ML21137A290

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7_AN2-2021-03 Final Outlines
ML21137A290
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 04/08/2021
From: Greg Werner
Operations Branch IV
To:
Entergy Operations
References
Download: ML21137A290 (45)
v • d • e


Text

ES-401 1

Form ES-401-2 1

Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total 1.

Emergency and Abnormal Plant Evolutions 1

3 4

3 N/A 3

3 N/A 2

18 6

2 2

1 1

1 1

3 9

4 Tier Totals 5

5 4

4 4

5 27 10 2.

Plant Systems 1

4 3

3 4

1 1

3 2

2 2

3 28 5

2 0

0 1

2 2

2 1

1 1

0 0

10 3

Tier Totals 4

3 4

6 3

3 4

3 3

2 3

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 3

2 2

Note: 1.

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

2.

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

3.

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

4.

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

5.

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

6.

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

7.

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

8.

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

9.

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

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

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

ES-401 2

Form ES-401-2 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)

Reactor Trip, Stabilization, Recovery / 1 X

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) 3.3 1

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

AK2.02 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and detectors (CFR 41.7 / 45.7) 2.7 2

000009 (EPE 9) Small Break LOCA / 3 SRO X

EK2.03 Knowledge of the interrelations between the small break LOCA and the following: S/Gs (CFR 41.7 / 45.7) 3.0 3

000011 (EPE 11) Large Break LOCA / 3 X

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) 4.3 4

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

AK3.07 Knowledge of the reasons for the 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) 4.1 5

000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X

AA2.01 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Whether charging line leak exists (CFR 43.5/ 45.13) 3.2 6

000025 (APE 25) Loss of Residual Heat Removal System / 4 SRO 000026 (APE 26) Loss of Component Cooling Water / 8 X

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) 3.6 7

ES-401 3

Form ES-401-2 3

000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X

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) 3.2 8

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

EK2.06 Knowledge of the interrelations between the and the following an ATWS:

Breakers, relays, and disconnects (CFR 41.7 / 45.7) 2.9 9

000038 (EPE 38) Steam Generator Tube Rupture / 3 SRO X

EK3.01 Knowledge of the reasons for the following responses as the apply to the SGTR: Equalizing pressure on primary and secondary sides of ruptured S/G (CFR 41.5 / 41.10 / 45.6 / 45.13) 4.1 10 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 X

AK1.06 Knowledge of the operational 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) 3.7 11 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

2.1.20 Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12) 4.6 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 following as they apply to the Loss of Offsite Power: AFW flow control valve operating switches (CFR 41.7 / 45.5 / 45.6) 3.5 13 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X

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) 4.1 14 000058 (APE 58) Loss of DC Power / 6 X

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) 2.8 15

ES-401 4

Form ES-401-2 4

000062 (APE 62) Loss of Nuclear Service Water / 4 X

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) 2.5 16 000065 (APE 65) Loss of Instrument Air / 8 X

2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) 4.4 17 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 SRO X

AK2.07 Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Turbine /

generator control (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8) 3.6 18 (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 NA K/A Category Totals:

3 4

3 3

3 2

Group Point Total:

18

ES-401 5

Form ES-401-2 5

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

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) 3.1 19 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 X

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) 4.2 20 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 SRO 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 System 033 rejected (See 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

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) 3.9 21 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X

2.1.30 Ability to locate and operate components, including local controls.

(CFR: 41.7 / 45.7) 4.4 22 000060 (APE 60) Accidental Gaseous Radwaste Release / 9

ES-401 6

Form ES-401-2 6

000061 (APE 61) Area Radiation Monitoring System Alarms

/ 7 X

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

(CFR: 41.10 / 43.5 / 45.3 /

45.12) 4.1 23 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 X

AK2.03 Knowledge of the interrelations between the Control Room Evacuation and the following: Controllers and positioners (CFR 41.7 / 45.7) 2.9 24 000069 (APE 69; W E14) Loss of Containment Integrity / 5 SRO 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /

4 X

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) 3.8 25 000076 (APE 76) High Reactor Coolant Activity / 9 X

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) 2.9 26 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

ES-401 7

Form ES-401-2 7

(CE A16) Excess RCS Leakage / 2 X

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) 3.0 27 (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

ES-401 8

Form ES-401-2 8

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

System # / Name K1 K2 K 3

K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump X2 X

A4.06 Ability to manually operate and/or monitor in the control room: RCP parameters (CFR: 41.7 / 45.5 to 45.8) 2.9 28 003 (SF4P RCP) Reactor Coolant Pump X2 X

K4.03 Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following: Adequate lubrication of the RCP (CFR: 41.7) 2.5 29 004 (SF1; SF2 CVCS) Chemical and Volume Control X2 X

K1.19 Knowledge of the physical connections and/or cause-effect relationships between the CVCS and the following systems: Primary grade water supply (CFR: 41.2 to 41.9 / 45.7 to 45.8) 2.7 30 004 (SF1; SF2 CVCS) Chemical and Volume Control X2 X

K5.50 Knowledge of the operational implications of the following concepts as they apply to the CVCS: Design basis letdown system temperatures: resin integrity (CFR: 41.5/45.7) 2.6 31 005 (SF4P RHR) Residual Heat Removal SRO X

K2.03 Knowledge of bus power supplies to the following: RCS pressure boundary motor-operated valves (CFR: 41.7) 2.7 32 006 (SF2; SF3 ECCS) Emergency Core Cooling X2 X

A2.13 Ability to (a) predict the impacts of the following malfunctions or operations 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) 3.9 33 006 (SF2; SF3 ECCS) Emergency Core Cooling X2 X

K6.10 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Valves (CFR: 41.7 / 45.7) 2.6 34

ES-401 9

Form ES-401-2 9

007 (SF5 PRTS) Pressurizer Relief/Quench Tank X

K3.01 Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment (CFR: 41.7 / 45.6) 3.3 35 008 (SF8 CCW) Component Cooling Water X2 X

A3.08 Ability to monitor automatic operation of the CCWS, including:

Automatic actions associated with the CCWS that occur as a result of a safety injection signal (CFR: 41.7 / 45.5) 3.6 36 008 (SF8 CCW) Component Cooling Water X2 X

A1.04 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including:

Surge tank level (CFR: 41.5 / 45.5) 3.1 37 010 (SF3 PZR PCS) Pressurizer Pressure Control X

K4.01 Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following: Spray valve warm-up (CFR: 41.7) 2.7 38 012 (SF7 RPS) Reactor Protection SRO X

K2.01 Knowledge of bus power supplies to the following: RPS channels, components, and interconnections (CFR: 41.7) 3.3 39 013 (SF2 ESFAS) Engineered Safety Features Actuation SRO X

2.4.39 Knowledge of RO responsibilities in emergency plan implementation.

(CFR: 41.10 / 45.11) 3.9 40 022 (SF5 CCS) Containment Cooling X2 X

A3.01 Ability to monitor automatic operation of the CCS, including: Initiation of safeguards mode of operation (CFR: 41.7 / 45.5) 4.1 41 022 (SF5 CCS) Containment Cooling X2 X

K1.01 Knowledge of the physical 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) 3.5 42 025 (SF5 ICE) Ice Condenser NA

ES-401 10 Form ES-401-2 10 026 (SF5 CSS) Containment Spray X

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) 3.8 43 039 (SF4S MSS) Main and Reheat Steam X

2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) 4.4 44 059 (SF4S MFW) Main Feedwater X

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) 2.7 45 061 (SF4S AFW)

Auxiliary/Emergency Feedwater X2 X

A2.03 Ability to (a) predict the impacts of the following malfunctions or operations 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) 3.1 46 061 (SF4S AFW)

Auxiliary/Emergency Feedwater X2 X

A3.03 Ability to monitor automatic operation of the AFW, including: AFW S/G level control on automatic start (CFR: 41.7 / 45.5) 3.9 47 062 (SF6 ED AC) AC Electrical Distribution SRO X

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

(CFR: 41.10 / 43.5 / 45.13) 2.6 48 063 (SF6 ED DC) DC Electrical Distribution X

A1.01 Ability to predict and/or monitor 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) 2.5 49

ES-401 11 Form ES-401-2 11 064 (SF6 EDG) Emergency Diesel Generator SRO X

A4.01 Ability to manually operate and/or monitor in the control room: Local and remote operation of the ED/G (CFR: 41.7 / 45.5 to 45.8) 4.0 50 073 (SF7 PRM) Process Radiation Monitoring X

K1.01 Knowledge of the physical 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) 3.6 51 076 (SF4S SW) Service Water X

K2.01 Knowledge of bus power supplies to the following: Service water (CFR: 41.7) 2.7 52 078 (SF8 IAS) Instrument Air X2 X

K4.02 Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: Cross-over to other air systems (CFR: 41.7) 3.2 53 078 (SF8 IAS) Instrument Air X2 X

K3.02 Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Systems having pneumatic valves and controls (CFR: 41.7 / 45.6) 3.4 54 103 (SF5 CNT) Containment X

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) 2.8 55 053 (SF1; SF4P ICS*) Integrated Control NA K/A Category Point Totals:

4 3 3 4 1 1 3 2 2 2 3 Group Point Total:

28

ES-401 12 Form ES-401-2 12 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 Coolant X

K6.06 Knowledge of the effect or a loss or malfunction on the following RCS components: Sensors and Detectors (CFR: 41.7 / 45.7 2.5 56 011 (SF2 PZR LCS) Pressurizer Level Control X

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) 3.7 57 014 (SF1 RPI) Rod Position Indication SRO 015 (SF7 NI) Nuclear Instrumentation X

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) 3.1 58 016 (SF7 NNI) Nonnuclear Instrumentation X

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) 2.8 59 017 (SF7 ITM) In-Core Temperature Monitor SRO 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X

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) 3.4 60 029 (SF8 CPS) Containment Purge X

A3.01 Ability to monitor automatic operation of the Containment Purge System including: CPS isolation (CFR: 41.7 / 45.5) 3.8 61 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment SRO

ES-401 13 Form ES-401-2 13 035 (SF 4P SG) Steam Generator X

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) 3.6 62 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X

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) 2.8 63 045 (SF 4S MTG) Main Turbine Generator X

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) 2.9 64 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate X

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) 2.6 65 068 (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 Ventilation NA K/A Category Point Totals:

0 0

1 2

2 2

1 1

1 0

0 Group Point Total:

10

ES-401 Generic Knowledge and Abilities Outline (Tier 3) RO Form ES-401-3 14 Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.21 Ability to verify the controlled procedure copy.

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

(CFR: 41.10 / 45.12) 3.4 67 2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.

(CFR: 41.10 / 45.12) 2.7 68 Subtotal 3

2. Equipment Control 2.2.13 Knowledge of tagging and clearance procedures.

(CFR: 41.10 / 45.13) 4.1 69 2.2.14 Knowledge of the process for controlling equipment configuration or status.

(CFR: 41.10 / 43.3 / 45.13) 3.9 70 2.2.43 Knowledge of the process used to track inoperable alarms.

(CFR: 41.10 / 43.5 / 45.13) 3.0 71 Subtotal 3

3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10) 3.2 72 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 / 45.9 / 45.10) 3.2 73 Subtotal 2

4. Emergency Procedures/Plan 2.4.25 Knowledge of fire protection procedures.

(CFR: 41.10 / 43.5 / 45.13) 3.3 74 2.4.3 Ability to identify post-accident instrumentation.

(CFR: 41.6 / 45.4) 3.7 75 Subtotal 2

Tier 3 Point Total 10

ES-401 15 Form ES-401-2 15 Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total 1.

Emergency and Abnormal Plant Evolutions 1

N/A N/A 18 3

3 6

2 9

3 1

4 Tier Totals 27 6

4 10 2.

Plant Systems 1

28 3

2 5

2 10 2

1 3

Tier Totals 38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

2 2

1 2

Note: 1.

