ML17137A184

2017-03 Draft Outlines - Do Not Release Until 04/30/2019

ML17137A184
Person / Time
Site:	Waterford
Issue date:	04/05/2017
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML17137A184 (71)

Text

ES-401PWR Examination OutlineForm ES-401-2Facility: Waterford 3 (RO Exam) Date of Exam: March 27, 2017TierGroupRO 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 &AbnormalPlantEvolutions 1 3 3 3N/A3 3N/A 3 18 6 2 2 2 11 2 1 9 4Tier Totals 5 5 44 5 4 27 10 2.PlantSystems 13 3 2 3 2 2 2 2 3 3 3 28 5 21 0 1 1 1 1 1 1 1 1 1 10 3Tier Totals 4 3 3 4 3 3 3 3 4 4 4 38 83. Generic Knowledge and AbilitiesCategories 1 3 2 3 3 2 4 2 10 1 2 3 4 7Note: 1.Ensure that at least two topics from every applicable K/A category are sampled within each tier of the ROand SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" ineach K/A category shall not be less than two).(One Tier 3 Radiation Control K/A is allowed if the K/A isreplaced by a K/A from another Tier 3 Category).

2.The point total for each group and tier in the proposed outline must match that specified in the table. Thefinal point total for each group and tier may deviate by +/-1 from that specified in the table based on NRCrevisions. 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 thatdo not apply at the facility should be deleted with justification; operationally important, site-specificsystems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401for 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 thegroup 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 beselected. 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 topicsmust be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicableK/As.8.On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importanceratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enterthe group and tier totals for each category in the table above; if fuel handling equipment is sampled in acategory other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 forTier 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.

G*Generic K/As ES-401 2Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G*K/A Topic(s)

IR#000007 (BW/E02&E10; CE/E02) ReactorTrip - Stabilization - Recovery / 1 X CE/E02, EA1.2: Ability to operate and

/ or monitor the following as they apply to the (Reactor Trip Recovery):

Operating behavior characteristics of the facility (CFR: 41.7 / 45.5 / 45.6) 3.3 1000008 Pressurizer Vapor SpaceAccident / 3 X AK2.01: Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Valves (CFR 41.7 / 45.7) 2.7 2000009 Small Break LOCA / 3 X EK3.04: Knowledge of the reasons for the following responses as the apply to the small break LOCA:

Starting additional Charging Pumps (CFR 41.5 /

41.10 / 45.6 / 45.13) 4.1 3000011 Large Break LOCA / 3 X EA2.03: Ability to determine or interpret the following as they apply to a Large Break LOCA: Consequences of managing LOCA with loss of CCW (CFR 43.5 / 45.13) 3.7 4000015/17 RCP Malfunctions / 4 X AK2.07: Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following:

RCP seals (CFR 41.7 / 45.7) 2.9 5000022 Loss of Rx Coolant Makeup / 2 X AK1.03: Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Relationship between charging flow and PZR level (CFR 41.8 / 41.10 / 45.3) 3.0 6000025 Loss of RHR System / 4 X 2.1.27: Knowledge of system purpose and/or function. (CFR: 41.7).

3.9 7000026 Loss of Component CoolingWater / 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. (CFR: 43.5 / 45.13) 2.9 16000027 Pressurizer Pressure ControlSystem Malfunction / 3 X 2.2.44: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions (CFR: 41.5 / 43.5 / 45.12) 4.2 8000029 ATWS / 1 X EA1.12: Ability to operate and monitor the following as they apply to a ATWS: M/G set power supply and reactor trip breakers (CFR 41.7 / 45.5

/ 45.6)4.1 9000038 Steam Gen. Tube Rupture / 3 X EK3.02: Knowledge of the reasons for the following responses as the apply to the SGTR: Prevention of secondary PORV cycling (CFR 41.5 / 41.10 / 45.6

/ 45.13)4.4 10 000040 (BW/E05; CE/E05; W/E12)Steam Line Rupture - Excessive HeatTransfer / 4 X CE/E05, EK1.2: Knowledge of the operational implications of the following concepts as they apply to the (Excess Steam Demand): Normal, abnormal and emergency operating procedures associated with (Excess Steam Demand)(CFR: 41.8 / 41.10 /

45.3)3.2 11000054 (CE/E06) Loss of MainFeedwater / 4 X CE/E06, EK1.1: Knowledge of the operational implications of the following concepts as they apply to the (Loss of Feedwater): Components, capacity, and function of emergency systems.(CFR: 41.8 / 41.10 / 45.3) 3.2 12000055 Station Blackout / 6 X EA2.03: Ability to determine or interpret the following as they apply to a Station Blackout: Actions necessary to restore power (CFR 43.5 /

45.13)3.9 13000056 Loss of Off-site Power / 6 X 2.4.18: Knowledge of the specific bases of the EOPs (CFR: 41.10 / 43.1 /

45.13)3.3 14000057 Loss of Vital AC Inst. Bus / 6000058 Loss of DC Power / 6 X AK3.01: Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: Use of dc control power by D/Gs (CFR 41.5,41.10

/ 45.6 / 45.1) 3.4 15000062 Loss of Nuclear Svc Water / 4000065 Loss of Instrument Air / 8 X AA1.01: Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air:

Remote manual loaders 2.7 17W/E04 LOCA Outside Containment / 3W/E11 Loss of Emergency CoolantRecirc. / 4BW/E04; W/E05 Inadequate HeatTransfer - Loss of Secondary Heat Sink / 4000077 Generator Voltage and ElectricGrid Disturbances / 6 X AK2.07: Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following:

Turbine/Generator Control (CFR: 41.4, 41.5, 41.7, 41.10

/ 45.8)3.6 18K/A Category Totals:

3 3 3 33 3Group Point Total:

18/6 ES-401 3Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / 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 / 1000003 Dropped Control Rod / 1000005 Inoperable/Stuck Control Rod / 1000024 Emergency Boration / 1 X AK2.01: Knowledge of the interrelations between Emergency Boration and the following:

Valves (CFR 41.7 / 45.7) 2.7 20000028 Pressurizer Level Malfunction / 2 X AA2.12: Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: Cause for PZR level deviation alarm:

controller malfunction or other instrumentation malfunction (CFR:

43.5 / 45.13) 3.1 21000032 Loss of Source Range NI / 7000033 Loss of Intermediate Range NI / 7000036 (BW/A08) Fuel Handling Accident / 8000037 Steam Generator Tube Leak / 3 X AK1.02: Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak:

Leak rate vs. pressure drop (CFR 41.8 / 41.10 / 45.3) 3.5 23000051 Loss of Condenser Vacuum / 4 X AK3.01: Knowledge of the reasons for the following responses as they apply to the Loss of Condeser Vacuum: Loss of Steam Dump capability upon loss of condenser vacuum. (CFR 41.5,41.10

/ 45.6 / 45.13) 2.8 22000059 Accidental Liquid Radwaste Rel. / 9000060 Accidental Gaseous Radwaste Rel. / 9 X 2.4.31: Knowledge of annunciator alarms, indications, or response procedures (CFR: 41.10 / 45.3) 4.2 24000061 ARM System Alarms / 7000067 Plant Fire On-site / 8000068 (BW/A06) Control Room Evac. / 8 X AK2.03 Knowledge of the interrelations between the Control Room Evacuation and the following:

Controllers and positioners (CFR 41.7 / 45.7) 2.9 25000069 (W/E14) Loss of CTMT Integrity / 5000074 (W/E06&E07) Inad. Core Cooling / 4 X EA1.01: Ability to operate and monitor the following as they apply to a Inadequate Core Cooling: RCS Water Inventory (CFR 41.7 / 45.5 / 45.6) 4.2 26000076 High Reactor Coolant Activity / 9W/EO1 & E02 Rediagnosis & SI Termination / 3W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5W/E16 High Containment Radiation / 9BW/A01 Plant Runback / 1BW/A02&A03 Loss of NNI-X/Y / 7BW/A04 Turbine Trip / 4BW/A05 Emergency Diesel Actuation / 6BW/A07 Flooding / 8BW/E03 Inadequate Subcooling Margin / 4BW/E08; W/E03 LOCA Cooldown - Depress. / 4BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4BW/E13&E14 EOP Rules and EnclosuresCE/A11; W/E08 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 and operation within the limitations in the facility's license and amendments (CFR: 43.5

/ 45.13)3.0 27CE/A16 Excess RCS Leakage / 2 X AK1.1: Knowledge of the operational implications of the following concepts as they apply to the (Excess RCS Leakage):

Components, capacity, and function of emergency systems (CFR 41.8 / 41.10 / 45.3) 3.2 19CE/E09 Functional RecoveryK/A Category Point Totals:

2 2 1 1 2 1Group Point Total:

9/4 ES-401 4Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Plant Systems

-Tier 2/Group 1 (RO/ 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 X K6.14: Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: Starting requirements (CFR: 41.7 / 45/5) 2.6 28004 Chemical and VolumeControl X X A4.07: Ability to manually operate and/or monitor in the control room:

Boration/dilution (CFR: 41/7

/ 45.5 to 45.8)

K5.30: Knowledge of the operational implications of the following concepts as they apply to the CVCS:

Relationship between temperature and pressure in CVCS components during solid plant operation 3.9 3.8 29 51005 Residual Heat Removal X X A4.02: Ability to manually operate and/or monitor in the control room:

Heat exchanger bypass flow control (CFR: 41.7 / 45.5 to 45.8) 2.1.20: Ability to interpret and execute procedure steps.(CFR: 41.10 / 43.5 / 45.12) 3.4 4.6 30 31006 Emergency Core Cooling X K4.17: Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Safety Injection valve interlocks (CFR: 41.7) 3.8 32007 Pressurizer Relief/QuenchTank X A2.03: 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:

Overpressurization of the PZR (CFR: 41.5 / 43.5 / 45.3 /

45.13)3.6 33 008 Component Cooling Water X X K1.05: Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: Sources of makeup water(CFR: 41.2 to 41.9 /

45.7 to 45.9)

A3.08: Ability to monitor automatic operation of the CCWS, including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal (CFR: 41.7 / 45.5) 3.0 3.6 34 35010 Pressurizer Pressure Control X X K6.01: Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:

Pressure detection systems (CFR: 41.7 / 45.7)

A1.01: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including:

PZR and RCS boron concentrations (CFR: 41.5 /

45.5)2.7 2.8 36 37012 Reactor Protection X K1.02: Knowledge of the physical connections and/or cause effect relationships between the RPS and the following systems: 125V dc system (CFR:

41.2 to 41.9 / 45.7 to 45.8) 3.4 38013 Engineered Safety FeaturesActuation X X K2.01: Knowledge of bus power supplies to the following:

ESFAS/safeguards equipment control (CFR: 41.7)

A2.02: Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Excess steam demand (CFR: 41.5 / 43.5 /

45.3 / 45.13) 3.6 4.3 39 40022 Containment Cooling X K3.02: Knowledge of the effect that a loss or malfunction of the CCS will have on the following: Containment instrumentation readings (CFR: 41.7 / 45.6) 3.0 41025 Ice Condenser026 Containment Spray X K2.01: Knowledge of bus power supplies to the following:

Containment spray pumps (CFR:

41.7)3.4 42 039 Main and Reheat Steam X K5.08: Knowledge of the operational implications of the following concepts as the apply to the MRSS:

Effect of steam removal on reactivity (CFR: 441.5 /

45.7)3.6 43059 Main Feedwater X K4.18: Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Automatic feedwater reduction on plant trip (CFR: 41.7) 2.8 44061 Auxiliary/EmergencyFeedwater X X K2.02: Knowledge of bus power supplies to the following:

AFW electric drive pumps(CFR:

41.7)A3.03: Ability to monitor automatic operation of the AFW, including: AFW S/G level control on automatic start (CFR: 41.7 / 45.5) 3.7 3.9 46 47062 AC Electrical Distribution X A3.05: Ability to monitor automatic operation of the ac distribution system, including: Safety-related indicators and controls (CFR:

41.7 / 45.5) 3.5 48063 DC Electrical Distribution X A4.02: Ability to manually operate and/or monitor in the control room: Battery voltage indicator (CFR: 41.7 / 45.5 to 45.8)2.8 49064 Emergency Diesel Generator X K4.10: Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following:

Automatic load sequencer:

blackout (CFR: 41.7) 3.5 50073 Process Radiation Monitoring X 2.4.4: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. l(CFR: 41.10 / 43.2 / 45.6) 4.5 52 076 Service Water X X K1.19: Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: SWS Emergency heat loads (CFR:

41.2 to 41.9 / 45.7 to 45.8)

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 3.6 2.6 53 45078 Instrument Air X K3.01: Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Containment air system (CFR: 41.7 / 45.6) 3.1 54103 Containment X 2.2.44: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions (CFR: 41.5 / 43.5 / 45.12) 4.2 55K/A Category Point Totals:

3 3 2 3 2 2 2 2 3 3 3Group Point Total:28/5 ES-401 5Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Plant Systems

-Tier 2/Group 2 (RO/ 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 X K5.06: Knowledge of the following operational implications as they apply to the CRDS: Effects of control rod motion on axial offset(CFR:

41.5/45.7) 3.8 56002 Reactor Coolant X K6.02: Knowledge of the effect or a loss or malfunction on the following RCS components: RCP(CFR: 41.7 /

45.7)3.6 57011 Pressurizer Level Control014 Rod Position Indication015 Nuclear Instrumentation X K4.07: Knowledge of NIS design feature(s) and/or interlock(s) provide for the following:

Permissives (CFR: 41.7) 3.7 58016 Non-Nuclear Instrumentation017 In-Core Temperature Monitor X A4.02: Ability to manually operate and/or monitor in the control room:

Temperature values used to determine RCS/RCP operation during inadequate core cooling (i.e., if applicable, average of five highest values)(CFR: 41.7 / 45.5 to 45.8) 3.8 59027 Containment Iodine Removal028 Hydrogen Recombiner andPurge Control029 Containment Purge033 Spent Fuel Pool Cooling X A1.01: Ability to predict and/or monitor changes in parameters(to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including:

Spent fuel pool water level (CFR: 41.5 /

45.5)2.7 60034 Fuel Handling Equipment035 Steam Generator041 Steam Dump/Turbine BypassControl X A3.02: Ability to monitor automatic operation of the SDS, including: RCS Pressure, RCS Temperature, and reactor power (CFR: 41.7 / 45.5) 3.3 61 045 Main Turbine Generator X A2.08: 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 correct, control, or mitigate the consequences of those malfunctions or operations: Steam dumps are not cycling properly at low load, or stick open at higher load (isolate and use atmospheric reliefs when necessary) (CFR:

41.5 / 43.5 / 45.3 / 45.5) 2.8 62055 Condenser Air Removal X K3.01: Knowledge of the effect that a loss or malfunction of the CARS will have on the following: Main condenser (CFR:

41.7 / 45.6) 2.5 63056 Condensate068 Liquid Radwaste071 Waste Gas Disposal X K1.04: Knowledge of the physical connections and/or cause-effect relationships between the Waste Gas Disposal System and the following systems: Station ventilation (CFR: 41.2 to 41.9 /

45.7 to 45.8) 2.7 64072 Area Radiation Monitoring075 Circulating Water X 2.1.30: Ability to locate and operate components, including local controls. (CFR: 41.7 /

45.7)4.4 65079 Station Air086 Fire ProtectionK/A Category Point Totals:

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

10/3 ES-401Generic Knowledge and Abilities Outline (Tier 3)Form ES-401-3Facility: Waterford 3 (RO)Date of Exam: September 14, 2015CategoryK/A #Topic ROSRO-Only IR#IR#1.Conduct ofOperations 2.1.3Knowledge of shift or short-term relief turnover practices.(CFR: 41.10 / 45.13) 3.7 66 2.1.19Ability to use plant computers to evaluate system orcomponent status. (CFR: 41.10 / 45.12) 3.9 67 2.1.44Knowledge of RO duties in the control room during fuelhandling, such as responding to alarms from the fuelhandling area, communication with the fuel storagefacility, systems operated from the control room insupport of fueling operations, and supportinginstrumentation. (CFR: 41.10 / 43.7 / 45.12) 3.9 68 2.1.Subtotal 2.EquipmentControl 2.2.6Knowledge of the process for making changes toprocedures. (CFR: 41.10 / 43.3 / 45.13) 3.0 69 2.2.35Ability to determine Technical Specification Mode ofOperation. (CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.6 70 2.2.39Knowledge of less than or equal to one hour TechnicalSpecification action statements for systems.(CFR: 41.7 /41.10 / 43.2 / 45.13) 3.9 71Subtotal 3.RadiationControl 2.3.13Knowledge of radiological safety procedures pertaining tolicensed operator duties, such as response to radiationmonitor alarms, containment entry requirements, fuelhandling responsibilities, access to locked high-radiationareas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 /

45.10)3.4 72 2.3.15Knowledge of radiation monitoring systems, such as fixedradiation monitors and alarms, portable surveyinstruments, personnel monitoring equipment, etc. (CFR:41.12 / 43.4 / 45.9) 2.9 73 2.3.Subtotal 4.EmergencyProcedures /

Plan 2.4.8Knowledge of how abnormal operating procedures areused in conjunction with EOPs.(CFR: 41.10 / 43.5 /

45.13)3.8 74 2.4.25Knowledge of fire protection procedures.(CFR: 41.10 / 43.5 / 45.13) 3.3 75 2.4.2.4.SubtotalTier 3 Point Total 10 7 ES-401Record of Rejected K/AsForm ES-401-4Tier /GroupRandomlySelected K/AReason for Rejection 1/1 RO 3009 EK3.04009 EK 3.17 rejected because suitable distractors were notavailable for knowledge of the reason for a containmentisolation. All the distractors we came up with were notplausible.

1/1RO 14056 2.4.18056 2.4.45 rejected because we could not develop a questionwith plausible distractors for a loss of offsite power that wouldrequire a prioritization of annunciators (first part of the K/A) 1/1RO 16026 AA2.01062 AA2.02 The only possible loss of ACCW is a pump trip.The only ACCW pump trips is overcurrent and undervoltage.A question was developed on the ACCW pump failure to starton the EDG sequencer but discovered it overlapped aquestion on one of the previous two exams. Could notdevelop a question that did not involve the EDG sequencerthat would have plausible distractors.

1/1 RO17065 AA1.01065 AA1.04 is related to an Emergency Air Compressor for aLoss of Instrument Air. W3 does not have an Emergency AirCompressor.

1/1RO 18077 AK2.07077 AK2.02 Could not link breakers, relays with generatorvoltage and grid disturbance that could be supported withadequate technical reference.

1/2RO 19CE/A16 AK1.10003 AK1.19 rejected because we could not identify plausibledistractors when creating a question. There is no referencesin OPS procedures that discussed differential rod worthduring a dropped CEA.

1/2RO 22051 AK3.01032 AK3.01 rejected because we could not develop a loss ofsource range question that did not duplicate existing ROquestions on this exam (RO68) or the previous two exams.

1/2RO 26074 EA1.01074 EA1.13 and the K/A for question number 59 were bothasking about subcooling monitors. Rejected the K/A for RO26 because we could not develop independent questions.

2/1RO 28003 K6.14003 K6.04 rejected because we could not develop a questionwhere a fault of a CIV would affect a RCP. CBO flow has twoCIV's but isolating them has no effect due to flow beingdirected to another location, and the CCW valves to theRCPs fail in the open (no effect) position.

2/1RO 38012 K1.02012 K1.04 rejected. The rod position indicating system doesnot have a direct physical connection with the W3 RPSsystem. Could not develop a question that effectivelymatched the K/A.

2/1RO 45076 A1.02059 A1.03 rejected because a question could not bedeveloped that did not require a reference and a direct look up.2/1RO 51004 K5.30073 K5.01 rejected because we could only find questions forthis K/A that pertained to radiation theory. We could not applythis theory to the Process Radiation Monitors and haveadequate technical references.

2/1RO 53076 K1.19076 K1.05 rejected because the ACCW system does nothave a physical connection with the EDG system.

2/2RO 61041 A3.02041 A3.05 rejected. Two independent Main Steam crossoverpressures would have to fail such that the SDS is affected(permissive and demand). The question developed for thisK/A was not plausible.

