ML14155A502

Arkansas Nuclear One, Unit 2-2014-02-Final Outlines

ML14155A502
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/27/2014
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
laura hurley
References
50-313/14-02, 50-368/14-02 50-313/OL-14, 50-368/OL-14
Download: ML14155A502 (28)

Text

Revision 2 ES-401PWR Examination Outline FORM ES-401-2 Facility Name:Arkansas Nuclear One Unit 2 Date of Exam:2/21/2014 K1K2K3K4K5K6A1A2A3A4G*TotalTotal1424341186221221194Tier Totals636 5522710132342234122 285211111001121 10113Tier Totals43453235243 38812342212Note: 1.

Note:2.Note:3.Note:4.Note:5.Note:6.Note:7.*Note:8.Note:9.ES-401, 21 of 33 2.Plant Systems 13SRO-Only Points RO K/A Category Points A2G*1. Emergency

&AbnormalPlantEvolutions TierGroup5N/AN/A314For 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.

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.

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

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.

23Select 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.

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).

324333. Generic Knowledge and Abilities Categories 103361227Systems/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 Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

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.

Revision 2 Revision 2 ES-4012Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)

E/APE # / Name / Safety Function K1K2K3A1A2GK/A Topic(s)

IR#000007 Reactor Trip / 1 CE/E02 Reactor Trip Recovery / 1 03Annunciators and conditions indicating signals, and remedialactions associated with the (Reactor Trip Recovery).

3.0000008 Pressurizer Vapor Space Accident / 3 30Inadequate core cooling 4.31000009 Small Break LOCA / 3 0000011 Large Break LOCA / 3 01.20Ability to interpret and execute procedure steps.

4.61000015 RCP Malfunctions / 4 000017 RCP Malfunctions (Loss of RC Flow) / 4 01Natural circulation in a nuclear reactor power plant 4.41000022 Loss of Rx Coolant Makeup / 2 03Relationship between charging flow and PZR level 3.01000025 Loss of RHR System / 4 03LPI pumps 3.41000026 Loss of Component Cooling Water / 8 02The automatic actions (alignments) within the CCWS resultingfrom the actuation of the ESFAS 3.61000027 Pressurizer Pressure Control System Malfunction / 3 03Controllers and positioners 2.61000029 ATWS / 1 12M/G set power supply and reactor trip breakers 4.11000038 Steam Gen. Tube Rupture / 3 44Level operating limits for S/Gs 3.41000040 Steam Line Rupture / 4 CE/E05 Excessive Steam Demand / 4 02Adherence to appropriate procedures and operation within thelimitations in the Facility's license and amendments.

3.4000054 Loss of Main Feedwater / 4 CE/E06 Loss of Feedwater / 4 01Components, and functions of control and safety systems,including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

3.3000055 Station Blackout / 6 02Actions contained in EOP for loss of offsite and onsite power 4.31000056 Loss of Off-site Power / 6 47Proper operation of the ED/G load sequencer 3.81000057 Loss of Vital AC Inst. Bus / 6 04ESF system panel alarm annunciators and channel status indicators 3.71000058 Loss of DC Power / 6 02Actions contained in EOP for loss of dc power 4.01000062 Loss of Nuclear Svc Water / 4 0000065 Loss of Instrument Air / 8 03Knowing effects on plant operation of isolating certain equipment from instrument air 2.91000077 Generator Voltage and Electric Grid Disturbances / 6 03Under-excitation 3.31K/A Category Totals:

42434118ES-401, 22 of 33 Group Point Total:

111Revision 2 Revision 2 ES-4013Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)

E/APE # / Name / Safety Function K1K2K3A1A2GK/A Topic(s)

IR#000001 Continuous Rod Withdrawal / 1 0000003 Dropped Control Rod / 1 05Reactor power - turbine power 4.11000005 Inoperable/Stuck Control Rod / 1 0000024 Emergency Boration / 1 04Pumps2.61000028 Pressurizer Level Malfunction / 2 0000032 Loss of Source Range NI / 7 01Startup termination on source-range loss 3.21000033 Loss of Intermediate Range NI / 7 0000036 Fuel Handling Accident / 8 02SDM3.41000037 Steam Generator Tube Leak / 3 11When to isolate one or more S/Gs 3.81000051 Loss of Condenser Vacuum / 4 0000059 Accidental Liquid RadWaste Rel. / 9 0000060 Accidental Gaseous Radwaste Rel. / 9 0000061 ARM System Alarms / 7 0000067 Plant Fire On-site / 9 8 04.31Knowledge of annunciator alarms, indications, or response procedures.

4.21000068 Control Room Evac. / 8 18Actions contained in EOP for control room evacuation emergency task4.21000069 Loss of CTMT Integrity / 5 0000074 Inad. Core Cooling / 4 27ECCS valve control switches and indicators 4.21000076 High Reactor Coolant Activity / 9 0CE/A13 Natural Circ. / 4 0CE/A11 RCS Overcooling / 4 0CE/A16 Excess RCS Leakage / 2 0CE/E09 Functional Recovery 02Normal, abnormal and emergency operating proceduresassociated with (Functional Recovery).

3.21000000000K/A Category Totals:

2122119ES-401, 23 of 33 Group Point Total:

Revision 2 ES-4014Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO)

System # / Name K1K2K3K4K5K6A1A2A3A4GK/A Topic(s)

IR#003 Reactor Coolant Pump 02S/G3.51004 Chemical and Volume Control 30Relationship between temperature and pressure in CVCS components during solid plant operation 3.81005 Residual Heat Removal 0101RHR pumps; Heatup/cooldown rates 3; 3.52006 Emergency Core Cooling 0913Valve positioning on safety injection signal; Inadvertent SIS actuation 3.9;3.92007 Pressurizer Relief/Quench Tank 02Maintaining quench tank pressure 2.71008 Component Cooling Water 0502.44Sources of makeup water; Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect 3; 4.22010 Pressurizer Pressure Control 01Pressure detection systems 2.71012 Reactor Protection 03Channel blocks and bypasses 3.61013 Engineered Safety Features Actuation 01Fuel4.41022 Containment Cooling 04Cooling of control rod drive motors 2.81025 Ice Condenser 0026 Containment Spray 0208Cooling water; Automatic swapover to containment sump suction for recirculation phase after LOCA (RWST low-low level alarm) 4.1;4.12039 Main and Reheat Steam 0504Increasing steam demand, its relationship to increases in reactor power; Emergency feedwater pump turbines 3.3;3.82059 Main Feedwater 0307S/Gs; Tripping of MFW pump turbine 3.5; 32061 Auxiliary/Emergency Feedwater 02Decay heat sources and magnitude 3.21062 AC Electrical Distribution 0112Major system loads; Restoration of power to a system with a fault on it 3.3;3.22063 DC Electrical Distribution 02Breaker interlocks, permissives, bypasses and cross-ties 2.91064 Emergency Diesel Generator 08Fuel oil storage tanks 3.21073 Process Radiation Monitoring 01Those systems served by PRMs 3.61076 Service Water 02Emergency heat loads 3.71078 Instrument Air 01.30Ability to locate and operate components, including local controls.

