ML052710137

09-2005-DRAFT-Outline

ML052710137
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/15/2005
From:	Gody A Operations Branch IV
To:	Forbes J Entergy Operations
References
50-313/05-301, ES-401-2, NUREG-1021, Draft Rev 9 50-313/05-301
Download: ML052710137 (37)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Arkansas Nuclear One Unit 1 RO Written Outline Date of Exam: 09/12/2005 RO K/A Category Points SRO - Only Points Tier Group K K K K K K A A A A G Total A2 G* TOTAL 1 2 3 4 5 6 1 2 3 4 *

1. 1 2 4 3 3 3 3 18 6 Emergency &

Abnormal Plant Evolutions 2 1 1 N/A 2 2 N/A 1 9 4 2

Tier Totals 4 5 4 5 5 4 27 10 1 2 3 3 3 2 2 3 2 3 3 2 28 5 2.

Plant 1 0 1 1 1 1 1 1 1 1 1 10 3 2

Systems Tier Totals 3 3 4 4 3 3 4 3 4 4 3 38 8

3. Generic Knowledge and Abilities Catergories 1 2 3 4 1 2 3 4 10 7 3 3 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and the 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 that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from 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.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IR) 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. 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 number, descriptions, importance ratings, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43 NUREG-1021, Revision 9 1 of 9

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier1 /Group1 (RO /SRO)

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

1 2 3 1 2 00007 (BW/E02 & E10; CE/E02) Reactor Trip X EK3.01 - Knowledge of the reasons for the following as they 4.0 1

- Stabilization - Recovery / 1 apply to Reactor Trip: Actions contained in EOP for reactor trip 00008 Pressurizer Vapor Space Accident / 3 X AK2.01 - Knowledge of the interrelations between the 2.7* 1 Pressurizer Vapor Space Accident and the following: Valves.

000009 Small Break LOCA / 3 X EA2.01 - Ability to determine or interpret the following as they 4.2 1 apply to a Small Break LOCA: Actions to be taken, based on RCS temperature and pressure, saturated and superheated.

000011 Large Break LOCA / 3 X EK2.02 - Knowledge of the interrelations between the Large 2.6* 1 Break LOCA and the following: Pumps.

000015/17 RCP Malfunctions / 4 X AK2.07 - Knowledge of the interrelations between the Reactor 2.9 1 Coolant Pump Malfunctions (Loss of RC Flow) and the following:

RCP seals.

000022 Loss of Rx Coolant Makeup / 2 X AK3.04 - Knowledge of the reasons for the following responses 3.2 1 as they apply to Loss of Rx Coolant Makeup: isolating letdown.

000025 Loss of RHR System / 4 X 2.4.10 - Knowledge of annunciator response procedures. 3.0 1 000026 Loss of Component Cooling X AA2.03 - Ability to determine and interpret the following as they 2.6 1 Water / 8 apply to the Loss of Component Cooling Water: The valve lineups necessary to restart the CCWS while bypassing the portion of the system causing the abnormal condition.

000027 Pressurizer Pressure Control X AK2.02 - Knowledge of the interrelations between the 2.4 1 System Malfunction / 3 Pressurizer Pressure Control Malfunctions and the following:

Sensors and detectors.

Justification for K/A <2.5:

Knowledge of interrelationship between sensors/detectors and control systems is important to a Reactor Operators duties of monitoring the control panels.

000029 ATWS / 1 Not selected.

000038 Steam Gen. Tube Rupture / 3 X EA1.32 - Ability to operate and monitor the following as they 4.6 1 apply to SGTR: Isolation of a ruptured S/G.

000040 (BW/E05; CE/E05; W/E12) X AK1.01 - Knowledge of the operational implications of the 4.1 1 Steam Line Rupture - Excessive Heat following concepts as they apply to Steam Line Rupture:

Transfer / 4 Consequences of PTS.

000054 (CE/E06) Loss of Main Not selected.

Feedwater / 4 000055 Station Blackout / 6 X 2.4.1 Knowledge of EOP entry conditions and immediate action 4.3 1 steps.

000056 Loss of Off-site Power / 6 X AK1.01 - Knowledge of the operational implications of the 3.7 1 following concepts as they apply to Loss of Off-site Power:

Principle of cooling by natural convection.

NUREG-1021, Revision 9 2 of 9

ES-401 2 Form ES-401-2 000057 Loss of Vital AC Inst. Bus / 6 X AA1.01 - Ability to operate and/or monitor the following as they 3.7 1 apply to the Loss of Vital AC Instrument Bus: Manual inverter swapping.

AK3.02 Knowledge of the reasons for the following responses as 000058 Loss of DC Power / 6 X they apply to the Loss of DC Power: Actions contained in EOP 4.0 1 for loss of DC power.

000062 Loss of Nuclear Svc Water / 4 X 2.4.9 - Knowledge of low power/shutdown implications in 3.3 1 accident (e.g. LOCA or loss of RHR) mitigation strategies.

AA2.05 - Ability to determine and interpret the following as they 000065 Loss of Instrument Air / 8 X apply to the Loss of Instrument Air: When to commence plant 3.4* 1 shutdown if instrument air pressure is decreasing.

W/E04 LOCA Outside Containment / 3 Not applicable to this unit.

W/E11 Loss of Emergency Coolant Not applicable to this unit.

Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer X EA1.2 - Ability to operate and/or monitor the following as they 3.4 1

- Loss of Secondary Heat Sink / 4 apply to the (Inadequate Heat Transfer): Operating behavior characteristics of the facility.

K/A Category Totals: 2 4 3 3 3 3 Group Point Total: 18 NUREG-1021, Revision 9 3 of 9

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier1 /Group2 (RO /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 Not selected.

000003 Dropped Control Rod / 1 X AA2.02 Ability to determine and interpret the following as 2.7 1 they apply to the Dropped Control Rod: Signal inputs to rod control system.

000005 Inoperable/Stuck Control Rod / 1 Not selected.

000024 Emergency Boration / 1 X AA1.03 Ability to operate and/or monitor the following as 3.5 1 they apply to the Emergency Boration: Boric acid controller.

000028 Pressurizer Level Malfunction / 2 Not selected.

000032 Loss of Source Range NI / 7 Not selected.

000033 Loss of Intermediate Range NI / 7 Not selected.

AK2.02 Knowledge of the interrelations between the Fuel 000036 (BW/A08) Fuel Handling Accident / 8 X Handling Accident and the following: Radiation monitoring 3.4 1 equipment (portable and installed).