ES-401PWR Examination OutlineForm ES-401-2Facility: Waterford 3 (SRO) Date of Exam: March 27, 2017TierGroupRO 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 &AbnormalPlantEvolutions 1N/AN/A 18 3 3 6 2 9 2 2 4Tier Totals 27 5 5 10 2.PlantSystems 1 28 3 2 5 2 10 0 1 2 3Tier Totals 38 4 4 83. Generic Knowledge and AbilitiesCategories 1 2 3 4 10 1 2 2 2 3 1 4 2 7Note: 1.Ensure that at least two topics from every applicable K/A category are sampled within each tier of the ROand SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" ineach K/A category shall not be less than two).(One Tier 3 Radiation Control K/A is allowed if the K/A isreplaced by a K/A from another Tier 3 Category).

2.The point total for each group and tier in the proposed outline must match that specified in the table. Thefinal point total for each group and tier may deviate by +/-1 from that specified in the table based on NRCrevisions. 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 thatdo not apply at the facility should be deleted with justification; operationally important, site-specificsystems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401for 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 thegroup 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 beselected. 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 topicsmust be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicableK/As.8.On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importanceratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enterthe group and tier totals for each category in the table above; if fuel handling equipment is sampled in acategory other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 forTier 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.

G*Generic K/As ES-401 2Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO /SRO)E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G*K/A Topic(s)

IR#000007 (BW/E02&E10; CE/E02) ReactorTrip - Stabilization - Recovery / 1000008 Pressurizer Vapor SpaceAccident / 3000009 Small Break LOCA / 3 XEA2.14: Ability to determine and interpret thefollowing as they apply to a small break LOCA:Actions to be taken if PTS limits are violated (CFR43.5)4.4 76000011 Large Break LOCA / 3000015/17 RCP Malfunctions / 4000022 Loss of Rx Coolant Makeup / 2 X2.2.22: Knowledge of limiting conditions foroperations and safety limits.(CFR: 41.5 / 43.2 /45.2)4.7 81000025 Loss of RHR System / 4 X2.4.9: Knowledge of low power/shutdownimplications in accident (LOCA or loss of heatremoval) mitigation strategies (CFR 41.10, 43.5,45.13)4.2 77000026 Loss of Component CoolingWater / 8000027 Pressurizer Pressure ControlSystem Malfunction / 3000029 ATWS / 1000038 Steam Gen. Tube Rupture / 3000040 (BW/E05; CE/E05; W/E12)Steam Line Rupture - Excessive HeatTransfer / 4 XAA2.01 Ability to determine and interpret thefollowing as they apply to the Steam Line Rupture:Occurrence and location of a steam line rupturefrom pressure and flow indications (CFR43.5,45.13)4.7 78000054 (CE/E06) Loss of MainFeedwater / 4000055 Station Blackout / 6 X2.4.6: Knowledge of EOP mitigation strategies.(CFR: 41.10 / 43.5 / 45.13)4.7 79000056 Loss of Off-site Power / 6000057 Loss of Vital AC Inst. Bus / 6000058 Loss of DC Power / 6 XAA2.03: Ability to determine and interpret thefollowing as they apply to the Loss of DC Power:DC loads lost; impact on ability to operate andmonitor plant systems(CFR: 43.5) 3.9 80000062 Loss of Nuclear Svc Water / 4000065 Loss of Instrument Air / 8W/E04 LOCA Outside Containment / 3W/E11 Loss of Emergency CoolantRecirc. / 4 BW/E04; W/E05 Inadequate HeatTransfer - Loss of Secondary Heat Sink / 4000077 Generator Voltage and ElectricGrid Disturbances / 6K/A Category Totals: 3 3Group Point Total:

18/6 ES-401 3Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO /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 / 1000003 Dropped Control Rod / 1000005 Inoperable/Stuck Control Rod / 1000024 Emergency Boration / 1 XAA2.05: Amount of boron to add to achieverequired shutdown margin (CFR 43.5) 3.9 82000028 Pressurizer Level Malfunction / 2000032 Loss of Source Range NI / 7000033 Loss of Intermediate Range NI / 7000036 (BW/A08) Fuel Handling Accident / 8 X2.3.12 Knowledge of radiological safetyprinciples pertaining to licensed-operatorduties, such as containment entryrequirements, fuel handling responsibilities,access to locked high rad areas, aligningfilters, etc. (CFR 43.7) 3.7 84000037 Steam Generator Tube Leak / 3000051 Loss of Condenser Vacuum / 4000059 Accidental Liquid Radwaste Rel. / 9000060 Accidental Gaseous Radwaste Rel. / 9000061 ARM System Alarms / 7000067 Plant Fire On-site / 8 XAA2.17: Ability to determine and interpretthe following as they apply to plant fire onsite: systems that may be affected by the fire(CFR 43.5) 4.3 85000068 (BW/A06) Control Room Evac. / 8000069 (W/E14) Loss of CTMT Integrity / 5000074 (W/E06&E07) Inad. Core Cooling / 4000076 High Reactor Coolant Activity / 9W/EO1 & E02 Rediagnosis & SI Termination / 3W/E13 Steam Generator Over-pressure / 4W/E15 Containment Flooding / 5W/E16 High Containment Radiation / 9BW/A01 Plant Runback / 1BW/A02&A03 Loss of NNI-X/Y / 7BW/A04 Turbine Trip / 4BW/A05 Emergency Diesel Actuation / 6BW/A07 Flooding / 8BW/E03 Inadequate Subcooling Margin / 4BW/E08; W/E03 LOCA Cooldown - Depress. / 4BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4BW/E13&E14 EOP Rules and EnclosuresCE/A11; W/E08 RCS Overcooling - PTS / 4CE/A16 Excess RCS Leakage / 2CE/E09 Functional Recovery X2.2.37: Ability to determine operability and/oravailability of safety related equipment.(CFR 43.5) 4.6 83 K/A Category Point Totals:

2 2Group Point Total:

9/4 ES-401 4Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Plant Systems - Tier 2/Group 1 (RO /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 Pump004 Chemical and VolumeControl XA2.15: Ability to (a) predict the impactsof the following malfunctions oroperations on the CVCS; and (b) basedon those predictions, use procedures tocorrect, control, or mitigate theconsequences of those malfunctions oroperations: High or low PZR level (CFR43.5)3.7 86005 Residual Heat Removal006 Emergency Core Cooling007 Pressurizer Relief/QuenchTank008 Component Cooling Water X2.2.40: Ability to apply Technicalspecifications for a system.(CFR: 43.2 /43.5)4.7 87010 Pressurizer Pressure Control012 Reactor Protection013 Engineered Safety FeaturesActuation XA2.06: Ability to (a) predict the impactsof the following malfunctions oroperations on the ESFAS; and (b)based on those predictions, useprocedures to correct, control, ormitigate the consequences of thosemalfunctions or operations: InadvertentESFAS Actuation (CFR 43.5)4.0 88022 Containment Cooling025 Ice Condenser026 Containment Spray039 Main and Reheat Steam059 Main Feedwater061 Auxiliary/EmergencyFeedwater XA2.04: Ability to (a) predict the impactsof the following malfunctions oroperations on the AFW; and (b) basedon those predictions, use procedures tocorrect, control, or mitigate theconsequences of those malfunctions oroperations: pump failure or improperoperation (CFR 43.5) 3.8 90062 AC Electrical Distribution063 DC Electrical Distribution064 Emergency Diesel Generator X2.2.25 Knowledge of the bases inTechnical Specifications for limitingconditions for operation and safety limits(CFR 43.2) 4.2 89073 Process Radiation Monitoring076 Service Water078 Instrument Air 103 ContainmentK/A Category Point Totals:

3 2Group Point Total:

28/5 ES-401 5Form ES-401-2ES-401PWR Examination OutlineForm ES-401-2Plant Systems - Tier 2/Group 2 (RO /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 Drive002 Reactor Coolant011 Pressurizer Level Control014 Rod Position Indication X2.4.11 Knowledge of abnormal conditionprocedures.(CFR: 41.10 / 43.5 / 45.13) 4.2 91015 Nuclear Instrumentation016 Non-Nuclear Instrumentation017 In-Core Temperature Monitor027 Containment Iodine Removal028 Hydrogen Recombiner andPurge Control029 Containment Purge XA2.03: Ability to (a) predict the impacts ofthe following malfunctions or operations onthe Containment Purge System; and (b)based on those predictions, useprocedures to correct, control, or mitigatethe consequences of those malfunctions oroperations: Startup operations and theassociated required valve lineups. (CFR43.5)3.1 92033 Spent Fuel Pool Cooling034 Fuel Handling Equipment035 Steam Generator041 Steam Dump/Turbine BypassControl045 Main Turbine Generator055 Condenser Air Removal056 Condensate068 Liquid Radwaste071 Waste Gas Disposal072 Area Radiation Monitoring075 Circulating Water079 Station Air X2.4.8 Knowledge of how abnormaloperating procedures are used inconjunction with EOPs (CFR 43.5) 4.5 93086 Fire Protection K/A Category Point Totals:

1 2Group Point Total:

10/3 ES-401Generic Knowledge and Abilities Outline (Tier 3)Form ES-401-3Facility: Waterford 3 (SRO)Date of Exam: September 14, 2015CategoryK/A #Topic ROSRO-Only IR#IR#1.Conduct ofOperations 2.1.4Knowledge of individual licensed operator responsibilitiesrelated to shift staffing, such as medical requirements, "nosolo" operation, maintenance of active license status,10CFR55, etc. (CFR 43.2) 3.8 94 2.1.45Ability to identify and interpret diverse indications tovalidate the response of another indication (CFR 43.5) 4.3 95 2.1.Subtotal 2.EquipmentControl 2.2.13Knowledge of tagging and clearance procedures 4.3 962.2.12 Knowledge of surveillance procedures (CFR 43.5) 4.1 97 2.2.Subtotal 3.RadiationControl 2.3.11Ability to control releases (CFR 43.4) 4.3 98 2.3.2.3.Subtotal 4.EmergencyProcedures /

Plan 2.4.21Knowledge of the parameters and logic used to assessthe status of safety functions, such as reactivity control,core cooling and heat removal, reactor coolant integrity,containment conditions, radioactivity release control, etc.(CFR 43.4) 4.6 99 2.4.30Knowledge of events related to system operation/statusthat must be reported to internal organizations or externalagencies, such as the State, the NRC, or thetransmission system operator (CFR 43.5) 4.1 100 2.4.SubtotalTier 3 Point Total 10 7 ES-401Record of Rejected K/AsForm ES-401-4Tier /GroupRandomlySelected K/AReason for Rejection 1/1 SRO 3(Q78)040 AA2.01038 EA2.14 rejected. Could not make an SRO question fromthe original K/A that did not conflict with the K/A for RO10.They were too much alike and one of the K/As had to berejected.1/1 SRO 5(Q80)058 AA2.03058 AA2.02 rejected. Could not develop an SRO onlyquestion associated with the determination and interpretationof a low DC voltage.