4.41103 Containment 01Containment pressure, temperature, and humidity 3.71K/A Category Totals:32342234122 28ES-401, 24 of 33 Group Point Total:

Revision 2 ES-4015Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO)

System # / Name K1K2K3K4K5K6A1A2A3A4GK/A Topic(s)

IR#001 Control Rod Drive 0002 Reactor Coolant 04.11Knowledge of abnormal condition procedures.

4.01011 Pressurizer Level Control 02PZR heaters 3.11014 Rod Position Indication 0015 Nuclear Instrumentation 19Heat balance 2.91016 Non-nuclear Instrumentation 02PZR LCS3.41017 In-core Temperature Monitor 0027 Containment Iodine Removal 0028 Hydrogen Recombiner and Purge Control 03The hydrogen air concentration in excess of limit flame propagation or detonation with resulting equipment damage in containment 3.41029 Containment Purge 03Automatic purge isolation 3.21033 Spent Fuel Pool Cooling 0034 Fuel Handling Equipment 01Radiation levels 3.31035 Steam Generator 0041 Steam Dump/Turbine Bypass Control 01RCS T-ave. meter (cooldown rate) 3.21045 Main Turbine Generator 06RCS, during steam valve test 2.61055 Condenser Air Removal 0056 Condensate 0068 Liquid Radwaste 0071 Waste Gas Disposal 0072 Area Radiation Monitoring 0075 Circulating Water 0079 Station Air 0086 Fire Protection 05Deluge valves 3.01K/A Category Totals:11111001121 10ES-401, 25 of 33 Group Point Total:

Revision 2 Revision 2 ES-4012Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)

E/APE # / Name / Safety Function K1K2K3A1A2GK/A Topic(s)

IR#000007 Reactor Trip / 1 CE/E02 Reactor Trip Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 0000009 Small Break LOCA / 3 06Whether PZR water inventory loss is imminent 4.31000011 Large Break LOCA / 3 0000015 RCP Malfunctions / 4 000017 RCP Malfunctions (Loss of RC Flow) / 4 0000022 Loss of Rx Coolant Makeup / 2 0000025 Loss of RHR System / 4 0000026 Loss of Component Cooling Water / 8 0000027 Pressurizer Pressure Control System Malfunction / 3 0000029 ATWS / 1 0000038 Steam Gen. Tube Rupture / 3 0000040 Steam Line Rupture / 4 04.02Knowledge of system set points, interlocks and automatic actionsassociated with EOP entry conditions.

4.6CE/E05 Excessive Steam Demand / 4 000054 Loss of Main Feedwater / 4 CE/E06 Loss of Feedwater / 4 000055 Station Blackout / 6 01Existing valve positioning on a loss of instrument air system 3.71000056 Loss of Off-site Power / 6 0000057 Loss of Vital AC Inst. Bus / 6 0000058 Loss of DC Power / 6 03DC loads lost; impact on to operate and monitor plant systems 3.91000062 Loss of Nuclear Svc Water / 4 01.32Ability to explain and apply system limits and precautions.

4.01000065 Loss of Instrument Air / 8 0000077 Generator Voltage and Electric Grid Disturbances / 6 02.36Ability to analyze the effect of maintenance activities, such asdegraded power sources, on the status of limiting conditions for operations.

4.21K/A Category Totals:

0000336ES-401, 22 of 33 Group Point Total:

001Revision 2 Revision 2 ES-4013Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)

E/APE # / Name / Safety Function K1K2K3A1A2GK/A Topic(s)

IR#000001 Continuous Rod Withdrawal / 1 0000003 Dropped Control Rod / 1 0000005 Inoperable/Stuck Control Rod / 1 02.22Knowledge of limiting conditions for operations and safety limits.

4.71000024 Emergency Boration / 1 01Whether boron flow and/or MOVs are malfunctioning, from plant conditions 4.11000028 Pressurizer Level Malfunction / 2 0000032 Loss of Source Range NI / 7 0000033 Loss of Intermediate Range NI / 7 0000036 Fuel Handling Accident / 8 0000037 Steam Generator Tube Leak / 3 0000051 Loss of Condenser Vacuum / 4 0000059 Accidental Liquid RadWaste Rel. / 9 0000060 Accidental Gaseous Radwaste Rel. / 9 0000061 ARM System Alarms / 7 0000067 Plant Fire On-site / 9 8 0000068 Control Room Evac. / 8 0000069 Loss of CTMT Integrity / 5 02Verification of automatic and manual means of restoring integrity 4.41000074 Inad. Core Cooling / 4 0000076 High Reactor Coolant Activity / 9 0CE/A13 Natural Circ. / 4 0CE/A11 RCS Overcooling / 4 01Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

3.31CE/A16 Excess RCS Leakage / 2 0CE/E09 Functional Recovery 0K/A Category Totals:

0000314ES-401, 23 of 33 Group Point Total:

Revision 2 Revision 2 ES-4014Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (SRO)

System # / Name K1K2K3K4K5K6A1A2A3A4GK/A Topic(s)

IR#003 Reactor Coolant Pump 0004 Chemical and Volume Control 0005 Residual Heat Removal 0006 Emergency Core Cooling 0007 Pressurizer Relief/Quench Tank 0008 Component Cooling Water 0010 Pressurizer Pressure Control 0012 Reactor Protection 03Incorrect channel bypassing 3.71013 Engineered Safety Features Actuation 02.22Knowledge of limiting conditions for operations and safety limits.4.71022 Containment Cooling 0025 Ice Condenser 0026 Containment Spray 01.20Ability to interpret and execute procedure steps.

4.61039 Main and Reheat Steam 0059 Main Feedwater 0061 Auxiliary/Emergency Feedwater 04pump failure or improper operation 3.81062 AC Electrical Distribution 0063 DC Electrical Distribution 0064 Emergency Diesel Generator 01Failure modes of water, oil, and air valves 3.31073 Process Radiation Monitoring 0076 Service Water 0078 Instrument Air 0103 Containment 0K/A Category Totals:00000003002 5ES-401, 24 of 33 Group Point Total:

Revision 2 Revision 2 ES-4015Form ES-401-2 ES-401PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (SRO)

System # / Name K1K2K3K4K5K6A1A2A3A4GK/A Topic(s)

IR#001 Control Rod Drive 18Incorrect rod stepping sequence 3.81002 Reactor Coolant 0011 Pressurizer Level Control 0014 Rod Position Indication 0015 Nuclear Instrumentation 0016 Non-nuclear Instrumentation 01.07Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

4.71017 In-core Temperature Monitor 0027 Containment Iodine Removal 0028 Hydrogen Recombiner and Purge Control 0029 Containment Purge 0033 Spent Fuel Pool Cooling 0034 Fuel Handling Equipment 01Fuel protection from binding and dropping 3.41035 Steam Generator 0041 Steam Dump/Turbine Bypass Control 0045 Main Turbine Generator 0055 Condenser Air Removal 0056 Condensate 0068 Liquid Radwaste 0071 Waste Gas Disposal 0072 Area Radiation Monitoring 0075 Circulating Water 0079 Station Air 0086 Fire Protection 0K/A Category Totals:00010001001 3ES-401, 25 of 33 Group Point Total:

Revision 2 Revision 2 ES-401Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3Facility Name:Arkansas Nuclear One Unit 2 Date of Exam:2/21/2014 IR#IR#2.1.07Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

2.912.1.21Ability to verify the controlled procedure copy.

