Text

09 - Final Outlines
	ML082750588
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/05/2008
From:	Brian Larson; NRC Region 4
To:	;
References
	Download: ML082750588 (30)
	v • d • e

ES-401 PWR Examination Outline Revision 1 Form ES-401-2 Facility: Arkansas Nuclear One - Unit 1 Date of Exam: 9/5/2008 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

Total

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

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

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 2 3 2 1 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only 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 1 Form ES-401-2

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

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

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) X EA2.03 - Reactor trip breaker position 4.2 1 Reactor Trip - Stabilization -

Recovery / 1 AA1.07 - Reseating of code safety and PORV.

000008 Pressurizer Vapor Space X 4.0 2 Accident / 3 000009 Small Break LOCA / 3 X EK2.03 - S/Gs 3.0 3 000011 Large Break LOCA / 3 X EA2.11 - Conditions for throttling or stopping HPI. 3.9 4 000015/17 RCP Malfunctions / 4 X AK1.04 - Basic steady state thermodynamic 2.9 5 relationship between RCS loops and S/Gs resulting from unbalanced RCS flow.

000022 Loss of Rx Coolant Makeup / X AK3.01 - Adjustment of RCP seal backpressure 2.7 6 2 regulator valve to obtain normal flow.

AK1.01 - Loss of RHRS during all modes of 000025 Loss of RHR System / 4 X operation. 3.9 7 000026 Loss of Component Cooling Not selected n/a n/a Water / 8 AA2.07 - Makeup flow indication.

000027 Pressurizer Pressure Control X System Malfunction / 3 Changed KA to AA2.12 - Pressurizer Level 3.7 8 000029 ATWS / 1 X EA1.11 - Manual Opening of the CRD Breakers 3.9 9 000038 Steam Gen. Tube Rupture / X EK3.08 - Criteria for securing an RCP 4.1 10 3

000040 (BW/E05; CE/E05; W/E12) X EK2.1 - Components, and functions of control and 3.8 11 Steam Line Rupture - Excessive safety systems, including instrumentation, signals, Heat Transfer / 4 interlocks, failure modes, and automatic and manual features.

2.2.40 - Ability to apply Technical Specifications for 000054 (CE/E06) Loss of Main X a system 3.4 12 Feedwater / 4 2.4.45 - Ability to prioritize and interpret the 000055 Station Blackout / 6 X significance of each annunciator or alarm. 4.1 13 AK1.04 - Definition of saturation conditions, 000056 Loss of Off-site Power / 6 X implication for the systems 3.1 14 2.1.28 - Knowledge of the purpose and function of 000057 Loss of Vital AC Inst. Bus / 6 X major system components and controls. 4.1 15 AK3.01 - Use of dc control power by ED/Gs 000058 Loss of DC Power / 6 X 3.4 16 000062 Loss of Nuclear Svc Water / Not selected n/a n/a 4

AA1.02 - Components served by instrument air to 000065 Loss of Instrument Air / 8 X minimize drain on system 2.6 17 W/E04 LOCA Outside Containment / Suppressed n/a n/a 3

ES-401 2 Form ES-401-2

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

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

1 2 3 1 2 W/E11 Loss of Emergency Coolant Suppressed n/a n/a Recirc. / 4 EK2.02 - Facility's heat removal systems, including BW/E04; W/E05 Inadequate Heat X primary coolant, emergency coolant, the decay heat 4.2 18 Transfer - Loss of Secondary Heat Sink / 4 removal systems, and relations between the proper operation of these systems to the operation of the facility 000077 Generator Voltage and Electric Not selected n/a n/a Grid Disturbances / 6 K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18 ES-401 3 Form ES-401-2

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

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

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 Not selected n/a n/a 000003 Dropped Control Rod / 1 Not selected n/a n/a AK2.01 - Controllers and positioners 2.5 19 000005 Inoperable/Stuck Control Rod / 1 X 000024 Emergency Boration / 1 Not selected n/a n/a 000028 Pressurizer Level Malfunction / 2 Not selected n/a n/a 000032 Loss of Source Range NI / 7 Not selected n/a n/a 000033 Loss of Intermediate Range NI / 7 Not selected n/a n/a 2.1.28 - Knowledge of the purpose and function n/a n/a 000036 (BW/A08) Fuel Handling Accident / 8 of major system components and controls.

Changed to System 076 2.1.31 AA2.15 - Magnitude of atmospheric radioactive 000037 Steam Generator Tube Leak / 3 X release if cool-down must be completed using steam dump or atmospheric reliefs Changed to AA2.01 Unusual readings of the monitors; steps needed 3.0 20 to verify readings.

000051 Loss of Condenser Vacuum / 4 Not selected n/a n/a AK1.05 - The calculation of offsite doses due to n/a n/a 000059 Accidental Liquid RadWaste Rel. / 9 a release from the power plant Changed to System 067 AK1.02 000060 Accidental Gaseous Radwaste Rel. / 9 Not selected n/a n/a 000061 ARM System Alarms / 7 Not selected n/a n/a 000067 Plant Fire On-site / 8 X AK1.02 - Fire Fighting 3.1 21 AA1.13 - Charging Pump controllers (to 4.1 22 000068 (BW/A06) Control Room Evac. / 8 X maintain Pressurizer level)

Not selected 000069 (W/E14) Loss of CTMT Integrity / 5 n/a n/a Not selected 000074 (W/E06&E07) Inad. Core Cooling / 4 n/a n/a 000076 High Reactor Coolant Activity / 9 X 2.1.31 - Ability to locate control switches, 4.6 23 controls, and indications, and to determine they correctly reflect the desired plant lineup.

