ML101270492

WT-2009-10-Draft Outlines

ML101270492
Person / Time
Site:	Waterford
Issue date:	10/05/2009
From:	Brian Larson NRC Region 4
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Download: ML101270492 (44)

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ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 Date of Exam: October 14, 2009 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 1 4 3 4 3 18 4 2 6 Emergency &

2 1 2 2 1 2 1 9 2 2 4 Abnormal N/A N/A Plant 4 3 6 4 6 4 27 6 4 10 Evolutions Tier Totals 1 3 3 2 4 2 1 2 3 2 2 4 28 3 2 5 2.

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 2 3 2 2 2 1 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

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

7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

2009 WF3 NRC Written Outline, rev 0 ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 RO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 CE/E02 Reactor Trip - Stabilization - X EK1.3 Knowledge of the operational implications of the following 3.0 1 Recovery / 1 concepts as they apply to the (Reactor Trip Recovery):

Annunciators and conditions indicating signals, and remedial actions associated with the (Reactor Trip Recovery).

000008 Pressurizer Vapor Space X AA2.12 Ability to determine and interpret the following as they apply 3.4 2 Accident / 3 to a Pressurizer Vapor Space Accident: PZR level indications.

000009 Small Break LOCA / 3 X EA2.34 Ability to determine or interpret the following as they apply to 3.6 3 a small break LOCA: Conditions for throttling or stopping HPI 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 X AA1.22 Ability to operate and / or monitor the following as they apply 4.0 4 to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

RCP seal failure/malfunction Knowledge of the reasons for the following responses as 000022 Loss of Rx Coolant Makeup / 2 X AK3.04 3.2 5 they apply to the Loss of Reactor Coolant Makeup: Isolating letdown 000025 Loss of RHR System / 4 X AA2.05 Ability to determine and interpret the following as they apply 3.1 6 to the Loss of Residual Heat Removal System: Limitations on LPI flow and temperature rates of change Conduct of Operations: Ability to locate and operate 000026 Loss of Component Cooling X G2.1.30 4.4 7 Water / 8 components, including local controls.

000027 Pressurizer Pressure Control X AK3.03 Knowledge of the reasons for the following responses as 3.7 8 System Malfunction / 3 they apply to the Pressurizer Pressure Control Malfunctions:

Actions contained in EOP for PZR PCS malfunction 000029 ATWS / 1 X EA1.11 Ability to operate and monitor the following as they apply to a 3.9 9 ATWS: Manual opening of the CRDS breakers 2009 WF3 NRC Written Outline, rev 0 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 000038 Steam Gen. Tube Rupture / 3 X EK3.05 Knowledge of the reasons for the following responses as the 4.0 10 apply to the SGTR: Normal operating precautions to preclude or minimize SGTR 000040 (CE/E05) Steam Line Rupture - X EA2.4 Ability to determine and interpret the following as they apply 4.5 11 Excessive Heat Transfer / 4 to the Steam Line Rupture: Conditions requiring ESFAS initiation CE/E06 Loss of Main Feedwater / 4 X EK3.2 Knowledge of the reasons for the following responses as 3.2 12 they apply to the (Loss of Main Feedwater): Normal, abnormal and emergency operating procedures associated with (Loss of Feedwater).

000055 Station Blackout / 6 X EK1.01 Knowledge of the operational implications of the following 3.3 13 concepts as they apply to the Station Blackout : Effect of battery discharge rates on capacity 000056 Loss of Off-site Power / 6 X AA1.07 Ability to operate and / or monitor the following as they apply 3.2 14 to the Loss of Offsite Power: Service water pump.

Equipment Control: Ability to apply Technical Specifications for 000057 Loss of Vital AC Inst. Bus / 6 X G2.2.40 3.4 15 a system.

000058 Loss of DC Power / 6 X AK1.01 Knowledge of the operational implications of the following 2.8 16 concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation.

000062 Loss of Nuclear Svc Water / 4 Emergency Procedures / Plan: Ability to recognize abnormal 000065 Loss of Instrument Air / 8 X G2.4.4 indications for system operating parameters that are entry-level 4.5 17 conditions for emergency and abnormal operating procedures.

000077 Generator Voltage and Electric X AK2.03 Knowledge of the interrelations between Generator Voltage 3.0 18 Grid Disturbances / 6 and Electric Grid Disturbances and the following: Sensors, detectors, indicators.

K/A Category Totals: 3 1 4 3 4 3 Group Point Total: 18 2009 WF3 NRC Written Outline, rev 0 ES-401 3 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 000003 Dropped Control Rod / 1 X AA1.07 Ability to operate and / or monitor the following as they 3.8 19 apply to the Dropped Control Rod: In-core and ex-core instrumentation 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 X AK2.03 Knowledge of the interrelations between the 2.6 20 Pressurizer Level Control Malfunctions and the following: Controllers and positioners 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 Fuel Handling Accident / 8 X AK1.02 Knowledge of the operational implications of the 3.4 21 following concepts as they apply to Fuel Handling Incidents : SDM 000037 Steam Generator Tube Leak / 3 G2.4.47 Emergency Procedures / Plan: Ability to diagnose and 4.2 22 X

recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

000051 Loss of Condenser Vacuum / 4 X AA2.02 Ability to determine and interpret the following as they 3.9 23 apply to the Loss of Condenser Vacuum: Conditions requiring reactor and/or turbine trip 000059 Accidental Liquid RadWaste Rel. / 9 X AK3.04 Knowledge of the reasons for the following responses 3.8 24 as they apply to the Accidental Liquid Radwaste Release: Actions contained in EOP for accidental liquid radioactive-waste release.

000060 Accidental Gaseous Radwaste Rel. / 9 2009 WF3 NRC Written Outline, rev 0 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 000061 ARM System Alarms / 7 X AA2.06 Ability to determine and interpret the following as they 3.2 25 apply to the Area Radiation Monitoring (ARM) System Alarms: Required actions if alarm channel is out of service 000067 Plant Fire On-site / 8 000068 Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 X AK2.03 Knowledge of the interrelations between the Loss of 2.8 26 Containment Integrity and the following: Personnel access hatch and emergency access hatch 000074 Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 CE/A13 Natural Circ. / 4 CE/A11 RCS Overcooling - PTS / 4 AK3.2 Knowledge of the reasons for the following responses 2.9 27 X

as they apply to the (RCS Overcooling): Normal, abnormal and emergency operating procedures associated with (RCS Overcooling).

CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 1 2 2 1 2 1 Group Point Total: 9 2009 WF3 NRC Written Outline, rev 0 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump X K4.04 Knowledge of RCPS design feature(s) and/or 2.8 28 interlock(s) which provide for the following: Adequate cooling of RCP motor and seals Conduct of Operations: Ability to explain and apply 003 Reactor Coolant Pump X G2.1.32 3.8 29 system limits and precautions.

