecting each channel and then depressing the source check button on the RM11 console.

S8 This is a NEW ALTERNATE PATH Time Critical JPM. The applicant will perform actions of OTN-EG-00001 section 5.4 to alternate 'B' Train CCW pumps. After the pump swap, malfunctions occur resulting in a loss of all CCW. The applicant will start either 'A' Train CCW pump (EG HIS 21 or 23 to Start), establish service water to the 'A' CCW HX by opening EF HIS 51&59 and shift CCW loads to the

'A' Train by opening EG HS-15 and closing EG HS-16 10 minutes from the loss of both 'B' Train CCW pumps.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In Plant JPMs P1 This is an ALTERNATE PATH, BANK JPM. This original JPM was used on the 2017 NRC ILT Exam. This JPM was modified by the addition of NG breaker status cue sheets and by making one step that was previously not critical a critical step (but this was not considered a significant modification and therefore listed as a BANK JPM). Upon completion of this JPM, the applicant will have closed the following breakers thereby loading equipment onto the NB01 bus:

NG0301 for NG03 bus NG0303 for Battery Chargers NK23 NK51-20 for Control Room emergency lighting P2 This is a BANK JPM that has not been used on at least the last 3 initial license exams. 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.

P3 This is a NEW JPM. The applicant will enter the RCA and perform actions of OTN-EC-00001, Addendum 6, Filling the Spent Fuel Pool, Step 5.2.6. To initiate SFP makeup, the applicant will fully open ECV0076 and throttling open ECV0128 to achieve a SFP makeup flow of ~65,000 lbm/hr.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Callaway Date of Examination: 8/31/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 2020-1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function Sim1. 024AA2.01 Emergency Boration / Loss of Shutdown Margin in N, S, L, A 1 Mode 4 - Emergency Borate from the RWST using 'B' SI Pump Sim2. 006A4.02 Emergency Core Cooling System / Open EN, N, S 2 EMHV8802A ('A' SI Pump Discharge to Hot Leg Inj) in Mode 1

,2 or 3 Sim3. 010A4.01 Pressurizer Pressure Control System (BB) / D, S, A 3 Respond to a Master Pressure Controller Failure Sim4. 002A2.03 Reactor Coolant System / Perform ES-0.2, Natural N, S, L, A 4P Circulation Cooldown, Steps #7 and 8 to initiate a RCS cooldown Sim5. 103A4.04 Containment / Perform OTO-SA-00001, Attachment EN, N, S 5 C, Containment Isolation Phase A Recovery Sim7. 073A4.03 Process Radiation Monitoring System / Source M, S 7 Check BM-RE-52 per OSP-SP-00001 Sim8. 008A4.01 CCW System / Alternate CCW pumps in a single N, S, A 8 train then respond to a loss of both pumps and swap CCW service loop to the other CCW train within 10 minutes.

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. 055EK3.02 Loss of Offsite and Onsite Power / Manually load A, D, E, P1 6 equipment on to an AC Bus

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 P2. 061K1.07 Auxiliary Feedwater System / Emergency Makeup D, E 4S Water to CST per EOP Addendum 23 P3. 033A2.03 Spent Fuel Pool Cooling System / Perform Local N, R 8 actions to fill the SFP with Reactor Makeup Water Per OTN-EC-00001 Addendum 6

• All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1/ 1/ 1 (control room system)

(L)ow-Power/Shutdown 1/ 1/ 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 (P)revious 2 exams 3/ 3/ 2 (randomly selected)

(R)CA 1/ 1/ 1 (S)imulator Note 1: The JPMs from the 2017 exam were randomly selected by placing 11 slips of paper labeled A through K in a container. No JPMs from the 2019 NRC exam were available for random selection as those JPMs will be used as a part of 2020 Audit Exam.

Simulator JPMs S1 This is a NEW ALTERNATE PATH JPM. The applicant will perform actions of OTO-ZZ-00003, Loss of Shutdown Margin, and borate the RCS using the B SI Pump and associated piping. Both CCPs and the flowpath to the A SI pump are not available and/or successful in establishing emergency boration thereby requiring the use of the B SI Pump and associated flowpath.

S2 This is a NEW JPM. The applicant will perform actions of OTN-EM-00001, 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.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S3 This is an ALTERNATE PATH, BANK JPM. The JPM (URO-SBB-04-C166J(A))

was used on the 2016 ILT NRC Exam. The applicant will be directed to equalize RCS and Pressurizer Boron Concentration using OTG-ZZ-00004, Addendum 03.

When the master pressure controller is taken to AUTO the PZR spray valves fail open requiring the applicant to manually close the spray valves. Upon completion of this JPM, the master pressure controller failure has been addressed prior to a Reactor Trip being generated on low pressurizer pressure.

S4 This is a NEW ALTERNATE PATH JPM. The applicant will perform actions of ES-0.2, Natural Circulation Cooldown, steps 7 & 8. Conditions are such that the applicant will have to perform the RNO actions of both steps, starting 4 CRDM fans and using the more than one SG ASDs (as the applicant must diagnosis that the condenser is unavailable) to establish a cooldown rate without generating a Steam Line SI nor exceed the cooldown rate.

S5 This is a NEW JPM. The applicant will perform actions of OTO-SA-00001, Attachment C, CTMT Isolation Phase A Recovery. The applicant will have to reset the Phase A signal, establish RCP seal water return, restore sample flow to CTMT monitors, restore reactor makeup water, and provide a CTMT sump discharge flowpath.

S7 This is a MODIFIED BANK JPM. This JPM was developed and submitted for use on the 2017 ILT NRC exam but not used due to the final class makeup (No ROs in 2017). It was modified as the original JPM source checked a different monitor, BM-RE-52. The applicant will perform actions of OSP-SP-00001 and have successfully source checked all three channels of GH-RE-10B (105, 108, 109) by selecting each channel and then depressing the source check button on the RM11 console.

S8 This is a NEW ALTERNATE PATH Time Critical JPM. The applicant will perform actions of OTN-EG-00001 section 5.4 to alternate 'B' Train CCW pumps. After the pump swap, malfunctions occur resulting in a loss of all CCW. The applicant will start either 'A' Train CCW pump (EG HIS 21 or 23 to Start), establish service water to the 'A' CCW HX by opening EF HIS 51&59 and shift CCW loads to the

'A' Train by opening EG HS-15 and closing EG HS-16 10 minutes from the loss of both 'B' Train CCW pumps.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In Plant JPMs P1 This is an ALTERNATE PATH, BANK JPM. This original JPM was used on the 2017 NRC ILT Exam. This JPM was modified by the addition of NG breaker status cue sheets and by making one step that was previously not critical a critical step (but this was not considered a significant modification and therefore listed as a BANK JPM). Upon completion of this JPM, the applicant will have closed the following breakers thereby loading equipment onto the NB01 bus:

NG0301 for NG03 bus NG0303 for Battery Chargers NK23 NK51-20 for Control Room emergency lighting P2 This is a BANK JPM that has not been used on at least the last 3 initial license exams. 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.

