ML100670657

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AN2-2009-08-Final Outlines
ML100670657
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 08/07/2009
From: Brian Larson
Operations Branch IV
To:
Entergy Operations
References
Download: ML100670657 (37)
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Text

ES-401 PW R RO Examination Outline Form ES-401-2 Rev. 1 Facility: Arkansas Nuclear One Unit 2 Date of Exam: 08/07/2009 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 3 3 3 3 3 3 18 0 0 0 Emergency 2 1 2 1 N/A 1 2 2 9 0 0 0

& N/A Abnormal Tier Plant Totals 4 5 4 4 5 5 27 0 0 0 Evolutions 1 2 2 3 3 3 2 2 3 2 3 3 28 0 0 0 2.

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

3. Generic Knowledge And 10 0 Abilities Categories 2 3 3 2 0 0 0 0 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).
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 and justified; 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 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.

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

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

000008 Pressurizer Vapor Space X AK3.01 - Knowledge of the reasons 1 3.7 1 Accident / 3 for the following responses as they apply to the Pressurizer Vapor Space Accident: - Why PZR level may come back on scale if RCS is saturated 000009 Small Break LOCA / 3 X EA2.09 - Ability to determine and 2 2.8 1 interpret the following as they apply to a small break LOCA: - Low-pressure SWS activity monitor 000011 Large Break LOCA / 3 X EK1.01 - Knowledge of the 3 4.1 1 operational implications of the following concepts as they apply to the Large Break LOCA: - Natural circulation and cooling, including reflux boiling 000015/000017 RCP X AK2.10 - Knowledge of the 4 2.8 1 Malfunctions / 4 interrelations between the Reactor Coo lant Pum p Malfunctions and the following: - RCP indicators and controls 000022 Loss of Rx Coolant X AK3.05 - Knowledge of the reasons 5 3.2 1 Makeup / 2 for the following responses as they apply to the Loss of Reactor Coolant Makeup: - Need to avoi d plant transients 000026 Loss of Component X 2.2.44 - Equipment Control - Ability to 6 4.2 1 Cooling Water / 8 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.

000027 Pressurizer Pressure X AK1.03 - Knowledge of the 7 2.6 1 Control System Malfunction / 3 operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: - Latent heat of vaporization/condensation 000029 ATWS / 1 X 2.2.36 - Equipment Control - Ability to 8 3.1 1 analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

000038 Steam Gen. Tube X EK3.02 - Knowledge of the reasons 9 4.4 1 Rupture / 3 for the following responses as they apply to the SGTR: - Prevention of secondary PORV cycling 000054 Loss of Main Feedwater X 2.4.6 - Emergency Procedures/Plan - 10 3.7 1

/4 Knowledge of EOP mitigation strategies.

ES-401, 22 of 33

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

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

000055 Station Blackout / 6 X EA1.05 - Ability to operate and/or 11 3.3 1 monitor the following as they apply to a Station Blackout: - Battery, when approaching fully discharged 000056 Loss of Off-site Power / 6 X AK1.03 - Knowledge of the 12 3.1 1 operational implications of the following concepts as they apply to Loss of Offsite Power: - Definition of subcooling: use of steam tables to determine it 000057 Loss of Vital AC Inst. X AA1.05 - Ability to operate and/or 13 3.2 1 Bus / 6 monitor the following as they apply to the Loss of Vital AC Instrument Bus: -

Backup instrument indications 000058 Loss of DC Power / 6 X AA2.02 - Ability to determine and 14 3.3 1 interpret the following as they apply to the Loss of DC Power: - 125V dc bus voltage, low/critical low, alarm 000062 Loss of Nuclear Svc X AA1.03 - Ability to operate and/or 15 3.6 1 Water / 4 monitor the following as they apply to the Loss of Nuclear Service Water (SWS): - SWS as a backup to the CCWS 000077 Generator Voltage and X AK2.06 - Knowledge of the 16 3.9 1 Electric Grid Disturbances / 6 interrelations between Generator Voltage and Electric Grid Disturbances and the following: -

Reactor power CE/E02 Reactor Trip - X EK2.1 - Knowledge of the 17 3.3 1 Stabilization - Recovery / 1 interrelations between the (Reactor Trip Recovery) and the following: -

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features CE/E05 Steam Line Rupture - X EA2.2 - Ability to determine and 18 3.4 1 Excessive Heat Transfer / 4 interpret the following as they apply to the (Excess Steam Demand): -

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments K/A Category Totals: 3 3 3 3 3 3 Group Point Totals: 18 ES-401, 22 of 33

ES -401 3 Form ES -401-2 Rev. 1 PWR Examination Outline Emergency and Abnormal Evolutions - Tier 1/Group 2 (RO)

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

000001 Continuous Rod X AA2.05 - Ability to determine and 19 4.4 1 Withdrawal / 1 interpret the following as they apply to the Continuous Rod Withdrawal: -

Uncontrolled rod withdrawal, from available indications 000003 Dropped Control Rod / 1 X AK2.05 - Knowledge of the 20 2.5 1 interrelations between the Dropped Control Rod and the following: -

Control rod drive power supplies and logic circuits 000032 Loss of Source Range X 2.1.28 - Conduct of Operations - 21 4.1 1 NI / 7 Knowledge of the purpose and function of major system components and controls.

000036 Fuel Handling Accident / X AA1.03 - Ability to operate and/or 22 3.5 1 8 monitor the following as they apply to the Fuel Handling Incidents: - Reactor building containment evacuation alarm enable switch 000051 Loss of Condenser X AA2.02 - Ability to determine and 23 3.9 1 Vacuum / 4 interpret the following as they apply to the Loss of Condenser Vacuum: -

Conditions requiring reactor and/or turbine trip 000060 Accidental Gaseous X 2.1.27 - Conduct of Operations - 24 3.9 1 Radwaste Rel. / 9 Knowledge of system purpose and/or function.

000074 Inad. Core Cooling / 4 X EK3.05 - Knowledge of the reasons 25 4.2 1 for the following responses as they apply to the Inadequate Core Cooling: - Activating the HPI system CE/A11 RCS Overcooling - PTS X EK2.1 - Knowledge of the 26 3.2 1

/4 interrelations 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 CE/E09 Functional Recovery X EK1.1 - Knowledge of the operational 27 3.4 1 implications of the following concepts as they apply to the (Functional Recover): - Components, capacity, and function of emergency systems K/A Category Totals: 1 2 1 1 2 2 Group Point Totals: 9 ES-401, 23 of 33

ES -401 4 Form ES -401-2 Rev. 1 PWR Examination Outline Plant Systems - Tier 2/Group 1 (RO)

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

003 Reactor Coolant Pump X K5.03 - Knowledge of the 28 3.1 1 operational implications of the following concepts as they apply to the RCPS: - Effects of RCP shutdown on T-ave.,

including the reason for the unreliability of T-ave.

in the shutdown loop 004 Chemical and Volume X K1.17 - Knowledge of the 29 3.4 1 Control physical connections and/or cause-effect relationships between the CVCS and the following systems: - PZR 005 Residual Heat X K6.03 - Knowledge of the 30 2.5 1 Removal effect of a loss or malfunction of the following will have on the RHRS: - RHR heat exchanger 006 Emergency Core X K1.03 - Knowledge of the 31 4.2 1 Cooling physical connections and/or cause-effect relationships between the ECCS and the following systems: - RCS 007 Pressurizer X K4.01 - Knowledge of 32 2.6 1 Relief/Quench Tank PRTS design feature(s) and/or interlock(s) which provide for the following: -

Quench tank cooling 008 Component Cooling X A1.01 - Ability to predict 33 2.8 1 Water and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including: - CCW flow rate 010 Pressurizer Pressure X K4.02 - Knowledge of 34 3.0 1 Control PZR PCS design feature(s) and/or interlock(s) which provide for the following: -

Prevention of uncovering PZR heaters 010 Pressurizer Pressure X A4.01 - Ability to 35 3.7 1 Control manually operate and/or monitor in the control room: - PZR spray valve ES-401, 24 of 33

ES -401 4 Form ES -401-2 Rev. 1 PWR Examination Outline Plant Systems - Tier 2/Group 1 (RO)

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

012 Reactor Protection X K2.01 - Knowledge of bus 36 3.3 1 power supplies to the following: - RPS channels, components, and interconnections 013 Engineered Safety X 2.2.44 - Equipment 37 4.2 1 Features Actuation 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.