1/1 SRO 6(Q81)022 2.2.22077 2.2.22 rejected. There is a 2015 RO question (RO18)written on voltage and Electrical Grid Disturbance basisinformation. Could not devise an SRO question independentof that one.

1/2 SRO 8(Q83)CE/E09033 2.2.37 rejected. The intermediate range detectors at W3are the log channel instruments. There is a failed log channelon the audit simulator exam (scenario 1). Was not able tocreate a question that would not create overlap with the auditexam.2/1SRO 11(Q86)004 A2.15004 A2.05 rejected. RCP seal failures do not have an effecton CVCS other than CBO going to the VCT. Could notdevelop a question based solely on rising CBO temperatureor flow rate.

2/1SRO 13(Q88)013 A2.06013 A2.02 rejected. This K/A was an exact duplicate of theK/A for RO40.

2/1SRO 14(Q89)064 2.2.25062 2.2.25 rejected. The basis section for AC distribution(TRM 3.8.1.1) was used as an RO question on the 2015 ROexam (RO18). Could not develop a question that did notduplicate the RO question.

2/1SRO15(Q90)061 A2.04103 A2.03 rejected. W3 does not utilize a Phase A and PhaseB Isolation system.

2/2SRO 17(Q92)029 A2.03017 A2.02 rejected. Could not locate enough proceduralguidance on the effects that core damage would have onCore Exit Thermocouples to develop an SRO test question.

3/2SRO21(Q96)2.2.132.2.5 rejected. The steps in the W3 procedure for makingdesign changes to the facility are cumbersome and designinga specific question on these detailed steps resulted inquestions that are considered minutia.

ES-301Administrative Topics OutlineForm ES-301-12017 NRC Revision 0Facility:Waterford 3Date of Examination:Mar 27, 2017Examination Level:

RO SROOperating Test Number:

1Administrative Topic (see Note)Type Code*Describe activity to be performed A1Conduct of OperationsK/A Importance: 3.6D, R2.1.18, Ability to make accurate, clear, and concise logs,records, status boards, and reports.Complete OP-004-005, Core Operating Limits SupervisorySystem Operation, Attachment 11.6, Calculation ofCharging and Letdown Parameters.

A2Conduct of OperationsK/A Importance: 4.3N,R2.1.23 Ability to perform specific system and integratedplant procedures during all modes of plant operation.Determine times to boil and core uncovery in accordancewith OP-901-131, Shutdown Cooling Malfunction.

A3Equipment ControlK/A Importance: 3.7N,R2.2.12, Knowledge of Surveillance Procedures.Perform Keff Calculation in accordance with OP-903-090,Shutdown Margin, Section 7.5, Keff Calculation.

A4Radiation ControlK/A Importance: 3.8P,D,R2.3.11, Ability to control radiation releases.Evaluate Meteorological conditions for gaseous releasefrom the Gaseous Waste Management System inaccordance with OP-007-003, Gaseous WasteManagement.(From 2014 NRC Exam)Emergency PlanNot SelectedNOTE:All items (five total) are required for SROs. RO applicants require only four items unless theyare retaking only the administrative topics (which would require all five items).* 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-301Administrative Topics OutlineForm ES-301-12017 NRC Revision 0Facility:Waterford 3 Date of Examination:Mar 27, 2017Examination Level:

RO SRO Operating Test Number:

1Administrative Topic (see Note)Type Code*Describe activity to be performed A5Conduct of OperationsK/A Importance: 3.8D,R2.1.18, Ability to make accurate, clear, and concise logs,records, status boards, and reports.Review and approve completed OP-903-117, EmergencyDiesel Generator Fuel Oil Transfer Pump Operability Check,Attachment 10.1, Fuel Oil Transfer Pump A IST Data.

A6Conduct of OperationsK/A Importance: 4.4N,R2.1.23 Ability to perform specific system and integratedplant procedures during all modes of plant operation.Determine time to boil and identify containment closurerequirements in accordance with OP-901-131, ShutdownCooling Malfunction.

A7Equipment ControlK/A Importance: 4.1N,R2.2.12, Knowledge of Surveillance ProceduresReview Keff Calculation in accordance with OP-903-090,Shutdown Margin, Section 7.5, Keff Calculation. Applicantdetermines Keff does not meet Tech Spec 3.1.2.9requirements and identifies required corrective actions.

A8Radiation ControlK/A Importance: 3.8P,M,R2.3.14, Knowledge of radiation or contamination hazardsthat may arise during normal, abnormal, or emergencyconditions or activities.Calculate dose and assign non-licensed operators to ventSafety Injection piping in Safeguards Room A. Given doserate with and without shielding installed, time to installshielding, and job completion time using 1 team or using 2teams, determine proper job assignment.(Modified from 2014 NRC Exam)

A9Emergency PlanK/A Importance: 4.6N,R2.4.41, Knowledge of the emergency action level thresholdsand classifications.Determine appropriate Emergency Plan action level inaccordance with EP-001-001, Recognition andClassification of Emergency Conditions.NOTE:All items (five total) are required for SROs. RO applicants require only four items unless theyare retaking only the administrative topics (which would require all five items).* 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-301Control Room/In-Plant Systems OutlineForm ES-301-2 12017 NRC Revision 0Facility:Waterford 3 Date of Examination:Mar 27, 2017Exam Level RO SRO-I SRO-U Operating Test No.:

1Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-USystem / JPM TitleType Code*SafetyFunction S1001 Control Rod Drive SystemPlace Reactor Cutback (RXC) in service and perform ImmediateOperator Actions following a Cutback with unanalyzed rodconfiguration.Alt. Path: Feedwater pump will trip resulting in a RXC. During thecutback, an incorrect CEA will drop.A,D,S 1GEN 2.4.49 The ability to perform withoutreference to procedures those actions thatrequire immediate operation of systemcomponents and controls.RO - 4.6, SRO - 4.4 S2006 Emergency Core Cooling SystemReduce RCS pressure and use High Pressure Safety Injection Pumpsto restore Pressurizer level in accordance with OP-901-112, Chargingor Letdown Malfunction.D,L,S 3A1.18 PZR level and pressureRO - 4.0, SRO - 4.3 S3003 Reactor Coolant Pump SystemPerform a Reactor Coolant Pump Shutdown in accordance with OP-001-002, Reactor Coolant Pump Operation.

(2014 NRC Exam)Alt. Path: Reactor Coolant pump reverse rotates requiring stopping ofremaining Reactor Coolant Pumps.(W3 OE)A,D,L,P,S 4PA2.02 Conditions which exist for an abnormalshutdown of an RCP in comparison to anormal shutdown of an RCPRO - 3.7, SRO - 3.9 S4061 Emergency Feedwater SystemReset EFW Pump AB after Overspeed Trip in accordance with OP-009-003, Emergency Feedwater (Control Room actions)EN,L,N,S 4SGEN EPE 074 EA1.07 AFW SystemRO - 4.2, SRO - 4.3 S5022 Containment Cooling SystemPerform OP-903-037, Containment Cooling Fans OperabilityVerificationD,S 5A4.01 CCS FansRO - 3.6, SRO - 3.6 S6064 Emergency Diesel Generator (ED/G) SystemParallel Emergency Diesel Generator A for EDG testing in accordancewith OP-009-002, Emergency Diesel Generator.Alt. Path: After EDG A load is raised, EDG A load will rise withoutmanipulation requiring a trip of EDG A.A,D,S 6A4.06 Manual start, loading, and stopping ofthe ED/GRO - 3.9, SRO - 3.9 ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 22017 NRC Revision 0 S7012 Reactor Protection SystemReset High Containment Pressure ESFAS trip in accordance with OP-902-009, EOP Standard Appendices, Att. 5-D.EN,L,N,S 7A4.04 Bistable, trips, reset and test switchesRO - 3.3, SRO - 3.3 S8008 Component Cooling Water SystemSplit CCW headers in accordance with OP-901-510, CCW SystemMalfunction, Section E2, step 8.N,S 8A4.01 CCW indications and controlsRO - 3.3, SRO - 3.1In-Plant Systems * (3 for RO; (3 for SRO-I); (3 or 2 for SRO-U)

P1076 Service Water System (ACCW)Transfer EFW Pump Suctions to Wet Cooling Tower after CondensateStorage Pool Depletion using EOP OP-902-009, Standard Appendices,Attachment 10(Top 10 PSA Action)D,E,L,R 4SK1.20 AFWRO - 3.4, SRO - 3.4 P2064 Electrical Diesel GeneratorsReset EDG A following an overspeed trip with a LOOP in accordancewith OP-009-002, Emergency Diesel Generator, Section 8.8.D,E,L,R 6EPE 055 EA1.06 Restoration of power withone ED/GRO - 4.1, SRO - 4.5 P3006 Emergency Core Cooling SystemIsolate RWSP from Purification in accordance with OP-902-009, EOPStandard Appendices, Att. 40.Alt. Path: FS-423, RWSP Suction Isolation is unable to be closedA,E,L,N,R 2EPE 011 EK3.12 Actions contained in EOPfor emergency LOCA (large break)RO - 4.4, SRO - 4.6

• All RO and SRO-I control room (and in-plant) systems must be different and serve differentsafety functions; all five SRO-U systems must serve different safety functions; in-plant systemsand functions may overlap those tested in the control room.* Type CodesCriteria for RO / SRO-I / SRO-U(A)lternate path4-6 / 4-6 / 2-3 4(C)ontrol room 0(D)irect from bank 9 / 8 / 4 7(E)mergency or abnormal in-plant 1 / 1 / 1 3(EN)gineered safety feature1 / 1 / 1 (control room system) 2(L)ow-Power / Shutdown 1 / 1 / 1 7(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 3(S)imulator 8

ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 32017 NRC Revision 0Facility:Waterford 3 Date of Examination:Mar 27, 2017Exam Level RO SRO-I SRO-U Operating Test No.:

1Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-USystem / JPM TitleType Code*SafetyFunction S1001 Control Rod Drive SystemPlace Reactor Cutback (RXC) in service and perform ImmediateOperator Actions following a Cutback with unanalyzed rodconfiguration.Alt. Path: Feedwater pump will trip resulting in a RXC. During thecutback, an incorrect CEA will drop.A,D,S 1GEN 2.4.49 The ability to perform withoutreference to procedures those actions thatrequire immediate operation of systemcomponents and controls.RO - 4.6, SRO - 4.4 S2006 Emergency Core Cooling SystemReduce RCS pressure and use High Pressure Safety Injection Pumpsto restore Pressurizer level in accordance with OP-901-112, Chargingor Letdown Malfunction.D,L,S 3A1.18 PZR level and pressureRO - 4.0, SRO - 4.3 S3003 Reactor Coolant Pump SystemPerform a Reactor Coolant Pump Shutdown in accordance with OP-001-002, Reactor Coolant Pump Operation.

