ML21137A284

1_AN2-2021-03 Draft Outlines

ML21137A284
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/08/2021
From:	Greg Werner Operations Branch IV
To:	Entergy Operations
References
Download: ML21137A284 (59)

Text

ES-401 1 Form ES-401-2 Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 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 3 4 3 3 3 2 18 6 Emergency and Abnormal Plant 2 2 1 1 N/A 1 1 N/A 3 9 4 Evolutions Tier Totals 5 5 4 4 4 5 27 10 1 4 3 3 4 1 1 3 2 2 2 3 28 5 2.

Plant 2 0 0 1 2 2 2 1 1 1 0 0 10 3 Systems Tier Totals 4 3 4 6 3 3 4 3 3 2 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.

1

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 #

000007 (EPE 7; BW E02&E10; CE E02) X 3.3 1 Reactor Trip, Stabilization, Recovery / 1 EK1.05 Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Decay power as a function of time (CFR 41.8 / 41.10 / 45.3) 000008 (APE 8) Pressurizer Vapor Space X 2.7 2 Accident / 3 AK2.02 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and detectors (CFR 41.7 / 45.7) 000009 (EPE 9) Small Break LOCA / 3 X 3.0 3 EK2.03 Knowledge of the interrelations SRO between the small break LOCA and the following: S/Gs (CFR 41.7 / 45.7) 000011 (EPE 11) Large Break LOCA / 3 X 4.3 4 EA1.09 Ability to operate and monitor the following as they apply to a Large Break LOCA: Core flood tank initiation (CFR 41.7 / 45.5 / 45.6) 000015 (APE 15) Reactor Coolant Pump X 4.1 5 AK3.07 Knowledge of the reasons for the Malfunctions / 4 following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow) : Ensuring that S/G levels are controlled properly for natural circulation enhancement (CFR 41.5,41.10 / 45.6 / 45.13) 000022 (APE 22) Loss of Reactor Coolant X Makeup / 2 AA2.01 Ability to determine and interpret the 3.2 6 following as they apply to the Loss of Reactor Coolant Makeup: Whether charging line leak exists (CFR 43.5/ 45.13) 000025 (APE 25) Loss of Residual Heat Removal System / 4 SRO 000026 (APE 26) Loss of Component X 3.6 7 Cooling Water / 8 AA1.07 Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: Flow rates to the components and systems that are serviced by the CCWS; interactions among the components (CFR 41.7 / 45.5 / 45.6) 2

ES-401 3 Form ES-401-2 000027 (APE 27) Pressurizer Pressure X 3.2 8 Control System Malfunction / 3 AA2.05 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: PZR heater setpoints (CFR: 43.5 / 45.13) 000029 (EPE 29) Anticipated Transient X 2.9 9 Without Scram / 1 EK2.06 Knowledge of the interrelations between the and the following an ATWS:

SRO Breakers, relays, and disconnects (CFR 41.7 / 45.7) 000038 (EPE 38) Steam Generator Tube X 4.1 10 Rupture / 3 EK3.01 Knowledge of the reasons for the following responses as the apply to the SRO SGTR: Equalizing pressure on primary and secondary sides of ruptured S/G (CFR 41.5 / 41.10 / 45.6 / 45.13) 000040 (APE 40; BW E05; CE E05; W E12) X 3.7 11 Steam Line RuptureExcessive Heat AK1.06 Knowledge of the operational Transfer / 4 implications of the following concepts as they apply to Steam Line Rupture: High-energy steam line break considerations (CFR 41.8 / 41.10 / 45.3) 000054 (APE 54; CE E06) Loss of Main X 4.6 12 2.1.20 Ability to interpret and execute Feedwater /4 procedure steps.

(CFR: 41.10 / 43.5 / 45.12) 000055 (EPE 55) Station Blackout / 6 SRO 000056 (APE 56) Loss of Offsite Power / 6 X AA1.30 Ability to operate and / or monitor the 3.5 13 following as they apply to the Loss of Offsite Power: AFW flow control valve operating switches (CFR 41.7 / 45.5 / 45.6) 000057 (APE 57) Loss of Vital AC X 4.1 14 Instrument Bus / 6 AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus (CFR 41.5,41.10 / 45.6 / 45.13) 000058 (APE 58) Loss of DC Power / 6 X 2.8 15 AK1.01 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation (CFR 41.8 / 41.10 / 45.3) 3

ES-401 4 Form ES-401-2 000062 (APE 62) Loss of Nuclear Service X 2.5 16 Water / 4 AA2.04 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The normal values and upper limits for the temperatures of the components cooled by SWS (CFR: 43.5 / 45.13) 000065 (APE 65) Loss of Instrument Air / 8 X 4.4 17 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) 000077 (APE 77) Generator Voltage and X 3.6 18 AK2.07 Knowledge of the interrelations Electric Grid Disturbances / 6 between Generator Voltage and Electric Grid SRO Disturbances and the following: Turbine /

generator control (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

NA (W E04) LOCA Outside Containment / 3 NA (W E11) Loss of Emergency Coolant Recirculation / 4 NA (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals: 3 4 3 3 3 2 Group Point Total: 18 4

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 SRO 000005 (APE 5) Inoperable/Stuck Control Rod / 1 X 3.1 19 AK1.01 Knowledge of the operational implications of the following concepts as they apply to Inoperable / Stuck Control Rod: Axial power imbalance (CFR 41.8 / 41.10 / 45.3) 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control X 4.2 20 Malfunction / 2 2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12) 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 SRO 000033 (APE 33) Loss of Intermediate Range Nuclear System 033 rejected (See Instrumentation / 7 attached ES-401-4) for original QID #20. System 028 above selected for a replacement with same K/A 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 SRO 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 X 3.9 21 AA2.02 Ability to determine and 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 X 3.9 22 2.1.30 Ability to locate and operate components, including local controls.

(CFR: 41.7 / 45.7) 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 5

ES-401 6 Form ES-401-2 000061 (APE 61) Area Radiation Monitoring System Alarms X 4.1 23

/7 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 /

45.12) 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 X 2.9 24 AK2.03 Knowledge of the interrelations between the Control Room Evacuation and the following: Controllers and positioners (CFR 41.7 / 45.7) 000069 (APE 69; W E14) Loss of Containment Integrity / 5 SRO 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling / X 3.8 25 4 EA1.25 Ability to operate and monitor the following as they apply to Inadequate Core Cooling: Atmospheric dump valve controllers and indicators (CFR 41.7 / 45.5 / 45.6) 000076 (APE 76) High Reactor Coolant Activity / 9 X 2.9 26 AK3.05 Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity: Corrective actions as a result of high fission-product radioactivity level in the RCS (CFR 41.5,41.10 / 45.6 /

45.13) 000078 (APE 78*) RCS Leak / 3 NA (W E01 & E02) Rediagnosis & SI Termination / 3 NA (W E13) Steam Generator Overpressure / 4 NA (W E15) Containment Flooding / 5 NA (W E16) High Containment Radiation /9 NA (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 NA (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 6

ES-401 7 Form ES-401-2 (CE A16) Excess RCS Leakage / 2 X 3.0 27 AK1.2 Knowledge of the operational implications of the following concepts as they apply to the (Excess RCS Leakage): Normal, abnormal and emergency operating procedures associated with Excess RCS Leakage.

(CFR: 41.8 / 41.10 / 45.3)

(CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 NA K/A Category Point Totals: 2 1 1 1 1 3 Group Point Total: 9 7

ES-401 8 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 #

3 003 (SF4P RCP) Reactor Coolant X A4.06 Ability to manually operate and/or 2.9 28 Pump monitor in the control room: RCP X2 parameters (CFR: 41.7 / 45.5 to 45.8) 003 (SF4P RCP) Reactor Coolant X 2.5 29 K4.03 Knowledge of RCPS design Pump feature(s) and/or interlock(s) which X2 provide for the following: Adequate lubrication of the RCP (CFR: 41.7) 004 (SF1; SF2 CVCS) Chemical and X 2.7 30 K1.19 Knowledge of the physical Volume Control connections and/or cause-effect X2 relationships between the CVCS and the following systems: Primary grade water supply (CFR: 41.2 to 41.9 / 45.7 to 45.8) 004 (SF1; SF2 CVCS) Chemical and X 2.6 31 K5.50 Knowledge of the operational Volume Control implications of the following concepts as X2 they apply to the CVCS: Design basis letdown system temperatures: resin integrity (CFR: 41.5/45.7) 005 (SF4P RHR) Residual Heat X K2.03 Knowledge of bus power supplies 2.7 32 Removal to the following: RCS pressure boundary SRO motor-operated valves (CFR: 41.7) 006 (SF2; SF3 ECCS) Emergency X A2.13 Ability to (a) predict the impacts of 3.9 33 Core Cooling the following malfunctions or operations X2 on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadvertent SIS actuation (CFR: 41.5 / 45.5) 006 (SF2; SF3 ECCS) Emergency X 2.6 34 K6.10 Knowledge of the effect of a loss Core Cooling or malfunction on the following will have X2 on the ECCS: Valves (CFR: 41.7 / 45.7) 8

ES-401 9 Form ES-401-2 007 (SF5 PRTS) Pressurizer X K3.01 Knowledge of the effect that a loss 3.3 35 Relief/Quench Tank or malfunction of the PRTS will have on the following: Containment (CFR: 41.7 / 45.6) 008 (SF8 CCW) Component Cooling X 3.6 36 A3.08 Ability to monitor automatic Water operation of the CCWS, including:

X2 Automatic actions associated with the CCWS that occur as a result of a safety injection signal (CFR: 41.7 / 45.5) 008 (SF8 CCW) Component Cooling X 3.1 37 A1.04 Ability to predict and/or monitor Water changes in parameters (to prevent X2 exceeding design limits) associated with operating the CCWS controls including:

Surge tank level (CFR: 41.5 / 45.5) 010 (SF3 PZR PCS) Pressurizer X 2.7 38 K4.01 Knowledge of PZR PCS design Pressure Control feature(s) and/or interlock(s) which provide for the following: Spray valve warm-up (CFR: 41.7) 012 (SF7 RPS) Reactor Protection X K2.01 Knowledge of bus power supplies 3.3 39 SRO to the following: RPS channels, components, and interconnections (CFR: 41.7) 013 (SF2 ESFAS) Engineered X 3.9 40 2.4.39 Knowledge of RO responsibilities Safety Features Actuation in emergency plan implementation.