P3 This is a NEW JPM. The applicant will enter the RCA and perform actions of OTN-EC-00001, Addendum 6, Filling the Spent Fuel Pool, Step 5.2.6. To initiate SFP makeup, the applicant will fully open ECV0076 and throttling open ECV0128 to achieve a SFP makeup flow of ~65,000 lbm/hr.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Callaway Date of Examination: 8/31/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 2020-1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function Sim1. 024AA2.01 Emergency Boration / Loss of Shutdown Margin in N, S, L, A 1 Mode 4 - Emergency Borate from the RWST using 'B' SI Pump Sim2. 006A4.02 Emergency Core Cooling System / Open EN, N, S 2 EMHV8802A ('A' SI Pump Discharge to Hot Leg Inj) in Mode 1 ,2 or 3 Sim4. 002A2.03 Reactor Coolant System / Perform ES-0.2, Natural N, S, L, A 4P Circulation Cooldown, Steps #7 and 8 to initiate a RCS cooldown In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. 055EK3.02 Loss of Offsite and Onsite Power / Manually load A, D, E, P1 6 equipment on to an AC Bus P3. 033A2.03 Spent Fuel Pool Cooling System / Perform Local N, R 8 actions to fill the SFP with Reactor Makeup Water Per OTN-EC-00001 Addendum 6

• All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1/ 1/ 1 (control room system)

(L)ow-Power/Shutdown 1/ 1/ 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 (P)revious 2 exams 3/ 3/ 2 (randomly selected)

(R)CA 1/ 1/ 1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Note 1: The JPMs from the 2017 exam were randomly selected by placing 11 slips of paper labeled A through K in a container. No JPMs from the 2019 NRC exam were available for random selection as those JPMs will be used as a part of 2020 Audit Exam.

Simulator JPMs S1 This is a NEW ALTERNATE PATH JPM. The applicant will perform actions of OTO-ZZ-00003, Loss of Shutdown Margin, and borate the RCS using the B SI Pump and associated piping. Both CCPs and the flowpath to the A SI pump are not available and/or successful in establishing emergency boration thereby requiring the use of the B SI Pump and associated flowpath.

S2 This is a NEW JPM. The applicant will perform actions of OTN-EM-00001, 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.

S4 This is a NEW ALTERNATE PATH JPM. The applicant will perform actions of ES-0.2, Natural Circulation Cooldown, steps 7 & 8. Conditions are such that the applicant will have to perform the RNO actions of both steps, starting 4 CRDM fans and using the more than one SG ASDs (as the applicant must diagnosis that the condenser is unavailable) to establish a cooldown rate without generating a Steam Line SI nor exceed the cooldown rate.

In Plant JPMs P1 This is an ALTERNATE PATH, BANK JPM. This original JPM was used on the 2017 NRC ILT Exam. This JPM was modified by the addition of NG breaker status cue sheets and by making one step that was previously not critical a critical step (but this was not considered a significant modification and therefore listed as a BANK JPM). Upon completion of this JPM, the applicant will have closed the following breakers thereby loading equipment onto the NB01 bus:

NG0301 for NG03 bus NG0303 for Battery Chargers NK23 NK51-20 for Control Room emergency lighting P3 This is a NEW JPM. The applicant will enter the RCA and perform actions of OTN-EC-00001, Addendum 6, Filling the Spent Fuel Pool, Step 5.2.6. To initiate SFP makeup, the applicant will fully open ECV0076 and throttling open ECV0128 to achieve a SFP makeup flow of ~65,000 lbm/hr.

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.1, Rev 0 Op-Test No.: 2020-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100%, MOC Turnover: No equipment out of service. After the crew takes the watch, park Control Bank A, B, &C at 228 steps per OTG-ZZ-00004, Attachment 3, Parking Control Rods at 228 steps.

Event Malf. No. Event Event No. Type* Description SRO (R) Park Control Bank A, B, & C at 228 steps per OTG-ZZ-00004, 1 N/A RO (R) Attachment 3, Parking Control Rods at 228 steps.

EA / PEA2101B = 1 EA / SRO (C) Service Water pump lockout with a failure of the standby pump, 2 BOP (C)

EA09PB12104TF_ Essential Service Water placed in service per Annunciator 12A.

Open = 1 SRO (I) CVCS Letdown (LTDN) HX Temperature Controls fails upscale X01A103P = 0.8 3 causing LTDN temperature to slowly rise requiring manual ramp 180 sec RO (I) control to stabilize.

'A' SG Pressure Instrument, (AB PT514) fails downscale AB / ABPT514=0.1 SRO (I) 4 requiring the crew to enter OTO-AE-00002. Technical ramp 120 sec BOP (I) Specification determination.

BB / BB005 = 100, RCS Excessive Leakage (30 gpm) requiring the crew to enter ramp 60 sec SRO (C) OTO-BB-00003. Technical Specification determination and 5 BG / BG002 = 30, RO (C) then the leak grows to over 50 gpm requiring a manual reactor no ramp trip. E-0 Reactor Trip or Safety Injection.

SA / SRO (I) Safety Injection fails to Automatically Actuate SIS_A_Block_Auto RO (I) 6 =0 SIS_B_Block_Auto CT-2, Manually actuate SI

=0 RCS LOCA with 'A' SI pump autostart failure, 'B' ESW Pump BB / BB005 = 500, trip, 'A' RHR Pump trip, E-1 Loss of Primary or Secondary ramp 60 sec SRO (M) Coolant.

7 EM / PEM01A=1 RO (M)

EF / PEF01B=1 BOP (M)

EJ / PEJ01A=1 CT-6, Establish flow from at least one Charging/SI pump CT-16, Manually Trip RCPs

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 6

Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes

1. Total malfunctions (5-8) 8

2. Malfunctions after EOP entry (1-2) 2

3. Abnormal events (2-4) 2

4. Major transients (1-2) 1

5. EOPs entered/requiring substantive actions (1-2) 1

6. EOP contingencies requiring substantive actions (0-2) 0

7. Critical tasks (2-3) 3 Page 2 of 6

Scenario #1 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 The Plant is stable at 100% with no equipment out of service. Once the crew has taken the watch, they will park Control Bank A, B, & C at 228 steps per OTG-ZZ-00004, Attachment 3, Parking Control Rods at 228 Steps.

After Control Banks A, B, & C are parked at 228 steps, the 'B' Service Water pump trips and the standby service water pump did not auto start and can't be manually started. With only 1 Service Water pump running, the crew should place both trains of Essential Service Water (ESW) in service to restore Service Water pressure per Annunciator 12A, Service Water Pump Lockout.

After ESW is placed in service, Letdown Temperature HX controller, BGTK130, slowly fails causing letdown temperature to slowly rise. The crew should utilize Annunciator 39B to take manual control of BGTK130 and stabilize the plant.

After addressing the LTDN HX controller failure, 'A' SG Pressure Instrument (AB PT-514) will fail downscale. AB-PT-514 compensates AB FT-512 and the crew should utilize OTO-AE-00002, SG Water Level Control Malfunction, to remove the faulted instruments from control. This failure will result in Technical Specification 3.3.1, and 3.3.2 not being met.

