ML103080010

2010-10-Final Outlines

ML103080010
Person / Time
Site:	Waterford
Issue date:	11/04/2010
From:	Apger G Operations Branch IV
To:	Entergy Operations
References
50-382/10-302, ES-401, ES-401-2 50-382/10-302
Download: ML103080010 (49)

Text

E S-401 PWR Examination Outline Form ES-401-22010 Waterford 3 Examination Outline Revision 2 1 Facility: Waterford 3 Date of Exam:October 13, 2010 Tier Group RO K/A Category PointsSRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total 1. Emergency & Abnormal Plant Evolutions 1 2 1 4 3 4 4 18 3 3 6 2 1 2 1 2 2 1 9 3 1 4 Tier Totals 3 3 5 5 6 5 27 6 4 10 2. Plant Systems 1 3 2 3 3 2 2 3 2 3 2 3 28 3 2 5 2 2 0 1 1 0 0 0 2 1 1 2 10 0 2 1 3 Tier Totals 5 2 4 4 2 2 3 4 4 3 5 38 5 3 8 3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 3 2 2 3 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each K/A category shall not be less than two). 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points. 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements. 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution. 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selecte

d. Use the RO and SRO ratings for the RO and SRO-only portions, respectively. 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

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

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

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

ES-401 2 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 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 1 K 2 K 3 A 1 A 2 GK/A Topic(s)

IR # 000007 (CE/E02) Reactor Trip - Recovery / 1 X EA2.2 Ability to determine and interpret the following as they apply to the (Reactor Trip Recovery): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

3.0 39 000008 Pressurizer Vapor Space Accident / 3 X AK1.01 Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: Thermodynamics and flow characteristics of open or leaking valves 3.2 40 000009 Small Break LOCA / 3 X EK3.12 Knowledge of the reasons for the following responses as the apply to the small break LOCA: Letdown isolation 3.4 41 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 X AA2.10 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

When to secure RCPs on loss of cooling or seal injection 3.7 42 000022 Loss of Rx Coolant Makeup / 2 X AA1.01 Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup: CVCS letdown and charging 3.4 43 000025 Loss of RHR System / 4 X2.2.38 Knowledge of conditions and limitations in the facility license.

3.6 44 000026 Loss of Component Cooling Water / 8 X AA2.01 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: Location of a leak in the CCW system 2.9 45 000027 Pressurizer Pressure Control System Malfunction / 3 X AK3.03 Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: Actions contained in EOP for PZR PCS malfunction 3.7 46 000029 ATWS / 1 X2.4.11 Knowledge of abnormal condition procedures.

4.0 47 2010 Waterford 3 Examination Outline Revision 2 3 Tier 1 / Group 1 Reactor Operator (cont.)

000038 Steam Gen. Tube Rupture / 3 X EK1.02 Knowledge of the operational implications of the following concepts as they apply to the SGTR: Leak rate vs. pressure drop 3.2 48 000040 (CE/E05) Steam Line Rupture - Excessive Steam Demand / 4 X EK2.2 Knowledge of the interrelations between the (Excess Steam Demand) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the ope ration of the facility.

3.7 49 000054 (CE/E06) Loss of Main Feedwater / 4 X 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

4.4 50 000055 Station Blackout / 6 X EA1.06 Ability to operate and monitor the following as they apply to a Station Blackout: Restoration of power with one ED/G 4.1 51 000056 Loss of Off-site Power / 6 X AA1.37 Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power: Instrument air 3.4 52 000057 Loss of Vital AC Inst. Bus

/ 6 X AA2.19 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus 4.0 53 000058 Loss of DC Power / 6 X AK3.02 Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: Actions contained in EOP for loss of dc power 4.0 54 000062 Loss of Nuclear Svc Water / 4 X2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

4.0 55 000065 Loss of Instrument Air / 8 X AK3.04 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Cross-over to backup air supplies 3.0 56 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals:

2 1 4 34 4 Group Point Total: 18 ES-401 3 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 4 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 1 K 2 K 3 A 1 A 2 G K/A Topic(s)

IR # 000001 Continuous Rod Withdrawal / 1 X AK2.08 Knowledge of the interrelations between the Continuous Rod Withdrawal and the following: Individual rod display lights and indications 3.1 57 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 Fuel Handling Accident / 8 X AA1.02 Ability to operate and / or monitor the following as they apply to the Fuel Handling Incidents: ARM system 3.1 58 000037 Steam Generator Tube Leak / 3 X AK1.01 Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak: Use of steam tables 2.9 59 000051 Loss of Condenser Vacuum / 4 X AA2.02 Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum:

Conditions requiring reactor and/or turbine trip 3.9 60 000059 Accidental Liquid RadWaste Rel.

/ 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7

2010 Waterford 3 Examination Outline Revision 2 5 Tier 1 / Group 2 Reactor Operator (cont.)

000067 Plant Fire On-site / 8 000068 Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity /

5 X AA1.01 Ability to operate and / or monitor the following as they apply to the Loss of Containment Integrity:

Isolation valves, dampers, and electropneumatic devices 3.5 63 000074 Inad. Core Cooling / 4 X EK3.04 Knowledge of the reasons for the following responses as they apply to the Inadequate Core Cooling: Tripping RCPs 3.9 62 000076 High Reactor Coolant Activity / 9 X2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

3.9 64 CE/A13 Natural Circ. / 4 CE/A11 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 X AA2.1 Ability to determine and interpret the following as they apply to the (Excess RCS Leakage):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

2.7 65 CE/E09 Functional Recovery X EK2.1 Knowledge of the interrelations between the Functional Recovery and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

3.6 61 K/A Category Point Totals:

1 2 1 2 2 1Group Point Total:

9 ES-401 4 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 6 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 RO System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G K/A Topic(s)

IR # 003 Reactor Coolant Pump X K3.01 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: RCS 3.7 1 003 Reactor Coolant Pump X K5.05 Knowledge of the operational implications of the following concepts as they apply to the RCPS: The dependency of RCS flow rates upon the number of operating RCPs 2.8 2 004 Chemical and Volume

Control X A3.14 Ability to monitor automatic operation of the CVCS, including: Letdown and charging flows 3.4 3 005 Residual Heat Removal X K1.06 Knowledge of the physical connections and/or cause effect relationships between the RHRS and the following systems:

ECCS 3.5 4 006 Emergency Core Cooling X K2.04 Knowledge of bus power supplies to the following:

ESFAS-operated valves 3.6 5 007 Pressurizer Relief/Quench Tank X A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Stuck-open PORV or code safety 3.9 6 008 Component Cooling Water X K4.09 Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following:

The "standby" feature for the CCW pumps; 2.7 7 008 Component Cooling Water X A3.04 Ability to monitor automati c operation of the CCWS, including: Requirements on and for the CCWS for different conditions of the power plant 2.9 8 2010 Waterford 3 Examination Outline Revision 2 7 Tier 2 / Group 1 Reactor Operator (cont.)