3.912.1.44Knowledge of RO duties in the control room during fuel handling such as responding to alarms from thefuel handling area, communication with the fuel storage facility, systems operated from the control roomin support of fueling operations, and supporting instrumentation.

3.912.1.2.1.01Knowledge of conduct of operations requirements.

4.212.1.39Knowledge of conservative decision making practices.

4.31Subtotal322.2.06Knowledge of the process for making changes to procedures.

3.012.2.07Knowledge of the process for conducting special or infrequent tests.

2.912.2.43Knowledge of the process used to track inoperable alarms.

3.012.2.14Knowledge of the process for controlling equipment configuration or status.

4.312.2.17Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

3.812.2.Subtotal322.3.11Ability to control radiation releases.

3.812.3.13Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response toradiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to lockedhigh-radiation areas, aligning filters, etc.

3.412.3.2.3.14Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.812.3.2.3.Subtotal212.4.05Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

3.712.4.32Knowledge of operator response to loss of all annunciators.

3.612.4.2.4.14Knowledge of general guidelines for EOP usage.

4.512.4.28Knowledge of procedures relating to a security event.

4.112.4.Subtotal22107ES-401, Page 26 of 33 1.Conduct of Operations ROSRO-OnlyK/A #Tier 3 Point Total 3.Radiation Control4.Emergency Procedures

/ Plan2.Equipment ControlCategoryTopicRevision 2 ES-401Record of Rejected K/AsForm ES-401-4ES-401, Page 27 of 33Tier /GroupRandomlySelected K/AReason for Rejection ROT1/G1QID # 50022 G2.4.1There are no immediate actions (as defined by plant procedures)associated with Loss of Reactor Coolant Makeup (Loss of Charging AOP)Selected 0022 K1.03 as the replacement K/A ROT1/G1QID #100038 G2.2.42Overlap with Operating exam (there is a SGTR event with Tech Spec Calls)Selected 0038 A1.44 as the replacement K/A ROT1/G1QID #16058 A1.01Question does not apply. Unit does not have the ability to cross tie Vital DCbuses. Selected 058 AK3.02 as the replacement K/A ROT1/G2QID #21033 AK 3.01 Startup Channels serve the purpose of Intermediate range instrumentationduring a reactor startup. Selected 032 AK3.01 as the replacement K/A ROT2/G1QID #390013 A3.02System over sample concerns between Tier 1 and Tier 2. Selected 026K4.08 as the replacement K/A ROT2/G1QID #46061 A3.03System over sample concerns. Selected 005 A1.01 as the replacement K/A ROT2/G1QID #51064 A4.06System over sample concerns. Selected 039 A2.05 as the replacement K/A ROT2/G2QID #65086 A1.01Unit 2 does not control the stations fire pumps or have indications in thecontrol room for the fire water system. The fire water system is operated byUnit 1. Selected 086 A 4.05 as the replacement K/A ROT3QID #66G2.1.5Rejected original G2.1.5 due to being an SRO duty. Selected G2.1.7 as thereplacement K/A ROT3QID #67G2.1.19Does not lend itself to a generic question (directs monitoring plantcomponents or systems). Selected G2.1.21 as the replacement K/A ROT3QID #71G2.2.42Does not lend itself to a generic question (specific system parameters thatare entry level conditions for Tech Specs). Selected G2.2.43 as thereplacement K/ASROT1/G2QID #83024 G2.4.41There is not an E-Plan associated with Emergency Boration. Selected 024AA2.01 as the replacement K/ASROT3QID #99G2.4.23Difficultly of matching K/A. Selected G2.4.14 as the replacement K/A Rev 2 Revision 1ES-301Administrative Topics OutlineForm ES-301-1Facility: Arkansas Nuclear One Unit 2Date of Examination: 02/10/2014Examination Level: RO X SROOperating Test Number: 2014-1Administrative Topic (see Note)TypeCode*Describe activity to be performedA1. Conduct of Operations2.1.20 RO(4.6)D/RSpent Fuel Pool Makeup CalculationANO-2-JPM-NRC-ADMIN-SFPMU2A2. Conduct of Operations2.1.23 RO (4.3)D/P/RCalculate Time to Boil using Computer ProgramANO-2-JPM-NRC-ADMIN-TTBCROA3. Equipment Control2.2.12 RO (3.7)N/REvaluate Containment Atmospheric ConditionsANO-2-JPM-NRC-ADMIN-CNTMTA4. Radiation Control2.3.7 RO (3.5)D/RReview Emergency RWP and Perform EvolutionANO-2-JPM-NRC-ADMIN-RWP2Emergency Procedures/PlanNOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they areretaking only the administrative topics, when all 5 are required.* Type Codes & Criteria:(C)ontrol room, (S)imulator, or Class(R)oom(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)(N)ew or (M)odified from bank ( 1)(P)revious 2 exams ( 1; randomly selected)

Revision 2ES-301Administrative Topics OutlineForm ES-301-1Facility: Arkansas Nuclear One Unit 2Date of Examination: 02/10/2014Examination Level: RO SRO XOperating Test Number: 2014-1Administrative Topic (see Note)TypeCode*Describe activity to be performedA5. Conduct of Operations2.1.20 SRO (4.6)D/RReview and Approve Spent Fuel Pool MakeupCalculationANO-2-JPM-NRC-ADMIN-SFPMUA6. Conduct of Operations2.1.40 SRO (3.9)N/RDetermine Shutdown Operations Protection PlanConditionANO-2-JPM-NRC-ADMIN-SOPP1A7. Equipment Control2.2.14 SRO (4.3)D/P/RSupervisory Review of Maintenance Activities forConfiguration ControlANO-2-JPM-NRC-ADMIN-MAINTA8. Radiation Control2.3.7 SRO (3.6)

M/RReview Emergency RWPANO-2-JPM-NRC-ADMIN-RWP3A9. EmergencyProcedures/Plan2.4.38 SRO (4.4)N/REOF Evacuation DeterminationANO-2-JPM-NRC-ADMIN-EOFEVACNOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they areretaking only the administrative topics, when all 5 are required.* Type Codes & Criteria:(C)ontrol room, (S)imulator, or Class(R)oom(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)(N)ew or (M)odified from bank ( 1)(P)revious 2 exams ( 1; randomly selected)

Revision 2ES-301Control Room/In-Plant Systems OutlineForm ES-301-2Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014Exam Level: RO X SRO-I SRO-UOperating Test No.: 2014-1Control Room Systems

