ML15266A070

2015-09 - Final Outlines

ML15266A070
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/09/2015
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML15266A070 (41)

Text

ES-401 PWR Examination Outline (RO) Form ES-401-2 Facility: Arkansas Nuclear One, Unit 2 Date of Exam: August 21, 2015 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 3 3 6 2 3 1 18 6 Emergency &

Abnormal Plant 2 1 1 2 N/A 3 1 N/A 1 9 4 Evolutions Tier Totals 4 4 8 5 4 2 27 10 1 2 2 2 3 1 2 4 4 2 2 4 28 5 2.

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

3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 3 2 2 3 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A 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 associated 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 importance ratings (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 Rev 3

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

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000007 Reactor Trip - Stabilization - X EA2.03 Ability to determine or interpret the 4.2 1 Recovery / 1 following as they apply to a reactor trip: Reactor trip breaker position (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000008 Pressurizer Vapor Space X AA2.28 Ability to determine and interpret the 3.3 2 Accident / 3 following as they apply to the Pressurizer Vapor Space Accident: Safety parameter display system indications (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000009 Small Break LOCA / 3 X EK3.22 Knowledge of the reasons for the following 4.4 3 responses as the apply to the small break LOCA:

Maintenance of heat sink (CFR: 41.5 / 41.10 / 45.6 / 45.13) 000011 Large Break LOCA / 3 X 2.4.8 Knowledge of how abnormal operating 3.8 4 procedures are used in conjunction with EOPs.

(CFR: 41.10 / 43.5 / 45.13) 000015/17 RCP Malfunctions / 4 X AK3.07 Knowledge of the reasons for the 4.1 5 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 Loss of Rx Coolant Makeup / 2 X AK1.03 Knowledge of the operational implications 3.0 6 of the following concepts as they apply to Loss of Reactor Coolant Makeup: Relationship between charging flow and PZR level (CFR: 41.8 / 41.10 / 45.3) 000025 Loss of RHR System / 4 X AK1.01 Knowledge of the operational implications 3.9 7 of the following concepts as they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation (CFR: 41.8 / 41.10 / 45.3) 000026 Loss of Component Cooling X AK3.02 Knowledge of the reasons for the 3.6 8 Water / 8 following responses as they apply to the Loss of Component Cooling Water: The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS (CFR: 41.5 / 41.10 / 45.6 / 45.13) 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 X EK2.06 Knowledge of the interrelations between 2.9* 9 the and the following an ATWS: Breakers, relays, and disconnects (CFR: 41.7 / 45.7) 000038 Steam Gen. Tube Rupture / 3 X EK3.01 Knowledge of the reasons for the 4.1 10 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)

Rev 3

ES-401 3 Form ES-401-2 CE/E05 Steam Line Rupture - Excessive X EK2.2 Knowledge of the interrelations between 3.7 11 Heat Transfer / 4 the (Excessive Heat Transfer) and the following:

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

(CFR: 41.10 / 45.13) 000054 Loss of Main Feedwater / 4 X AA2.05 Ability to determine and interpret the 3.5 12 following as they apply to the Loss of Main Feedwater (MFW): Status of MFW pumps, regulating and stop valves (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000055 Station Blackout / 6 X EK3.02 Knowledge of the reasons for the 4.3 13 following responses as the apply to the Station Blackout: Actions contained in EOP for loss of offsite and onsite power (CFR: 41.5 / 41.10 / 45.6 / 45.13) 000056 Loss of Off-site Power / 6 X AK3.01 Knowledge of the reasons for the 3.5 14 following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer (CFR: 41.5 / 41.10 / 45.6 / 45.13) 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 X AK1.01 Knowledge of the operational implications 2.8 15 of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation (CFR: 41.8 / 41.10 / 45.3) 000062 Loss of Nuclear Svc Water / 4 X AA1.07 Ability to operate and / or monitor the 2.9 16 following as they apply to the Loss of Nuclear Service Water (SWS): Flow rates to the components and systems that are serviced by the SWS; interactions among the components (CFR: 41.7 / 45.5 / 45.6) 000065 Loss of Instrument Air / 8 X AA1.02 Ability to operate and/or monitor the 2.6 17 following as they apply to the Loss of Instrument Air:

Components served by instrument air to minimize drain on system (CFR: 41.7 / 45.5 / 45.6) 000077 Generator Voltage and Electric X AK2.02 Knowledge of the interrelations between 3.1 18 Grid Disturbances / 6 Generator Voltage and Electric Grid Disturbances and the following: Breakers, relays (CFR: 41.4 / 41.5 / 41.7 / 41.10 / 45.8)

K/A Category Totals: 3 3 6 2 3 1 Group Point Total: 18 Rev 3

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 X AA1.07 Ability to operate and / or monitor 3.3 19 the following as they apply to the Continuous Rod Withdrawal: RPI (CFR: 41.7 / 45.5 / 45.6) 000003 Dropped Control Rod / 1 X AK3.07 Knowledge of the reasons for the 3.8* 20 following responses as they apply to the Dropped Control Rod: Tech-Spec limits for T-ave (CFR: 41.5 / 41.10 / 45.6 / 45.13) 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 X AK3.03 Knowledge of the reasons for the 3.1 21 following responses as they apply to the Steam Generator Tube Leak: Comparison of makeup flow and letdown flow for various modes of operation (CFR: 41.5 / 41.10 / 45.6 / 45.13) 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 X AA2.05 Ability to determine and interpret 3.6 22 the following as they apply to the Accidental Liquid Radwaste Release: The occurrence of automatic safety actions as a result of a high PRM system signal (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 X AK1.02 Knowledge of the operational 3.1 24 implications of the following concepts as they apply to Plant Fire on Site: Fire fighting (CFR: 41.8 / 41.10 / 45.3) 000068 Control Room Evac. / 8 X AA1.28 Ability to operate and / or monitor 3.8 25 the following as they apply to the Control Room Evacuation: PZR level control and pressure control (CFR: 41.7 / 45.5 / 45.6) 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 AA1.04 Ability to operate and/or monitor the 3.2 23 following as they apply to the High Reactor X Coolant Activity: - Failed fuel-monitoring equipment (CFR: 41.7 / 45.5 / 45.6)

CE/A13 Natural Circ. / 4 X 2.1.19 Ability to use plant computers to 3.9 26 evaluate system or component status.

(CFR: 41.10 / 45.12)

Rev 3

ES-401 5 Form ES-401-2 CE/A11RCS Overcooling - PTS / 4 X AK2.1 Knowledge of the interrelations 3.2 27 between the (RCS Overcooling) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(CFR: 41.7 / 45.7)

CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 1 1 2 3 1 1 Group Point Total: 9 Rev 3

ES-401 6 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump X 2.4.11 Knowledge of abnormal 4.0 28 condition procedures.