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

2.

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

3.

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

4.

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

5.

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

6.

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

7.

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

8.

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

9.

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

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

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

ES-401 16 Form ES-401-2 16 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)

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

K/A Topic(s)

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

Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 X

EA2.01 Ability to determine or interpret the 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) 4.8 76 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 X

AA2.06 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Existence of proper RHR overpressure protection (CFR: 43.5 / 45.13) 3.4 77 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient Without Scram / 1 X

2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13) 4.5 78 000038 (EPE 38) Steam Generator Tube Rupture / 3 X

2.4.6 Knowledge of EOP mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) 4.7 79 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

2.1.32 Ability to explain and apply system limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 4.0 80 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6

ES-401 17 Form ES-401-2 17 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X

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) 3.8 81 (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 NA K/A Category Totals:

3 3

Group Point Total:

6

ES-401 18 Form ES-401-2 18 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

2.4.6 Knowledge of EOP mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) 4.7 82 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 X

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) 4.1 83 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 X

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) 4.2 84 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

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) 4.3 85 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

ES-401 19 Form ES-401-2 19 (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

ES-401 20 Form ES-401-2 20 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 Removal X

A2.03 Ability to (a) predict the impacts of 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: RHR pump/motor malfunction (CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.1 86 006 (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 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) 3.2 87 013 (SF2 ESFAS) Engineered Safety Features Actuation X

2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13) 4.2 88 022 (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 061 (SF4S AFW)

Auxiliary/Emergency Feedwater

ES-401 21 Form ES-401-2 21 062 (SF6 ED AC) AC Electrical Distribution X

2.2.22 Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2) 4.7 89 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel Generator X

A2.03 Ability to (a) predict the impacts of 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) 3.1 90 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control NA K/A Category Point Totals:

3 2

Group Point Total:

5

ES-401 22 Form ES-401-2 22 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (SRO)

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

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication X

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) 2.9 91 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor X

2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) 4.7 92 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment X

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) 4.0 93 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator

ES-401 23 Form ES-401-2 23 055 (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 Ventilation NA K/A Category Point Totals:

2 1

Group Point Total:

3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) SRO Form ES-401-3 24 Facility: Arkansas Nuclear One Unit 2 Date of Exam: April 07, 2021 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.34 Knowledge of primary and secondary plant chemistry limits.

(CFR: 41.10 / 43.5 / 45.12) 3.5 94 2.1.35 Knowledge of the fuel-handling responsibilities of SROs.

(CFR: 41.10 / 43.7) 3.9 95 Subtotal 2

2. Equipment Control 2.2.5 Knowledge of the process for making design or operating changes to the facility.

(CFR: 41.10 / 43.3 / 45.13) 3.2 96 2.2.20 2.2.20 Knowledge of the process for managing troubleshooting activities.

(CFR: 41.10 / 43.5 / 45.13) 3.8 97 Subtotal 2

3. Radiation Control 2.3.11 Ability to control radiation releases.

(CFR: 41.11 / 43.4 / 45.10) 4.3 98 Subtotal 1

4. Emergency Procedures/Plan 2.4.37 Knowledge of the lines of authority during implementation of the emergency plan.

(CFR: 41.10 / 43.1 / 45.13) 4.1 99 2.4.23 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

(CFR: 41.10 / 43.5 / 45.13) 4.4 100 Subtotal 2

Tier 3 Point Total 7

ES-401 Record of Rejected K/As Form ES-401-4 25 ANO Unit 2 April 2021 RO/SRO Exam Record of Rejected K/As Tier /

Group Randomly Selected K/A Reason for Rejection RO Tier 1 Group 1 QID# 7 026 AA1.03 (Original)

Loss of Component Cooling Water (CCW)

Ability to operate and/or monitor the following as they apply to the Loss of Component Cooling Water: - SWS as a backup to the CCWS 026 AA1.07 (New Recommendation)

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 rates to the components and systems that are serviced by the CCWS; interactions among the components Need to reject the original K/A for the following reasons: The Service Water System (SWS) does not provide a backup to any component cooled by CCW.

The SWS does provide cooling to the CCW System Heat Exchangers but a question on this knowledge would overlap another QID (#36) on the RO exam.

We recommend K/A 026 AA1.07 as a replacement as the flow of CCW to the RCPs inside Containment is vital to plant operations.

NRC APPROVED REPLACEMENT ON 09/14/2020 Tier /

Group Randomly Selected K/A Reason for Rejection RO Tier 1 Group 2 QID# 20 033 2.2.44 (Original)

Loss of Intermediate Range Nuclear Instrumentation Equipment Control - Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.

028 2.2.44 (New Recommendation)

Pressurizer Level Control Malfunction Equipment Control - Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.

Need to reject the original 033 system for the following reasons: ANO Unit 2 does not have intermediate range Nuclear Instrumentation. Our normal Power Nuclear Instruments are (3) stacked fission chambers with the middle chamber being a Log channel detector that monitors the intermediate power range during a Reactor Startup. I have another RO QID (#58) that requires knowledge of the Log Power Channel outputs. I also have an SRO QID (83) on the Loss of Source Range Nuclear Instrument - System 032 Therefore, we recommend K/A 028 2.2.44 as a replacement as the Pressurizer Level Control Malfunction system 028 has not been used and has a lot of Control Room Indications that must be interpreted accurately during a failure of the system.

NRC APPROVED REPLACEMENT ON 09/14/2020

ES-401 Record of Rejected K/As Form ES-401-4 26 Tier /

Group Randomly Selected K/A Reason for Rejection RO Tier 1 Group 2 QID# 22 059 2.1.19 (Original)

Accidental Liquid Radwaste Release Conduct of Operations - Ability to use plant computers to evaluate system or component status.

059 2.1.30 (New Recommendation)

Accidental Liquid Radwaste Release Conduct of Operations - Ability to locate and operate components, including local controls.