(2014 NRC Exam)Alt. Path: Reactor Coolant pump reverse rotates requiring stopping ofremaining Reactor Coolant Pumps.A,D,L,P,S 4PA2.02 Conditions which exist for an abnormalshutdown of an RCP in comparison to anormal shutdown of an RCPRO - 3.7, SRO - 3.9 S4061 Emergency Feedwater SystemReset EFW Pump AB after Overspeed Trip in accordance with OP-009-003, Emergency Feedwater (Control Room actions)EN,L,N,S 4SGEN EPE 074 EA1.07 AFW SystemRO - 4.2, SRO - 4.3 S5 S6064 Emergency Diesel Generator (ED/G) SystemParallel Emergency Diesel Generator A for EDG testing in accordancewith OP-009-002, Emergency Diesel Generator.Alt. Path: After EDG A load is raised, EDG A load will rise withoutmanipulation requiring a trip of EDG A.A,D,S 6A4.06 Manual start, loading, and stopping ofthe ED/GRO - 3.9, SRO - 3.9 S7012 Reactor Protection SystemReset High Containment Pressure ESFAS trip in accordance with OP-902-009, EOP Standard Appendices, Att. 5-D.EN,L,N,S 7A4.04 Bistable, trips, reset and test switchesRO - 3.3, SRO - 3.3 ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 42017 NRC Revision 0 S8008 Component Cooling Water SystemSplit CCW headers in accordance with OP-901-510, CCW SystemMalfunction, Section E2, step 8.N,S 8A4.01 CCW indications and controlsRO - 3.3, SRO - 3.1In-Plant Systems * (3 for RO; (3 for SRO-I); (3 or 2 for SRO-U)

P1076 Service Water System (ACCW)Transfer EFW Pump Suctions to Wet Cooling Tower after CondensateStorage Pool Depletion using EOP OP-902-009, Standard Appendices,Attachment 10(Top 10 PSA Action)D,E,L,R 4SK1.20 AFWRO - 3.4, SRO - 3.4 P2064 Electrical Diesel GeneratorsReset EDG A following an overspeed trip with a LOOP in accordancewith OP-009-002, Emergency Diesel Generator, Section 8.8.D,E,L,R 6EPE 055 EA1.06 Restoration of power withone ED/GRO - 4.1, SRO - 4.5 P3006 Emergency Core Cooling SystemIsolate RWSP from Purification in accordance with OP-902-009, EOPStandard Appendices, Att. 40.Alt. Path: FS-423, RWSP Suction Isolation is unable to be closedA,E,L,N,R 2EPE 011 EK3.12 Actions contained in EOPfor emergency LOCA (large break)RO - 4.4, SRO - 4.6

• All RO and SRO-I control room (and in-plant) systems must be different and serve differentsafety functions; all 5 SRO-U systems must serve different safety functions; in-plant systemsand functions may overlap those tested in the control room.* Type CodesCriteria for RO /SRO-I / SRO-U(A)lternate path4-6 / 4-6 / 2-3 4(C)ontrol room 0(D)irect from bank 9 / 8 / 4 6(E)mergency or abnormal in-plant 1 / 1 / 1 3(EN)gineered safety feature1 / 1 / 1 (control room system) 2(L)ow-Power / Shutdown 1 / 1 / 1 7(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 3(S)imulator 7

ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 52017 NRC Revision 0Facility:Waterford 3 Date of Examination:Mar 27, 2017Exam Level RO SRO-I SRO-U Operating Test No.:

1Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-USystem / JPM TitleType Code*SafetyFunction S1001 Control Rod Drive SystemPlace Reactor Cutback (RXC) in service and perform ImmediateOperator Actions following a Cutback with unanalyzed rodconfiguration.Alt. Path: Feedwater pump will trip resulting in a RXC. During thecutback, an incorrect CEA will drop.A,D,S 1GEN 2.4.49 The ability to perform withoutreference to procedures those actions thatrequire immediate operation of systemcomponents and controls.RO - 4.6, SRO - 4.4 S2 S3 S4061 Emergency Feedwater SystemReset EFW Pump AB after Overspeed Trip in accordance with OP-009-003, Emergency Feedwater (Control Room actions)EN,L,N,S 4SGEN EPE 074 EA1.07 AFW SystemRO - 4.2, SRO - 4.3 S5 S6 S7012 Reactor Protection SystemReset High Containment Pressure ESFAS trip in accordance with OP-902-009, EOP Standard Appendices, Att. 5-D.EN,L,N,S 7A4.04 Bistable, trips, reset and test switchesRO - 3.3, SRO - 3.3 S8 ES-301Control Room/In-Plant Systems OutlineForm ES-301-2 62017 NRC Revision 0In-Plant Systems * (3 for RO; (3 for SRO-I); (3 or 2 for SRO-U)

P1 P2064 Electrical Diesel GeneratorsReset EDG A following an overspeed trip with a LOOP in accordancewith OP-009-002, Emergency Diesel Generator, Section 8.8.D,E,L,R 6EPE 055 EA1.06 Restoration of power withone ED/GRO - 4.1, SRO - 4.5 P3006 Emergency Core Cooling SystemIsolate RWSP from Purification in accordance with OP-902-009, EOPStandard Appendices, Att. 40.Alt. Path: FS-423, RWSP Suction Isolation is unable to be closedA,E,L,N,R 2EPE 011 EK3.12 Actions contained in EOPfor emergency LOCA (large break)RO - 4.4, SRO - 4.6

• All RO and SRO-I control room (and in-plant) systems must be different and serve differentsafety functions; all 5 SRO-U systems must serve different safety functions; in-plant systemsand functions may overlap those tested in the control room.* Type CodesCriteria for RO / SRO-I /

SRO-U(A)lternate path4-6 / 4-6 / 2-3 2(C)ontrol room 0(D)irect from bank 9 / 8 / 4 2(E)mergency or abnormal in-plant 1 / 1 / 1 2(EN)gineered safety feature1 / 1 / 1 (control room system) 2(L)ow-Power / Shutdown 1 / 1 / 1 4(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 2(S)imulator 3

ES-301Transient and Event ChecklistForm ES-301 1 -2017 NRC Revision 1Facility:Waterford 3Date of Exam:March 27, 2017Operating Test No.

1 A P P L I C A N T E V E N T T Y P EScenarios1 (SPARE)2 3 4 T O T A L M I N I M U M(*)CREWPOSITIONCREWPOSITIONCREWPOSITIONCREWPOSITION 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 S R O A T C B O PR I U R1 RX 1 11 1 0NOR 1,5 4 31 1 1 I/C 2,3,4, 8 2,5 3,7,9 94 4 2 MAJ 7 6 6,8 42 2 1 TS 00 2 2 R2 & R5 RX 4 11 1 0NOR 1 11 1 1 I/C 3,4,8 2,5 54 4 2 MAJ 6 6,8 32 2 1 TS 00 2 2 R3 RX 5 11 1 0NOR 4 11 1 1 I/C 2,6 3,7,9 54 4 2 MAJ 7 6,8 32 2 1 TS 00 2 2 R4 & R7 RX 1 11 1 0NOR 1,5 21 1 1 I/C 2,3,4, 8 2,5 64 4 2 MAJ 7 6 22 2 1 TS 00 2 2Instructions:1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each eventtype; 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 (SRO-I) must serve in both the SRO and the ATC positions, includingat least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-Iadditionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctionsrequired for the ATC position.2. Reactivity manipulations may be conducted under normal orcontrolled abnormal conditions (refer toSection D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions maybe 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 requireverifiable actions that provide insight to the applicant's competence count toward the minimum requirementsspecified for the applicant's license level in the right-hand columns.4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may placeSRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

ES-301Transient and Event ChecklistForm ES-301 2 -2017 NRC Revision 1Facility:Waterford 3Date of Exam:March 27, 2017Operating Test No.