013 Engineered Safety X K5.02 - Knowledge of the 38 2.9 1 Features Actuation operational implications of the following concepts as they apply to the ESFAS: - Safety system logic and reliability 022 Containment Cooling X A4.02 - Ability to 39 3.2 1 manually operate and/or monitor in the control room: - CCS pumps 022 Containment Cooling X A2.01 - Ability to (a) 40 2.5 1 predict the impacts 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 over-current 026 Containment Spray X K3.02 - Knowledge of the 41 4.2 1 effect that a loss or malfunction of the CSS will have on the following: - Recirculation spray system 039 Main and Reheat X K3.04 - Knowledge of the 42 2.5 1 Steam effect that a loss or malfunction of the MRSS will have on the following: - MFW pumps ES-401, 24 of 33

ES -401 4 Form ES -401-2 Rev. 1 PWR Examination Outline Plant Systems - Tier 2/Group 1 (RO)

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

059 Main Feedwater X K4.19 - Knowledge of 43 3.2 1 MFW System design feature(s) and/or interlock(s) which provide for the following: -

Automatic feedwater isolation of MFW 059 Main Feedwater X A3.03 - Ability to monitor 44 2.5 1 automatic operation of the MFW System, including: - Feedwater pump suction flow pressure 061 Auxiliary/Emergency X K5.02 - Knowledge of the 45 3.2 1 Feedwater operational implications of the following concepts as they apply to the AFW System: - Decay heat sources and magnitude 062 AC Electrical X A1.03 - Ability to predict 46 2.5 1 Distribution and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the A.C. Distribution System controls including: - Effect on instrumentation and controls of switching power supplies 063 DC Electrical X 2.1.27 - Conduct of 47 3.9 1 Distribution Operations - Knowledge of system purpose and/or function.

063 DC Electrical X A3.01 - Ability to monitor 48 2.7 1 Distribution automatic operation of the D.C. Electrical System, including: -

Meters, annunciators, dials, recorders, and indicating lights 064 Emergency Diesel X A4.01 - Ability to 49 4.0 1 Generator manually operate and/or monitor in the control room: - Local and remote operation of the ED/G 064 Emergency Diesel X K6.08 - Knowledge of the 50 3.2 1 Generator effect of a loss or malfunction of the following will have on the ED/G System: - Fuel oil storage tanks ES-401, 24 of 33

ES -401 4 Form ES -401-2 Rev. 1 PWR Examination Outline Plant Systems - Tier 2/Group 1 (RO)

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

073 Process Radiation X K3.01 - Knowledge of the 51 3.6 1 Monitoring effect that a loss or malfunction of the PRM System will have on the following: - Radioactive effluent releases 076 Service Water X A2.02 - Ability to (a) 52 2.7 1 predict the impacts of the following malfunctions or operations on the SWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Service water header pressure 078 Instrument Air X K2.01 - Knowledge of bus 53 2.7 1 power supplies to the following: - Instrument air compressor 103 Containment X A2.05 - Ability to (a) 54 2.9 1 predict the impacts 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 103 Containment X 2.1.23 - Conduct of 55 4.3 1 Operations - Ability to perform specific system and integrated plant procedures during all modes of plant operation.

K/A Category Totals: 2 2 3 3 3 2 2 3 2 3 3 Group Point Totals: 28 ES-401, 24 of 33

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

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

001 Control Rod Drive X K2.05 - Knowledge of bus 56 3.1 1 power supplies to the following: - M/G sets 014 Rod Position Indication X K4.06 - Knowledge of 57 3.4 1 RPIS design feature(s) and/or interlock(s) which provide for the following: -

Individual and group misalignment 016 Non-nuclear X A4.02 - Ability to 58 2.7 1 Instrumentation manually operate and/or monitor in the control room: - Recorders 017 In-core Temperature X K6.01 - Knowledge of the 59 2.7 1 Monitor effect of a loss or malfunction of the following will have on the ITM System components: - Sensors and detectors 028 Hydrogen Recombiner X A1.02 - Ability to predict 60 3.4 1 and Purge Control and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the HRPS controls including: - Containment pressure 034 Fuel Handling X A3.03 - Ability to monitor 61 2.9 1 Equipment automatic operation of the Fuel Handling System, including: - High flux at shutdown 045 Main Turbine X K3.01 - Knowledge of the 62 2.9 1 Generator effect that a loss or malfunction of the MT/G System will have on the following: - Remainder of the plant 068 Liquid Radwaste X A2.02 - Ability to (a) 63 2.7 1 predict the impacts of the following malfunctions or operations on the Liquid Radwaste System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Lack of tank recirculation prior to release ES-401, 25 of 33

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

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

075 Circulating Water X 2.1.28 - Conduct of 64 4.1 1 Operations - Knowledge of the purpose and function of major system components and controls.

086 Fire Protection X K5.04 - Knowledge of the 65 2.9 1 operational implications of the following concepts as they apply to the Fire Protection System: -

Hazards to personnel as a result of fire type and methods of protection K/A Category Totals: 0 1 1 1 1 1 1 1 1 1 1 Group Point Totals: 10 ES-401, 25 of 33

ES -401 Generic Knowledge and Abilities Outline (Tier 3) Form ES -401-3 Rev. 1 Facility: Date of Exam:

Category K/A # Topic RO SRO Only QID IR # IR #

1 - Conduct of 2.1.15 Knowledge of administrative requirements for temporary 66 2.7 1 Operations management directives, such as standing orders, night orders, Operations memos, etc.

2.1.18 Ability to make accurate, clear, and concise logs, records, status 67 3.6 1 boards, and reports.

Subtotal 2 2 - Equipment 2.2.7 Knowledge of the process for conducting special or infrequent 68 2.9 1 Control tests.

2.2.21 Knowledge of pre- and post-maintenance operability requirements. 69 2.9 1 2.2.23 Ability to track Technical Specification limiting conditions for 70 3.1 1 operations.

Subtotal 3 3 - Radiological 2.3.4 Knowledge of radiation exposure limits under normal or 71 3.2 1 Controls emergency conditions.

2.3.11 Ability to control radiation releases. 72 3.8 1 2.3.12 Knowledge of radiological safety principles pertaining to licensed 73 3.2 1 operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Subtotal 3 4 - Emergency 2.4.6 Knowledge of EOP mitigation strategies. 74 3.7 1 Procedures/Plan 2.4.26 Knowledge of facility protection requirements, including fire 75 3.1 1 brigade and portable fire fighting equipment usage.