SRO (CFR: 41.10 / 45.11) 022 (SF5 CCS) Containment Cooling X 4.1 41 A3.01 Ability to monitor automatic X2 operation of the CCS, including: Initiation of safeguards mode of operation (CFR: 41.7 / 45.5) 022 (SF5 CCS) Containment Cooling X 3.5 42 K1.01 Knowledge of the physical X2 connections and/or cause-effect relationships between the CCS and the following systems: SWS/cooling system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 025 (SF5 ICE) Ice Condenser NA 9

ES-401 10 Form ES-401-2 026 (SF5 CSS) Containment Spray X 3.8 43 K4.07 Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following: Adequate level in containment sump for suction (interlock)

(CFR: 41.7) 039 (SF4S MSS) Main and Reheat X 4.4 44 2.1.7 Ability to evaluate plant Steam performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) 059 (SF4S MFW) Main Feedwater X 2.7 45 A1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including:

Power level restrictions for operation of MFW pumps and valves (CFR: 41.5 / 45.5) 061 (SF4S AFW) X A2.03 Ability to (a) predict the impacts of 3.2 46 Auxiliary/Emergency Feedwater the following malfunctions or operations X2 on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of dc power (CFR: 41.5 / 43.5 / 45.3 / 45.13) 061 (SF4S AFW) X 3.9 47 A3.03 Ability to monitor automatic Auxiliary/Emergency Feedwater operation of the AFW, including: AFW X2 S/G level control on automatic start (CFR: 41.7 / 45.5) 062 (SF6 ED AC) AC Electrical X 2.6 48 2.2.17 Knowledge of the process for Distribution managing maintenance activities during SRO power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

(CFR: 41.10 / 43.5 / 45.13) 063 (SF6 ED DC) DC Electrical X 2.5 49 A1.01 Ability to predict and/or monitor Distribution changes in parameters associated with operating the DC electrical system controls including: Battery capacity as it is affected by discharge rate (CFR: 41.5 / 45.5) 10

ES-401 11 Form ES-401-2 064 (SF6 EDG) Emergency Diesel X A4.01 Ability to manually operate and/or 4.0 50 Generator monitor in the control room: Local and SRO remote operation of the ED/G (CFR: 41.7 / 45.5 to 45.8) 073 (SF7 PRM) Process Radiation X 3.6 51 K1.01 Knowledge of the physical Monitoring connections and/or cause-effect relationships between the PRM system and the following systems: Those systems served by PRMs (CFR: 41.2 to 41.9 / 45.7 to 45.8) 076 (SF4S SW) Service Water X K2.01 Knowledge of bus power supplies 2.7 52 to the following: Service water (CFR: 41.7) 078 (SF8 IAS) Instrument Air X 3.2 53 K4.02 Knowledge of IAS design X2 feature(s) and/or interlock(s) which provide for the following: Cross-over to other air systems (CFR: 41.7) 078 (SF8 IAS) Instrument Air X K3.02 Knowledge of the effect that a loss 3.4 54 X2 or 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 2.8 55 K1.05 Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems:

Personnel access hatch and emergency access hatch (CFR: 41.2 to 41.9 / 45.7 to 45.8) 053 (SF1; SF4P ICS*) Integrated NA Control K/A Category Point Totals: 4 3 3 4 1 1 3 2 2 2 3 Group Point Total: 28 11

ES-401 12 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 002 (SF2; SF4P RCS) Reactor X 2.5 56 Coolant K6.06 Knowledge of the effect or a loss or malfunction on the following RCS components: Sensors and Detectors (CFR: 41.7 / 45.7 011 (SF2 PZR LCS) Pressurizer X 3.7 57 Level Control K5.10 Knowledge of the operational implications of the following concepts as they apply to the PZR LCS: Indications of reactor vessel bubble (CFR: 41.5 / 45.7) 014 (SF1 RPI) Rod Position SRO Indication 015 (SF7 NI) Nuclear X 3.1 58 Instrumentation K6.04 Knowledge of the effect of a loss or malfunction on the following will have on the NIS: Bistables and logic circuits (CFR: 41.7 / 45.7) 016 (SF7 NNI) Nonnuclear X 2.8 59 Instrumentation K4.01 Knowledge of NNIS design feature(s) and/or interlock(s) which provide for the following: Reading of NNIS channel values outside control room (CFR: 41.7) 017 (SF7 ITM) In-Core Temperature SRO Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen X 3.4 60 Recombiner and Purge Control K5.01 Knowledge of the operational implications of the following concepts as they apply to the HRPS: Explosive hydrogen concentration (CFR: 41.5 / 45.7) 029 (SF8 CPS) Containment Purge X 3.8 61 A3.01 Ability to monitor automatic operation of the Containment Purge System including: CPS isolation (CFR: 41.7 / 45.5) 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling SRO Equipment 12

ES-401 13 Form ES-401-2 035 (SF 4P SG) Steam Generator X 3.6 62 A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the S/GS controls including:

S/G wide and narrow range level during startup, shutdown, and normal operations (CFR: 41.5 / 45.5) 041 (SF4S SDS) Steam X 2.8 63 Dump/Turbine Bypass Control K4.11 Knowledge of SDS design feature(s) and/or interlock(s) which provide for the following: T-ave./T-ref.

program (CFR: 41.7) 045 (SF 4S MTG) Main Turbine X 2.9 64 Generator K3.01 Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following: Remainder of the plant (CFR: 41.7 / 45.6) 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate X 2.6 65 A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of condensate pumps (CFR: 41.5 / 43.5 / 45.3 / 45.13) 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room NA Ventilation K/A Category Point Totals: 0 0 1 2 2 2 1 1 1 0 0 Group Point Total: 10 13

ES-401 Generic Knowledge and Abilities Outline (Tier 3) RO Form ES-401-3 Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 Category K/A # Topic RO SRO-only IR # IR #

Ability to verify the controlled procedure copy.

2.1.21 3.5 66 (CFR: 41.10 / 45.10 / 45.13) 2.1.26 Knowledge of industrial safety procedures (such as 3.4 67 rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

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

Knowledge of administrative requirements for 2.1.15 2.7 68 temporary management directives, such as standing orders, night orders, Operations memos, etc.

(CFR: 41.10 / 45.12)

Subtotal 3 Knowledge of tagging and clearance procedures.

2.2.13 4.1 69 (CFR: 41.10 / 45.13)

Knowledge of the process for controlling equipment 2.2.14 3.9 70 configuration or status.

2. Equipment (CFR: 41.10 / 43.3 / 45.13)

Control 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)

Knowledge of radiological safety principles pertaining

3. Radiation 2.3.12 3.2 73 to licensed operator duties, such as containment entry Control requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 / 45.9 / 45.10)

Subtotal 2 Knowledge of fire protection procedures.

2.4.25 3.3 74 (CFR: 41.10 / 43.5 / 45.13)

4. Emergency 2.4.3 Ability to identify post-accident instrumentation. 3.7 75 Procedures/Plan (CFR: 41.6 / 45.4)

Subtotal 2 Tier 3 Point Total 10 14

ES-401 15 Form ES-401-2 Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 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 3 1 4 Evolutions Tier Totals 27 6 4 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.

15

ES-401 16 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 EA2.01 Ability to determine or interpret the 000009 (EPE 9) Small Break LOCA / 3 X 4.8 76 following as they apply to a small break LOCA: Actions to be taken, based on RCS temperature and pressure, saturated and superheated (CFR 43.5 / 45.13) 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 X AA2.06 Ability to determine and interpret the 3.4 77 Removal System / 4 following as they apply to the Loss of Residual Heat Removal System: Existence of proper RHR overpressure protection (CFR: 43.5 / 45.13) 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient X 4.5 78 2.2.38 Knowledge of conditions and Without Scram / 1 limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13) 000038 (EPE 38) Steam Generator Tube X Rupture / 3 2.4.6 Knowledge of EOP mitigation strategies. 4.7 79 (CFR: 41.10 / 43.5 / 45.13) 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 X 4.0 80 2.1.32 Ability to explain and apply system limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 16

ES-401 17 Form ES-401-2 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and X 3.8 81 Electric Grid Disturbances / 6 AA2.05 Ability to determine and interpret the 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 NA (W E11) Loss of Emergency Coolant NA Recirculation / 4 (BW E04; W E05) Inadequate Heat NA TransferLoss of Secondary Heat Sink / 4 K/A Category Totals: 3 3 Group Point Total: 6 17

ES-401 18 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 X 4.7 82 2.4.6 Knowledge of EOP mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear X 3.4 83 Instrumentation / 7 AA2.06 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: -

Confirmation of reactor trip (CFR: 43.5 / 45.13) 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 X 4.2 84 AA2.03 Ability to determine and interpret the following as they apply to the Fuel Handling Incidents: Magnitude of potential radioactive release (CFR: 43.5 / 45.13) 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms

/7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 X 4.3 85 AA2.01 Ability to determine and interpret the following as they apply to the Loss of Containment Integrity: Loss of containment integrity (CFR: 43.5 / 45.13) 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /

4 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 NA (W E01 & E02) Rediagnosis & SI Termination / 3 NA 18

ES-401 19 Form ES-401-2 (W E13) Steam Generator Overpressure / 4 NA (W E15) Containment Flooding / 5 NA (W E16) High Containment Radiation /9 NA (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 NA (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 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 NA K/A Category Point Totals: 3 1 Group Point Total: 4 19

ES-401 20 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 Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control 005 (SF4P RHR) Residual Heat X A2.01 Ability to (a) predict the impacts of 2.9 86 Removal the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation (CFR: 41.5 / 43.5 / 45.3 / 45.13) 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection X A2.05 Ability to (a) predict the impacts of 3.2 87 the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty or erratic operation of detectors and function generators (CFR: 41.5 / 43.5 / 45.3 / 45.5) 013 (SF2 ESFAS) Engineered X 4.2 88 2.2.36 Ability to analyze the effect of Safety Features Actuation maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13) 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser NA 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 20

ES-401 21 Form ES-401-2 061 (SF4S AFW)

Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical X 4.7 89 2.2.22 Knowledge of limiting conditions Distribution for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2) 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel X A2.03 Ability to (a) predict the impacts of 3.1 90 Generator the 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: Parallel operation of ED/Gs (CFR: 41.5 / 43.5 / 45.3 / 45.13) 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated NA Control K/A Category Point Totals: 3 2 Group Point Total: 5 21

ES-401 22 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 X 2.9 91 Indication A2.07 Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of reed switch (CFR: 41.5 / 43.5 / 45.3 / 45.13) 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature X 4.7 92 Monitor 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) 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 4.0 93 Equipment A2.03 Ability to (a) predict the impacts of the 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 22

ES-401 23 Form ES-401-2 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room NA Ventilation K/A Category Point Totals: 2 1 Group Point Total: 3 23

ES-401 Generic Knowledge and Abilities Outline (Tier 3) SRO Form ES-401-3 Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 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 Knowledge of the fuel-handling responsibilities of SROs.

2.1.35 3.9 95 (CFR: 41.10 / 43.7)

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

(CFR: 41.10 / 43.3 / 45.13)

2. Equipment 2.2.20 Knowledge of the process for managing Control 2.2.20 3.8 97 troubleshooting activities.

(CFR: 41.10 / 43.5 / 45.13)

Subtotal 2 Ability to control radiation releases.