Suppressed W/EO1 & E02 Rediagnosis & SI Termination / 3 n/a n/a Suppressed W/E13 Steam Generator Over-pressure / 4 n/a n/a Suppressed W/E15 Containment Flooding / 5 n/a n/a Suppressed W/E16 High Containment Radiation / 9 n/a n/a BW/A01 Plant Runback / 1 Not selected n/a n/a BW/A02&A03 Loss of NNI-X/Y / 7 Not selected n/a n/a AK3.2 - Normal, abnormal and emergency 3.4 24 BW/A04 Turbine Trip / 4 X operating procedures associated with (Turbine Trip)

BW/A05 Emergency Diesel Actuation / 6 Not selected n/a n/a AA2.2 - Adherence to appropriate procedures 3.3 25 BW/A07 Flooding / 8 X and operation within the limitations in the facility's license and amendments BW/E03 Inadequate Subcooling Margin / 4 Not selected n/a n/a BW/E08; W/E03 LOCA Cooldown - Depress. / 4 Not selected n/a n/a EK2.1 - Components, and functions of control 3.7 26 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 X and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features ES-401 4 Form ES-401-2

ES-401 PWR Examination Outline Revision 1 Form ES-401-2 EK3.2- Normal, Abnormal and Emergency 3.2 27 BW/E13&E14 EOP Rules and Enclosures X Operating Procedures associated with (EOP Rules and EOP Enclosures)

Suppressed CE/A11; W/E08 RCS Overcooling - PTS / 4 n/a n/a Suppressed CE/A16 Excess RCS Leakage / 2 n/a n/a Suppressed CE/E09 Functional Recovery n/a n/a K/A Category Point Totals: 1 2 2 1 2 1 Group Point Total: 9 ES-401 5 Form ES-401-2

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

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 K5.03 - Effects of RCP shutdown on T-003 Reactor Coolant Pump X X ave., including the reason for the unreliability of T-ave. in the shutdown loop 28 Changed to K5.02 - Effects of RCP 2.8 coastdown on RCS parameters.

A3.01 - Seal Injection flow 3.3 29 K6.09 - Purpose of VCT divert valve 004 Chemical and Volume X 2.8 30 Control 2.2.38 - Knowledge of conditions and 005 Residual Heat Removal X X limitations in the facility license. 3.6 31 K3.01 - RCS 3.9 32 K4.08 - Recirculation flowpath of reactor 006 Emergency Core Cooling X building sump 3.4 33 K1.01 - Containment System 007 Pressurizer Relief/Quench X 2.9 34 Tank K1.02 - Loads cooled by CCWS 008 Component Cooling Water X 3.3 35 K4.02 - Prevention of uncovering PZR 010 Pressurizer Pressure Control X heaters 3.0 36 K6.04 - Bypass block circuits 012 Reactor Protection X 3.3 37 K5.02 - Safety system logic and reliability 013 Engineered Safety Features X 2.9 38 Actuation K2.02 - Chillers 022 Containment Cooling X 2.5 39 025 Ice Condenser Suppressed n/a n/a K2.02 - MOVs 026 Containment Spray X X 2.7 40 A4.01 - CSS control 4.5 41 K1.06 - Condenser steam dump 039 Main and Reheat Steam X X 3.1 42 A4.07 - Steam dump valves 2.8 43 A1.07 - Feed Pump speed, including normal 059 Main Feedwater X control speed for ICS 2.5 44 K3.02 - S/G 061 Auxiliary/Emergency X X 4.2 45 Feedwater A3.02 - RCS Cooldown during AFW operations. 4.0 46 A2.16 - Degraded system voltages 062 AC Electrical Distribution X 2.5 47 ES-401 6 Form ES-401-2

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

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 A2.01 - Grounds 063 DC Electrical Distribution X 2.5 48 X 2.2.42 - Ability to recognize system 3.4 49 parameters that are entry level condition for Technical Specificationn K6.08 - Fuel Oil Storage Tanks 064 Emergency Diesel Generator X 3.2 50 A1.01 - Radiation levels 073 Process Radiation X 3.2 51 Monitoring K4.01 - Conditions initiating automatic 076 Service Water X closure of closed cooling water auxiliary 2.5 52 building header supply and return valves

K3.01 - Containment Air System 078 Instrument Air X X Changed KA to K3.02 - Pneumatic Valves and Controls 3.4 53 2.1.32 - Ability to explain and apply system 3.8 54 limits and precautions.

A1.01 - Containment pressure, temperature, 103 Containment X and humidity 3.7 55 K/A Category Point Totals: 3 2 3 3 2 3 3 2 2 2 3 Group Point Total: 28 ES-401 7 Form ES-401-2

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

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

1 2 3 4 5 6 1 2 3 4 K4.01 - Rod Position Indication 3.5 56 001 Control Rod Drive X 002 Reactor Coolant Not selected n/a n/a A3.01 - Boration/dilution 2.8 57 011 Pressurizer Level Control X

014 Rod Position Indication Not selected n/a n/a 015 Nuclear Instrumentation X K5.10 - Excore Detector Operation 2.8 58 016 Non-nuclear Instrumentation Not selected n/a n/a 2.1.27 - Knowledge of system purpose and / or 3.9 59 017 In-core Temperature Monitor X function.