Subtotal 2 Tier 3 Point Total 10 ES-401, 26 of 33

ES-401 PW R SRO Examination Outline Form ES-401-2 Rev. 2 Facility: Arkansas Nuclear One Unit 2 Date of Exam: 08/07/2009 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 0 0 0 0 0 0 0 3 3 6 Emergency 2 0 0 0 N/A 0 0 0 0 2 2 4

& N/A Abnormal Tier Plant Totals 0 0 0 0 0 0 0 5 5 10 Evolutions 1 0 0 0 0 0 0 0 0 0 0 0 0 3 2 5 2.

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

3. Generic Knowledge And 0 7 Abilities Categories 0 0 0 0 2 2 2 1 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).
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 and justified; 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 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.

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

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

000007 Reactor Trip - X 2.4.45 - Emergency Procedures/Plan - 76 4.3 1 Stabilization - Recovery / 1 Ability to prioritize and interpret the significance of each annunciator or alarm.

000025 Loss of RHR System / 4 X AA2.04 - Ability to determine and 77 3.6 1 interpret the following as they apply to the Loss of Residual Heat Removal System: - Location and isolability of leaks 000040 Steam Line Rupture - X AA2.01 - Ability to determine and 78 4.7 1 Excessive Heat Transfer / 4 interpret the following as they apply to the Steam Line Rupture: - Occurrence and location of a steam line rupture from pressure and flow indications 000058 Loss of DC Power / 6 X 2.4.20 - Emergency Procedures/Plan - 79 4.3 1 Knowledge of operational implications of EOP warnings, cautions, and notes.

000065 Loss of Instrum ent Air / 8 X AA2.01 - Ability to determ ine and 80 3.2 1 interpret the following as they apply to the Loss of Instrum ent Air: - Cause and effect of low pressure instrument air alarm CE/E06 Loss of Main Feedwater X 2.1.23 - Conduct of Operations - 81 4.4 1

/4 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

K/A Category Totals: 0 0 0 0 3 3 Group Point Totals: 6 ES-401, 22 of 33

ES -401 3 Form ES -401-2 Rev. 2 PWR Examination Outline Emergency and Abnormal Evolutions - Tier 1/Group 2 (SRO)

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

000037 Steam Generator Tube X AA2.06 - Ability to determine and 82 4.5 1 Leak / 3 interpret the following as they apply to the Steam Generator Tube Leak: -

S/G tube failure 000059 Accidental Liquid X 2.4.41 - Emergency Procedures/Plan - 83 4.6 1 RadWaste Rel. / 9 Knowledge of the emergency action level thresholds and classifications.

000068 Control Room Evac. / 8 X AA2.05 - Ability to determine and 84 4.3 1 interpret the following as they apply to the Control Room Evacuation: -

Availability of heat sink 000069 Loss of CTMT Integrity / X 2.4.46 - Emergency Procedures/Plan - 85 4.2 1 5 Ability to verify that the alarms are consistent with the plant conditions.

K/A Category Totals: 0 0 0 0 2 2 Group Point Totals: 4 ES-401, 23 of 33

ES -401 4 Form ES -401-2 Rev. 2 PWR Examination Outline Plant Systems - Tier 2/Group 1 (SRO)

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

006 Emergency Core X A2.07 - Ability to (a) 86 3.1 1 Cooling 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: - Loss of heat tracing 012 Reactor Protection X A2.01 - Ability to (a) 87 3.6 1 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 bistable operation 039 Main and Reheat X 2.4.11 - Emergency 88 4.2 1 Steam Procedures/Plan -

Knowledge of abnormal condition procedures.

061 Auxiliary/Emergency X A2.03 - Ability to (a) 89 3.4 1 Feedwater 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 078 Instrument Air X 2.2.36 - Equipment 90 4.2 1 Control - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

K/A Category Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Totals: 5 ES-401, 24 of 33

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

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

071 Waste Gas Disposal X A2.09 - Ability to (a) 91 3.5 1 predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Stuck-open relief valve 072 Area Radiation X 2.2.40 - Equipment 92 4.7 1 Monitoring Control - Ability to apply Technical Specifications for a system.

029 Containment Purge X A2.01 - Ability to (a) 93 3.6 1 predict the impacts of the following malfunctions or operations on the Containment Purge System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: -

Maintenance or other activity taking place inside containment K/A Category Totals: 0 0 0 0 0 0 0 2 0 0 1 Group Point Totals: 3 ES-401, 25 of 33

ES -401 Generic Knowledge and Abilities Outline (Tier 3) Form ES -401-3 Rev. 2 Facility: Date of Exam:

Category K/A # Topic RO SRO Only Q IR # IR #

1 - Conduct of 2.1.7 Ability to evaluate plant performance and make operational 94 4.7 1 Operations judgments based on operating characteristics, reactor behavior, and instrument interpretation.

2.1.23 Ability to perform specific system and integrated plant procedures 95 4.4 1 during all modes of plant operation.

Subtotal 2 2 - Equipment 2.2.11 Knowledge of the process for controlling temporary design changes. 96 3.3 1 Control 2.2.43 Knowledge of the process used to track inoperable alarms. 97 3.3 1 Subtotal 2 3 - Radiological 2.3.6 Ability to approve release permits. 98 3.8 1 Controls 2.3.12 Knowledge of radiation safety principles pertaining to licensed 99 3.7 1 duties, such as containment entry requirements, fuel handling responsibilities, access to high-radiation areas, aligning filters, etc.

Subtotal 2 4 - Emergency 2.4.6 Knowledge of EOP mitigation strategies. 100 4.7 1 Procedures/Plan Subtotal 1 Tier 3 Point Total 7 ES-401, 26 of 33

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 1 Tier / Randomly Reason for Rejection Group Selected K/A QID #6 This K&A is more directed to the SRO applicant and is not one of 026 Loss of CCW the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-410 section D.1.b. therefore, it was rejected Generic 2.2.17 - and another generic 2.2 K&A was randomly selected form the list Knowledge of the in NUREG 1021 ES-401 Section D1.b.

maintenance activities during Generic 2.2.44 - Ability to interpret control room indications to power operations, verify the status and operation of a system, and understand how such as risk operator actions and directives affect plant and system conditions assessments, work was randomly selected from the Generic K&A list in NUREG prioritation. 1021 ES-401 Section D1.b. for this category.

QID #9 038 Steam This K&A is not applicable to ANO since the RCS at ANO does Generator Tube not have RCS loop isolation valves therefore, a credible and Rupture operationally valid question can not be developed on this K&A.

RO Exam Tier 1 EK3.07 - Knowledge EK3.02 - Knowledge of the reasons for the following responses as Group 1 of the reasons for the they apply to the SGTR: - Prevention of secondary PORV cycling following responses was randomly selected as a replacement from the other EK3 K&A as they apply to the statements under this category with a RO importance rating of >

SGTR: - RCS loop 2.5.

isolation values QID #10 054 Loss of Main This K&A does not directly tie to the MFW system and is not one Feedwater of the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-401 section D.1.b. therefore, it was 2.3.7 - Ability to rejected and another generic K&A was randomly selected form the comply with radiation list in NUREG 1021 ES-401 Section D1.b.

work permit requirements during 2.4.6 Emergency Procedures/Plan - Knowledge of EOP mitigation normal or abnormal strategies was randomly selected from the Generic K&A list in conditions. NUREG 1021 ES-401 Section D1.b. for this category.