000037 Steam Generator Tube Leak / 3 X AK1.02 Knowledge of the operational implications of the 3.5 1 following concepts as they apply to the Steam Generator Tube Leak: Leak rate vs. pressure drop.

000051 Loss of Condenser Vacuum / 4 X 2.4.11 3.4 1 Knowledge of abnormal condition procedures.

Not selected.

000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 X AK3.02 Knowledge of the reasons for the following 3.3 1 responses as they apply to the Accidental Gaseous Radwaste Release: Isolation of the auxiliary building ventilation.

000061 ARM System Alarms / 7 Not selected.

000067 Plant Fire On-site / 8 Not selected.

000068 (BW/A06) Control Room Evac. / 8 X AA1.02 Ability to operate and/or monitor the following as 4.3 1 they apply to the Control Room Evacuation: AFW emergency pump.

000069 (W/E14) Loss of CTMT Integrity / 5 Not selected.

000074 (W/E06&E07) Inad. Core Cooling / 4 Not selected.

000076 High Reactor Coolant Activity / 9 Not Selected.

W/EO1 & E02 Rediagnosis & SI Termination / 3 Not applicable to this Unit.

W/E13 Steam Generator Over-pressure / 4 Not applicable to this Unit.

W/E15 Containment Flooding / 5 Not applicable to this Unit.

W/E16 High Containment Radiation / 9 Not applicable to this Unit.

BW/A01 Plant Runback / 1 Not selected.

BW/A02&A03 Loss of NNI-X/Y / 7 Not selected.

Not selected.

BW/A04 Turbine Trip / 4 NUREG-1021, Revision 9 5 of 9

ES-401 3 Form ES-401-2 BW/A05 Emergency Diesel Actuation / 6 X AK1.3 Knowledge of the operational implications of the 3.8 1 following concepts as they apply to the (Emergency Diesel Actuation): Annunciators and conditions indicating signals, and remedial actions associated with the (Emergency Diesel Actuation).

BW/A07 Flooding / 8 Not selected.

BW/E03 Inadequate Subcooling Margin / 4 X EA2.1 Ability to determine and interpret the following as they 3.0 1 apply to the (Inadequate Subcooling Margin): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

BW/E08; W/E03 LOCA Cooldown - Depress. / 4 Not selected.

BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 Not selected.

BW/E13&E14 EOP Rules and Enclosures Not selected.

CE/A11; W/E08 RCS Overcooling - PTS / 4 Not applicable to this Unit.

CE/A16 Excess RCS Leakage / 2 Not applicable to this Unit.

CE/E09 Functional Recovery Not applicable to this Unit.

K/A Category Point Totals: 2 1 1 2 2 1 9 NUREG-1021, Revision 9 5 of 9

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant systems - Tier 2/Group 1 (RO / 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 X K6.14 Knowledge of the effect of a loss or 2.6 1 malfunction of the following will have on the RCPS: Starting requirements.

K3.05 Knowledge of the effect that a loss or 004 Chemical and Volume Control X malfunction of the CVCS will have on the 3.8 1 following: PZR LCS.

A1.01 Ability to predict and/or monitor changes 005 Residual Heat Removal X in parameters (to prevent exceeding design 3.5 1 limits) associated with operating the RHRS controls including: heatup/cooldown rates.

A4.02 Ability to manually operate and/or monitor 005 Residual Heat Removal X in the control room: Heat exchanger bypass flow 3.4 1 control.

006 Emergency Core Cooling X K6.02 Knowledge of the effect of a loss or 3.4 1 malfunction of the following will have on the ECCS system: Core flood tanks (accumulators).

006 Emergency Core Cooling X 2.4.6 Knowledge of symptom based EOP 3.1 1 mitigation strategies.

007 Pressurizer Relief/ Quench Tank X K5.02 Knowledge of the operational implications 3.1 1 of the following concepts as they apply to the PRTS: Method of forming a steam bubble in the PZR.

007 Pressurizer Relief/ Quench Tank X A1.01 Ability to predict and/or monitor changes 2.9 1 in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining Quench Tank water level within limits.

008 Component Cooling Water X K2.02 Knowledge of bus power supplies to the 3.0 1 following: CCW pump, including emergency backup.

010 Pressurizer Pressure Control X A2.03 Ability to (a) predict the impacts of the 4.1 1 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: PORV failures.

012 Reactor Protection X K5.01 Knowledge of the operational implications 3.3 1 of the following concepts as they apply to the RPS: DNB.

012 Reactor Protection X A4.04 Ability to manually operate and/or monitor 3.3 1 in the control room: Bistable trips, reset and test switches.

013 Engineered Safety Features Actuation X K1.06 Knowledge of the physical connections 4.2 1 and/or cause-effect relationships between the ESFAS and the following systems: ECCS.

2.4.50 Ability to verify system alarm setpoints 013 Engineered Safety Features Actuation X and operate controls identified in the alarm 3.3 1 response manual.

A3.01 Ability to monitor automatic operation of 022 Containment Cooling X the CCS, including: Initiation of safeguards mode 4.1 1 of operation.

025 Ice Condenser Not applicable to this Unit.

026 Containment Spray X K4.06 Knowledge of CSS design feature(s) 2.8 1 and/or interlock(s) which provide for the following: Iodine scavenging via the CSS.

NUREG-1021, Revision 9 7 of 9

ES-401 4 Form ES-401-2 A1.06 Ability to predict and/or monitor changes 039 Main and Reheat Steam X in parameters (to prevent exceeding design 3.0 1 limits) associated with operating the MRSS controls including: Main Steam pressure.

059 Main Feedwater X K1.07 Knowledge of the physical connections 3.2 1 and/or cause-effect relationships between the MFW and the following systems: ICS (FWCS).

059 Main Feedwater X A3.04 Ability to monitor automatic operation of 2.5 1 the MFW, including: Turbine driven feed pump.

061 Auxiliary/Emergency Feedwater X K2.02 Knowledge of bus power supplies to the 3.7 1 following: AFW electric drive pumps.

062 AC Electrical Distribution X A2.04 Ability to (a) predict the impacts of the 3.1 1 following malfunctions or operations on the AC Distribution System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effect on plant of de-energizing a bus.

062 AC Electrical Distribution X A4.01 Ability to manually operate and/or monitor 3.3 1 in the control room: All breakers (including available switchyard).