1 A P P L I C A N T E V E N T T Y P EScenarios1 (SPARE)2 3 4 T O T A L M I N I M U M(*)CREWPOSITIONCREWPOSITIONCREWPOSITIONCREWPOSITION 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 S R O A T C B O PR I U R6 RX 5 11 1 0NOR 1 11 1 1 I/C 2,6 3,4,8 54 4 2 MAJ 7 6 22 2 1 TS 00 2 2 I1 RX 5 11 1 0NOR 1 4 21 1 1 I/C 2,6 2,3,4, 5,8 2,3,5, 7,9 124 4 2 MAJ 7 6 6,8 42 2 1 TS 3,4 1,4 40 2 2 U1 RX 01 1 0NOR 1,5 4 31 1 1 I/C 2,3,4, 6,8 2,3,5, 7,9 104 4 2 MAJ 7 6,8 32 2 1 TS 1,4 1,4 40 2 2 U2 & U3 RX 01 1 0NOR 1,5 1 31 1 1 I/C 2,3,4, 6,8 2,3,4, 5,8 104 4 2 MAJ 7 6 22 2 1 TS 1,4 3,4 40 2 2Instructions:1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each eventtype; 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 (SRO-I) must serve in both the SRO and the ATC positions, includingat least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-Iadditionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctionsrequired for the ATC position.2. Reactivity manipulations may be conducted under normal orcontrolled abnormal conditions (refer toSection D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions maybe 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 requireverifiable actions that provide insight to the applicant's competence count toward the minimum requirementsspecified for the applicant's license level in the right-hand columns.4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may placeSRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

Appendix DScenario OutlineForm ES-D 1 -2017 NRC Exam Scenario 1 D-1 Rev 2Facility:Waterford 3Scenario No.:

1Op Test No.:

1Examiners:Operators:Initial Conditions: Reactor power is 100%. AB Buses are aligned to Train B.Turnover: Protected Train is B; Maintain 100%.High Pressure Safety Injection Pump A is out of service.EventNo.Malf.No.EventType*EventDescription 1CV12A1I - ATCI - SROVCT level instrument CVC-ILT-0227 Fails highdiverting letdown to the Boron Management system.OP-901-113, Volume Control Tank Makeup ControlMalfunction 2 RC19CI - BOPI - SROTS - SROSafety Channel C RCS Cold Leg instrument RC-ITI-0102CC (Loop T112C) fails high requiring TS 3.3.1entry and bypassing affected bistables.

3 CV01BC - ATC C - SROTS - SROCharging Pump B trips on overcurrent requiringimplementation of OP-901-112, Charging or Letdownmalfunction. (TS 3.1.2.4) 4SG05BI - BOPI - SROSteam Generator 2 Level Control Transmitter,SG-ILT-1106, fails low requiring implementation ofOP-901-201, Steam Generator Level ControlMalfunction and manual control of SG level.

5FW21AFW21AA R- ATC N-BOPN-SROLowering Main Condenser vacuum requiringimplementation of OP-901-220, Loss of CondenserVacuum and a plant power reduction in accordancewith OP-901-212, Rapid Plant Power Reduction.

6RP02A-D RC03CM - AllRCP 2A sustains a locked rotor and an automaticreactor trip does not occur. Manual action is neededto trip the reactor (CT 1, manually trip the reactor

)7MS13AM - AllMain Steam Line Break outside Containment, SG 1,OP-902-004, Excess Steam Demand Recovery. (CT2, stabilize RCS temperature and pressure

)8 RP08CI - ATCI - BOPI - SRORelay K202A fails, CVC-401, CVC-109, IA-909, andFP-601A fail to close automatically

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event DescriptionNRC Scenario 1- 2 -2017 NRC Exam Scenario 1 D-1 Rev 2The crew assumes the shift at 100% power with instructions to maintain 100% power.High PressureSafety Injection (HPSI) pump A is out of service.After taking the shift, Volume Control Tank (VCT) level instrument CVC-ILT-0227 fails high resulting invalve CVC-169 diverting letdown to the Boron Management System. The SRO should enter intoprocedure OP-901-113, Volume Control Tank Makeup Control Malfunction, and direct the ATC to placevalve CVC-169 to the VCT position.After the crew addresses the VCT instrument malfunction,RCS Cold Leg instrument RC-ITI-0102CC onCP-7 fails high. The crew will enter TS 3.3.1 action 2 and bypass bistables 3 & 4 on channel C on CP-10.After Technical Specifications are addressed and Channel C bistables bypassed, Charging Pump B tripson overcurrent. The SRO will implement OP-901-112, Charging or Letdown Malfunction, Section E 1 ,Charging Malfunction. The SRO should direct the ATC to start a standby charging pump after verifying asuction path available or isolate Letdown using CVC-101, Letdown Stop Valve. If Letdown is isolated,Charging and Letdown will be re-initiated using Attachment 2 of OP-901-112. The SRO should reviewand enter Technical Specification 3.1.2.4. Technical Specification 3.1.2.4 may be exited after aligningCharging Pump AB to replace Charging Pump B.The SRO may implement EN-OP-200, TransientResponse Rules.After the crew addresses the Charging pump malfunction, Steam Generator 2 Level Control Transmitter,SG-ILT-1106 fails low. The SRO should direct the BOP to take manual control of SG2 level 50-70%Narrow Range and establish contingency actions. The SRO will enter OP-901-201, Steam GeneratorLevel Control Malfunction and implement Attachment 1, General Actions. Manual action by the BOP tocontrol SG2 level will be required during the subsequent plant shutdown and reactor trip.After the crew completes actions in OP-901-201, a leak in the Main Condenser develops and MainCondenser vacuum begins to drop. The SRO will enter OP-901-220, Loss of Condenser Vacuum. MainCondenser vacuum will drop below 25 inches, requiring a rapid plant power reduction. The SRO willenter OP-901-212, Rapid Plant Power Reduction and should implement EN-OP-200, Transient ResponseRules. Vacuum will drop below 25 inches but remain above 20 inches, the procedure trigger for trippingthe Reactor. For the power reduction, the ATC will perform direct boration to the RCS as well as ASIcontrol with CEAs and Pressurizer boron equalization. The BOP will manipulate the controls to reduceMain Turbine load.After the reactivity manipulation is satisfied, Reactor Coolant Pump 2A rotor seizes and the RCP breakertrips. The Reactor Protection System fails to open the required Reactor Trip Breakers and an ATWScondition exists. The ATC should recognize that an automatic protection system has failed to occur andmanually trip the reactor by depressing both Reactor Trip pushbuttons (A and D) on CP-2(CRITICALTASK 1). The Reactor will be successfully tripped from CP-2 and the SRO will enter OP-902-000,Standard Post Trip Actions.During the performance of Standard Post Trip Actions (RCS Heat Removal checks), an excess steamdemand event will occur on SG 1 outside containment upstream of the MSIV. The SRO will direct theATC/BOP to initiate Safety Injection, Containment Isolation and Main Steam Isolation. The SRO will directaction to establish RCS Temperature and Pressure control using SG 2(CRITICAL TASK 2)when CETtemperature and PZR pressure begins to rise indicating a blown dry SG. Relay K202A will fail to actuateresulting in CVC-109, Letdown Outside Containment Isolation, CVC-401, Controlled Bleedoff OutsideContainment Isolation, IA-909, Instrument Air Containment Isolation and FP-601A, Fire Water AContainment Isolation, valves not going to their required positions. The ATC and BOP will take action toclose these valves. The crew should diagnose to OP-902-004, Excess Steam Demand Recovery andisolate Steam Generator 1.The scenario can be terminated after the crew has isolated Steam Generator 1 or at the lead examiner'sdiscretion.

NRC Scenario 1- 3 -2017 NRC Exam Scenario 1 D-1 Rev 2Critical TaskNumberDescriptionBasis 1Establish Reactivity ControlThis task is satisfied by manuallytripping the reactor using the manualpushbuttons, Diverse Reactor Trip, orde-energizing bus 32A and 32B within 1minute of exceeding a PPS limit. Thistask becomes applicable following theRCP trip.(OP-902-000, 1.a.1)Failure to trip the reactor when an automaticPPS signal has failed to actuate can lead todegradation of fission product barriers. OPSManagement Expectation of 1 minute isdetermined to be a reasonable time limit toidentify and take action for satisfactoryperformance.(TM-OP-100-03, CT-1) 2Establish RCS Pressure andTemperature ControlThis task is satisfied by manually feedingand steaming the unaffected SteamGenerator to stabilize RCS temperatureand pressure prior to exiting the step tostabilize RCS temperature in OP-902-004, Excess Steam Demand Recoveryand take action to achieve and maintainless than 1600 PSID across the affectedSteam Generator. This task becomesapplicable once CET temperature andPZR pressure begins to rise followingthe ESDE. (OP-902-004, step 18 or OP-902-009, App. 13)An ESDE will result in a rapid cooldown anddepressurization of the RCS. After the SteamGenerator dries out, RCS temperature andpressure will begin to recover. Operator actionis required to stabilize RCS pressure andtemperature to prevent a situation that maycause pressurized thermal shock which couldjeopardize the RCS integrity. A large D/Pacross the Steam Generator tubes will make asubsequent SGTR more likely.(TM-OP-100-03, CT-7)

• Critical Task (As defined in NUREG 1021 Appendix D)** Per NUREG-1021, Appendix D, If an operator or the crew significantly deviates from or fails to followprocedures that affect the maintenance of basic safety functions, those actions may form the basis of aCT identified in the post-scenario review.Scenario Quantitative Attributes1. Malfunctions after EOP entry (1-2)12. Abnormal events (2-4)43. Major transients (1-2)24. EOPs entered/requiring substantive actions (1-2)15. EOP contingencies requiring substantive actions (0-2)06. EOP based Critical tasks (2-3)2 NRC Scenario 1- 4 -2017 NRC Exam Scenario 1 D-1 Rev 2SCENARIO SETUPA. Reset Simulator to IC-161.B. Verify Scenario Malfunctions, Remotes, and Overrides are loaded, as listed in the Scenario Timeline.C. Verify HPSI pump A is removed from service as follows:1. InsertSIR29 (HPSI pump A breaker) to RKOUT2. Place C/S in OFF with a Danger Tag.D. Verify all EFW Flow Control Valves are in Auto and Caution Tags removed.E. Ensure Protected Train B sign is placed in SM office window.F. Verify EOOS is 8.7 Yellow with HPSI pump A out of service.G. Protected Equipment covers on running SFP pump and HPSI Pump B control switches.H. Complete the simulator setup checklist.I. Start Insight, open file Crew Performance.tis.