Not selected n/a n/a 027 Containment Iodine Removal K2.01 - Hydrogen Recombiners 2.5 60 028 Hydrogen Recombiner X and Purge Control 029 Containment Purge Not selected n/a n/a 033 Spent Fuel Pool Cooling Not selected n/a n/a K1.05 - Shutdown monitor 2.5 61 034 Fuel Handling Equipment X

K6.01 - MSIVs 3.2 62 035 Steam Generator X 041 Steam Dump/Turbine Not selected n/a n/a Bypass Control 045 Main Turbine Generator X K3.01 - Remander of the plant 2.9 63 055 Condenser Air Removal Not selected n/a n/a 056 Condensate Not selected n/a n/a Not selected n/a n/a 068 Liquid Radwaste A4.26 - Authorized waste gas release, 3.1 64 071 Waste Gas Disposal X conducted in compliance with radioactive gas discharge permit A1.01 - Radiation levels 3.4 65 072 Area Radiation Monitoring X 075 Circulating Water Not selected n/a n/a 079 Station Air Not selected n/a n/a 086 Fire Protection Not selected n/a n/a K/A Category Point Totals: 1 1 1 1 1 1 1 0 1 1 1 Group Point Total: 10 ES-401 8 Form ES-401-2

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Arkansas Nuclear One - Unit 1 Date of Exam: 9/5/2008 Category K/A # Topic RO SRO-Only IR # IR #

2.1.21 Ability to obtain and verity controlled procedure copy. 3.5 66 2.1.29 Knowledge of how to conduct and verify valve lineups. 4.1 67 1.

Conduct 2.1.

of Operations 2.1.

2.1.

Subtotal 2 2.2.15 Ability to determine the expected plant configuration using 3.9 68 design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

2.

2.2.13 Knowledge of tagging and clearance procedures 4.1 69 Equipment Control 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Changed KA to 2.9 70 Knowledge of pre- and post-maintenance operability 2.2.21 requirements.

2.2.

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

2.3.13 Knowledge of radiological safety procedures pertaining to 3.4 72 3.

licensed operator duties, such as response to radiation monitor Radiation alarms, containment entry requirements, fuel handling Control responsibilities, access to locked high-radiation areas, aligning filters, etc.

2.3.

Subtotal 2 2.4.11 Knowledge of abnormal condition procedures. 4.0 73

4. 2.4.21 Knowledge of the parameters and logic used to assess the status 4.0 74 Emergency of safety functions, such as reactivity control, core cooling and Procedures / heat removal, reactor coolant system integrity, containment Plan conditions, radioactivity release control, etc.

2.4.49 Ability to perform without reference to procedures those 4.6 75 actions that require immediate operation of system components and controls.

2.4.

Subtotal 3 Tier 3 Point Total 10

ES-401 Record of RO Rejected K/As Revision 1 Form ES-401-4 Tier / Randomly Selected K/A Reason for Rejection Group 1/1 027 Pressurizer Pressure No credible tie for this K/A exists for the System.

Control System Malfunction Replaced with same system - AA2.12 Pressurizer Level.

- AA2.07 Makeup flow indication.

1/2 036 Fuel Handling Accident Could not write a credible question since only RO action is

- 2.1.28 Knowledge of the to suspend fuel movement and exit.

purpose and function of major system components Randomly selected new system 076 High Reactor Coolant and controls. Activity - 2.1.31 Ability to locate control switches, controls, and indications, and to determine they correctly reflect the desired plant lineup.

1/2 037 Steam Generator Tube No credible CFR 41 RO tie exists for this K/A. Replaced Leak - AA2.15 Magnitude of with same system - AA2.01 Unusual readings of the atmospheric radioactive monitors; steps needed to verify readings.

release if cool-down must be completed using steam dump or atmospheric reliefs.

1/2 059 Accidental Liquid Not possible to prepare a psychometrically sound question Radwaste Release - AK1.05 related to the subject K/A.

The calculation of offsite does due to a release from Randomly selected new system 067 Plant Fire On-site -

the power plant. KA AK1.02 Fire Fighting.

2/1 003 Reactor Coolant Pump - Posed a double jeopardy with Tier 1 Group 1 015/17 RCP K5.03 Effects of RCP Malfunctions - AK1.04 Basic steady state thermodynamic shutdown on T-ave, relationship between RCS loops and S/Gs resulting from including the reason for the unbalanced RCS flow.

unreliability of T-ave, in the shutdown loop. Randomly selected same system - KA 5.02 Effects of RCP coastdown on RCS parameters.

2/1 078 Instrument Air System Rejected due to Low Operational value for discriminatory (IAS) - K3.01 Containment SRO/RO level question. Containment Air system is used Air System. for Breathing Air or Tool Service Air and serves no other functions.

Randomly selected same system - K3.02 Pneumatic Valves and Controls.

3 Equipment Control - Generic RO would not perform this function. At ANO normally 2.2.36 Ability to analyze the performed by an SRO.

effect of maintenance Randomly selected same category - 2.2.21 Knowledge of activities, such as degraded pre- and post- maintenance operability requirements.

power sources, on the status of limiting conditions for operations.

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

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

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor X 2.4.2- Knowledge of system setpoints, interlocks 4.6 76 Trip - Stabilization - Recovery / 1 and automatic actions associated with EOP entry conditions.