063 DC Electrical Distribution X K3.02 Knowledge of the effect that a loss or 3.5 1 malfunction of the DC Electrical Distribution System will have on the following: Components using DC control power.

064 Emergency Diesel Generator X A3.04 Ability to monitor automatic operation of 3.1 1 the ED/G system, including: Number of starts available with an air compressor.

073 Process Radiation Monitoring X K4.01 Knowledge of PRM System design 4.0 1 feature(s) and/or interlock(s) which provide for the following: Release termination when radiation exceeds setpoint.

076 Service Water X K2.01 Knowledge of bus power supplies to the 2.7 1 following: Service Water.

078 Instrument Air X K4.02 Knowledge of IAS design feature(s) 3.2 1 and/or interlock(s) which provide for the following: Cross-over to other air systems.

103 Containment X K3.02 Knowledge of the effect that a loss or 3.8 1 malfunction of the Containment System will have on the following: Loss of containment integrity under normal operations.

K/A Category Point Totals: 2 3 3 3 2 2 3 2 3 3 2 Group Point Total: 28 NUREG-1021, Revision 9 7 of 9

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant systems - Tier 2/Group 2 (RO / 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 X A3.07 Ability to monitor automatic operation of the 4.1 1 CRDS, including: Boration/dilution.

002 Reactor Coolant X A2.03 Ability to (a) predict the impacts of the following 4.1 1 malfunctions or operations on the RCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of forced circulation.

011 Pressurizer Level Control Not selected.

014 Rod Position Indication Not selected.

015 Nuclear Instrumentation X K3.01 Knowledge of the effect that a loss or 3.9 1 malfunction of the NIS will have on the following:

RPS.

016 Non-nuclear Instrumentation Not selected.

017 In-core Temperature Monitor X 2.1.33 Ability to recognize indications for system 3.4 1 operating parameters which are entry-level conditions for technical specifications.

027 Containment Iodine Removal Not selected.

028 Hydrogen Recombiner and Purge Control Not selected.

029 Containment Purge Not selected.

033 Spent Fuel Pool Cooling X K1.02 Knowledge of the physical connections and/or 2.5 1 cause-effect relationships between the Spent Fuel Cooling System and the following systems: RHRS.

034 Fuel Handling Equipment Not selected.

035 Steam Generator K4.01 Knowledge of S/GS design feature(s) and/or X interlock(s) which provide for the following: S/G level 3.6 1 control.

041 Steam Dump/Turbine Bypass Control X K5.07 Knowledge of the operational implications of 3.1 1 the following concepts as they apply to the SDS:

Reactivity feedback effects.

045 Main Turbine Generator X A1.06 Ability to predict and/or monitor changes in 3.3 1 parameters (to prevent exceeding design limits) associated with operating the MT/G System controls including: Expected response of secondary plant parameters following a T/G trip.

055 Condenser Air Removal Not selected.

056 Condensate Not selected.

068 Liquid Radwaste Not selected.

071 Waste Gas Disposal X A4.25 Ability to manually operate and/or monitor in 3.2 1 the control room: Setting of process radiation monitor alarms, automatic functions, and adjustment of setpoints.

072 Area Radiation Monitoring Not selected.

075 Circulating Water Not selected.

079 Station Air Not selected.

086 Fire Protection X K6.04 Knowledge of the effect of a loss or malfunction 2.6 1 of the following will have on the Fire Protection System: Fire, smoke, and heat detectors.

K/A Category Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10 NUREG-1021, Revision 9 8 of 9

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-2 Facility: Arkansas Nuclear One Unit 2 RO Written Outline Date of Exam: 01/21/2005 Category K/A # Topic RO IR #

2.1 2.1.7 Ability to evaluate plant performance and make operational judgments 3.7 1 based on operating characteristics, reactor behavior, and instrument interpretation.

1.

Conduct of 2.1 2.1.20 Ability to execute procedure steps. 4.3 1 Operations 2.1 2.1.22 Ability to determine mode of operation. 2.8 1 2.1 2.1 2.1 Subtotal 3 2.2 2.2.1 Ability to perform pre-startup procedures for the facility, including 3.7 1 operating those controls associated with plant equipment that could affect reactivity.

2. 2.2 2.2.13 Knowledge of tagging and clearance procedures. 3.6 1 Equipment 2.2 2.2.33 Knowledge of control rod programming. 2.5 1 Control 2.2 2.2 Subtotal 3 2.3 2.3.4 Knowledge of radiation exposure limits and contamination control, 2.5 1 including permissible levels in excess of those authorized.

2.3 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and 2.9 1

3. guard against personnel exposure.

Radiation Control 2.3 2.3 2.3 2.3 Subtotal 2 2.4 2.4.17 Knowledge of EOP terms and definitions. 3.1 1 2.4 2.4.29 Knowledge of the emergency plan. 2.6 1 2.4 4.

Emergency 2.4 Procedures/ Plan 2.4 Subtotal 2 Tier 3 Point Total 10 NUREG-1021, Revision 9 9 of 9

ES-401 PWR Examination Outline Form ES-401-2 Facility: Arkansas Nuclear One Unit 1 SRO Written Outline Date of Exam: 09/12/2005 RO K/A Category Points SRO - Only Points Tier Group K K K K K K A A A A G Total A2 G* TOTAL 1 2 3 4 5 6 1 2 3 4 *

1. 1 4 2 6 Emergency &

Abnormal Plant Evolutions N/A N/A 3 4 2 1 Tier Totals 7 3 10 1 4 1 5 2.

Plant 2 3 2 1 Systems Tier Totals 6 2 8

3. Generic Knowledge and Abilities Catergories 1 2 3 4 7 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 Outline and the 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 that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from 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.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IR) 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. 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 number, descriptions, importance ratings, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43 NUREG-1021, Revision 9 1 of 8

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier1 /Group1 (RO /SRO)

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

1 2 3 1 2 00007 (BW/E02 & E10; CE/E02) Reactor Trip Not selected.

- Stabilization - Recovery / 1 Not selected.

00008 Pressurizer Vapor Space Accident / 3 2.4.9 Knowledge of low power / shutdown implications in 000009 Small Break LOCA / 3 X accident (e.g. LOCA or loss of RHR) mitigation strategies. 3.9 76 000011 Large Break LOCA / 3 Not selected.

Not selected.

000015/17 RCP Malfunctions / 4 Not selected.