@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)System / JPM TitleType Code*SafetyFunctionS1.ANO-2-JPM-NRC-CNTCL022 A4.03 RO-3.2/SRO-3.2Verify Containment Coolers in Emergency ModeA/D/EN/L/S 5Containment S2.ANO-2-JPM-NRC-ELEC06062 A4.01 RO-3.3/SRO-3.1Transfer Auxiliaries from SU#2 to SU#3 for 2A-1 A/M/S6Electrical S3. ANO-2-JPM-NRC-CVCS2004 A4.07 RO-3.9/SRO3.7Perform Emergency BorationA/D/L/S1Reactivity controlS4.ANO-2-JPM-NRC-EFW01061 A1.01 RO-3.9/SRO4.2Shutdown EFW Train 'A' with EFAS Signal PresentD/EN/L/S4Heat RemovalSecondary S5.ANO-2-JPM-NRC-FWCS1035 A4.01 RO-3.7/SRO-3.6Place Feedwater Control system in AutomaticD/S4Heat RemovalPrimaryS6.ANO-2-JPM-NRC-CVCS12004 A4.06 RO-3.6/SRO-3.1Verification of Minimum Letdown FlowN/S2Inventory Control S7.ANO-2-JPM-NRC-EOP6012 A2.06 RO-4.4/SRO-4.7Manually Trip the ReactorA/D/S7Instrumentation S8.ANO-2-JPM-NRC-PZR01010 A4.01 RO-3.7/SRO-3.5Equalize RCS and Pressurizer BoronD/S3Pressure ControlIn-Plant Systems

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

P1.ANO-2-JPM-NRC-PRHTR068 AA1.07 RO-4.1/SRO-4.2Perform Local Operations of the Proportional HeatersD/E/L3Pressure ControlP2.ANO-2-JPM-NRC-EDDCS064 A4.01 RO-4.0/SRO-4.3Startup Diesel Generator Without DC Control Power (2K-4A)D/E/L6ElectricalP3.ANO-2-JPM-NRC-WGDTR071 A2.02 RO-3.3/SRO-3.6Perform Waste Gas Decay Tank ReleaseA/N/R9Rad Control

@All RO and SRO-I control room (and in-plant) systems must be different and serve different safetyfunctions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions mayoverlap those tested in the control room.* Type CodesCriteria for RO / SRO-I / SRO-U(A)lternate path(C)ontrol room(D)irect from bank(E)mergency or abnormal in-plant(EN)gineered safety feature(L)ow-Power / Shutdown(N)ew or (M)odified from bank including 1(A)(P)revious 2 exams(R)CA(S)imulator4-6 / 4-6 / 2-3 9 / 8 / 4 1 / 1 / 1 - / - / 1 (control room system) 1 / 1 / 1 2 / 2 / 1 3 / 3 / 2 (randomly selected) 1 / 1 / 1 Revision 2ES-301Control Room/In-Plant Systems OutlineForm ES-301-2Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014Exam Level: RO SRO-I X SRO-UOperating Test No.: 2014-1Control Room Systems

@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)System / JPM TitleType Code*SafetyFunctionS1.ANO-2-JPM-NRC-CNTCL022 A4.03 RO-3.2/SRO-3.2Verify Containment Coolers in Emergency ModeA/D/EN/L/S 5Containment S2.ANO-2-JPM-NRC-ELEC06062 A4.01 RO-3.3/SRO-3.1Transfer Auxiliaries from SU#2 to SU#3 for 2A-1 A/M/S6Electrical S3. ANO-2-JPM-NRC-CVCS2004 A4.07 RO-3.9/SRO3.7Perform Emergency BorationA/D/L/S1Reactivity controlS4.ANO-2-JPM-NRC-EFW01061 A1.01 RO-3.9/SRO4.2Shutdown EFW Train 'A' with EFAS Signal PresentD/EN/L/S4Heat RemovalSecondary S5.ANO-2-JPM-NRC-FWCS1035 A4.01 RO-3.7/SRO-3.6Place Feedwater Control system in AutomaticD/S4Heat RemovalPrimaryS6.ANO-2-JPM-NRC-CVCS12004 A4.06 RO-3.6/SRO-3.1Verification of Minimum Letdown FlowN/S2Inventory Control S7.ANO-2-JPM-NRC-EOP6012 A2.06 RO-4.4/SRO-4.7Manually Trip the ReactorA/D/S7Instrumentation S8.In-Plant Systems

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

P1.ANO-2-JPM-NRC-PRHTR068 AA1.07 RO-4.1/SRO-4.2Perform Local Operations of the Proportional HeatersD/E/L3Pressure ControlP2.ANO-2-JPM-NRC-EDDCS064 A4.01 RO-4.0/SRO-4.3Startup Diesel Generator Without DC Control Power (2K-4A)D/E/L6ElectricalP3.ANO-2-JPM-NRC-WGDTR071 A2.02 RO-3.3/SRO-3.6Perform Waste Gas Decay Tank ReleaseA/N/R9Rad Control

@All RO and SRO-I control room (and in-plant) systems must be different and serve different safetyfunctions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions mayoverlap those tested in the control room.* Type CodesCriteria for RO / SRO-I / SRO-U(A)lternate path(C)ontrol room(D)irect from bank(E)mergency or abnormal in-plant(EN)gineered safety feature(L)ow-Power / Shutdown(N)ew or (M)odified from bank including 1(A)(P)revious 2 exams(R)CA(S)imulator4-6 / 4-6 / 2-3 9 / 8 / 4 1 / 1 / 1 - / - / 1 (control room system) 1 / 1 / 1 2 / 2 / 1 3 / 3 / 2 (randomly selected) 1 / 1 / 1 Revision 2ES-301Control Room/In-Plant Systems OutlineForm ES-301-2Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014Exam Level: RO SRO-I SRO-U XOperating Test No.: 2014-1Control Room Systems

@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)System / JPM TitleType Code*SafetyFunctionS1.ANO-2-JPM-NRC-CNTCL022 A4.03 RO-3.2/SRO-3.2Verify Containment Coolers in Emergency ModeA/D/EN/L/S 5Containment S2.S3.S4.ANO-2-JPM-NRC-EFW01061 A1.01 RO-3.9/SRO4.2Shutdown EFW Train 'A' with EFAS Signal PresentD/EN/L/S4Heat RemovalSecondary S5.S6.S7.S8.In-Plant Systems

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

P1.ANO-2-JPM-NRC-PRHTR068 AA1.07 RO-4.1/SRO-4.2Perform Local Operations of the Proportional HeatersD/E/L3Pressure ControlP2.ANO-2-JPM-NRC-EDDCS064 A4.01 RO-4.0/SRO-4.3Startup Diesel Generator Without DC Control Power (2K-4A)D/E/L6ElectricalP3.ANO-2-JPM-NRC-WGDTR071 A2.02 RO-3.3/SRO-3.6Perform Waste Gas Decay Tank ReleaseA/N/R9Rad Control

@All RO and SRO-I control room (and in-plant) systems must be different and serve different safetyfunctions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions mayoverlap those tested in the control room.* Type CodesCriteria for RO / SRO-I / SRO-U(A)lternate path(C)ontrol room(D)irect from bank(E)mergency or abnormal in-plant(EN)gineered safety feature(L)ow-Power / Shutdown(N)ew or (M)odified from bank including 1(A)(P)revious 2 exams(R)CA(S)imulator4-6 / 4-6 / 2-3 9 / 8 / 4 1 / 1 / 1 - / - / 1 (control room system) 1 / 1 / 1 2 / 2 / 1 3 / 3 / 2 (randomly selected) 1 / 1 / 1 Appendix D Scenario 1Form ES-D-1Revision 3Page 1 of 51Facility: ANO-2Scenario No.: 1 (New)Op-Test No.: 2014-1Examiners:Operators:Initial Conditions:100%, 260 EFPD. RED Train Maintenance Week.Turnover:EOOS indicates 'Minimal Risk'.Evolution scheduled: Shift Control Element Drive Mechanism (CEDM) fans from 2VSF-35D to 2VSF-35C IAW2104.033 starting with 10.6.EventNo.Malf. No.Event Type*EventDescription 1N (BOP)N (SRO)Shift Control Elements Drive Mechanism (CEDM) fans.OP-2104.033, Containment Atmosphere Control.