004 Chemical and Volume X K5.04 Knowledge of the operational implications of the 2.8 30 Control following concepts as they apply to the CVCS Reason for hydrogen cover gas in VCT (oxygen scavenge) 005 Residual Heat Removal X K2.03 Knowledge of bus power supplies to the following: 2.7 31 RCS pressure boundary motor-operated valves 006 Emergency Core Cooling X K5.06 Knowledge of the operational implications of the 3.5 32 following concepts as they apply to ECCS:

Relationship between ECCS flow and RCS pressure 007 Pressurizer Relief/Quench X A1.02 Ability to predict and/or monitor changes in 2.7 33 Tank parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining quench tank pressure 008 Component Cooling Water X K1.05 Knowledge of the physical connections and/or cause- 3.0 34 effect relationships between the CCWS and the following systems: Sources of makeup water 008 Component Cooling Water X A3.08 Ability to monitor automatic operation of the CCWS, 3.6 35 including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal 010 Pressurizer Pressure Control X K4.03 Knowledge of PZR PCS design feature(s) and/or 3.8 36 interlock(s) which provide for the following: Over pressure control 2009 WF3 NRC Written Outline, rev 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 Equipment Control: Knowledge of less than or equal to 012 Reactor Protection X G2.2.39 3.9 37 one hour Technical Specification action statements for systems.

013 Engineered Safety Features X K2.01 Knowledge of bus power supplies to the following: 3.6 38 Actuation ESFAS/safeguards equipment control 013 Engineered Safety Features X A4.02 Ability to manually operate and/or monitor in the 4.3 39 Actuation control room: Reset of ESFAS channels Knowledge of the effect that a loss or malfunction of 022 Containment Cooling X K3.02 3.0 40 the CCS will have on the following: Containment instrumentation readings 026 Containment Spray X K2.01 Knowledge of bus power supplies to the following: 3.4 41 Containment spray pumps 026 Containment Spray X A1.02 Ability to predict and/or monitor changes in 3.6 42 parameters (to prevent exceeding design limits) associated with operating the CSS controls including:

Containment temperature 039 Main and Reheat Steam X K1.08 Knowledge of the physical connections and/or cause- 2.7 43 effect relationships between the MRSS and the following systems: MFW Ability to manually operate and monitor in the control 059 Main Feedwater X A4.08 3.0 44 room: Feed regulating valve controller Emergency Procedures / Plan: Knowledge of abnormal 059 Main Feedwater X G2.4.11 4.0 45 condition procedures.

061 Auxiliary/Emergency X K4.06 Knowledge of AFW design feature(s) and/or 4.0 46 Feedwater interlock(s) which provide for the following: AFW startup permissives 2009 WF3 NRC Written Outline, rev 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 061 Auxiliary/Emergency X A2.05 Ability to (a) predict the impacts of the following 3.1 47 Feedwater malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Automatic control malfunction 062 AC Electrical Distribution X A2.12 Ability to (a) predict the impacts of the following 3.2 48 malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Restoration of power to a system with a fault on it 063 DC Electrical Distribution X K3.02 Knowledge of the effect that a loss or malfunction of 3.7 49 the DC electrical system will have on the following:

Components using DC control power 064 Emergency Diesel Generator X K6.08 Knowledge of the effect of a loss or malfunction of 3.2 50 the following will have on the ED/G system: Fuel oil storage tanks Emergency Procedures / Plan: Ability to recognize 064 Emergency Diesel Generator X G2.4.4 4.5 51 abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

073 Process Radiation X A2.02 Ability to (a) predict the impacts of the following 2.7 52 Monitoring malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure 076 Service Water X K4.02 Knowledge of SWS design feature(s) and/or 2.9 53 interlock(s) which provide for the following: Automatic start features associated with SWS pump controls 2009 WF3 NRC Written Outline, rev 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 078 Instrument Air X A3.01 Ability to monitor automatic operation of the IAS, 3.1 54 including: Air pressure 103 Containment X K1.03 Knowledge of the physical connections and/or cause 3.1 55 effect relationships between the containment system and the following systems: Shield building vent system K/A Category Point Totals: 3 3 2 4 2 1 2 3 2 2 4 Group Point Total: 28 2009 WF3 NRC Written Outline, rev 0 ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant X K6.12 Knowledge of the effect or a loss or malfunction on the 3.0 56 following RCS components: Code Safety valves 011 Pressurizer Level Control X A4.05 Ability to manually operate and/or monitor in the 3.2 57 control room: Letdown flow controller 014 Rod Position Indication X A2.04 Ability to (a) predict the impacts of the following 3.4 63 malfunctions or operations on the RPIS; and (b) based on those on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Misaligned Rod 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor X A1.01 Ability to predict and/or monitor changes in parameters 3.7 58 (to prevent exceeding design limits) associated with operating the ITM system controls including: Core exit temperature 027 Containment Iodine Removal X K1.01 Knowledge of the physical connections and/or cause 3.4 59 effect relationships between the CIRS and the following systems: CSS 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X K4.01 Knowledge of design feature(s) and/or interlock(s) 2.9 60 which provide for the following: Maintenance of spent fuel level 2009 WF3 NRC Written Outline, rev 0 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 4 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine X K3.04 Knowledge of the effect that a loss or malfunction of 3.5 61 Bypass Control the SDS will have on the following: Reactor power 045 Main Turbine Generator X A1.05 Ability to predict and/or monitor changes in parameters 3.8 62 (to prevent exceeding design limits) associated with operating the MT/G system controls including:

Expected response of primary plant parameters (temperature and pressure) following T/G trip 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal X K1.06 Knowledge of the physical connections and/or cause 3.1 64 effect relationships between the Waste Gas Disposal System and the following systems: ARM and PRM systems 072 Area Radiation Monitoring X A2.01 Ability to (a) predict the impacts of the following 2.7 65 malfunctions or operations on the ARM system- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 2 0 1 1 0 1 2 2 0 1 0 Group Point Total: 10 2009 WF3 NRC Written Outline, rev 0 ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 SRO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000007 (CE/E02) Reactor Trip - X EA2.1 Ability to determine and interpret the following as they apply 3.7 S1 Stabilization - Recovery / 1 to the Reactor Trip Recovery: Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 Conduct of Operations: Ability to evaluate plant performance 000011 Large Break LOCA / 3 X G2.1.7 4.7 S2 and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 X AA2.07 Ability to determine and interpret the following as they apply 3.7 S3 to the Loss of Residual Heat Removal System: Pump cavitation 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 000038 Steam Gen. Tube Rupture / 3 CE/E05 Steam Line Rupture - Excessive X EA2.2 Ability to determine and interpret the following as they apply 4.2 S4 Heat Transfer / 4 to the (Excess Steam Demand): Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

000054 (CE/E06) Loss of Main Feedwater / 4 2009 WF3 NRC Written Outline, rev 0 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 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 Emergency Procedures / Plan: Ability to prioritize and interpret 000057 Loss of Vital AC Inst. Bus / 6 X G2.4.45 4.3 S5 the significance of each annunciator or alarm.