000022 Loss of Rx Coolant Makeup / 2 Not selected.

000025 Loss of RHR System / 4 Not selected.

000026 Loss of Component Cooling Water / 8 Not selected.

000027 Pressurizer Pressure Control System Malfunction / 3 EA2.02 Ability to determine and interpret the following as 000029 ATWS / 1 X they apply to the ATWS: Reactor trip alarm. 4.4 77 EA2.01 - Ability to determine or interpret the following as 000038 Steam Gen. Tube Rupture / 3 X they apply to the SGTR: When to isolate one or more 4.7 78 S/Gs.

000040 (BW/E05; CE/E05; W/E12) X 2.4.6 - Knowledge of symptom based EOP mitigation 4.0 79 Steam Line Rupture - Excessive Heat strategies.

Transfer / 4 AA2.06 - Ability to determine and interpret the following 000054 (CE/E06) Loss of Main X as they apply to the Loss of Main Feedwater (MFW): 4.3 80 Feedwater / 4 AFW adjustments needed to maintain proper T-ave and S/G level.

000055 Station Blackout / 6 Not selected.

000056 Loss of Off-site Power / 6 Not selected.

000057 Loss of Vital AC Inst. Bus / 6 X AA2.02 - Ability to determine and interpret the following 3.8 81 as they apply to the Loss of Vital AC Inst. Bus: Core flood tank pressure and level indicators.

000058 Loss of DC Power / 6 Not selected.

000062 Loss of Nuclear Svc Water / 4 Not selected.

000065 Loss of Instrument Air / 8 Not selected.

Not applicable to this Unit.

W/E04 LOCA Outside Containment / 3 Not applicable to this Unit.

W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer Not selected.

- Loss of Secondary Heat Sink / 4 K/A Category Totals: 4 2 Group Point Total: 6 NUREG-1021, Revision 9 2 of 8

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier1 /Group2 (RO /SRO)

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

1 2 3 1 2 Not selected.

000001 Continuous Rod Withdrawal / 1 Not selected.

000003 Dropped Control Rod / 1 AA2.03 - Ability to determine and interpret the following 000005 Inoperable/Stuck Control Rod / 1 X as they apply to the Inoperable/Stuck Control Rod: 4.4 82 Required actions if more than one rod is stuck or inoperable.

Not selected.

000024 Emergency Boration / 1 Not selected.

000028 Pressurizer Level Malfunction / 2 Not selected.

000032 Loss of Source Range NI / 7 AA2.10 - Ability to determine and interpret the following 000033 Loss of Intermediate Range NI / 7 X as they apply to the Loss of Intermediate Range Nuclear 3.8 83 Instrumentation: Tech-spec limits if both intermediate-range channels have failed.

Not selected.

000036 (BW/A08) Fuel Handling Accident / 8 Not selected.

000037 Steam Generator Tube Leak / 3 Not selected.

000051 Loss of Condenser Vacuum / 4 Not selected.

000059 Accidental Liquid RadWaste Rel. / 9 Not selected.

000060 Accidental Gaseous Radwaste Rel. / 9 Not selected.

000061 ARM System Alarms / 7 Not selected.

000067 Plant Fire On-site / 8 Not selected.

000068 (BW/A06) Control Room Evac. / 8 Not selected.

000069 (W/E14) Loss of CTMT Integrity / 5 Not selected.

000074 (W/E06&E07) Inad. Core Cooling / 4 Not selected.

000076 High Reactor Coolant Activity / 9 Not applicable to this Unit.

W/EO1 & E02 Rediagnosis & SI Termination / 3 Not applicable to this Unit.

W/E13 Steam Generator Over-pressure / 4 Not applicable to this Unit.

W/E15 Containment Flooding / 5 Not applicable to this Unit.

W/E16 High Containment Radiation / 9 Not selected.

BW/A01 Plant Runback / 1 AA2.1 - Ability to determine and interpret the following BW/A02&A03 Loss of NNI-X/Y / 7 X as they apply to the (NNI-X): Facilty conditions and 4.0 84 selection of appropriate procedures during abnormal and emergency operations.

Not selected..

BW/A04 Turbine Trip / 4 Not selected.

BW/A05 Emergency Diesel Actuation / 6 Not selected.

BW/A07 Flooding / 8 NUREG-1021, Revision 9 4 of 8

ES-401 2 Form ES-401-2 Not selected.

BW/E03 Inadequate Subcooling Margin / 4 2.4.16 Knowledge of EOP implementation hierarchy and BW/E08; W/E03 LOCA Cooldown - Depress. / 4 X coordination with other support procedures. 4.0 85 Not selected.

BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 Not selected.

BW/E13&E14 EOP Rules and Enclosures Not applicable to this Unit.

CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 Not applicable to this Unit.

CE/E09 Functional Recovery Not applicable to this Unit.

K/A Category Point Totals: 3 1 Group Point Total: 4 NUREG-1021, Revision 9 4 of 8

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant systems - Tier 2/Group 1 (RO / 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 Not selected.

004 Chemical and Volume Control Not selected.

005 Residual Heat Removal X A2.02 - Ability to (a) predict the impacts of 3.7 86 the following malfunctions or operations on the RHRS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Pressure transient protection during cold shutdown.

006 Emergency Core Cooling Not selected.

007 Pressurizer Relief/ Quench Tank 2.2.22 - Knowledge of limiting conditions X for operation and safety limits. 4.1 87 008 Component Cooling Water Not selected.

010 Pressurizer Pressure Control Not selected.

012 Reactor Protection A2.03 Ability to (a) predict the impacts of X the following malfunctions or operations 3.7 88 on the RPS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Incorrect channel bypassing.

013 Engineered Safety Features Actuation Not selected.

Not selected.

022 Containment Cooling 025 Ice Condenser Not applicable to this unit.

026 Containment Spray Not selected.

039 Main and Reheat Steam Not selected.

059 Main Feedwater Not selected 061 Auxiliary/Emergency Feedwater Not selected.

062 AC Electrical Distribution Not selected.

063 DC Electrical Distribution Not selected.

064 Emergency Diesel Generator X A2.02 Ability to (a) predict the impacts of 2.9 89 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: Load, VARS, pressure on air compressor, speed droop, frequency, voltage, fuel oil level, temperatures.

073 Process Radiation Monitoring Not selected.

076 Service Water A2.01 Ability to (a) predict the impacts of X the following malfunctions or operations 3.7 90 on the SWS System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of SWS.

