ML110970438

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WT-2011-03 Final Outlines
ML110970438
Person / Time
Site: Waterford Entergy icon.png
Issue date: 03/21/2011
From:
NRC Region 4
To:
Entergy Operations
References
50-382/11-301
Download: ML110970438 (57)
v • d • e


Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 Date of Exam: March 29, 2011 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

  • Total
1. 1 3 3 3 3 3 3 18 3 3 6 Emergency &

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

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

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

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

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

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.

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

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

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

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

2011 Waterford 3 Examination Outline Revision 2 1

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)

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

Function 1 2 3 1 2 Knowledge of the interrelations between the (Reactor Trip Recovery) and the following: Components, and functions of 000007 (CE/E02) Reactor control and safety systems, including instrumentation, signals, R1 Trip - Recovery / 1 X EK2.1 interlocks, failure modes, and automatic and manual features. 3.3 1 000008 Pressurizer Vapor Knowledge of the interrelations between the Pressurizer R2 Space Accident / 3 X AK2.01 Vapor Space Accident and the following: Valves. 2.7* 1 000009 Small Break LOCA Conduct of Operations: Ability to interpret and execute R3 /3 X 2.1.20 procedure steps. 4.6 1 000011 Large Break LOCA Emergency Procedures / Plan: Knowledge of the specific R4 /3 X 2.4.18 bases for EOPs. 3.3 1 Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow): Basic steady state 000015/17 RCP thermodynamic relationship between RCS loops and S/Gs R5 Malfunctions / 4 X AK1.04 resulting from unbalanced RCS flow. 2.9 1 Ability to determine and interpret the following as they 000022 Loss of Rx Coolant apply to the Loss of Reactor Coolant Makeup: Whether R6 Makeup / 2 X AA2.01 charging line leak exists. 3.2 1 000025 Loss of RHR N/A System / 4 N/A Randomly Deselected. N/A 0 000026 Loss of Ability to operate and / or monitor the following as they Component Cooling Water apply to the Loss of Component Cooling Water: Loads on R7 /8 X AA1.02 the CCWS in the control room. 3.2 1 000027 Pressurizer Ability to operate and / or monitor the following as they Pressure Control System apply to the Pressurizer Pressure Control Malfunctions:

R8 Malfunction / 3 X AA1.04 Pressure recovery, using emergency-only heaters. 3.9* 1 Knowledge of the interrelations between the ATWS and R9 000029 ATWS / 1 X EK2.06 the following: Breakers, relays, and disconnects. 2.9* 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 2

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

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

Function 1 2 3 1 2 000038 Steam Gen. Tube Knowledge of the reasons for the following responses as R10 Rupture / 3 X EK3.08 they apply to the SGTR: Criteria for securing RCP. 4.1 1 Knowledge of the operational implications of the 000040 (CE/E05) Steam following concepts as they apply to the (Excess Steam Line Rupture - Excessive Demand): Normal, abnormal and emergency operating R11 Steam Demand / 4 X EK1.2 procedures associated with (Excess Steam Demand). 3.2 1 Knowledge of the operational implications of the following concepts as they apply to Loss of Main 000054 (CE/E06) Loss Feedwater (MFW): Effects of feedwater introduction on dry R12 of Main Feedwater / 4 X AK1.02 S/G. 3.6 1 Ability to determine or interpret the following as they 000055 Station Blackout / apply to a Station Blackout: Actions necessary to restore R13 6 X EA2.03 power. 3.9 1 Knowledge of the reasons for the following responses as 000056 Loss of Off-site they apply to the Loss of Offsite Power: Order and time to R14 Power / 6 X AK3.01 initiation of power for the load sequencer. 3.5 1 Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus:

000057 Loss of Vital AC Actions contained in EOP for loss of vital ac electrical R15 Inst. Bus / 6 X AK3.01 instrument bus. 4.1 1 Ability to determine and interpret the following as they 000058 Loss of DC Power apply to the Loss of DC Power: DC loads lost; impact on to R16 /6 X AA2.03 operate and monitor plant systems. 3.5 1 Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): Flow 000062 Loss of Nuclear rates to the components and systems that are serviced by the R17 Svc Water / 4 X AA1.07 SWS; interactions among the components. 2.9 1 000065 Loss of Instrument Emergency Procedures / Plan: Knowledge of EOP R18 Air / 8 X 2.4.6 mitigation strategies. 3.7 1 000077 Generator Voltage and Electric Grid N/A Disturbances / 6 N/A Randomly deselected. N/A 0 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 3

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

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

Function 1 2 3 1 2 K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 4

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)

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

1 2 3 1 2 000001 Continuous Rod N/A Withdrawal / 1 N/A Randomly deselected. N/A 0 Ability to operate and / or monitor the following as they apply to the Dropped Control Rod: Reactor R19 000003 Dropped Control Rod / 1 X AA1.05 power - turbine power. 4.1 1 Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck 000005 Inoperable/Stuck Control Control Rod: Actions contained in EOP for R20 Rod / 1 X AK3.06 inoperable/stuck control rod. 3.9 1 N/A 000024 Emergency Boration / 1 N/A Randomly deselected. N/A 0 000028 Pressurizer Level Equipment Control: Ability to determine operability R21 Malfunction / 2 X 2.2.37 and/or availability of safety related equipment. 3.6 1 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear 000032 Loss of Source Range NI / Instrumentation: Expected values of source range R22 7 X AA2.03 indication when high voltage is automatically removed. 2.8 1 000033 Loss of Intermediate N/A Range NI / 7 N/A Randomly deselected. N/A 0 N/A 000036 Fuel Handling Accident / 8 N/A Randomly deselected. N/A 0 Knowledge of the reasons for the following 000037 Steam Generator Tube responses as they apply to the Steam Generator R23 Leak / 3 X AK3.09 Tube Leak: Maximum load change capability of facility. 2.7* 1 000051 Loss of Condenser N/A Vacuum / 4 N/A Randomly deselected. N/A 0 000059 Accidental Liquid N/A RadWaste Rel. / 9 N/A Randomly deselected. N/A 0 000060 Accidental Gaseous N/A Radwaste Rel. / 9 N/A Randomly deselected. N/A 0 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 5

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

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

1 2 3 1 2 N/A 000061 ARM System Alarms / 7 N/A Randomly deselected. N/A 0 N/A 000067 Plant Fire On-site / 8 N/A Randomly deselected. N/A 0 N/A 000068 Control Room Evac. / 8 N/A Randomly deselected. N/A 0 Knowledge of the interrelations between the Loss of 000069 (W/E14) Loss of CTMT Containment Integrity and the following: Personnel R24 Integrity / 5 X AK2.03 access hatch and emergency access hatch. 2.8* 1 N/A 000074 Inad. Core Cooling / 4 N/A Randomly deselected. N/A 0 000076 High Reactor Coolant N/A Activity / 9 N/A Randomly deselected. N/A 0 N/A CE/A13 Natural Circ. / 4 N/A Randomly deselected. N/A 0 Equipment Control: Knowledge of the bases in Technical Specifications for limiting conditions for R25 CE/A11 RCS Overcooling - PTS / 4 X 2.2.25 operations and safety limits. 3.2 1 Knowledge of the operational implications of the following concepts as they apply to the (Excess RCS Leakage): Annunciators and conditions indicating signals, and remedial actions associated with the R26 CE/A16 Excess RCS Leakage / 2 X AK1.3 (Excess RCS Leakage). 3.2 1 Knowledge of the interrelations between the (Functional Recovery) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, R27 CE/E09 Functional Recovery X EK2.1 failure modes, and automatic and manual features. 3.6 1 K/A Category Point Totals: 1 2 2 1 1 2 Group Point Total: 9 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 6

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO)

Q# 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 Ability to monitor automatic operation of the RCPS, R28 Pump X A3.05 including: RCP lube oil and bearing lift pumps. 2.7* 1 Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of 003 Reactor Coolant those malfunctions or operations: Problems with R29 Pump X A2.01 RCP seals, especially rates of seal leak-off. 3.5 1 Emergency Procedures / Plan: Ability to recognize abnormal indications for system operating parameters 004 Chemical that are entry-level conditions for emergency and R30 and Volume Control X 2.4.4 abnormal operating procedures. 4.5 1 Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following:

004 Chemical Interlock between letdown isolation valve and flow R31 and Volume Control X K4.13 control valve. 3.2* 1 Knowledge of the physical connections and/or 005 Residual Heat cause-effect relationships between the RHRS and R32 Removal X K1.01 the following systems: CCWS. 3.2 1 Knowledge of the effect of a loss or malfunction on 006 Emergency Core the following will have on the ECCS: HPI/LPI R33 Cooling X K6.05 cooling water. 3.0 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls 007 Pressurizer including: Maintaining quench tank water level within R34 Relief/Quench Tank X A1.01 limits. 2.9 1 008 Component Knowledge of bus power supplies to the following:

R35 Cooling Water X K2.02 CCW pump, including emergency backup. 3.0* 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 7

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

Q# 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 4 010 Pressurizer Knowledge of the effect that a loss or malfunction R36 Pressure Control X K3.02 of the PZR PCS will have on the following: RPS. 4.0 1 Knowledge of RPS design feature(s) and/or 012 Reactor interlock(s) which provide for the following:

R37 Protection X K4.06 Automatic or manual enable/disable of RPS trips. 3.2 1 012 Reactor Ability to manually operate and/or monitor in the R38 Protection X A4.04 control room: Bistable, trips, reset and test switches. 3.3* 1 013 Engineered Safety Features Knowledge of bus power supplies to the following:

R39 Actuation X K2.01 ESFAS/safeguards equipment control. 3.6* 1 013 Engineered Safety Features Emergency Procedures / Plan: Knowledge of R40 Actuation X 2.4.11 abnormal condition procedures. 4.0 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) 022 Containment associated with operating the CCS controls R41 Cooling X A1.04 including: Cooling water flow. 3.2 1 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of 026 Containment those malfunctions or operations: Failure of R42 Spray X A2.02 automatic recirculation transfer. 4.2* 1 Knowledge of the operational implications of the 039 Main following concepts as they apply to the MRSS:

R43 and Reheat Steam X K5.05 Bases for RCS cooldown limits. 2.7 1 039 Main Ability to manually operate and/or monitor in the R44 and Reheat Steam X A4.01 control room: Main steam supply valves. 2.9* 1 Ability to monitor automatic operation of the MFW, R45 059 Main Feedwater X A3.04 including: Turbine driven feed pump. 2.5* 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 8

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

Q# 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 Knowledge of the operational implications of the Auxiliary/Emergency following concepts as they apply to the AFW: Feed R46 Feedwater X K5.05 line voiding and water hammer. 2.7 1 Knowledge of the physical connections and/or cause-effect relationships between the ac 062 AC Electrical distribution system and the following systems: DC R47 Distribution X K1.03 distribution. 3.5 1 Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the 063 DC Electrical consequences of those malfunctions or R48 Distribution X A2.01 operations: Grounds. 2.5 1 Knowledge of the effect that a loss or malfunction 064 Emergency of the ED/G system will have on the following:

R49 Diesel Generator X K3.02 ESFAS controlled or actuated systems. 4.2 1 Knowledge of the effect of a loss or malfunction of 064 Emergency the following will have on the ED/G system: Air R50 Diesel Generator X K6.07 receivers. 2.7 1 073 Process Ability to manually operate and/or monitor in the Radiation control room: Radiation monitoring system control R51 Monitoring X A4.02 panel. 3.7 1 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 R52 076 Service Water X A1.02 water temperatures. 2.6* 1 Conduct of Operations: Ability to interpret reference R53 076 Service Water X 2.1.25 materials, such as graphs, curves, tables, etc. 3.9 1 Ability to monitor automatic operation of the IAS, R54 078 Instrument Air X A3.01 including: Air pressure. 3.1 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 9

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

Q# 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 Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under R55 103 Containment X K3.03 refueling operations. 3.7 1 K/A Category Point Totals: 2 2 3 2 2 2 3 3 3 3 3 Group Point Total: 28 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 10

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO)

Q# 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 Knowledge of the physical connections and/or 001 Control Rod cause-effect relationships between the CRDS and R56 Drive X K1.01 the following systems: CCW. 3.0* 1 Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following:

R57 002 Reactor Coolant X K4.03 Venting the RCS. 2.9 1 011 Pressurizer N/A Level Control N/A Randomly deselected. N/A 0 014 Rod Position N/A Indication N/A Randomly deselected. N/A 0 015 Nuclear Knowledge of bus power supplies to the following:

R58 Instrumentation X K2.01 NIS channels, components, and interconnections. 3.3 1 016 Non-nuclear N/A Instrumentation N/A Randomly deselected. N/A 0 Knowledge of the operational implications of the 017 In-core following concepts as they apply to the ITM R59 Temperature Monitor X K5.01 system: Temperature at which cladding and fuel melt. 3.1 1 027 Containment N/A Iodine Removal N/A Randomly deselected. N/A 0 Ability to manually operate and/or monitor in the 028 Hydrogen control room: Location and operation of hydrogen Recombiner sampling and analysis of containment atmosphere, R60 and Purge Control X A4.03 including alarms and indications. 3.1 1 029 Containment N/A Purge N/A Randomly deselected. N/A 0 Knowledge of the effect that a loss or malfunction 033 Spent Fuel Pool of the Spent Fuel Pool Cooling System will have on R61 Cooling X K3.03 the following: Spent fuel temperature. 3.0 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 11

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

Q# 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 N/A Equipment N/A Randomly deselected. N/A 0 Emergency Procedures / Plan: Ability to perform without reference to procedures those actions that 035 Steam require l immediate operation of system components R62 Generator X 2.4.49 and controls. 4.6 1 Ability to predict and/or monitor changes in 041 Steam parameters (to prevent exceeding design limits)

Dump/Turbine associated with operating the SDS controls R63 Bypass Control X A1.02 including: Steam pressure. 3.1 1 045 Main Turbine N/A Generator N/A Randomly deselected. N/A 0 055 Condenser Air N/A Removal N/A Randomly deselected. N/A 0 N/A 056 Condensate N/A Randomly deselected. N/A 0 Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste R64 068 Liquid Radwaste X K6.10 System: Radiation monitors. 2.5 1 071 Waste Gas N/A Disposal N/A Randomly deselected. N/A 0 072 Area Radiation N/A Monitoring N/A Randomly deselected. N/A 0 075 Circulating N/A Water N/A Randomly deselected. N/A 0 N/A 079 Station Air N/A Randomly deselected. N/A 0 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 12

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

Q# 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 Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure to actuate the FPS when required, R65 086 Fire Protection X A2.04 resulting in fire damage. 3.3 1 K/A Category Point Totals: 1 1 1 1 1 1 1 1 0 1 1 Group Point Total: 10 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 13

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)

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

Function 1 2 3 1 2 000007 (CE/E02) Reactor Trip - Stabilization -

N/A Recovery / 1 N/A Randomly deselected. N/A 0 000008 Pressurizer Vapor N/A Space Accident / 3 N/A Randomly deselected. N/A 0 000009 Small Break LOCA N/A /3 N/A Randomly deselected. N/A 0 000011 Large Break LOCA N/A /3 N/A Randomly deselected. N/A 0 000015/17 RCP N/A Malfunctions / 4 N/A Randomly deselected. N/A 0 000022 Loss of Rx Coolant N/A Makeup / 2 N/A Randomly deselected. N/A 0 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

000025 Loss of RHR Leakage of reactor coolant from RHR into closed cooling S1 System / 4 X AA2.02 water system or into reactor building atmosphere. 3.8 1 000026 Loss of Component Cooling N/A Water / 8 N/A Randomly deselected. N/A 0 000027 Pressurizer Pressure Control System N/A Malfunction / 3 N/A Randomly deselected. N/A 0 N/A 000029 ATWS / 1 N/A Randomly deselected. N/A 0 Ability to determine or interpret the following as they 000038 Steam Gen. Tube apply to a SGTR: Viable alternatives for placing plant in S2 Rupture / 3 X EA2.08 safe condition when condenser is not available. 4.4 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 14