000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 000077 Generator Voltage and Electric X AA2.05 Ability to determine and interpret the following as they apply 3.8 S6 Grid Disturbances / 6 to Generator Voltage and Electric Grid Disturbances:

Operational status of offsite circuit.

K/A Category Totals: 4 2 Group Point Total: 6 2009 WF3 NRC Written Outline, rev 0 ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 SRO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 X AA2.03 Ability to determine and interpret the following as 4.4 S7 they apply to the Inoperable / Stuck Control Rod:

Required actions if more than one rod is stuck or inoperable.

000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 G2.2.38 Equipment Control: Knowledge of conditions and 4.5 S8 X

limitations in the facility license.

000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 Control Room Evac. / 8 000069 Loss of CTMT Integrity / 5 X AA2.01 Ability to determine and interpret the following as 4.3 S9 they apply to the Loss of Containment Integrity: Loss of containment integrity.

000074 Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 2009 WF3 NRC Written Outline, rev 0 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 CE/A13 Natural Circ. / 4 CE/A11 RCS Overcooling - PTS / 4 X G2.4.18 Emergency Procedures / Plan: Knowledge of the 4.0 S10 specific bases for EOPs.

CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 2 2 Group Point Total: 4 2009 WF3 NRC Written Outline, rev 0 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump 004 Chemical and Volume X A2.25 Ability to (a) predict the impacts of the following 4.3 S11 Control malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Uncontrolled boration or dilution 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water Conduct of Operations: Ability to interpret reference 010 Pressurizer Pressure Control X G2.1.25 4.2 S12 materials, such as graphs, curves, tables, etc.

012 Reactor Protection X A2.01 Ability to (a) predict the impacts of the following 3.6 S13 malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty bistable operation 013 Engineered Safety Features Actuation 022 Containment Cooling 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 2009 WF3 NRC Written Outline, rev 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 061 Auxiliary/Emergency X A2.04 Ability to (a) predict the impacts of the following 3.8 S14 Feedwater malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: pump failure or improper operation 062 AC Electrical Distribution 063 DC Electrical Distribution Emergency Procedures / Plan: Ability to prioritize and 064 Emergency Diesel Generator X G2.4.45 4.3 S15 interpret the significance of each annunciator or alarm.

073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals: 3 2 Group Point Total: 5 2009 WF3 NRC Written Outline, rev 0 ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation Equipment Control: Knowledge of limiting conditions for 017 In-core Temperature Monitor X G2.2..22 4.7 S16 operations and safety limits.

027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment X A2.01 Ability to (a) predict the impacts of the following 4.4 S17 malfunctions or operations on the Fuel Handling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Dropped fuel element 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 2009 WF3 NRC Written Outline, rev 0 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 4 056 Condensate 068 Liquid Radwaste X A2.04 Ability to (a) predict the impacts of the following 3.3 S18 malfunctions or operations on the Liquid Radwaste System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure of automatic isolation 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 2 1 Group Point Total: 3 2009 WF3 NRC Written Outline, rev 0 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 Date of Exam: October 14, 2009 Category K/A # Topic RO SRO-Only IR # IR #

2.1.25 Ability to interpret reference materials, such as graphs, curves, 3.9 66

1. tables, etc.

Conduct 2.1.32 of Operations Ability to explain and apply system limits and precautions. 3.8 67 2.1.43 Ability to use procedures to determine the effects on reactivity of 4.3 S19 plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

2.1.6 Ability to manage the control room crew during plant transients. 4.8 S20 Subtotal 2 2 2.2.41 Ability to obtain and interpret station electrical and mechanical 3.5 68 drawings.

2. 2.2.14 Equipment Knowledge of the process for controlling equipment configuration 3.9 69 Control or status.

2.2.38 Knowledge of conditions and limitations in the facility license. 3.6 70 2.2.20 Knowledge of the process for managing troubleshooting activities. 3.8 S21 2.2.42 Ability to recognize system parameters that are entry-level 4.6 S22 conditions for Technical Specifications.

Subtotal 3 2 2009 WF3 NRC Written Outline, rev 0 2.3.14 Knowledge of radiation or contamination hazards that may arise 3.4 71 during normal, abnormal, or emergency conditions or activities.

3. 2.3.7 Radiation Ability to comply with radiation work permit requirements during 3.5 72 Control normal or abnormal conditions.

2.3.13 Knowledge of radiological safety procedures pertaining to licensed 3.8 S23 operator l duties, such as response to radiation monitor alarms, containment entry l requirements, fuel handling responsibilities, access to locked high-radiation l areas, aligning filters, etc.

2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation 3.1 S24 monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Subtotal 2 2 2.4.4 Ability to recognize abnormal indications for system operating 4.5 73

4. parameters that are entry-level conditions for emergency and Emergency abnormal operating procedures.

Procedures /

Plan 2.4.2 Knowledge of system set points, interlocks and automatic actions 4.5 74 associated with EOP entry conditions.

2.4.30 Knowledge of events related to system operation/status that must 2.7 75 be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

2.4.38 Ability to take actions called for in the facility emergency plan, 4.4 S25 including supporting or acting as emergency coordinator if required.

Subtotal 3 1 Tier 3 Point Total 10 7 2009 WF3 NRC Written Outline, rev 0 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Selected K/A Reason for Rejection Group RO 029 ATWS K/A selected directed questioning on the ability to operate and monitor the Rod Control Function Switch as it applies to ATWS. Waterford does not have a Rod 1/1 A1.10 Control Function Switch, and a similar switch, the CEA Mode Select Switch, has no interrelation with ATWS. Randomly selected another K/A under A1 from those items in section A1 that were applicable to Waterford for an ATWS condition.

K/A A1.11 was drawn.

RO 038 S/G Tube Rupture K/A selected directed questioning on the reasons for automatic actions associated with high activity in the S/G sample lines as it applies to SGTR. Waterford does 1/1 K3.03 not have any automatic actions occur associated with high activity in the S/G sample lines. Randomly selected another K/A under K3 from those items in section K3 that were applicable to Waterford for a S/G Tube Rupture.

K/A K3.05 was drawn.

RO 065 Loss of Instrument Air K/A selected directed questioning on the generic Emergency Procedures/Plan:

Knowledge of SRO responsibilities in the Emergency Plan. Waterford does not 1/1 G2.4.40 have any SRO responsibilities related to loss of Instrument Air in the Emergency Plan. Additionally, this question is contained in the RO section of the exam and the K/A pertains to SRO responsibilities. Randomly selected another generic K/A from section 2.4 from those items in section 2.4 that were applicable to Loss of Instrument Air.

K/A G2.4.4 was drawn.

K/A selected directed questioning on the knowledge of the effect that a loss or RO 063 DC Electrical malfunction of the DC electrical system will have on the EDGs. Waterford is 2/1 K3.01 unable to write a question of adequate difficulty for this K/A. Randomly selected another K/A under K3 from those items in section K3 that were applicable to Waterford for a loss or malfunction of DC electrical power.

K/A K3.02 was drawn.