010 Pressurizer Pressure Control X A1.06 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: RCS heatup and cooldown effect on pressure 3.1 9 010 Pressurizer Pressure Control X2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

3.9 10 012 Reactor Protection X A1.01 Ability to predict and/or monitor Changes in parameters (to prevent exceeding design limits) associated with operating the RPS controls including: Trip setpoint adjustment 2.9 11 013 Engineered Safety Features Actuation X K1.12 Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems:

ED/G 4.1 12 013 Engineered Safety Features Actuation X K6.01 Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS: Sensors and detectors 2.7 13 022 Containment Cooling X A4.03 Ability to manually operate and/or monitor in the control room: Dampers in the CCS 3.2 14 026 Containment Spray X K3.02 Knowledge of the effect that a loss or malfunction of the CSS will have on the following: Recirculation spray system 4.2 15 039 Main and Reheat Steam X K1.08 Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: MFW 2.7 16 039 Main and Reheat Steam X A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunctioning steam dump 3.4 17 059 Main Feedwater X K4.05 Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Control of speed of MFW pump turbine 2.5 18 061 Auxiliary/Emergency Feedwater X K6.02 Knowledge of the effect of a loss or malfunction of the following will have on the AFW components:

Pumps 2.6 19 2010 Waterford 3 Examination Outline Revision 2 8 Tier 2 / Group 1 Reactor Operator (cont.)

061 Auxiliary/Emergency Feedwater X2.1.28 Knowledge of the purpose and function of major system components and controls.

4.1 20 062 AC Electrical Distribution X2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

4.2 21 063 DC Electrical Distribution X K2.01 Knowledge of bus power supplies to the following: Major DC loads 2.9 22 064 Emergency Diesel Generator X A3.07 Ability to monitor automati c operation of the ED/G system, including: Load sequencing 3.6 23 064 Emergency Diesel Generator X A4.01 Ability to manually operate and/or monitor in the control room: Local and remote operation of the EDG 4.0 27 073 Process Radiation Monitoring X K5.03 Knowledge of the operational implications as they apply to concepts as they apply to the PRM system:

Relationship between radiation intensity and exposure limits 2.9 24 076 Service Water X A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls

including: Reactor and turbine building closed cooling water temperatures 2.6 25 078 Instrument Air X K3.02 Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Systems having pneumatic valves and controls 3.4 26 103 Containment X K4.01 Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following: Vacuum breaker protection 3.0 28 K/A Category Point Totals:

3 2 3 3 2 2 3 2 3 2 3 Group Point Total:

28 ES-401 5 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 9 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 RO System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 GK/A Topic(s)

IR # 001 Control Rod Drive X K1.05 Knowledge of the physical connections and/or cause effect relationships between the CRDS and the following systems: NIS and RPS 4.5 29 002 Reactor Coolant 011 Pressurizer Level Control X A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of one, two or three charging pumps 3.5 37 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation X K4.03 Knowledge of NNIS design feature(s) and/or interlock(s) which provide for the following: Input to control systems 2.8 30 017 In-core Temperature

Monitor X A3.01 Ability to monitor automati c operation of the ITM system including: Indications of normal, natural, and interrupted circulation of RCS 3.6 31 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control X A2.03 Malfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: The hydrogen air concentration in excess of limit flame propagation or detonation with resulting equipment damage in containment 3.4 32 029 Containment Purge X2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

2.7 33 2010 Waterford 3 Examination Outline Revision 2 10 Tier 2 / Group 2 Reactor Operator (cont.)

033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator X K3.01 Knowledge of the effect that a loss or malfunction of the S/GS will have on the following: RCS 4.4 34 041 Steam Dump/Turbine Bypass Control X 2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

4.5 35 045 Main Turbine Generator 055 Condenser Air Removal X K1.06 Knowledge of the physical connections and/or cause effect relationships between the CARS and the following PRM system 2.6 36 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection X A4.05 Ability to manually operate and/or monitor in the control room: Deluge valves 3.0 38 K/A Category Point Totals:

2 0 1 1 0 0 0 2 1 1 2Group Point Total: 10 ES-401 2 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 11 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 1 K 2 K 3 A 1 A 2 GK/A Topic(s)

IR # 000007 (CE/E02) Reactor Trip -

Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 X AA2.03 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Failures of flow control valve or controller 3.6 9 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 X2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm 4.3 10 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 X2.4.41 Knowledge of the emergency action level thresholds and classifications.

4.6 11 000038 Steam Gen. Tube Rupture / 3 CE/E05 Steam Line Rupture -

Excessive Heat Transfer / 4

2010 Waterford 3 Examination Outline Revision 2 12 Tier 1 / Group 1 Senior Reactor Operator (cont.)

000054 (CE/E06) Loss of Main Feedwater / 4 X EA2.2 Ability to determine and interpret the following as they apply to the (Loss of Feedwater): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

4.2 12 000055 Station Blackout / 6 X2.1.20 Ability to interpret and execute procedure steps.

4.6 13 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus

/ 6 X AA2.04 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: ESF system panel alarm annunciators and channel status indicators 4.0 14 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals:

0 0 0 03 3 Group Point Total:

6 ES-401 3 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 13 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 1 K 2 K 3 A 1 A 2 G K/A Topic(s)

IR # 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 X AA2.03 Ability to determine and interpret the following as they apply to the Dropped Control Rod: Dropped rod, using in-core/ex-core instrumentation, in-core or loop temperature measurements 3.8 15 000005 Inoperable/Stuck Control Rod / 1 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 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel.

/ 9 000060 Accidental Gaseous Radwaste

Rel. / 9 000061 ARM System Alarms / 7

2010 Waterford 3 Examination Outline Revision 2 14 Tier 1 / Group 2 Senior Reactor Operator (cont.)

000067 Plant Fire On-site / 8 X AA2.12 Ability to determine and interpret the following as they apply to the Plant Fire on Site: Location of vital equipment within fire zone 3.9 17 000068 Control Room Evac. / 8 X2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

4.1 24 000069 Loss of CTMT Integrity / 5 000074 Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 CE/A13 Natural Circ. / 4 CE/A11 RCS Overcooling - PTS / 4 X AA2.2 Ability to determine and interpret the following as they apply to the (RCS Overcooling): Adherence to appropriate procedures an d operation within the limitations in the facility*s license and amendments.