NUREG-1021, Revision 9 Supplement 1 Page 1 of 6

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 1 Tier / Randomly Reason for Rejection Group Selected K/A QID #37 This K&A is not one of the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-401 section 013 ESF Actuation D.1.b. therefore, it was rejected and another generic K&A was randomly selected form the list in NUREG 1021 ES-401 Section 2.3.7 Ability to comply D1.b.

with radiation work 2.2.44 Equipment Control - Ability to interpret control room permit requirements indications to verify the status and operation of a system, and during normal or understand how operator actions and directives affect plant and abnormal conditions. system conditions was randomly selected from the Generic K&A list in NUREG 1021 ES-401 Section D1.b. for this category.

ANO Unit 2 does not have the capability to operate the QID #49 Emergency Diesel Generator air compressors remotely; 064 Emergency Diesel therefore, a credible and operationally valid question can not be Generator. developed on this K&A; therefore, it was rejected and another generic K&A was randomly selected from the A4 section of A4.04 - Ability to K&As for EDGs manually operate and/or monitor in the A4.01 - Ability to manually operate and/or monitor in the control room: - Remote control room: - Local and remote operation of the ED/G was operation of the air randomly selected as a replacement from the other A4 K&A RO Exam compressor switch statements under this category with a RO importance rating of >

(different modes) 2.5.

Tier 2 Group 1 This K&A statement was randomly selected on the last 2 NRC QID #50 exams given to Unit 2 and 2 different questions on EDG air 064 Emergency Diesel receivers was used. This K&A was rejected due to over Generator. sampling of this K&A statement on the 2 previous NRC Exams and no additional operationally valid questions could be K6.07 Knowledge of generated.

the effect of a loss or malfunction of the K6.08 - Knowledge of the effect of a loss or malfunction of the following will have on following will have on the ED/G System: - Fuel oil storage tanks the ED/G System: - Air was selected as a replacement from the other K6 K&A receivers statements under this category with a RO importance rating of >

2.5.

QID #53 ANO Unit 2 does not have an emergency air compressor for an IA backup. Unit 2 and Unit 1 IA systems are normally cross tied 078 Instrument Air to allow for backup air in case of a problem on the opposite System unit.; therefore, a credible and operationally valid question can K2.02 Knowledge of not be developed on this K&A.

bus power supplies to K2.01 - Knowledge of bus power supplies to the following: -

emergency air Instrument air compressor was selected as a replacement from compressors.

the other K2 K&A statements under this category with a RO importance rating of > 2.5.

NUREG-1021, Revision 9 Supplement 1 Page 2 of 6

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 1 QID #54 103 Containment System A2.04 Ability to (a) predict the impacts of This K&A was rejected due to the similarities between question the following #22 K&A 036 AA1.03 and over sampling of the containment malfunctions or evacuation alarm.

operations on the Containment System A2. 05- Ability to (a) predict the impacts of the following and (b) based on those malfunctions or operations on the Containment System and (b) predictions, use based on those predictions, use procedures to correct, control, procedures to correct, or mitigate the consequences of those malfunctions or control, or mitigate the operations: - Emergency Containment Entry was randomly RO Exam consequences of those selected as a replacement from the other A2 K&A statements malfunctions or under this category with a RO importance rating of > 2.5.

Tier 2 operations: -

Group 1 Containment evacuation (including recognition of the alarm)

QID #55 This K&A is not one of the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-401 section 103 Containment D.1.b. therefore, it was rejected and another generic K&A was System randomly selected form the list in NUREG 1021 ES-401 Section 2.1.34 Conduct of D1.b.

Operations -

2.1.23 Conduct of Operations - Ability to perform specific Knowledge of primary system and integrated plant procedures during all modes of and secondary plant plant operation was randomly selected from the Generic K&A chemistry limits.

list in NUREG 1021 ES-401 Section D1.b. for this category.

QID #64 RO Exam 075 Circulating Water Although there are 2 units here at ANO, they are not of the same Tier 2 vender design and a Unit 2 operator cannot operate the Unit 1 Group 2 Generic 2.2.3 -

Equipment Control - plant; therefore, this K&A statement does not apply and is being (multi-unit license) rejected as directed by NUREG 1021 ES-401 section D.1.b. This Knowledge of the K&A was rejected and another generic K&A was randomly design, procedural, selected form the list in NUREG 1021 ES-401 Section D1.b.

and operational Generic 2.1.28 - Conduct of Operations - Knowledge of the differences between purpose and function of major system components and control units. was randomly selected from the Generic K&A list in NUREG 1021 ES-401 Section D1.b. for this category.

NUREG-1021, Revision 9 Supplement 1 Page 3 of 6

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 1 Tier / Randomly Reason for Rejection Group Selected K/A QID #76 This K&A is not one of the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-401 section 007 Reactor Trip - D.1.b. therefore, it was rejected and another generic K&A was Stabilization - randomly selected form the list in NUREG 1021 ES-401 Section Recovery D1.b.

Generic 2.2.19 - Generic 2.4.45 - Emergency Procedures/Plan - Ability to Knowledge of prioritize and interpret the significance of each annunciator or maintenance work alarm was randomly selected from the Generic K&A list in order requirements NUREG 1021 ES-401 Section D1.b. for this category.

QID #79 058 Loss of DC Power Generic 2.4.16 - This K&A is not one of the required Generic K&As to be used Knowledge of EOP in Tier 1 and 2 as directed by NUREG 1021 ES-401 section implementation D.1.b. therefore, it was rejected and another generic K&A was hierarchy and randomly selected form the list in NUREG 1021 ES-401 Section coordination with other D1.b.

support procedures or Generic 2.4.20 - Emergency Procedures/Plan - Knowledge of guidelines such as, operational implications of EOP warnings, cautions, and notes operating procedures, was randomly selected from the Generic K&A list in NUREG abnormal operating 1021 ES-401 Section D1.b. for this category.

procedures, and severe SRO Exam accident management guidelines.

Tier 1 Group 1 QID #80 This K&A statement was randomly selected on the last 2 NRC CE/E06 Loss of exams given to Unit 2 and 2 different questions on when to Instrument Air shutdown on loss of IA was used. This K&A was rejected due to over sampling of this K&A statement on the 2 previous NRC Ability to determine Exams and no additional operationally valid questions could be and interpret the generated.

following as they apply to the Loss of AA2.01 Ability to determine and interpret the following as they Instrument Air: - When apply to the Loss of Instrument Air: - Cause and effect of low-to commence plant pressure instrument air alarm was randomly selected as a shutdown if instrument replacement from the other AA2 K&A statements under this air pressure is category with a SRO importance rating of > 2.5.

decreasing QID #81 This K&A is not one of the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-401 section CE/E06 Loss of Main D.1.b. therefore, it was rejected and another generic K&A was Feedwater randomly selected form the list in NUREG 1021 ES-401 Section Generic 2.2.23 - D1.b.

Equipment Control - Generic 2.1.23 - Conduct of Operations - Ability to perform Ability to track specific system and integrated plant procedures during all Technical Specification modes of plant operation was randomly selected from the limiting conditions for Generic K&A list in NUREG 1021 ES-401 Section D1.b. for operations. this category.