078 Instrument Air Not selected.

NUREG-1021, Revision 9 5 of 8

ES-401 4 Form ES-401-2 103 Containment Not selected.

K/A Category Point Totals: 4 1 Group Point Total: 5 NUREG-1021, Revision 9 5 of 8

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant systems - Tier 2/Group 2 (RO / 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 Not selected.

002 Reactor Coolant Not selected.

011 Pressurizer Level Control Not selected.

014 Rod Position Indication A2.05 Ability to (a) predict the impacts of X the following malfunctions or operations 4.1 91 on the RPIS System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Reactor Trip.

015 Nuclear Instrumentation Not selected.

016 Non-nuclear Instrumentation 2.1.12 Ability to apply technical X specifications for a system. 4.0 92 017 In-core Temperature Monitor Not selected.

027 Containment Iodine Removal Not selected.

028 Hydrogen Recombiner and Purge Control Not selected.

029 Containment Purge Not selected.

033 Spent Fuel Pool Cooling Not selected.

034 Fuel Handling Equipment X A2.05 Ability to (a) predict the impacts of 4.4 93 the following malfunctions or operations on the Fuel Handling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped fuel element.

035 Steam Generator Not selected.

041 Steam Dump/Turbine Not selected.

Bypass Control 045 Main Turbine Generator Not selected.

055 Condenser Air Removal Not selected.

056 Condensate Not selected.

068 Liquid Radwaste Not selected.

071 Waste Gas Disposal Not selected.

072 Area Radiation Monitoring Not selected.

075 Circulating Water Not selected.

079 Station Air Not selected.

086 Fire Protection Not selected.

K/A Category Totals: 2 1 Group Point Total: 3 NUREG-1021, Revision 9 6 of 8

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Arkansas Nuclear One Unit 2 SRO Written Outline Date of Exam: 01/21/2005 Category K/A # Topic SRO-Only IR #

2.1 2.1.7 Ability to evaluate plant performance and make operational judgments based on 4.4 94 operating characteristics, reactor behavior, and instrument interpretation.

2.1 2.1.11 Knowledge of less than one hour technical specification action statements for 3.8 95

1. systems.

Conduct of Operations 2.1 2.1 2.1 2.1 Subtotal 2 2.2 2.2.11 Knowledge of the process for controlling temporary changes. 3.4 96 2.2 2.2.22 Knowledge of limiting conditions for operations and safety limits. 4.1 97 2.2 2.

Equipment 2.2 Control 2.2 2.2 Subtotal 2 2.3 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements. 3.0 98 2.3 2.3.8 Knowledge of the process for performing a planned gaseous radioactive release. 3.2 99 2.3 3.

Radiation Control 2.3 2.3 2.3 Subtotal 2 2.4 2.4. 22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency 4.0 100 operations.

2.4

4. 2.4 Emergency 2.4 Procedures/ Plan 2.4 2.4 Subtotal 1 Tier 3 Point Total 7 NUREG-1021, Revision 9 7 of 8

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Group Selected K/A Reason for Rejection 1/2 000001 AA2.02 ANO-1 does not have Emergency Boration valves.

Replaced with randomly selected 000003 AA2.02.

1/2 000068 AA1.15 Replaced due to double jeopardy with K/A in Tier 1 Group

1. Replaced with randomly selected 068 AA1.02.

1/2 CE/A13 EK2.2 ANO-1 is a B&W unit, not CE. Replaced with randomly selected BW/A03 EA2.1.

1/2 CE/E09 EK1.1 ANO-1 is a B&W unit, not CE. Replaced with randomly selected BW/A05 AK1.1.

2/1 007 A3.01 Replaced due to clues to answer being given in questions to be developed for other K/As. Replaced with randomly selected 2/1 013 K1.18 ANO-1 cannot prematurely reset ESF. Replaced with randomly selected 013 K1.06.

2/1 026 2.2.30 ANO-1 Containment Spray is not related to fuel handling.

Replaced with 006 Gen 2.4.6 since 026 already had a KA randomly selected.

2/1 059 A3.02 ANO-1 does not have programmed levels for SGs. Replaced with randomly selected 059 A3.04.

2/1 061 K2.03 ANO-1 does not have diesel driven AFW pumps. Replaced with randomly selected 061 K2.02.

2/1 062 A4.02 ANO-1 does not have Control Room remote racking capability for breakers. Replaced with randomly selected 062 A4.01.

2/1 073 K4.02 ANO-1 does not have automatic Letdown isolation on high RCS activity. Replaced with randomly selected 073 K4.01.

2/1 073 2.2.30 This KA was rejected since it was too similar to a KA selected for an administrative JPM and thus would compromise another part of the exam. Replaced with 005 A4.02 since 026 already had a KA randomly selected.

2/2 001 K2.03 Could not write multiple choice question on one line diagrams. Replaced with randomly selected 001 A3.07.

2/2 027 A2.01 ANO-1 does not have a Containment Iodine Removal filter system. Replaced with randomly selected 002 A2.03.

3 2.1.16 An adequate written exam question could not be constructed on the ability to operate plant phone.

NUREG-1021, Draft Revision 9

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Group Selected K/A Reason for Rejection 1/1 0040 2.2.31 Generic K/A concerned fuel loading which is not consistent with Steam Line Rupture. Replaced with 2.4.6.

1/2 B/W E08 2.2.29 Generic K/A concerned fuel handling which is not consistent with LOCA cooldown. Replaced with 2.4.16.

2/2 016 2.3.2 Generic K/A concerned ALARA which is not consistent with Non-nuclear instrumentation. Replaced with 2.1.12.

3 2.1.25 This K/A does not lend itself to construction of a closed reference since it is based on the ability to interpret drawings, graphs, etc. Replaced with 2.1.11.

3 2.2.25 This K/A was too similar to the other randomly selected K/A for this group, 2.2.22. Replaced with 2.2.11.

3 2.4.49 This K/A does not lend itself to construction of a closed reference for SROs since it is based on the ability to perform immediate action steps which an RO must know. Replaced with 2.4.22.

ES-401, Page 27 of 33

ES-301 Administrative Topics Outline Form ES-301-1 Facility: ANO-1 Date of Examination: 9-12-05 Examination Level (circle one): RO / SRO Operating Test Number: 1 Type Administrative Topic Describe activity to be performed Code*

(see note)

Conduct of Operations Ability to use plant computer to obtain and evaluate M/S parametric information on system or component status.