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

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

Function 1 2 3 1 2 000040 (CE/E05) Steam Line Rupture - Excessive Emergency Procedures / Plan: Knowledge of EOP S3 Heat Transfer / 4 X 2.4.6 mitigation strategies. 4.7 1 Emergency Procedures / Plan: Knowledge of low power/shutdown implications in accident (e.g., loss of 000054 (CE/E06) Loss of coolant accident or loss of residual heat removal) mitigation S4 Main Feedwater / 4 X 2.4.9 strategies. 4.2 1 000055 Station Blackout /

N/A 6 N/A Randomly deselected. N/A 0 000056 Loss of Off-site N/A Power / 6 N/A Randomly deselected. N/A 0 000057 Loss of Vital AC Conduct of Operations: Ability to explain and apply S5 Inst. Bus / 6 X 2.1.32 system limits and precautions. 4.0 1 000058 Loss of DC Power N/A /6 N/A Randomly deselected. N/A 0 000062 Loss of Nuclear N/A Svc Water / 4 N/A Randomly deselected. N/A 0 000065 Loss of Instrument N/A Air / 8 N/A Randomly deselected. N/A 0 000077 Generator Voltage Ability to determine and interpret the following as they and Electric Grid apply to Generator Voltage and Electric Grid S6 Disturbances / 6 X AA2.04 Disturbances: VARs outside the capability curve. 3.6 1 K/A Category Totals: 3 3 Group Point Total: 6 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 15

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)

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

1 2 3 1 2 000001 Continuous Rod N/A Withdrawal / 1 N/A Randomly deselected. N/A 0 N/A 000003 Dropped Control Rod / 1 N/A Randomly deselected. N/A 0 000005 Inoperable/Stuck Control N/A Rod / 1 N/A Randomly deselected. N/A 0 N/A 000024 Emergency Boration / 1 N/A Randomly deselected. N/A 0 000028 Pressurizer Level N/A Malfunction / 2 N/A Randomly deselected. N/A 0 000032 Loss of Source Range NI /

N/A 7 N/A Randomly deselected. N/A 0 000033 Loss of Intermediate Conduct of Operations: Ability to interpret and S7 Range NI / 7 X 2.1.20 execute procedure steps. 4.6 1 Ability to determine and interpret the following as 000036 (BW/A08) Fuel Handling they apply to the Fuel Handling Incidents:

S8 Accident / 8 X AA2.02 Occurrence of a fuel handling incident. 4.1 1 000037 Steam Generator Tube N/A Leak / 3 N/A Randomly deselected. N/A 0 Ability to determine and interpret the following as 000051 Loss of Condenser they apply to the Loss of Condenser Vacuum:

S9 Vacuum / 4 X AA2.02 Conditions requiring reactor and/or turbine trip. 4.1 1 000059 Accidental Liquid N/A RadWaste Rel. / 9 N/A Randomly deselected. N/A 0 000060 Accidental Gaseous N/A Radwaste Rel. / 9 N/A Randomly deselected. N/A 0 Equipment Control: Knowledge of the bases in Technical Specifications for limiting conditions for S10 000061 ARM System Alarms / 7 X 2.2.25 operations and safety limits. 4.2 1 N/A 000067 Plant Fire On-site / 8 N/A Randomly deselected. N/A 0 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 16

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

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

1 2 3 1 2 N/A 000068 Control Room Evac. / 8 N/A Randomly deselected. N/A 0 N/A 000069 Loss of CTMT Integrity / 5 N/A Randomly deselected. N/A 0 N/A 000074 Inad. Core Cooling / 4 N/A Randomly deselected. N/A 0 000076 High Reactor Coolant N/A Activity / 9 N/A Randomly deselected. N/A 0 N/A CE/A13 Natural Circ. / 4 N/A Randomly deselected. N/A 0 N/A CE/A11 RCS Overcooling - PTS / 4 N/A Randomly deselected. N/A 0 N/A CE/A16 Excess RCS Leakage / 2 N/A Randomly deselected. N/A 0 N/A CE/E09 Functional Recovery N/A Randomly deselected. N/A 0 K/A Category Point Totals: 2 2 Group Point Total: 4 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 17

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (SRO)

Q# 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 Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Problems associated with RCP motors, including faulty motors 003 Reactor and current, and winding and bearing temperature S11 Coolant Pump X A2.03 problems. 3.1 1 004 Chemical and Volume N/A Control N/A Randomly deselected. N/A 0 005 Residual Heat N/A Removal N/A Randomly deselected. N/A 0 006 Emergency N/A Core Cooling N/A Randomly deselected. N/A 0 007 Pressurizer N/A Relief/Quench Tank N/A Randomly deselected. N/A 0 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of 008 Component those malfunctions or operations: Loss of CCW S12 Cooling Water X A2.01 Pump. 3.6 1 Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of 010 Pressurizer those malfunctions or operations: Spray valve S13 Pressure Control X A2.02 failures. 3.9 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 18

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

Q# 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 4 012 Reactor N/A Protection N/A Randomly deselected. N/A 0 013 Engineered Safety Features N/A Actuation N/A Randomly deselected. N/A 0 022 Containment N/A Cooling N/A Randomly deselected. N/A 0 026 Containment N/A Spray N/A Randomly deselected. N/A 0 039 Main N/A and Reheat Steam N/A Randomly deselected. N/A 0 Conduct of Operations: Ability to perform specific 059 Main system and integrated plant procedures during all S14 Feedwater X 2.1.23 modes of plant operation. 4.4 1 061 Auxiliary/

Emergency N/A Feedwater N/A Randomly deselected. N/A 0 Emergency Procedures / Plan: Ability to diagnose 062 AC Electrical and recognize trends in an accurate and timely manner S15 Distribution X 2.4.47 utilizing the appropriate control room reference material. 4.2 1 063 DC Electrical N/A Distribution N/A Randomly deselected. N/A 0 064 Emergency N/A Diesel Generator N/A Randomly deselected. N/A 0 073 Process Radiation N/A Monitoring N/A Randomly deselected. N/A 0 N/A 076 Service Water N/A Randomly deselected. N/A 0 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 19

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

Q# 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 N/A 078 Instrument Air N/A Randomly deselected. N/A 0 N/A 103 Containment N/A Randomly deselected. N/A 0 K/A Category Point Totals: 3 2 Group Point Total: 5 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 20

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (SRO)

Q# 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 N/A Drive N/A Randomly deselected. N/A 0 002 Reactor N/A Coolant N/A Randomly deselected. N/A 0 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 011 Pressurizer those malfunctions or operations: Loss of one, two or S16 Level Control X A2.04 three charging pumps. 3.7 1 014 Rod Position N/A Indication N/A Randomly deselected. N/A 0 015 Nuclear N/A Instrumentation N/A Randomly deselected. N/A 0 016 Non-nuclear N/A Instrumentation N/A Randomly deselected. N/A 0 017 In-core Temperature N/A Monitor N/A Randomly deselected. N/A 0 027 Containment N/A Iodine Removal N/A Randomly deselected. N/A 0 028 Hydrogen Recombiner N/A and Purge Control N/A Randomly deselected. N/A 0 029 Containment N/A Purge N/A Randomly deselected. N/A 0 033 Spent Fuel N/A Pool Cooling N/A Randomly deselected. N/A 0 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 21

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

Q# 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 Conduct of Operations: Knowledge of refueling S17 Equipment X 2.1.40 administrative requirements. 3.9 1 035 Steam N/A Generator N/A Randomly deselected. N/A 0 041 Steam Dump/Turbine N/A Bypass Control N/A Randomly deselected. N/A 0 Ability to (a) predict the impacts of the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to 045 Main Turbine correct, control, or mitigate the consequences of S18 Generator X A2.15 those malfunctions or operations: Turbine overspeed. 2.6* 1 055 Condenser Air N/A Removal N/A Randomly deselected. N/A 0 N/A 056 Condensate N/A Randomly deselected. N/A 0 068 Liquid N/A Radwaste N/A Randomly deselected. N/A 0 071 Waste Gas N/A Disposal N/A Randomly deselected. N/A 0 072 Area Radiation N/A Monitoring N/A Randomly deselected. N/A 0 075 Circulating N/A Water N/A Randomly deselected. N/A 0 N/A 079 Station Air N/A Randomly deselected. N/A 0 N/A 086 Fire Protection N/A Randomly deselected. N/A 0 K/A Category Point Totals: 2 1 Group Point Total: 3 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 22