NUREG-1021, Revision 9 Supplement 1 Page 4 of 6

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 1 QID #83 059 Accidental Liquid This K&A is not one of the required Generic K&As to be used Rad waste Release. in Tier 1 and 2 as directed by NUREG 1021 ES-401 section Generic 2.4.5 - D.1.b. therefore, it was rejected and another generic K&A was SRO Exam Emergency randomly selected form the list in NUREG 1021 ES-401 Section Tier 1 Procedures/Plan - D1.b.

Group 2 Knowledge of the Generic 2.4.41 - Emergency Procedures/Plan - Knowledge of organization of the the emergency action level thresholds and classifications was operating procedures randomly selected from the Generic K&A list in NUREG 1021 network for normal, ES-401 Section D1.b. for this category.

abnormal, and emergency evolutions.

QID #88 This K&A is not one of the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-401 section 039 Main and Reheat D.1.b. therefore, it was rejected and another generic K&A was SRO Exam Steam randomly selected form the list in NUREG 1021 ES-401 Section Tier 2 Generic 2.2.6 - D1.b.

Group 1 Equipment Control - Generic 2.4.11 - Emergency Procedures/Plan - Knowledge of Knowledge of the abnormal condition procedures was randomly selected from the process for making Generic K&A list in NUREG 1021 ES-401 Section D1.b. for changes to procedures. this category.

QID #92 This K&A is not one of the required Generic K&As to be used in Tier 1 and 2 as directed by NUREG 1021 ES-401 section 072 Area Radiation D.1.b. therefore, it was rejected and another generic K&A was Monitoring randomly selected form the list in NUREG 1021 ES-401 Section Generic 2.2.35 D1.b.

Equipment Control - Generic 2.2.40 - Equipment Control - Ability to apply Technical Ability to determine Specifications for a system was randomly selected from the Technical Specification Generic K&A list in NUREG 1021 ES-401 Section D1.b. for Mode of Operation. this category.

QID #93 SRO Exam There is no credible 10 CFR 43 SRO tie between this K&A 079 Station Air System statement because Unit 2 does not have a cross connection with Tier 2 A2.01 Ability to (a) IA and Station air therefore, a credible and operationally valid Group 2 question can not be developed on this K&A statement or any predict the impacts of the following others Station Air K&A statements. (most have an importance malfunctions or rating of less than 2.5) operations on the SAS Based on this selection, the Containment Purge System was and (b) based on those randomly selected from the Tier 2 Group 2 systems and the predictions, use originally selected A2.01 K&A statement was retained: Ability procedures to correct, to (a) predict the impacts of the following malfunctions or control, or mitigate the operations on the Containment Purge System and (b) based on consequences of those those predictions, use procedures to correct, control, or mitigate malfunctions or the consequences of those malfunctions or operations: -

operations: - Cross- Maintenance or other activity taking place inside Containment.

connection with IAS.

NUREG-1021, Revision 9 Supplement 1 Page 5 of 6

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 1 QID # 99 This K&A is not SRO only knowledge and does not allow an Generic 2.3.15 SRO only question to be developed thus this K&A was rejected.

Radiological Controls -

Generic K&A 2.3.12, Radiological Controls - Knowledge of Knowledge of radiation radiological safety principles pertaining to licensed operator monitoring systems, duties, such as containment entry requirements, fuel handling such as fixed radiation responsibilities, access to locked high-radiation areas, aligning monitors and alarms, filters, etc. was randomly selected as a replacement from the portable survey required section 2.3 generic K&A statements.

instruments, personnel SRO Exam monitoring equipment, etc.

Tier 3 QID #100 This K&A is not SRO only knowledge and does not allow an Generic 2.4.35 SRO only question to be developed thus this K&A was rejected.

Emergency Generic K&A 2.4.6, Emergency Procedures/Plan - Knowledge Procedures/Plan - of EOP mitigation strategies was randomly selected as a Knowledge of local replacement from the required section 2.4 generic K&A auxiliary operator statements.

tasks during an emergency and the resultant operational effects.

NUREG-1021, Revision 9 Supplement 1 Page 6 of 6

ES-301 Administrative Topics Outline Form ES-301-1 Facility: _______ Arkansas Nuclear One_____________ Date of Examination: ____8-10-09_

Examination Level: RO X SRO Operating Test Number: 2009-1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine volume of Boric acid and DI water to makeup to RWT.

A1. Conduct of Operations P/R ANO-2-JPM-NRC-ADMIN RWT 2.1.23 RO (4.3)

Determine limits for CEA positions using the COLR PDIL.

A2. Conduct of Operations N/R 2.1.25 RO (3.9) ANO-2-JPM-NRC-ADMIN-PDIL Surveillance review A3. Equipment Control D/R ANO-2-JPM-NRC-2P-35A REVIEW 2.2.12 RO (3.7)

Review RWP and determine requirements for tagging B LPSI pump.

A4. Radiation Control D/R ANO-2-JPM-NRC-RWP 2.3.7 RO (3.5)

Emergency Procedures/Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • 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 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility: _______ Arkansas Nuclear One_____________ Date of Examination: ____8-10-09_

Examination Level: RO SRO X Operating Test Number: 2009-1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine if COLR DNBR limit is being met during a loss of COLSS and what action is required.

A5. Conduct of Operations N/R ANO-2-JPM-NRC-ADMIN-DNBRL 2.1.25 SRO (4.2)

Review and approve calculation of volume needed to raise SFP level.

A6. Conduct of Operations D/R ANO-2-JPM-NRC-ADMIN-SFPMU 2.1.23 SRO (4.4)

Determine RPS trip set point due to inoperable MSSV is correct using Technical Specifications.

A7. Equipment Control P/R Ability to apply technical specification for a system 2.2.40 SRO (4.7)

ANO-2-JPM-NRC-ADMIN-MSSVINOP Approve administration of Potassium Iodide A8. Radiation Control D/R ANO-2-JPM-NRC-KI 2.3.14 SRO (3.8)

Issue Protective Action Recommendation to Offsite Authorities (Time Critical)

A9. Emergency D/R ANO-2-JPM-NRC-PAR3 Procedures/Plan 2.4.44 SRO (4.4)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • 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 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: __Arkansas Nuclear One______________ Date of Examination: ____8-10-09___

Exam Level: RO X SRO-I SRO-U Operating Test No.: 2009-1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-EOP01 5 026 A2.03 RO 4.1 SRO 4.4 Secure Containment Spray System (inadvertent)

EN/L/P/S Containment Integrity S2. ANO-2-JPM-NRC-RCP04 4 003 A2.02 RO-3.7/SRO-3.9 A/L/M/S Heat Removal RCP Shutdown S3. ANO-2-JPM-NRC-SIT01 2 006 A2.03 RO-3.3 SRO-3.7 A/P/S RCS Inventory High pressure fill of Safety Injection Tank S4. ANO-2-JPM-NRC-PZR01 3 010 A4.01 RO-3.7/SRO-3.5 D/S Pressure Control Equalize Pressurizer boron S5. ANO-2-JPM-NRC- VCTMU 1 004 A4.15 RO-3.6 SRO-3.7 D/S Reactivity control Perform Manual Makeup to the VCT S6. ANO-2-JPM-NRC-ELECXT 6 062 A4.01 RO-3.3/SRO-3.1 A/M/S Electrical Cross connect 2B-1 and 2B-2.