(CFR: 41.10 / 43.5 / 45.13) 004 Chemical and Volume X K3.04 Knowledge of the effect that a 3.7 29 Control loss or malfunction of the CVCS will have on the following: RCPS (CFR: 41.7 / 45.6) 004 Chemical and Volume X K6.09 Knowledge of the effect of a 2.8 30 Control loss or malfunction on the following CVCS components: Purpose of VCT divert valve (CFR: 41.7 / 45.7) 005 Residual Heat Removal X K2.01 Knowledge of bus power 3.0 31 supplies to the following: RHR pumps (CFR: 41.7) 006 Emergency Core Cooling X A2.02 Ability to (a) predict the 3.9 32 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: Loss of flow path (CFR: 41.5 / 43.5 / 45.3 / 45.13) 007 Pressurizer Relief/Quench X K5.02 Knowledge of the operational 3.1 33 Tank implications of the following concepts as the apply to PRTS: Method of forming a steam bubble in the PZR (CFR: 41.5 / 45.7) 008 Component Cooling Water X A2.04 Ability to (a) predict the 3.3 34 impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PRMS alarm (CFR: 41.5 / 43.5 / 45.3 / 45.13) 008 Component Cooling Water X K1.02 Knowledge of the physical 3.3 35 connections and/or cause-effect relationships between the CCWS and the following systems: Loads cooled by CCWS (CFR: 41.2 to 41.9 / 45.7 to 45.8) 010 Pressurizer Pressure Control X K4.01 Knowledge of PZR PCS 2.7 36 design feature(s) and/or interlock(s) which provide for the following: Spray valve warm-up (CFR: 41.7)

Rev 3

ES-401 7 Form ES-401-2 012 Reactor Protection X A4.02 Ability to manually operate 3.3 37 and/or monitor in the control room:

Components for individual channels (CFR: 41.7 / 45.5 to 45.8) 013 Engineered Safety Features X A1.09 Ability to predict and/or 3.4 38 Actuation monitor changes in parameters (to Prevent exceeding design limits) associated with operating the ESFAS controls including: T-hot (CFR: 41.5 / 45.5) 022 Containment Cooling X A2.03 Ability to (a) predict the impacts 2.6 39 of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Fan motor thermal overload/high-speed operation (CFR: 41.5 / 43.5 / 45.3 / 45.13) 022 Containment Cooling X K2.02 Knowledge of power supplies to 2.5* 40 the following: Chillers (CFR: 41.7) 025 Ice Condenser 026 Containment Spray X 2.4.21 Knowledge of the parameters 4.0 41 and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12) 026 Containment Spray X K4.07 Knowledge of CSS design 3.8* 42 feature(s) and/or interlock(s) which provide for the following: Adequate level in containment sump for suction (interlock)

(CFR: 41.7) 039 Main and Reheat Steam X 2.4.34 Knowledge of RO tasks 4.2 43 performed outside the main control room during an emergency and the resultant operational effects.

(CFR: 41.10 / 43.5 / 45.13) 059 Main Feedwater X A3.02 Ability to monitor automatic 2.9 44 operation of the MFW, including:

Programmed levels of the S/G (CFR: 41.7 / 45.5) 059 Main Feedwater X A4.03 Ability to manually operate and 2.9* 45 monitor in the control room: Feedwater control during power increase and decrease (CFR: 41.7 / 45.5 to 45.8)

Rev 3

ES-401 8 Form ES-401-2 061 Auxiliary/Emergency X A1.04 Ability to predict and/or 3.9 46 Feedwater monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including: AFW source tank level (CFR: 41.5 / 45.5) 062 AC Electrical Distribution X K1.03 Knowledge of the physical 3.5 47 connections and/or cause-effect relationships between the ac distribution system and the following systems: DC distribution (CFR: 41.2 to 41.9 / 45.7 to 45.8) 063 DC Electrical Distribution X A2.01 Ability to (a) predict the 2.5 48 impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Grounds (CFR: 41.5 / 43.5 / 45.3 / 45.13) 063 DC Electrical Distribution X A1.01 Ability to predict and/or 2.5 49 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) 064 Emergency Diesel Generator X K6.08 Knowledge of the effect of a 3.2 50 loss or malfunction of the following will have on the ED/G system: Fuel oil storage tanks (CFR: 41.7 / 45.7) 064 Emergency Diesel Generator X K4.11 Knowledge of ED/G system 3.5 51 design feature(s) and/or interlock(s) which provide for the following:

Automatic load sequencer: safeguards (CFR: 41.7) 073 Process Radiation X K3.01 Knowledge of the effect that a 3.6 52 Monitoring loss or malfunction of the PRM system will have on the following: Radioactive effluent releases (CFR: 41.7 / 45.6) 076 Service Water X A1.02 Ability to predict and/or 2.6* 53 monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures (CFR: 41.5 / 45.5) 078 Instrument Air X 2.1.28 Knowledge of the purpose and 4.1 54 function of major system components and controls.

(CFR: 41.7)

Rev 3

ES-401 9 Form ES-401-2 103 Containment X A4.06 Ability to manually operate and/or 2.7* 55 monitor in the control room: Operation of the containment personnel airlock door (CFR: 41.7 / 45.5 to 45.8)

K/A Category Point Totals: 2 2 3 3 1 2 4 4 2 2 4 Group Point Total: 28 Rev 3

ES-401 10 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive X K6.02 Knowledge of the effect of a loss 2.8 56 or malfunction on the following CRDS components: Purpose and operation of sensors feeding into the CRDS (CFR: 41.7 / 45.7) 002 Reactor Coolant X A4.07 Ability to manually operate 2.8 57 and/or monitor in the control room: Flow path linking the RWST through the RHR system to the RCS hot legs for gravity refilling of the refueling cavity (CFR: 41.7 / 45.5 to 45.8) 011 Pressurizer Level Control 014 Rod Position Indication X A2.03 Ability to (a) predict the impacts 3.6 58 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: Dropped rod (CFR: 41.5 / 43.5 / 45.3 / 45.13) 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal X K1.01 Knowledge of the physical 3.4* 59 connections and/or cause-effect relationships between the CIRS and the following systems: CSS (CFR: 41.2 to 41.9 / 45.7 to 45.8) 028 Hydrogen Recombiner and Purge Control 029 Containment Purge X A3.01 Ability to monitor automatic 3.8 60 operation of the Containment Purge System including: CPS isolation (CFR: 41.7 / 45.5) 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment X A4.02 Ability to manually operate 3.5 61 and/or monitor in the control room: -

Neutron levels (CFR: 41.7 / 45.5 to 45.8) 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator Rev 3

ES-401 11 Form ES-401-2 055 Condenser Air Removal X K3.01 Knowledge of the effect that a 2.5 62 loss or malfunction of the CARS will have on the following: Main condenser (CFR: 41.7 / 45.6) 056 Condensate 068 Liquid Radwaste X K5.04 Knowledge of the operational 3.2 63 implication of the following concepts as they apply to the Liquid Radwaste System:

Biological hazards of radiation and the resulting goal of ALARA (CFR: 41.5 / 45.7) 071 Waste Gas Disposal 072 Area Radiation Monitoring X 2.1.23 Ability to perform specific system 4.3 64 and integrated plant procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6) 075 Circulating Water X K2.03 Knowledge of bus power 2.6* 65 supplies to the following:

Emergency/essential SWS pumps (CFR: 41.7) 079 Station Air 086 Fire Protection K/A Category Point Totals: 1 1 1 0 1 1 0 1 1 2 1 Group Point Total: 10 Rev 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Arkansas Nuclear One, Unit 2 Date of Exam: August 21, 2015 Category K/A # Topic RO SRO-Only IR # IR #

2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, A no-solo@

3.3 66

1. operation, maintenance of active license status, 10CFR55, etc.

Conduct (CFR: 41.10 / 43.2) of Operations 2.1.29 Knowledge of how to conduct system lineups, such as valves, 4.1 67 breakers, switches, etc. (CFR: 41.10 / 45.1 / 45.12) 2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, 2.7 68 Operations memos, etc. (CFR: 41.10 / 45.12)

Subtotal 3 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 3.7 69 2.2.41 Ability to obtain and interpret station electrical and mechanical 3.5 70

2. drawings. (CFR: 41.10 / 45.12 / 45.13)

Equipment Control Subtotal 2 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10) 3.8 71 2.3.15 Knowledge of radiation monitoring systems, such as fixed

3. radiation monitors and alarms, portable survey instruments, 2.9 72 Radiation Control personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 / 45.9)

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

3.7 73 45.4) 4.

Emergency 2.4.13 Knowledge of crew roles and responsibilities during EOP usage.

4.0 74 Procedures / Plan (CFR: 41.10 / 45.12) 2.4.32 Knowledge of operator response to loss of all annunciators.

3.6 75 (CFR: 41.10 / 43.5 / 45.13)

Subtotal 3 Tier 3 Point Total 10 Rev 3

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 008 Pressurizer Vapor Limited applicability to computer alarms for ANO2. Also, some overlap with Q76 Space Accident on the SRO outline.