Once the Technical Specification for the SG pressure instrument are addressed, a small non isolable leak inside containment (30 gpm) will develop on the letdown line. The crew should respond by entering OTO-BB-00003, RCS Excessive Leakage. This failure will result in Technical Specification 3.4.13, RCS Operational Leakage, not being met. Once this Technical Specification is addressed, a RCS leak on surge line will develop requiring a Reactor Trip and Safety Injection. The automatic Safety Injection (SI) will not work requiring the crew to manually initiate SI.

During the immediate actions of E-0, LOCA increases on the PZR surge line. Additionally, the

'B' ESW pump will trip, the 'A' SI pump will fail to autostart when manual safety injection is initiated and the 'A' RHR pump will trip during the performance of E-0. As the 'B' ESW is the heat sink for the 'B' Train ECCS, this effectively removes the 'B' train from service (Note, if the crew does not secure the 'B' Train ECCS pumps, these pumps begin tripping due to high temperature with various time delays.) While the leak size is relatively small (RCS pressure does not lower to below RHR pump head), the 'A' CCP cannot recover RCS pressure and level, therefore requiring the manual start of the 'A' SI pump to stabilize the plant. RCS pressure does lower far enough to meet the RCP trip criteria. The loss of both RHR pumps will require the crew to transition to ECA-1.1 at E-1 step #12 RNO.

The scenario is complete when the crew has transitioned to ECA-1.1 at step #12 of E-1, secured the RCPs, and started the 'A' SI pump.

Page 3 of 6

Scenario #1 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 Critical Tasks:

Critical Tasks Trip all RCPs within 5 minutes of meeting RCP trip criteria. (optional) Establish RCS Injection flow from 'A' SI pump before completion of E-0 Attachment A EVENT 7 7 Safety Failure to trip the RCPs under the postulated plant conditions leads to core uncovery The acceptable results obtained in the FSAR analysis of a small-break LOCA are significance and to fuel cladding temperatures in excess of 2200°F, which is the limit specified in predicated on the assumption of minimum ECCS pumped injection. The analysis the ECCS acceptance criteria. Thus, failure to perform the task represents assumes that a minimum pumped ECCS flow rate, which varies with RCS pressure, is misoperation or incorrect crew performance in which the crew has failed to prevent injected into the core. The flow rate values assumed for minimum pumped injection are degradation of...{the fuel cladding} ...barrier to fission product release and which based on operation of one each of the following ECCS pumps: Charging/SI pump (HP leads to violation of the facility license condition. plants only), high-head SI pump, and low-head SI pump. Operation of this minimum required complement of ECCS injection pumps is consistent with the FSAR assumption that only minimum safeguards are actuated. Because compliance with the assumptions of the FSAR is part of the facility license condition, failure to perform the critical task (under the postulated plant conditions) constitutes a violation of the license condition.

Cueing Indications of a SBLOCA Indication and/or annunciation that Charging/SI pump injection is required AND

• SI actuation Indication and/or annunciation of safety injection

• RCS pressure below the shutoff head of the Charging/SI pump AND Indication and/or annunciation that no Charging/SI pump is injecting into the core Indication and/or annunciation that at least one CCP/SI pump is running

• Control switch indication that the circuit breakers or contactors for both AND Charging/SI pumps are open Indication that the RCP trip criteria are met

• All Charging/SI pump discharge pressure indicators read zero

• All flow rate indicators for Charging/SI pump injection read zero Performance Manipulation of controls as required to trip all RCPs Starting the 'A' SI pump indicator

• RCP breaker position lights indicate breaker open Performance Indication that all RCPs are stopped With the 'B' train of ECCS have lost its cooling water pump and the size of the LOCA, feedback

• RCP breaker position lights the 'A' CCP will not be able to restore RCS water level while RCS pressure remains

• RCP flow decreasing greater than the 'A' RHR pump head. As RCS Pressures lowers below the SI pump

• RCP motor amps decreasing shutoff head, the 'A' SI pump will be required and once the 'A' SI pump is started, Indication and/or annunciation that the A SI is injecting and Flow rate indication of injection from the A SI pump Justification for In a letter to the NRC titled Justification of the Manual RCP Trip for Small Break LOCA before completion of Attachment A of E-0 is in accordance with the PWR Owners the chosen Events (OG-117, March 1984) (also known as the Sheppard letter), the WOG Group Emergency Response Guidelines. It allows enough time for the crew to take the performance limit provided the required assurance based on the results of the analyses performed in correct action while at the same time preventing avoidable adverse consequences.

conjunction with WCAP-9584. The WOG showed that for all Westinghouse plants, more than two minutes were available between onset of the trip criteria and depletion of RCS inventory to the critical inventory. In fact, additional analyses sponsored by the WOG in connection with OG-117 conservatively showed that manual RCP trip could be delayed for five minutes beyond the onset of the RCP trip criteria without incurring any adverse consequence.

PWR Owners CT-16, Manually Trip RCPs CT-6, Establish flow from at least one Charging/SI pump Group Appendix Page 4 of 6

Scenario #1 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 Critical Tasks Manually actuate at least one train of SIS-actuated safeguards equipment before transition to E-1, E-2 or E-3 series or transition to any FRG.

EVENT 6 Safety Failure to manually actuate SI under the postulated conditions constitutes mis-significance 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 Cueing Indication and/or annunciation that that Sl is required

• PRZR pressure or SG pressure less than SI actuation setpoint

• Containment pressure greater than SI actuation setpoint

• 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 Performance Manipulation of controls as required to actuate at least one train of Sl indicator

• SB HS-27

• SB HS-28 Performance Indication that both Trains of SI - Actuated feedback

• LOCA Sequencer annunciator 30A - Lit

• LOCA Sequencer annunciator 30B - Lit

• SB069 SI Actuate Red Light - Lit SOLID (NOT blinking)

Justification for The crew has had ample opportunity to recognize the need for Sl and the fact that Sl the chosen has not automatically actuated.

performance limit Given the postulated plant conditions, transition from E-0 to ES-0.1 constitutes an error in using the E-0 procedure. The crew is in the wrong procedure; however, the crew is allowed to recover from this error up through Step 3.a of ES-0.1.

The ERG network is designed to "catch" errors in procedure usage. Step 3.a is designed to get the crew back to E-0, if that is in fact where the crew should be. If the crew members pass through Step 3.a and remain in ES-0.1, they have missed the last step that would return them to the correct procedure.

PWR Owners CT-2, Manually actuate SI Group Appendix NOTE: (Per NUREG-1021, Appendix D) If an operator or the Crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Page 5 of 6

Scenario #1 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 References OTG-ZZ-00004, Power Operations OTA-RK-00014, Addendum 12A, Service Water Pump Lockout OTA-RK-00018, Addendum 39B, Letdown Heat Exchanger Discharge Temperature High OTO-AE-00002, SG Water Level Control Malfunctions OTO-BB-00003, RSC Excessive Leakage E-0, Reactor Trip or Safety Injection E-1, Loss of Primary or Secondary Reactor Coolant Technical Specification 3.4.13, RCS operational Leakage Technical Specification 3.3.1, RTS Instrumentation Technical Specification 3.3.2, ESFAS Instrumentation ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions:

1. Medium Break LOCA (19% contribution to CDF)

Page 6 of 6

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.2, Rev 0 Op-Test No.: 2020-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 50 to 55%, BOC Turnover: No equipment out of service. The plant is performing a beginning of Cycle startup. Once the crew has the watch, the crew will start a second feed pump per OTN-AE-00001 and OTG-ZZ-00004.