S7. ANO-2-JPM-NRC-CCW02 8 008 A4.01 RO 3.3/SRO 3.1 Shift running CCW pumps M/S Plant service systems S8. ANO-2-JPM-NRC-CPC02 7 015 A3.03 RO-3.9/SRO-3.9 D/S Instrumentation DNBR/LPD limits with COLSS out of service In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. A2JPM-RO-RSD 6 062 A2.01 RO-3.4/SRO-3.9 A//D/E/L Electrical Perform a remote shutdown as an CBOT P2. ANO-2-JPM-NRC-XFCEA 1 001 A2.14 RO-3.7/SRO-3.9 D/E Reactivity control Transfer a CEA to the hold bus P3. ANO-2-JPM-NRC-69REL 9 068 A4.02 RO-3.2/SRO-3.1 N/R Radioactivity Perform a release of 2T-69A Boric Acid Condensate Tank Release

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 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 (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 3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: __Arkansas Nuclear One______________ Date of Examination: ____8-10-09___

Exam Level: RO SRO-I X SRO-U Operating Test No.: 2009-1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-EOP01 5 026 A2.03 RO 4.1 SRO 4.4 Secure Containment Spray System (inadvertent)

EN/L/P/S Containment Integrity S2. ANO-2-JPM-NRC-RCP04 4 003 A2.02 RO-3.7/SRO-3.9 A/L/M/S Heat Removal RCP Shutdown S3. ANO-2-JPM-NRC-SIT01 2 006 A2.03 RO-3.3 SRO-3.7 A/D/P/S RCS Inventory High pressure fill of Safety Injection Tank S4. ANO-2-JPM-NRC-PZR01 3 010 A4.01 RO-3.7/SRO-3.5 D/S Pressure Control Equalize Pressurizer boron S5. ANO-2-JPM-NRC- VCTMU 1 004 A4.15 RO-3.6 SRO-3.7 D/S Reactivity control Perform Manual Makeup to the VCT S6. ANO-2-JPM-NRC-ELECXT 6 062 A4.01 RO-3.3/SRO-3.1 A/M/S Electrical Cross connect 2B-1 and 2B-2.

S7. ANO-2-JPM-NRC-CCW02 8 008 A4.01 RO 3.3/SRO 3.1 M/S Plant service Shift running CCW pumps systems In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. A2JPM-RO-RSD 6 062 A2.01 RO-3.4/SRO-3.9 A/D/E/L Electrical Perform a remote shutdown as an CBOT P2. ANO-2-JPM-NRC-XFCEA 1 001 A2.14 RO-3.7/SRO-3.9 D/E Reactivity control Transfer a CEA to the hold bus P3. ANO-2-JPM-NRC-69REL 9 068 A4.02 RO-3.2/SRO-3.1 N/R Radioactivity Perform a release of 2T-69A Boric Acid Condensate Tank Release

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 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 (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 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: __Arkansas Nuclear One______________ Date of Examination: ____8-10-09___

Exam Level: RO SRO-I SRO-U X Operating Test No.: 2009-1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-EOP01 5 026 A2.03 RO 4.1 SRO 4.4 Secure Containment Spray System (inadvertent)

EN/L/P/S Containment Integrity S2. ANO-2-JPM-NRC-RCP04 4 003 A2.02 RO-3.7/SRO-3.9 A/L/M/S Heat Removal RCP Shutdown In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. A2JPM-RO-RSD 6 062 A2.01 RO-3.4/SRO-3.9 A/D/E/L Electrical Perform a remote shutdown as an CBOT P2. ANO-2-JPM-NRC-XFCEA 1 001 A2.14 RO-3.7/SRO-3.9 D/E Reactivity control Transfer a CEA to the hold bus P3. ANO-2-JPM-NRC-69REL 9

068 A4.02 RO-3.2/SRO-3.1 Perform a release of 2T-69A Boric Acid Condensate Tank N/R Radioactivity Release

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 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 (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 2

Appendix D Scenario 1 Form ES-D-1 Facility: ANO-2 Scenario No.: 1 (New) Op-Test No.: 2009-1 Examiners: Operators:

Initial Conditions:

100% MOL, All Engineered Safety Features systems are in standby. Gland Seal pressure control valve 2PCV-0231 is jacked closed. Green Train Maintenance Week.

Turnover:

100%. 260 EFPD. EOOS indicates Minimal Risk. 2PCV-0231 jacked closed, manually control seal steam pressure. Green Train Maintenance Week.

Event Malf. No. Event Event No. Type* Description 1 XSG2PT10411 I (BOP) A Steam generator safety channel pressure transmitter I (SRO) fails low. Tech Spec for SRO.

2 XCVLDNHXOU I (ATC) Letdown temperature control valve input fails low.

I (SRO) 3 CVC2P36BFAL C (ATC) B Charging pump supply breaker trips due to a motor C (SRO) fault.

4 RCSLOCATCB N (BOP) Shutdown required due to 15 gpm Reactor Coolant R (ATC) system leak on B cold leg. Tech Spec for SRO.

N (SRO) 5 MTGTRPLOCKO M (ALL) Main Turbine Generator trip causing a reactor trip and a FAILSU3 Startup #3 transformer lockout causing natural circulation of the RCS.

6 CV15042 C (BOP) #2 Emergency Diesel Generator service water outlet C (SRO) valve fails to open. This will result in de-energizing green train vital 4160VAC (2A-4) and vital 480VAC (2B-6).

7 RCSLOCATCB M (ALL) Reactor Coolant system B cold leg leakage ramps up to 225 gpm over 10 minutes.

8 HPISI10A C (BOP) 2P-89A Safety injection pump shaft shear.

C (SRO)

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total malfunctions. = 8, Malfunctions after EOP entry = 2, Abnormal events = 4, Major transient = 2, EOPs with substantive actions =1, EOP Contingencies = 0, Critical tasks = 2.

Revision 3 Page 1 of 28

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

1) Evaluate individual response to a failure of a Steam Generator Safety Channel Pressure detector.
2) Evaluate individual response to a failure of a Letdown temperature controller.
3) Evaluate individual response to a failure of a Coolant Charging pump.
4) Evaluate individual response to a small Reactor Coolant system leak.
5) Evaluate individual ability to perform a reduction in plant power.
6) Evaluate individual ability to respond to Degraded Electrical power.
7) Evaluate individual ability to mitigate a LOCA.
8) Evaluate individual response to a failure of cooling water to an Emergency Diesel Generator.
9) Evaluate individual ability to monitor operation of Emergency Core Cooling Equipment.

SCENARIO #1 NARRATIVE Simulator session begins with the plant at 100% power. Gland seal regulator, 2PCV-0231 is jacked closed.

When the crew has completed their control room walk down and brief, the A Steam Generator pressure safety channel pressure instrument, 2PT-1041-1, will fail low. This will trip one of the four PPS channels for low SG pressure trip and MSIS. Alarms for MSIS pre-trip, RPS channel trip/pre-trip, and channel A operator insert (2C03) trip and pre-trip lights will be lit. The SRO will refer to the ACA 2203.012D and tech specs 3.3.1.1, 3.3.2.1, 3.3.3.5 and 3.3.3.6 for guidance. The BOP will place Channel A PPS in bypass for point 11, SG pressure low, point 19, A SG delta-P for EFAS 1, and point 20, B SG delta-P for EFAS 2 for maintenance and trouble shooting. The crew will have one hour to place these points in bypass before exceeding the tech spec LCO.