A1JPM-RO-PMS1 Rev. 2 2.1.19 (Imp 3.0)

Ability to perform specific and integrated plant Conduct of Operations D/S procedures during all modes of operation.

A1JPM-RO-RBAL3 Rev. 1 2.1.23 (Imp 3.9)

Equipment Control Knowledge of refueling administration requirements.

N/S 2.2.26 A1JPM-RO-REFUL1 Rev. 0 (Imp 2.5)

Radiation Control Ability to control radiation releases.

D/P/S A1JPM-RO-RAD1 Rev. 2 2.3.11 (Imp 2.7)

Emergency 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 (D)irect from bank ( 3 for ROs; 4 for SROs)

(N)ew or (M)odified from bank ( 1 (P)revious ( 1; randomly selected)

(S)imulator ES-301, Page 22 of 27

ES-301 Administrative Topics Outline Form ES-301-1 Facility: ANO-1 Date of Examination: 9-12-05 Examination Level (circle one): RO / SRO Operating Test Number: 1 Type Administrative Topic Describe activity to be performed Code*

(see note)

Conduct of Operations Ability to apply technical specifications for a system.

N/S New admin JPM A1JPM-SRO-TS1 2.1.12 (Imp 4.0)

Ability to maintain primary and secondary plant Conduct of Operations N/S chemistry within allowable limits.

2.1.34 New admin JPM A1JPM-SRO-CHEM1 (Imp2.9)

Equipment Control Knowledge of the process for controlling temporary N/S changes.

2.2.11 New admin JPM A1JPM-SRO-TALT2 (Imp 3.4)

Radiation Control Ability to perform procedures to reduce excessive M/S levels of radiation and guard against personnel 2.3.10 exposure.

(Imp 3.3) Modified A1JPM-RO-RC22 Emergency Plan Knowledge of the emergency plan.

N/S New admin JPM based on scenario developed.

2.4.29 (Imp 4.0)

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 (D)irect from bank ( 3 for ROs; 4 for SROs)

(N)ew or (M)odified from bank ( 1 (P)revious ( 1; randomly selected)

(S)imulator ES-301, Page 22 of 27

ES-301 Control Room / in Plant Systems Outline Form ES-301-2 Facility: _ANO-1______________ Date of Examination: _ 9/12/2005 ____

Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___One________

Control Room Systems@ ( 8 for RO; 7 for SRO-I; 2 or 3 for SRO- U)

System / JPM Title Type Code* Safety Function

a. A1JPM-RO-EOP07, Perform Reactor Trip Immediate Actions A/D/S 1 007 EK3.01 (RO 4.0/SRO 4.6) Reactivity

b. A1JPM-RO-LTOP1, Establish LTOP protection during cool down of RCS D/L/S 3 002 K4.10 (RO 4.2/SRO 4.4) Reactor Pressure Control

c. A1JPM-RO-RCP05, Shutdown of P-32C and P-32D after Decay Heat in service D/L/S 4 003 A4.03 (RO 2.8/SRO 2.5) Reactor Heat Removal (Primary)

d. A1JPM-RO-EOP16, Perform actions required to correct overcooling of the RCS A/D/S 4 039 A2.04 (RO 3.4/SRO 3.7) Reactor Heat Removal (Secondary)

e. A1JPM-RO-HYD03, Place Hydrogen Recombiner M-55B in operation C/D/L/S 5 028 A4.01 (RO 4.0/SRO 4.0) Containment Integrity

f. A1JPM-RO-EDO08, Shift buses A1, A2, H1, H2 from SU#1 to Unit Aux A/D/S 6 062 A4.07 (RO 3.1/SRO 3.1) Electrical

g. A1JPM-RO-ARM01, Respond to Area Rad Monitor alarm C/N/S 7 072 A4.01 (RO 3.0/SRO 3.3) Instrumentation

h. A1JPM-RO-AOP28, Respond to lo-lo Instrument Air pressure D/S 8 065 AK3.08 (RO 3.7/SRO 3.9) Plant Service Systems In- Plant Systems@ ( 3 for RO; 3 for SRO-I; 3 or 2 for SRO- U)

i. A1JPM-RO-CA01, Borate via alternate path bypassing batch controller N/R 1 004 K6.13 (RO 3.1/SRO 3.3) Reactivity

j. A1JP-RO-EFW02, Manually control P-7A at turbine D/E/R 4 061 A2.05 (RO 3.1/SRO 3.4) Reactor Heat Removal (Secondary)

k. A1JPM-RO-EDO30, Place battery charger D-03B in service A/D/P 6 2.1.30 (RO 3.9/SRO 3.4) Electrical

@ All control room (and in-plant) systems must be different and serve different safety functions; Type Codes Criteria for RO /SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)direct from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 (P)revious 2 Exams 3 / 3 / 2 (randomly selected)

(R)CA 1 / 1 / 1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room / in Plant Systems Outline Form ES-301-2 Facility: _ANO-1______________ Date of Examination: _ 9/12/2005 ____

Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___One________

Control Room Systems@ ( 8 for RO; 7 for SRO-I; 2 or 3 for SRO- U)

System / JPM Title Type Code* Safety Function

a. A1JPM-RO-EOP07, Perform Reactor Trip Immediate Actions A/D/S 1 007 EK3.01 (RO 4.0/SRO 4.6) Reactivity

b. A1JPM-RO-LTOP1, Establish LTOP protection during cool down of RCS D/L/S 3 002 K4.10 (RO 4.2/SRO 4.4) Reactor Pressure Control

c. A1JPM-RO-RCP05, Shutdown of P-32C and P-32D after Decay Heat in service D/L/S 4 003 A4.03 (RO 2.8/SRO 2.5) Reactor Heat Removal (Primary)

d. A1JPM-RO-EOP16, Perform actions required to correct overcooling of the RCS A/D/S 4 039 A2.04 (RO 3.4/SRO 3.7) Reactor Heat Removal (Secondary)

e. A1JPM-RO-HYD03, Place Hydrogen Recombiner M-55B in operation C/D/L/S 5 028 A4.01 (RO 4.0/SRO 4.0) Containment Integrity

f. A1JPM-RO-EDO08, Shift buses A1, A2, H1, H2 from SU#1 to Unit Aux A/D/S 6 062 A4.07 (RO 3.1/SRO 3.1) Electrical

g. A1JPM-RO-ARM01, Respond to Area Rad Monitor alarm C/N/S 7 072 A4.01 (RO 3.0/SRO 3.3) Instrumentation h.