000008 Pressurizer Vapor Space Not selected n/a n/a Accident / 3 000009 Small Break LOCA / 3 Not selected n/a n/a 000011 Large Break LOCA / 3 Not selected n/a n/a 000015/17 RCP Malfunctions / 4 Not selected n/a n/a 000022 Loss of Rx Coolant Makeup / 2 Not selected n/a n/a 000025 Loss of RHR System / 4 Not selected n/a n/a 000026 Loss of Component Cooling Not selected n/a n/a Water / 8 000027 Pressurizer Pressure Control Not selected n/a n/a System Malfunction / 3 000029 ATWS / 1 X 2.2.39 Knowledge of less than or equal to one hour 4.5 77 Technical Specification action statements for systems.

000038 Steam Gen. Tube Rupture / 3 Not selected n/a n/a 000040 (BW/E05; CE/E05; W/E12) X EA2.2 - Adherence to appropriate procedures and 4.0 78 Steam Line Rupture - Excessive Heat operation within the limitations in the facilitys Transfer / 4 license and amendments.

000054 (CE/E06) Loss of Main Not selected n/a n/a Feedwater / 4 000055 Station Blackout / 6 Not selected n/a n/a 000056 Loss of Off-site Power / 6 Not selected n/a n/a 000057 Loss of Vital AC Inst. Bus / 6 X AA2.16 - Normal and Abnormal Pressurizer Level 3.1 79 for various modes of plant operation.

000058 Loss of DC Power / 6 Not selected n/a n/a 000062 Loss of Nuclear Svc Water / 4 Not selected n/a n/a 000065 Loss of Instrument Air / 8 Not selected n/a n/a W/E04 LOCA Outside Containment / 3 Suppressed n/a n/a W/E11 Loss of Emergency Coolant Suppressed n/a n/a Recirc. / 4 2.4.46 - Ability to verify that the alarms are BW/E04; W/E05 Inadequate Heat X consistent with the plant conditions. 4.2 80 Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric X AA2.03 - Generator current outside the capability 3.6 81 Grid Disturbances / 6 curve.

K/A Category Totals: 3 3 Group Point Total: 6 ES-401 1 Form ES-401-2

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

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

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 Not selected n/a n/a 000003 Dropped Control Rod / 1 Not selected n/a n/a 000005 Inoperable/Stuck Control Rod / 1 Not selected n/a n/a 000024 Emergency Boration / 1 Not selected n/a n/a 000028 Pressurizer Level Malfunction / 2 Not selected n/a n/a 000032 Loss of Source Range NI / 7 Not selected n/a n/a 000033 Loss of Intermediate Range NI / 7 Not selected n/a n/a 000036 (BW/A08) Fuel Handling Accident / 8 Not selected n/a n/a 000037 Steam Generator Tube Leak / 3 Not selected n/a n/a 000051 Loss of Condenser Vacuum / 4 X 2.4.21 - Knowledge of the parameters and logic 4.6 82 used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

000059 Accidental Liquid RadWaste Rel. / 9 Not selected n/a n/a 000060 Accidental Gaseous Radwaste Rel. / 9 Not selected n/a n/a 000061 ARM System Alarms / 7 Not selected n/a n/a 000067 Plant Fire On-site / 8 Not selected n/a n/a 000068 (BW/A06) Control Room Evac. / 8 Not selected n/a n/a Not selected 000069 (W/E14) Loss of CTMT Integrity / 5 n/a n/a Not selected 000074 (W/E06&E07) Inad. Core Cooling / 4 n/a n/a 000076 High Reactor Coolant Activity / 9 X AA2.02 - Corrective Actions required for High 3.4 83 Fission Product Activity in RCS.

Suppressed W/EO1 & E02 Rediagnosis & SI Termination / 3 n/a n/a Suppressed W/E13 Steam Generator Over-pressure / 4 n/a n/a Suppressed W/E15 Containment Flooding / 5 n/a n/a Suppressed W/E16 High Containment Radiation / 9 n/a n/a BW/A01 Plant Runback / 1 Not selected n/a n/a BW/A02&A03 Loss of NNI-X/Y / 7 Not selected n/a n/a BW/A04 Turbine Trip / 4 Not selected n/a n/a BW/A05 Emergency Diesel Actuation / 6 Not selected n/a n/a BW/A07 Flooding / 8 Not selected n/a n/a BW/E03 Inadequate Subcooling Margin / 4 X EA2.1 - Facility conditions and selection of 4.0 84 appropriate procedures during abnormal and emergency operations.

BW/E08; W/E03 LOCA Cooldown - Depress. / 4 Not selected n/a n/a 2.1.7 - Ability to evaluate plant performance and 4.7 85 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 X make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

BW/E13&E14 EOP Rules and Enclosures Not selected n/a n/a Suppressed CE/A11; W/E08 RCS Overcooling - PTS / 4 n/a n/a Suppressed CE/A16 Excess RCS Leakage / 2 n/a n/a Suppressed CE/E09 Functional Recovery n/a n/a K/A Category Point Totals: 0 0 0 0 2 2 Group Point Total: 4 ES-401 2 Form ES-401-2

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

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 A2.03 - Problems associated with RCP 003 Reactor Coolant Pump X motors, including faulty motors and 3.1 86 current, and winding and bearing temperature problems.

2.4.1 - Knowledge of EOP entry conditions 004 Chemical and Volume Control and immediate action steps.