3.4 16 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

0 0 0 0 3 1Group Point Total:

4 ES-401 4 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 15 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 SRO System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G K/A Topic(s)

IR # 003 Reactor Coolant Pump 004 Chemical and Volume

Control X A2.15 Ability to (a) predict the impacts of the following 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: High or low PZR level 3.7 1 005 Residual Heat Removal X2.2.37 Ability to determine operability and/or availability of safety related equipment.

4.6 2 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water X2.2.37 Ability to determine operability and/or availability of safety related equipment.

3.6 3 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features Actuation 022 Containment Cooling

2010 Waterford 3 Examination Outline Revision 2 16 Tier 2 / Group 1 Senior Reactor Operator (cont.)

026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater X A2.04 Ability to (a) predict the impacts of the following 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 operations 3.8 5 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator X A2.11 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Conditions (minimum load) required for unloading an ED/G 2.9 4 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals:

0 0 0 0 0 0 0 3 00 2Group Point Total:

5 ES-401 5 Form ES-401-22010 Waterford 3 Examination Outline Revision 2 17 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 SRO System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G K/A Topic(s)

IR # 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation X A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effects on axial flux density of control rod alignment and sequencing, xenon production and decay, and boron vs. control rod reactivity changes 3.8 6 016 Non-nuclear Instrumentation 017 In-core Temperature

Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

4.7 7 2010 Waterford 3 Examination Outline Revision 2 18 Tier 1 / Group 2 Senior Reactor Operator (cont.)

034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal X A2.08 Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Meteorological changes 2.8 8 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals:

0 0 0 0 0 0 0 2 0 0 1Group Point Total:

3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-32010 Waterford 3 Examination Outline Revision 2 19 Facility: Waterford 3 Date of Exam: October 13, 2010 Category K/A # Topic RO SRO-Only IR # IR # 1. Conduct of Operations 2.1.40 Knowledge of refueling administrative requirements.

2.8 66 2.1.19 Ability to use plant computers to evaluate system or component status.

3.9 67 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

4.3 68 2.1.36 Knowledge of procedures and limitations involved in core alterations.

4.1 18 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

4.6 19 3 2

2. Equipment Control 2.2.38 Knowledge of conditions and limitations in the facility license.

3.6 69 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings.

3.5 70 2.2.14 Knowledge of the process for controlling equipment configuration or status.

4.3 20 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems.

4.5 21 2 2

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

2.9 71 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

3.2 72 2.3.6 Ability to approve release permits.

3.8 22 2 1 2010 Waterford 3 Examination Outline Revision 2 20 4. Emergency Procedures /

Plan 2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

4.2 73 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

2.7 74 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. 4.0 75 2.4.18 Knowledge of the specific bases for EOPs.

4.0 23 2.4.44 Knowledge of emergency plan protective action recommendations.

4.4 25 3 2 Tier 3 Point Total 10 7 ES-401 Record of Rejected K/As Form ES-401-42010 Waterford 3 Examination Outline Revision 2 21 Tier / Group Randomly Selected K/A Reason for Rejection 1 / 1 2 / 1 RO A4.01 There is excessive coverage of Instrument Air in the exam. Topics drawn include 000056, 000065, and 000078 twice. This was all in the RO portion of the exam. 000056 (question #52), 000065 (question #56), and the first 000078(question #26) were left as selected. The second 000078 (question #27), K/A A4.01, was randomly selected as another system in Tier 2, Group 1. The new system selected was 064 Emergen cy Diesel Generator. Left the K/A as the original A4

.01 (still question #27). 1 / 1 RO AA2.06 There is an overlap issue between 015/17 RCP (question #42) and 000026 (question #45) related to CCW. 015/17 RCP (question #42) was left as selected and 000026 (question #45) was re-drawn to K/A AA2.01.

1 / 2 RO AK3.04 Waterford has no actions on fire situations in the EOPs. There is already 1 other fire question in RO and 1 in SRO. Replaced system 000067 Plant Fire on Site (question #62). Reselected system 000074, Inadequate Core Cooling and kept original K/A AK3.04 (still question #62).

1 / 2 2 / 1 2 / 2 RO There is excessive coverage of radiation monitors in the exam. Topics drawn include:

000036, Fuel handling Accident Question 58 Left as selected.

000060, Accidental Gas Radwaste Release Question 61 Reselected CE/E09, Functional Recovery, and used the same K/A. 000073, Process Rad Monitoring Question 24 Left as selected. 000055, Condenser Air Removal Question 36 Left as selected. 068, Liquid Rad Waste Question 37 Reselected 011, Pressurizer level Control, and used the same K/A. This was all in the RO portion of the exam. These moves also removed potential overlap between the written exam and operating exam.

1 / 1 SRO AA2.06 Addressed overlap between RO test, 015/017 (question #42) and 026 (question #45) in the RO test and 026 (question #10) in the SRO test. Reselected K/A 2.4.45 for 026, SRO question #10.

2010 Waterford 3 Examination Outline Revision 2 22 2 / 1 RO K3.02 Addressed overlap between RO question #1 and #2. Both were against system 003, Reactor Coolant Pump. Question #1 was K/A K3.02 and question #2 was K5.04. Both were testing the same concept for the same components. Reselected new K/A for question #1. The new K/A is K3.01.

2 / 1 SRO A2.04 There is excessive coverage of Component Cooling in the exam. Topics drawn include 000008, 000076, and 000026. This was all in the SRO portion of the exam. Questions 000008 (question #3) and 000026 (question #10) were left as selected. The question 000076 (question #5), K/A A2.01, was removed and randomly selected a new system in Tier 2, Group 1. The new system selected was 061 Auxiliary/Emergency Feedwater and randomly selected a new K/A (still question #5).

2 / 1 RO K5.05 RO question 2: Efforts to write a discriminatory question against this K/A did not yield a suitable question. The Waterford 2 loop, 4 RCP design did not fit well with this K/A. There is already 1 other RCP question in the RO section on RCS affects. Randomly reselected K/A to replace K5.04.

2 / 1 SRO SRO question 1 listed the K/A 2.10 with the description for 2.01. K/A 2.01 was the correct K/A for this question. No actual K/A change, typo correction.