2.3.11 4.3 98

3. Radiation Control (CFR: 41.11 / 43.4 / 45.10)

Subtotal 1 Knowledge of the lines of authority during 2.4.37 4.1 99 implementation of the emergency plan.

(CFR: 41.10 / 45.13)

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

(CFR: 41.10 / 43.5 / 45.13)

Subtotal 2 Tier 3 Point Total 7 24

ES-401 Record of Rejected K/As Form ES-401-4 ANO Unit 2 April 2021 RO/SRO Exam Record of Rejected K/As Tier / Randomly Selected K/A Reason for Rejection Group RO Tier 1 026 AA1.03 (Original) Need to reject the original K/A for the following Group 1 Loss of Component Cooling Water reasons: The Service Water System (SWS) does not (CCW) provide a backup to any component cooled by CCW.

QID# 7 The SWS does provide cooling to the CCW System Ability to operate and/or monitor the Heat Exchangers but a question on this knowledge following as they apply to the Loss of would overlap another QID (#36) on the RO exam.

Component Cooling Water: - SWS as a backup to the CCWS We recommend K/A 026 AA1.07 as a replacement as 026 AA1.07 (New the flow of CCW to the RCPs inside Containment is Recommendation) vital to plant operations.

Loss of Component Cooling Water (CCW Ability to operate and/or monitor the following as they apply to the Loss of Component Cooling Water: - Flow NRC APPROVED REPLACEMENT ON 09/14/2020 rates to the components and systems that are serviced by the CCWS; interactions among the components Tier / Randomly Selected K/A Reason for Rejection Group RO Tier 1 033 2.2.44 (Original) Need to reject the original 033 system for the following Group 2 Loss of Intermediate Range reasons: ANO Unit 2 does not have intermediate Nuclear Instrumentation range Nuclear Instrumentation. Our normal Power QID# 20 Nuclear Instruments are (3) stacked fission chambers Equipment Control - Ability to with the middle chamber being a Log channel detector interpret control room indications to that monitors the intermediate power range during a verify the status and operation of a Reactor Startup. I have another RO QID (#58) that system and understand how operator actions and directives affect plant requires knowledge of the Log Power Channel and system conditions. outputs. I also have an SRO QID (83) on the Loss of Source Range Nuclear Instrument - System 032 028 2.2.44 (New Recommendation)

Therefore, we recommend K/A 028 2.2.44 as a Pressurizer Level Control replacement as the Pressurizer Level Control Malfunction Malfunction system 028 has not been used and has a Equipment Control - Ability to lot of Control Room Indications that must be interpret control room indications to interpreted accurately during a failure of the system.

verify the status and operation of a system and understand how operator actions and directives affect plant NRC APPROVED REPLACEMENT ON 09/14/2020 and system conditions.

25

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Selected K/A Reason for Rejection Group RO Tier 1 059 2.1.19 (Original) Need to reject the original K/A for the following Group 2 Accidental Liquid Radwaste reasons: We do not use the Plant Computer system to Release monitor a Liquid Rad Waste release. We use alarms, QID# 22 trending Chart recorders and local status of Conduct of Operations - Ability to use pumps/valves to monitor and evaluate the releases.

plant computers to evaluate system or component status. We recommend K/A 059 2.1.30 as a replacement as a replacement for this QID because several alarms, 059 2.1.30 (New Recommendation) indications and controls for liquid rad waste releases Accidental Liquid Radwaste are local with a common trouble alarm in the control Release room.

Conduct of Operations - Ability to locate and operate components, including local controls. NRC APPROVED REPLACEMENT ON 09/14/2020 Tier / Randomly Selected K/A Reason for Rejection Group RO Tier 2 061 A2.02 (Original) Need to reject the original K/A for the following Group 1 Auxiliary / Emergency Feedwater reasons: The Steam supply valves for the steam (AFW) System driven EFW 2P-7A Pump Turbine are motor operated QID# 46 valves MOVs. The turbine startup valves are also Ability to (a) predict the impacts of MOVs thus a Loss of air has no effect on the startup of the following malfunctions or the steam driven EFW 2P-7A Pump Turbine.

operations on the AFW System and (b) based on those predictions, use We recommend K/A 061 A2.03 as a replacement as procedures to correct, control, or the startup MOVs for the steam driven EFW 2P-7A mitigate the consequences of those Pump Turbine are powered form DC and this power is malfunctions or operations: - Loss of needed to start the steam driven EFW 2P-7A Pump air to steam supply valve Turbine.

061 A2.03 (New Recommendation)

Ability to (a) predict the impacts of the following malfunctions or operations on the AFW System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Loss of NRC APPROVED REPLACEMENT ON 08/24/2020 dc power.

26

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Selected K/A Reason for Rejection Group SRO Tier 1 032 AA2.07 (Original) Need to reject the original K/A for the following Group 2 Loss of Source Range Nuclear reasons. The Maximum allowable channel Instrumentation disagreement is performed by I&C maintenance during QID# 83 the Source Range NI Functional testing that is Ability to determine and interpret the required to be completed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to Core following as they apply to the Loss of Alterations. This is not an item the RO or SRO would Source Range Nuclear check or see in real time. They would verify that the Instrumentation: - Maximum allowable channel disagreement instruments have power and are reading out properly.

I do not believe I can write a discriminatory SRO Only question since there is no AOP/EOP associated with 032 AA2.01 (1st Recommendation) the detector channel disagreement.

Loss of Source Range Nuclear We recommend K/A 032 AA2.01 as a replacement as Instrumentation this is something the SRO can see in real time and provide the appropriate direction given a set of plant Ability to determine and interpret the conditions associated with Fuel handling activities and following as they apply to the Loss of Source Range Nuclear procedures.

Instrumentation: - Normal/abnormal NRC APPROVED 1st RECOMMENDED power supply operation REPLACEMENT ON 07/23/2020 032 AA2.06 (2nd Recommendation)

Need to reject the 1st recommended K/A change due Loss of Source Range Nuclear the inability to generate a discriminatory SRO Only Instrumentation level question for K/A 032 AA2.01.

Ability to determine and interpret the We recommend K/A 032 AA2.06 as a replacement as following as they apply to the Loss of an SRO needs to know the correct procedure section Source Range Nuclear to implement if confirmation of a reactor trip CANNOT Instrumentation: - Confirmation of reactor trip be confirmed in SPTAs.

27

ES-301 Administrative Topics Outline Form ES-301-1 Facility: ANO-2 Date of Examination: 3/29/2021 Examination Level: RO SRO Operating Test Number: 2021 Administrative Topic (see Note) Type Describe activity to be performed Code*

D/R Determine CEA#1 Upper Gripper Coil A1. Conduct of Operations Temperature 2.1.23 RO (4.3)

A2JPM-NRC-ADMIN-CEA N/R Determine acceptable control room loading.

A2. Conduct of Operations 2.1.9 RO (2.9) A2JPM-NRC-ADMIN-CRLD A3. Equipment Control P/D/R Perform identification of boundary isolations and electrical power to tagout a Boric Acid 2.2.15 RO (3.9) Makeup Pump A2JPM-NRC-ADMIN-HCRD2 D/R Calculate Containment Purge Release Setpoint A4. Radiation Control 2.3.11 RO (3.8) A2JPM-NRC-ADMIN-PURGE1 Emergency Plan 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)

Revision 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: ANO-2 Date of Examination: 3/29/2021 Examination Level: RO SRO Operating Test Number: 2021 Administrative Topic (see Note) Type Describe activity to be performed Code*

D/R Review CEA#1 Upper Gripper Coil A5. Conduct of Operations Temperature Calculation 2.1.23 SRO (4.4)

A2JPM-NRC-ADMIN-XCEA P/D/R Verify RPS trip set point determination for inoperable MSSV A6. Conduct of Operations A2JPM-NRC-ADMIN-MSSVINOP 2.1.25 SRO (4.2)

M/R Determine operability of Emergency Feedwater A7. Equipment Control System 2.2.37 SRO (4.6)

A2JPM-NRC-ADMIN-EFWTS3 D/R Calculate expected dose for entry during an A8. Radiation Control emergency and determine if entry is allowed.

2.3.4 SRO (3.7)

A2JPM-NRC-ADMIN-EMGRESPSRO N/R Determine Emergency Action Level, Time A9. Emergency Plan Critical (Rev 6 EAL) 2.4.41 SRO (4.6)

A2JPM-NRC-ADMIN-EAL16 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)

Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: ANO-2 Date of Examination: _3/29/2021___

Exam Level: RO X SRO-I SRO-U Operating Test No.: 2021 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1. A2JPM-NRC-CVCS2 1 004 A4.07; RO 3.9 / SRO 3.7 A/M/EN/L/S Perform Emergency Boration Reactivity Control S2. A2JPM-NRC-SDBC1 4 041 A4.05; RO 3.1 / SRO 3.3 D/L/S Heat Removal Perform a restart and reset of SDBCS after power interruption Secondary S3. A2JPM-NRC-CSASV 5 026 A4.01; RO 4.5 / SRO 4.3 A/N/EN/L/S Verify CSAS Attachment 41 Containment S4. A2JPM-NRC-ELECXT5 6 062 A4.01; RO 3.3 / SRO 3.1 D/S Cross Connect 2B3 and 2B4 Electrical S5. A2JPM-RO-RCP04 4 003 A2.02; RO-3.7 / SRO-3.9 A/P/D/L/S Heat Removal Perform a normal RCP shutdown Primary S6. A2JPM-RO-SIT02 3 006 A1.07; RO 3.3 / SRO 3.6 D/EN/S Pressure Control Adjust SIT 2T-2A Pressure (Raise)

S7. A2JPM-NRC-CCW04 8 008 A4.01; RO 3.3 / SRO 3.1 N/S Splitting out loop 1 and 2 CCW Plant Service Systems S8.. A2JPM-NRC-CEA02 7 012 A4.06; RO 4.3 / SRO 4.3 D/S Test a Reactor Trip Circuit Breaker Instrumentation Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. A2JPM-NRC-IA04 8 065 AA2.01; RO-2.9 / SRO-3.2 A/P/D/E Respond to lowering Instrument Air Pressure Plant Service Systems P2. A2JPM-NRC-2D31BSWAP2 6 063 A3.01: RO-2.7 / SRO 3.1- A/L/M Align 2D-31B Battery Charger to Green Train and place it in service Electrical Distribution P3. A2JPM-NRC-69REL2 P/D/R 9 2.3.11; RO-3.8 / SRO-4.3 Perform a release of 2T-69A Boric Acid Condensate Tank Radioactivity Release

• 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; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U Revision 0

(A)lternate path (5) 4-6 / 4-6 / 2-3 (C)ontrol room (0)

(D)irect from bank (7) 9/8/4 (E)mergency or abnormal in-plant (1) 1/1/1 (EN)gineered safety feature (3) 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown (5) 1/1/1 (N)ew or (M)odified from bank including 1(A) (4) 2/2/1 (P)revious 2 exams (3) 3 / 3 / 2 (randomly selected)