Event Malf. No. Event Event No. Type* Description SRO (N) 1 N/A Start a second feed pump per OTN-AE-00001 Addendum 2.

BOP (N)

BB / BBLT0460 = SRO (I) PZR level instrument (BBLT460) fails downscale isolating 2 466.1, ramp=60 letdown. OTO-BG-00001 to remove failed instrument and sec RO (I) restore letdown. Technical Specification Determination.

EG / EG004=300, SRO (C) CCW Leak on the supply to Radwaste. OTO-EG-00001, CCW 3 BOP (C) System Malfunction to isolate the leak.

ramp = 60sec SRO (R) Tube Leak on Steam Generator 'B',

4 EBB01B=20 RO (R) OTO-BB-00001, Steam Generator Tube Leak. Technical BOP (R) Specification Determination.

Tube Rupture in Steam Generator 'B' SRO (M) E-3, Steam Generator Tube Rupture EBB01B=600, 5 RO (M) ramp = 30sec BOP (M) CT-18, Isolate the Ruptured SG CT-19, Control initial RCS cooldown SA / SRO (C) 6 Failure of 'B' MSIV (AB HIS-17) to close, fast close all MSIVs SAS9XX_2=1 BOP (C)

BB /

BBPCV0455B_2= Failure of normal PZR Sprays valves during E-3 performance -

SRO (C) 0.1 requires PZR PORVS to depressurize RCS.

7 RO (C)

BB /

BBPCV0455C_2 BOP (C)

CT 20, Depressurize RCS to E-3 SI termination criteria

=0.1

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 6

Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes

1. Total malfunctions (5-8) 6

2. Malfunctions after EOP entry (1-2) 2

3. Abnormal events (2-4) 3

4. Major transients (1-2) 1

5. EOPs entered/requiring substantive actions (1-2) 1

6. EOP contingencies requiring substantive actions (0-2) 0

7. Critical tasks (2-3) 3 Page 2 of 6

Scenario #2 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 The Plant is stable at 50-55% with no equipment out of service. A beginning of cycle startup is in progress per OTG-ZZ-00004, Power Operations. Once the crew has the watch, the crew is to startup a second MFP per OTG-ZZ-00004 step 5.5.1.b and OTN-AE-00001, Addendum 2, MFP Operations.

After the 2nd MFP is placed in service, PZR level instrument (BBLT460) will fail downscale and isolate letdown. (BBLT460 will be the bottom selected PZR level instrument which service an alarm function and letdown isolation; in this initial configuration, it does not input into the PZR level controller.) The crew should enter OTO-BG-00001, remove the failed instrument from control and restore letdown. This failure will result in Technical Specification 3.3.1 not being met.

After letdown is restored, a CCW leak on the header to radwaste occurs causing Annunciator 52F to alarm. The crew should enter OTO-EG-00001 and utilize Attachment B to find the leak and isolate it.

Once the crew has initiated restoration of control rod positions, a 20 gpm tube leak develops in Steam Generator 'B'. The crew should enter OTO-BB-00001, Steam Generator Tube Leak, and quantify the leak to be greater than 150 gpd. The crew should initiate a load reduction to below 50% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using OTO-MA-00008, Rapid Load Reduction. This failure will result in Technical Specification 3.4.13 and 3.4.17 not being met.

After a measurable downpower (>2%), the Tube Leak will grow into a tube rupture. The crew should manually trip the reactor and initiate Safety Injection. The crew should implement E-0 and transition to E-3, Steam Generator Tube Rupture at E-0 step #15. The crew should direct an Operations Technician to locally close ABV0085, Steam Supply to Turbine Driven AFW pump.

During the performance of E-3, the 'B' MSIV will fail to close which will require the crew to fast close all MSIVs at E-3 Step 3.g RNO. This will remove the option to cooldown the RCS by dump steam to the condenser at E-3 step 6 requiring the crew to use 'A', 'C', and 'D' ASDs per step 6.d RNO.

When the crew attempts to depressurize the RCS to minimize break flow at step #16 of E-3, the crew should determine that both normal PZR Spray valves have failed close and will not reopen.

This will result in the crew proceeding to Step #17 (per step 16 RNO) and utilizing the PZR PORVs to depressurize the RCS to E-3 SI termination criteria are met.

The scenario is complete when the crew has completed the initial depressurization using one PORV until conditions are met in E-3 step 17.b to close the PORV and the crew closed the open PORV per E-3 step 17.c.

Page 3 of 6

Scenario #2 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 Critical Tasks:

Critical Tasks Isolate feedwater flow into and steam flow from the ruptured SG before a transition to Establish/maintain an RCS temperature so that transition from E-3 does not occur ECA-3.1 occurs. because the RCS temperature is in either of the following conditions:

• Too high to maintain minimum required subcooling OR

• Below the RCS temperature that causes an extreme (RED path) or a severe (ORANGE path) challenge to the subcriticality and/or the integrity CSF EVENT 5, 6 5 Safety Isolating the ruptured SG maintains a differential pressure between the ruptured SG Failure to establish and maintain the correct RCS temperature during a SGTR leads to significance and the intact SGs. The differential pressure (250 psi) ensures that minimum RCS a transition from E-3 to a contingency ERG. This failure constitutes an incorrect subcooling remains after RCS depressurization. performance that necessitates the crew taking compensating action that would complicate the event mitigation strategy....

Cueing All of the following: All of the following:

• Indication and/or annunciation of SGTR in one SG

• Indication and/or annunciation of SGTR in one SG o Increasing SG water level o Increasing SG water level / Radiation o Radiation

• Indication and/or annunciation of reactor trip

• Indication and/or annunciation of reactor trip

• Indication and/or annunciation of SI

• Indication and/or annunciation of SI

• Indication of ruptured SG pressure greater than minimum required pressure Performance Manipulation of controls as required to isolate the ruptured 'B' SG Manipulation of controls as required to establish and maintain RCS temperature indicator

• Close TDAFP Steam Supply from Main Steam Loop Manual Isolation valve

• Steam dump valve position lamps and/or indicators indicate closed o ABV0085 (SG B)

• SG PORV valve position lamps and/or indicators indicate closed

• Close Steam line low point drain valve from ruptured SG o AB HIS-8 (SG B)

• Fast Close all remaining MSIVs (A ,C & D MSIVs) and Bypass valves:

o AB HS-79 and AB HS-80

• Stop Auxiliary feed flow to ruptured SG o CLOSE AL HK-9A and AL HK-10A Performance Crew will observe the following: Indication of steam flow rate greater than zero feedback

• Indication of stable or increasing pressure in the ruptured SG

• Indication of RCS temperature decreasing

• Indication of decreasing or zero feedwater flow rate in the ruptured SG OR

• Indication of RCS temperature less than target value Justification for When the crew cannot maintain the 250 psi differential, the ERGs require a transition Terminating the RCS cooldown before reaching the target temperature prevents the chosen to contingency ERG ECA-3.1. This transition unnecessarily delays the sequence of achieving the minimum RCS subcooling. Failure to achieve the required RCS performance limit actions leading to RCS depressurization and Sl termination. subcooling results in a condition that forces the crew to transition to contingency ERG ECA-3.1, thereby delaying the RCS depressurization and SI termination. Such a delay allows the excessive inventory increase of the ruptured SG to continue until the SG overpressure components release water or until SG overfill occurs.