After the A PPS channel is placed in bypass and cued by lead examiner, the Letdown heat exchanger temperature input will fail low. The ATC will report that LETDOWN HX 2E29 OUTLET TEMP HI alarm is in and the letdown heat exchanger temperature is reading low on the hand indicating controller but the computer point and control board indication are reading higher than normal. The SRO will direct the ATC to take manual control of the Letdown heat exchanger temperature control valve and manually control temperature. The SRO will also refer to the ACA for letdown radiation monitor flow low 2K12 J1 RADMONITOR FLOW LO and restore letdown radiation monitor flow.

After the ATC has taken manual control of the letdown temperature control valve, the B Coolant Charging Pump (CCP) will trip. The SRO will enter AOP 2203.036, Loss of Charging. The SRO will direct the ATC to manually start a standby CCP. The SRO should refer to the TRM and determine no LCOs are applicable.

Revision 3 Page 2 of 28

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

When a backup Charging pump has been started or at the lead examiners cue, a 15 gpm RCS leak will start. The SRO will enter the excess RCS leakage AOP, OP 2203.016. ATC and BOP will perform RCS Leak rate determinations. The SRO will enter Tech Spec 3.4.6.2. The SRO will direct to ATC to maintain pressurizer level within 5% of set point by starting additional charging pumps as needed. The SRO will also direct the ATC to isolate letdown to determine the leak location.

After the crew has determined the leak is not in letdown, they will restore letdown and the crew will perform a plant shutdown.

After the ATC has completed the required reactivity manipulation or cued by lead examiner, a Main Turbine Trip will occur and the crew will verify the reactor has tripped. When the reactor trips Start up transformer #3 will lock out. The RCPs and Circulating Water pumps lose power which will cause natural circulation conditions in the RCS and require steaming to atmosphere. During SPTAs, the BOP will report that #2 EDG service water outlet valve is not open requiring the #2 EDG to be secured locally.

The Crew will implement Standard Post Trip Actions (SPTA), OP 2202.001. After the reactor trips the RCS leakrate will ramp to 225 gpm over 10 minutes.

The SRO will diagnose a loss of coolant accident and enter EOP 2202.003, Loss of Coolant Accident (LOCA).

The ATC will cool down the RCS using the atmospheric dump valves. The BOP will override and open the Service Water to Component Cooling Water and Auxiliary Cooling Water valves. The ATC will control RCS pressure to recover pressurizer level. The BOP will take manual control of Emergency Feedwater flow and restore steam generator levels to normal.

Post SIAS, the A High Pressure Safety Injection (HPSI) pump will shear the shaft (red train) and the B high pressure safety injection pump (green train) will not be available due to loss of power.

The BOP will start the C HPSI pump to satisfy the required HSPI flow requirements.

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Appendix D Scenario 2 Form ES-D-1 Facility: ANO-2 Scenario No.: 2 (New) Op-Test No.: 2009-1 Examiners: Operators:

Initial Conditions:

100% MOL, All Engineered Safety Features systems are in standby. Gland Seal steam pressure control valve 2PCV-0231 jacked closed. RED Train Maintenance Week. 2T-6A Boric Acid Make up tank is aligned for acid reducing shut down chemistry for Unit 1 Turnover:

100%. 260 EFPD. EOOS indicates Minimal Risk. 2PCV-0231 jacked closed, manually control seal steam pressure. RED Train Maintenance Week. 2T-6A Boric Acid Make up tank is aligned for acid reducing shut down chemistry for Unit 1 refueling outage that starts next week.

Evolution scheduled: Drain Containment Sump to 50% level. Steps 20.1.1 and 20.1.2 of OP-2104.014 have been completed.

Event Malf. No. Event Event No. Type* Description 1 CV20612 C (BOP) Containment sump drain valve fails open during normal C (SRO) drain evolution. Tech Spec for SRO 2 SW2P-4A C (BOP) 2P-4A Service water pump breaker trips C (SRO) 3 XRCCHAPLVL I (ATC) A Pressurizer level channel fails High. Tech Spec for I (SRO) SRO 4 R (ATC) System Dispatcher call with a request to reduce power N (BOP) to ~ 850MWe(~80%) within 30 min.

N (SRO) 5 CV10101 M (ALL) A Steam Generator MSIV 2CV-1010-1 fails closed.

SGAMSIVBEF Then 4 minutes after trip an Excess Steam Demand on A S/G outside containment commences.

6 CEA38STUCK C (ATC) Control Element Assembly #38 remains fully CV48731 C (SRO) withdrawn when the reactor trips. The VCT outlet will not close and 2P-39B will trip when the reactor trips.

CVC2P39B 7 ESFCCAS2 C (BOP) Green train Containment coolers fail to automatically C (SRO) actuate.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total malfunctions. = 6, Malfunctions after EOP entry = 2, Abnormal events = 3, Major transient = 1, EOPs with substantive actions =1, EOP Contingencies = 1, Critical tasks = 2.

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

1) Evaluate individual response to a failure of a Containment isolation valve.
2) Evaluate individual ability to operate safety related equipment.
3) Evaluate individual response to a failure of a Service Water pump.
4) Evaluate individual response to a failure of a Pressurizer level control channel.
5) Evaluate individual ability to perform a reduction in plant power.
6) Evaluate individual response to a stuck out CEA after a Reactor Trip.
7) Evaluate individual response to a failed closed Main Steam Isolation Valve.
8) Evaluate individual ability to mitigate an Excess Steam Demand Outside containment.
9) Evaluate individual ability to monitor operation of Engineered Safety Features equipment.

SCENARIO #2 NARRATIVE Simulator session begins with the plant at 100% power steady state Gland seal regulator, 2PCV-0231 is jacked closed and 2T-6A is aligned for acid reducing chemistry.

When the crew has completed their control room walk down and brief, The BOP will drain the Containment sump to the Auxiliary building sump using the normal drain method. When the BOP attempts to close 2CV-2061-2 it remains open. The BOP should ensure the series sump drain valve 2CV-2060-1 is closed and sump level is stable. The SRO will determine that Tech Spec 3.6.3.1 is applicable and will enter Tech Spec 3.6.3.1.

After the Crew has entered the appropriate Tech Spec, verified sump level is stable and cued by lead examiner; the A Service Water Pump (2P-4A) breaker will trip open. The SRO will enter AOP, 2203.022 Loss of Service Water. The SRO will direct the BOP to start 2P-4B Service Water Pump.

After the B Service Water pump is placed in service and cued by lead examiner, the in service Pressurizer (PZR) control channel level instrument will fail high. This will result in letdown going to maximum flow, the two backup CCPs will automatically stop, all PZR heaters energizing and actual PZR level lowers. AOP 2203.028, PZR Systems Malfunctions, will be entered and actions directed by the SRO. The ATC will take the letdown flow controller to manual and control PZR level. The ATC will also verify that the other level control channel is reading correctly and select that channel for control and place the PZR low level cutoff switch to the unaffected channel. The SRO will enter Tech Spec 3.3.3.6 Post Accident Instrumentation. When the auto and manual signals are matched, the ATC will place the letdown flow controller to automatic.

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

After the ATC has placed B Channel pressurizer level control in service and restored letdown to automatic, The Pine Bluff Systems Operations Center (SOC) will call the Control Room with a Transmission Loading Relief (TLR) to reduce plant output by 200MWe. The SOC will also report that all limits of EN-DC-199 are still met. If Contacted, Unit 1 will be unable to maneuver due to a planned refueling outage.