2 / 1 SRO 2.2.37 SRO question 3 was K/A 2.4.9 for a LOCA or loss of Shutdown Cooling situation related to CCW. The main interrelation with CCW and LOCA is the RCPs. There is already a question on the RO portion of the test related to this topic. Efforts to create a question related to Shutdown Cooling and CCW did not produce a suitable SRO level question. Reselected another generic K/A.

3 SRO 2.1.36 Question 18 previous K/A was 2.1.13 related to vital access control. Efforts to create a suitable SRO level question on this K/A were not successful. Reselected another K/A in section 2.1.

3 SRO 2.4.44 SRO question 11 was selected for K/A 2.4.4; a SRO question could not be written for this K/A for ATWS. Reselected K/A 2.4.41, which was to classify an event for E Plan. This K/A was already selected for question 25. To avoid having too many questions about classifying an E Plan event, used K/A 2.4.41 for question 11 and reselected K/A 2.4.44 for question 25.

1 / 1 RO AA2.19 RO question 53: Efforts to write a discriminatory question against this K/A did not yield a suitable question. Randomly reselected another K/A from section AA2 to replace AA2.05.

1 / 2 SRO AA2.2 Replaced due to overlap. Topics not already sampled in Tier 1, group 2 were assembled, and CE/A11 was drawn. Used A2 to stay consistent with original K/A number.

2 / 1 SRO A2.15 SRO question 1: Efforts to write a discriminatory question against this K/A did not yield a suitable question. Reselected from A2.01 to A2.15 without changing system.

2010 Waterford 3 Examination Outline Revision 2 23 Question Number Description of Change Revision RO 21 Reselected K/A 2.4.46 based on NRC feedback 1 RO 64 Reselected K/A 2.1.25 based on NRC feedback 1 SRO 7 Reselected K/A 2.4.4 based on NRC feedback 1 RO 39 Added K/A root statement to document.

1 RO 49 Added K/A root statement to document.

1 RO65 Added K/A root statement to document.

1 SRO 12 Added K/A root statement to document.

1 SRO 5 Reselected system 061and K/A A2.04 1 RO 2 Reselected K/A K5.05 1 SRO 1 Corrected typo.

1 SRO 3 Reselected K/A 2.2.37 1 SRO 18 Reselected K/A 2.1.36.

1 SRO 11 / 25 Reselected K/A 2.4.41 for question 11, reselected 2.4.44 for question 25.

1 RO 53 Reselected K/A AA2.19 1 SRO 16 Reselected K/A AA2.2 2 SRO 1 Reselected K/A A2.15.

2 ES-301 Administrative Topics Outline Form ES-301-1 NRC Revision 1 Facility: WATERFORD 3 Date of Examination: October 4, 2010 Examination Level:

SRO Operating Test Number: NRC Administrative Topic (see Note)

Type Code* Describe activity to be performed A1 Conduct of Operations K/A Importance:

4.4 S, N 2.1.23, Ability to perform specific system and integrated plant procedures during all modes of plant operation. Review and approve completed OP-903-117, Emergency Diesel Generator Fuel Oil Transfer Pump Operability Check, Attachment 10.1, Fuel Oil Transfer Pump A IST Data.

A2 Conduct of Operations K/A Importance:

4.6 S, M 2.1.20, Ability to interpret and execute procedure steps. Review COLSS constant calculation in accordance with OP-004-005, Core Operating Limits Supervisory System Operation A3 Equipment Control K/A Importance:

4.7 S, N 2.2.40, Ability to apply Technical Specifications for a system. Evaluate Safety Injection System voiding in accordance with OP-903-026, Emergency Core Cooling System Valve Lineup Verification.

A4 Radiation Control K/A Importance:

3.8 S, D, P 2.3.6, Ability to approve release permits. Review and approve a liquid release permit in accordance with OP-007-001, Boron Management.

A5 Emergency Plan K/A Importance:

4.6 S, M 2.4.41, Knowledge of the emergency action level thresholds and classifications. Determine appropriate Emergency Plan action level in accordance with EP-001-001, Recognition and Classification of Emergency Conditions.

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)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: October 4, 2010 Exam Level SRO - Instant Operating Test No.:

NRC 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 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 final reactor trip contingency from EOP OP-902-000, Standard Post Trip Actions. A2.13 ATWS RO - 4.4, SRO - 4.6 A, D, P, S 1 S2 004 Chemical and Volume Control System; Perform actions associated with a faulted Pressurizer level setpoint in accordance with OP-901-110, Pressurizer Level Control Malfunction. A2.22 Mismatch of Letdown and Charging flows RO - 3.2, SRO - 3.1 A, M, S 2 S3 006 Emergency Core Cooling System; Reduce RCS pressure and use High Pressure Safety Injection Pumps to restore Pressurizer level in accordance with OP-901-112, Charging or Letdown Malfunction. A1.18 PZR level and pressure RO - 4.0, SRO - 4.3 N, S 3 S4 005 Shutdown Cooling System; 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. A4.01 Controls and indication for RHR pumps RO - 3.6, SRO - 3.4 A, D, L, P, S 4 - P S5 062 A.C. Electrical Distribution, Synchronize the Main Generator to the grid in accordance with OP-010-004, Power Operations. Fault: Auto Synchronization will fail to function, requiring manual synchronizing of the Generator Output Breakers. A4.07 Synchronizing and paralleling of different AC supplies RO - 3.1, SRO - 3.1 A, N, S 6 S6 012 Reactor Protection System Reset CSAS in accordance with OP-902-009, Standard Appendices, Section 5 - E A4.04 Bistable, trips, reset and test switches RO - 3.3, SRO - 3.3 D, EN, L, P, S 7 S7. 068 Liquid Radwaste System; Discharge WCT A to the Circulating Water System in accordance with OP-007-004, Liquid Waste Management System Fault: Upon initiation of flow, controller fails in raise, exceeding maximum flow allowed, requiring the operator to manually close the isolation valves. A4.03 Stoppage of release if limits exceeded RO - 3.9, SRO - 3.8 A, M, S 9 1 NRC Revision 2 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 System; Reset overspeed device on Emergency Feedwater Pump AB in accordance with OP-902-005, Station Blackout Recovery. A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 D, E, L, R 4 - S P2 008 Component Cooling Water System; Restore Power to the DCT Sump Pumps Following a Loss of Off Site Power in accordance with OP-902-009, Standard Appendices G2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