2009 WF3 NRC Written Outline, rev 0 Tier / Randomly Selected K/A Reason for Rejection Group RO 056 Condensate K/A selected directed questioning on the ability to (a) predict the impacts of the A2.04 following malfunctions or operations on the Condensate System; and (b) based on 2/2 those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations for Loss of condensate pumps. This K/A was selected for the reactor operator portion of the written exam, but NO procedural actions address a loss of condensate pump(s). Randomly selected another system for A2 based upon NO other available K/As for 056 system.

K/A 014 A2.04 RO Conduct of Operations K/A selected directed questioning on the ability to direct personnel activities inside the control room. This K/A was selected for the reactor operator portion of the 3/1 2.1.9 written exam, but does not apply to reactor operator duties. Randomly selected another generic K/A from section 2.1.

K/A 2.1.25 was drawn.

END of RO Section 2009 WF3 NRC Written Outline, rev 0 K/A selected directed questioning on the ability to determine and interpret the SRO 077 Electrical Grid following as they apply to Generator Voltage and Electric Grid Disturbances:

Disturbance 1/1 Generator Frequency limitations. Waterford does not monitor Frequency as a AA2.06 procedural action but rather requires the operator to monitor Volts, MVARs, and MW parameters. Randomly selected another K/A under A2 from those items in section A2 that were applicable to Waterford for a Grid disturbance condition.

K/A AA2.05 was drawn.

SRO 026 Containment Spray K/A selected directed questioning on the ability to predict the impact of a malfunction or operations on the CSS and (b) based on those predictions, use 2/1 A2.07 procedures to correct, control, or mitigate the consequences of those malfunctions or operations. This item is being addressed in the operating test section which requires selecting a new system.

Topic 012, Reactor Protection, was randomly drawn as a replacement.

K/A 012 A2.01 SRO 034 Fuel Handling K/A selected directed questioning on the ability to (a) predict the impact of the Equipment following malfunctions or operations on the Fuel Handling System; and (b) based 2/2 on those predictions, use procedures to correct, control, or mitigate the A2.03 consequences of those malfunctions or operations: Mispositioned fuel element.

Waterford does not have a comprehensive procedural action but rather requires the operator to stop and contact Engineering for an evaluation/recommendation of corrective action. Randomly selected another K/A under A2 from those items in section A2 that were applicable to Waterford for a Fuel Handling condition.

K/A A2.01 was drawn.

END of SRO Section 2009 WF3 NRC Written Outline, rev 0 ES-301 Administrative Topics Outline Form ES-301-1 Facility: WATERFORD 3 Date of Examination: October 5, 2009 Examination Level: RO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

2.1.23, Ability to perform specific system and A1 S, D integrated plant procedures during all modes of plant operation.

Conduct of Operations Calculation of Primary Makeup Water Volume for Direct Dilution or VCT Dilute Makeup Mode in accordance with OP-002-005, Attachment 11.7.

Perform in conjunction with Control Room System JPM S2.

A2 R, M 2.1.18, Ability to make accurate, clear, and concise Conduct of Operations logs, records, status boards, and reports.

Complete OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data.

A3 R, N 2.2.12, Knowledge of surveillance procedures Equipment Control Complete surveillance OP-903-008, Reactor Coolant System Isolation Leakage Test, Attachment 10.11 for SI-329 A.

A4 R, N 2.3.4, Knowledge of radiation exposure limits under Radiation Control normal and emergency conditions.

Calculate stay time to perform a tagout verification in the Regen Heat Exchanger Room. Room dose rate &

operators yearly dose provided.

Emergency Plan Not selected NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 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 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: WATERFORD 3 Date of Examination: October 5, 2009 Examination Level: SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

2.1.23, Ability to perform specific system and A5 R, N integrated plant procedures during all modes of plant operation.

Conduct of Operations Verify Core Protection Calculator, Plant Protection System, and Calorimetric power are within limits during power ascension in accordance with OP-010-004 Power Operations, and OP-903-001, Technical Specification Surveillance Logs.

2.1.20, Ability to interpret and execute procedure A6 R, M steps Conduct of Operations Perform SM/CRS review OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction, Attachments 1, 2, and 3 following a PMC failure.

2.2.40, Ability to apply Technical Specifications for a A7 R, N system.

Equipment Control Review surveillance OP-903-008, Reactor Coolant System Isolation Leakage Test, Attachment 10.11 for SI-329 A.

2.3.4, Knowledge of radiation exposure limits under A8 R, N normal and emergency conditions.

Radiation Control Calculate dose and assign non-licensed operators to vent Safety Injection piping in Safeguards Room A.

Given dose rate with and without shielding installed, time to install shielding, and job completion time using 1 operator or using 2 operators, determine proper job assignment.

2.4.41, Knowledge of the emergency action level A9 S, M thresholds and classifications.

Emergency Plan Determine appropriate Emergency Plan EAL immediately following Scenario 1.

NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 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 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: October 5, 2009 Exam Level (circle one): RO Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1 001 Control Rod Drive; ATC Operator Immediate Operator Actions A, S, D 1 on 2 Dropped CEAs from OP-901-102, CEA or CEDMCS Malfunction Fault: The first and second reactor trip options do not function, requires performance of the 2nd reactor trip contingency from EOP OP-902-000, Standard Post Trip Actions.

S2 004 Chemical and Volume Control; VCT Makeup Using the Dilute A, L, S, M 2 makeup Mode Admin JPM A1 is performed in conjunction with this JPM as step 6.9.2 of OP-002-005, Chemical and Volume Control.

Fault: PMU-144, Primary Makeup Water Control Valve, will not auto close when Primary Makeup Water Batch Counter counts down to zero.

S3 006 Emergency Core Cooling System; BOP Operator Actions on L, P, S, EN, D 3 RAS This is a time critical task performed in EOP OP-902-002, Loss of Coolant Accident Recovery Procedure.

S4 005 Shutdown Cooling System / 0025 E/APE Loss of Shutdown A, L, S, M 4-P Cooling; Place Shutdown Cooling Train B in Service Fault: After LPSI Pump B is running, SI-405 B will fail closed, requiring the operator to take immediate operator actions IAW OP-903-130, Shutdown Cooling Malfunction, to secure LPSI Pump B.

S5 022 Containment Cooling System; Perform OP-903-037, S, D 5 Containment Cooling Fans Operability Verification S6 062 AC Electrical Distribution System, Energize 4.16 KV Safety Bus L, S, D 6 from Offsite Power This task will re-energize the 3A Bus with EDG A powering the 3A Bus using OP-902-009, Standard Appendices, Attachment 12-B.

S7. 012 Reactor Protection System; Remove Reactor Trip on Turbine A, S, M 7 Trip from Service using OP-004-015, Reactor Power Cutback System, and place Reactor Power Cutback in Service.

Fault: When Reactor Power Cutback is placed in service, a Reactor Power Cutback will occur. The student will then need to take the immediate operator actions for Reactor Power Cutback.