(R)CA (1) 1/1/1 (S)imulator (8)

Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: ANO-2 Date of Examination: _3/29/2021_

Exam Level: RO SRO-I X SRO-U Operating Test No.: 2021 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1. A2JPM-NRC-CVCS2 1 004 A4.07; RO 3.9 / SRO 3.7 A/M/EN/L/S Perform Emergency Boration Reactivity Control S2. A2JPM-NRC-SDBC1 4 041 A4.05; RO 3.1 / SRO 3.3 D/L/S Heat Removal Perform a restart and reset of SDBCS after power interruption Secondary S3. A2JPM-NRC-CSASV 5 026 A4.01; RO 4.5 / SRO 4.3 A/N/EN/L/S Verify CSAS Attachment 41 Containment S4. A2JPM-NRC-ELECXT5 6 062 A4.01; RO 3.3 / SRO 3.1 D/S Cross Connect 2B3 and 2B4 Electrical S5. A2JPM-RO-RCP04 4 003 A2.02; RO-3.7 / SRO-3.9 A/P/D/L/S Heat Removal Perform a normal RCP shutdown Primary S6. A2JPM-RO-SIT02 3 006 A1.07; RO 3.3 / SRO 3.6 D/EN/S Adjust SIT 2T-2A Pressure Pressure Control S7. A2JPM-NRC-CCW03 8

008 A4.01; RO 3.3 / SRO 3.1 N/S Splitting out loop 1 and 2 CCW Plant Service Systems Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. A2JPM-NRC-IA04 8 065 AA2.01; RO-2.9 / SRO-3.2 A/P/D/E Respond to lowering Instrument Air Pressure Plant Service Systems P2. A2JPM-NRC-2D31BSWAP2 6 063 A3.01: RO-2.7 / SRO 3.1- A/L/M Align 2D-31B Battery Chargers to Green Train and place it in service Electrical Distribution P3. A2JPM-NRC-69REL2 P/D/R 9 2.3.11; RO-3.8 / SRO-4.3 Perform a release of 2T-69A Boric Acid Condensate Tank Radioactivity Release

• 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; in-plant systems and functions may overlap those tested in the control room.

Revision 0

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (5) 4-6 / 4-6 / 2-3 (C)ontrol room (0)

(D)irect from bank (6) 9/8/4 (E)mergency or abnormal in-plant (1) 1/1/1 (EN)gineered safety feature (3) 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown (5) 1/1/1 (N)ew or (M)odified from bank including 1(A) (4) 2/2/1 (P)revious 2 exams (3) 3 / 3 / 2 (randomly selected)

(R)CA (1) 1/1/1 (S)imulator (7)

Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: ANO-2 Date of Examination: _3/29/2021__

Exam Level: RO SRO-I SRO-U X Operating Test No.: 2021 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1. A2JPM-NRC-CVCS2 1 004 A4.07; RO 3.9 / SRO 3.7 A/M/EN/L/S Perform Emergency Boration Reactivity Control S2. A2JPM-NRC-SDBC1 4 041 A4.05; RO 3.1 / SRO 3.3 D/L/S Heat Removal Perform a restart and reset of SDBCS after power interruption Secondary S3.

S4.

S5.

S6.

S7.

S8.

Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. A2JPM-NRC-IA04 8 065 AA2.01; RO-2.9 / SRO-3.2 A/P/D/E Respond to lowering Instrument Air Pressure Plant Service Systems P2. A2JPM-NRC-2D31BSWAP2 6 063 A3.01: RO-2.7 / SRO 3.1- A/L/M Align 2D-31B Battery Chargers to Green Train and place it in service Electrical Distribution P3. A2JPM-NRC-69REL2 P/D/R 9 2.3.11; RO-3.8 / SRO-4.3 Perform a release of 2T-69A Boric Acid Condensate Tank Radioactivity Release

• 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; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (3) 4-6 / 4-6 / 2-3 (C)ontrol room (0)

(D)irect from bank (3) 9/8/4 (E)mergency or abnormal in-plant (1) 1/1/1 (EN)gineered safety feature (1) 1 / 1 / 1 (control room system)

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

(R)CA (1) 1/1/1 (S)imulator (2)

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Appendix D Scenario # 2 Form ES-D-1 Facility: ANO-2 Scenario No.: 2 Op-Test No.: 2021 Examiners: ___________________________ Operators: ____________________________

Initial Conditions: 100%, MOL, Red Train Maintenance Week.

Turnover: 260 EFPD. EOOS indicates Minimal Risk. Red Train Maintenance Week.

Scheduled evolution: None.

Critical Tasks: Isolate A SG (2202.010 Attachment 10 completed) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the Reactor trip, Establish and then maintain RCS subcooling, and Secure A RCP within 10 min of the reactor trip.

Event Malf. No. Event Type* Event No. Description 1 NIBLOGPWR C (BOP / CRS) B channel Log power fails low then fails high.

TS (CRS) OP-2203.026, NI malfunction AOP.

2 K12-C05 C (ATC / CRS) B Boric Acid Makeup Tank Temperature Lo OP-2203.012L, Annunciator 2K12 Corrective Action 3 SGATUBE R (ATC) Shutdown required due to 4 gpm steam generator tube leak on A steam generator.

C (CRS / BOP)

OP-2203.038 Primary to Secondary Leakage AOP.

TS (CRS) 4 CNDVACPPA C (BOP / CRS) 2C-5A Vacuum pump breaker trip and 2C-5B Vacuum CND2C5B pump failure to auto start.

CNDAIRLEAKHI OP-2203.019 Loss of Condenser Vacuum AOP.

OP-2203.012C, Annunciator 2K03 Corrective Action 5 RCP2P32ALOS C (ATC / CRS) A RCP oil leak. CT-1 OP-2203.025, RCP Emergencies AOP.

6 SGBTUBE M (ALL) A Steam Generator Tube Rupture of 275 gpm. CT-2 OP-2202.001, Standard Post Trip Actions (SPTAs),

OP-2202.004, Steam Generator Tube Rupture.

7 CV0231 C (BOP / CRS) Gland seal regulator 2PCV-0231 fails closed.

2203.012B, Annunciator 2K-02 Corrective Action (ACA) 8 CV48202 C (ATC / CRS) Letdown fails to isolate on SIAS.

CV48211 OP-2202.010, Standard Attachments EOP.

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Appendix D Scenario # 2 Form ES-D-1 Target Quantitative Attributes (Section D.5.d) Actual Attributes Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 4 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per scenario 0 set)

Critical Tasks (2) 2 Critical Task Justification Cueing Measurable Performance References Performance Feedback Indicators CT-1: Exceeding 2K11 F1 A Secured A A RCP amps EN-OP-123 operating limits RCP RCP by indicate zero. Time Critical Secure A RCP has the potential Upper/Lower Oil securing Operation within 10 min of Green light on to degrade the RSVR Level Lo. placing Actions, the reactor trip. and Red light RCS pressure handswitch to Attachment 4 RCP bearing off above A boundary. stop or PTL.

temperature RCP CE EPGB RCPs should be rising >18 handswitch. Simulator CTs:

maintained in an degrees per CT-23, Trip any available minute on RCP exceeding condition for last computer trends operating limits resort use if or (SGTR-03) needed.

2K11 B1 A RCP Upper Thrust BRG Metal Temperature Hi in alarm.

CT-2: Isolating the SG Procedural Completion of A SG CE EPGB will minimize the direction when Standard Att. Pressure Simulator CTs:

Isolate A SG potential loss of RCS Thot is less 10. All valves trending up CT-14, Isolate (2202.010 the containment than 535 are closed by while B SG most affected 0 boundary, thus degrees F on A placing pressure SG (SGTR-09).

completed) preventing an SG. handswitches continues to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SAR Section offsite release to closed lower from the after the Reactor 15.1.18 and exceeding except valves cooldown.

trip.

10CFR100 that are EN-OP-123 Green light Assumption is exposure limits allowed to Time Critical on, Red Light that the operator at the site remain open Operation off above will diagnose boundary. by CRS Actions, valves listed in within 30 directions. Attachment 4 Att. 10.

minutes and then isolate EOP 2202.004, within next 30 SGTR Tech minutes after Guide entry into 2202.004, SGTR EOP Note: Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.

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Appendix D Scenario # 2 Form ES-D-1 Scenario #2 Objectives

1) Evaluate individual and crew response to a Log power channel failure.

2) Evaluate individual and crew response to low Boric Acid make-up temperature.

3) Evaluate individual response to a Primary to Secondary leak.

4) Evaluate individual response to a Condenser Vacuum pump trip and standby pump failure to start.

5) Evaluate individual response to RCP oil leak

6) Evaluate individual and crew ability to perform standard post trip actions.

7) Evaluate individual and crew ability to respond to a Steam Generator Tube rupture.

8) Evaluate individual response to letdown failing to isolate on SIAS.

9) Evaluate individual response to a gland seal regulator failure.

Scenario #2 Narrative Simulator session begins with the plant at 100% power steady state.

When the crew has completed their Control Room walk down and brief, B Log power will fail low to remove log power from bypass then fail high. The Excore will not be affected. The failure is in the log power circuit. The CRS will enter the OP-2203.026, NI Malfunction AOP and the crew should determine that B channel linear power is operable, but log power is failed by monitoring output for the three Excore chambers and log power indication. The CRS will also enter 3.3.3.5 for Remote Shutdown Instrumentation. The BOP will bypass point 2 on channel B PPS. [Site OE: CR-ANO 2002-693, D Excore failure.]

When the crew bypassed point 2 on channel B PPS or at the lead examiners cue, the B Boric Acid Makeup Tank (BAM tank) low temperature alarm will annunciate. The crew will place the BAM tank on recirc and manually energize the BAM tank Heaters to restore tank temperature to clear the alarm. [Site OE: CR-ANO-2-2011-03642, Boric Acid tank temperature control out of cal.]

After the ATC has placed the BAM tank on recirc and energized the heaters, and cued by the lead examiner, a primary to secondary leak will start. The CRS will enter the primary to secondary leakage AOP, OP 2203.038. ATC and BOP will perform RCS Leak rate determinations. The CRS will enter Tech Spec 3.4.6.2 action a. The CRS will direct the NLOs to control secondary contamination using standard attachment 19 and direct the chemists to sample the SGs for activity.

The crew will isolate Main steam from A Steam Generator to 2P-7A EFW pump and commence a plant shutdown. [Site OE: Primary to secondary leak, CR-ANO-2-2005-0344, Industry OE: SEN 16, SER 33-87, SOER 83-2, SOER 93-1]

When the crew has commenced power reduction or at the lead examiners cue, the A Vacuum pump will trip, and B Vacuum pump will fail to auto start. The crew will either enter the Loss of Vacuum AOP, OP-2203.019 or refer to the ACA and manually start B Vacuum pump. [Site OE:

Vacuum pump trip, CR-ANO-2-2009-92]

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Appendix D Scenario # 2 Form ES-D-1 Scenario #2 Narrative (Continued)

After the crew has manually started B Vacuum pump or at the lead examiners cue, A RCP oil leak will start that causes oil level to lower and when oil level lowers enough the bearing temperatures will start to rise. The CRS will enter OP-2203.025, RCP Emergencies AOP. The crew will monitor the A RCP oil level trend and bearing temperatures. After bearing temperatures begin to rise (trip criteria >180F/min or greater than 1950F.) the ATC should trip the reactor and secure the A RCP.