2XCVLDNHXOUK12D01I (ATC)I (SRO)The temperature input to the letdown HX temperaturecontroller (2TIC-4815) fails Hi.OP-2203.012L, Annuciator 2K-12 Corrective Action (ACA)3CT2VSF1DC (BOP)C (SRO)TS (SRO)2VSF-1D Containment cooler trips. TS for SRO.OP-2203.012D/E, 2K-04 and 2K05 ACAs 4CEA43DROPR (ATC)C (BOP)C (SRO)TS (SRO)CEA 43 fully inserts. TS for SRO.OP-2203.003, CEA Malfunction AOP 5RCP2P32ALOSC (ATC)C (SRO)'A' RCP oil leak.OP-2203.025, RCP Emergencies AOP 6MSSGBLKM (ALL)Excess Steam Demand inside containment on 'B' SG.OP-2202.001, Standard Post Trip Actions (SPTAs) EOPand OP-2202.009, Functional Recovery EOP.

7CV4652C (ATC)C (SRO)'B' RCP normal spray valve fails open.OP-2202.010, Standard Attachments EOP orOP-2203.0028, Pressurizer System Malfunction AOP 8EFW2P7BFLTEFW2P7ACOUM (ALL)2P-7B EFW pump motor fault on start, 2P-7A EFW pumpcoupling failure.OP-2202.009, Functional Recovery EOP.

9CV0760DO_CV_0760_1DO_CV_0760_2 CV0761DO_CV_0760_1DO_CV_0760_2C (BOP)C (SRO)The selected AFW flow path discharge valve (2CV-0760 or2CV-0761) breaker trip.OP-2202.010, Standard Attachments EOP.EndpointFeedwater is restored to 'A' SG.* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D Scenario 1Form ES-D-1Revision 3Page 2 of 51Target Quantitative Attributes (Section D.5.d)Actual AttributesTotal Malfunctions (5-8) 8Malfunctions after EOP entry (1-2) 2Abnormal Events (2-4) 2Major Transients (1-2) 2EOPs entered requiring substantive actions (1-2) 1EOP contingencies requiring substantive actions (0-1) 1Critical Tasks (2-3) 3Critical TaskJustificationReferences'A' RCP must be secured within10 min of the reactor trip.Exceeding operating limits hasthe potential to degrade the RCSpressure boundary. RCPs shouldbe maintained in an availablecondition for last-resort use ifneeded.1015.050 Time CriticalOperation action program,Attachment CCE EPGB Simulator CTs:CT-23, Trip any RCP exceedingoperating limits (ESDE-03,FRG-04)Stabilize and control RCStemperature after the ESDblowdown terminates. MaintainRCS pressure within thePressure-Temperature limits of200°F and 30°F Margin toSaturation throughoutimplementation of SPTAs andFunctional Recovery EOP.If RCS heatup is allowed afterSG blowdown, the RCS couldover pressurize and result inlifting PZR and SG safeties.These pressure stresses added tothermal stresses of rapidcooldown could present PTSconcerns.CE EPGB Simulator CTs:CT-07, Establish RCStemperature Control (SPTA-07,ESDE-05, HR-05)Restore Feedwater prior to bothSG levels reaching 70" widerange.Inventory in the unaffected SG isrequired to remove decay heatfrom the reactor core (core meltpotential).CE EPGB Simulator CTs:CT-08, Establish RCS HeatRemoval (ESDE-08, HR-01)EOP 2202.009 FunctionalRecoveryEOP 2202.006 Loss ofFeedwater EOP Tech GuideScenario #1 Objectives 1)Evaluate individual ability to transfer CEDM fans.

2)Evaluate individual response to a failure of a temperature input to the letdown heat exchangerand ability to manually control temperature.

3)Evaluate individual response to a trip of a Containment Cooling fan.

4)Evaluate individual response to a CEA Malfunction.

5)Evaluate individual response to a Reactor Coolant pump oil leak (RCP emergencies).

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

7)Evaluate crew ability to mitigate a Loss of Feedwater.

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

9)Evaluate individual ability to respond to a failure of an AFW pump discharge valve.

10)Evaluate individual ability to respond to RCP spray valve failing open.

Appendix D Scenario 1Form ES-D-1Revision 3Page 3 of 51 SCENARIO #1 NARRATIVESimulator session begins with the plant at 100% power steady state.When the crew has completed their control room walk down and brief, the BOP will shift ControlElements Drive Mechanism (CEDM) fans from 2VSF-35D to 2VSF-35C.When the CEDM fans have been shifted or cued by lead examiner, the temperature input (2TE-4815) to the letdown heat exchanger temperature controller will fail high. The ATC will report thatthe letdown heat exchanger temperature is reading high on the hand indicating controller but thecomputer point and control board indication are reading lower than normal due to excessive coolingflow. The SRO will direct the ATC to take manual control of the Letdown heat exchangertemperature control valve and manually control temperature for the duration of the scenario.After the letdown temperature controller has been placed in manual and cued by the lead examiner,2VSF-1D containment cooler will trip. The BOP will determine that 2VSF-1D containment coolerhas tripped and refer to OP-2203.012D/E, 2K04 and 2K05 Annunciator Corrective Actions. TheBOP will start the idle containment cooler to maintain containment temperature and pressure in theacceptable region of operation. The SRO will enter Tech Spec 3.6.2.3 Action a. [Site OE: CR-ANO-2-2006-2444, 2VSF-1A motor failure and breaker trip.]After the BOP has started the idle containment cooling fan and cued by lead examiner, CEA 43 willdrop into the core due to faulty timing card. The SRO will enter OP-2203.003, CEA malfunctionAOP. The SRO should check that less than 2 CEAs are inserted and then commence a down powerwithin 15 minutes. The BOP should complete attachment C DNBR/LPD log. The SRO will enterTech Specs for CEA position (3.1.3.1 Action d) and Aztilt (3.2.3). [Site and industry OE: CR-ANO-2-2007-0127 dropped CEA, and NRC Event # 49601 Palo Verde dropped CEA.]After the crew has completed the required reactivity manipulation, entered the appropriate techspecs, and cued by the lead examiner, 'A' RCP oil leak will start that causes oil level to lower andbearing temperatures to rise. The CRS will enter OP-2203.025, RCP Emergencies AOP. The crewwill monitor the 'A' RCP oil level trend and bearing temperatures. After bearing temperaturesbegin to rise (trip criteria >18 0F/min.) the ATC should trip the reactor and secure the 'A' RCP. Thecrew may elect to secure a RCP in the 'B' S/G loop to balance flows. Securing a RCP notsatisfying operating limits is a time critical operator action per OP-1015.050 Time Critical OperatorAction Program. [Site OE: RCP oil leaks CR-ANO-2-2013-1602, CR-ANO-2-2013-587, CR-ANO-2-2013-58.]The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. After the reactortrips a Main Steam line break ('B' SG) inside containment will cause an Excess Steam Demand.Main Steam Isolation (MSIS) and Containment Spray (CSAS) will actuate tripping Main Feedwaterpumps, Condensate pumps, AFW pump, closing the MSIVs and feedwater block valves. The 2P-7B EFW pump motor will fail to start and 2P-7A EFW pump coupling will break causing a loss offeedwater event. The ATC will secure all the Reactor Coolant pumps due to the ContainmentSpray actuation. When the 'B' RCP spray valve (2CV-4652) handswitch is placed in manual, thevalve will fail open. The ATC must recognize this and isolate the spray valve using the associatedblock valve. If the spray valve is not isolated, the ATC's ability to control RCS pressure will belimited. [Industry OE for Excess Steam Demand, SOER 82-7, Reator Vessel Pressurized ThermalShock.]