S8 029 Containment Purge System; Perform Containment Purge S, D 8 Initiation with RAB Normal Ventilation using OP-002-010, Reactor Auxiliary Building HVAC and Containment Purge 1 Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Emergency Feedwater; Transfer EFW Pump Suctions to Wet E, L, R, D 4-S Cooling Tower after Condensate Storage Pool Depletion using EOP OP-902-009, Standard Appendices, Attachment 10 P2 064 Electrical Diesel Generators, Reset EDG A following an E, L, R, D 6 overspeed trip with a LOOP.

Reset is accomplished with OP-009-002, Emergency Diesel Generator, Section 8.8.

P3 006 Emergency Core Cooling System (ECCS), Align HPSI Pump AB A, R, M 2 for performance of OP-903-030, Safety Injection Pump Operability Verification.

Fault: Reach rod for SI-208 A will bottom out during valve alignment, requiring contingencies of EN-OP-115, Conduct of Operations.

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature - / - / 1 (control room system)

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

(R)CA 1/ 1/ 1 (S)imulator 2 Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: October 5, 2009 Exam Level (circle one): SRO - I Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1 001 Control Rod Drive; ATC Operator Immediate Operator Actions A, S, D 1 on 2 Dropped CEAs from OP-901-102, CEA or CEDMCS Malfunction Fault: The first and second reactor trip options do not function, requires performance of the 2nd reactor trip contingency from EOP OP-902-000, Standard Post Trip Actions.

S2 S3 006 Emergency Core Cooling System; BOP Operator Actions on L, P, S, EN, D 3 RAS This is a time critical task performed in EOP OP-902-002, Loss of Coolant Accident Recovery Procedure.

S4 005 Shutdown Cooling System / 0025 E/APE Loss of Shutdown A, L, S, M 4-P Cooling; Place Shutdown Cooling Train B in Service Fault: After LPSI Pump B is running, SI-405 B will fail closed, requiring the operator to take immediate operator actions IAW OP-903-130, Shutdown Cooling Malfunction, to secure LPSI Pump B.

S5 022 Containment Cooling System; Perform OP-903-037, S, D 5 Containment Cooling Fans Operability Verification S6 062 AC Electrical Distribution System, Energize 4.16 KV Safety Bus L, S, D 6 from Offsite Power This task will re-energize the 3A Bus with EDG A powering the 3A Bus using OP-902-009, Standard Appendices, Attachment 12-B.

S7. 012 Reactor Protection System; Remove Reactor Trip on Turbine A, S, M 7 Trip from Service using OP-004-015, Reactor Power Cutback System, and place Reactor Power Cutback in Service.

Fault: When Reactor Power Cutback is placed in service, a Reactor Power Cutback will occur. The student will then need to take the immediate operator actions for Reactor Power Cutback.

S8 029 Containment Purge System; Perform Containment Purge S, D 8 Initiation with RAB Normal Ventilation using OP-002-010, Reactor Auxiliary Building HVAC and Containment Purge 3 Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Emergency Feedwater; Transfer EFW Pump Suctions to Wet E, L, R, D 4-S Cooling Tower after Condensate Storage Pool Depletion using EOP OP-902-009, Standard Appendices, Attachment 10 P2 064 Electrical Diesel Generators, Reset EDG A following an E, L, R, D 6 overspeed trip with a LOOP.

Reset is accomplished with OP-009-002, Emergency Diesel Generator, Section 8.8.

P3 006 Emergency Core Cooling System (ECCS), Align HPSI Pump AB A, R, M 2 for performance of OP-903-030, Safety Injection Pump Operability Verification.

Fault: Reach rod for SI-208 A will bottom out during valve alignment, requiring contingencies of EN-OP-115, Conduct of Operations.

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature - / - / 1 (control room system)

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

(R)CA 1/ 1/ 1 (S)imulator 4 Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: October 5, 2009 Exam Level (circle one): SRO - U Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1 S2 S3 006 Emergency Core Cooling System; BOP Operator Actions on L, P, S, EN, D 3 RAS This is a time critical task performed in EOP OP-902-002, Loss of Coolant Accident Recovery Procedure.

S4 005 Shutdown Cooling System / 0025 E/APE Loss of Shutdown A, L, S, M 4-P Cooling; Place Shutdown Cooling Train B in Service Fault: After LPSI Pump B is running, SI-405 B will fail closed, requiring the operator to take immediate operator actions IAW OP-903-130, Shutdown Cooling Malfunction, to secure LPSI Pump B.

S5 S6 S7. 012 Reactor Protection System; Remove Reactor Trip on Turbine A, S, M 7 Trip from Service using OP-004-015, Reactor Power Cutback System, and place Reactor Power Cutback in Service.

Fault: When Reactor Power Cutback is placed in service, a Reactor Power Cutback will occur. The student will then need to take the immediate operator actions for Reactor Power Cutback.

S8 5 Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 P2 064 Electrical Diesel Generators, Reset EDG A following an E, L, R, D 6 overspeed trip with a LOOP.

Reset is accomplished with OP-009-002, Emergency Diesel Generator, Section 8.8.

P3 006 Emergency Core Cooling System (ECCS), Align HPSI Pump AB A, R, M 2 for performance of OP-903-030, Safety Injection Pump Operability Verification.

Fault: Reach rod for SI-208 A will bottom out during valve alignment, requiring contingencies of EN-OP-115, Conduct of Operations.

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature - / - / 1 (control room system)

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

(R)CA 1/ 1/ 1 (S)imulator 6 Revision 0

ES-301 Simulator Scenario Quality Checklist Form ES-301-5 Facility: Waterford 3 Date of Exam: October 5, 2009 Operating Test No. 1 A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION I

C S A B S A B S A B S A B A M

A T R T O R T O R T O R T O L U N Y O C P O C P O C P O C P M(*)

T P E R I U RX 4 1 1 1 0 NOR 1 1 1 1 1 RO 1 I/C 1,6 3,4,7 5 4 4 2 MAJ 5,7 6 3 2 2 1 TS 0 0 2 2 RX 4 1 1 1 0 NOR 4 1 1 1 1 RO 2 I/C 1,3, 1,6 6 4 4 2 6,7 MAJ 5 5,7 3 2 2 1 TS 0 0 2 2 RX 0 1 1 0 NOR 4 4 2 1 1 1 RO 3 I/C 1,3, 2,3,8 2,5,8 10 4 4 2 6,7 MAJ 5 5,7 6 4 2 2 1 TS 0 0 2 2 RX 4 1 1 1 0 RO 4 NOR 4 1 1 1 1 I/C 2,3 2,3,8 5 4 4 2 MAJ 5 5,7 3 2 2 1 TS 0 0 2 2 RX 0 1 1 0 NOR 4 4 2 1 1 1 RO 5 I/C 1,3, 2,3,8 2,5,8 10 4 4 2 6,7 MAJ 5 5,7 6 4 2 2 1 TS 0 0 2 2

A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION I

C S A B S A B S A B S A B A M

A T R T O R T O R T O R T O L U N Y O C P O C P O C P O C P M(*)