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 Date of Exam: March 28, 2011 Category Q# K/A # Topic RO SRO-Only IR # IR #

Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance R66 2.1.4 of active license status, 10CFR55, etc. 3.3 1 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and

1. R67 2.1.26 hydrogen). 3.4 1 Conduct of Operations R68 2.1.40 Knowledge of refueling administrative requirements. 2.8 1 Knowledge of procedures, guidelines, or limitations associated with S19 2.1.37 reactivity management. 4.6 1 S20 2.1.42 Knowledge of new and spent fuel movement procedures. 3.4 1 Category 1 Point Total 3 2 R69 2.2.21 Knowledge of pre- and post-maintenance operability requirements 2.9 1
2. Knowledge of the bases in Technical Specifications for limiting conditions Equipment R70 2.2.25 for operations and safety limits. 3.2 1 Control S21 2.2.23 Ability to track Technical Specification limiting conditions for operations. 4.6 1 Category 2 Point Total 2 1 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 23

Facility: Waterford 3 Date of Exam: March 28, 2011 Category Q# K/A # Topic RO SRO-Only IR # IR #

Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling R71 2.3.12 responsibilities, access to locked high-radiation areas, aligning filters, etc. 3.2 1 Knowledge of radiation monitoring systems, such as fixed radiation

3. monitors and alarms, portable survey instruments, personnel monitoring Radiation R72 2.3.15 equipment, etc. 2.9 1 Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personal monitoring S22 2.3.5 equipment, etc. 2.9 1 S23 2.3.6 Ability to approve release permits. 3.8 1 Category 3 Point Total 2 2 R73 2.4.13 Knowledge of crew roles and responsibilities during EOP usage. 4.0 1 R74 2.4.29 Knowledge of the emergency plan. 3.1 1 4.

Emergency Knowledge of RO tasks performed outside the main control room during Procedures / R75 2.4.34 an emergency and the resultant operational effects. 4.2 1 Plan S24 2.4.29 Knowledge of the emergency plan. 4.4 1 S25 2.4.44 Knowledge of emergency plan protective action recommendations. 4.4 1 Category 4 Point Total 3 2 Tier 3 Point Total 10 7 Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 24

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Group Selected K/A Reason for Rejection Initial Outline Submittal W3 does not have any controllers or positioners that would interrelate with a Pressurizer Vapor Space Accident.

1/1 RO 000008 AK2.03 Randomly reselected to K/A 000008 AK2.01.

Unable to write an operationally valid question to this generic K/A for this subject.. Randomly reselected to K/A 1/1 RO 000009 G2.1.19 000009 G2.1.20.

1/1 RO 000038 EK3.07 W3 does not have RCS loop isolation valves. Randomly reselected to K/A 000038 EK3.08.

W3 does not perform feed and bleed operations for a S/G tube leak. Randomly reselected to K/A 000037 1/2 RO 000037 AK3.04 AK3.09.

Reselected system and K/A due to too much overlap concerning radiation monitors. Randomly reselected to 1/2 RO 000059 AK2.02 CE/E09 EK2.1.

1/2 RO 000069 AK1.01 Unable to write a psychometrically sound question for this topic. Randomly reselected to K/A 000069 AK2.03 At W3 the CVC system has no post accident instrumentation with the exception of containment isolation valve indication. Unable to write a psychometrically sound question for this topic. Randomly reselected to K/A 004 2/1 RO 004 G2.4.3 G2.4.4.

W3 does not have a quench tank cooling system. No other K4 K/As 2.5 or greater in this system. Randomly 2/1 RO 007 K4.01 selected to 007 A1.01.

CE procedures do not address reflux boiling pressure spike when going on recirculation. Unable to write a 2/1 RO 026 A2.01 psychometrically sound question for this topic. Randomly reselected to K/A 026 A2.02.

To address excessive overlap with DC system randomly reselected another system and K/A. Randomly 2/1 RO 063 A4.02 reselected System and K/A to 012 A4.04.

Randomly reselected System 003 and retained K/A A2.01 to address excessive overlap issues with Radiation 2/1 RO 073 A2.01 Monitoring systems.

Randomly reselected system due to excessive overlap for radiation monitors. Additionally there are no immediate 2/2 RO 072 G2.4.49 operator actions associated with radiation monitors. Reselected to System 035 and retained K/A G2.4.49.

This K/A is more appropriate for an SRO candidate. The RO does not manage the Control Room crew during 3/1 RO 2.1.6 transients. Randomly reselected K/A to 2.1.40.

3/4 RO 2.4.26 At W3 the RO is not involved in fire brigade or fire fighting equipment usage. Randomly reselected to K/A 2.4.34.

Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 25

Tier / Randomly Group Selected K/A Reason for Rejection Unable to write a question at the SRO level for actions for this topic. Majority of ATWS actions are immediate operator actions required to be known by the RO. Randomly reselected new system and K/A. Reselected to K/A 1/1 SRO 000007 EA2.04 000038 EA2.08.

1/1 SRO 000040 G2.4.34 Unable to write discriminatory question at the SRO level for this K/A. Randomly reselected K/A to 000040 G2.4.6.

Unable to write discriminatory question to this K/A. The Loss of Vital AC procedures do not contain references to 1/1 SRO 000057 G2.1.23 support this K/A. Randomly reselected to K/A 000057 G2.1.32.

Unable to write a discriminatory question to this K/A at the SRO level. Radiation monitor in letdown system for 1/2 SRO 000076 AA2.04 detecting high RCS activity is no longer used. Randomly reselected to K/A 000036 AA2.02.

2/1 SRO 010 A2.03 W3 does not have PORVs. Randomly reselected to K/A 010 A2.02.

Unable to write a discriminatory question at the SRO level for this K/A. Randomly reselected system and K/A to 2/2 SRO 079 A2.01 045 A2.15.

3/4 SRO 2.4.1 Unable to write an SRO level question to this K/A. The K/A is RO knowledge. Randomly reselected to 2.4.29.

Initial Examination Submittal Replace K/A due to overlap with the Operating Test. This K/A tested knowledge required to performed during the Control Room JPM for the Hydrogen Recombiner startup. Randomly reselected another K/A within K4 category 1/2 RO 028 A4.01 for System 028. Replaced by K/A 028 A4.03.

Unable to write a psychometrically sound question that an RO is required to know for this K/A. Randomly 3/2 RO 2.2.21 reselected K/A 2.2.14.

Final Exam Submittal Per NRC initial written exam submittal comments replaced K/A and question. Randomly selected K/A 003 A3.05 2/1 RO 003 K5.02 as the replacement.

Per NRC initial written exam submittal comments replaced K/A and question. Randomly selected K/A 034 2/2 SRO 034 G2.1.32 G2.1.40 as the replacement.

Facility: Waterford 3 Date of Exam: March 29, 2011 Revision 2 26

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

Initial Conditions: Reactor power is 100%

Protected Train is A AB Bus is aligned to Train A Turnover: Maintain 100% power Event Malf. No. Event Type* Event No. Description I - ATC Pressurizer level instrument RC-ILI-0110 X 1 RC15A2 I - SRO fails low. OP-901-110, Pressurizer Level TS - SRO Control Malfunction.

Steam Generator #1 Feedwater flow I - BOP instrument FW-IFR-1111 fails low. OP-901-2 FW26A I - SRO 201, Steam Generator Level Control TS - SRO Malfunction.

C - BOP CEA 52 Drops into the core 3 RD02A52 C - SRP OP-901-102, CEA or CEDMCS Malfunction TS - SRO R - ATC OP-901-212, Rapid Plant Power Reduction.

3 N/A R - BOP N - SRO Loss of Coolant Accident, OP-902-002, Loss RC23A of Coolant Accident Recovery.

4 M - All CS04A CS-125 A fails closed Secure RCPs (Critical Task 1)

C - ATC Charging Pump A fails to auto-start.