Need to reject the original K/A for the following reasons: We do not use the Plant Computer system to monitor a Liquid Rad Waste release. We use alarms, trending Chart recorders and local status of pumps/valves to monitor and evaluate the releases.

We recommend K/A 059 2.1.30 as a replacement as a replacement for this QID because several alarms, indications and controls for liquid rad waste releases are local with a common trouble alarm in the control room.

NRC APPROVED REPLACEMENT ON 09/14/2020 Tier /

Group Randomly Selected K/A Reason for Rejection RO Tier 2 Group 1 QID# 46 061 A2.02 (Original)

Auxiliary / Emergency Feedwater (AFW) System 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 air to steam supply valve 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 dc power.

Need to reject the original K/A for the following reasons: The Steam supply valves for the steam driven EFW 2P-7A Pump Turbine are motor operated valves MOVs. The turbine startup valves are also MOVs thus a Loss of air has no effect on the startup of the steam driven EFW 2P-7A Pump Turbine.

We recommend K/A 061 A2.03 as a replacement as the startup MOVs for the steam driven EFW 2P-7A Pump Turbine are powered form DC and this power is needed to start the steam driven EFW 2P-7A Pump Turbine.

NRC APPROVED REPLACEMENT ON 08/24/2020

ES-401 Record of Rejected K/As Form ES-401-4 27 Tier /

Group Randomly Selected K/A Reason for Rejection SRO Tier 1 Group 2 QID# 83 032 AA2.07 (Original)

Loss of Source Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: - Maximum allowable channel disagreement 032 AA2.01 (1st Recommendation)

Loss of Source Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: - Normal/abnormal power supply operation 032 AA2.06 (2nd Recommendation)

Loss of Source Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: - Confirmation of reactor trip Need to reject the original K/A for the following reasons. The Maximum allowable channel disagreement is performed by I&C maintenance during the Source Range NI Functional testing that is 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 Alterations. This is not an item the RO or SRO would check or see in real time. They would verify that the 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 the detector channel disagreement.

We recommend K/A 032 AA2.01 as a replacement as this is something the SRO can see in real time and provide the appropriate direction given a set of plant conditions associated with Fuel handling activities and procedures.

NRC APPROVED 1st RECOMMENDED REPLACEMENT ON 07/23/2020 Need to reject the 1st recommended K/A change due the inability to generate a discriminatory SRO Only level question for K/A 032 AA2.01.

We recommend K/A 032 AA2.06 as a replacement as an SRO needs to know the correct procedure section to implement if confirmation of a reactor trip CANNOT be confirmed in SPTAs.

NRC APPROVED 2nd RECOMMENDED REPLACEMENT ON 11/23/2020

ES-401 Record of Rejected K/As Form ES-401-4 28 Tier /

Group Randomly Selected K/A Reason for Rejection SRO Tier 2 Group 1 QID# 86 005.A2.01 (Original)

Residual Heat Removal System (RHRS)

Ability to (a) predict the impacts of 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.

005.A2.03 (New Recommendation)

Residual Heat Removal System (RHRS)

Ability to (a) predict the impacts of 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: - RHR pump/motor malfunction Need to reject the original K/A for the following reasons. The original question written for the original K/A was determined to be an unsat question due to minutia based on AOP step memorization and not very discriminating for an SRO only question. Also it is difficult to generate a question on the original K/A that would not potentially overlap with similar QIDs 17 and 77 on this exam.

We recommend K/A 005.A2.01 as a replacement K/A as this can expand the possibilities of question development that will be more procedurally driven to evaluate the impact following a malfunction, and then use procedures to mitigate the consequences of a loss of the RHR pump.

NRC APPROVED REPLACEMENT ON 02/04/2021

ES-301 Administrative Topics Outline Form ES-301-1 Revision 0 Facility:

ANO-2 Date of Examination:

3/29/2021 Examination Level: RO SRO Operating Test Number:

2021 Administrative Topic (see Note)

Type Code*

Describe activity to be performed A1. Conduct of Operations 2.1.23 RO (4.3)

D/R Determine CEA#1 Upper Gripper Coil Temperature A2JPM-NRC-ADMIN-CEA A2. Conduct of Operations 2.1.9 RO (2.9)

N/R Determine acceptable control room loading.

A2JPM-NRC-ADMIN-CRLD A3. Equipment Control 2.2.15 RO (3.9)

P/D/R Perform identification of boundary isolations and electrical power to tagout a Boric Acid Makeup Pump A2JPM-NRC-ADMIN-HCRD2 A4. Radiation Control 2.3.11 RO (3.8)

D/R Calculate Containment Purge Release Setpoint 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)

ES-301 Administrative Topics Outline Form ES-301-1 Revision 0 Facility:

ANO-2 Date of Examination:

3/29/2021 Examination Level: RO SRO Operating Test Number:

2021 Administrative Topic (see Note)

Type Code*

Describe activity to be performed A5. Conduct of Operations 2.1.23 SRO (4.4)

D/R Review CEA#1 Upper Gripper Coil Temperature Calculation A2JPM-NRC-ADMIN-XCEA A6. Conduct of Operations 2.1.25 SRO (4.2)

P/D/R Verify RPS trip set point determination for inoperable MSSV A2JPM-NRC-ADMIN-MSSVINOP A7. Equipment Control 2.2.37 SRO (4.6)

M/R Determine operability of Emergency Feedwater System A2JPM-NRC-ADMIN-EFWTS3 A8. Radiation Control 2.3.4 SRO (3.7)

D/R Calculate expected dose for entry during an emergency and determine if entry is allowed.

A2JPM-NRC-ADMIN-EMGRESPSRO A9. Emergency Plan 2.4.41 SRO (4.6)

N/R Determine Emergency Action Level, Time Critical (Rev 6 EAL)

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

Facility: ANO-2 Exam Level: RO X

SRO-I X 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 004 A4.07; RO 3.9 / SRO 3.7 Perform Emergency Boration, Alternate path due to dilution lineup failure requires securing the dilution lineup using an alternate isolation valve.