Changed to System 006 A2.03 n/a n/a 005 Residual Heat Removal Not selected n/a n/a 006 Emergency Core Cooling X A2.03 - System Leakage 3.7 87 007 Pressurizer Relief/Quench Tank Not selected n/a n/a 008 Component Cooling Water Not selected n/a n/a 010 Pressurizer Pressure Control Not selected n/a n/a 012 Reactor Protection Not selected n/a n/a 2.4.1 - Knowledge of EOP entry conditions 013 Engineered Safety Features X and immediate action steps. 4.8 88 Actuation 022 Containment Cooling Not selected n/a n/a 025 Ice Condenser Suppressed n/a n/a 026 Containment Spray X 2.4.21- Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Changed to 2.4.4 Ability to recognize abnormal indications for 4.7 89 system operating parameters that are entry level conditions for emergency and abnormal operating procedures.

039 Main and Reheat Steam Not selected n/a n/a 059 Main Feedwater Not selected n/a n/a 061 Auxiliary/Emergency Feedwater X A2.02 - Loss of air to steam supply valve Changed to A2.03 3.4 90 Loss of DC Power 062 AC Electrical Distribution Not selected n/a n/a 063 DC Electrical Distribution Not selected n/a n/a 064 Emergency Diesel Generator Not selected n/a n/a 073 Process Radiation Monitoring Not selected n/a n/a 076 Service Water Not selected n/a n/a 078 Instrument Air Not selected n/a n/a 103 Containment Not selected n/a n/a K/A Category Point Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5 ES-401 3 Form ES-401-2

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

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

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive Not selected n/a n/a 002 Reactor Coolant Not selected n/a n/a A2.12 - Operation of Auxiliary Spray 3.3 91 011 Pressurizer Level Control X 014 Rod Position Indication Not selected n/a n/a 015 Nuclear Instrumentation X 2.2.39 - Knowledge of less than or equal to one 4.5 92 hour0.00106 days 0.0256 hours 1.521164e-4 weeks 3.5006e-5 months Technical Specification action statements for systems.

016 Non-nuclear Instrumentation X A2.01 - Detector Failure. 3.1 93 017 In-core Temperature Monitor Not selected n/a n/a 027 Containment Iodine Removal Not selected n/a n/a 028 Hydrogen Recombiner Not selected n/a n/a and Purge Control 029 Containment Purge Not selected n/a n/a 033 Spent Fuel Pool Cooling Not selected n/a n/a 034 Fuel Handling Equipment Not selected n/a n/a 035 Steam Generator Not selected n/a n/a 041 Steam Dump/Turbine Not selected n/a n/a Bypass Control 045 Main Turbine Generator Not selected n/a n/a 055 Condenser Air Removal Not selected n/a n/a 056 Condensate Not selected n/a n/a 068 Liquid Radwaste Not selected n/a n/a 071 Waste Gas Disposal Not selected n/a n/a 072 Area Radiation Monitoring Not selected n/a n/a 075 Circulating Water Not selected n/a n/a 079 Station Air Not selected n/a n/a 086 Fire Protection Not selected n/a n/a K/A Category Point Totals: 0 0 0 0 0 0 0 2 0 0 1 Group Point Total: 3 ES-401 4 Form ES-401-2

ES-401 Generic Knowledge and Abilities Outline (Tier 3) (SRO) Revision 1 Form ES-401-3 Facility: ANO Unit 1 Date of Exam: 9/5/2008 Category K/A # Topic RO SRO-Only IR # IR #

2.1.5 Ability to use procedures related to shift staffing, such as 3.9 94 minimum crew complement, overtime limitations, etc.

1.

2.1.35 Knowledge of the fuel-handling responsibilities of SROs. 3.9 95 Conduct of Operations Subtotal 2 2.2.19 Knowledge of maintenance work order requirements. 3.4 96 2.

Equipment Control Subtotal 1 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personal monitoring equipment, etc.

3.

Changed KA to:

Radiation 2.3.4 3.7 97 Control Knowledge of radiation exposure limits under normal or emergency conditions.

2.3.11 Ability to control radiation releases 4.3 98 Subtotal 2 2.4.8 Knowledge of how abnormal operating procedures are used in 4.5 99 conjunction with EOPs.

4.

2.4.35 Knowledge of local auxiliary operator tasks during an 4.0 100 Emergency emergency and the resultant operational effects.

Procedures /

Plan Subtotal 2 Tier 3 Point Total 7 ES-401 1 Form ES-401-3

ES-401 Record of SRO Rejected K/As Revision 1 Form ES-401-4 Tier / Randomly Selected K/A Reason for Rejection Group 2/1 004 Chemical and Volume Control - 2.4.1 Two of the five questions in tier 2 group 1 were Knowledge of EOP entry conditions and KA 2.4.1. Felt the group was oversampled.

immediate action steps.

Randomly selected new system 006 Emergency Core Cooling - A2.03 System Leakage.

2/1 026 Containment Spray -2.4.21 Due to the simplicity of the system, no credible Knowledge of the parameters and logic distracters could be written for this system.

used to assess the status of safety Randomly selected same system - 2.4.4 Ability functions, such as reactivity control, core to recognize abnormal indications for system cooling and heat removal, reactor coolant operating parameters that are entry level system integrity, containment conditions, conditions for emergency and abnormal radioactivity release control, etc.

operating procedures.

2/1 061 Auxiliary / Emergency Feedwater - KA does not apply to ANO Unit 1. Additionally A2.02 Loss of air to steam supply valve. suppressed A2.02, 2.06, 2.07, 2.08 as they do not apply.