Terminating the cooldown too late challenges either the subcriticality CSF or the integrity CSF. Because the crew is directed to cool down at the maximum rate, late termination of cooldown could force the RCS temperature low enough to challenge the integrity CSF. The crew must then transition to one of the integrity FRGs. The transition also delays RCS depressurization and SI termination.

PWR Owners CT-18, Isolate the Ruptured SG CT-19, Control initial RCS cooldown Group Appendix Page 4 of 6

Scenario #2 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 Critical Tasks Depressurize RCS using one PORV per E-3 step#17 until E-3 SI termination criteria is met.

EVENT 7 Safety RCS depressurization decreases the RCS leakage into the SG, helping to mitigate the significance inventory increase in the ruptured SG. The RCS depressurization also helps the ECCS restore RCS inventory, which in turn allows SI termination. SI termination eliminates the remaining cause of leakage from the RCS into the SG.

Cueing All of the following:

• Indication and/or annunciation of SGTR in one SG

• Indication and/or annunciation of reactor trip and SI

• Indication that the RCS is cooled down to the target temperature Performance Manipulation of controls as required to depressurize the RCS indicator

• Valve position lamps show PRZR PORV open Performance Crew will observe the following:

feedback

• Indication of RCS pressure decreasing

• Indication of PRZR level increasing Justification for The intent is to depressurize to establish and maintain the criteria that allow the crew the chosen to terminate SI. Before depressurization, the crew has met most of the criteria for SI performance limit termination. The most likely criterion not met is adequate pressurizer level. The depressurization establishes pressurizer level within the range to allow termination.

However, if the crew depressurizes too much, the existing subcooling can be lost, inhibiting termination. In addition, if the crew fails to realign the controls after depressurization, RCS pressure will continue to decrease, also inhibiting termination.

PWR Owners CT-20, Depressurize RCS to E-3 SI termination criteria Group Appendix NOTE: (Per NUREG-1021, Appendix D) If an operator or the Crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Page 5 of 6

Scenario #2 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 References OTG-ZZ-00004, Power Operations OTN-AE-00001, Feedwater System OTN-AE-00001, Addendum 2, MFP Operations OTO-EG-00001, CCW System Malfunction OTO-BB-00001, Steam Generator Tube Leak OTO-MA-00008, Rapid Load Reduction E-0, Reactor Trip or Safety Injection E-3, Steam Generator Tube Rupture Technical Specification 3.3.1, RTS Instrumentation Technical Specification 3.4.13, RCS Operational Leakage Technical Specification 3.4.17, SG Tube Integrity ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions:

1. SG Tube Rupture (2% contribution to CDF)

Page 6 of 6

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.3, Rev 0 Op-Test No.: 2020-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: ~22% MOC Turnover: No equipment out of service. Boron equalization is in service. Once the crew has the watch, the crew will transfer SG Level Control from the MFRV Bypass Valves to the MFRVs per OTN-AE-00001, Feedwater System, Section 5.10.

Event Malf. No. Event Event No. Type* Description SRO (N) Transfer SG Level Control from the MFRV Bypass Valves 1 N/A to the MFRVs per OTN-AE-00001 section 5.10 starting at BOP (N) step 5.10.11.

PG / SRO (C) PZR Backup Heater Group 'A' feeder breaker trips and 2

PG2101_OCTF_FAIL=1 RO (C) locks out. Technical Specification Determination.

BB /

PBB01B_SWZWBRRG=1 SRO (C) 'B' RCP develops abnormal motor temperatures, OTO-BB-3 , 10 sec ramp RO (C) 00002, RCP Off Normal and secure 'B' RCP per BOP (C) Attachment E. Technical Specification Determination.

BBV0143=1, 120 delay, ramp=120 sec SRO (C)

Turbine Lube Oil Malfunction which leads to a Turbine Trip 4 PCB01=0(Fail) RO (C) below P9, OTO-AC-00001 (Turbine Trip below P-9)

BOP (C)

NB / NB01_F=1, PB / PB05=1 MD / MDCB1=1 MD / MDLC1=1, delay=1 MD / MT7=1, delay=2 SRO (C) Loss of Offsite AC power and NB01 and PB05 bus fault.

5 MD / MDMT8=1, delay=5 E-0 Reactor Trip or Safety Injection.

BOP (C)

MD / ESFB=1, delay=5 X06I78T=1, delay=1 CALBLAND1 and CALLC2=OPEN delay=7 MTGYCAL= OPEN delay=6 SRO (M)

PKJ06B=1 'B' EDG (NE02) vital trip. Loss of ALL AC power. ECA-6 RO (M)

KJL02=1018.5 0.0, Loss of ALL AC Power.

BOP (M)

'A' PZR PORV failed open during SBO.

X21I1490=1, delay = 30 SRO (C) 7 X21II1490 (delete) BOP (C) CT-22, Manually close 'A' PZR PORV during SBO.

Offsite Power becomes available. Restore NB02 with delIA MDCB1 2 SRO (C) offsite power per EOP Addendum 7.

8 delIA CALBLAND1 2 BOP (C)

CT-24, Energize NB02 emergency bus.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 5

Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes

1. Total malfunctions (5-8) 8

2. Malfunctions after EOP entry (1-2) 2

3. Abnormal events (2-4) 2

4. Major transients (1-2) 1

5. EOPs entered/requiring substantive actions (1-2) 1

6. EOP contingencies requiring substantive actions (0-2) 1

7. Critical tasks (2-3) 2 Page 2 of 5

Scenario #3 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 The Plant is stable at ~22% with boron equalization in service and no equipment out of service.

Once the crew has the watch, the crew is to transfer SG Level Control from the MFRV Bypass Valves to the MFRVs per OTN-AE-00001 section 5.10 starting at step 5.10.11 and OTG-ZZ-00003, step 5.4.6.

Once the transfer to MFRVs is complete, the 'A' Backup Pressurizer Heaters trip and lockout.

The crew should response per Annunciator 33E, PZR Heater Group Lockout. This failure will result in Technical Specification 3.4.9 not being met.

Once the plant and PZR have been stabilized, 'B' RCP motor develops high temperatures (due to a partial loss of CCW cooling flow to the motor). The crew should secure the 'B' RCP per OTO-BB-00002, RCP Off Normal, Attachment E. This failure will result in Technical Specification 3.4.4 not being met.