T P E R I U RX 4 1 1 1 0 NOR 4 1 2 1 1 1 SRO-I 1 I/C 1,2,3, 1,6 2,3,4, 13 4 4 2 6,7 5,7,8 MAJ 5 5,7 6 4 2 2 1 TS 1,2,3 1,3 5 0 2 2 RX 4 1 1 1 0 NOR 4 1 2 1 1 1 SRO-I 2 I/C 2,3 1,2,3, 3,4,7 10 4 4 2 6,8 MAJ 5 5,7 6 4 2 2 1 TS 1,2 2 0 2 2 RX 4 1 1 1 0 NOR 4 1 1 1 1 SRO-I 3 I/C 2,3 1,2,3, 7 4 4 2 6,8 MAJ 5 5,7 3 2 2 1 TS 1,2 2 0 2 2 RX 0 1 1 0 NOR 4 1 2 1 1 1 SRO-U 1 I/C 1,2,3, 2,3,4, 11 4 4 2 6,7 5,7,8 MAJ 5 6 2 2 2 1 TS 1,2,3 1,3 5 0 2 2 RX 0 1 1 0 NOR 4 4 2 1 1 1 SRO-U 2 I/C 1,2,3, 1,2,3, 10 4 4 2 6,7 6,8 MAJ 5 5,7 3 2 2 1 TS 1,2,3 1,2 5 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 1 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 100%, MOC, AB buses aligned to B side.

Protected Train is B Turnover: Maintain 100 % power Continue with Surveillance OP-903-094, section 7.20 Event Malf. No. Event Event No. Type* Description 1 Di08A04S08-1 C - BOP Perform surveillance OP-903-094, section 7.20.

TS - SRO BD-103 B fail to close.

2 RC15-A1 I - ATC Pressurizer level instrument RC-ILI-0110-X fails I - SRO high TS-SRO 3 CC12-E2 I - BOP Component Cooling Water Surge Tank level TS - SRO instrument CC-ILS-7013A fails low 4 FW21-A R- ATC Main Condenser leak with lowering Main N-BOP Condenser vacuum requiring a Rapid Plant Power Reduction N-SRO 5 RC23B M-All Small Break LOCA, SIAS and CIAS 3 CC12-E2 C-ATC Secure Reactor Coolant Pumps due to the C - SRO combination of event 3 and event 5.

6 SI02 C - BOP Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start 7 CS01-A C-BOP Containment Spray Pump A trip, OP-902-008, C-SRO Safety Function Recovery Procedure Alignment of LPSI Pump A to replace CS Pump A

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 1 Rev 0 Scenario Event Description NRC Scenario 1 The crew assumes the shift at 100% power with instructions to maintain 100% power.

Surveillance procedure OP-903-094, ESFAS Subgroup Relay Test - Operating, is in progress. The previous crew stopped at section 7.20, Train A Position 44, Relay K310 (BD-103 B). This crew should resume testing. The BOP will secure Blowdown flow for Steam Generator #2 and test BD-103 B, which will fail to close. The SRO should enter Tech Spec 3.6.3.

After briefing the failure, Pressurizer level instrument RC-ILI-0110X fails high. Due to the failure, Letdown flow goes to maximum flow of approximately 125 gpm and all Pressurizer Heaters energize. The SRO should enter OP-901-110, Pressurizer Level Control Malfunction. The crew should utilize sub section E1, Pressurizer Level Control Channel Malfunction. The ATC should take manual control of Pressurizer level and select the non-faulted channel. Using Tech Specs and OP-903-013, Monthly Channel Checks, the SRO should enter Tech Spec 3.3.3.5, a 7 day action requirement, and determine Tech Spec 3.3.3.6 entry is not required since QSPDS is operable and meeting the Pressurizer level channel check. SPDS indication of Pressurizer level is affected by this failure.

After the non-faulted channel is selected and Tech Specs are addressed, Component Cooling Water Surge Tank level instrument CC-ILS-7013A fails low. CCW Dry Cooling Tower A will bypass due to the failure. CCW Headers A and B will split, and CCW Loop AB supply and return from the A Header will close. The SRO should enter OP-901-510, Component Cooling Water System Malfunction. The BOP should use Attachment 1 to diagnose which instrument is failed. The crew should verify Auxiliary Component Cooling Water Pump A starts and control CCW system temperature with ACC-126 A.

CCW Train A should be declared inoperable and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action Tech Spec 3.7.3 entered as well as cascading Tech Specs. The SRO should address the need to accomplish surveillance OP-903-066, Electrical Breaker Alignment Checks, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with Tech Spec 3.8.1.1.b. They must also address the need to accomplish the requirements of Tech Spec 3.8.1.1.d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

After the crew has addressed Tech Specs, a leak in the Main Condenser develops and Main Condenser vacuum begins to drop. Off normal procedure OP-901-220, Loss of Condenser Vacuum, should be entered. Main Condenser vacuum will drop below 25 inches, requiring a rapid plant power reduction. The SRO should enter OP-901-212, Rapid Plant Power Reduction. Vacuum will drop below 25 inches but remain above 20 inches, the procedure trigger for tripping the Reactor. For the power reduction, the ATC will perform direct Boration to the RCS as well as ASI control with CEAs and Pressurizer boron equalization. The BOP will manipulate the controls to reduce Main Turbine load.

Scenario 1 Rev 0 Scenario Event Description NRC Scenario 1 Once the crew has commenced the power reduction and lowered power to ~ 90%, or at the lead examiners discretion, a small break loss of coolant accident will occur. The crew should diagnose Pressurizer level dropping with all available Charging Pumps operating, trip the Reactor, and initiate Safety Injection Actuation (SIAS) and Containment Isolation Actuation (CIAS). Because of the earlier CCW level instrument failure, all CCW flow will be lost to the Reactor Coolant Pumps; the pumps must be manually secured within 3 minutes of the loss of CCW flow. When Containment Spray is actuated, either manually or automatically, CS-125 B will fail to automatically open and will not open using the control switch. This does not create a need for action at this time, but Containment Spray flow will only be provided from Train A with CS-125 B failed closed. Low Pressure Safety Injection Pump A will fail to automatically start on SIAS, requiring the BOP operator to manually start LPSI Pump A.

After the crew completes OP-902-000, Standard Post Trip Actions and diagnoses into OP-902-002, Loss of Coolant Accident Recovery, Containment Spray Pump A will trip, resulting in no Containment Spray flow. The crew should recognize that they are not meeting the Safety Function Status Checklist of OP-902-002 and transition to OP-902-008, Safety function Recovery Procedure.

Prioritization in OP-902-008 should result in Containment Isolation being priority 1 and Containment Temperature and Pressure Control being priority 2. The crew should address Containment Isolation by overriding CS-125 A closed. The crew should address Containment Temperature and pressure Control by aligning Low Pressure Safety Injection Pump A to replace the failed Containment Spray Pump A.

The scenario can be terminated after Low Pressure Safety Injection Pump A is aligned for Containment Spray, or after the CRS gives the order to perform that alignment, at the lead examiners discretion.