Randomly selected same system - A2.03 Loss of DC Power.

3 3 Radiation Control - 2.3.5 Ability to use No credible CFR 43 SRO tie.

radiation monitoring systems, such as fixed radiation monitors and alarms, Randomly selected - 2.3.4 Knowledge of radiation exposure limits under normal or portable survey instruments, personnel emergency conditions.

monitoring equipment, etc.

ES-401 1 Form ES-401-4

Appendix D Scenario Outline Form ES-D-1 Facility: ANO-1 Scenario No.: 1-R5 Op-Test No.: 2008-1 Examiners: __________________________ Operators: __________________________

Initial Conditions:

90% power due to dispatcher direction for loss of 500KV.

C2A IA compressor is out of service for overhaul.

ULD is failed and will not lower power.

RPS is failed and will not initiate a RX trip.

Provide stop watch for surveillance.

Two rods will fail to insert on the RX trip.

Turnover:

Day shift normal working day.

90% power due to dispatcher direction for loss of 500KV line as a result of storm damage to the Mabelvale 500KV line. No storms are currently in the area.

C2A IA compressor is out of service for overhaul.

1104.005 Supplement 2 RB Spray Red Train Valves Quarterly Test is in progress complete through step 2.2.2. This is not the 18 month surveillance.

Crew will continue in surveillance after turnover.

Event Malf. No. Event Event No. Type* Description N (SRO, Perform 1104.005 Supplement 2 RB Spray Red Train 1 N/A BOP) Valves Quarterly Test N (SRO) Dispatcher directs power reduction to 700Mwe 2 DI_ICC0009L R (ATC) ULD is failed I (SRO, 3 AI_FIC1207 RCP total flow setpoint fails to high ATC)

C (SRO, 4 CV016 RCP seal failure BOP)

N (SRO) Power reduction 5 DI_ICC0009L R (ATC) ULD is failed DI_H24T M (ALL) Loss of H2 bus 6 If power <55%,

CO_P32A N/A Trip A RCP when H2 bus is lost (Contingency to ensure RPS trip setpoint is reached.)

C (SRO, RPS is failed 7 RP246,7,8,9 ATC) Manual RX trip (TS) (ATC-CT)

CONTINUED Page 1 of 19

Event Event Event Malf. No.

No. Type* Description RD362 C (SRO, Stuck rod (TS) 8 RD363 ATC) Stuck rod (ATC-CT)

CV020 B RCP seal failure RC006 M (ALL) ~700gpm RCS leak (TS) (BOP-CT) 9 ES259 C (BOP) ESAS Channel 1 fails to actuate (TS) (ATC-CT)

CV-1300 CV1300 Fails open

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 2 of 19

Appendix D Scenario Outline Form ES-D-1 SCENARIO #1-R5 NARRATIVE The crew will assume the watch at 90% power. The Mabelvale 500KV line is out of service due to storm related damage in south east Arkansas. The down power was completed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ago.

Reactor engineering has directed that rods can be used to account for the xenon transient.

1104.005 Supplement 2 RB Spray Red Train Valves Quarterly Test is complete through step 2.2.2. The crew should brief the remaining steps of the surveillance. The BOP will perform the stroke test (BOP-N) (SRO-N).

The dispatcher will call and direct U1 power be reduced to ~700Mwe in the next 10 minutes.

The CRS should calculate the required rate and direct power reduction per 1203.045 Rapid Plant Shutdown. The ATC should recognize that the ULD is not responding and power is not lowering. The ATC should reduce power to ~700 Mwe using the SG/RX Master station in hand.

(ATC-R) (SRO-N)

The controlling RCP total seal injection flow setpoint will fail high. (ATC-I) (SRO-I). This will cause CV-1207 RCP Seal Injection Control Valve to fail open raising total seal injection flow to maximum. Annunciator K08-D7 RCP Seal Cavity Press HI/LO will alarm due to the high seal injection flow. The CRS should reference 1203.012G K08 Annunciator Corrective Action. The ACA will direct the CRS to 1203.031 Reactor Coolant Pump and Motor Emergency. The crew should recognize high seal injection flow on C04 and take manual control of CV-1207 to establish 8-10gpm individual seal injection flow.

The RCP seal injection transient will cause the B RCP lower seal to fail (BOP-C) (SRO-C). The CRS should implement 1203.031 Reactor Coolant Pump and Motor Emergency and reduce power to ~60% in order to stop the B RCP using 1103.006 (ATC-R) (SRO-N) The ATC should reduce power to ~60% using the SG/RX Master station in hand.

At ~65% power or as directed by the lead evaluator a loss of the H2 bus will occur resulting in a loss of two RCPs (B and D) (BOP-M) (SRO-M) (ATC-M). If reactor power is <55% then the A RCP will be tripped along with the H2 bus by the simulator instructor to create the condition for the recognition of an RPS failure. The ATC should recognize the reactor should have tripped and manually trip the reactor.

(ATC-C) (SRO-C) (ATC-CT) (TS).

TS 3.3.1 Condition C CT criteria - The reactor should be manually tripped within 2 minutes of the loss of the H2 bus.

Two control rods will fail to fully insert on the Rx trip. (SRO-C) (ATC-C) The ATC should recognize and report this condition during the immediate action report. The CRS will direct 1202.012, RT-12 Emergency Boration to be performed. (ATC-CT)

TS 3.1.4 Condition C CT criteria - Emergency boration should be started within 5 minutes of the RX trip.