The crew may elect to secure an RCP in the B S/G loop to balance flows. Securing an RCP not satisfying operating limits is a time critical operator action per EN-OP-123 Time Critical Operator Action Program. [Site OE: RCP oil leaks CR-ANO-2-2013-1602, CR-ANO-2-2013-587, CR-ANO 2013-58, CR-ANO-2-2019-1211]

The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. After the reactor trips the Steam Generator tube rupture will degrade to 275 gpm. At the reactor trip, the gland seal regulator will fail to control gland seal pressure. This could cause a loss of condenser vacuum if not addressed by the crew. Also, Letdown will fail to isolate on SIAS and the crew will have to isolate letdown by closing the isolation valves. The crew will restore SW to CCW to ensure RCP seal integrity. [PRA item # 3 restore Service Water to CCW, PSA-ANO2-06-05] [Industry OE: SEN 16, SER 33-87, SOER 83-2, SOER 93-1]

The CRS should diagnose and enter the SGTR EOP. The CRS will direct an RCS cooldown to Th

<535 degrees Fahrenheit. When Th is <535 degrees Fahrenheit the BOP will isolate A S/G.

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Appendix D Scenario # 3 Form ES-D-1 Facility: ANO-2 Scenario No.: 3 (Modified) Op-Test No.: 2021 Examiners: ___________________________ Operators: ____________________________

Initial Conditions: ~74.5%, MOL, Red Train Maintenance Week.

Turnover: 260 EFPD. EOOS indicates Minimal Risk. 2P-3A circ water pump is secured for repair due to high vibrations. Red Train Maintenance Week. Reactor Power band 73 to 75% for 2P-3A circ water pump repair.

Scheduled evolution: Shift EH pumps from 2P-14A to 2P-14B in service.

Critical Tasks: Restore CCW to RCPs within 10 minutes of loss of flow, Isolate RCS leakage from leaving primary containment by closing CCW containment isolation valves prior to completion of SPTAs, Commence an RCS Cooldown within 30 minutes of entry into LOCA EOP.

Event Malf. No. Event Type* Event Description No.

1 N (BOP / CRS) Shift EH pumps from 2P-14A to 2P-14B OP-2106.012 Electrohydraulic Oil system.

2 XRCCHAPLVL I (ATC / CRS) A Channel Pressurizer Level channel fails low.

TS (CRS) OP-2203.028, Pressurizer System Malfunction AOP 3 CCW2P33BPWR C (BOP / ATC / 2P-33C CCW pump trips and 2P-33B CCW pump fails to CCW2P33CPWR CRS) start. CT-1 OP-2203.025, RCP Emergencies AOP 4 RCP2P32CSLK R (ATC) C Reactor Coolant Pump (RCP) develops an intersystem C (BOP / CRS) LOCA from the RCS to CCW of 15 gpm. TS for CRS.

TS (CRS) OP-2203.016, Excess RCS leakage AOP 5 RCP2P32CSLK M (All) C RCP intersystem LOCA degrades to 300 gpm. CCW to ESFK202AAF RCPs fail to auto close on CIAS. CT-2, & 3 ESFK202BAF OP-2202.001, Standard Post Trip Actions (SPTA), and OP-2202.003, Loss of Coolant Accident EOP 6 RCSHTRON C (ATC / CRS) Pressurizer Backup Heaters fail to de-energize on low pressurizer level.

OP-2202.010, Standard Attachments EOP.

7 SIS2P89AX C (BOP / CRS) 2P89A HPSI pump degradation.

OP-2202.010, Standard Attachments EOP.

End Point RCS cooldown commenced.

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Appendix D Scenario # 3 Form ES-D-1 Target Quantitative Attributes (Section D.5.d) Actual Attributes Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 3 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per 0 scenario set)

Critical Tasks (2) 3 Critical Safety Significance Cueing Measurable Performance References Task Performance Feedback Indicators CT-1: Exceeding operating 2K-11 A1, On Panel 2C14 2CV-5220 and EN-OP-123 Time limits has the A3, the BOP opens 2CV-5230 will Critical Action/Time Restore potential to degrade A5, A7 CCW the CCW open (Red light Sensitive Action CCW flow to the RCS pressure Disch flow lo crosstie valves on, green light Program.

the RCP boundary. RCPs alarms in. 2CV-5220 and off) and CCW CR-ANO-2-2010-948, within 10 should be maintained 2CV-5230. Disch flow lo Critical task criteria minutes of in an available alarms clear the loss of And Start a RCP emergencies condition for last (2K-11 A1, A3, CCW flow. backup CCW Tech guide resort use if needed. A5, A7) after pump restoring the CCW pump TDB580.0040 RCP flow. 2P-33A is started. Tech Manual preferred.

CT-2: Loss of pressurizer CCW surge On panel 2C16 2CV 5236 1, CE EPGB Simulator level takes away the tank level and 2C17 will 2CV 5254 2, & CTs:

Isolate RCS operator's most direct and radiation close the three 2CV 5255 1 CT-13, Isolate RCS leakage means of monitoring will trend up CCW to RCP closed (Red Leakage (LOCA-05) from leaving RCS inventory, indicating an CNTMT light off, green EN-OP-123 Time primary lowers pressurizer RCS leak into isolation light on) Critical Action/Time containment pressure, and CCW. The valves. Sensitive Action by closing CCW surge reduces RCS crew Program.

CCW 2CV 5236 1 tank level subcooling, all of assesses containment 2CV 5254 2 stops rising, CR-ANO-2-2010-948, which jeopardize the RCS leakage isolation 2CV 5255 1 and leakage is Critical task criteria RCS Inventory and rates.

valves prior indicated Pressure Control to Procedure inside safety functions. If the completion direction containment RCS mass loss is not of SPTAs directs the (dew point, isolated or reduced to operators to sump level, less than the makeup trip the radiation capacity of the SI reactor and trending up pumps, then core isolate the uncovery and core leakage > 44 damage will occur.

gpm.

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Appendix D Scenario # 3 Form ES-D-1 CT-3: Cooling down and Procedural On panel 2C02, SG pressure CE EPGB Simulator depressurizing the Direction the crew and RCS CTs: CT-20, Cool Commence RCS removes decay operates the temperature down and an RCS heat and lowers the turbine bypass indicators depressurize RCS cooldown DP at the break, valves to the begin lowering (LOCA-09) within 30 slowing the leak rate condenser to CR-ANO-2-2010-948, minutes of and reducing makeup lower SG Critical task criteria entry into volume required. pressure and OP-SDC entry conditions thus lower RCS 2202.003, are also required for temperature LOCA EOP.

long-term cooling.

Note: Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.

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Appendix D Scenario # 3 Form ES-D-1 Scenario #3 Objectives

1) Evaluate individual ability to shift Electrohydraulic pumps.

2) Evaluate individual and crew response to a PZR level instrument failure.

3) Evaluate individual and crew response to a loss of CCW pumps.

4) Evaluate individual and crew response to excess RCS leak (intersystem).

5) Evaluate individual ability to perform a power reduction.

6) Evaluate crews and individual ability to perform standard post trip actions.

7) Evaluate crews ability to respond to an intersystem LOCA.

8) Evaluate individual response to Pressurizer Heaters remaining energized with level below the low level cut out.

9) Evaluate individual response to a HPSI pump degradation.

Scenario #3 NARRATIVE Simulator session begins with the plant at 73-75% power steady state due to 2P-3A Circ water pump maintenance. [Site OE: CR-ANO-2-2020-2252, 2511, 2P-3B elevated vibs cause plant down power.

When the crew has taken the watch, they will use 2106.012 Electrohydraulic system operations procedure to swap lead EH pumps.

When the crew has completed the EH pump swap or at the lead examiners cue, Channel A pressurizer level control channel will fail low. This will cause letdown to lower to minimum output and both back up charging pumps to start. The CRS will enter the PZR System Malfunction AOP.

The ATC will take manual control of letdown. The ATC will shift level control channels from A to B and then restore letdown to automatic. The CRS will enter TS 3.3.3.6 for post-accident instrumentation. [Site OE: CR-ANO-2-2000-175, Pressurizer Channel 2 indication failed low.]

When letdown has been restored to automatic or cued by lead examiner, 2P-33C CCW pump will trip and 2P-33B CCW pump will fail to start automatically or manually. The CRS will enter OP-2203.025, RCP Emergencies AOP. The BOP should call NLOs to investigate the CCW pump trip.

The CRS will then direct opening all CCW crosstie valves and start 2P-33A CCW pump. [Site OE:

CR-ANO-2-2007-313, Trip of 2P-33B CCW pump with 2P-33C out of service for maintenance.]

[PRA item #1 to secure RCP if CCW can not be restored in 10 min.]

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Appendix D Scenario # 3 Form ES-D-1 SCENARIO #3 NARRATIVE (continued)

After the crew has restored CCW flow to the RCPs, and cued by the lead examiner, a 15 gpm RCS to CCW leak will start. The crew should notice that CCW Surge Tank level is rising. The crews recognition of the leak may be delayed because the B Surge Tank level would normally rise from the different pump configuration. Also, the CCW letdown radiation monitor will alarm indicating RCS to CCW leakage. The CRS will enter OP-2203.016, Excess RCS Leakage AOP, and direct the board operator actions. The crew should perform leak rates, isolate letdown to verify the leak is not in letdown and determine the need for a plant shutdown using normal boration. The CRS should enter Attachment A of Excess RCS Leakage, direct the BOP to perform attachment B, align the CCW surge tanks to the gas collection header and direct the NLO to control surge tank level.

The crew will perform a power reduction such that the plant will be taken offline. The CRS should enter Tech Spec 3.4.6.2 Action a for RCS leakage. The ATC will borate the RCS and reduce turbine load to maintain Tave-Tref within 2°F. The BOP will make preparations to remove secondary plant equipment from service as power is reduced. [Industry OE: NRC information notice 92-36 Intersystem LOCA outside containment. Industry OE: SEN-220, SEN-216, & SEN-182, RCS leakage events.]

After the required reactivity manipulations are complete and cued by the lead examiner, the RCS to CCW will degrade to 300 gpm. The CRS will direct the reactor to be tripped, actuate SIAS &

CCAS, secure RCPs, and isolate CCW to the RCPs. The CCW to RCPs valves will fail to auto close on a valid CIAS. The CRS should enter and direct the actions of SPTAs. [Industry OE: NRC information notice 92-36 Intersystem LOCA outside containment. Industry OE: SEN-220, SEN-216,

& SEN-182, RCS leakage events.] [PRA item #1 to secure RCP if CCW cannot be restored in 10 min.]