5 CV02A C - SRO C - BOP Low Pressure Safety Injection Pump A fails 6 SI02D to auto start on SIAS requiring manual start C - SRO Containment Spray Pump B trip, OP-902-C - BOP 008, Safety Function Recovery Procedure 7 CS01B C - SRO Alignment of LPSI Pump B to replace CS Pump B. (Critical Task 2)

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 1 Rev 2

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

After taking the shift, Pressurizer level instrument RC-ILI-0110X fails low. Due to the failure, Letdown flow goes to minimum flow and both backup Charging Pumps start.

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 on the Plant Monitoring Computer is affected by this failure.

After the non-faulted channel is selected and Tech Specs are addressed, Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low. The Feedwater Control System will respond by raising Feedwater flow to Steam Generator #1. The SRO should enter OP-901-201, Steam Generator Level Control Malfunction. The BOP will be required to take manual control and match Feedwater and Main Steam flow. The Ultrasonic Flow Meter will fail as a result of the instrument failure and require entry into TRM 3.3.5. The Feedwater controls for Steam Generator #1 will remain in manual as a result of this failure.

After the crew has addressed the Feedwater instrument failure, CEA 52 drops into the core. Off normal procedure OP-901-102, CEA or CEDMCS Malfunction, should be entered. The dropped CEA will require a rapid plant power reduction. The SRO should enter OP-901-212, Rapid Plant Power Reduction. Direct Boration should commence within 15 minutes of the dropped CEA. 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 and manipulate Feedwater to Steam Generator #1 in manual. The SRO should enter Tech Specs 3.2.3, 3.1.3.1, and 3.1.3.5.

Once the crew has commenced the power reduction and lowered power to ~ 90%, or at the lead examiners discretion, a loss of coolant accident will occur. Charging Pump A will fail to start on the lowering Pressurizer level. The crew should diagnose the Pressurizer level dropping with all available Charging Pumps operating, trip the Reactor, and initiate Safety Injection Actuation (SIAS) and Containment Isolation Actuation (CIAS). When Containment Spray is actuated, either manually or automatically, CS-125 A 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 B with CS-125 A 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.

Scenario 1 Rev 2

Scenario Event Description NRC Scenario 1 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 B 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 B closed. The crew should address Containment Temperature and Pressure Control by aligning Low Pressure Safety Injection Pump B to replace the failed Containment Spray Pump B. It is acceptable to pursue these tasks in parallel, since establishing flow with LPSI B to the Containment Spray header will also satisfy Containment Isolation concerns.

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

Scenario 1 Rev 2

NRC Scenario 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. This task becomes applicable after Containment Spray is initiated. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling.

2. Establish Containment temperature and pressure control.

This task is satisfied by aligning LPSI Pump B to replace CS Pump B prior to exiting the Containment Temperature and Pressure Control safety function in OP-902-008. This task becomes applicable following the failure of Containment Spray Pump B. The Functional Recovery procedure utilized following this failure will direct this activity to satisfy the Containment Pressure and Temperature Control safety function.

Scenario Quantitative Attributes

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

NRC Scenario 1 Scenario Notes:

A. Reset Simulator to IC-191.

B. Verify the following Scenario Malfunctions:

1. rc15a for Pressurizer level
2. fw26a for Steam Generator #1 Feedwater flow
3. rd02a52 for CEA 52
4. rc23a for LOCA
5. cv02a for Charging Pump A
6. si02d for Low Pressure Safety Injection Pump A
7. cs01b for Containment Spray Pump B
8. cs04a for CS-125 A C. Verify the following Override:
1. di-08a04s22-1 for CS-125 A D. Ensure Protected Train A sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

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

Scenario 1 Rev 2

NRC Scenario 1 Simulator Booth Instructions Event 1 Pressurizer Level Instrument RC-ILI-0110X Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 1.
2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.
3. If sent to LCP-43, report RC-ILI-0110 X1 is failed low.

Event 2 Steam Generator #1 Feedwater Flow Instrument FW-IFR-1111 Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 2.
2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 CEA 52 Drops, Rapid Plant Power Reduction

1. On Lead Examiner's cue, initiate Event Trigger 3.
2. If called to remove Condensate Polishers from service, acknowledge communication and report that you will perform actions requested.
3. If Work Week Manager or I&C is called, inform the caller that a and a team will be sent to the CEDMCS Alley to investigate.

Event 4 LOCA Inside Containment

1. On Lead Examiner's cue, initiate Event Trigger 4.
2. If called as RCA watch report CS-125 A appears to be mechanically bound, the stem looks bent.
3. If called as RAB watch to check the Emergency Diesel Generators, use remote EGR26 and 27. When EDG A & B Trouble alarms clear, report they are running satisfactorily.
4. If the Duty Plant Manager is called, inform the caller that he will make the necessary calls.

Event 5 Low Pressure Safety Injection Pump A fails to start

1. If called to check the LPSI Pump A breaker, report all indications are normal.
2. If called to check the LPSI Pump A locally, report all indications are normal.

Scenario 1 Rev 2

NRC Scenario 1 Event 6 Containment Spray Pump B Trips

1. After the crew has entered OP-902-002 and on the Lead Examiner's cue, initiate Event Trigger 7.
2. If called to check the Containment Spray Pump B breaker, report over-current flags are picked up on all 3 phases.
3. If called to check the Containment Spray Pump B, report that there are visible charring on the motor with an acrid smell, but no indications of a fire or smoke.
4. If called for TSC concurrence, report SM/EC has granted concurrence.
5. If called as RAB watch to come to the Control Room for over-ride key for CS-125 B, acknowledge communication. Report to the Control Room on lead examiners cue.
6. If crew does obtain key and over-rides CS-125 B closed, use remote CSR13B for the local key operation.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario1.cdf. Save the file into the folder for the appropriate crew.

Scenario 1 Rev 2

NRC Scenario 1 Scenario Timeline:

Ramp Event Malfunction Severity Delay Trigger HH:MM:SS 1 RC15A2 0 N/A N/A 1 Pressurizer level instrument RC-ILI-0110 X fails low 2 FW26A 0 N/A N/A 2 Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low 3 RD02A52 N/A N/A N/A 3 CEA 52 Drops into the core 4 RC23A 3.0 % 8:00 N/A 4 Loss of Coolant Accident 5 CV02A N/A N/A N/A 5 Charging Pump A fails to auto-start 6 SI02D N/A N/A N/A N/A Low Pressure Safety Injection Pump A fails to auto start 7 CS04A N/A N/A N/A N/A DI-08a04s22-1 CS-125 A Fails to open, will not open manually.

7 CS01B N/A N/A N/A 7 Containment Spray Pump B trip Scenario 1 Rev 2

NRC Scenario 1

REFERENCES:

Event Procedures 1 OP-901-110, Pressurizer Level Control Malfunction OP-903-013, Monthly Channel Checks Tech Spec 3.3.3.5 2 OP-901-201, Steam Generator Level Control Malfunction Tech Requirement Manual 3.3.5 3 OP-901-102, CEA or CEDMCS Malfunction OP-901-212, Rapid Plant Power Reduction OP-004-004, Control Element Drive Tech Spec 3.2.3, 3.1.3.1, 3.1.3.5 4 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-002, Loss of Coolant Accident Recovery 5 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 6 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 7 OP-902-008, Safety Function Recovery Procedure OP-902-009, Standard Appendices, Appendix 28, Aligning LPSI to Replace CS Scenario 1 Rev 2

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

Initial Conditions: Reactor power is 77%

Protected Train is B AB Bus is aligned to Train B Turnover: Charging Pumps A & B are operating Boron Equalization is in progress Event Malf. No. Event Type* Event No. Description Pressurizer pressure instrument RC-IPR-I - ATC 1 RX14A 0100 X fails low, OP-901-120, Pressurizer I - SRO Pressure Control Malfunction I - BOP RCP 1A speed instrument failure, Channel B, 2 RC16B I - SRO Core Protection Calculator B trip TS - SRO Letdown Back Pressure controller CVC-IPIC-I - ATC 3 DI-04a3a02e-5 0201 setpoint fails to 100% output. OP-901-I - SRO 112, Charging or Letdown Malfunction.