A/M/EN/L/S 1

Reactivity Control S2.

A2JPM-NRC-SDBC1 041 A4.05; RO 3.1 / SRO 3.3 Perform a restart and reset of SDBCS after power interruption D/L/S 4

Heat Removal Secondary S3.

A2JPM-NRC-CSASV 026 A4.01; RO 4.5 / SRO 4.3 Verify CSAS Attachment 41, Alternate path, CSAS signal fails to auto actuate, after manual actuation some of the required components do not auto align to required position, Manual operation using a verification lineup procedure is required.

A/N/EN/L/S 5

Containment S4.

A2JPM-NRC-ELECXT5 062 A4.01; RO 3.3 / SRO 3.1 Cross Connect 2B3 and 2B4 D/S 6

Electrical S5.

A2JPM-RO-RCP04 003 A2.02; RO-3.7 / SRO-3.9 Perform a normal RCP shutdown Alternate path. When the RCP is secured the anti-rotation device will fail bring in an alarm, the annunciator procedure will be used to secure all RCPs.

A/P/D/L/S 4

Heat Removal Primary S6.

A2JPM-RO-SIT02 006 A1.07; RO 3.3 / SRO 3.6 Adjust SIT 2T-2A Pressure (Raise)

D/EN/S 3

Pressure Control S7.

A2JPM-NRC-CCW04 008 A4.01; RO 3.3 / SRO 3.1 Splitting out loop 1 and 2 CCW N/S 8

Plant Service Systems S8..

A2JPM-NRC-CEA02 012 A4.06; RO 4.3 / SRO 4.3 Test a Reactor Trip Circuit Breaker D/S 7

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

P1.

A2JPM-NRC-IA04 065 AA2.01; RO-2.9 / SRO-3.2 Respond to lowering Instrument Air Pressure. Alternate path, when performing the loss of IA procedure it requires checking DP that has several components between the two pressure indicators, the contingency column must be used to determine the component with high DP and the alternate component placed in service.

A/P/D/E 8

Plant Service Systems P2.

A2JPM-NRC-2D31BSWAP2 063 A3.01: RO-2.7 / SRO 3.1-Align 2D-31B Battery Charger to Green Train and place it in service, Alternate path. 2D-31B is aligned from the alternate train power supply, and placed in service, after the off going battery charger is secure 2D-31B does not maintain load. The alternate charger must be returned to service.

A/L/M 6

Electrical Distribution P3.

A2JPM-NRC-69REL2 2.3.11; RO-3.8 / SRO-4.3 Perform a release of 2T-69A Boric Acid Condensate Tank P/D/R 9

Radioactivity Release

Revision 1 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.

SROU JPMS are bold SROI do all JPMs except S8 (italics

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator (5 / 5 / 3) 4-6 / 4-6 / 2-3 (0 / 0 / 0)

(7 / 6 / 3) 9 / 8 / 4 (1 / 1 / 1) 1 / 1 / 1 (3 / 3 / 1) 1 / 1 / 1 (control room system)

(5 / 5 / 3) 1 / 1 / 1 (4 / 4 / 2) 2 / 2 / 1 (3 / 3 / 2) 3 / 3 / 2 (randomly selected)

(1 / 1 / 1) 1 / 1 / 1 (8 / 7 / 2)

Revision 1 This is an un-official (for examiner info only) overview of the JPMs selected for the outline.

Simulator / In-plant JPMs:

S1: Perform Emergency Boration. This is an alternate path JPM. When initiating emergency boration, 2CV-4927 Reactor Makeup Water Flow Control Valve will fail open aligning non-borated water to the charging pump suction, thereby diluting the RCS. The applicant must then perform a contingency action of closing 2CV-4941-2 VCT Makeup Isolation Valve to ensure that no reactor makeup water is diluting the emergency boration line up. This is a Modified JPM. The original JPM the VCT outlet valve failed to close and the applicant had to use a Boric Acid Makeup pump for emergency boration.

S2: Perform a restart/reset of SDBCS after a power interruption. A temporary loss of power has caused the Steam Dump Bypass Control System (SDBCS) to revert to manual operation. The SDBCS will be setup in manual steaming to the condenser through 2CV-0303 controlling SG pressure. The operator must align the SDBCS panel for a self-test and then restore the system to automatic operation controlling SG pressure in automatic to the condenser.

S3: Ensure Containment Spray Actuation Signal (CSAS) actuation. This is an alternate path JPM. With a LOCA in progress, containment pressure is observed to be greater than CSAS actuation setpoint. The system fails to automatically actuate requiring manual actuation of CSAS and a board verification of actuated components. Two required components will not be in the required positions requiring examinee to start a room cooler and close a Main Steam Isolation Valve.

S4: Non-ESF 480V busses 2B-3 and 2B-4 will be cross connected for maintenance. This will involve synching the busses and ensuring loading limits are maintained. Then, opening the supply breakers from one of the supply busses.

S5: Perform a normal RCP shutdown. This JPM is alternate path. When securing the 1st RCP 2P-32A, a valid Reverse Rotation Alarm will come in which will require a transition to the Alarm Corrective Action procedure which requires securing all RCPs for a Valid alarm to stop the reverse rotation on 2P-32A. The alternate path portion is time critical based on EN-OP-123 Time critical and Time sensitive actions.

S6: Adjust SIT 2T-2A Pressure (RAISE). This JPM requires the applicant to align nitrogen to raise safety injection tank pressure while monitoring all of the safety injection tank pressures. When the desired pressure is reached, they will secure the nitrogen lineup ensure that safety injection tank low pressure alarm is clear and pressure is not above the allowed Tech Spec pressure.

S7: Splitting out Loop 1 and 2 CCW. The applicant will align the CCW system for 2 loop operation split out from 1 loop operation. They will close the system crosstie valves, ensure loop 2 CCW is still operating properly, direct a NLO to align manual valves in the field, direct a NLO to vent 2P-33A CCW pump and then start 2P-33A CCW pump to restore loop 1 CCW flow.