The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. The ATC should recognize that the pressurizer backup heaters failed to de-energize on low pressurizer level. Also, the crew should place the SDBCS master controller in Auto Local and lower the set point to maintain margin to saturation. [Site OE: Proportional Heater fails to de-energize on low level, CR-ANO-2-1029-1588] {Site OE: SDBCS system failures, CR-ANO-2-2020-2143, CR-ANO-2-2017-1155, CR-ANO-2-2013,2254}

The CRS will diagnose and enter OP-2202.003, Loss of Coolant Accident EOP. The crew will commence a cooldown to allow depressurization and refilling the pressurizer. The BOP will restore Service Water to Component Cooling Water and Auxiliary Cooling water. The BOP should recognize that 2P-89A discharge pressure is degraded and place 2P-89C in-service. [PRA item # 7 Start standby HPSI pump after failure of HPSI aligned for auto-start during small break LOCA, PSA-ANO2-06-05]

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Appendix D Scenario 4 Form ES-D-1 Facility: ANO-2 Scenario No.: 4 Op-Test No.: 2021 Examiners: __________________________ Operators: ____________________________

Initial Conditions: _100% power. 2P-89C is aligned to Green Train, Containment sump level ~77%

Unit 1 is offline for maintenance______________________________________________________

Turnover: Containment sump needs to be drained. Unit 1 is offline for outage and is in a Red Train Maintenance Window with H1 and A1 de-energized for maintenance_________________________

Evolution Scheduled: Drain the containment Sump from ~77% to 72%. Level is being maintained >

71% due to possible oil contamination.

Critical Tasks: When both S/G levels are less than or equal to 70 WR, then Once Through Cooling (OTC) shall be established prior to uncontrolled RCS heat up of 5°F based on average CET temperature, Establish minimum design HPSI Flow prior to initiating once through cooling, Maintain RCS Pressure within PT Curves) prior to establishing once through cooling.

Event Malf. No. Event Type* Event No. Description 1 DI_HS_2060_2 N (BOP/CRS) Drain the containment Sump. Containment sump TS(CRS) drain valve 2CV-2060-1 (containment penetration isolation) trips breaker with valve still open.

OP-2104.014, LRW and BMS Operations 2 K12-H04 C (ATC/CRS) 2K12-H4 Reactor Makeup Water Tank Temperature Lo OP-2203.012L Annunciator 2K12 corrective actions 3 XRRPZRLSP I (ATC/CRS) Reactor Reg. output to PZR level control program fails to 41%.

OP-2203.028, Pressurizer System Malfunction AOP 4 XFW2FIS0735 C (BOP/CRS) 2P-1A Feedwater pump suction flow transmitter fails.

OP-2203.012 Annunciator 2K03 Corrective Action 5 MFWPMPATRP C (BOP/CRS) 2P-1A Main Feedwater Pump R (ATC) OP-2203.027 Loss of Main Feedwater Pump AOP TS (CRS) 6 BUS2A1 M (ALL) 2A-1 bus lockout requiring a reactor trip.

7 500LOSE500 M (ALL) Loss of Offsite Power, 2P-7A EFW pump overspeed 500LOSE161 trips, and 2P-7B EFW pump has a motor fault causing a complete loss of feedwater. CT-1, & 2 CV0336 OP-2202.006 Loss of Feedwater EOP EFW2P7BFLT 8 HPI2P89BFAL C (BOP/CRS) 2P-89B High Pressure Safety Injection pump fails to ESFK409AAF start SIAS. 2CV-5075-1 Fails to open on SIAS due to a failed relay. CT-3 OP-2202.010 Standard Attachments EOP End Point At least one full train of HPSI is established, and Once Through Cooling has been initiated.

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Appendix D Scenario 4 Form ES-D-1 Target Quantitative Attributes (Section D.5.d) Actual Attributes Malfunctions after EOP entry (1-2) 1 Abnormal Events (2-4) 3 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per 0 scenario set)

Critical Tasks (2) 3 Critical Task Safety Significance Cueing Measurable Performance References Performance Feedback Indicators CT1: Adequate time to Procedural One full train of HPSI flow CE EPGB establish OTC Direction. HPSI placed in indication. Simulator CTs:

If both S/G CT-04 Establish following indications service:

levels are less once through than or equal The 5 degrees One HPSI cooling (HR-07)

Both Steam LTOP Valve to 70 WR, allows for small Pump is then Once temperature rises Generator running.

Position EOP 2202.006 Level Wide Indicates Open. Loss of Through due to fluctuations in Range level All four Cold Feedwater EOP Cooling (OTC) feedwater and steam indication Leg injection EOP 2202.006 shall be flow. This is Less than or Valves Verified Quench Tank Loss of established consistent with the equal to 70. Open. Level, Feedwater EOP prior to CEN-152 basis.

Pressure, Tech Guide uncontrolled If OTC is required, LTOP Isolation Temperature RCS heat up of this means that all Valves verified EN-OP-123 5F based on RCS CET Rising.

other available RCS Energized and Time Critical average CET Temperature and Core Heat Open on 2C-09. Operation uncontrolled temperature. Removal heat sinks Actions, heat up of 5F have been lost or on SPDS. Attachment 4 have been ineffective and that core temperatures have already, or will soon begin to, rise uncontrollably.

Unless OTC is established as a last-resort effort to remove core decay heat, core overheating and core damage are very likely. Therefore, it is essential that OTC be established promptly after meeting the OTC initiation criteria.

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Appendix D Scenario 4 Form ES-D-1 Critical Task Safety Significance Cueing Measurable Performance References Performance Feedback Indicators CT2: Loss of RCS Procedural Energizes PZR RCS Margin to CE EPGB pressure control low Direction. heaters and/or Saturation is Simulator CTs:

Maintain RCS CT-06 Establish will result in a loss of SPDS PT align for controlled Pressure within RCS Pressure subcooling. Loss of screen auxiliary spray between 30 and PT Curves (30 Control (LOAF-RCS pressure display. to control RCS 200 degrees to 200 degrees 06, SPTA-05) control high could pressure. displayed on MTS, less than 2K-10 E6, E7 result in a SPDS and RCS 2500 psia) CNTRL CH 1/2 pressurized thermal pressure does prior to Pressure shock condition. not exceed establishing HI/LO alarm.

2500 psia.

once through cooling.

CT3: SI flow keeps the Procedural On 2C16/2C17, On 2C16/2C17 CE EPGB core covered, Direction. Start HPSI indications of Simulator CTs:

Establish one CT-16 Establish cooled, and borated. pump(s) not HPSI Pressure.

full train of HPSI flow is required SI flow Inadequate HPSI running HPSI prior to throttled and (IC-03) flow could result in following SIAS.

initiating once HPSI throttle EN-OP-123 eventual core On 2C16/2C17 through criteria are Time Critical uncover and fuel Open indication cooling. NOT met. Operation damage. On 2C16/2C17 on all 4 cold leg HPSI Pump Open HPSI injection valves. Actions, not running Injection Attachment 4 after SIAS. Valve(s) not open following HPSI Injection SIAS.

Valves not open following SIAS.

Note: Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.

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Appendix D Scenario 4 Form ES-D-1 Scenario #4 Objectives

1) Evaluate individual ability to drain the containment sump.

2) Evaluate individual response to a malfunction of the heating steam supply to reactor makeup water tank.

3) Evaluate individual/crew response to a failure of a Pressurizer Systems Malfunction.

4) Evaluate individual/crew response to a failure of Feedwater Pump Flow Transmitter.

5) Evaluate individual/crew response to a Loss of Main Feedwater Pump.

6) Evaluate individual/crew ability to respond to bus lockout.

7) Evaluate individual ability to respond to a Loss of Feedwater.

8) Evaluate individual ability to monitor operation of Engineered Safety Features equipment.

9) Evaluate individual ability to establish once through cooling.

Scenario #4 NARRATIVE Simulator session begins with the plant at ~100% power.

When the crew has completed their control, room walk down/brief, they will use OP-2104.014 LRW and BMS Operations to drain the containment sump. When the sump is drained and they attempt to isolate the second drain valve, 2CV-2060-1 will trip the breaker. The CRS should enter Tech Spec 3.6.3.1 action a. [Site OE: CR-ANO-2-2005-386 Containment isolation MOV inoperable due to intermittent failure, CR-ANO-2-2005-2182, blown control power fuse for Containment cooler MOV.]

When containment drain activity completed, tech specs addressed, and cued by the lead examiner, a Reactor Makeup Water Tank Low Temperature Alarm will annunciate. The Crew should refer to the Annunciator corrective action and dispatch a field operator for local temperature indication. When temperature is determined to be low. Crew should align heating steam by opening 2CV-4962 Steam Inlet on 2C09 control room panel per the Annunciator 2K12 Corrective Action. [Site OE: CR-ANO 2013-438, RMWT heating steam valve stuck in intermediate position; CR-ANO-2-2013-560, RMWT heating steam temperature switch mis-calibration]

After the ATC has completed aligning the heating steam for the Reactor Makeup Water Tank and cued by lead examiner; the Reactor Reg Input to the Pressurizer Level Control Program will fail low which will cause Pressurizer Level setpoint to go from 60% to 41%. The CRS will enter PZR Systems Malfunction AOP. The CRS will direct placing letdown flow controller 2HIC-4817 in manual and stop backup chagrining pumps to control pressurizer level. The CRS will then direct the ATC to place the Pressurizer Level Controller in Local/Auto and adjust the setpoint based on Tave and referencing the Pressurizer Level Program.

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Appendix D Scenario 4 Form ES-D-1 SCENARIO #4 NARRATIVE (continued)

After Pressurizer Level Controller setpoint adjusted and cued by the lead examiner 2FIS-0735 2P-1A Feedwater pump (FWP) suction flow transmitter fails low. This will cause 2P-1A FWP recirc and the associated condensate recirc to fully open. This will drop condensate pressure and cause a condensate pump to auto start. The Crew should refer to the ACA place the FWP recirc and condensate recirc to manual and throttle/close to restore condensate header pressure to the normal band of 650 psig to 750 psig. [Site OE: CR-ANO-2-2020-1141, CR-ANO-2-2017-4175, CR-ANO 2015-4606, No flow indication through 2FIC-0735]

After Feedwater and condensate recircs are closed, and cued by the lead examiner, the 2P-1A Main Feedwater Pump will trip. The CRS will enter OP-2203.027 Loss of Main Feedwater Pump AOP. The CRS will direct the BOP to perform Exhibit 1 to lower turbine load and initiate Emergency Boration.