RO - 4.2, SRO - 4.1 D, R, E 8 P3 062 A.C. Electrical Distribution Transfer SUPS 014AB from Alternate to Normal AC Power in accordance with OP-006-005, Inverters and Distribution. Fault: After alignment, voltage will not be indicated on SUPS 014 AB inverter. A3.04 Operation of inverter RO - 2.7, SRO - 2.9 A, M 6 @ 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 6 (C)ontrol room (D)irect from bank 9 / 8 / 4 5 (E)mergency or abnormal in-plant 1 / 1 / 1 1 (EN)gineered safety feature - / - / 1 (control room system) 1 (L)ow-Power / Shutdown 1 / 1 / 1 3 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 5 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 3 (R)CA 1 / 1 / 1 2 (S)imulator 2 NRC Revision 2 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: October 4, 2010 Exam Level SRO - Upgrade Operating Test No.:

NRC 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 004 Chemical and Volume Control System; Perform actions associated with a faulted Pressurizer level setpoint in accordance with OP-901-110, Pressurizer Level Control Malfunction. A2.22 Mismatch of Letdown and Charging flows RO - 3.2, SRO - 3.1 A, M, S 2 S3 S4 S5 S6 012 Reactor Protection System Reset CSAS in accordance with OP-902-009, Standard Appendices, Section 5 - E A4.04 Bistable, trips, reset and test switches RO - 3.3, SRO - 3.3 D, EN, L, P, S 7 S7. 068 Liquid Radwaste System; Discharge WCT A to the Circulating Water System in accordance with OP-007-004, Liquid Waste Management System Fault: Upon initiation of flow, controller fails in raise, exceeding maximum flow allowed, requiring the operator to manually close the isolation valves. A4.03 Stoppage of release if limits exceeded RO - 3.9, SRO - 3.8 A, M, S 9 3 NRC Revision 2 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 System; Reset overspeed device on Emergency Feedwater Pump AB in accordance with OP-902-005, Station Blackout Recovery. A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 D, E, L, R 4 - S P2 P3 062 A.C. Electrical Distribution Transfer SUPS 014AB from Alternate to Normal AC Power in accordance with OP-006-005, Inverters and Distribution. Fault: After alignment, voltage will not be indicated on SUPS 014 AB inverter. A3.04 Operation of inverter RO - 2.7, SRO - 2.9 A, M 6 @ 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 3 (C)ontrol room (D)irect from bank 9 / 8 / 4 2 (E)mergency or abnormal in-plant 1 / 1 / 1 1 (EN)gineered safety feature - / - / 1 (control room system) 1 (L)ow-Power / Shutdown 1 / 1 / 1 3 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 3 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 1 (R)CA 1 / 1 / 1 1 (S)imulator 4 NRC Revision 2 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 5 NRC Revision 2 Description of Change Revision JPM S5, Procedure title was incorrect. Changed to OP-010-004, Power Operations.

1 JPM P2 replaced 2

Appendix D Scenario Outline Form ES-D-1 Scenario 1 Rev 1 Facility: WATERFORD 3 Scenar io No.: 1 Op Test No.:

NRC Examiners:

Operators:

Initial Conditions: 100%, 125 EFPD, AB buses aligned to Train B Protected Train is B Emergency Diesel Generator A is out of service.

Turnover: Maintain 100 % power Event No. Malf. No.

Event Type* Event Description 1 H_105 I - ATC I - SRO Remove Reactor Power Cutback from service in accordance with OP-004-015.

2 RC22F2 TS - SRO I - SRO Pressurizer pressure instrument RC-IPI-0102 B fails low.

3 SG01B CV02C TS - SRO C - SRO C - ATC Steam Generator 2 tube leakage Charging Pump AB fails to auto start 4 N/A R- ATC N-BOP N-SRO Rapid Plant Power Reduction due to Steam Generator tube leakage.

5 RC08A RC09A RC10A C - ATC C - SRO Reactor Coolant Pump 1A seal failure, manual reactor trip 6 SG01B M-All Steam Generator #2 Tube Rupture 7 RP08C C - ATC C - SRO Containment Isolation CVC-401 fails to auto close on CIAS 8 MS11B M - All Main Steam li ne break on Steam Generator 2 inside Containment. * (N)ormal, (R)eactivi ty, (I)nstrument, (C)omponent, (M)ajor

Scenario Event Description NRC Scenario 1 Scenario 1 Rev 1 The crew assumes the shift at 100% power with instructions to maintain 100% power.

After assuming the shift, annunciator H1005, Reactor Power Cutback Single Channel Trouble, alarms and Reactor Power Cutba ck auto actuation fails. The crew will be contacted by I&C maintenance after re viewing the annunciator response procedure

describing a failure of Reactor Power Cut back, requiring Reactor Power Cutback be removed from service in accordance with OP-004-015, Reactor Power Cutback. This manipulation is performed by the ATC operator at CP-2 and CP-7.

After Reactor Power Cutback is removed from service, Pressurizer pressure instrument RC-IPI-0102 B fails low. The AT C operator will receive the annunc iators for this failure. The CRS should evaluate Tech Specs and enter Tech Spec 3.3.1 and 3.3.2 and determine that Plant Protection System bistable 6 for low Pressurizer pressure must be bypassed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> on C hannel B. Tech Spec 3.

3.3.5 and 3.3.3.6 should be referenced but not entered. Plant Monito ring Computer group SPD S indication of Pressurizer pressure is affected by this failure.

After the CRS evaluates Tech Specs but before bistable 6 is bypassed, a Steam Generator tube leak develops in Steam Generator 2. Charging Pump AB will be aligned as the first backup Charging Pump and it will fail to auto start on the lowering

Pressurizer level. The ATC operator should make this diagnosis and start Charging Pump AB. The CRS should enter OP-901-202, Steam Generator Tube Leakage or High Activity. The size of the leak will require entering OP-901-212, Rapid Plant Power Reduction. For the power reduction, the ATC will perform Direct Boration to the RCS as well as Pressurizer boron equalization and ASI control with CEAs. The BOP will manipulate the controls to reduce Main Turbine load. The CRS should enter Tech Spec 3.4.5.2 for Steam Generator 2 leakage. The CRS may consider declaring Charging Pump AB inoperable due to it's failure to auto start, but no Tech Spec entry is required.

Once the crew has commenced the power reduc tion and lowered power to ~ 90%, or at the lead examiner's cue, RCP 1a will develop multiple seal failures. The second and third seal failures will occur after the crew has entered the appropriate off normal. The crew should enter off no rmal procedure OP-901-130, Reactor Coolant Pump Malfunction, before the subsequent seals fail. With the second and third seals failing, the crew will be required to trip the reactor.