The NRC examiner observing each SRO candidate will ask that candidate to determine the Emergency Plan classification. This will satisfy Administrative JPM A9. The candidate should determine the following: Site Area Emergency (FS1, Loss or Potential Loss of any 2 Barriers). The specific Barriers and EALs are RCB1, RCS Leak, for the RCS Barrier, and CNB1, Containment Pressure, Potential Loss c, for the Containment Barrier. The candidate may discuss that the classification was an Alert (FA1, Loss or Potential Loss of either Fuel Clad or RCS) due to RCB1, RCS Leak Rate, before the Containment Spray Pump A trip. If the crew aligned Low Pressure Safety Injection Pump A to replace the tripped Containment Spray Pump A in less than 15 minutes, the CRS could choose to classify as the Alert FA1.

Scenario 1 Rev 0 Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 2 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 100%, MOC, AB buses aligned to B side.

Protected Train is B Emergency Diesel Generator A is tagged out for planned maintenance.

Turnover: Maintain 100 % power Event Malf. No. Event Type* Event No. Description 1 C- ATC Swap Charging Pump using OP-002-005.

C - SRO Charging Pump B develops oil leak.

TS - SRO 2 RC22 B1 I - BOP Pressurizer narrow range safety pressure I - SRO instrument RC-IPI-0101 B fails high TS - SRO 3 SG05 B I - BOP Steam Generator #2 level instrument, I - SRO SG ILR1106, fails low.

4 TPR13, 14 R - ATC Main Generator Stator Coil Water temperature N - BOP high, normal plant downpower N - SRO 5 TU01A, D, M - All Main Turbine High Vibration and Reactor Trip R

6 RD11A-10 C-ATC 2 CEAs stuck out requiring Emergency Boration RD11A-22 C - SRO 7 ED01 M-All Loss of Off Site Power A, B, C, D 8 EG08B C- BOP EDG B fails to auto-start C - SRO

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 2 Rev 0 Scenario Event Description NRC Scenario 2 The crew assumes the shift at 100% power with instructions to maintain 100% power.

Oil filter replacement has been completed on Charging Pump B and maintenance is standing by to perform a retest. The shift manager has instructed the control room supervisor to swap Charging Pumps leaving Charging Pump B running and Charging Pump A secured and in auto.

After starting Charging Pump B, the watchstander will call and report an oil leak, recommending Charging Pump B be secured. With Charging Pump A control switch in off, the SRO should enter Tech Spec and TRM 3.1.2.4. Tech Spec 3.1.2.4 is a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action and TRM 3.1.2.4 is a 7 day action. Tech Spec 3.1.2.4 can be exited if the SRO directs the ATC to align Charging Pump AB to replace B.

After the ATC aligns Charging Pump AB or at the lead examiners direction, Pressurizer narrow range safety pressure instrument RC-IPI-0101 B fails high. After identifying the failure, the SRO should enter Tech Spec 3.3.1. The BOP should be directed to bypass the PPS bistables for High Pressurizer Pressure, Low DNBR, and High LPD within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

After the crew bypasses the appropriate bistables, Steam Generator #2 level instrument, SG ILR1106, Steam Generator 2 Downcomer Level (red pen), fails low.

The controllers for Main Feedwater Regulating Valve 2, Startup Feedwater Regulating Valve 2, and Main Feedwater Pump B transfer to manual. The crew should enter OP-901-201, Steam Generator Level Control Malfunction. No Tech Spec entries are required and no actions by the Balance of Plant operator are necessary at this time.

After the crew has completed their brief, PMC alarms will come in for Main Generator Stator Coil Water hose temperatures. The crew should enter OP-901-211, Generator Malfunction. Using Attachment 1, SCW High Temperature, the crew will determine the need to commence a normal plant shutdown in accordance with OP-010-005. Due to the earlier Steam Generator level instrument failure, the BOP operator will have to control Steam Generator level in manual for Steam Generator #2. The ATC will perform direct Boration to the RCS as well as ASI control with CEAs and Pressurizer boron equalization. The BOP will manipulate the Main Turbine controls to reduce load.

Once the crew has commenced the power reduction and lowered power to ~ 90%, or at the lead examiners discretion, high vibration alarms will come in on the Main Turbine.

Using annunciator response procedure OP-500-001, Control Room Cabinet A, and OP-901-210, Turbine Trip, the SRO should direct a Reactor trip. The crew should enter OP-902-000, Standard Post Trip Actions, and work this procedure concurrent with the Turbine Trip off normal procedure. OP-901-210 will direct breaking Main Condenser vacuum. On the Reactor Trip, 2 CEAs will stick out, requiring the ATC operator to Emergency Borate. The BOP will have to establish Feedwater Control Reactor Trip Override conditions manually on Steam Generator #2 due to the earlier level instrument failure.

Scenario 2 Rev 0 Scenario Event Description NRC Scenario 2 The SRO should direct the BOP to continue with the actions to break Main Condenser vacuum. The crew should diagnose into OP-902-006, Loss of Main Feedwater Recovery, and secure 2 Reactor Coolant Pumps. After 2 RCPs are secured and the BOP has commenced breaking vacuum, a loss of off site power occurs. Emergency Diesel Generator B will fail to auto-start on the LOOP and the BOP will be required to start EDG B. The crew will transition to OP-902-003, Loss of Off Site Power/Loss of Forced Flow Recovery procedure. During the scenario, environmental conditions will have rain occurring. After the LOOP, the high level alarms will come in for Dry Cooling Tower 1 and 2 Sumps. The CRS will direct the performance of OP-902-009, Standard Appendices, Appendix 20, Operation of DCT Sump Pumps.

The scenario can be terminated after the CRS orders the performance of OP-902-009 Appendix 20 or at the lead examiners discretion.

Scenario 2 Rev 0 Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 3 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 1.2 % Power Power ascension is being held pending Main Feedwater Pumps governor adjustment Preparations are being made to start Main Feedwater Pump A AB Bus is aligned to the A side Turnover: OP-903-052 for CVAS Train A will go late this shift. Complete OP-903-052, section 10.1.

OP-007-004, attachment 11.4 is in the field to discharge Waste Condensate Tank A.

Event Malf. No. Event Type* Event No. Description 1 DI-18A4s27-1 N - BOP During performance of OP-903-052, CVAS N - SRO Fan A will fail to start.

TS - SRO 2 AO-04A3a12-1 C - ATC Waste Condensate Tank A flow controller C - SRO LWM-IFIC-0647 fails high 3 CH08-A1 I - BOP Containment pressure Instrument I - SRO CB-IPT-6701-SMC fails high TS - SRO 4 AO-01A09A05-1 C - BOP Steam Bypass Valve MS-320 A fails open C - SRO 5 RX-14A I - ATC Pressurizer Pressure instrument RC-IPR-I - SRO 0100 X fails low 6 FW38B M - ALL Main Feedwater line break in Containment.