S8: Perform Reactor Trip Circuit Breaker Testing on TCB-02 following maintenance using OP-2105.009 Supplement 1. Section 1 initial conditions have been completed. The applicant will use Supplement 1 to test that TCB-2 will open by either shunt (energized trip) or UV (undervoltage trip). TCB-02 is open initially, the examinee will close TCB-02 on 2C23 in the back of the control room and then communicate with the local operator while testing both the Shunt and Undervoltage Trips on the TCB while observing appropriate response from the TCB opening.

P1: Respond to lowering Instrument Air pressure. This is an Alternate Path JPM. The lowering IA header pressure is due to high DP across the in-service IA Filter 2F-173A. When performing step 8 of the IA AOP and checking the DP of the IA header pressure and receiver tanks pressure, the DP will be greater than 10 psid. Step 8 Contingency Actions A and B will fail to lower the IA to receiver tanks DP thus per Contingency Action Step 8.C, a transition will be made to OP 2104.024, IA System OPS, and the standby IA Filter will be placed in service which will restore header pressure back to normal. The Loss of IA AOP procedure Step 7 and 8 should be handed out first. When examinee, gets to Contingency Action Step 8.C, then provide Section 14.1 of the IA System Operating Procedure OP-2104.024.

Revision 1 P2: Align 2D-31B Battery Chargers to Green Train and place it in service. This is an alternate path JPM. The plant is defueled and 2D-31B is to be powered from the opposite (green) train due to maintenance. The examinee will align 2D-31B to Green Train, and then proceed to perform a parallel transfer from 2D-31A to 2D-31B in service. 2D-31B will not function properly leading the examinee to re-start 2D-31A and place it back in service. This is Modified from a previous alternate path JPM. The changes include addition of aligning alternate power to 2D-31B and the previous JPM when 2D-31B was placed in service it never indicated amps. The modified JPM when 2D-31B is placed in service it indicated amp properly but when 2D-31A is secured and additional load is placed on 2D-31 it fails.

P3: Perform a release of 2T-69A Boric Acid Condensate Tank. The applicant will commence the release of 2T-69A by verifying that the pump 2P-47A is running then remove the red tag from the discharge isolation valve 2CV-2318 and opening the valve. The applicant will then align the automatic dump isolation valves 2CV-2330A and 2CV-2330B to OPEN. Next a manual isolation will be opened to commence the release. The applicant will then commence throttling the 2P-47A pump recirc isolation to achieve the desired release rate.

Appendix D Scenario # 2 Form ES-D-1 Revision 2 Page 1 of 58 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 No.

Malf. No.

Event Type*

Event Description 1

NIBLOGPWR C (BOP / CRS)

TS (CRS)

B channel Log power fails low then fails high.

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 C (CRS / BOP/

ATC)

TS (CRS)

Shutdown required due to 4 gpm steam generator tube leak on A steam generator.

OP-2203.038 Primary to Secondary Leakage AOP.

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

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 SGATUBE 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 CV48202 CV48211 C (ATC / CRS)

Letdown fails to isolate on SIAS.

OP-2202.010, Standard Attachments EOP.

8 CV0231 C (BOP / CRS)

Gland seal regulator 2PCV-0231 fails closed.

2203.012B, Annunciator 2K-02 Corrective Action (ACA)

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario # 2 Form ES-D-1 Revision 2 Page 2 of 58 Target Quantitative Attributes (Section D.5.d)

Actual Attributes Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 5 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per scenario set) 0 Critical Tasks (2) 2 Critical Task Justification Cueing Measurable Performance Indicators Performance Feedback References CT-1:

Secure A RCP within 10 min of the reactor trip.

Exceeding operating limits has the potential to degrade the RCS pressure boundary. RCPs should be maintained in an available condition for last resort use if needed.

2K11 F1 A RCP Upper/Lower Oil RSVR Level Lo.

RCP bearing temperature rising >18 degrees per minute on computer trends or 2K11 B1 A RCP Upper Thrust BRG Metal Temperature Hi in alarm.

Secured A RCP by securing placing handswitch to stop or PTL.

A RCP amps indicate zero.

Green light on and Red light off above A RCP handswitch.

CT-23, Trip any RCP exceeding operating limits (SGTR-03)

CT-2:

Isolate A SG (2202.010 0 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 or upon completion of 2202.010 Attachment 10 whichever comes first.

Isolating the SG will minimize the potential loss of the containment boundary, thus preventing an offsite release and exceeding 10CFR100 exposure limits at the site boundary.

Procedural direction when RCS Thot is less than 535 degrees F on A SG.

Completion of Standard Att. 10. All valves are closed by placing handswitches to closed.

MSIV HEADER #1 2SV 1010 1A, 2C17 MSIV HEADER #1.

2SV 1010 2A, 2C16 FEEDWATER BLOCK VALVE TO SG A 2CV 1024 1, 2C17 FEEDWATER BLOCK VALVE TO SG A 2CV 1023 2, 2C16 A SG Pressure trending up while B SG pressure continues to lower from the cooldown.

Green light on, Red Light off above valves listed in Att. 10.

CT-14, Isolate most affected SG (SGTR-09).

  • SAR Section 15.1.18
  • EOP 2202.004, SGTR Tech Guide 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.

Appendix D Scenario # 3 Form ES-D-1 Revision 1 Page 1 of 52 Facility: ANO-2 Scenario No.: 3 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 No.

Malf. No.

Event Type*

Event Description 1

N (BOP / CRS)

Shift EH pumps from 2P-14A to 2P-14B OP-2106.012 Electrohydraulic Oil system.

2 XRCCHAPLVL I (ATC / CRS)

TS (CRS)

A Channel Pressurizer Level channel fails low.

OP-2203.028, Pressurizer System Malfunction AOP 3

CCW2P33BPWR CCW2P33CPWR C (BOP / ATC /

CRS) 2P-33C CCW pump trips and 2P-33B CCW pump fails to start. CT-1 OP-2203.025, RCP Emergencies AOP 4

RCP2P32CSLK C (ATC / BOP /

CRS)

TS (CRS)

C Reactor Coolant Pump (RCP) develops an intersystem leak from the RCS to CCW of 15 gpm. TS for CRS.