Ability to determine and Re-sampled from available AA2 K/As in the same E/APE.

interpret the following as 1/1 they apply to the Pressurizer Vapor Space Accident:

Q2 AA2.04 High-temperature computer alarm and alarm type 065 Loss of Instrument Air This component, emergency air compressor, is not applicable at ANO2.

Ability to operate and / or Re-sampled from available AA1 K/As in the same E/APE.

1/1 monitor the following as they apply to the Loss of Q17 Instrument Air:

AA1.04 Emergency air compressor 001 Continuous Rod Difficulty level for available questions too low for RO level. Also similarities to Withdrawal typical questions for Q1.

Ability to operate and / or Re-sampled from remaining available AA1 K/As in the same E/APE.

1/2 monitor the following as they apply to the Continuous Rod Q19 Withdrawal:

AA1.05 Reactor trip switches 008 Component Cooling Overlap with Q34 due to both tie to makeup water to CCW .

Water Re-sampled from remaining K1 topics in the same system.

Knowledge of the physical 2/1 connections and/or cause-effect relationships between Q35 the CCWS and the following systems:

K1.05 Sources of makeup water 012 Reactor Protection Questions for this K/A are similar to questions used for the K/A selected for Q1.

2/1 Ability to manually operate Re-sampled from the A4 group.

and/or monitor in the control Q37 room:

A4.07 M/G set breakers 026 Containment Spray There are no immediate action ties to this system for ANO2.

Generic K/A 2.4.49 Re-sampled from the Generic K/A list in ES-401 D.1.b.

2/1 Ability to perform without reference to procedures Q41 those actions that require immediate operation of system components and controls.

Rev 3

ES-401 Record of Rejected K/As Form ES-401-4 078 Instrument Air There are no safety related components in the ANO2 Instrument Air system.

Generic K/A 2.2.37 Re-sampled from the Generic K/A list in ES-401 D.1.b.

2/1 Ability to determine Q54 operability and/or availability of safety related equipment.

Generic K/A 2.1.23 Unable to write a question for this K/A at the generic level.

3 Ability to perform specific Re-sampled from the generic Conduct of Operations (2.1) category.

system and integrated plant Q68 procedures during all modes of plant operation.

009 Small Break LOCA Had difficulty making applicable ties for small break to containment atmosphere limit challenges.

Knowledge of the reasons for the following responses Re-sampled from the available EK3 K/As in the same E/APE.

1/1 as the apply to the small break LOCA:

Q3 EK3.16 Containment temperature, pressure, humidity and level limits 038 Steam Gen. Tube Not applicable to ANO2. Process rad monitors related to SGTR do not Rupture automatically actuate. They are used for trending and alarm functions.

1/1 Knowledge of the reasons Re-sampled from the available EK3 K/As in the same E/APE.

for the following responses Q10 as the apply to the SGTR:

EK3.04 Automatic actions provided by each PRM 062 Loss of Nuclear Svc Limited applicability for ANO2 making it difficult to develop a discriminating Water question. Most temperature indications available are on the primary process fluid side and not SW.

Ability to operate and / or 1/1 monitor the following as they Re-sampled from remaining AA1 K/As in the same E/APE.

apply to the Loss of Nuclear Q16 Service Water (SWS):

AA1.01 Nuclear service water temperature indications 060 Accidental Gaseous Limited scope for selection and development of a discriminating question.

Radwaste Rel.

Remaining AA1 K/A in this system has same problem. Re-sampled from the un-Ability to operate and / or used E/APEs in this Tier and group and sampled from the AA1 K/As to preserve 1/2 monitor the following as they the original outline profile for A1 K/As.

Q23 apply to the Accidental Gaseous Radwaste:

AA1.01 Area radiation monitors 068 Liquid Radwaste The level of difficulty for this K/A is low and tends toward the general radworker making it difficult to develop a discriminatory question.

Knowledge of the operational implication of the Re-sampled from the remaining K5 K/As in this same system topic.

2/2 following concepts as they apply to the Liquid Radwaste Q63 System:

K5.03 Units of radiation, dose, and dose rate Rev 3

ES-401 Record of Rejected K/As Form ES-401-4 Generic K/A 2.1.21 Unable to develop a question at the discriminatory difficulty level.

3 Ability to verify the controlled Re-sampled from the generic Conduct of Operations (2.1) K/As.

Q67 procedure copy.

Generic K/A 2.4.2 Unable to write a question for this K/A at the generic level.

3 Knowledge of system set Re-sampled from the generic Emergency Procedures / Plan (2.4) K/As.

points, interlocks and Q73 automatic actions associated with EOP entry conditions.

034 Fuel Handling Not applicable to the RO job level for ANO2.

Equipment Re-sampled from RO applicable K/As in this system/topic.

2/2 Knowledge of design feature(s) and/or interlock(s)

Q61 which provide for the following:

K4.03 Overload Protection Written Exam Sample Plan: This sampling plan was developed using the manual sampling method described in ES 401, Attachment 1 of NUREG 1021, Revision 10. It applies this same manual method throughout the sampling process, including the sampling of the Generic KAs listed on page 4 of ES 401, section D.1.b, as well as any re sampling that is required for rejected KAs (i.e., KA swaps). Instead of tokens, the plan was developed using the web site random.org to generate the random number associated with each decision. Instead of bounding the possible selections by the number of tokens, the web site allows the user to specify the range of possible numbers for each choice.

Rev 3

ES-401 PWR Examination Outline (SRO) Form ES-401-2 Facility: Arkansas Nuclear One, Unit 2 Date of Exam: August 21, 2015 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 18 3 3 6 Emergency &

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

Plant 2 10 0 2 1 3 Systems 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A 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 associated 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 importance ratings (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 Rev 3

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

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space X 2.4.50 Ability to verify system alarm setpoints and 4.0 76 Accident / 3 operate controls identified in the alarm response manual.

(CFR: 41.10 / 43.5 / 45.3) 000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 X 2.1.20 Ability to interpret and execute procedure 4.6 77 steps.

(CFR: 41.10 / 43.5 / 45.12) 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control X AA2.04 Ability to determine and interpret the 4.3 78 System Malfunction / 3 following as they apply to the Pressurizer Pressure Control Malfunctions: Tech-Spec limits for RCS pressure (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000029 ATWS / 1 000038 Steam Gen. Tube Rupture / 3 000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture - Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main X 2.4.2 Knowledge of system set points, interlocks 4.6 79 Feedwater / 4 and automatic actions associated with EOP entry conditions.

(CFR: 41.7 / 45.7 / 45.8) 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 X AA2.18 Ability to determine and interpret the 3.1 80 following as they apply to the Loss of Vital AC Instrument Bus: The indicator, valve, breaker, or damper position which will occur on a loss of power (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 X AA2.03 Ability to determine and interpret the 2.9 81 following as they apply to the Loss of Nuclear Service Water: The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000065 Loss of Instrument Air / 8 Rev 3

ES-401 3 Form ES-401-2 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 0 0 0 0 3 3 Group Point Total: 6 Rev 3

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 X 2.4.47 Ability to diagnose and recognize 4.2 82 trends in an accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 / 43.5 / 45.12) 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 X AA2.02 Ability to determine and interpret 4.1 83 the following as they apply to the Fuel Handling Incidents: Occurrence of a fuel handling incident (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 X AA2.01 Ability to determine and interpret 4.3 84 the following as they apply to the Control Room Evacuation: S/G level (CFR: 41.7 / 41.10 / 43.5 / 45.13) 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 CE/A13 Natural Circ. / 4 X 2.1.31 Ability to locate control room 4.3 85 switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

(CFR: 41.10 / 45.12)

CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 2 2 Group Point Total: 4 Rev 3

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water X 2.4.6 Knowledge of EOP mitigation 4.7 86 strategies.