CONTINUED Page 4 of 19

Appendix D Scenario Outline Form ES-D-1 SCENARIO #1-R5 NARRATIVE CONTINUED

~10 minutes post trip a ~700gpm leak will occur in the reactor building from the failed RCP seal (ALL-M) (TS). Subcooling margin will be lost requiring the BOP to trip all running RCPs within 2 minutes. (BOP-CT) All ESAS channels will auto actuate except ES Ch.1. The crew should recognize the leak and recognize ES Ch. 1 failed to actuate. The ATC should manually actuate ES Ch.1 from C04 (ATC-CT). CV-1300 on ES Ch.2 fails to close. The BOP should recognize the open valve and attempt to close it manually. The CRS will direct operation per either 1202.002 Loss of Subcooling Margin or 1202.010 ESAS. Either procedure is acceptable. If the CRS enters ESAS he will eventually be directed to 1202.002 Loss of Subcooling Margin.

TS 3.4.13 Condition A CT-Criteria RCP should be tripped within 2 minutes of the loss of SCM.

CT criteria - ESAS Ch. 1 should be manually actuated before RT-10 is reported complete.

Page 5 of 19

Appendix D Scenario Outline Form ES-D-1 Facility: ANO-1 Scenario No.: 2-R5 Op-Test No.: 2008-1 Examiners: __________________________ Operators: __________________________

Initial Conditions:

100% Power EFIC is failed and will not auto actuate.

C2A IA compressor is out of service for overhaul.

Provide picture of RS-4 Turnover:

Day shift normal working day.

C2A IA compressor is out of service for overhaul.

Currently under a severe thunderstorm warning for the next hour. All actions of 1203.025 Natural Emergencies are complete.

Swap ICW pumps to have P33A and P33B running to allow visual inspection of P33C. The Inside AO has been briefed and is standing by the ICW pumps. P33B had not been drained.

Event Malf. No. Event Event No. Type* Description N (SRO, 1 N/A Swap operating ICW pumps BOP)

Lightning Lightning strike strike N/A 2 DI-DG2S #2 EDG auto start K01A3 #2 EDG Auto Start Alarm C (SRO, #2 EDG SW valve fails to open (TS).

CV-3807 BOP) #2 EDG Shutdown N (SRO) Dispatcher directs a power reduction to 700Mwe in the 3 N/A R (ATC) next 15 minutes.

Lightning Lightning strike strike I (SRO, 4 ATC)

ED451 Loss of the NNI Y power supply.

5 ED183 M (ALL) Loss of Offsite Power/Degraded Power DG175 #1 EDG will not auto start (BOP-CT)

C (SRO, 6 BOP)

DI_DG1_VR-LW #1 EDG voltage low (<4100V)

C (SRO, 7 FW621 EFIC fails (ATC-CT) (TS)

ATC)

DG173 M (ALL) #1 EDG will trip (TS) 8 A901 N/A Alternate AC Generator available. (BOP-CT)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 15

Appendix D Scenario Outline Form ES-D-1 SCENARIO #2-R5 NARRATIVE The crew will assume the watch at 100% power. C2A IA compressor is out of service for overhaul. It is a day shift normal working day. ANO is currently under a severe thunderstorm warning for the next hour. All actions of 1203.025 Natural Emergencies are complete.

The crew will start P33B and secure P33C per1104.028 ICW Operating Procedure step10.2 (BOP-N) (SRO-N). This is being performed in order to visually inspect the motor rotor. P33B had not been drained.

A lightning strike will cause the auto start of the #2 EDG. The CRS should direct operations per 1203.012A Annunciator K01 Corrective Actions. The crew should determine it is a spurious actuation. The #2 EDG SW valve CV-3807 will not open resulting in no cooling flow to the EDG (BOP-C) (SRO-C). The crew should take the #2 EDG to lockout to prevent an automatic trip of the EDG (TS).

TS 3.8.1. Condition B The dispatcher will call and direct U1 to reduce net generator output to 700Mwe (SRO-N)

(ATC-R).

A second lightning strike will result in the loss on NNI-Y power supply (ATC-I) SRO-I). The crew should recognize the loss of NNI-Y power. The CRS should implement 1203.047 Loss of NNI Power. The power supply breakers on RS-4 bkr 9 and Y01 bkr 39 will not reset. Breakers S-1 and S-2 on the NNI-Y power supply will not be tripped.

Letdown flow and pressure indication will be lost on C04. The letdown orifice bypass valve will fail to 50% reducing letdown flow. CFT pressure instrumentation will be lost along with NaOH tank temperature. All NNI-Y inputs to PMS/PDS will be lost or fail to mid scale.

A loss of offsite power will occur due to storm related grid instabilities (SRO-M) (ATC-M)

(BOP-M). The reactor will trip automatically. The CRS will direct operations per 1202.001 Rx Trip. After the immediate actions are complete the CRS will transition to either 1202.008 Blackout or 1202.007 Degraded Power depending on when the BOP manually starts the #1 EDG.

The #1 EDG will not auto start requiring the BOP to manually start the EDG (BOP-C) (BOP-CT).

The #1 EDG voltage will be <4100V requiring the BOP to raise EDG voltage.

CT criteria - The BOP should start the #1 EDG before the 15 minute criteria for declaring an SAE for a station blackout.