Appendix D Scenario 1Form ES-D-1Revision 3Page 4 of 51SCENARIO #1 NARRATIVE (continued)After completing SPTAs, The SRO will diagnose an Excess Steam Demand and Loss of Feedwaterevent and enter OP-2202.009, Functional Recovery EOP. The crew will maintain post blowdowntemperature and pressure of the RCS to prevent pressurized thermal shock. The BOP will steam'A' S/G using the upstream Atmospheric Dump valve when 'B' S/G blows dry. The ATC shoulduse Auxiliary Spray to maintain RCS pressure. The Crew will restore Feedwater from the AFWpump (2P-75) after removing the MSIS and CSAS trip. The selected feed path valve from AFWwill trip its breaker when the valve is opened requiring use of the alternate flow path. [Loss offeedwater events industry OE: SOER 86-01 Reliability of PWR Auxiliary feedwater systems, andPRA operator action # 3 Establish flow to SGs from AFW to the SGs given a los of both EFW andMFW flow to the SGs.]

Appendix D Scenario 2Form ES-D-1Revision 3Page 1 of 41Facility: ANO-2Scenario No.: 2 (New)Op-Test No.: 2014-1Examiners:Operators:Initial Conditions:~40 %. MOL. 'C' channel Excore has failed and PPS points 1 through 4 are in bypass. RED TrainMaintenance Week. 'B' Component Cooling Water CCW pump in service.Turnover:EOOS indicates 'Minimal Risk'. Hold power 39- 41 % until S/G Chloride less than 10 ppb. SGblowdown ~120 gpm per SG for cleanup. Reactor Engineering is developing reactivity plan for powerescalation. 'C' channel Excore has failed and PPS points 1 through 4 are in bypass and all requiredactions are complete (TS 3.3.1.1 action 2 entered).Evolution scheduled: Perform Red Train Proportional Heater test starting with step 2.1.EventNo.Malf. No.Event Type*EventDescription 1N (ATC)N (SRO)Perform Red Train Proportional Heater test.OP-2103.005 Pressurizer Operations.

2NIBUPPERC (BOP)C (SRO)TS (SRO)'B' channel Excore upper chamber fails high. TS for SRO.OP-2203.026, NI malfunction AOP.

3XRCCHBPCNTI (ATC)I (BOP)I (SRO)'B' Pressurizer pressure control channel fails high.OP-2203.028, Pressurizer System Malfunction AOP 4CCW2P33BPWRCCW2P33CPWRC (BOP)C (SRO)2P-33B CCW pump trips and 2P-33C CCW pump fails tostart.OP-2203.025, RCP Emergencies AOP 5RCP2P32CSLKR (ATC)N (BOP)N (SRO)TS (SRO)'C' Reactor Coolant Pump (RCP) develops an intersystemLOCA from the RCS to CCW of 15 gpm. TS for SRO.OP-2203.016, Excess RCS leakage AOP 6RCP2P32CSLKESFK202AAFESFK202BAFM (All)'C' RCP intersystem LOCA degrades to 250 gpm. CCW toRCPs fail to auto close on CIAS.OP-2202.001, Standard Post Trip Actions (SPTA), andOP-2202.003, Loss of Coolant Accident EOP 7RCSHTRONC (ATC)C (SRO)Pressurizer Backup Heaters fail to de-energize on lowpressurizer level.

8CV0231C (BOP)C (SRO)Gland seal regulator 2PCV-0231 fails closed.2203.012B, Annuciator 2K-02 Corrective Action (ACA)EndpointCCW to RCP has been isolated, a RCS cooldown has beenstarted and condenser vacuum maintain by operation of2CV-0233 gland seal regulator.* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D Scenario 2Form ES-D-1Revision 3Page 2 of 41Target Quantitative Attributes (Section D.5.d)Actual AttributesTotal Malfunctions (5-8) 7Malfunctions after EOP entry (1-2) 2Abnormal Events (2-4) 4Major Transients (1-2) 1EOPs entered requiring substantive actions (1-2) 1EOP contingencies requiring substantive actions (0-1) 0Critical Tasks (2-3) 3Critical TaskJustificationIsolate RCS from leavingContainment by closing CCW tocontainment valves within 10min. of the Reactor Trip.Isolating CCW to Containmentprevents release of radioactivityby bypassing Containment.Failure to establish a containmentboundary could result in violatingexposure limits.CE EPGB Simulator CTs: CT-09, Establish ContainmentIsolation (LOCA-07)10CFR2010CFR100Commence an RCS cooldownwithin 30 minutes of entry intoOP-2202.003, LOCA EOP.Cooling down and depressurizingthe RCS removes decay heat andlowers the DP at the break,slowing the leak rate and reducingmakeup volume required. SDCentry conditions are also requiredfor long-term cooling.CE EPGB Simulator CTs: CT-20, Cool down and depressurizeRCS (LOCA-09)CR-ANO-2-2010-948, Criticaltask timesEstablish RCS pressure control tomaintain RCS subcooling.Maintain pressure andtemperature within the PT limitsof <2000 F and >30 0F MTSthroughout implementation ofOP-2202.003, LOCA EOP.Once RCS subcooling is lost,PZR level is no longer a validindication of RCS inventory. Areactor head void can form, and ifleft uncontrolled, could result incore uncovery and fuel damage.CE EPGB Simulator CTs: CT-06, Establish RCS PressureControl (LOCA-12)Scenario #2 Objectives 1)Evaluate individual ability to perform Proportional heater surveillance.

2)Evaluate individual response to a failure of a Nuclear Instrument.

3)Evaluate individual response to a Pressurizer System Malfunction (Pressure channel failure).

4)Evaluate individual response to a failure of a Component Cooling water pump.

5)Evaluate individual response to an intersystem Loss of Coolant Accident. (LOCA) 6)Evaluate crew ability to mitigate an intersystem LOCA.