(CFR: 41.10 / 43.5 / 45.13) 010 Pressurizer Pressure Control X A2.01 Ability to (a) predict the impacts 3.6 87 of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater failures (CFR: 41.5 / 43.5 / 45.3 / 45.13) 012 Reactor Protection 013 Engineered Safety Features Actuation 022 Containment Cooling X 2.2.25 Knowledge of the bases in 4.2 88 Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.10 / 43.2) 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator X A2.01 Ability to (a) predict the impacts 3.3 89 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: Failure modes of water, oil, and air valves.

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

Rev 3

ES-401 6 Form ES-401-2 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment X A2.05 Ability to (a) predict the impacts 3.9 90 of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Emergency containment entry (CFR: 43.2/43.5)

K/A Category Point Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5 Rev 3

ES-401 7 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control X A2.07 Ability to (a) predict the impacts of 3.3 91 the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Isolation of letdown (CFR: 41.5 / 43.5 / 45.3 / 45.13) 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge X 2.4.11 Knowledge of abnormal condition 4.2 92 procedures.

(CFR: 41.10 / 43.5 / 45.13) 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring X A2.01 Ability to (a) predict the impacts of 2.9 93 the following malfunctions or operations on the ARM system- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply (CFR: 41.5 / 43.5 / 45.3 / 45.13) 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 0 0 0 0 0 0 0 2 0 0 1 Group Point Total: 3 Rev 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Arkansas Nuclear One, Unit 2 Date of Exam: August 21, 2015 Category K/A # Topic RO SRO-Only IR # IR #

2.1.8 Ability to coordinate personnel activities outside the control room. 4.1 94 (CFR: 41.10 / 45.5 / 45.12 / 45.13) 1.

Conduct 2.1.7 Ability to evaluate plant performance and make operational 4.7 95 of Operations judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

Subtotal 2 2.2.37 Ability to determine operability and/or availability of safety related 4.6 96 equipment. (CFR: 41.7 / 43.5 / 45.12)

2. 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 4.7 97 Equipment 41.10 / 43.2 / 43.5 / 45.3)

Control Subtotal 2 2.3.14 Knowledge of radiation or contamination hazards that may arise 3.8 98 during normal, abnormal, or emergency conditions or activities.

(CFR: 41.12 / 43.4 / 45.10) 3.

Radiation Control Subtotal 1 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 4.7 99 45.13) 4.

Emergency 2.4.8 Knowledge of how abnormal operating procedures are used in 4.5 100 Procedures / Plan conjunction with EOPs.

(CFR: 41.10 / 43.5 / 45.13)

Subtotal 2 Tier 3 Point Total 7 Rev 3

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 025 Loss of RHR System Not an SRO level K/A. Also, not a direct tie for this generic to the 025 system.

1/1 Generic 2.1.27 Re-sampled from the Generic K/A list in ES-401 D.1.b.

Q77 Knowledge of system purpose and/or function.

028 Pressurizer Level EALs not directly applicable to this E/APE for ANO2.

Malfunction Re-sampled from the Generic K/A list in ES-401 D.1.b.

1/2 Generic 2.4.41 Q82 Knowledge of the emergency action level thresholds and classifications.

022 Containment Cooling K/A does not have a clear tie to SRO level.

Generic 2.1.28 Re-sampled from the Generic K/A list in ES-401 D.1.b.

2/1 Knowledge of the purpose and Q88 function of major system components and controls.

011 Pressurizer Level Control Rejected due to overlap with the operating test section.

Ability to (a) predict the Re-sampled from the remaining K/As in the A2 group in the same system.

impacts of the following malfunctions or operations on 2/2 the PZR LCS; and (b) based on those predictions, use Q91 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.03 Loss of PZR level 029 Containment Purge K/A tie to this system is not applicable for ANO2.

2/2 Generic 2.4.20 Re-sampled from the Generic K/A list in ES-401 D.1.b.

Q92 Knowledge of the operational implications of EOP warnings, cautions, and notes.

064 Emergency Diesel Overlap with operating test scenario.

Generator Re-sampled from the remaining K/As in the A2 group for the same system.

Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those 2/1 predictions, use procedures to Q89 correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04 Unloading prior to securing an ED/G Generic 2.1.29 K/A does not have a clear tie to SRO level and unable to develop a question 3 with a 10CFR55.43 reference.

Knowledge of how to conduct Q95 system lineups, such as valves, Re-sampled from the Generic Conduct of Operations (2.1) K/As.

breakers, switches, etc.

Rev 3

ES-401 Record of Rejected K/As Form ES-401-4 Generic 2.1.32 Unable to develop a question for the new KA selected on Rev 2 outline at the 3

SRO level.

Ability to explain and apply Q95 system limits and precautions. Re-sampled from the Generic Conduct of Operations (2.1) K/As.

Written Exam Sample Plan: This sampling plan was developed using the manual sampling method described in ES 401, Attachment 1 of NUREG 1021, Revision 10. It applies this same manual method throughout the sampling process, including the sampling of the Generic KAs listed on page 4 of ES 401, section D.1.b, as well as any re sampling that is required for rejected KAs (i.e.,

KA swaps). Instead of tokens, the plan was developed using the web site random.org to generate the random number associated with each decision. Instead of bounding the possible selections by the number of tokens, the web site allows the user to specify the range of possible numbers for each choice.

Rev 3

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 08/24/2015 Examination Level: RO X SRO Operating Test Number: 2015-1 Administrative Topic (see Note) Type Describe activity to be performed Code*

Determine CEA#1 Upper Gripper Coil Temperature A1. Conduct of Operations D/R ANO-2-JPM-NRC-ADMIN-XTCEA 2.1.23 RO (4.3)

Determine time to start CNTMT evacuation and closure A2. Conduct of Operations N/R ANO-2-JPM-NRC-ADMIN-CNTMT2 2.1.25 RO (3.9)

Evaluate Containment Atmospheric Conditions.

A3. Equipment Control D/P/R ANO-2-JPM-NRC-ADMIN-CNTMT 2.2.12 RO (3.7)

Determine condenser off gas radiation monitor setting.

A4. Radiation Control M/R ANO-2-JPM-NRC-ADMIN-CRADMON 2.3.15 RO (2.9)

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 & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & 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: Arkansas Nuclear One Unit 2 Date of Examination: 08/24/2015 Examination Level: RO SRO X Operating Test Number: 2015-1 Administrative Topic (see Note) Type Describe activity to be performed Code*

Determine which operators are available for call out.

A5. Conduct of Operations N/R ANO-2-JPM-NRC-ADMIN-WORK 2.1.5 SRO (3.9)

Determine Shutdown Operations Protection Plan Condition A6. Conduct of Operations D/P/R ANO-2-JPM-NRC-ADMIN-SOPP1 2.1.40 SRO (3.9)

Verify RPS trip set point determination for M/R inoperable MSSV A7. Equipment Control ANO-2-JPM-NRC-ADMIN-MSSVINOP 2.2.40 SRO (4.7)

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

A8. Radiation Control ANO-2-JPM-NRC-ADMIN-EMGRESPSRO 2.3.4 SRO (3.7)

Determine protective action recommendations A9. Emergency Plan M/R ANO-2-JPM-NRC-ADMIN-PAR2 2.4.44 SRO (4.4)

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 & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & 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: Arkansas Nuclear One Unit 2 Date of Examination: 08/24/2015 Exam Level: RO X SRO-I SRO-U Operating Test No.: 2015-1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-HPSI1 2 006 A1.17; RO-4.2/SRO-4.3 A/D/EN/L/S Inventory Control Align HPSI for Hot leg injection S2. ANO-2-JPM-NRC-H2003 5 028 A4.01; RO-4.0/SRO-4.0 L/M/S Containment Start up a Hydrogen Recombiner S3. ANO-2-JPM-NRC-CVCS2 1 004 A4.07; RO-3.9/SRO3.7 A/D/L/P/S Reactivity control Perform Emergency Boration S4. ANO-2-JPM-NRC-EFW01 4 061 A1.01; RO-3.9/SRO4.2 D/EN/L/P/S Heat Removal Shutdown EFW Train A with EFAS Signal Present Secondary S5. ANO-2-JPM-NRC-PZR08 3 010 A2.02; RO-3.9/SRO-3.9 A/M/S Pressure Control Initiate Auxiliary Spray S6. ANO-2-JPM-NRC-FP02 8 086 A4.02; RO-3.5, SRO-3.5 N/S Plant Service systems Respond to a Fire Panel alarm.