The EFIC system is failed and will not automatically actuate EFW on the Loss of Offsite Power (ATC-C) (SRO-C). The ATC should manually actuate EFW from the remote switch matrix (ATC-CT) (TS). The ATC will perform 1202.012 Repetitive Tasks RT-5 Verify proper EFW actuation and control. EFW may be manually actuated before the step in the EOP that directs verifying EFW actuated.

TS 3.3.11 Condition B CT - EFW should be manually actuated before the ERV opens in automatic.

CONTINUED Page 3 of 15

Appendix D Scenario Outline Form ES-D-1 SCENARIO #2-R5 NARRATIVE CONTINUED The #1 EDG will trip putting the plant into a blackout (SRO-M) (ATC-M) (BOP-M) (TS). The CRS will direct operations per 1202.008 Blackout.

TS 3.8.1 Condition E, 3.0.3 The alternate AC generator will become available. The BOP should energize the A3 bus from the AAC generator (BOP-CT).

CT criteria - The AAC generator should be placed in service within 20 minutes of regaining the AAC Generator.

Page 4 of 15

Appendix D Scenario Outline Form ES-D-1 Facility: ANO-1 Scenario No.: 3-R5 Op-Test No.: 2008-1 Examiners: __________________________ Operators: __________________________

Initial Conditions:

100% power C2A IA compressor is out of service for overhaul.

RPS is failed RX trip P/B on C03 is failed Turnover:

C2A IA compressor is out of service for overhaul.

Swap operating EH oil pumps following maintenance on the standby pump. The AO has been briefed and is standing by the EH pump.

Event Malf. No. Event Event No. Type* Description N (BOP, 1 N/A Swap the operating EH oil pumps SRO)

C (SRO, FW086 P8A heater drain pump winding failure and trip.

BOP) 2&3 N (SRO)

N/A Power reduction R (ATC)

I (SRO, 4 TR580 ATC, Controlling Header Pressure fails low BOP)

CO_P14B Operating EH pump will trip 5 CO_P14A C (BOP)

Turbine trip >43%

DI_PB9201 RP246,7,8,9 C (SRO, RPS is failed (TS)

ATC) 6 DI_ICC0020 C (ATC) C03 manual trip P/B failed (TS)

M (ALL) Manual reactor trip (ATC-CT)

RC002 M (ALL) ~150 GPM tube leak in the B SG (TS) (BOP-CT) 7&8 N/A N (ATC) Plant cooldown and depressurization 9 IMF CV061 C (ALL) Operating HPI pump trip

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 15

Appendix D Scenario Outline Form ES-D-1 SCENARIO #3-R5 NARRATIVE The crew will assume the watch with the plant at 100% power. C2A IA compressor is out of service for overhaul.

The turn over sheet will direct the crew to swap operating turbine electro-hydraulic pumps per 1106.012 Electro-Hydraulic Oil System Operation Section 14 (SRO-N) (BOP-N).

P8A heater drain pump will experience a winding failure causing a high temperature alarm and P8A trip (SRO-C) (BOP-C). The CRS should reference 1203.012E Annunciator Corrective Action P8A/P8B FLOW LO and CONDENSATE PUMP AUTOSTART.

A plant power reduction is required to maintain suction pressure. The CRS should direct the power reduction per 1203.045 Rapid Plant Shutdown (ATC-R) (SRO-N).

The controlling Turbine Header Pressure instrument will fail low. (SRO-C) (ATC-C) (BOP-C)

This will result in the turbine lowering demand to raise header pressure. The reactor and feedwater will rise as a result of the header pressure error. A SASS mismatch alarm will be received. The CRS will direct operations per 1203.012F Annunciator Corrective actions for SASS mismatch alarm and 1203.001 ICS Abnormal Operation. The crew should verify the turbine in manual control SG/RX master and both turbine bypass valves in manual. Once the plant is stable the crew will verify the alternate instrument is good and select the good it on C04.

The crew should return ICS to automatic.

After the power reduction the operating EH oil will trip. The standby pump will not start (BOP-C).

The loss of both EH pumps will result in a turbine trip. (Crew may complete a manual RX trip prior to the turbine trip) The reactor will fail to trip do to a failure of RPS (ATC-C) (TS). The manual RX trip pushbutton is failed (ATC-C). The ATC should manually trip the reactor using the shunt trip breakers (ATC-CT) (ATC-M) (BOP-M) (SRO-M).

TS 3.3.1 Condition C TS 3.3.2 Condition A CT criteria - The reactor should be manually tripped before the pressurizer goes solid.

A ~150gpm tube leak will occur in the B SG (ATC-M) (BOP-M) (SRO-M) (TS). The CRS should direct operation per 1202.006 Tube Rupture. The leak will be large enough to require HPI be initiated (BOP-CT). An RCS depressurization and cooldown should be started (ATC-N).

TS 3.4.13 Condition B CT criteria - HPI should be initiated before SCM is lost.

The Operating HPI pump breaker trips due to a motor fault. (SRO-C) (ATC-C) (BOP-C) The CRS should direct operations per 1203.026 Loss of Reactor Coolant Makeup section 1 Loss of HPI Pump. The crew should diagnose the pump trip as a breaker fault and start the standby HPI pump. HPI should be restarted using the standby pump or raised using the ES pump. (BOP-CT)

(TS) (BOP-CT)

TS 3.5.2 Condition A CT criteria - HPI should be started or raised on the ES pump before SCM is lost.

The scenario may be terminated when HPI has been restarted or at the direction of the lead evaluator.

Page 3 of 15

ML082750588

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