7)Evaluate individual response to failure of a gland seal regulator.

8)Evaluate individual response to a failure of the pressurizer backup heaters to de-energize on lowlevel.SCENARIO #2 NARRATIVESimulator session begins with the plant at 40% power steady.When the crew has completed their control room walk down and brief, they will perform the RedTrain Proportional Heater surveillance.

Appendix D Scenario 2Form ES-D-1Revision 3Page 3 of 41SCENARIO #2 NARRATIVE (continued)When the Red Train Proportional Heater has been placed to auto or when cued by the leadexaminer, Channel B Excore upper chamber will fail high. The SRO will enter the OP-2203.026,NI Malfunction AOP and the crew should determine that B channel linear power is failed but logpower is still functional by monitoring output for the three chambers. The SRO will also enter TechSpec 3.3.1.1 Action 3 for Reactor Protection System. The BOP will trip points 1, 3, and 4 onchannel 'B' by using the linear calibrate switch. The points must be tripped because Channel C isin bypass. [Site OE: CR-ANO-2-2002-693, D Excore failure.]When the SDBCS permissives have been aligned and cued by the lead examiner, the 'B' pressurizerpressure control channel will fail high causing the spray valves to open and RCS pressure to lower.The CRS should enter the OP-2203.028, Pressurizer System Malfunction AOP. The crew willplace the other pressurizer pressure controller in service, verify that both spray valves close, and thepressurizer heaters restore RCS pressure. The BOP will place a maximum of one Steam Dump andBypass Control System (SDBCS) valve permissive in manual and all other permissives to off. [SiteOE: CR-ANO-2-2011-1605, Pressurizer pressure failing high.]After the BOP has tripped points 1, 3, and 4, and cued by lead examiner, 2P-33B CCW pump willtrip and 2P-33C CCW pump will fail to start automatically or manually. The SRO will enter OP-2203.025, RCP Emergencies AOP. The BOP should call NLOs to investigate the CCW pump trip.The SRO should direct the BOP to start 2P-33C CCW pump but it will fail to start. The SRO willthen direct opening all CCW cross-tie valves and start 2P-33A CCW pump. [Site OE: CR-ANO-2-2007-313, Trip of 2P-33B CCW pump with 2P-33C out of service for maintenance.]After the crew has restored CCW flow to the RCPs, and cued by the lead examiner, a 15 gpm RCSto CCW leak will start. The crew should notice that CCW Surge Tank level is rising. The crew'srecognition of the leak may be delayed because the 'B' Surge Tank level would normally rise fromthe different pump configuration. Also the CCW letdown radiation monitor will alarm indicatingRCS to CCW leakage. The SRO will enter OP-2203.016, Excess RCS Leakage AOP, and directthe board operator actions. The crew should perform leak rates, isolate letdown to verify the leak isnot in letdown and determine the need for a plant shutdown using normal boration. The SROshould enter Attachment A of Excess RCS Leakage, align the CCW surge tanks to the gascollection header and direct the NLO to control surge tank level. The crew will perform a powerreduction such that the plant will be taken off line. The SRO should enter Tech Spec 3.4.6.2 Actiona for RCS leakage. The ATC will boratethe RCS and reduce turbine load to maintain Tave-Trefwithin 2°F. The BOP will make preparations to remove secondary plant equipment from service aspower is reduced. [Industry OE: NRC information notice 92-36 Intersystem LOCA outsidecontainment. Industry OE: SEN-220, SEN-216, & SEN-182, RCS leakage events.]After the required reactivity manipulations are complete and cued by the lead examiner, the RCS toCCW will degrade to 250 gpm. The SRO will direct the reactor to be tripped, actuate SIAS &CCAS, secure RCPs, and isolate CCW to the RCPs. The CCW to RCPs valves will fail to autoclose on a valid CIAS. The SRO should enter and direct the actions of SPTAs.

Appendix D Scenario 2Form ES-D-1Revision 3Page 4 of 41SCENARIO #2 NARRATIVE (continued)The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. The ATC shouldrecognize that the pressurizer backup heaters failed to de-energize on low pressurizer level. Also,the crew should place the SDBCS master controller in Auto Local and lower the set point tomaintain margin to saturation.The SRO will diagnose and enter OP-2202.003, Loss of Coolant Accident EOP. After the crew hasentered the LOCA EOP and cued by the lead examiner, 2PCV-0231 gland seal pressure controlvalve will fail closed. The BOP will manually control 2CV-0233 gland seal bypass valve tomaintain gland seal header pressure and condenser vacuum. The crew will commence a cooldownto allow depressurization and refilling the pressurizer. The BOP will restore Service Water toComponent Cooling Water and Auxiliary Cooling water. [Site OE: for 2PCV-0231 gland sealpressure control valve CR-ANO-2-2009-719, CR-ANO-2-2009-311, and CR-ANO-2-2006-1406.]

Appendix D Scenario 3Form ES-D-1Revision 3Page 1 of 44Facility: ANO-2Scenario No.: 3 (New)Op-Test No.: 2014-1Examiners:Operators:Initial Conditions:98% MOL; RED Train Maintenance Week.Turnover:Mabelvale transmission line out of service and Unit 2 output is limited to 1035 MW gross, 995 MW net.EOOS indicates 'Minimal Risk'.Evolution scheduled: Shift running vacuum pumps.EventNo.Malf. No.EventType*EventDescription 1N (BOP)N (SRO)Shift running vacuum pumps.OP-2106.010 Condenser Vacuum System.