After the 'B' RCP is secured and Technical Specification addressed, a turbine lube oil malfunction leads to a Main Turbine trip. The crew should enter OTO-AC-00001, Turbine Trip below P-9, and stabilize the plant. When the crew has completed transferring Steam Dumps To Steam Pressure per OTO-AC-00001, a loss of offsite power occurs concurrent with a 4160 VAC NB01 bus lockouts due to a bus fault.

After the crew performs the immediate actions of E-0 and transitions to ES-0.1 at E-0 Step#4, the 'B' EDG (NE02) trips and cannot be restarted 4160 VAC NB01 and PB05 buses lockout due to a bus faults. This results in a loss of All AC Power and the crew should enter ECA-0.0, Loss of All AC Power. The PB05 bus fault is solely present to make COOP power unavailable for power restoration.

30 seconds after NB01 locks out, the 'A' PZR PORV fails open. The crew should take manual action to close the open PORV during a station blackout per step #3 of ECA-0.0.

Shortly after entry into ECA-0.0, offsite power (CAL - Bland1) becomes available. The crew should restore power to NB02 using EOP Addendum 7, Restoring Offsite Power per ECA-0.0 step 5.a.

The scenario is complete when NB02 bus is reenergized by offsite power.

Page 3 of 5

Scenario #3 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 Critical Tasks:

Critical Tasks Energize NB02 AC Emergency Bus using EOP Addendum 7, Restoring Offsite Power. Manually close the Open 'A' PORV before completing Step 3 of ECA-0.0.

EVENT 8 7 Safety In the scenario, failure to energize at least one ac emergency bus results in the The open PORV greatly increases the rate at which RCS inventory is depleted, at a significance needless continuation of a situation in which the pumped ECCS capacity and the time when the lost inventory cannot be replaced by active injection. Thus, failure to emergency power capacity are both in a completely degraded status, as are all other close the PORV defeats the basic purpose of ECA-0.0. Additionally, it is critical that active safeguards requiring electrical power. Although the completely degraded status the PORV be closed as soon as possible. Hence, manual closure of the PORV (when is not due to the crew's action (was not initiated by operator error), continuation in the the PORV is open and RCS pressure is less than [the setpoint for automatic closure]4) completely degraded status is a result of the crew's failure to energize at least one ac is imperative and urgent in order to ensure the effectiveness of subsequent actions in emergency bus. extending the time to core uncovery.

Cueing Indication and/or annunciation that all ac emergency buses are de-energized All of the following:

• Bus energized lamps extinguished

• Indication and/or annunciation of station blackout

• Circuit Breaker Position

• Valve position indication and/or annunciation that the PRZR PORV is open

• Bus Voltage

• Indication that RCS pressure is below [the setpoint at which the PRZR

• EDG status PORV should reclose automatically]

• Indication and/or annunciation of decreasing RCS pressure

• Indication and/or annunciation consistent with the discharge of PRZR fluid to the PRT o PRT temperature, level, pressure o PZR PORV Tailpipe RTDs Performance Manipulation of controls as required to energize NB02 from offsite power: Manipulation of controls as required to close the 'A' PRZR PORV indicator o PCB-V45 (BUS B CAL-BLAND-1)

• BB HIS-455A o Bus Tie Breaker, PCB-V43 o Startup Xfmr 1 Bus A, PCB-V41 o NB HS-11, NB02 Sync Scope Sel o NB HS-8, NB02 NORM SPLY SYNC TRANSFER o NB HIS-4, NB02 NORM SPLY BKR NB0209 Performance Indication that NB02 is energized: Indication that 'A' PZR PORV is closed feedback

• NB02 Bus energized light

• PRZR pressure stabilizes

• NB02 bus voltage * 'A' PRZR PORV indicates closed Justification for Failure to perform the critical task prior to the completion of EOP Addendum 39 results This performance standard is imposed because it is imperative and urgent that the the chosen in needless degradation of RCS barrier (and to fission product release, specifically of PRZR PORV be closed in order for the strategy of ECA-0.0 to succeed. The PORV performance limit the RCS barrier at the point of the RCP seals. Failure to perform the critical task constitutes a very large leakage path. Leaving it open causes rapid depletion of RCS means that RCS inventory lost through the RCP seals cannot be replaced. It also inventory at a time when that inventory cannot be replaced.

means that the RCP seals remain without cooling and gradually deteriorate. As the seals deteriorate the rate of RCS inventory loss increases. In step 3 of ECA-0.0, the crew is directed to check the major RCS outflow paths that could contribute to rapid depletion of RCS inventory. The PRZR PORVs offer the largest potential for RCS inventory loss.

Therefore, they are an outflow path that must be checked and, if necessary, closed.

PWR Owners CT - 24, Energize at least one ac emergency bus CT-22, Manually close an open PORV during SBO.

Group Appendix NOTE: (Per NUREG-1021, Appendix D) If an operator or the Crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Page 4 of 5

Scenario #3 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 References OTG-ZZ-00003, Plant startup Not Zero Power to 30% Power - IPTE OTN-AE-00001, Feedwater system OTA-RK-00018 Addendum 33E, Pressurizer Heater Group Lockout OTO-BB-00002, RCP Off Normal OTO-AC-00001, Turbine Trip below P-9 Technical Specification 3.4.9,Pressurizer Technical Specification 3.4.4, RCS Loops - Modes 1 and 2 E-0, Reactor Trip or Safety Injection ES-0.1, Reactor Trip Response ECA-0.0, Loss of All AC Power EOP Addendum 7, Restoring Offsite Power ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions:

1. Station Blackout (19% contribution to CDF)

Page 5 of 5

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.4, Rev 0 Op-Test No.: 2020-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 80% EOC Turnover: NSAFP (Non Safety Aux Feed Pump) is out of service. The crew is to maintain current plant conditions.

Event Malf. No. Event Event No. Type* Description SRO (R) AC PT 505, HP Turbine 1st Stage Pressure Indicator, fails Low.

AC / ACPT505 = 9, RO (R) Control Rods insertion due to failure and then manual 1

ramp =30 recovery. OTO-AC-00003 entry and Technical Specification BOP (I) Determination.

SRO (C) 'A' Train Control Room HVAC Fan trips. Place 'B' Train in GK /

2 service per the normal operating procedure. Technical GK05M1TSVP=1 BOP (C) Specification Determination.

BG / SRO (I) VCT Level Channel failure low resulting in a swap over to the 3 BGLT0112=0.1 RWST. OTO-BG-00004 entry and establish letdown.

RO (I) ramp=60 GN17RELAY_D251 278TVSP=1 SRO (C) CRDM fans trip resulting in only 1 CRDM fan running.

4 OTO-GN-00002 entry and start 1 CRDM fan.

GN17RL49TVSP=1 BOP (C)

AB / AVV0045 = 100 AB / ABV0065 = 100 SRO (M)

AB / ABV0075 = 100 Steam Generator Faults, transition to E-2, Faulted SG 5 RO (M) Isolation, at E-0 step#14.