Dry Cooling Tower Fan 8B failure 4 N/A TS - SRO DI-07a8s06-1 I - BOP Inadvertent Containment Spray Actuation 5 OP-901-504, Inadvertent ESFAS Actuation DI-07a8s12-1 I - SRO Main Steam line break inside Containment, S/G #2, OP-902-004, Excess Steam Demand 6 MS11B M - All Recovery (Critical Task 1, 3, and 4)

C - BOP Initiate Containment Spray flow 7 N/A (Critical Task 2)

C - SRO C - ATC Relay K301 failure, BAM-113 A and CVC-8 RP09E 183 fail to position on Safety Injection C - SRO

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 2 Rev 2

Scenario Event Description NRC Scenario 2 The crew assumes the shift at ~77% power with instructions to hold power pending planned Chemistry Department hold.

After assuming the shift, Pressurizer pressure instrument RC-IPR-0100 X will fail low.

Since Boron Equalization is in progress, the Main Spray valves will close. The SRO will enter OP-901-120, Pressurizer Pressure Control Malfunction, and select the non-faulted pressure channel.

After Channel Y has been selected for Pressurizer pressure control, Reactor Coolant Pump 1A speed sensor for Core Protection Calculator B will fail. CPC B will trip as a result of the failure. The SRO should enter Tech Spec 3.3.1 and have the BOP operator bypass bistables 3 and 4 on Channel B.

After the bypass operation is complete, the Letdown Back Pressure controller, CVC-IPIC-0201, setpoint fails to 700 PSIA, 100% scale. This causes the in service Letdown Back Pressure control valve to close and Letdown flow to go to 0 gpm. The CRS should enter OP-901-112, Charging or Letdown Malfunction, and use sub-section E2 to address the failure. The Letdown Flow controller and the Back Pressure controller will be placed in manual to control Letdown flow.

After the ATC has control of the Letdown System in manual, the Outside Watch will call and report an oil failure on Dry Cooling Tower Fan 8B. The SRO should enter Tech Spec 3.7.4 action d. His review of ambient temperature and Tech Spec 3.7.4 should conclude that Train B Ultimate Heat Sink remains operable and that Tech Spec 3.8.1.1 is being complied with.

After the Tech Spec review is complete, an inadvertent Containment Spray Actuation will occur. Component Cooling Water flow to the Reactor Coolant Pumps will be secured. The SRO should enter OP-901-504, Inadvertent ESFAS Actuation. The Containment Spray Pumps should be secured. If the Component Cooling Water Isolations to the Reactor Coolant Pumps are not restored within 3 minutes, the reactor should be tripped and the Reactor Coolant Pumps secured.

A Main Steam line break will develop on Steam Generator #2 after the preceding event.

If the crew restored CCW to the Reactor Coolant Pumps, the crew should perform a manual reactor trip due to the excess steam demand. If the crew tripped the reactor and secured Reactor Coolant Pumps on the previous event, then the Main Steam line break will ramp in after the reactor trip. Because the Containment Spray Pumps control switches maintain off, the BOP should re-start Containment Spray Pumps A and B after Containment pressure rises above 17.7 psia.

Relay K301 will not actuate and BAM-113 A will fail to open and CVC-183 will fail to close on the Safety Injection Actuation. The ATC operator should position these valves to ensure Emergency Boration. After Steam Generator #2 blows down, the crew will take action to maintain RCS temperature and pressure. The scenario can be terminated after these actions are complete.

Scenario 2 Rev 2

NRC Scenario 2 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. If the crew does not restore CCW flow to the RCPs after the inadvertent CSAS, then the 3 minute criteria starts at the time of that CSAS. If the crew restores CCW flow to the RCPs following the inadvertent CSAS, then the 3 minute criteria starts after the Main Steam line break.

2. Establish Containment temperature and pressure control This task is satisfied by manually starting at least 1 Containment Spray Pump following the Main Steam line break. This should be completed before completing the review of OP-902-000, Standard Post Trip Actions.
3. Establish RCS temperature control This task is satisfied by taking action to stabilize RCS temperature within the limits of the RCS P/T curve using ADV #1 and establishing EFW flow to Steam Generator #1.

Action to address this task should commence prior to RCS temperature exceeding 550 °F.

4. Establish RCS pressure control This task is satisfied by taking action to stabilize RCS pressure within the limits of the RCS P/T curve and additionally maintain RCS pressure within 1500-1600 psid of the faulted steam generator. Action to address this task should commence prior to RCS pressure exceeding 2250 PSIA.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 7
2. Malfunctions after EOP entry (1-2) 2
3. Abnormal events (2-4) 3
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) 4 Scenario 2 Rev 2

NRC Scenario 2 Scenario Notes:

A. Reset Simulator to IC-192.

B. Verify the following Scenario Malfunctions:

1. rx14-A for Pressurizer pressure instrument RC-IPT-0100 X
2. rc16b for RCP 1A speed
3. ms11b for Main Steam line break S/G #2
4. rp09e for Relay K301 C. Verify the following Overrides:
1. di-07a08s06-1 and di-07a08s12-1 for CSAS
2. di-04a3a02e-5 for Letdown Back Pressure Controller D. Ensure Protected Train B sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

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

Scenario 2 Rev 2

NRC Scenario 2 Simulator Booth Instructions Event 1 Pressurizer Pressure Instrument Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 1.
2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 RCP 1A Speed Instrument Failure

3. On Lead Examiner's cue, initiate Event Trigger 2.
4. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 Letdown Back Pressure Controller Setpoint Failure

1. On Lead Examiner's cue, initiate Event Trigger 3.
2. If called as the RCA Watch to locally monitor the following Letdown DP indications, report that all DP indications are normal.
3. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 4 Dry Cooling Tower Fan 8B Fan Failure

1. On Lead Examiner's cue, call the CRS as the Outside Watch and report that Dry Cooling Tower Fan 8B has no oil in the reduction gear sightglass. There is oil on the ground under the fan. This discovery is made during rounds.
2. If Work Week Manager or PMM is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 Inadvertent CSAS

1. On Lead Examiner's cue, initiate Event Trigger 5.
2. No communications should occur for this evolution.

Event 6 Main Steam Line Break S/G #2

1. On the Lead Examiner's cue, or after the reactor is manually tripped in the previous event, initiate Event Trigger 6.
2. When called as the Outside Watch to check Main Steam Safeties not lifting, report that no safety valves are lifting.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario 2.cdf. Save the file into the folder for the appropriate crew.

Scenario 2 Rev 2

NRC Scenario 2 Scenario Timeline:

Ramp Event Malfunction Severity Delay Trigger HH:MM:SS 1 RX14A 0% N/A N/A 1 Pressurizer pressure RC-IPR-0100 X fails low 2 RC16B N/A N/A N/A 2 RCP 1A Speed failure, Channel B 3 Di-04a3a02e-5 Push N/A N/A 3 Letdown Back Pressure controller setpoint failure 4 N/A N/A N/A N/A N/A Dry Cooling Tower Fan 8B failure 5 Di-07a8a06-1 N/A N/A N/A 5 DI-07a8s12-1 Inadvertent Containment Spray 6 MS11B 10% 3:00 N/A 6 Main Steam line break, S/G #2 7 N/A N/A N/A N/A N/A Initiate Containment Spray flow 8 RP09E N/A N/A N/A N/A Relay K301 failure Scenario 2 Rev 2

NRC Scenario 2

REFERENCES:

Event Procedures 1 OP-901-120, Pressurizer Pressure Control Malfunction 2 OP-009-007, Plant Protection System Tech Spec 3.3.1 3 OP-901-112, Charging or Letdown Malfunction 4 Tech Spec 3.7.4 and 3.8.1.1 OP-100-014, Technical Specification and Technical Requirements Compliance 5 OP-901-504, Inadvertent ESFAS Actuation 6 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-004, Excess Steam Demand Recovery 7 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

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

Guidance Scenario 2 Rev 2

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

Initial Conditions: Reactor power is 100%

Protected Train is B AB Bus is aligned to Train A Turnover: Maintain 100% power Event Malf. No. Event Type* Event No. Description SG10D C - BOP Steam Generator #1 level instrument SG-ILI-1 C - SRO 1113 D fails high.