When the reactor is tripped, the Steam Genera tor tube leak will degr ade to a rupture. The CRS should direct the ATC to initiate Sa fety Injection and Containment Isolation.

Containment Isolation valve CVC-401 at CP-4 will fail to close on the CIAS, requiring

the ATC operator to close the valve.

Scenario Event Description NRC Scenario 1 Scenario 1 Rev 1 After the crew completes OP-902-000, Standard Post Trip Actions and the CRS diagnoses into OP-902-007, Steam Generator Tube Rupture Recovery, the CRS should direct a rapid RCS cooldown to less than 520 °F T HOT. After this direction is given, a Main Steam Line break will develop on Stea m Generator 2 inside Containment. The CRS should exit OP-902-007 and enter OP-902-008, Functional Recovery Procedure.

Prioritization in OP-902-008 should result in Containment Isolation being priority 1. The crew should address Containment Isolation by using the steps in the Heat Removal section to isolate Steam Generator 2.

The scenario can be terminated after Steam G enerator 2 is isolat ed, or at the lead examiners discretion.

Scenario Event Description NRC Scenario 1 Scenario 1 Rev 1 Critical Tasks

1. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow. The required task becomes applicable after Containment Spray has been actuated. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling.

2. Prevent Opening the Main Steam Safety Valves.

This task is satisfied by the crew taking action to maintain Steam Generator #2 pressure below the safety valve setpoint by taking action to reduce RCS pressure to < 945 psia.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 8 2. Malfunctions after EOP entry (1-2) 2 3. Abnormal events (2-4) 3 4. Major transients (1-2) 2 5. EOPs entered/requiring substantive actions (1-2) 2 6. EOP contingencies requiring substantive actions (0-2) 1 7. Critical tasks (2-3) 2

NRC Scenario 1 Scenario 1 Rev 1 Scenario Notes:

A. Reset Simulator to IC-191.

B. Verify the following Scenario Malfunctions are loaded:

1. rc22f2 for Pressurizer le vel instrument RC-ILI-0102 B

2. sg01B for Steam Generator #2 tube leak

3. cv02c for Charging Pump AB

4. rc08a, 09a, and 10a for RCP 1A

5. rp08c for CVC-401

6. ms11b for Steam Generat or #2 steam line break

7. eg10a for EDG A overspeed device C. Verify the following remotes

1. egr27 for EDG B local alarm acknowledgement D. Verify the following overrides

1. di-08a05s14-1 for closing FP-601 A set to trigger 9

2. di-08a05s19-1 for closi ng IA-909 set to trigger 9

3. di-04a04s10-1 for closing CVC-109 set to trigger 9

4. di-02a05a2s34-0 for Reactor Power Cutback out of service set to trigger 1.

5. di-02a05a2s30-0 for Reactor Power Cutback lamp test.

E. Verify the following under Event Triggers:

1. zdirpciastrp(4).eq.1 is set on trigger 9 F. Verify the following C ontrol Board Conditions:

1. Danger tag placed on EDG A control switch

2. Danger tag placed on EDG A Output Breaker G. Verify EDG A output breaker is racked out and place danger tags on EDG A and its output breaker.

H. Ensure Protected Train B sign is placed in SM office window.

I. Verify EOOS is 8.5 Yellow

J. Complete the simulator setup checklist.

K. Start DCS, Record Data, select file PlantParameters.txt.

NRC Scenario 1 Scenario 1 Rev 1 Simulator Booth Instructions Event 1 Reactor Power Cutback Failure

1. On Lead Examiner's cue, initiate Event Trigger 1.

2. After the annunciator response procedure is referenced, call the CRS as the on shift I&C technician. R eport that during planned data collection, the Reactor Power Cutback Channel A has locked up and will not respond to any commands or signals.

3. If Work Week Manager is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 Pressurizer Pressure Instru ment RC-IPI-0102 B Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

3. If sent to LCP-43, report RCS pressu re indications are reading normal on all channels except for Channel B.

Event 3 / 4 Steam Generator #2 Tube Leak / Charging Pump AB Fails to Start / Rapid Plant Power Reduction

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If Chemistry or HP is called, acknowledge S/G tube leak and need to carry out the actions of UNT-005-032, Steam Gener ator Primary to Secondary Leakage.

3. If called as TGB watch to monitor Condensate System Polisher pressure, acknowledge and remove Polisher Vessel from service if necessary.

4. If called as the RCA Watch to check C harging Pump AB, report that it looks normal locally.

Event 5 Reactor Coolant Pump 1A seal failure

1. On Lead Examiner's cue, initiate Event Trigger 5.

2. If Engineering is called, acknowledge communication and report that engineering will investigate the RCP 1A failure.

3. If called as the RAB Watch to check RCP vibration in Switchgear room B, acknowledge communication, no report is necessary.

4. Initiate Event Trigger 12 after the crew has entered OP-901-130.

NRC Scenario 1 Scenario 1 Rev 1 Event 4 Steam Generator #2 Tube Rupture

1. On Lead Examiner's cue, initiate change the severity of malfunction SG01B to 7% and ramp it in over a 1 minute period.

2. If called as RAB watch to check Emergen cy Diesel Generator B, initiate Trigger 16, after EDG B Trouble alarm clear, report it is running satisfactorily.

Event 5 Steam Generator #2 Steam Line Break

1. After the crew has entered OP-902-007 and commenced the rapid cooldown and on the Lead Examiner's cue, in itiate Event Trigger 7.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2010 SRO Scenario1.cdf NRC Scenario 1 Scenario 1 Rev 1 Scenario Timeline: Event Malfunction Severity Ramp HH:MM:SS Delay Trigger 1 H_L05 N/A N/A N/A 1 Remove Reactor Power Cutback from service 1 DI-02A05A2S34-0 N/A N/A N/A 1 Reactor Power Cutback Auto Actuate 1 DI-02A05A2S30-0 N/A N/A N/A 1 Remove Reactor Power Cutback Lamp Test 2 RC22F2 N/A N/A N/A 2 Pressurizer pressure instru ment RC-IPI-0102 B fails low 3 SG01 B 0.6 % 00:3:00 NA 3 Steam Generator #2 Tube Leak 4 CV02C N/A N/A N/A N/A Charging Pump AB fails to auto start 5 RC08 RC09 RC10 A 100 % 00:30 00:30 00:30 N/A 5 12 12 Reactor Coolant Pump 1A seal failures 6 SG01 B 7.0 % 00:01:00 N/A N/A Steam Generator #2 Tube Rupture 7 RP08C N/A N/A N/A N/A CVC-401 fails to close on CIAS 8 MS11B 12 % 03:00 N/A 8 Main Steam Line Break S/G 2