OP-2203.016, Excess RCS leakage AOP 5

RCP2P32CSLK ESFK202AAF ESFK202BAF M (All)

C RCP intersystem leak degrades to 300 gpm. CCW to RCPs fail to auto close on CIAS. CT-2, & 3 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.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario # 3 Form ES-D-1 Revision 1 Page 2 of 52 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 scenario set) 0 Critical Tasks (2) 3 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback References CT-1:

Restore CCW flow to the RCP within 10 minutes of the loss of CCW flow.

Exceeding operating limits has the potential to degrade the RCS pressure boundary. RCPs should be maintained in an available condition for last resort use if needed.

2K-11 A1, A3, A5, A7 CCW Disch flow lo alarms in.

On Panel 2C14 the BOP opens the CCW crosstie valves 2CV-5220 and 2CV-5230.

And Start a backup CCW pump restoring flow. 2P-33A preferred.

2CV-5220 and 2CV-5230 will open (Red light on, green light off) and CCW Disch flow lo alarms clear (2K-11 A1, A3, A5, A7) after the CCW pump is started.

  • EN-OP-123 Time Critical Action/Time Sensitive Action Program.
  • CR-ANO-2-2010-948, Critical task criteria
  • RCP emergencies Tech guide
  • TDB580.0040 RCP Tech Manual CT-2:

Isolate RCS leakage from leaving primary containment by closing CCW containment isolation valves prior to completion of SPTAs Loss of pressurizer level takes away the operator's most direct means of monitoring RCS inventory, lowers pressurizer pressure, and reduces RCS subcooling, all of which jeopardize the RCS Inventory and Pressure Control safety functions. If the RCS mass loss is not isolated or reduced to less than the makeup capacity of the SI pumps, then core uncovery and core damage will occur.

CCW surge tank level and radiation will trend up indicating an RCS leak into CCW. The crew assesses RCS leakage rates.

Procedure direction directs the operators to trip the reactor and isolate the leakage > 44 gpm.

On panel 2C16 and 2C17 will close the three CCW to RCP CNTMT isolation valves.

2CV-5236-1 2CV-5254-2 2CV-5255-1 2CV-5236-1, 2CV-5254-2, &

2CV-5255-1 closed (Red light off, green light on)

CCW surge tank level stops rising, and leakage is indicated inside containment (dew point, sump level, radiation trending up

CT-13, Isolate RCS Leakage (LOCA-05)

  • EN-OP-123 Time Critical Action/Time Sensitive Action Program.
  • CR-ANO-2-2010-948, Critical task criteria

Appendix D Scenario # 3 Form ES-D-1 Revision 1 Page 3 of 52 CT-3:

Commence an RCS cooldown within 30 minutes of entry into OP-2202.003, LOCA EOP.

Cooling down and depressurizing the RCS removes decay heat and lowers the DP at the break, slowing the leak rate and reducing makeup volume required.

SDC entry conditions are also required for long-term cooling.

Procedural Direction On panel 2C02, the crew operates the turbine bypass valves to the condenser to lower SG pressure and thus lower RCS temperature SG pressure and RCS temperature indicators begin lowering

  • CE EPGB Simulator CTs: CT-20, Cool down and depressurize RCS (LOCA-09)
  • CR-ANO-2-2010-948, Critical task criteria 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.

Appendix D Scenario 4 Form ES-D-1 Revision 1 Page 1 of 44 Facility: ANO-2 Scenario No.: 4 Op-Test No.: 2021 Examiners: __________________________ Operators:

Initial Conditions: _100% power. 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. TRM 3.7.6 action d for A1 unavailable to CFW was entered 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

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 No.

Malf. No.

Event Type*

Event Description 1

N (BOP/CRS)

Drain the containment Sump.

OP-2104.014, LRW and BMS Operations 2

DI_HS_2060_2 TS(CRS)

Containment sump drain valve 2CV-2060-1 (containment penetration isolation) trips breaker with valve still open.

OP-2203.012G Annunciator 2K07 Corrective Actions.

3 K12-H04 C (ATC/CRS) 2K12-H4 Reactor Makeup Water Tank Temperature Lo OP-2203.012L Annunciator 2K12 Corrective Actions 4

XRRPZRLSP I (ATC/CRS)

Reactor Reg. output to PZR level control program fails to 41%.

OP-2203.028, Pressurizer System Malfunction AOP 5

XFW2FIS0735 C (BOP/CRS) 2P-1A Feedwater pump suction flow transmitter fails.

OP-2203.012 Annunciator 2K03 Corrective Action 6

MFWPMPATRP C (ALL)

TS (CRS) 2P-1A Main Feedwater Pump OP-2203.027 Loss of Main Feedwater Pump AOP 7

BUS2A1 C (ALL) 2A-1 bus lockout requiring a reactor trip.

8 500LOSE500 500LOSE161 CV0336 EFW2P7BFLT M (ALL)

Loss of Offsite Power, 2P-7A EFW pump overspeed trips, and 2P-7B EFW pump has a motor fault causing a complete loss of feedwater. CT-1, & 2 OP-2202.006 Loss of Feedwater EOP 9

HPI2P89BFAL ESFK409AAF C (BOP/CRS) 2P-89B High Pressure Safety Injection pump fails to 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.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario 4 Form ES-D-1 Revision 1 Page 2 of 44 Target Quantitative Attributes (Section D.5.d)

Actual Attributes Malfunctions after EOP entry (1-2) 1 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 set) 1 Critical Tasks (2) 3 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback References CT-1:

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 5F based on average CET temperature.

Adequate time to establish OTC following indications The 5 degrees allows for small temperature rises due to fluctuations in feedwater and steam flow. This is consistent with the CEN-152 basis.