The CRS will then Direct the ATC to insert group 6 and P CEAs as necessary to reduce power. When feed flow is greater than steam flow and steam generator levels have stopped lowering, turbine load reduction, emergency boration, and CEA insertion will be secured. CRS will enter T.S. 3.2.6 for exceeding the Reactor Coolant Cold Leg Temperature of 554.7 degrees F. [Site OE: CR-ANO-2-2009-3744, Loss of A Main Feedwater pump, CR-ANO-2-2020-0678 Loss of Main Feedwater pump]

[Industry OE: SER 29-85 Loss of Main and Auxiliary Feedwater, IER L3-11-44 Reactor Scram following a loss of Main Feedwater.]

After emergency boration is secured and cued by the lead examiner, 2A-1 4160V Non-Vital Bus will have a lockout. The loss of 2A-1 will de-energize 2 Condensate Pumps which will cause the crew to take the plant offline or the reactor to trip on low steam generator level. [Industry OE: IER L3-12-88, Automatic Reactor Scram caused by a 4.16-kV bus lockout during maintenance, DEN 126 Fire in 4.16 KV switchgear]

The Crew will implement Standard Post Trip Actions (SPTA), OP 2202.001. When the Reactor trips a Loss of Offsite Power will occur and when 2P-7A Emergency Feedwater pump starts, it will overspeed trip and 2P-7B Emergency Feedwater Pump will have a motor fault on start causing a complete loss of feedwater. The BOP will close Main Steam Isolation Valves (MSIVs) due to a loss of power causing overcooling. The ATC will align for aux spray to control RCS pressure during the heat up caused by closing the MSIVs. The BOP will isolate Steam Generator (SG) blowdown to conserve SG inventory.

The BOP should align upstream atmospheric dump isolation valve to control SG pressure 950 to 1050 psia. [Industry OE: SOER 99-1 Loss of Grid, SER 29-85 Loss of Main and Auxiliary Feedwater, SER 77-81 Complete loss of Auxiliary Feedwater]

The CRS will diagnose Loss of Feedwater EOP, 2202.006. The CRS will direct energizing the LTOP relief isolation valves and commence trying to restore Feedwater to the Steam Generators. 2P-7A EFW pump will not be able to be reset, 2P-7B EFW pump will be unavailable due to motor fault, 2P-75 AFW pump is unavailable due to the 2A-1 lockout, Main Feedwater and Condensate pumps are not available due to the loss of offsite power. Common Feedwater will not be available due to loss of off-site power and Unit 1 A1 being de-energized for maintenance. The Crew will align for once through cooling. When SIAS is actuated 2P-89B HPSI pump will fail to auto start and an ESF relay will fail to open red train injection valve 2CV-5075-1. The BOP should start 2P-89C or 2P-89B HPSI pump and open 2CV-5075-1. After once through cooling has been initiated, the crew will transition to the Functional Recovery EOP, 2202.009. [PRA item # 7 Start standby HPSI pump after failure of HPSI aligned for auto-start during small break LOCA, PSA-ANO2-06-05]

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Appendix D Scenario 5 Form ES-D-1 Facility: ANO-2 Scenario No.: 5 Op-Test No.: 2021 Examiners: __________________________ Operators: ____________________________

Initial Conditions: ~4.5% MOL; RED Train Maintenance Week.

Turnover: ~4.5%. 260 EFPD. EOOS indicates Minimal Risk. RED Train Maintenance Week. Steam Bypass valve in auto local setpoint of 990 psia. Reactor power was reduced to ~4.5% for Turbine CV EH leak repair and DEFAS cabinet repair. Power is being maintained at 4 to 4.9%. Trip Criteria established per Power Operation Low Power guidelines is 2%.

Evolution scheduled: Shift in service Hold Up Tanks from 2T-12B to 2T-12C.

Critical Tasks: Crew should establish and maintain the RCS within the limits of the PT Curve, energize at least 1 vital 4160V bus prior to completion of SPTAs, Restore CCW to the RCPS within 10 minutes of the loss of CCW.

Event Malf. No. Event Type* Event No. Description 1 N (BOP/ CRS) Swap in service 2T-12 Holdup Tanks OP-2104.014 LRW & BMS Operations 2 CVC4817DEM I (ATC/ CRS) Letdown flow controller auto signal drifts high.

OP-2203.012L Annunciator 2K12 Corrective Action (ACA).

3 XSG2PT10412 I (BOP/ CRS) 2PT-1041-2 SG-A pressure detector fails low.

TS (CRS) OP-2203.012D, Annunciator 2K04 Corrective Action OP-2105.001, CPC/CEAC Operations 4 CV4652 C (ATC/ CRS) B RCP normal Spray Valve drifts partially open.

OP-2203.028 Pressurizer System Malfunction AOP 5 DI_C40_S72B C (BOP/ ATC/ Inadvertent Containment Isolation Actuation Signal (CIAS)

CRS) on the Green Train. CT-1 TS(CRS) OP-2203.039, Inadvertent CIAS 6 LOSE161 M (ALL) Loss of Offsite Power. CT-2 LOSE500 OP-2202.001, Standard Post Trip Actions (SPTAs) EOP OP-2202.007, Loss of Offsite Power EOP 7 EDG2AUTOFAIL C (BOP/ CRS) #2 EDG fails to auto-start. CT-3 OP-2202.001, Standard Post Trip Actions (SPTAs) EOP 8 CV15031 C (BOP/ CRS) Service water Outlet valve binds and will not open for #1 EDG requiring the crew to secure it during SPTAs.

Alternate AC Diesel and aligned to 2A-3.

OP-2202.007, Loss of Offsite Power EOP End Point Alternate AC Diesel Generator Started and aligned to 2A-3 with pressurizer inventory and pressure control established.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Revision 0 Page 1 of 47

Appendix D Scenario 5 Form ES-D-1 Target Quantitative Attributes (Section D.5.d) Actual Attributes Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 4 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per scenario 0 set)

Critical Tasks (2) 3 Critical Task Safety Significance Cueing Measurable Performance References Performance Feedback Indicators CT-1: Exceeding operating 2K-11 A1, A3, On Panel 2C- 2CV-5254-2 will EN-OP-123 limits has the A5, A7 CCW 16 place the open (Red light Attachment 4 Component potential to degrade Disch flow lo CCW to RCP on) and CCW ANO Specific Cooling Water the RCS pressure alarms in. valve hand Disch flow lo TCA (CCW) to boundary. RCPs switch to close alarms clear CE EPGB RCPs must be should be then to open for (2K-11 A1, A3, Simulator CTs:

restored within maintained in an 2CV-5254-2 A5, A7) CT-23, Trip any 10 minutes of available condition RCP exceeding the loss of for last resort use if operating limits cooling water.

needed. (LOCA-04)

AOP OP-If RCPs are allowed 2203.039 to operate for 10 Inadvertent minutes without CIAS.

CCW flow. OP-1015.050 requires RCPs not meeting operating limits to be secured within 10 minutes.

CT-2: Loss of RCS Procedural Energizes PZR RCS Margin to CE EPGB pressure control will Direction. heaters and/or Saturation is Simulator CT Crew should result in a loss of SPDS PT align for controlled CT-06 Establish establish and RCS subcooling and screen auxiliary spray between 30 and RCS Pressure maintain the a reactor head void display. to control RCS 200 degrees Control (LOOP-RCS within the can form, both of pressure. displayed on 04) limits of the PT 2K-10 E6, E7 which complicate the SPDS and RCS CR-ANO Curve (<2000 F CNTRL CH 1/2 event recovery. pressure does 2010-948, and >300F Pressure Uncontrolled void not exceed Critical Task MTS. Also HI/LO alarm.

growth could result 2500 psia. Criteria RCS pressure in eventual core does not uncover and fuel exceed 2500 damage.

psia.)

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Appendix D Scenario 5 Form ES-D-1 Critical Task Safety Significance Cueing Measurable Performance References Performance Feedback Indicators CE EPGB CT-3: Without any AC 2A-4 bus is #2 EDG start 2K09-A3 alarm Simulator CTs:

power available for de-energized Handswitch is clears and 2A-4 Crew should CT-03, ESF pumps, the zero volts on taken to start. bus voltmeter ensure that at Energize at ability to maintain the 2A4 due #2 indicating least one vital least one vital plant in a safe state EDG failure to approximately 4160V bus is AC bus (SPTA-is severely degraded Auto start. 4160 V.

energized prior 02) since no makeup to completion 2K09-A3 water can be added CR-ANO of SPTAs. alarm in and to the RCS for 2010-948, 2A-4 bus inventory control Critical Task voltmeter purposes. Therefore, Criteria indicating zero energizing at least one vital (4.16 kV) 2A-3 bus is 10 CFR 50.63, AC bus in the SPTAs de-energized Station is essential if it can zero volts due Blackout be done quickly. to securing #1 EDG because of not service water cooling.

2K08-A3 alarm in and 2A-3 bus voltmeter indicating zero Note: Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.

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Appendix D Scenario 5 Form ES-D-1 Scenario #5 Objectives

1) Evaluate individual ability to shift Hold Up Tanks.

2) Evaluate individual and crew response to a failure of a Letdown flow controller.

3) Evaluate individual response to a failure of Steam Generator pressure transmitter.

4) Evaluate individual response to a failure of a Pressurizer Spray Valve

5) Evaluate individual response to Inadvertent Containment Injection Actuation Signal

6) Evaluate individual and crews ability to mitigate a Loss of Offsite Power event

7) Evaluate individual ability to monitor operation of Engineered Safety Features equipment and respond to failure of #2 EDG failure to auto start.

8) Evaluate individual ability to monitor operation of Engineered Safety Features equipment and restore power with Alternate AC Diesel Generator.

SCENARIO #5 NARRATIVE Simulator session begins with the plant at ~4% power. [Site OE: CR-ANO-2-2016-1993 EH Leak causes power reduction and manual turbine trip. ICES # 323174]

When the crew has completed their control room walk down and brief, the BOP will shift in service 2T-12 Holdup Tanks using OP-2104.014 LRW & BMS Operations. The BOP will shift from 2T-12B to 2T-12C in service.

After the BOP has shifted the in-service holdup tanks, the letdown flow controller signal will drift high. This will cause elevated letdown flow. The ATC should recognize elevated letdown flow.

The ATC should take manual control of the letdown flow controller and adjust letdown to restore PZR level near setpoint. The crew should follow up with the Annunciator corrective action. [Site OE: Letdown flow and pressure oscillations, CR-ANO-2-2001-0078, After the ATC has control of letdown flow and is restoring PZR level to setpoint, the A Steam Generator pressure safety channel pressure instrument, 2PT-1041-2, will fail low. This will trip one of the four PPS channels for low SG pressure trip. Alarms for RPS channel trip/pre-trip, MSIS pre-trip and channel B operator insert (2C03) trip and pre-trip light will be lit. The CRS will refer to the ACA 2203.012D and enter tech specs 3.3.1.1 action 2, 3.3.2.1 action 10, and 3.3.3.6 action 1. The BOP will place Channel B PPS in bypass for point 11 SG pressure low, point 19 SG1 delta-P high, and point 20 SG2 delta-P. The crew will have one hour to place these points in bypass before exceeding the tech spec LCO. [Site OE: CR-ANO-2-1988-0025, CR-ANO-2-1994-398, Steam Generator pressure transmitter failed low.]