NRC Scenario 1 Scenario 1 Rev 1

REFERENCES:

Event Procedures 1 OP-004-015, Reactor Power Cutback System 2 OP-009-007, Plant Protection System Tech Spec 3.3.1 and 3.3.2 3 OP-901-202, Steam Generator Tube Leakage or High Activity Tech Spec 3.4.5.2 4 OP-901-212, Rapid Plant Power Reduction Tech Spec 3.1.3.6 Regulating an d Group P CEA Insertion Limits 5 OP-901-130, Reactor C oolant Pump Malfunction 6 OP-902-000, Standard Post Trip Actions OP-902-007, Steam Generator Tube Rupture Recovery OP-902-009, Standard Appendices, Append ix 1, Diagnostic Flow Chart 8 OP-902-008, Safety Function Recovery Procedure OP-902-009, Standard Appendic es, Appendix 13, Stabilize RCS Temperature

Appendix D Scenario Outline Form ES-D-1 Scenario 3 Rev 1 Facility: WATERFORD 3 Scenar io No.: 3 Op Test No.:

NRC Examiners:

Operators:

Initial Conditions: Mode 2 with 2 Charging Pumps in operation Protected Train is B AB Bus is aligned to Train B Reactor power is 1%

Turnover: Dilute to 5-15% power Event No. Malf. No. Event Type*

Event Description 1 N/A R - ATC N - SRO Dilute to 5-10% power, perform 100 gallon PMU addition.

2 CB08B1 TS - SRO Containment pressure instrument CB-IPI-6702-SMC C fails high.

3 CV35A CVR101 C - ATC C - BOP C - SRO During dilution, PMU counter fails to secure flow OP-901-104, Inadvertent Positive Reactivity Addition 4 CC02A TS - SRO Auxiliary Component Cooling Water Pump A trip 5 RX14A I - ATC I - SRO Pressurizer Pressure RC-IPR-0100 X fails high, Main Spray Valves open 6 RC23A M - ALL Small break loss of coolant accident 7 RP05 A3, B3, C3, D3 I - ATC I - SRO Containment Spray fails to actuate

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Scenario Event Description NRC Scenario 3 Scenario 3 Rev 1 The crew assumes the shift with the reactor at 1% power following a forced outage. The turnover will include instructions to perform RCS dilution to 5 - 15% power.

The reactivity plan will include instructions to dilute in multiple PMU batches. The initial batch will be 100 gallons of PMU. Each subsequent batch will be 50 gallons of PMU.

This will allow for an observable power rise without concern for a reactor trip on the PMU failure.

After the first 100 gallons of PMU are added, Containment pr essure instrument CB-IPI-6702-SMC Channel C input to Containment Spray Actuation fails high. The CRS should enter Tech Spec 3.3.2 and direct bypassing high Containment Pressure - Hi Hi bistable on Channel C.

During the second dilution, the Primary Water counter will fail to secure dilution. The ATC should attempt to secure Primar y Water Flow by operating PMU-144 and CVC-510. Neither of these acti ons will secure flow. The CRS should enter OP-901-104, Inadvertent Positive Reactivity Additi on, and secure Primary Makeup Pump A.

After these actions are completed, Auxiliary Component Cooling Water Pump A will trip. The crew should verify Component Cooling Water temperature is being controlled by Dry Cooling Tower Fans. The SRO should declare ACCW Pump 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 surv eillance 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 Tech Specs are evaluated, Pre ssurizer pressure instrument RC-IPR-0100 X fails high. This causes both Main Spray valves to open. The SRO should enter OP-

901-120, Pressurizer Pressure Malfunction. The ATC will align the non-faulted Pressurizer pressure channel.

After the Pressurizer Pressure Control Channel Y is selected, a small break LOCA event will occur. Containment Spray Actuation will fail requi ring manual actuation. All Reactor Coolant Pumps will be required to be secured after the CSAS.

After the crew has completed OP-902-000, St andard Post Trip Actions, the CRS should diagnose into OP-902-002, Loss of Coolant Accident Recovery. The scenario can be

terminated after the crew has commenced an RCS cooldown using Atmospheric Dump Valves.

NRC Scenario 3 Scenario 3 Rev 1 Critical Tasks

1. Establish Containment temperature and pressure control This task is satisfied by manually actuating Containment Spray prior to completing the review of step 8 in OP-902-000, Standard Post Trip Actions, or before completing the review of step 15 in OP-902-002, Loss of Coolant Accident Recovery, if OP-902-000 is exited before Containment pressure exceeds 17.7 psia.

2. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow. The required task becomes applicable after Containment Spray has been actuated. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 6 2. Malfunctions after EOP entry (1-2) 1 3. Abnormal events (2-4) 2 4. Major transients (1-2) 1 5. EOPs entered/requiring substantive actions (1-2) 1 6. EOP contingencies requiring substantive actions (0-2) 0 7. Critical tasks (2-3) 2

NRC Scenario 3 Scenario 3 Rev 1 Scenario Notes:

A. Reset Simulator to IC-193.

1. Use keys 165 - 168 for S/G high level bypass setup.

B. Verify the following Scenario Malfunctions:

1. cv35a for the PMU Dilution counter

2. rx14-A for Pressurizer pressure instrument RC-IPT-0100 X

3. cb08b1 for Containment Pressu re instrument CB-IPI-6702-SMC

4. cc02a for Auxiliary Component Cooling Water Pump A

5. rc23a for RCS break

6. rp05a3, b3, c3, d3 for C ontainment Spray failure C. Verify the following Remotes:

1. cvr101 set at 2% for PMU-140

2. anr04h for EDG A local alarm acknowledgement.

3. anr04i for EDG B loca l alarm acknowledgement.

D. Verify the following Event Trigger

1. zdifwpmuecs1357(1).eq.1 sets irf cvr101 0.