AB001_A = 250 BOP (M)

SA / SAS9XX_1=1 SA / SAS9XX_2=1 SRO (I) Failure of the automatic SL isolation, fast close MSIVs and 6

SA / SAS9XX_3=1 BOP (I) Bypass valves isolates 'A' SG fault.

SA / SAS9XX_4=1 AL / PAL02_1=1 delay =30, condition = rec0009 eq 0 AL / PAL01A_1 SRO (M) condition = Loss of Auxiliary Feedwater resulting in a Loss of Secondary 7 hwx06d55v LE 0.05 RO (M) Heat Sink, FR-H.1. RCS Bleed and Feed successful.

AL/ALHV0034_MT BOP (M)

VFAILSP = 0.05 AL/ALHV0035_MT VFAILSP = 0.05

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 5

Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes

1. Total malfunctions (5-8) 7

2. Malfunctions after EOP entry (1-2) 2

3. Abnormal events (2-4) 3

4. Major transients (1-2) 2

5. EOPs entered/requiring substantive actions (1-2) 2

6. EOP contingencies requiring substantive actions (0-2) 1

7. Critical tasks (2-3) 2 Page 2 of 5

Scenario #4 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 The Plant is stable at ~80% End of Cycle. The Non Safety Auxiliary Feed Pump (NSAFP) is out of service for bearing maintenance.

After the crew has taken the watch and at the direction of the lead evaluator, AC PT-505, HP Turbine 1st Stage Pressure Indicator, fails low. This will result in control rods stepping in. The crew may enter OTO-SF-00001 and after verifying no Main Turbine Setback or Runback is in progress, the Reactor Operator will place Control Rods in Manual and then transition to OTO-AC-00003. (Note: the crew may enter OTO-AC-00003 directly and place rod control to manual per step #1: either application of offnormal procedures is correct). The BOP will select away from the failed instrument and the RO will restore Tavg to within band by withdrawing control rods back to their original position. This failure will result in Technical Specification 3.3.1 Condition A and T not being met.

After rods are restored to Auto and Technical Specification addressed, 'A' Train Control Room HVAC fans trips. The crew should place 'B' Train Control Room HVAC in service per OTN-GK-00001. This failure will result in Technical Specification 3.7.11 Condition A not being met.

After the 'B' Train is placed in service and Technical Specification addressed, BG LT-112 fails low resulting in a partial swapover from the VCT to the RWST and a RCS boration. The crew should enter OTO-BG-00004, isolate letdown, reduce charging to seals only, and establish Excess Letdown. I&C is available to place a jumper per OTO-BG-00004 step #A1 to allow the boration to be terminated. Once BNLCV0112D, CCP A suction from RSWT ISO VLV, is closed (which effectively stops the boration event), proceed to the next event.

CRDM fans trip resulting in only 1 CRDM fan running. The crew should enter OTO-GN-00002 and start a CRDM fan to ensure a total of 2 CRDM fans are in service.

After 2 CRDM Fans are verified in service, 'A' SG develops a fault inside containment. The crew should insert a manually reactor trip, enter E-0 and transition to E-2 at step #14.

SLIS will not automatically occur and the crew should fast close all MSIVs and bypass valves per E-2 step #1 RNO. (Note: MSIVs may be fast closed during the performance of E-0 per the foldout page.) This will isolate 'A' SG fault from the other SGs.

After MSIVs and bypass valves are fast closed, a series of malfunctions ('A' & 'B' MDAFP suction clogging, TDAFP trip) occur resulting is a loss of a secondary heat sink. The crew should transition to FR-H.1, Loss of Secondary Heat Sink, due to a RED Path condition. At this point, 'D' SG will maintain pressure and level. 'B' and 'C' will being losing level (due to one faulted SG Safety Valve in each SG; specifically V075, V065, V045). Actions to restore main feedwater or establish condensate flow to 'B', 'C', and 'D' SGs will be unsuccessful. The crew will be in a "do loop" of FR-H.1 step #3 to #11 and when 3 SG Wide Range levels ('A', 'B' and 'C' SGs) are <27% [42%], they should stop all RCPs and implement RCS bleed and feed by actuating or verifying SI, verify a feed path, and open both PZR PORVS, BB HIS-455A/456A.

The scenario is complete once RCS bleed and feed is established.

Page 3 of 5

Scenario #4 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 Critical Tasks:

Critical Tasks Establish RCS Bleed Feed by opening both PZR PORVS before SGs 'B', 'C' and 'D' Isolate the faulted A SG before transition out of E-2 reach dryout conditions (10% WR level).

EVENT 7 6 Safety For the HP and LP plants, failure to initiate RCS bleed and feed results in significant and Failure to isolate a faulted SG that can be isolated causes challenges to CSFs significance sustained core uncovery. If bleed and feed is successfully initiated, then core uncovery is beyond those irreparably introduced by the postulated conditions.

prevented or minimized. For the RCS feed path to be effective, the operator should Failure to isolate a faulted SG can result in challenges to the following CSFs:

ensure that at least one charging/SI pump is injecting into the RCS and at least one high-

• Integrity head SI pump is running with valves properly aligned for maximum injection flow.

• Subcriticality

• Containment (if the break is inside containment)

Cueing

• Extreme (RED path) challenge to the heat sink CSF Both of the following:

AND

• Steam pressure and flow rate indications that make it possible to

• Indication that RCS pressure is greater than the pressure in any non-faulted identify A SG as faulted SG AND AND

• Valve position and flow rate indication that AFW continues to be

• Indication that RCS temperature is greater than the highest temperature at delivered to the faulted A SG which the RHR system can be placed in service in the shutdown cooling mode AND

• Indication and/or annunciation that no AFW is available AND

• SG level is below 7% NR Performance Manipulation of controls as required to initiate RCS bleed and feed, including stopping of ISOLATE AFW flow to faulted SG(s):

indicator all RCPs

• CLOSE associated MD AFP Flow Control Valve(s):

• RCP breaker position lights indicate breaker open o AL HK-7A (SG A)

• Circuit breaker position indication (closed) for all available charging/SI pumps

• CLOSE associated TD AFP Flow Control Valve(s):

• Flow rate indication of RCS feed from charging/SI pumps o AL HK-8A (SG A)

• Circuit breaker position indication (closed) for [all available] high-head ECCS pumps

• CLOSE Steamline Low Point Drain valve from faulted SG(s):

• Open both PZR PORVs: BB HIS-455A / 456A o AB HIS-9 (SG A)

• Flow rate indication of RCS feed from high-head ECCS pumps FAST CLOSE all MSIVs and Bypass valves:

o AB HS79 o AB HS80 Performance

• Indication that all RCPs are stopped Crew will observe the following:

feedback

• Indication of decreasing RCS pressure

• Any depressurization of intact SGs stops

• Indication of PORV flow to PRT

• AFW flow rate indication to faulted SG of zero Justification for Successful bleed-and-feed cooling of the RCS is that the core-exit vapor temperature not before transition out of E-2 is in accordance with the PWR Owners Group the chosen exceed 1200°F on the average fuel rod channel, which is considered appropriate for a Emergency Response Guidelines. It allows enough time for the crew to take the performance limit beyond-design-basis event. Analysis showed that when this criterion is met, long-term correct action while at the same time preventing avoidable adverse core cooling is sustained through RCS bleed-and-feed heat removal. Before all 3 faulted consequences.