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

i. A1JPM-RO-CA01, Borate via alternate path bypassing batch controller N/R 1 004 K6.13 (RO 3.1/SRO 3.3) Reactivity

j. A1JP-RO-EFW02, Manually control P-7A at turbine D/E/R 4 061 A2.05 (RO 3.1/SRO 3.4) Reactor Heat Removal (Secondary)

k. A1JPM-RO-EDO30, Place battery charger D-03B in service A/D/P 6 2.1.30 (RO 3.9/SRO 3.4) Electrical

@ All control room (and in-plant) systems must be different and serve different safety functions; Type Codes Criteria for RO /SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)direct from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 (P)revious 2 Exams 3 / 3 / 2 (randomly selected)

(R)CA 1 / 1 / 1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room / in Plant Systems Outline Form ES-301-2 Facility: _ANO-1______________ Date of Examination: _ 9/12/2005 ____

Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ___One________

Control Room Systems@ ( 8 for RO; 7 for SRO-I; 2 or 3 for SRO- U)

System / JPM Title Type Code* Safety Function

a. A1JPM-RO-EOP07, Perform Reactor Trip Immediate Actions A/D/S 1 007 EK3.01 (RO 4.0/SRO 4.6) Reactivity

b. A1JPM-RO-LTOP1, Establish LTOP protection during cool down of RCS D/L/S 3 002 K4.10 (RO 4.2/SRO 4.4) Reactor Pressure Control C/N/S 7

c. A1JPM-RO-ARM01, Respond to Area Rad Monitor alarm 072 A4.01 (RO 3.0/SRO 3.3) Instrumentation d.

e.

f.

g.

h.

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

i. A1JP-RO-EFW02, Manually control P-7A at turbine D/E/R 4 061 A2.05 (RO 3.1/SRO 3.4) Reactor Heat Removal (Secondary)

j. A1JPM-RO-EDO30, Place battery charger D-03B in service A/D/P 6 2.1.30 (RO 3.9/SRO 3.4) Electrical k.

@ All control room (and in-plant) systems must be different and serve different safety functions; Type Codes Criteria for RO /SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)direct from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 (P)revious 2 Exams 3 / 3 / 2 (randomly selected)

(R)CA 1 / 1 / 1 (S)imulator ES-301, Page 23 of 27

Appendix D Scenario 1 Form ES-D-1 Facility: ANO-1 Scenario No.: 1 Op-Test No.:2005-1 Page 1 Examiners: Operators:

Initial Conditions:

70% power holding for the start of P8B (2nd heater drain pump)

Power escalation to 100% in progress following maintenance to the A MFW pump.

Idle condensate pump handswitch is in P-T-L.

Turnover:

70% power holding for the start of P8B (2nd heater drain pump). Heater Drain Pump, P8A has just been placed in service per 1106.016, section 16.0 through 16.19. Step 16.20 is ready to be performed and P8B placed in service.

Power escalation to 100% in progress following maintenance to the A MFW pump.

CV1207, Seal Injection control valve, is in manual due to oscillations of seal injection when in auto.

Event Malf. No. Event Event No. Type* Description 1 N/A N (CBOT) Start a heater drain pump during power escalation 2 N/A R (CBOR) Power escalation toward 100% following maintenance to main Feedwater pump and after Heater Drain Pumps in service.

3 TR565 520 I (CBOR) RCS Thot slowly fails low causing changing input Ramp=4 Min. signal to ICS and requires operator intervention to stop transient.

4 N/A N (CBOR) Place the ICS in AUTO.

5 ED191 C (CBOT) Loss of non vital bus B3 requiring crew to restart redundant equipment to support plant operation.

6 ED183 M (All) Random electrical grid upsets result in loss of offsite power. Reactor trip. Degraded Power.

7 EG176 C (CBOT) EDG #2 fails to autostart on command. Manual attempt to start at panel C10 fails.

8 FW611 C (CBOR) EFIC Loss of Fill Rate Control after EFW actuation.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 3

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

1) Evaluate individual response to input signal failures to the Integrated Control System

2) Evaluate individual response to electrical system abnormal conditions

3) Evaluate individual response to a loss of offsite electrical power

4) Evaluate individual ability to start and control components of the feed and condensate system

5) Evaluate individual ability to perform a power escalation in accordance with plant procedures

6) Evaluate individual ability to recognize abnormal conditions associated with automatically actuated systems and components

7) Evaluate individual response to loss of automatic control of EFW to control OTSG levels SCENARIO #1 NARRATIVE SCENARIO #1 NARRATIVE (continued)

Page 2 of 3

Appendix D Scenario 1 Form ES-D-1 Simulator Instructions for Scenario 1 Event Time Malf. No. Value/ Ramp Event No. Time Description Page 3 of 3

Appendix D Scenario 2 Form ES-D-1 Facility: ANO-1 Scenario No.: 2 Op-Test No.:2005-1 Page 1 Examiners: Operators:

Initial Conditions:

100% power; Equilibrium xenon Power has been stable at 100% for last two days following return to full power after maintenance to the B main feedwater pump.

Chemistry is performing routine Tech Spec chemistry and sampling is aligned from pressurizer water space per 1104.002, Section 19.2.

Turnover:

100% power; Equilibrium xenon Power has been stable at 100% for last two days following return to full power after maintenance to the B main feedwater pump.

Chemistry is performing routine Tech Spec chemistry and sampling is aligned from pressurizer water space per 1104.002, Section 19.2.

CV1207, Seal Injection control valve, is in manual due to oscillations of seal injection when in auto.

Event Malf. No. Event Event No. Type* Description 1 N/A N (CBOR) Chemistry calls and reports boron samples with a 55 PPM difference. Equalize RCS/Pzr boron concentration 2 FW087 C (CBOT) Heater drain pump B bearing heat up resulting in need to trip the pump. Pump will trip if no action is taken. (2 min.15 sec.- alarm, 3 min. 15 sec. -trip) 3 N/A R (CBOR) Lower power to ~70% in response to a trip of a heater drain pump 4 DI_ICC0009L C (CBOR) ULD station fails to lower demand signal in manual False requiring operator action to manually reduce power.

5 TR049 0 I (CBOR) Controlling Pressurizer level transmitter fails low 6 RC001 M (ALL) OTSG tube leak progressing to a ~150 gpm.