TS - SRO FW03A C - ATC Main Feedwater Pump A trips, Reactor 2 C - SRO Power Cutback TS - SRO OP-901-101, Reactor Power Cutback RD07D R - ATC Regulating Group 4 CEAs fail to insert in 3 automatic following Reactor Power Cutback TP01A C - BOP Turbine Cooling Water Pump A trips, Turbine 4 TP08B C - SRO Cooling Water Pump B fails to auto start OP-901-512, Loss of Turbine Cooling Water FW03B M - All Main Feedwater Pump B trips, manual 5 FW07A N - SRO reactor trip, Emergency Feedwater Pump A fails to run RP03 C - BOP Main Turbine fails to trip following the reactor 6 C - SRO trip RD11A C - ATC 3 CEAs fail to insert following the reactor trip, 7 28, 37, 79 C - SRO Emergency Boration (Critical task 1)

FW05 C - BOP Emergency Feedwater Pump AB trip on 8 C - ATC overspeed C - SRO (Critical task 2)

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

Scenario Event Description NRC Scenario 4 The crew assumes the shift at 100% power with instructions to maintain 100% power.

After assuming the shift, Steam Generator #1 level instrument SG-ILI-1113 D fails high.

The SRO should review Tech Specs and enter Tech Spec 3.3.1 and 3.3.2 and TRM 3.3.1. The SRO should direct the BOP operator to bypass the bistables for low Steam Generator #1 level, high level and Steam Generator #1 differential pressure for channel D. This instrument does apply to Tech Spec 3.3.3.6 for Accident Monitoring, but the minimum channel requirements are met using other channels.

After the proper bistables are bypassed, Main Feedwater Pump A will trip. A Reactor Power Cutback will occur. The ATC should perform the immediate operator actions.

The SRO should enter OP-901-101, Reactor Power Cutback. Following the Cutback, Regulating Group 4 CEAs will fail to insert in automatic. The SRO should enter Tech Spec 3.2.4 for DNBR and 3.2.7 for ASI. The crew should take action to address the DNBR power operating limit within 15 minutes by performing ASI control with Group P CEAs.

After the crew has addressed Tech Specs and commenced ASI control with Group P CEAs, Turbine Cooling Water Pump A trips. Turbine Cooling Water Pump B fails to start. The SRO should enter OP-901-512, Loss of Turbine Cooling Water, and start Turbine Cooling Water Pump B. The Plant Monitoring Computer will display an overload condition for TCW Pump A After Turbine Cooling Water Pump B is running, Main Feedwater Pump B will trip. The crew should perform a manual reactor trip based on this failure. On the Emergency Feedwater Actuation, Emergency Feedwater Pump A will fail to start and will not start manually. The Main Turbine will fail to trip on the reactor trip. The BOP should manually trip the Main Turbine. 3 CEAs will fail to insert on the reactor trip. The ATC operator should perform Emergency Boration due to this condition. The SRO should enter OP-902-006, Loss of Main Feedwater Recovery. The ATC operator should secure 2 Reactor Coolant Pumps.

After 2 Reactor Coolant Pumps are secured, Emergency Feedwater Pump AB will trip due to operator error locally. The crew should remain in OP-902-006 and secure the remaining Reactor Coolant Pumps. On investigation, the local watchstander will report Emergency Feedwater Pump AB is ready to be reset. The BOP operator should perform the necessary actions for resetting Emergency Feedwater Pump AB.

The scenario can be terminated after Emergency Feedwater Pump AB is reset.

Scenario 4 Rev 2

NRC Scenario 4 Critical Tasks

1. Establish reactivity control.

This task is satisfied by establishing Emergency Boration prior to completing Standard Post Trip Actions Reactivity Control verification. The required task becomes applicable after the Reactor is tripped and 3 CEAs remain stuck out.

2. Establish a primary to secondary heat sink This task is satisfied by securing all RCPs after Emergency Feedwater Pump AB trips. With Emergency Feedwater Pump A off, Emergency Feedwater Pump B does not have the capacity to provide necessary Emergency Feedwater flow. The requirement is that all RCPs be secured within 30 minutes of the loss of Main Feedwater, the time of the reactor trip.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 8
2. Malfunctions after EOP entry (1-2) 3
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 Scenario 4 Rev 2

NRC Scenario 4 Scenario Notes:

A. Reset Simulator to IC-194.

B. Verify the following Scenario Malfunctions:

1. sg10d for S/G #1 level instrument
2. tp01a for TCW Pump A
3. tp08b for TCW Pump B
4. fw03a for Main Feedwater Pump A
5. rd07d for Regulating Group 4 CEAs
6. fw03b for Main Feedwater Pump B
7. fw07a for EFW Pump A
8. rp03 for the Main Turbine failure
9. rd11a28, 37, and 79 for CEAs 28, 37, and 79
10. fw05 for EFW Pump AB C. Verify the following Override:
1. di-08a04s09-1 for EFW Pump A D. Ensure Protected Train B sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

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

Scenario 4 Rev 2

NRC Scenario 4 Simulator Booth Instructions Event 1 Steam Generator #1 level instrument failure

1. On the Lead Examiner's cue, initiate Event Trigger 1.
2. If directed to check the remote shutdown panel, report that Channel D S/G #1 level reads 67%.
3. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2/3 Main Feedwater Pump A trip, Reactor Power Cutback/Reg Group 4 Failure

1. On the Lead Examiner's cue, initiate Event Trigger 3.
2. If directed to check Main Feedwater Pump A locally, report there are no abnormal indications locally.

Event 4 Turbine Cooling Water Pump A trip

1. On the Lead Examiner's cue, initiate Event Trigger 2.
2. If directed to check Turbine Cooling Water Pumps locally, report TCW Pump A has over-current flags tripped and that TCW Pump B looks normal.
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 5 MFW Pump B trip, Reactor trip, Emergency Feedwater Pump A trip

1. On the Lead Examiner's cue, initiate Event Trigger 5.
2. If directed to check Main Feedwater Pump B locally, report indications of broken linkages on the governor assembly.
3. If directed to check EFW Pump A locally, report indications of a broken breaker for EFW Pump A at Switchgear 3A.

Event 8 Emergency Feedwater Pump AB trip

1. On the Lead Examiner's cue, initiate Event Trigger 8.
2. After the remaining Reactor Coolant Pumps are tripped, call as the RCA watch and report that the Emergency Feedwater Pump AB tripped on overspeed due to his activities while checking the pump. Recommend performing actions to reset EFW Pump AB.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario 4.cdf. Save the file into the folder for the appropriate crew.

Scenario 4 Rev 2

NRC Scenario 4 Scenario Timeline:

Ramp Event Malfunction Severity Delay Trigger HH:MM:SS 1 SG10D 100% N/A N/A 1 S/G #1 level instrument channel D fails high FW03A N/A N/A N/A 2 MFW Pump A trips 3 RD07D N/A N/A N/A N/A Regulating Group 4 fails to auto insert 4 TP01A N/A N/A N/A 4 TP08B TCW Pump A trips, TCW Pump B fails to auto-start 5 FW03B N/A N/A N/A 5 FW07A DI-08a04s09-1 MFW Pump B trips, EFW Pump A fails to run 6 RP03 N/A N/A N/A N/A Main Turbine fails to trip on reactor trip 7 RD11A N/A N/A N/A N/A 28, 37, 79 CEAs 28, 37, 79 fail to insert 8 FW05 N/A N/A N/A 8 EFW Pump AB trips Scenario 4 Rev 2

NRC Scenario 4

REFERENCES:

Event Procedures 1 OP-009-007, Plant Protection System OP-903-013, Monthly Channel Checks Tech Spec 3.3.1 and 3.3.2 3&4 OP-901-101, Reactor Power Cutback Tech Spec 3.2.1 2 OP-901-512, Loss of Turbine Cooling Water 5 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-006, Loss of Main Feedwater Recovery 6 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 7 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 8 OP-902-006, Loss of Main Feedwater Recovery Scenario 4 Rev 2

ES-301 Administrative Topics Outline Form ES-301-1 Facility: WATERFORD 3 Date of Examination: March 21, 2011 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 R, D integrated plant procedures during all modes of plant operation.