NRC Scenario 1- 5 -2017 NRC Exam Scenario 1 D-1 Rev 2SIMULATOR BOOTH INSTRUCTIONSEvent 1 VCT level instrument, CVC-ILT-0227, Fails High1. On Lead Examiner's cue, initiate EventTrigger 1.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.Event 2 Safety Channel C RCS Cold Leg Temperature, RC-ITI-0112CC fails high1. On Lead Examiner's cue, initiate EventTrigger 2.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.3. If sent to LCP-43, wait 3 minutes and report all Cold Leg temperatures on LCP-43 readapproximately 545F.Event 3 Charging Pump B Trip1. On Lead Examiner's cue, initiate EventTrigger 3.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Charging Pump room and breaker.3. If called as NAO to investigate the breaker, wait 3 minutes and report overcurrent flags aredropped for all 3 phases for Charging Pump B4. If called as NAO to investigate the pump, wait 3 minutes and report that there are someindications of charring at the motor vent area, and an acrid odor is present but there is no fire.5. If directed to perform prestart checks for the A or AB Charging pump, wait 2 minutes and reportthe following for directed pump:a. Suction and discharge valves are openb. Proper oil level existsc. Motor vents unobstructedd. All personnel clear of the pump6. If directed to check a started Charging pump for proper operation following start, wait 1 minuteand report the following:a. Suction and discharge valves are openb. Proper oil pressure and seal water flow existc. No abnormal vibrations or noises presentEvent 4 Steam Generator 2 Level Control Transmitter, SG-ILT-1106, fails low1. On Lead Examiner's cue, initiate EventTrigger 4.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.

NRC Scenario 1- 6 -2017 NRC Exam Scenario 1 D-1 Rev 2Event 5 Main Condenser Leak, Rapid Power Reduction1. On Lead Examiner's cue, initiate EventTrigger 5.2. If called as TGB watch report all Air Evacuation Pumps look normal, Vacuum pump separatorsare greater than 1/2 full and there are no indications of a leak.3. Approximately 5 minutes after being called to investigate, TGB watch should report finding a non-isolable leak up-stream of AE-401 A, Condenser Vacuum Breaker A. Location of failure ispreventing any successful repair efforts.4. If called as other watch standers to assist, respond that you are going to the TGB to assist.5. If Work Week Manager is called, inform the caller that a team will be sent to the Turbine Buildingto assist.Event 6 RCP 2A locked rotorand an automatic Reactor trip does not occur1. On Lead Examiner's cue, initiate EventTrigger 6.2. No expected communications for this event.Event 7 Main Steam Line Break outside Containment, SG 11. On Lead Examiner's cue, initiate EventTrigger 7.2. If the Duty Plant Manager is called, inform the caller that you will make the necessary calls.3. If Chemistry is called to perform samples acknowledge the request.4. If requested to check Emergency Diesel Generators (EDG), wait 3 minutes and report EDGs areoperating properly. Initiate event triggers 20 & 21 to acknowledge local annunciator panels.5. If called as an NAO to check for steam outside, wait 2 minutes, report that a large amount ofsteam is issuing from the west MSIV area.Event 8 Relay K202A fails, CVC-401, CVC-109, IA-909, and FP-601A fail to close automatically1. No communications should occur for this event.At the end of the scenario, before resetting, end data collection and save the file as 2017Scenario 1-(start-end time).tid. Export to .csv file. Save the file into the folder for theappropriate crew.

.

NRC Scenario 1- 7 -2017 NRC Exam Scenario 1 D-1 Rev 2SCENARIO TIMELINE EVENT KEYDESCRIPTIONTRIGGER DELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTION 1CV12A1VCT LEVEL XMTR CVC-ILIC-0227 FAILS HI 1 00:00:00 00:00:00ACTIVEVCT LEVEL TRANSMITTER FAILS HIGH 2RC19CRCS COLD LEG 1A SAFETY TT 0112C FAILS (0-100%)

2 00:00:00 00:00:00 100SAFETY CHANNEL C RCS COLD LEG TEMPERATURE (RC-ITI-0102CC) 3 CV01BCHARGING PUMP B TRIPPED 3 00:00:00 00:00:00ACTIVECHARGING PUMP B TRIP 4 SG05BSG LEVEL ILT-1106 FAIL (0-100%)

4 00:00:00 00:00:10 0STEAM GENERATOR 2 LEVEL CONTROL TRANSMITTER, SG-ILT-1106, FAILS LOW 5FW21AFW21AACONDENSER A AIR INLEAK (100%=100% OF VAC BKR)CONDENSER A AIR INLEAK VACUUM SETPOINT 5 5 00:00:00 00:00:00 00:03:00 00:03:00 20 23.3MAIN CONDENSER LEAK, RAPID POWER REDUCTION 6RC03C RP02A RP02B RP02C RP02DRCP RC-MPMP-0002A SHAFT SEIZURERPS CH A AUTO TRIP FAILURERPS CH B AUTO TRIP FAILURERPS CH C AUTO TRIP FAILURERPS CH D AUTO TRIP FAILURE 6N/AN/AN/AN/A 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00 00:00:00ACTIVEACTIVEACTIVEACTIVEACTIVERCP 2A LOCKED ROTOR AND AN AUTOMATIC REACTOR TRIP DOES NOT OCCUR 7 MS13AMS A BREAK OUTSIDE CNTMT BEFORE MSIV (0-100%)

7 00:00:00 00:00:00 8%MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT, SG 1 8 RP08CRELAY K202 FAILED, CIAS TRAIN A (CVC/IA/FP)N/A 00:00:00 00:00:00ACTIVECVC-401, CVC-109, IA-909, AND FP-601A FAIL TO CLOSE AUTOMATICALLYN/A EGR26EDG A LOCAL ANNUN ACK 20 00:00:00 00:00:00 ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGE NRC Scenario 1- 8 -2017 NRC Exam Scenario 1 D-1 Rev 2 EVENT KEYDESCRIPTIONTRIGGER DELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTIONN/A EGR27EDG B LOCAL ANNUN ACK 21 00:00:00 00:00:00 ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGEN/A SIR29HPSI PUMP AN/A 00:00:00 00:00:00RKOUTHPSI PUMP A BREAKER NRC Scenario 1- 9 -2017 NRC Exam Scenario 1 D-1 Rev 2REFERENCESEventProcedures 1OP-901-113, Volume Control Tank Makeup Control Malfunction, Rev. 302 2OP-009-007, Plant Protection System, Rev. 17OP-903-013, Monthly Channel Checks, Rev. 18Technical Specification 3.3.1 3OP-901-112, Charging or Letdown Malfunction, Rev. 6OP-002-005, Chemical Volume Control, Rev. 56Technical Specification 3.1.2.4 4OP-901-201, Steam Generator Level Control Malfunction, Rev. 6 5OP-901-220, Loss of Condenser Vacuum, Rev. 302OP-002-005, Chemical Volume Control, Rev. 56OP-004-004, Control Element Drive, Rev. 23OP-901-212, Rapid Plant Power Reduction, Rev. 8 6OP-902-000, Standard Post Trip Actions, Rev. 16 7OP-902-004, Excess Steam Demand Recovery, Rev. 16OP-902-009, Standard Appendices, Rev. 315, Appendix 2, FiguresOP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart 8OP-902-004, Excess Steam Demand Recovery, Rev. 16GEN EN-OP-115, Conduct of Operations, Rev. 17EN-OP-115-08, Annunciator Response, Rev. 4EN-OP-200, Plant Transient Response Rules, Rev. 3OI-038-000, EOP Operations Expectations / Guidance, Rev. 14 Appendix DScenario OutlineForm ES-D 1 -2017 NRC Exam Scenario 2 D-1 Rev 1Facility:Waterford 3Scenario No.:

2Op Test No.:

1Examiners:Operators:Initial Conditions: Reactor power is 100%. AB Buses are aligned to Train B.Turnover: Protected Train is B; EFW Pump A operability check is in progress; Maintain 100%. LPSI pump A is out of service.EventNo.Malf.No.EventType*EventDescription 1FW06AN - BOP N - SROTS - SROManually start EFW Pump A. EFW Pump A failsduring operability check. (TS 3.7.1.2) 2 RC21AI - AllHot Leg 1 Temperature, RC-ITI-0111X, fails lowaffecting PZR level setpoint. OP-901-110, PressurizerLevel Control Malfunction (Sect. E2).

3 RC08CC - BOP C - SROReactor Coolant Pump 2A Lower Seal fails.OP-901-130, Reactor Coolant Pump Malfunction.

4SG04EI - BOPI - SROTS - SROSteam Generator 1 Pressure Instrument,SG-IPT-1013A, fails low requiring TechnicalSpecification entry and bypass of multiple PlantProtection System A trip bistables. (TS 3.3.1, 3.3.2, &3.3.3.5)5FW35BR - ATCN - BOP N - SROFeedwater Heater 5B tube leak from Condensate toheater shell causing isolation of the Low Pressureheater string. OP-901-221, Secondary SystemTransient (Sect. E1) and OP-901-212, Rapid PlantPower Reduction to 72% power.