After the ATC has completed the required reactivity manipulation and cued by lead examiner, 2CV-1010-1 A Steam Generator Main Steam Isolation Valve will fail closed. The crew will verify the reactor is tripped.

The Crew will implement Standard Post Trip Actions (SPTA), OP 2202.001. When the reactor trips one CEA will remain withdrawn which requires emergency boration. The VCT outlet will remain open and 2P-39B Boric Acid Makeup pump will trip when the reactor trips. Reactivity control will not be satisfied. During SPTAs, an Excess Steam Demand event outside of Containment will commence. The Crew will manually actuate Main Steam Isolation Signal (MSIS) or verify that a Main Steam Isolation signal automatically actuates. The Crew will secure and/or verify that Emergency Feedwater (EFW) is not feeding A Steam generator. The ATC will secure two Reactor coolant Pumps when RCS pressure goes below 1400 psia.

The SRO will diagnose Functional Recovery due to reactivity control not being met and enter OP 2202.009, Functional Recovery.

The crew will manually control Reactor Coolant Temperature once the A SG boils dry using the upstream Atmospheric Dump Valve on B SG. The ATC will control RCS pressure using normal spray.

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Appendix D Scenario 3 Form ES-D-1 Facility: ANO-2 Scenario No.: 3 (Modified) Op-Test No.: 2009-1 Examiners: Operators:

Initial Conditions:

~60% MOL, Condenser outer waterboxes placed in service after tube plugging. All Engineered Safety Features systems are in standby. 2PCV-0231 jacked closed. RED Train Maintenance Week.

Turnover:

~60% power Condenser outer waterboxes placed in service after tube plugging. 260 EFPD. EOOS indicates Minimal Risk. 2PCV-0231 jacked closed, manually control seal steam pressure. RED Train Maintenance Week. The previous shift has completed steps 10.1.1 through 10.1.6 and section 10.2 of starting the second heater drain pump and B Heater drain pump needs to be placed in service. 2 Condensate pumps are running with Condensate recirculation valves in manual.

Commence power escalation using applicable 2104.002 step 9.1 through 9.11 are complete and steps 9.13 to 9.18 and 9.20 to 9.28 are not applicable or complete. Reactor engineering recommends a 12%

per hour rate.

Event Malf. No. Event Event No. Type* Description 1 (New) N (SRO) Power escalation following a forced outage.

R (ATC) 2 (New) N (BOP) Place the B Heater Drain pump in service.

N (SRO) 3 NIBLINEPWR I (BOP) B channel nuclear instrument fails high. Tech Spec for (Modified) I (SRO) SRO.

4 XRCCHBPCNT I (ATC) B Pressurizer control channel pressure fails high.

(New) I (SRO) 5 SGBTUBE M (ALL) B Steam Generator tube leakage ramps up to 275 gpm (Modified) over 25 minutes. Manual reactor trip criteria when greater than 44 gpm. Tech Spec for SRO 6 RPSRXMAN C (ATC) Failure of the reactor protection system to automatically RPSRXAUTO C (SRO) or manually trip the reactor.

(New) 7 OP5130 C (BOP) Generator output breakers fail to open automatically OP5134 (New) C (SRO) 8 CNDVACPPA C (BOP) 2C-5A Vacuum pump trip CND2C5B C (SRO) 2C-5B Vacuum pump fails to auto start.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total malfunctions. = 6, Malfunctions after EOP entry = 2, Abnormal events = 2, Major transient = 1, EOPs with substantive actions =1, EOP Contingencies = 0, Critical tasks = 2.

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

1) Evaluate individual ability to perform a power escalation.
2) Evaluate individual ability to place the second heater drain pump in service.
3) Evaluate individual response to a failure of a Nuclear Instrumentation channel.
4) Evaluate individual response to a failure of a Pressurizer pressure control channel.
5) Evaluate individual response to a steam generator tube rupture.
6) Evaluate individual response to a failure of a RPS to manually or automatically trip the reactor.
7) Evaluate individual response to the generator output breakers failure to automatically open.
8) Evaluate individual ability to mitigate a steam generator tube rupture.
9) Evaluate individual response to a vacuum pump trip.
10) Evaluate individual response to a failure to the backup vacuum pump to auto start.

SCENARIO #3 NARRATIVE Simulator session begins with the plant at 60% power Condenser outer waterboxes placed in service after tube plugging. Gland seal regulator, 2PCV-0231 is jacked closed. The previous shift has completed steps 10.1.1 through 10.1.6 of OP-2106.016 (Condensate and Feedwater operations) for placing the B Heater Drain pump in service.

When the crew has completed their control room walk down and brief, the ATC will commence a power escalation and the BOP will place the B Heater drain pump in service.

After the B Heater Drain pump is placed in service, the ATC has completed all required reactivity manipulations and cued by lead examiner, B nuclear instrumentation channel will fail high. This will result in a PPS channel trip and pre-trip, CPC sensor failure alarms and the NI linear and log power instruments on B channel failing to High. The SRO will enter the Nuclear Instrument Malfunction AOP, AOP 2203.026. The SRO will direct the BOP to bypass points 1, 2, 3, and 4 on B PPS. TS 3.3.1.1 and 3.3.3.5 will be entered due to the excore failure.

After the BOP has bypassed points 1, 2, 3, and 4 on B PPS and cued by lead examiner, the B channel Pressurizer (PZR) control channel will fail high. This will result in both main spray valves to automatically open and both proportional heater banks to go to minimum fire. Actual PZR pressure will drop and will result in an automatic reactor trip, if actions are not taken to mitigate the event. The SRO will enter Abnormal operating procedure OP 2203.028, PZR systems malfunctions.

The SRO will direct selecting the opposite control channel.

After the Crew has taken all actions of OP 2203.028 PZR systems malfunctions and cued by lead examiner, a primary to secondary leak will start. The SRO will enter the primary to secondary leakage AOP, OP 2203.038. ATC and BOP will perform RCS Leak rate determinations. The SRO will enter Tech Spec 3.4.6.2. The SRO will direct the NLOs to control secondary contamination using standard attachment 19 and direct the chemists to sample the SGs for activity. When the crew has determined that the primary to secondary leak is greater than 44 gpm, the ATC will trip the reactor.

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

When the ATC attempts to manually trip the reactor RPS will not work and the SRO will direct the ATC to trip the reactor using the DSS system. The Crew will implement Standard Post Trip Actions (SPTA), OP 2202.001. During SPTAs the generator output breakers will fail to automatically open and require the BOP to manually open the breakers. The Crew should actuate safety injection (SIAS) and containment cooling actuation signal (CCAS) or verified that SIAS and CCAS automatically actuates. The Crew will lower Steam Dump master controller setpoint during SPTAs to aid in maintaining margin to saturation. The ATC will secure two Reactor coolant Pumps when RCS pressure goes below 1400 psia.

The SRO will diagnose a Steam Generator Tube Rupture event and enter OP 2202.004, Steam Generator Tube Rupture. The ATC will cool down the RCS to less than 535°F using the bypass valves to the condenser. The BOP will override SW to CCW and ACW. The BOP will isolate the B SG using standard attachment 10 when RCS Thot is less than 535°F.

After the EOP is entered, the RCS cooldown is started and at the lead examiners discretion, A vacuum pump will trip and B vacuum pump will NOT automatically start. The BOP will manually start B vacuum pump. If this vacuum pump is not started, the SDBCS bypass valves will close and the crew will have to steam the ruptured SG to atmosphere resulting in an off site radiological release.

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