S7. ANO-2-JPM-NRC-ELECXT 6 062 A4.01; RO-3.3/SRO-3.1 A/D/S Electrical Cross connect 2B-1 and 2B-2.

S8. ANO-2-JPM-NRC-CEA02 7 012 A4.06 ; RO-4.3/SRO-4.3 D/S Instrumentation Test a Reactor Trip Circuit Breaker In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. ANO-2-JPM-NRC-SFPFL 8 033 A2.03: RO-3.1/SRO3.5 E/D/R Plant Service systems Line up to fill the spent fuel pool from CVCS.

P2. ANO-2-JPM-NRC-DC01 6 063 A4.01; RO-2.8/SRO-3.1 A/N Electrical Swap in-service Battery Chargers.

P3. ANO-2-JPM-NRC-TLOF 4 CE E06 EA2.2; RO-3.0/SRO-4.2 D/E/L Heat Removal Perform Local Actions to start D Condensate pump during a Loss of Feedwater. Secondary

• 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 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA 1/ 1/ 1 (S)imulator Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 08/24/2015 Exam Level: RO SRO-I X SRO-U Operating Test No.: 2015-1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-HPSI1 2 006 A1.17; RO-4.2/SRO-4.3 A/D/EN/L/S Inventory Control Align HPSI for Hot leg injection S2. ANO-2-JPM-NRC-H2003 5 028 A4.01; RO-4.0/SRO-4.0 L/M/S Containment Start up a Hydrogen Recombiner S3. ANO-2-JPM-NRC-CVCS2 1 004 A4.07; RO-3.9/SRO3.7 A/D/L/P/S Reactivity control Perform Emergency Boration S4. ANO-2-JPM-NRC-EFW01 4 061 A1.01; RO-3.9/SRO4.2 D/EN/L/P/S Heat Removal Shutdown EFW Train A with EFAS Signal Present Secondary S5. ANO-2-JPM-NRC-PZR08 3 010 A2.02; RO-3.9/SRO-3.9 A/M/S Pressure Control Initiate Auxiliary Spray S6. ANO-2-JPM-NRC-FP02 8 086 A4.02; RO-3.5, SRO-3.5 N/S Plant Service systems Respond to a Fire Panel alarm.

S7. ANO-2-JPM-NRC-ELECXT 6 062 A4.01; RO-3.3/SRO-3.1 A/D/S Electrical Cross connect 2B-1 and 2B-2.

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

P1. ANO-2-JPM-NRC-SFPFL 8 033 A2.03: RO-3.1/SRO3.5 E/D/R Plant Service systems Line up to fill the spent fuel pool from CVCS.

P2. ANO-2-JPM-NRC-DC01 6 063 A4.01; RO-2.8/SRO-3.1 A/N Electrical Swap in-service Battery Chargers.

P3. ANO-2-JPM-NRC-TLOF 4 CE E06 EA2.2; RO-3.0/SRO-4.2 D/E/L Heat Removal Perform Local Actions to start D Condensate pump during a Loss of Feedwater. Secondary

• 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 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA 1/ 1/ 1 (S)imulator Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 08/24/2015 Exam Level: RO SRO-I SRO-U X Operating Test No.: 2015-1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-HPSI1 2 006 A1.17; RO-4.2/SRO-4.3 A/D/EN/L/S Inventory Control Align HPSI for Hot leg injection S2. ANO-2-JPM-NRC-H2003 5

028 A4.01; RO-4.0/SRO-4.0 Start up a Hydrogen Recombiner L/M/S Containment S3. ANO-2-JPM-NRC-CVCS2 1 004 A4.07; RO-3.9/SRO3.7 A/D/L/P/S Reactivity control Perform Emergency Boration In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. ANO-2-JPM-NRC-SFPFL 8 033 A2.03: RO-3.1/SRO3.5 E/D/R Plant Service systems Line up to fill the spent fuel pool from CVCS.

P2. ANO-2-JPM-NRC-DC01 6 063 A4.01; RO-2.8/SRO-3.1 A/N Electrical Swap in-service Battery Chargers.

• 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 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA 1/ 1/ 1 (S)imulator Revision 0

Appendix D Scenario Outline Form ES-D-1 Facility: _ANO-2_________ Scenario No.: ___1 (New)______ Op-Test No.: _2015-1___

Examiners: ___________________________ Operators: ____________________________

Initial Conditions: 100%, 250 EFPD.

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

Evolution scheduled: Add N2 to A Safety injection tank using 2104.001 section 11. NLO is standing by to align Nitrogen.

Event Malf. No. Event Event No. Type* Description 1 N (BOP) Add Nitrogen to A Safety Injection Tank.

N (SRO) OP-2104.001, Safety Injections Tank Operations.

2 XRCCHAPCNT I (ATC) A Pressurizer pressure control channel fails low.

I (BOP) OP-2203.028, Pressurizer System Malfunction AOP I (SRO) 3 K05-H05 C (BOP) 2A-4 Vital 4160 Bus Room cooler belts break.

C (SRO) OP-2203.012E Annunciator 2K05 Corrective Action.

4 SGATUBE R (ATC) A Steam Generator Tube leak ramps up to 12 gpm over 5 C (BOP) min.

C (SRO) OP-2203.038, Primary to Secondary leakage AOP TS (SRO) 5 CEA49STUCK C (ATC) CEA 49 fails to respond to insertion command.

C (SRO) OP-2203.003, CEA Malfunction AOP TS (SRO) 6 SGATUBE A Steam Generator Tube leak ramps up from 12 gpm to M (All) 100 gpm over 5 min causes Reactor Trip OP-2203.038, Primary to Secondary leakage AOP and OP-2202.001, Standard Post Trip Actions (SPTAs) EOP 7 FW2PW5AAFT A Main Feed water line breaks inside containment.

M (All) OP-2202.001, Standard Post Trip Actions (SPTAs) EOP and OP-2202.009, Functional Recovery EOP.

8 BS2P35AFAL C (BOP) Red Train Containment Spray pump fails to auto start.

C (SRO) OP-2202.010 Standard Attachments EOP.

End Post Blowdown RCS temperature and pressure have been point controlled and A SG has been isolated.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Revision 1 Page 1 of 66

Appendix D Scenario Outline Form ES-D-1 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) 2 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-1) 1 Critical Tasks (2-3) 3 Critical Task Justification References Component cooling water to Exceeding operating limits has 1015.050 Time Critical RCPs must be restored within 10 the potential to degrade the RCS Operation action program, minutes of CIAS or All RCPs pressure boundary. RCPs Attachment C must be secured within the next should be maintained in an CE EPGB Simulator CTs:

10 minutes. available condition for last-resort CT-23, Trip any RCP exceeding use if needed. operating limits (FRG-04)

EOP OP-2202.001 Standard If RCPs are allowed to operate Post Trip Actions.

for 10 minutes without CCW AOP OP-2203.025 RCP flow. OP-1015.050 requires Emergencies.

RCPs not meeting operating limits to be secured within 10 minutes.

Stabilize and control RCS If RCS heatup is allowed after CE EPGB Simulator CTs:

temperature after the ESD SG blowdown, the RCS could CT-07, Establish RCS blowdown terminates. Maintain over pressurize and result in temperature Control (SPTA-07, RCS pressure within the lifting PZR and SG safeties. ESDE-05, HR-05)

Pressure-Temperature limits of These pressure stresses added 200°F and 30°F Margin to to thermal stresses of rapid Saturation throughout cooldown could present PTS implementation of SPTAs and concerns.