2XRRPZRLSPI (ATC)I (SRO)Reactor Reg. output to PZR level control program failsto 41%.OP-2203.028, Pressurizer System MalfunctionAOP3DO_HS_8259_GCV82591XRI2RITS8231A DO_RITS8231_10C (BOP)C (SRO)TS (SRO)2RITS-8271-2 Containment Atmosphere Monitor(CAMS) coupling fails and 2RITS-8231-1 CAMSparticulate detector fails. TS for SRO.OP-2203.012J, Annunciator 2K-10 Corrective Action(ACA), and OP-2203.012K, 2K-11 ACA 4MFWPMPBTRPR (ATC)C (BOP)C (SRO)TS (SRO)'B' Main Feed Water pump trips. TS (Tcold out ofrange high) for SRO.OP-2203.027, Loss of Main Feedwater pump AOP 5SGBTUBEM (ALL)TS (SRO)'B' Steam Generator Tube Rupture ramps up to 300gpm over 20 min. TS for SROOP-2203.038, Primary to Secondary leakage AOP, OP-2201.001 Standard Post Trip Actions EOP, and 2202.004Steam Generator Tube Rupture EOP 6ESFSIAS2CV48211C (ATC)C (BOP)C (SRO)Green Train SIAS fails to actuate and letdown isolation2CV-4821-1 fails open.OP-2202.010 Standard Attachments EOP 7CV0302CV0303CV0306C (ATC)C (SRO)Steam dump turbine bypass valve fails closed.OP-2105.008, SDBCS operations EndPoint'B' Steam Generator is isolated.* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D Scenario 3Form ES-D-1Revision 3Page 2 of 44Target Quantitative Attributes (Section D.5.d)Actual AttributesTotal Malfunctions (5-8) 6Malfunctions after EOP entry (1-2) 2Abnormal Events (2-4) 3Major Transients (1-2) 1EOPs entered requiring substantive actions (1-2) 1EOP contingencies requiring substantive actions (0-1) 0Critical Tasks (2-3) 2Critical TaskJustificationPerform one or more of thefollowing to maintain/restoreMargin to Saturation (MTS) > 30degrees F.Start the Green train HPSIpump and open HPSI injectionvalve(s)Isolate letdownAdjust RCS Cooldown rateMTS must be restored >30degrees F within 10 min.Once RCS subcooling is lost,PZR level is no longer a validindication of RCS inventory. Areactor head void can form, and ifleft uncontrolled, could result incore uncovery and fuel damage.RCP operating limits requireMTS to be >30 0F.CE EPGB Simulator CTs:CT-06, Establish RCS PressureControl (SGTR-10)1015.050 Time CriticalOperation Actions,Attachment CConduct an RCS cooldown toThot <535 0F and isolate 'B' SG(2202.010 Attachment 10completed) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after theReactor trip.Assumption is that the operatorwill diagnose within 30 minutesand then isolate within next 30minutes after entry into 2202.004,SGTR EOPReduce Thot below 535 0F isnecessary to prevent a MSSVfrom lifting (535 0F), thuspreventing an offsite release andexceeding 10CFR100 exposurelimits at the site boundary.CE EPGB Simulator CTs:CT-20, Cooldown anddepressurize RCS (SGTR-05)CT-14, Isolate most affected SG(SGTR-09).SAR Section 15.1.181015.050 Time CriticalOperation Actions,Attachment CEOP 2202.004, SGTR Tech GuideScenario #3 Objectives 1)Evaluate individual ability to perform a vacuum pump swap.

2)Evaluate individual response to a failure of a Containment Air monitor sample pump.

3)Evaluate individual response to a failure of a Containment Air monitor radiation monitor.

4)Evaluate individual response to a Pressurizer system malfunction involving pressurizer levelfailing high.

5)Evaluate individual response to a failure of loss of main feedwater pump.

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

7)Evaluate individual response to Green Train SIAS failure to actuate.

8)Evaluate individual response to a failure of letdown to automatically isolate.

9)Evaluate individual response to a steam dump turbine bypass valve failing closed.

Appendix D Scenario 3Form ES-D-1Revision 3Page 3 of 44 SCENARIO #3 NARRATIVESimulator session begins with the plant at ~98% power.When the crew has completed their control room walk down and brief, they will shift runningvacuum pumps.When the vacuum pumps have been shifted or when cued by the lead examiner, the Reactor Regpressurizer level program output will fail to minimum (41%). The SRO will enter the OP-2203.028, PZR System Malfunctions AOP. The ATC will take manual control of letdown tocontrol pressurizer level. The ATC must take control of PZR heaters to control RCS pressure (Allheaters will be energized) The ATC should place the PZR level controller to Auto and Local thenadjust the setpoint to programmed setpoint. Then Letdown should be placed back in automatic.This failure will also prevent manual start of back up charging pumps if needed to control PZRlevel.When letdown has been restored to automatic or cued by lead examiner, the in-service ContainmentAir Monitor System (CAMS) unit 2RITS-8271-2 coupling will fail. The ATC should report the2K-11 H10 CNTMT Air Monitor trouble alarm and refer to the ACA. The BOP should investigateand determine that 2RITS-8271-2 has low flow. When contacted, the NLO will report the couplinghas failed. The BOP should use OP-2104.033 Ventilation System Operations to place the standbyCAMS unit in service. When the standby CAMS unit is placed in service, the particulate detectorwill fail requiring entry into Tech Spec 3.4.6.1 Action a. [Site OE: CR-ANO-2-2013-1880, CAMSparticulate detector failure, CR-ANO-2-2011-2691, for CAMS unit low air flow and CR-ANO-2-2006-1191, for sample motor failure.]When all actions due to the CAMS failure have been completed, or cued by the lead examiner, 'B'MFWP will trip. The SRO will enter and implement OP-2203.027, Loss of Main Feedwater PumpAOP. This will result in steam flow exceeding feed flow and SG levels lowering. The crew willmanually and rapidly reduce turbine load, insert group 6 and group P CEAs, borate usingemergency boration to the RCS until feed flow is greater than steam flow. The SRO will berequired to enter Tech Spec 3.2.6 for Tc out of range high. Then start a normal boration powerreduction to less than 80%. [Industry OE: INPO event # OE31445 Loss of a Main feedwater pump.Site OE: CR-ANO-2-2009-3744, 'B' Main Feedwater pump trip.]After the crew has restored feedwater flow greater than steam flow or cued by lead examiner, aSteam Generator Tube Leak will occur on 'B' Steam Generator. The SRO will enter OP 2203.038,Primary to Secondary Leakage AOP. The SRO should enter TS 3.4.6.2 Action a, RCS leakage, andTS 3.7.1.2 for EFW when steam is isolated to 2P-7A EFW pump. When the leak rate exceeds 44gpm, the crew should determine that the leak rate is greater than 44 gpm. They will trip the reactor,actuate SIAS, and CCAS. [Industry OE: SOER 83-2, Steam Generator Tube Ruptures. SteamGenerator Tube Rupture response is a time critical operator action per OP-1015.050 Time CriticalOperator action program.]

Appendix D Scenario 3Form ES-D-1Revision 3Page 4 of 44SCENARIO #3 NARRATIVE (continued)The Crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. When SIAS isactuated the Green train components will fail to reposition. The crew should recognize the failureof Green train SIAS to actuate. The BOP should have a NLO check the breaker and pump motorsfor the Green train High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection pumps(LPSI) pumps. After the NLO report, the BOP should manually start 2P-89B HPSI pump and openall injection valves. Also, 2CV-4821-1 Red train letdown isolation valve will fail to close leavingletdown aligned. The ATC should recognize that letdown is aligned and close a Green trainisolation to help maintain RCS inventory. The crew will align Service Water to CCW to maintainforced circulation. The crew may lower Steam Dump Master Controller setpoint during SPTAs toaid in maintaining margin to saturation. [Industry OE: SOER 83-9, Valve inoperability cause bymotor operator failures for 2CV-4821-1.]The SRO will diagnose and enter OP-2202.004, Steam Generator Tube Rupture (SGTR) EOP. TheATC should commence cool down of the RCS to allow isolation of 'B' steam generator. The BOPwill override Service Water to Auxiliary Cooling Water to maintain condenser vacuum. During thecooldown, 2CV-0303, 2CV-0302, or 2CV-0306 Steam dump valve (depending on which is beingused) will fail closed impacting the cooldown rate. The ATC will notice that the cooldown hasstopped and adjust the cooldown rate to ensure the steam generator is isolated within the 30 minuterequired time. Once T hot is less than 535 degrees F, the BOP should isolate 'B' steam generator.[Site OE: CR-ANO 2010-558, CR-ANO 2009-3780, CR-ANO 2008-1190, Steam dumpvalve failure to open.]