E. Ensure Protected Train B sign is placed in SM office window.

F. Verify EOOS is 10.0 Green

G. Complete the simulator setup checklist.

H. Remove Caution Tag from PMU-141.

I. Remove PMC point D39502 from scan.

J. Start DCS, Record Data, select file PlantParameters.txt.

NRC Scenario 3 Simulator Booth Instructions Event 1 Dilute to 5-10% power, Perf orm 100 gallon PMU addition

1. No communications should occur for this evolution.

Event 2 Containment pressure instrum ent CB-IPi-6702 SMC fails high

1. After the first 100 gallon addition is co mpleted, or on Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 PMU flow malfunction

1. After the Containment Pressure instru ment has failed, the crew should make their second PMU addition. Insert Event Trigger 3 after the ATC has established PMU flow.

2. If called to operate valves listed in OP-901-104, acknowledge communication and report that you will work on valve list.

Event 4 Auxiliary Component Cooling Water Pump A trip

1. After PMU Pump A is secured and on the Lead Examiner's cue, initiate Event Trigger 4.

2. If called as the watchstander and sent to ACCW Pump A, report that the pump looks normal locally.

3. If called as the watchstander and sent to ACCW Pump A breaker, report that the breaker indicates open and that there are various breaker parts on the floor of the cubicle.

4. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 Pressurizer pressure instrum ent RC-IPR-0100 X fails high

1. After Tech Specs for Event 4 are addresse d, or on Lead Examiner's cue, initiate Event Trigger 5.

2. If Work Week Manager or I&C is called, inform the ca ller that a work package will be assembled and a team will be sent to the Control Room.

Scenario 3 Rev 1 NRC Scenario 3 Event 6 Loss of Coolant Accident

1. After the actions in OP-901-120 are complete and on the Lead Examiner's cue, initiate Event Trigger 6.

2. If called as RAB watch to check EDG A & B, initiate Trigger 10, and when the EDG A & B Trouble alarms are clear, report that they are running satisfactorily.

Event 7 Containment Spray fails to actuate

1. No communications should occur for this malfunction.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2010 SRO Scenario3.cdf

Scenario 3 Rev 1 NRC Scenario 3 Scenario Timeline: Event Malfunction Severity Ramp HH:MM:SS Delay Trigger 1 N/A N/A N/A N/A N/A Dilute to critical boron concentration 2 CH08 B1 N/A N/A N/A 2 Containment Pressure Instrument CB-IPI-6702-SMC fails high 3 CVR101 2% N/A N/A 3 PMU - 140 throttled open 3 CV35 A N/A N/A N/A 3 PMU counter fails to secure dilution 4 CC02 A N/A N/A N/A 4 Auxiliary Component Cooling Water Pump A trips 5 RX14 A 100% N/A N/A 5 Pressurizer pressure instru ment RC-IPR-0100 X fails high 6 RC23 A 0.05% 5:00 N/A 6 Small Break LOCA 7 RP05 A3, B3, C3, D3 N/A N/A N/A N/A Containment Spray fails to actuate Scenario 3 Rev 1 NRC Scenario 3 Scenario 3 Rev 1

REFERENCES:

Event Procedures 1 OP-010-003, Plant Startup OP-002-005, Chemical and Volume Control 2 OP-009-007, Plant Protection System OP-903-013, Monthl y Channel Checks Tech Spec 3.3.2 3 OP-901-104, Inadvertent Positive Reactivity Addition 4 Tech Spec 3.7.3 OP-100-014 5 OP-901-120, Pressurizer Pressure Control Malfunction Tech Spec 3.2.8 6 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Append ix 1, Diagnostic Flow Chart OP-902-002, Loss of Coolant Accident Recovery 7 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance 8 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operat ing Procedures Operations Expectations /

Guidance ES-301 Simulator Scenario Quality Checklist Form ES-301-5 Facility: Waterford 3 Date of Exam:

October 4, 2010 Operating Test No.NRC A P P L I C A N T E V E N T T Y P

E Scenarios 1 2 3 T O T A L M I N I M U M(*) CREW POSITION CREW POSITION CREW POSITION CREW POSITION S A T C B O P S R O A T C B O P S R O A T C B O P S R O A T C B O P R O R I U SRO-I 1 Name RX 4 1 1 1 0 NOR 1 1 1 1 1 I/C 1,3,5, 7 3,5,7 7 4 4 2 MAJ 6,8 6 3 2 2 1 TS 2,4 2 0 2 2 SRO-I 2 Name RX 4 1 1 1 0 NOR 1 1 1 1 1 I/C 1,3,5, 7 3,5,7 7 4 4 2 MAJ 6,8 6 3 2 2 1 TS 2,4 2 0 2 2 SRO-I 3 Name RX 1 1 1 1 0 NOR 4 1 1 1 1 I/C 1,3,5, 7 3,5,7 7 4 4 2 MAJ 6,8 6 3 2 2 1 TS 2,3 2 0 2 2 SRO-U Name RX 1 1 1 1 0 NOR 4 1 1 1 1 I/C 1,3,5, 7 3,5,7 7 4 4 2 MAJ 6,8 6 3 2 2 1 TS 2,3 2 0 2 2 Spare RX 3 NOR 3 I/C 1,4,5, 7 1,4,5 MAJ 4,6 4,6 TS 1,2 NRC Revision 2 NRC Revision 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 applicant's competence count toward the minimum requirements specified for the applicant's license level in the right-hand columns.

E S-301 Competencies Checklist Form ES-301-6 Facility: Waterford 3 Date of Exam: October 4, 2010 Operating Test No.: NRC Competencies APPLICANTS

CRS

ATC

SCENARIO SCENARIO SCENARIO 1 2 3 1 2 3 Interpret/Diagnose Events and Conditions 1, 2, 3, 5, 6, 8 1, 2, 3, 4, 5, 6, 7 2, 3, 4, 5, 6, 7 1, 3, 5, 6, 7, 8 1, 4, 5, 6 3, 5, 6, 7 Comply With and Use Procedures (1)

All 1, 3, 4, 5, 6, 8 1, 3, 4, 5, 6 1, 3, 5, 6, 7 Operate Control Boards (2)

N/A 1, 3, 4, 5, 6, 7 1, 3, 4, 5 1, 3, 5, 6, 4 Communicate and Interact All 1, 2, 3, 4, 5, 6, 7, 8 1, 3, 4, 5, 6 1, 2, 3, 5, 6, 7 Demonstrate Supervisory Ability (3)

All N/A N/A Comply With and Use Tech. Specs. (3) 2,3 1,2 2,4 N/A N/A Notes: (1) Includes Technical Specification compliance for an RO. (2) Optional for an SRO-U. (3) Only applicable to SROs.

Instructions:

Check the applicants' license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.