Functional Recovery EOP.

Isolate A SG (2202.010 Isolating the SG will minimize CE EPGB Simulator CTs:

Attachment 10 completed) within the potential loss of the CT-14, Isolate most affected SG 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. containment boundary, thus (HR-03).

Assumption is that the operator preventing an offsite release and SAR Section 15.1.18 will diagnose within 30 minutes exceeding 10CFR100 exposure 1015.050 Time Critical and then isolate within next 30 limits at the site boundary. Operation Actions, minutes after entry into Attachment C 2202.009, Functional Recovery EOP 2202.009, Functional EOP Recovery Tech Guide Scenario #1 Objectives

1) Evaluate individual ability to add Nitrogen to Safety Injection Tanks.

2) Evaluate individual response to a failure of the in-service Pressurizer Pressure Channel.

3) Evaluate individual response to a failure of vital electrical room cooler.

4) Evaluate individual response to a Steam Generator Tube leak.

5) Evaluate individual response to a failure of CEAs to respond.

6) Evaluate crew ability to mitigate a Steam Generator Tube Rupture.

7) Evaluate crew ability to mitigate an Excess Steam Demand.

8) Evaluate individual ability to respond to a failure of Green Train Containment Spray to Actuate.

9) Evaluate individual ability to combat events using the Functional Recovery procedure.

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Appendix D Scenario Outline Form ES-D-1 SCENARIO #1 NARRATIVE Simulator session begins with the plant at 100% power steady state.

When the crew has completed their control room walk down and brief, the BOP will add Nitrogen to the A Safety Injection Tank.

When the Nitrogen has been added or cued by lead examiner, the A pressurizer pressure control channel fails low. This will cause all backup heaters to energize raising RCS pressure and the permissive controller for the SDBCS to calculate the permissive setpoint incorrectly. The RCS pressure rise will cause reactor power to increase. The ATC will report that controlling pressurizer pressure channel has failed low and actual pressure is rising. The SRO will enter the Pressurizer System Malfunction AOP. The SRO will direct the ATC to swap pressurizer pressure control channels and BOP to align the Steam Dump Bypass Control System (SDBCS) for the A Pressurizer Pressure control channel failure. [Site OE: CR-ANO-2-2011-1605, Pressurizer pressure failing high, CR-ANO-2-2011-1575, Pressurizer level transmitter failed low due to a reference line failure.]

After the B pressurizer pressure control channel has been placed in service and the SDBCS is aligned with one permissive in manual and cued by the lead examiner, 2VUC-2A, 2A-4 Vital 4160V bus room cooler belts will break. This will cause a 2A-4 room cooler trouble alarm in the control room and actual room temperature to rise. The BOP will refer to OP-2203.012E, 2K05 Annunciator Corrective Actions (ACA). The BOP dispatches a NLO to investigate 2VUC-2A. The NLO will report broken belts on 2VUC-2A and the BOP will use the ACA to place 2VUC-2B room cooler in service. [Site OE: CR-ANO-2-2014-1955, Fan belt broken on room cooler]

After the BOP has started the idle vital bus room cooler and cued by lead examiner, a Steam Generator (SG) Tube Leak will occur on A Steam Generator. The SRO will enter OP 2203.038, Primary to Secondary Leakage AOP. The SRO will direct the ATC to perform power reduction to take the unit offline. He will also direct the BOP to isolate steam to A EFW pump from the A steam generator. The SRO should enter TS 3.4.6.2 Action a, RCS leakage, and TS 3.7.1.2 for EFW when steam is isolated to 2P-7A EFW pump. [Industry OE: SOER 83-2, Steam Generator Tube Ruptures.]

During the power reduction CEA 49 will fail to insert. The ATC will notice CEA 49 failed to respond to the insertion command in Manual Group. The crew may attempt to align the CEA in Manual individual but it will not respond. The SRO will enter OP-2203.003, CEA malfunction AOP. The crew will contact I&C to perform CEA traces and the SRO should enter TS 3.1.3.1 b due to immovable but aligned CEA. I&C will report a Control Element Drive System problem that can be fixed. The ATC should use group P CEA to control ASI until CEA 49 is repaired. [Site OE: CR-ANO-2-2007-0128, CEA 49 fails to respond to insertion commands]

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

After the crew has completed the required reactivity manipulation, entered the appropriate tech specs, and cued by the lead examiner, The Steam Generator Tube leak will get larger. The CRS will perform the continuous action step in Primary to Secondary Leakage AOP to trip the reactor, actuate Safety Injection Actuation Signal (SIAS), actuate Containment Cooling Actuation Signal (CCAS), and go to Standard Post Trip Actions (SPTAs). [Industry OE: SOER 83-2, Steam Generator Tube Ruptures. Steam Generator Tube Rupture response is a time critical operator action per OP-1015.050 Time Critical Operator action program.]

The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. The crew will assess safety functions. After the reactor trips a Main Feedwater line break (A SG due to pressure surge from reactor trip causing the feedwater check valve cap to leak) inside containment will cause an Excess Steam Demand. Main Steam Isolation (MSIS) and Containment Spray (CSAS) will actuate tripping Main Feedwater pumps, Condensate pumps, AFW pump, closing the MSIVs and feedwater block valves. The crew should recognize that red train Containment Spray pump failed to start. The crew should send a NLO to the breaker and to the pump. After, the NLOs report no issues the BOP should start the red train spray pump 2P-35A. Containment Isolation Action Signal (CIAS) will occur causing a loss of Component Cooling Water (CCW) to the Reactor Coolant Pumps (RCPs) the crew will secure RCPs due to the loss of CCW flow causing natural circulation of the RCS. [Industry OE for Excess Steam Demand, SOER 82-7, Reactor Vessel Pressurized Thermal Shock. CR-ANO-2-2009-375, 2P-35A Spray pump failed to respond to handswitch. CR-ANO-2-2006-0848, Component failed to respond to SIAS signal. PRA item # 5 Trip RCPs after loss of CCW in order to avert RCP seal LOCA.]

After completing SPTAs, The SRO will diagnose Excess Steam Demand and Steam Generator Tube Rupture events and enter OP-2202.009, Functional Recovery EOP. The crew will maintain post blowdown temperature and pressure of the RCS to prevent pressurized thermal shock. The BOP will steam B SG using the upstream Atmospheric Dump valve when A SG blows dry. The ATC should use Auxiliary Spray to maintain RCS pressure. The Crew will isolate the A SG using OP-2202.010 Standard Attachment 10.

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Appendix D Scenario Outline Form ES-D-1 Facility: _ANO-2_________ Scenario No.: ___2 (New)______ Op-Test No.: _2015-1___

Examiners: ___________________________ Operators: ____________________________

Initial Conditions: 100%, 250 EFPD. B Pressurizer pressure and level aligned to B channel and 2P-89C aligned to green.

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

Evolution scheduled: Unload and secure #1 EDG using OP-2104.036, Emergency Diesel Generator Operations.

Event Malf. No. Event Event No. Type* Description 1 N (BOP)

Unload and secure #1 EDG.

N (SRO) OP-2104.036, Emergency Diesel Generator Operations.

TS (SRO) 2 CVC4817DEM I (ATC) Letdown flow controller auto signal drifts high.

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

3 FW2P8BSS C (BOP) B Heater Drain pump shaft shear inside the casing.

C (SRO) OP-2203.012C Annunciator 2K03 Corrective Action.

4 CVC2P36CFAL C (ATC) 2P-36C charging pump breaker trip.

C (SRO) OP-2203.036, Loss of Charging AOP 5 DI_HS_3810_2 C (BOP) A Main chiller trip.

K13-C03 C (SRO) OP-2203.012M Annunciator 2K13 Corrective Action (ACA).

6 CV0252 C (ATC) Turbine Control Valve fails closed.

C (SRO) OP-2203.024 Loss of Turbine Load AOP.