(.325)

Ramp 20 min.

7 DI_H15C C (CBOT) H1 feeder breaker from SU#1 transformer fails to False close when transferring auxiliaries.

8 IRF CO_P75 C (CBOT) P75 will not start causing crew to take contingency actions in the EOP Page 1 of 4

Appendix D Scenario 2 Form ES-D-1

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 2 of 4

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

1) Evaluate individual response to a feedwater and condensate system component failure.

2) Evaluate individual response to an Integrated Control System failure.

3) Evaluate individual response to a transmitter signal input failure to a controlling function.

4) Evaluate individual response to an electrical breaker failure

5) Evaluate individual response to an OTSG tube leak/rupture.

6) Evaluate individual ability to lower plant load in accordance with plant procedures.

7) Evaluate individual ability to operate controls to equalize boron concentrations between Reactor Coolant System and Pressurizer.

SCENARIO #1 NARRATIVE SCENARIO #1 NARRATIVE (continued)

Page 3 of 4

Appendix D Scenario 2 Form ES-D-1 Simulator Instructions for Scenario 1 Event Time Malf. No. Value/ Ramp Event No. Time Description Page 4 of 4

Appendix D Scenario 1 Form ES-D-1 Facility: ANO-1 Scenario No.: 3 Op-Test No.:2005-1 Page 1 Examiners: Operators:

Initial Conditions:

ICS in Automatic at ~25% power following a mid-cycle shutdown.

Power escalation is in progress per the Startup and Power Operations procedures.

A MFW pump is in service and B MFW pump is running at minimum speed.

Turnover:

ICS in Automatic at ~25% power following a mid-cycle shutdown.

Power escalation is in progress per the Startup and Power Operations procedures.

A MFW pump is in service and B MFW pump is running at minimum speed.

CV1207, Seal Injection control valve, is in manual due to oscillations of seal injection when in auto.

Event Malf. No. Event Event No. Type* Description 1 N/A R (CBOR) Power escalation from ~25% power to 35% power to place B MFP in service 2 AI_TIC4018S C (CBOT) Automatic control of CV4018 fails to maintain

(.7) generator temperatures 3 N/A N (CBOR) Place the second main feedwater pump in service 4 TR575 0 R10 I (CBOR) B OTSG startup level transmitter fails low causing a Feedwater transient requiring operator intervention.

5 RC464 2.5 M (ALL) LOCA- Leak on an HPI line inside containment Ramp 5 min.

6 RP245 C (CBOR) RPS fails to trip the reactor automatically on a valid RP246 RPS trip setpoint. The Manualtrip button on C03 fails to perform a reactor trip. (The backup RP247 pushbuttons on C03 must be depressed to RP248 complete a reactor trip)

DI_ICC0020 (False) 7 CV1407 C (CBOT) BWST outlet valve CV1407 fails to open on ESAS signal. Valve must be manually opened in the field.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 3

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

1) Evaluate individual response to component failures affecting cooling of the main turbine generator.

2) Evaluate individual response to input failures to the Integrated Control System.

3) Evaluate individual response to a loss of reactor coolant accident

4) Evaluate individual response to failure of automatic actuation systems.

5) Evaluate individual response to failure of Emergency Core Cooling System components.

6) Evaluate individual ability to maneuver the plant in accordance with plant procedures.

7) Evaluate individual ability to start and operate feedwater and condensate system components in accordance with plant procedures.

SCENARIO #1 NARRATIVE SCENARIO #1 NARRATIVE (continued)

Page 2 of 3

Appendix D Scenario 1 Form ES-D-1 Simulator Instructions for Scenario 1 Event Time Malf. No. Value/ Ramp Event No. Time Description Page 3 of 3

Appendix D Scenario 4 Form ES-D-1 Facility: ANO-1 Scenario No.: 4 (Spare) Op-Test No.:2005-1 Page 1 Examiners: Operators:

Initial Conditions: IC 2 100% Power, Equilibrium Xenon, Turnover:

100% Power, Equilibrium Xenon, MOL (250 EFPD)

An RCS delithiation is anticipated for this shift. Chemistry will call to provide the duration of the evolution.

CV1207, Seal Injection control valve, is in manual due to oscillations of seal injection when in auto.

Event Malf. No. Event Event No. Type* Description 1 BAT CRD.txt I(CBOR) CRD position indication faulty with a CRD W/D (Batch file that creates inhibit this condition) 2 N/A N (CBOR) Perform 5 minute RCS delithiation at the request of chemistry. (Idle purification DI is ~65 PPM above RCS boron concentration) 3 BAT ES19_2.txt C (CBOT) Condenser vacuum leak caused by the failure of (Batch file that creates the #5 turbine bearing gland seal regulator.

this condition) 4 N/A R (CBOR) Power reduction to stabilize vacuum 5 CV098 C (CBOT) Operating MU/HPI pump experiences high winding temperature and causes the pump to trip. The standby pump has no oil indicated in one of the oil bubblers and must have oil added prior to start.

6 MS143 M (ALL) Main steam safety valve associated with B OTSG MS134 .4 experiences structural failure and lifts. Steam Leak significant to warrant manual reactor trip.

7 TU155 C (CBOT) The main turbine fails to fully trip when the unit is TU156 tripped. One governor valve and one throttle valve fail to close, requiring the crew to shut the MSIV from the affected OTSG.

8 CV2648 C (CBOR) EFW valves to B OTSG from P7B fail open and CV2626 will not close from the handswitches.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 3

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

1) Evaluate individual ability to perform an RCS delithiation in accordance with plant procedures.

2) Evaluate individual ability to perform a plant power reduction and stabilize the plant in accordance with plant procedures.

3) Evaluate individual response to faulty control rod position indication.

4) Evaluate individual response to a loss of condenser vacuum.

5) Evaluate individual response to loss of RCS makeup due to an HPI pump trip.

6) Evaluate individual response to a main steam line break/overcooling event.

7) Evaluate individual response to failure of the main turbine steam valves to close on a turbine trip.

8) Evaluate individual response to a failure of Emergency Feedwater System components.

SCENARIO #1 NARRATIVE SCENARIO #1 NARRATIVE (continued)

Page 2 of 3

Appendix D Scenario 4 Form ES-D-1 Simulator Instructions for Scenario 1 Event Time Malf. No. Value/ Ramp Event No. Time Description Page 3 of 3