When the appropriate Tech spec has been entered and B PPS channel is placed in bypass and cued by lead examiner, 2CV-4652 Pressurizer Spray Valve drifts partially open. CRS will enter Pressurizer System Malfunction AOP. BOP will place the Pressurizer Spray valve in manual and attempt to close. When the spray valve will not close, BOP will isolate the spray valve by closing Block Valve 2CV-4654. [Industry OE: Pressurizer spray valve failures, SER 4-93, SER 14-81, SER 12-81, SEN 230]

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Appendix D Scenario 5 Form ES-D-1 SCENARIO #5 NARRATIVE (continued)

When pressurizer spray valve has been isolated and cued by lead examiner, An Inadvertent Containment Isolation will occur on the green train causing the green train CCW to RCP valve and the Main Chilled water to containment valves to close. The CRS will enter Inadvertent CIAS AOP, OP 2203.039. The crew should restore Component Cooling Water (CCW) to RCPs. The CRS will enter Tech Spec 3.6.3.1 for the overridden Containment Isolation valve. The ATC will cycle charging pumps to control pressurizer level. The crew should minimize CEA movement due to the loss of cooling. The BOP will start all containment coolers and align Service Water to maintain Containment temperature and pressure in the required band. The CRS should call for maintenance assistance to correct inadvertent green train Containment isolation. [Industry OE:

SEN 268 Invalid Safety Injection with Failure to Reset, Site OE: CR-ANO-2-2013-005 Inadvertent SIAS, CCAS, And CIAS.]

When the actions of inadvertent CIAS have been completed or at the lead examiners cue, a loss of all offsite power will occur. [Industry OE: 2020 Brunswick Unit 1 Loss of Offsite Power ICES 080420, SER 10-91, SEN 17, SEN 128, IER L2-12-27, IER L2-14-26]

The Crew will implement Standard Post Trip Actions (SPTA) EOP, 2202.001. During SPTAs the

1. 2 EDG will fail to auto start and the Service Water outlet for #1 EDG will fail to open. The BOP will recognize and manually start #2 EDG and coordinate field operators to investigate and then locally lockout the #1 EDG thus de-energizing 2A-3 vital 4160v bus. The loss of off-site power, and loss of red 4160V AC causes a loss of ability to precisely control B Steam Generator level (loss of 2P-7B electric Emergency Feedwater Pump) or pressure (loss of power to upstream steam dump isolation). The ATC must align the pressure level channel select, Lo level cutout switch and Pressure control channel selector switch to the B channel to allow restoration of PZR heaters and also align for auxiliary spray using Att. 48. [SOER 03-1 Emergency power reliability]

The CRS will diagnose and enter Loss of Offsite Power. BOP will start the Alternate AC Diesel Generator and tie it to 2A-3 vital 4160v bus to restore vital 4160 volt and 480-volt power. Once the Alternate AC Diesel is running the crew will start a Service Water (SW) pump to raise SW header pressure to clear the low pressure alarms. The crew align the SW return header to ECP to protect SW flow from being isolated by the SW lake return valve failing closed after the local IA receiver is depleted. The crew should also align SW to the containment coolers due to the loss of main chill water. [PRA item # 6 start and align alternate AC generator following failure of an EDG, PSA-ANO2-06-05]

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ES-301 Transient and Event Checklist Form ES-301-5 Crew A Facility: ANO-2 Date of Exam: March 2021 Operating Test No.: 2021 A E Scenarios P V 2 3 4 5 T M P E O I L N CREW POSITION CREW POSITION CREW POSITION CREW POSITION T N I T S A B S A B S A B S A B I C A R T O R T O R T O R T O M A T L O C P O C P O C P O C P U N Y M(*)

T P E R I U RO RX 0 1 1 0 NOR 0 1 1 1 SRO-I I/C 1,2,3, 7 4 4 2 4,5,7, SRO-U 8

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

TS 2, 4 2 0 2 2 RO RX 3 1 1 1 0 X NOR 1 1 1 1 1 SRO-I I/C 2,5,8 3,4,7 6 4 4 2 SRO-U MAJ 6 5 2 2 2 1 TS 0 0 2 2 RO RX 4 1 1 1 0 X NOR 0 1 1 1 SRO-I I/C 1,3,4, 2,3, 7 4 4 2 7 6 SRO-U MAJ 6 5 2 2 2 1 TS 0 0 2 2 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.

ES-301 Transient and Event Checklist Form ES-301-5 Crew B Facility: ANO-2 Date of Exam: March 2021 Operating Test No.: 2021 A E Scenarios P V 2 3 4 5 T M P E O I L N CREW POSITION CREW POSITION CREW POSITION CREW POSITION T N I T S A B S A B S A B S A B I C A R T O R T O R T O R T O M A T L O C P O C P O C P O C P U N Y M(*)

T P E R I U RO RX 0 1 1 0 NOR 0 1 1 1 SRO-I I/C 1,2,3, 7 4 4 2 4,5,7, SRO-U 8

X MAJ 6 1 2 2 1 TS 1,3 2 0 2 2 RX 3 1 1 1 0 RO NOR 1 1 2 1 1 1 SRO-I I/C 2,5,8 2,3,4, 2,3,4, 13 4 4 2 X 6,7 5,8 SRO-U MAJ 6 5 6,7 4 2 2 1 TS 2, 4 1,5 4 0 2 2 RO RX 4 1 1 1 0 X NOR 1 1 1 1 1 SRO-I I/C 1,3,4, 2,3, 4,5,8 10 4 4 2 7 6 SRO-U MAJ 6 5 6,7 4 2 2 1 TS 0 0 2 2 RO RX 5 1 1 1 0 X NOR 1 1 1 1 1 SRO-I I/C 3,4,7 2,3 5 4 4 2 SRO-U MAJ 5 6,7 3 2 2 1 TS 0 0 2 2 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.

ES-301 Transient and Event Checklist Form ES-301-5 Crew C Facility: ANO-2 Date of Exam: March 2021 Operating Test No.: 2021 A E Scenarios P V 2 3 4 5 T M P E O I L N CREW POSITION CREW POSITION CREW POSITION CREW POSITION T N I T S A B S A B S A B S A B I C A R T O R T O R T O R T O M A T L O C P O C P O C P O C P U N Y M(*)

T P E R I U RO RX 0 1 1 0 NOR 0 1 1 1 SRO-I I/C 1,2,3, 7 4 4 2 4,5,7, SRO-U 8

X MAJ 6 1 2 2 1 TS 1,3 2 0 2 2 RX 3 1 1 1 0 RO NOR 1 1 2 1 1 1 SRO-I I/C 2,5,8 2,3,4, 2,3,4, 14 4 4 2 X 5,8 5,7,8 SRO-U MAJ 6 6,7 6 4 2 2 1 TS 1,5 3,5 4 0 2 2 RO RX 5 1 1 1 0 X NOR 1 1 1 1 1 SRO-I I/C 1,3,4, 2,3 3,5,7, 10 4 4 2 7 8 SRO-U MAJ 6 6,7 6 4 2 2 1 TS 0 0 2 2 RO RX 0 1 1 0 X NOR 1 1 1 1 1 SRO-I I/C 4,5,8 2,4,5 6 4 4 2 SRO-U MAJ 6,7 6 3 2 2 1 TS 0 0 2 2 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.

ES-301 Transient and Event Checklist Form ES-301-5 Crew D Facility: ANO-2 Date of Exam: March 2021 Operating Test No.: 2021 A E Scenarios P V 2 3 4 5 T M P E O I L N CREW POSITION CREW POSITION CREW POSITION CREW POSITION T N I T S A B S A B S A B S A B I C A R T O R T O R T O R T O M A T L O C P O C P O C P O C P U N Y M(*)

T P E R I U RO RX 0 1 1 0 NOR 1 1 1 1 1 SRO-I I/C 1,2,3, 2,3,4, 13 4 4 2 4,5,7, 5,7,8 SRO-U 8

X MAJ 6 6 2 2 2 1 TS 1,3 3,5 4 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 RO RX 3 1 1 1 0 X NOR 1 1 1 1 1 SRO-I I/C 2,5,8 3,5,7, 7 4 4 2 8

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

SRO-U MAJ 6 6 2 2 2 1 TS 0 0 2 2 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.

ES-301 Transient and Event Checklist Form ES-301-5 Crew E Facility: ANO-2 Date of Exam: March 2021 Operating Test No.: 2021 A E Scenarios P V 2 3 4 5 T M P E O I L N CREW POSITION CREW POSITION CREW POSITION CREW POSITION T N I T S A B S A B S A B S A B I C A R T O R T O R T O R T O M A T L O C P O C P O C P O C P U N Y M(*)

T P E R I U RO RX 5 1 1 1 0 NOR 1 1 2 1 1 1 SRO-I X I/C 2,3,4, 2,3 2,3,4, 13 4 4 2 6,7 5,7,8 SRO-U MAJ 5 6,7 6 4 2 2 1 TS 2, 4 3,5 4 0 2 2 RX 4 1 1 1 0 RO NOR 1 1 2 1 1 1 SRO-I I/C 2,3, 2,3,4, 3,5,7, 12 4 4 2 X 6 5,8 8 SRO-U MAJ 5 6,7 6 4 2 2 1 TS 1,5 2 0 2 2 RO RX 0 1 1 0 X NOR 1 1 2 1 1 1 SRO-I I/C 3,4,7 4,5,8 2,4,5 9 4 4 2 SRO-U MAJ 5 6,7 6 4 2 2 1 TS 0 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 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.

ES-301 Transient and Event Checklist Form ES-301-5 Crew F Facility: ANO-2 Date of Exam: March 2021 Operating Test No.: 2021 A E Scenarios P V 2 3 4 5 T M P E O I L N CREW POSITION CREW POSITION CREW POSITION CREW POSITION T N I T S A B S A B S A B S A B I C A R T O R T O R T O R T O M A T L O C P O C P O C P O C P U N Y M(*)

T P E R I U RO RX 0 1 1 0 NOR 1 1 2 1 1 1 SRO-I X I/C 2,3,4, 4,5,8 2,4,5 11 4 4 2 6,7 SRO-U MAJ 5 6,7 6 4 2 2 1 TS 2, 4 2 0 2 2 RX 4 1 1 1 0 RO NOR 1 1 2 1 1 1 SRO-I I/C 2,3, 2,3,4, 3,5,7, 12 4 4 2 X 6 5,8 8 SRO-U MAJ 5 6,7 6 4 2 2 1 TS 1,5 2 0 2 2 RO RX 5 1 1 1 0 NOR 1 1 2 1 1 1 SRO-I X I/C 3,4,7 2,3 2,3,4, 11 4 4 2 5,7,8 SRO-U MAJ 5 6,7 6 4 2 2 1 TS 3,5 2 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 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.