TS (SRO) 7 M (All) Inadvertent Red train Main Steam Isolation Signal causing ESFMSIS1I a reactor trip.

OP-2202.001, Standard Post Trip Actions (SPTAs) EOP 8 M (All) Loss of Coolant accident.

RCSLOCATHB OP-2202.003, Loss of Coolant Accident EOP.

9 SIS2P89BDEG C (BOP) 2P-89B Motor overload.

ESFK311AAF C (SRO) 2CV-5035-1 High pressure safety injection and 2CV-5037-1 Low pressure safety injection valves fail to open.

OP-2202.010 Standard Attachments EOP and OP-2203.012E 2K05 ACA.

10 ESFCCAS2 C (BOP) Green train Containment Cooling fails to actuate.

C (SRO) OP-2202.010 Standard Attachments EOP End RCS cooldown in progress and Margin to Saturation > 30 point degrees.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Revision 2 Page 1 of 46

Appendix D Scenario Outline 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) 2 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-1) 0 Critical Tasks (2-3) 3 Critical Task Justification Commence an RCS cooldown Cooling down and depressurizing CE EPGB Simulator CTs:

within 30 minutes of entry into the RCS removes decay heat and CT-20, Cool down and OP-2202.003, LOCA EOP. lowers the DP at the break, depressurize RCS (LOCA-slowing the leak rate and 09) reducing makeup volume CR-ANO-2-2010-948, required. SDC entry conditions Critical task criteria are also required for long-term cooling.

Perform one or more of the SI flow keeps the core covered, CE EPGB Simulator CTs:

following to maintain OR restore cooled, and borated. Inadequate CT-16, Establish required Margin to Saturation (MTS) > 30 SI flow could result in a net loss of SI flow (LOCA-02) degrees F. RCS inventory, pressure control, 1015.050 Time Critical Start a Green train HPSI pump and subcooling. Once subcooling Operation Actions, 2P-89C is lost, pressurizer level is no Attachment C Throttle open red train HPSI longer a valid indication of RCS valve 2CV-5035-1. mass inventory, and a reactor head void can form, both of which If MTS lowers < 30 degrees F, it complicate the event recovery.

must be restored >30 degrees F within 10 min. RCP operating limits require MTS to be >300F.

Establish RCS pressure control to Once RCS subcooling is lost, CE EPGB Simulator CTs:

maintain RCS subcooling. After PZR level is no longer a valid CT-06, Establish RCS the HPSI failures have been indication of RCS inventory. A Pressure Control (LOCA-addressed and MTS restored to reactor head void can form, and if 12)

>30 degrees F then maintain left uncontrolled, could result in pressure and temperature within core uncovery and fuel damage.

the PT limits of <2000 F and

>300F MTS throughout implementation of OP-2202.003, LOCA EOP.

Scenario #2 Objectives

1) Evaluate individual ability to unload and secure #1 Emergency Diesel Generator.

2) Evaluate individual response to a Heater Drain Pump shaft shear.

3) Evaluate individual response to a Charging pump trip.

4) Evaluate individual response to a Loss of a Main Chiller.

5) Evaluate individual response to a failure of a Letdown flow controller.

6) Evaluate a crews response to turbine control valve failing closed.

7) Evaluate a crews response to an inadvertent Main Steam Isolation Signal Actuation.

8) Evaluate crew ability to mitigate a LOCA.

9) Evaluate individual response to Safety Injection valve and pump failures.

10) Evaluate individual response to Containment cooling failures.

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Appendix D Scenario Outline Form ES-D-1 SCENARIO #2 NARRATIVE Simulator session begins with the plant at 100% power steady state. The BOP will unload and secure the #1 Emergency Diesel Generator. The SRO will have to enter TS 3.8.1.1 action b and TS 3.4.4 action b.

When the #1 EDG is secured and the SRO has entered the appropriate TS or when cued by the lead examiner, 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. Not credited as an abnormal event since they may take action prior to alarm annunciation.

After the ATC has control of letdown flow and is restoring PZR level to setpoint, and cued by the lead examiner, 2P-8B Heater Drain pump shaft will shear. The BOP will report alarms for low flow and differential pressure on 2P-8B Heater Drain pump. The SRO should refer to the Annunciator Corrective Action and direct the BOPs actions. The BOP should investigate and determine that flow and differential pressure indicate zero. The BOP should secure 2P-8B and reduce turbine load as necessary to reactor power less than 100%. [Site OE: CR-ANO-1-2012-864, Unit 1 Service water pump shaft shear, CR-ANO-1-2013-2745, Heater drain pump degradation and failure.]

When the B Heater Drain Pump has been secured and power is stabilized <100% and cued by the lead examiner, C Charging pump will trip. The SRO will enter the Loss of Charging AOP.

The ATC will check for a suction source and discharge flow path. The ATC will then start a backup charging pump by moving the lead charging pump selector switch. [Site OE: CR-ANO-2-2015-0432, 2P-36A charging pump stopped running, CR-ANO-2-2001-0685, 2P-36C tripped.]

When a backup charging pump is in service or cued by the lead examiner, A Main Chiller will trip.

This will cause alarms on 2K-13 in the back of the control room. The BOP should assess the alarms using the Annunciator Corrective Action and direct a NLO to investigate. The loss of the Main Chiller will cause a loss of cooling to the control element drives and the Containment building coolers. The BOP should align Service Water to B Main Chiller and direct the NLO to start the B Main Chiller.

After the BOP has aligned the B Main Chiller and directed a NLO to start a Main chiller, or cued by lead examiner, #4 turbine control valve will fail closed. This will lower steam flow, raise steam pressure, raise RCS temperature and lower reactor power. The crew should recognize the signs of load rejection and determine that #4 control valve (#4CV) has failed closed. The SRO will enter OP-2203.024, Loss of Turbine Load AOP. The ATC will commence normal boration to lower Tave to Tref. The crew should contact a NLO to determine the reason control valve #4 CV has failed.

The SRO should enter TS 3.2.6 due to Tcold being out band high. The crew should contact I&C to fail #4 CV closed to prevent it from opening. [Site OE: CR-ANO-2-2009-109, Turbine Control Valve failed closed during power ascension.]

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Appendix D Scenario Outline Form ES-D-1 SCENARIO #2 NARRATIVE (continued)

After, the ATC has completed the required reactivity manipulation and cued by the lead examiner, an Inadvertent red train Main Steam Isolation Signal (MSIS) will occur. The MSIS will close the MSIVs, trip the Main Feedwater pumps, and Condensate pumps. RCS pressure will rise causing an automatic plant trip if the crew does not manually trip the reactor. [Site OE: CR-ANO-2-2013-005, Inadvertent SIAS, CIAS, and CCAS.]

The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. The crew will assess safety functions. The crew should recognize the signs of LOCA and actuate Safety Injection Actuation Signal (SIAS) and Containment Cooling Actuation Signal (CCAS) and the SRO should diagnose and enter OP-2202.003, Loss of Coolant Accident EOP. The BOP should recognize that two Safety Injection valves failed to open and open them. The B High Pressure Safety Injection (HPSI) pump motor will degrade and the BOP should secure the B HPSI pump.

They should also start the swing HPSI pump 2P-89C. The BOP should also recognize that green train Containment Cooling did not actuate as designed and place the green train containment coolers in emergency mode. After the crew has entered the LOCA EOP, the crew will commence a cooldown. [Industry OE: SEN-220, SEN-216, & SEN-182, RCS leakage events.]

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Appendix D Scenario Outline Form ES-D-1 Facility: _ANO-2_________ Scenario No.: ___3 (New)______ Op-Test No.: _2015-1___

Examiners: ___________________________ Operators: ____________________________

Initial Conditions: ~2% power, 250 EFPD. B Channel Pressurizer Pressure and Level in service. 2P-89C and 2P-4B aligned to green power.