If OTC is required, this means that all other available RCS and Core Heat Removal heat sinks have been lost or 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.

Procedural Direction.

Both Steam Generator Level Wide Range level indication Less than or equal to 70.

RCS CET Temperature uncontrolled heat up of 5F on SPDS.

One full train of HPSI placed in service:

One HPSI Pump is running.

All four Cold Leg injection Valves Verified Open.

ECCS Vent Valves verified Energized and Open on 2C-09.

HPSI flow indication.

ECCS Vent Valves 2CV-4698-1 and 2CV-4740-1 Position Indicate Open.

Quench Tank Level,

Pressure, Temperature Rising.

CT-04 Establish once through cooling (HR-07)

Appendix D Scenario 4 Form ES-D-1 Revision 1 Page 3 of 44 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback References CT-2:

Maintain RCS Pressure within PT Curves (30 to 200 degrees MTS, less than 2500 psia) prior to establishing once through cooling.

Loss of RCS pressure control low will result in a loss of subcooling. Loss of RCS pressure control high could result in a pressurized thermal shock condition.

Procedural Direction.

SPDS PT screen display.

2K-10 E6, E7 CNTRL CH 1/2 Pressure HI/LO alarm.

Energizes PZR heaters and/or align for auxiliary spray to control RCS pressure.

RCS Margin to Saturation is controlled between 30 and 200 degrees displayed on SPDS and RCS pressure does not exceed 2500 psia.

CT-06 Establish RCS Pressure Control (LOAF-06, SPTA-05)

CT-3:

Establish one full train of HPSI prior to zero degrees MTS.

SI flow keeps the core covered, cooled, and borated.

Inadequate HPSI flow could result in eventual core uncover and fuel damage.

Procedural Direction.

HPSI flow is throttled and HPSI throttle criteria are NOT met.

HPSI Pump not running after SIAS.

HPSI Injection Valves not open following SIAS.

On 2C16/2C17, Start HPSI pump(s) not running following SIAS.

On 2C16/2C17 Open HPSI Injection Valve(s) not open following SIAS.

On 2C16/2C17 indications of HPSI Pressure.

On 2C16/2C17 Open indication on all 4 cold leg injection valves.

CT-16 Establish required SI flow (IC-03)

  • EN-OP-123 Time Critical Operation Actions, 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.

Appendix D Scenario 5 Form ES-D-1 Revision 1 Page 1 of 45 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 LVDT repair and DEFAS cabinet repair. Power is being maintained at 3 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 No.

Malf. No.

Event Type*

Event 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)

TS (CRS) 2PT-1041-2 SG-A pressure detector fails low.

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/

CRS)

TS(CRS)

Inadvertent Containment Isolation Actuation Signal (CIAS) on the Green Train. CT-1 OP-2203.039, Inadvertent CIAS 6

LOSE161 LOSE500 M (ALL)

Loss of Offsite Power. CT-2 OP-2202.001, Standard Post Trip Actions (SPTAs) EOP OP-2202.007, Loss of Offsite Power EOP 7

EDG2AUTOFAIL C (BOP/ CRS)

  1. 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

Appendix D Scenario 5 Form ES-D-1 Revision 1 Page 2 of 45 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 set) 0 Critical Tasks (2) 3 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback References CT-1:

Component Cooling Water (CCW) to RCPs must be restored within 10 minutes of the loss of cooling water.

Exceeding operating limits has the potential to degrade the RCS pressure boundary. RCPs should be maintained in an available condition for last resort use if needed.

If RCPs are allowed to operate for 10 minutes without CCW flow. OP-1015.050 requires RCPs not meeting operating limits to be secured within 10 minutes.

2K-11 A1, A3, A5, A7 CCW Disch flow lo alarms in.

On Panel 2C-16 place the CCW to RCP valve hand switch to close then to open for 2CV-5254-2 2CV-5254-2 will open (Red light on) and CCW Disch flow lo alarms clear (2K-11 A1, A3, A5, A7)

CT-23, Trip any RCP exceeding operating limits (LOCA-04)

  • AOP OP-2203.039 Inadvertent CIAS.

CT-2:

Crew should establish and maintain the RCS within the limits of the PT Curve (<2000 F and >300F MTS. Also RCS pressure does not exceed 2500 psia.)

Loss of RCS pressure control will result in a loss of RCS subcooling and a reactor head void can form, both of which complicate the event recovery.

Uncontrolled void growth could result in eventual core uncover and fuel damage.

Procedural Direction.

SPDS PT screen display.

2K-10 E6, E7 CNTRL CH 1/2 Pressure HI/LO alarm.

Energizes PZR heaters and/or align for auxiliary spray to control RCS pressure.

RCS Margin to Saturation is controlled between 30 and 200 degrees displayed on SPDS and RCS pressure does not exceed 2500 psia.

  • CE EPGB Simulator CT CT-06 Establish RCS Pressure Control (LOOP-
04)

Appendix D Scenario 5 Form ES-D-1 Revision 1 Page 3 of 45 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback References CT-3:

Crew should ensure that at least one vital 4160V bus is energized prior to completion of SPTAs.

Without any AC power available for ESF pumps, the ability to maintain the plant in a safe state is severely degraded since no makeup water can be added to the RCS for inventory control purposes. Therefore, energizing at least one vital (4.16 kV)

AC bus in the SPTAs is essential if it can be done quickly.

2A-4 bus is de-energized zero volts on 2A4 due #2 EDG failure to Auto start.

2K09-A3 alarm in and 2A-4 bus voltmeter indicating zero 2A-3 bus is de-energized zero volts due to securing #1 EDG because of not service water cooling.

2K08-A3 alarm in and 2A-3 bus voltmeter indicating zero

  1. 2 EDG start Handswitch is taken to start.

OR AACDG Started and aligned to 2A-3 2K09-A3 alarm clears and 2A-4 bus voltmeter indicating approximately 4160 V.

CT-03, Energize at least one vital AC bus (SPTA-

02)
  • 10 CFR 50.63, Station Blackout 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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