SG reach dryout (10% WR) is an acceptable limit as it represents a loss of the secondary heat sink.

PWR Owners CT-46, Initiate RCS bleed and feed for successful ECCS injection CT-17 Isolate faulted SG Group Appendix NOTE: (Per NUREG-1021, Appendix D) If an operator or the Crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Page 4 of 5

Scenario #4 Event Description Callaway 2020-1 NRC ES-D-1, Rev. 0 References OTO-AC-00003, Turbine Impulse Pressure Channel Failure, Rev 13 OTO-GN-00002, CRDM Cooling Fan Malfunctions, Rev 6 OTN-GK-00001, Control Building HVAC system, Rev 60 OTO-BG-00004, VCT Level Channel Failures, Rev 20 Technical Specification 3.3.1,Reactor Trip System Instrumentation Technical Specification 3.7.11, Control Room Air Conditioning System (CRACS)

E-0, Reactor Trip or Safety Injection E-2, Faulted Steam Generator Isolation, Rev 11 FR-H.1, Respond to Loss of Secondary Heat Sink, Rev 18 CSF-1, Critical Safety Function Status Trees(CSFST), Rev 13 PRA Systems, Events or Operator Actions:

1. Secondary Line Breaks (10% contribution to CDF)

Top 10 Callaway Risk Important Systems - #1) Auxiliary Feedwater Top 10 Risk Reduction Operator Actions - #8) Establish RCS Feed and bleed following a loss of secondary heat sink Page 5 of 5

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Callaway Date of Exam: 8/31/2020 Operating Test No.:2020-1 A E Scenarios - Team 1: U1, R1, R2 P V P E 1@ 2 3 4# T M L N O I CREW POSITION CREW POSITION CREW POSITION CREW POSITION I T T N C S A B S A B S A B S A B I A

A T R T O R T O R T O R T O M O C P O C P O C P O C P L U N Y T P M(*)

E @ @ @

U1 R2 R1 U1 R1 R2 # # # R I U SRO-U RX 1 1 4 1 1 1 1 1 0 U1 NOR 1 1 1 1 1 1 I/C 2,3,4 3,5,6 2,4 2,3,6 2,3,4 2,3,4 3 1,2,4 10 4 4 2

,5,6 ,7 ,5,7, ,6 ,6 8

MAJ 7 7 7 5 6 5,7 5,7 5,7 2 2 2 1 TS 4,5 2,4 2,3 1,2 4 0 2 2 RO RX 4 1 1 1 0 R1 NOR 1 1 1 1 1 I/C 3,6,7 2,3,4 6 4 4 2 MAJ 5 6 2 2 2 1 TS 0 0 2 2 RO RX 4 1 1 1 0 R2 NOR 1 1 1 1 1 I/C 2,7 3,4,5 7 4 4 2

,7,8 MAJ 5 6 2 2 2 1 TS 0 0 2 2

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Callaway Date of Exam: 8/31/2020 Operating Test No.:2020-1 A E Scenarios - Team 2: I1, I2, R3 P V P E 1 2 3 4# T M L N O I CREW POSITION CREW POSITION CREW POSITION CREW POSITION I T T N C S A B S A B S A B S A B I A

A T R T O R T O R T O R T O M N Y O C P O C P O C P O C P L U T P M(*)

E I1 I2 R3 I2 I1 R3 I1 R3 I2 # # # R I U SRO-I RX 1 4 1 1 2 1 1 0 I1 NOR 1 1 1 1 1 I/C 2,3,4 2,7 2,3,4 2,3,4 3 1,2,4 13 4 4 2

,5,6 ,5,7, ,6 ,6 8

MAJ 7 5 6 5,7 5,7 5,7 3 2 2 1 TS 4,5 2,3 1,2 4 0 2 2 SRO I RX 1 4 2 1 1 0 I2 NOR 1 1 1 1 1 1 I/C 3,5,6 2,3,6 3,4,5 12 4 4 2

,7 ,7,8 MAJ 7 5 6 3 2 2 1 TS 2,4 2 0 2 2 RO RX 4 1 1 1 0 R3 NOR 1 1 1 1 1 I/C 2,4 3,6,7 2,3,4 8 4 4 2 MAJ 7 5 6 3 2 2 1 TS 0 0 2 2

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Callaway Date of Exam: 8/31/2020 Operating Test No.:2020-1 A E Scenarios - Team 3: I3, R4, R5, U2 P V P E 1 2 3 4# T M L N O I CREW POSITION CREW POSITION CREW POSITION CREW POSITION I T T N C S A B S A B S A B S A B I A

A T R T O R T O R T O R T O M N Y O C P O C P O C P O C P L U T P M(*)

E I3 R4 R5 I3 R5 R4 U2 I3 R5 # # # R I U SRO-I RX 1 4 1 1 2 1 1 0 I3 NOR 1 1 1 1 1 I/C 2,3,4 2,3,6 2,3,4 2,3,4 3 1,2,4 12 4 4 2

,5,6 ,7 ,6 ,6 MAJ 7 5 6 5,7 5,7 5,7 3 2 2 1 TS 4,5 2,4 1,2 4 0 2 2 SRO-U RX 0 1 1 0 U2 NOR 1 1 1 1 1 I/C 2,3,4 6 4 4 2

,5,7, 8

MAJ 6 1 2 2 1 TS 2,3 2 0 2 2 RO RX 1 4 2 1 1 0 R4 NOR 1 1 1 1 1 I/C 3,5,6 3,6,7 6 4 4 2 MAJ 7 5 2 2 2 1 TS 0 0 2 2 RO RX 4 1 1 1 0 R5 NOR 1 1 1 1 1 I/C 2,4 2,7 3,4,5 9 4 4 2

,7,8 MAJ 7 5 6 3 2 2 1 TS 0 0 2 2

ES-301 Transient and Event Checklist Form ES-301-5 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one-for-one basis.

3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4. For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

NOTES:

1. All 4 scenarios and their attributes are listed as they are labeled for ease of comparison. The Total Columns is summed for Scenario #1 through 3 (or 2 as appropriate for Team 1) while Scenario #4 and its attributes are shown as the spare. This in no way means that Callaway Energy Center desires Scenario #4 as the spare; specifically Callaway Energy Center would prefer the Chief Examiner to determine which scenario to designate as the spare based on the ES-D1's provided and on site validation. Callaway Energy Center will then update this ES-301-5 per NRC direction.

@Team 1 will require 2 scenarios. All 4 scenario attributes are listed with the total present for Scenario #2 and #3. This in no way means that Callaway Energy Center desires Team 1 to take Scenarios #2 and #3 and have Scenario#1 or #4 as the spare; specifically Callaway Energy Center would prefer the Chief Examiner to determine which scenarios Team 1 will take based on the ES-D1's provided and on site validation. The number of normal and reactivity events combined with the number of malfunctions will need to be reevaluated for R2. Callaway Energy Center will then update this ES-301-5 per NRC direction.