7 RP08G C - BOP Main Feedwater Isolation Valve #1 fails to C - SRO automatically close on MSIS.

8 CV34a1 C - ATC CVC-109 fails to auto close on CIAS.

C - SRO

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 3 Rev 0 Scenario Event Description NRC Scenario 3 The crew assumes the shift at 1.2 % power. Reactor Engineer has completed Low Power Physics Testing. I&C is making adjustments to Main Feedwater Pumps A & B governors based on vendor recommendations. The estimated time to completion is less than 60 minutes. When this is complete, Main Feedwater Pump A will be started and power ascension will commence.

Last shift, it was discovered that OP-903-052, CVAS Operability Test, will go late this shift. You have been directed to start CVAS Train A in accordance with OP-903-052.

This surveillance will have the BOP operator secure RAB Normal Supply and Normal Exhaust Fans A and start CVAS Fan A. After securing both normal ventilation fans, CVAS Fan A will fail to start. This will require entering Tech Spec 3.7.7, a 7 day action requirement. RAB Normal Supply and Normal Exhaust Fans A will have to be re-started.

After the failure of CVAS Fan A, the RCA watch will call and report that he has completed his lineup and is ready for the ATC to perform his actions to discharge Waste Condensate Tank A and is ready for the ATC to continue with step 6.10.7. When the ATC initiates flow on step 6.10.10, LWM-IFIC-0647 will fail high, raising flow in excess of 50 gpm, the discharge permit limit. The ATC should close LWM-441 and LWM-442 from CP-4 to terminate the release.

After the release is secured, Containment pressure instrument CB-IPT-6701 SMC fails high. The SRO should enter Tech Spec 3.3.1 and 3.3.2 and the BOP should bypass PPS bistables 13 and 16.

After the appropriate bistables are bypassed, Steam Bypass valve MS-320 A controller will begin to fail high. The crew should respond to the cooldown and reactivity effects by taking manual control of MS-320 A and closing it.

After MS-320 A is closed, Pressurizer pressure instrument RC-IPR-0100 X fails low.

This causes both Pressurizer heaters to energize. The SRO should enter OP-901-120, Pressurizer Pressure Malfunction. The ATC will select the non-faulted Pressurizer pressure channel.

After the Pressurizer Pressure Control Channel Y is selected, a Main Feedwater line break occurs in Containment. The Main Feedwater Isolation Valve #1 fails to automatically close on the MSIS and must be closed manually by the BOP operator.

CVC-109 fails to automatically close on the CIAS and must be manually closed by the ATC operator. The crew should enter OP-902-004, Excess Steam Demand Recovery.

Actions to address pressurized thermal shock should be taken when CET temperature and Pressurizer pressure start to rise. This can be accomplished using OP-902-009, Appendix 13 or with OP-902-004. The scenario can be terminated after PTS actions have been accomplished or at the lead examiners discretion.

The conditions in this scenario do not warrant declaration of any Emergency Plan Classification.

Scenario 3 Rev 0 Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 4 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 30% Power on RCS chemistry hold Main Feedwater Pump B is running Turnover: Hold power until directed by plant management Start ACCW Pump A for chemical mixing Event Malf. No. Event Event No. Type* Description 1 DI-33A04S36-1 C - BOP Start Auxiliary Component Cooling Water Pump C - SRO A for chemical mixing. ACC-110 A will fail to auto open.

TS - SRO 2 SG11C I - BOP Steam Generator Level #2 level instrument SG-TS - SRO ILT1123 C fails low 3 CV05B2 C - ATC Letdown Backpressure Control Valve CVC-C - SRO 123B, fails closed 4 R - ATC Direction given to raise power to < 50% using N - BOP OP-010-004, Power Operations.

N - SRO 5 RP04A3 I - BOP Inadvertent Containment Spray Actuation Signal, RP04B3 I - SRO secure Containment Spray Pumps 6 M - All Manual Reactor trip 7 DI-08A07S26-1 I - ATC CC-641 will fail to reopen, Secure all Reactor I - SRO Coolant Pumps on loss of Component Cooling Water flow 8 SG01A M - All Steam Generator #1 tube rupture ramps in over 3 minute period following reactor trip 8 C - BOP Isolate Steam Generator #1 when < 520 °F hot C - SRO leg temperature 8 C - ATC Reduce RCS pressure using Auxiliary Spray C - SRO while maintaining sub-cooled margin.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 4 Rev 0 Scenario Event Description NRC Scenario 4 The crew assumes the shift at 30% power with instructions to maintain power.

The crew is directed to start Auxiliary Component Cooling Water Pump A for basin chemical mixing. During the start, ACC-110 A will not auto open. And the BOP operator will have to take its control switch to open. The SRO should declare ACC-110 A inoperable and enter a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action for Tech Spec 3.7.3 as well as cascading Tech Specs. The SRO should address the need to accomplish surveillance OP-903-066, Electrical Breaker Alignment Checks, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with Tech Spec 3.8.1.1.b. They must also address the need to accomplish the requirements of Tech Spec 3.8.1.1.d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

After Tech Specs have been addressed, Steam Generator #2 Level instrument SG-ILT-1123 C fails low. The SRO should enter Tech Spec 3.3.1 and 3.3.2. PPS bistables for Channel C Steam Generator Level Low, Steam Generator Level High, and Steam Generator Differential Pressure for Steam Generator #2 should be placed in bypass within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

After the appropriate bistables have been bypassed, CVC-123 B, Chemical and Volume Control Backpressure Control Valve B will fail closed. Letdown flow will go to 0 gpm. The SRO should enter OP-901-112, Charging or Letdown Malfunction, and transition to sub-section E2, Letdown Malfunction. The ATC operator should place the standby Letdown Backpressure Control Valve in service and restore Letdown flow.

After Letdown has been restored, the SRO will be given direction to raise power to 50% for placing Main Feedwater Pump A in service. The SRO should use OP-010-004, Power Operations to direct the power ascension. The ATC operator will add Primary Makeup Water to the Volume Control Tank and the BOP operator will raise Main Turbine load.

At the direction of the lead examiner, an inadvertent Containment Spray Actuation Signal will be generated. The SRO should enter OP-901-504, Inadvertent ESFAS Actuation. The BOP should be directed to secure Containment Spray Pumps A and B. The BOP operator will be directed to restore CCW flow to the Reactor Coolant Pumps. CC-641, RCP Inlet Outside Isolation will fail to reopen when attempted by the BOP operator. The SRO should direct the ATC to trip the reactor and secure Reactor Coolant Pumps.

After the reactor trip, a Steam Generator tube rupture ramps in for Steam Generator #1. The crew should diagnose into OP-902-007, Steam Generator Tube Rupture Recovery. The SRO will direct a rapid RCS cooldown to < 520 °F hot leg temperature. Following the rapid cooldown, the BOP should be directed to isolate Steam Generator #1 and the ATC operator should be directed to lower RCS pressure using Auxiliary Spray within the RCS temperature and pressure limits.

The scenario can be terminated after Steam Generator #1 is isolated and the crew has taken action to reduce RCS pressure.

Scenario 4 Rev 0