Conduct of Operations Perform a Shutdown Margin with an immoveable CEA in accordance with OP-903-090, Shutdown Margin, K/A Importance: section 7.3, Shutdown Margin Verification -

4.3 Untrippable CEA.

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

Perform OP-903-001, Technical Specification K/A Importance: Surveillance Logs, Attachment 11.18, Adjustment of 3.6 CPC and Excore Nuclear Instrumentation Data.

A3 S, N 2.2.12, Knowledge of surveillance procedures Equipment Control Complete surveillance OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring K/A Importance: Instrumentation Channel Checks.

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

Calculate stay time to perform a tagout verification.

K/A Importance: Room dose rate & operators yearly dose provided.

3.2 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 1 RO

ES-301 Administrative Topics Outline Form ES-301-1 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Examination Level: SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A5 2.1.25, Ability to interpret reference materials, such as R, D graphs, curves, tables, etc.

Conduct of Operations Review and approve a completed Shutdown Margin with K/A Importance:

an immoveable CEA in accordance with OP-903-090, 3.9 Shutdown Margin, section 7.3, Shutdown Margin Verification - Un-trippable CEA.

A6 2.1.18, Ability to make accurate, clear, and concise logs, S, M records, status boards, and reports.

Conduct of Operations Review and approve OP-903-001, Technical K/A Importance:

Specification Surveillance Logs, Attachment 11.18, 3.8 Adjustment of CPC and Excore Nuclear Instrumentation Data.

A7 2.2.37, Ability to determine operability and/or availability S, M of safety related equipment.

Equipment Control Review and approve a completed Equipment Out of K/A Importance:

Service document in accordance with OP-100-010, 4.6 Equipment Out of Service.

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

Radiation Control Calculate dose and assign non-licensed operators to K/A Importance:

perform work in radiological restricted areas. Given 3.7 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.

A9 2.4.38, Ability to take actions called for in the facility S, M emergency plan, including supporting or acting as Emergency Plan emergency coordinator if required.

K/A Importance:

Determine appropriate classification and actions based 4.4 on a toxic gas release in accordance with EP-004-010, Toxic Chemical Contingency Procedure.

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 1 SRO

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Exam Level Reactor Operator 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 Safety Code* Function S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in A, D, S 1 accordance with OP-903-005, Control Element Assembly Operability Check.

Fault: CEA 20 will insert after initially moved, CEA will subsequently drop, the combination requiring a reactor trip.

A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 S2 004 Chemical and Volume Control System; Makeup to the Volume A, M, S 2 Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control.

Fault: The Boric Acid counter will fail to secure the Boric Acid addition, requiring the applicant to manually secure Boric Acid flow. The applicant will then need to add the Primary Makeup Water for the initial calculation, plus the additional based on the extra boric acid added.

A4.07 Boration/dilution RO - 3.9, SRO - 3.7 S3 005 Shutdown Cooling System; Secure Shutdown Cooling Train B and D, L, S 4-P place it in standby in accordance with OP-009-005, Shutdown Cooling.

A4.01 Controls and indication for RHR pumps RO - 3.6, SRO - 3.4 S4 039 Main and Reheat Steam System; BOP operator immediate operator A, M, S 4-S actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 028 Hydrogen Recombiner and Purge Control System D, L, P, S 5 Start Hydrogen Recombiner A in accordance with OP-008-006.

A4.01 HRPS controls RO - 4.0, SRO - 4.0 S6 064 Emergency Diesel Generator (ED/G) System; Parallel Emergency A, D, S 6 Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator.

Fault: After EDG A load is raised, EDG A load will rise without manipulation requiring a trip of EDG A.

A4.06 Manual start, loading, and stopping of the ED/G RO - 3.9, SRO - 3.9 S7. 029 Containment Purge System; Perform surveillance OP-903-052, N, S 8 Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A.

K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 S8. 012 Reactor Protection System; Place Reactor Power Cutback in service D, S 7 and remove reactor trip on turbine trip in accordance with OP-004-015, Reactor Power Cutback.

A4.03 Channel blocks and bypasses RO - 3.6, SRO - 3.6 1 NRC 2011 Revision 1

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 D, E, L, 4-S Emergency Feedwater Pump AB in accordance with OP-902-005, P, R Station Blackout Recovery.

A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency A, D, R 6 Diesel Generator B locally.

Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B.

K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 P3 068 Control Room Evacuation E, L, N 2 Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0

@ 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 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 7 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature - / - / 1 (control room system) -

(L)ow-Power / Shutdown 1/ 1/ 1 4 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 4 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 2 (R)CA 1/ 1/ 1 2 (S)imulator 8 2 NRC 2011 Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: March 21, 2011 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 Safety Code* Function S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in A, D, S 1 accordance with OP-903-005, Control Element Assembly Operability Check.

Fault: CEA 20 will insert after initially moved, CEA will subsequently drop, the combination requiring a reactor trip.

A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 S2 004 Chemical and Volume Control System; Makeup to the Volume A, M, S 2 Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control.

Fault: The Boric Acid counter will fail to secure the Boric Acid addition, requiring the applicant to manually secure Boric Acid flow. The applicant will then need to add the Primary Makeup Water for the initial calculation, plus the additional based on the extra boric acid added.

A4.07 Boration/dilution RO - 3.9, SRO - 3.7 S3 005 Shutdown Cooling System; Secure Shutdown Cooling Train B and D, L, S 4-P place it in standby in accordance with OP-009-005, Shutdown Cooling.

A4.01 Controls and indication for RHR pumps RO - 3.6, SRO - 3.4 S4 039 Main and Reheat Steam System; BOP operator immediate operator A, M, S 4-S actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 S6 064 Emergency Diesel Generator (ED/G) System; Parallel Emergency A, D, S 6 Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator.

Fault: After EDG A load is raised, EDG A load will rise without manipulation requiring a trip of EDG A.

A4.06 Manual start, loading, and stopping of the ED/G RO - 3.9, SRO - 3.9 S7. 029 Containment Purge System; Perform surveillance OP-903-052, N, S 8 Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A.

K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 S8. 012 Reactor Protection System; Place Reactor Power Cutback in service D, S 7 and remove reactor trip on turbine trip in accordance with OP-004-015, Reactor Power Cutback.

A4.03 Channel blocks and bypasses RO - 3.6, SRO - 3.6 3 NRC 2011 Revision 1

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 D, E, L, 4-S Emergency Feedwater Pump AB in accordance with OP-902-005, P, R Station Blackout Recovery.

A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency A, D, R 6 Diesel Generator B locally.

Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B.

K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 P3 068 Control Room Evacuation E, L, N 2 Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0

@ 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 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 6 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature - / - / 1 (control room system) -

(L)ow-Power / Shutdown 1/ 1/ 1 3 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 4 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 1 (R)CA 1/ 1/ 1 2 (S)imulator 7 4 NRC 2011 Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: March 21, 2011 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 Safety Code* Function S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in A, D, S 1 accordance with OP-903-005, Control Element Assembly Operability Check.

Fault: CEA 20 will insert after initially moved, CEA will subsequently drop, the combination requiring a reactor trip.

A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 S2 S3 S4 039 Main and Reheat Steam System; BOP operator immediate operator A, M, S 4-S actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 S6 S7. 029 Containment Purge System; Perform surveillance OP-903-052, N, EN, S 8 Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A.

K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 S8.

5 NRC 2011 Revision 1

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 Emergency Diesel Generator (ED/G) System; Trip Emergency A, D, R 6 Diesel Generator B locally.

Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B.

K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 P3 068 Control Room Evacuation E, L, N 2 Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0

@ 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 0 (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 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 3 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 0 (R)CA 1/ 1/ 1 1 (S)imulator 3 6 NRC 2011 Revision 1
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