Turnover: Red Train Maintenance Week. EOOS indicates Minimal Risk. 2% power B Main Feedwater pump idling. Turbine warm up complete. A Main Feedwater pump is tagged out for Maintenance. Expected duration 2 hrs. Test Team will roll and sync turbine and idle the A Main Feedwater pump after maintenance. SDBCS is balancing steam flow between condensers with 2CV-0302/2CV-0306 in manual and 2CV-0303 in auto.

Evolution scheduled: Place the B Main Feedwater pump in service using OP-2106.007 Main Feedwater Pump and FWCS Operations starting with step 13.8. Then commence power escalation to 15% for turbine tie on using OP-2102.004 Power Operations section 8 beginning with step 8.2. Logs will be completed by the test team.

Event Malf. No. Event Event No. Type* Description 1 N (BOP) Place the B Main Feedwater pump in service.

N (SRO) OP-2106.007, Main Feedwater Pump and FWCS Operations.

2 R (ATC) Power escalation.

N (SRO) OP-2102.004, Power Operations.

3 XRCCHBPLVL I (ATC) B Pressurizer level control channel fails high.

I (SRO) OP-2203.028, Pressurizer System Malfunction AOP TS (SRO) 4 SW2P4C C (BOP) 2P-4C Service water pump trip.

C (SRO) OP-2203.022, Loss of Service Water AOP 5 ESFSIAS2I C (ATC) Inadvertent Green train safety injection.

K04-G08 C (BOP) OP-2203.018, Inadvertent SIAS AOP K04-H08 C (SRO)

TS (SRO) 6 RCP2P32APV A RCP High Vibrations.

RCP2P32AMV M (All) OP-2203.025 RCP Emergencies AOP.

OP-2202.001, Standard Post Trip Actions (SPTAs) EOP 7 LOSE161 Loss of Offsite power.

M (All)

LOSE500 OP-2202.007, Loss of Offsite Power EOP 8 C (BOP) #1 Emergency Diesel Generator low voltage.

C (SRO) OP-2202.001, Standard Post Trip Actions (SPTAs) EOP 9 A308 #1 Emergency Diesel Generator output breaker will trip open.

C (BOP)

EDGDG2LOIL #2 Emergency Diesel Generator will trip due to a loss of lube oil.

C (SRO)

OP-2202.008, Station Blackout EOP One vital bus restored.

End point

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Revision 1 Page 1 of 48

Appendix D Scenario Outline 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) 2 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-1) 0 Critical Tasks (2-3) 3 Critical Task Justification References Maintain RCS pressure within the RCS pressure must be CE EPGB Simulator CTs:

Pressure-Temperature limits of maintained in these limits to allow CT-05, Establish RCS Pressure 200°F and 30°F Margin to natural circulation of the RCS and Control (LOOP-02)

Saturation and less than 2500 prevent over pressurizing the EOP 2202.007 Loss of Offsite psia. RCS boundary. Power EOP.

EOP 2202.007 Loss of Offsite power EOP Tech Guide A RCP must be secured within The out-of-limits condition could 1015.050 Time Critical 10 min of the reactor trip. result in shaft seal damage, and Operation action program, then shaft seal failure could result Attachment C in increased RCS leakage out the CE EPGB Simulator CTs:

seal to the containment CT-23, Trip any RCP exceeding atmosphere, which would worsen operating limits.

the event severity. CR-ANO-2-2010-948, Critical task criteria Energize at least one vital AC bus Without any AC power available CE EPGB Simulator CTs:

with in 10 minutes of entering the for ESF pumps, the ability to CT-03, Energize at least one station blackout EOP. maintain the plant in a safe state vital AC bus.

is severely degraded since no EOP 2202.007 Loss of Offsite makeup water can be added to Power EOP.

the RCS for inventory control EOP 2202.008 Station Blackout purposes. EOP.

ANO-2 SAR table 8.3.7, Reg Guide 1.155 and DCP 92-2011.

Scenario #3 Objectives

1) Evaluate individual ability to place the first Main Feedwater pump in service.

2) Evaluate individual ability to perform a power escalation.

3) Evaluate individual response to a Pressurizer level control channel failing high.

4) Evaluate individual response to a loss of a service water pump.

5) Evaluate individual response to an Inadvertent Safety Injection Actuation Signal.

6) Evaluate individual and crews response to Reactor Coolant Pump High Vibrations.

7) Evaluate individual response to a low voltage on an Emergency Diesel Generator.

8) Evaluate individual response to a loss of both vital 4160 busses.

9) Evaluate crews ability to mitigate a Loss of Offsite power and Station blackout event.

Revision 1 Page 2 of 48

Appendix D Scenario Outline Form ES-D-1 SCENARIO #3 NARRATIVE Simulator session begins with the plant at ~2% power.

When the crew has completed their control room walk down and brief, the BOP will place the B Main Feedwater pump in service to allow power escalation. After the B Main Feedwater pump has been placed in service the crew will commence a power escalation by withdrawing control rods and diluting as designated by the reactivity plan.

When the required reactivity manipulation has been completed or when cued by the lead examiner, Channel B pressurizer level control channel will fail high. This will cause letdown to rise to maximum output. The SRO will enter the PZR System Malfunction AOP. The ATC will take manual control of letdown and lower flow. The ATC will shift level control channels from B to A and then restore letdown to automatic. The SRO will enter TS 3.3.3.5 for remote shutdown and 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-4C service water pump will trip. The ATC/BOP should report 2P-4C breaker trip alarm and announce that 2P-4C has tripped. The SRO will enter the OP-2203.022 Loss of Service Water AOP. The BOP will verify the B Service Water pump aligned to supply Loop 2 Service Water. He will then start 2P-4B Service Water pump to restore green train service water. [Site OE: CR-ANO-2-2003-178, 2P-4C Service Water pump trip.]

When 2P-4B is supplying Loop 2 service water, or cued by the lead examiner, a Green Train Safety Injection Actuation Signal (SIAS) will occur. The SRO will enter Inadvertent SIAS AOP OP-2203.028. The BOP will override and open the auxiliary cooling water valve and component cooling water valves. The ATC will secure all charging pumps. The SRO will enter Tech Specs for ECCS components 3.5.2 and for 3.7.3.1 for service water. [Site OE: CR-ANO-2-2013-005, Inadvertent SIAS, CIAS, and CCAS. Industry OE: ICES # 140474, failure of relay cause auto ESF start of AFW]

When the crew has restored Service water to ACW/CCW, secured charging, and entered the appropriate Tech Specs or at the lead examiners cue, A Reactor Coolant Pump (RCP) vibrations will increase requiring a reactor trip and the pump to be secured. The crew will trip the reactor and secure A RCP. After A RCP is secured a loss of Offsite Power (LOOP) will occur. The SRO will enter OP 2202.001, Standard Post Trip Actions (SPTAs). The crew will assess safety functions.

Due to the LOOP both Emergency Diesel Generators will start. #1 Emergency Diesel Generator will have low voltage and the BOP should recognize the low voltage condition and raise voltage allowing the output breaker to close and supply 2A-3 4160 volt vital bus. The ATC will have to align for Auxiliary spray due to the loss of RCPs and control RCS pressure which will be trending up due to steam generator pressure rising and loss of forced circulation. The crew will actuate Emergency Feedwater Actuation Signal (EFAS). [Industry OE: ICES-295297 Manual shutdown due to RCP vibrations, ICES- 160561 High RCP vibs, SEN-274 RCP seal failure due to high vibs.

SOER 99-01 Loss of grid}

The SRO will diagnose and enter OP-2202.007, Loss of Offsite Power EOP. Once LOOP has been entered #1 EDG output breaker will trip open due to an internal issue and #2 Emergency Diesel will trip due to a lube oil gasket rupture. The SRO should re-diagnose and enter Station Blackout. The BOP will restore power to either 2A-3 or 2A-4 Vital 4160 volt bus from the Alternate AC generator. [Industry OE: SOER 03-01 emergency power reliability.]

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