6 RC09CC - ATC C - SROReactor Coolant Pump 2A Middle Seal fails, requiringa manual reactor trip, and securing of Reactor CoolantPump 2A.7 RC11A1M - AllPressurizer Code Safety, RC-317A, fails open. OP-902-002, Loss of Coolant Accident Recovery. AllReactor Coolant Pumps must be secured.(CT 1, TripRCPs exceeding operating limits) 8SI02BC - BOP C - SROHPSI Pump B fails to AUTO start on the SafetyInjection Actuation Signal requiring a manual start.(CT 2, Inventory Control) 9SI19ASI01AN/AHigh Press Safety Injection (HPSI) Pump A degradesinternally and trips.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event DescriptionNRC Scenario 2- 2 -2017 NRC Exam Scenario 2 D-1 Rev 1The crew assumes the shift at 100% power with instructions to maintain 100% power. Low PressureSafety Injection pump A is out of service. The crew turnover includes instructions to complete OP-903-046, Emergency Feed Pump Operability, for Emergency Feedwater (EFW) Pump A. EFW pump A will tripon overcurrent shortly after it is started. The SRO should declare EFW pump A inoperable and enter TechSpec 3.7.1.2.d.After Tech Specs are addressed, Loop 1 T hot instrument, RC-ITI-0111X, fails low. This affects theReactor Regulating System Tave calculation and the Pressurizer Level Setpoint. The SRO should enterOP-901-110, Pressurizer Level Control Malfunction and implement Section E2, Pressurizer LevelSetpoint Malfunction. The crew should take manual control of Pressurizer Level, select the non-faulted T hot instrument (Loop 2) in both Reactor Regulating System cabinets, verify normal setpoint is restoredand restore Pressurizer Level Control to Auto after returning Pressurizer Level to setpoint.After Pressurizer Level control is in automatic, Reactor Coolant Pump 2A Lower Seal fails. The crewshould enter OP-901-130, Reactor Coolant Pump Malfunction and implement Section E1, Seal Failure.After the crew is in Section E1 of OP-901-130 AND the BOP has adjusted Component Cooling WaterTemperature, Steam Generator 1 Pressure Instrument, SG-IPT-1013A, fails low. The SRO should reviewand enter Technical Specifications 3.3.1 action 2, 3.3.2 actions 13 and 19 and 3.3.3.5 action a. The SROwill direct the BOP to bypass Steam Generator 1 Pressure Lo, Steam Generator 1 DP, and SteamGenerator 2 DP trip bistables (11, 19 & 20) in Plant Protection System Channel A within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, inaccordance with OP-009-007, Plant Protection System. The SRO should review Technical Specifications3.3.3.5 and 3.3.3.6 using OP-903-013, Monthly Channel Checks, and determine that TechnicalSpecification 3.3.3.5 is applicable and 3.3.3.6 is not.Once the SRO has addressed Technical Specifications and trip bistables are bypassed, a tube leakoccurs in Feedwater Heater 5B, causing Condensate flow to isolate through Low Pressure FeedwaterHeaters 5B and 6B. The crew should enter OP-901-221, Secondary System Transient, and implementSection E1, Loss of Feedwater Preheating. This also requires a power reduction to < 72% power usingOP-901-212, Rapid Plant Power Reduction. The SRO should implement EN-OP-200, TransientResponse Rules.After the reactivity manipulation is satisfied, Reactor Coolant Pump 2A Middle Seal fails. The crew shouldtrip the reactor, implement OP-902-000, Standard Post Trip Actions AND secure Reactor Coolant Pump 2A.After Reactor Coolant Pump 2A is secured, Pressurizer Code Safety, RC-317A, fails open. The crewshould diagnose to OP-902-002, Loss of Coolant Accident Recovery. The crew should secure anadditional Reactor Coolant Pump in the opposite loop (preferably 1A) when RCS Pressure lowers to<1621 PSIA and secure all Reactor Coolant Pumps exceeding NPSH limits as indicated by high vibrationor within 3 minutes of the Containment Spray Actuation(CRITICAL TASK 1)

.When Safety Injection occurs, either manually or automatically, HPSI Pump B fails to Auto Start. HighPressure Safety Injection (HPSI) pump A will run for about three minutes, degrade internally and trip. TheBOP should manually start High Pressure Safety Injection Pump B(CRITICAL TASK 2)

.The scenario can be terminated after the crew starts a cooldown or at the lead examiner's discretion.

NRC Scenario 2- 3 -2017 NRC Exam Scenario 2 D-1 Rev 1Critical TaskNumberDescriptionBasis 1Trip Any RCP Exceeding OperatingLimitsThis task is satisfied by stopping allrunning RCPs within 3 minutes of lossof Component Cooling Water flow orprior to completing the step that verifiesRCP operating limits. This taskbecomes applicable after either runningRCP Vibration alarms actuate ORContainment Spray is initiated,whichever occurs first. (OP-902-002,9.b or 9.d.1)The time requirement of 3 minutes is based onthe RCP operating limit of 3 minutes withoutCCW cooling. Continued operation of RCPwithout CCW or outside of the operating limitscould lead to a failure of the RCS pressureboundary at the RCP seal.(TM-OP-100-03, CT-23; ECS98-001, D.10) 2Establish RCS Inventory ControlThis task is satisfied by starting HighPressure Safety Injection Pump B toestablish Reactor Coolant Systeminventory control before exiting the stepto verify Safety Injection Actuation SignalActuation. This task becomes applicablefollowing the initiation of a SafetyInjection Actuation Signal. (OP-902-002,step 7)Based on minimum required flow per the flowdelivery curve in OP-902-009, Appendix 2E.Failure to take action to establish the minimumrequired Safety Injection flow during a LOCAwould degrade the inventory available tomaintain the fuel covered.Adequate SI flowensures RCS Inventory Control and Core HeatRemoval safety functions are satisfied.(TM-OP-100-03, CT-16; ECS98-001, A.02)

• Critical Task (As defined in NUREG 1021 Appendix D)** Per NUREG-1021, Appendix D, If an operator or the crew significantly deviates from or fails to followprocedures that affect the maintenance of basic safety functions, those actions may form the basis of aCT identifiedin the post

-scenario review.Scenario Quantitative Attributes1. Malfunctions after EOP entry (1-2)22. Abnormal events (2-4)33. Major transients (1-2)14. EOPs entered/requiring substantive actions (1-2)15. EOP contingencies requiring substantive actions (0-2)06. EOP based Critical tasks (2-3)2 NRC Scenario 2- 4 -2017 NRC Exam Scenario 2 D-1 Rev 1SCENARIO SETUPA. Reset Simulator to IC-162.B. Verify Scenario Malfunctions, Remotes, and Overrides are loaded, as listed in the Scenario Timeline.C. Verify Event Trigger 8 is set to "PZR Press < 1684 psia".D. Verify LPSI pump A is removed from service as follows:1. InsertSIR32 to RKOUT2. Place C/S in OFF with a Danger TagE. Place a copy of OP-903-046, EFW Operability Check, Section 7.1 on the BOP desk. Section 7.1should be place-kept with step 7.1.5 (Start EFW pump A) circled. Previous steps should be circled-slashed and step 7.1.3 (Check valve test) N/A'd. A copy of Attachment 10.1, EFW Pump A IST Data,should also be available with step 7.1.1 (Group B Test selected) filled in. Shift turnover should statethat the NAO is standing by the pump with the required paperwork in hand.F. Verify all EFW Flow Control Valves are in Auto and Caution Tags removed.G. Ensure Protected Train B sign is placed in SM office window.H. Verify EOOS is 10.0 Green with LPSI pump A out of service.I. Place Protected Equipment covers on running SFP pump and LPSI pump B control switches.J. Complete the simulator setup checklist.K. Start Insight, open file Crew Performance.tis.

NRC Scenario 2- 5 -2017 NRC Exam Scenario 2 D-1 Rev 1SIMULATOR BOOTH INSTRUCTIONSEvent 1 EFW Pump A trips on overcurrent during operability check1. Approximately 1 minute after the crew starts EFW Pump A, initiate EventTrigger 1.2. If Work Week Manager or PME are called, inform the caller that a work package will beassembled and a team will be sent to the Control Room.3. If sent to the breaker, wait 2 minutes and report overcurrent flags on all three phases.4. If sent to the pump, wait 5 minutes and report an acrid odor in the room but no signs of fire.Event 2 Hot Leg 1 Temperature, RC-ITI-0111X, Fails Low1. On Lead Examiner's cue, initiate EventTrigger 2.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.Event 3 RCP 2A Lower Seal Fails1. On Lead Examiner's cue, initiate EventTrigger 3.2. If the Duty Engineering or RCP Engineer is called inform the caller that you will monitor RCP 2Afor further degradation.3. If the Work Week Manager or PMM are called, inform the caller that a work package will beassembled for the next forced outage.Event 4 Steam Generator Pressure Instrument, SG-IPT-1013A, Fails Low1. On Lead Examiner's cue, initiate EventTrigger 4.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.3. If sent to LCP-43, report that SG-IPI-1013-A1 reads 0 PSIA. Observe other indications of SGpressure using Extreme View on LCP-43 and report actual pressure if asked.Event 5 Feedwater Heater 5B Tube Leak, Rapid Plant Power Reduction1. On Lead Examiner's cue, initiate EventTrigger 5.2. If called to verify Low Pressure Heater levels, verify levels using the PMC and report levels to theControl Room.3. If called to verify position of the Normal and Alternate Control Valves, verify valve positions usingthe PMC and report the position of the valves to the Control Room.4. If requested to monitor Polisher Vessel D/P and remove as necessary, acknowledge the report.5. If Work Week Manager or PMM are called, inform the caller that a work package will beassembled.6. If Chemistry is called to sample the RCS for Dose Equivalent Iodine due to the down power,acknowledge and report that samples will be taken 2-6 hours from notification time and if askedtell the caller your name is Joe Chemist.

NRC Scenario 2- 6 -2017 NRC Exam Scenario 2 D-1 Rev 1Event 6 RCP 2A Middle Seal Fails1. After the reactivity manipulation is satisfied and on lead examiner's cue, initiate EventTrigger 6.2. If the Duty Engineering or RCP Engineer is called inform the caller that you will monitor RCP 2Afor further degradation.3. If the Work Week Manager or PMM are called, inform the caller that a work package will beassembled.Event 7 Pressurizer Code Safety, RC-317A, Fails Open1. After the crew secures RCP 2A, initiate EventTrigger 7.2. If Chemistry is called to perform samples acknowledge the request.3. If called as NAO to verify proper operation of unloaded Emergency Diesel Generators, then wait 2minutes and manually initiate EventTrigger 20. Wait an additional minute and manually initiateEventTrigger 21 to acknowledge local EDG panels. Report that both A and B EDGs are runningproperly and unloaded.Event 8 HPSI Pump B Fails To AUTO Start1. External communications are not expected for this event.Event 9 HPSI Pump B Fails To AUTO Start & HPSI Pump A Degrades & Trips1. EventTrigger 8(for event 9) is automatically triggered when PZR Pressure is <1684 psia.2. If Work Week Manager or PME are called, inform the caller that a work package will beassembled and a team will be sent to the Control Room.3. If sent to the HPSI Pump A breaker, wait 2 minutes and report overcurrent flags on all three phases.4. If sent to HPSI Pump A, wait 5 minutes and report the pump is not running and there is nothingelse abnormal.At the end of the scenario, before resetting, end data collection and save the file as 2017Scenario 2-(start-end time).tid. Export to .csv file. Save the file into the folder for theappropriate crew.

NRC Scenario 2- 7 -2017 NRC Exam Scenario 2 D-1 Rev 1SCENARIO TIMELINE EVENT KEYDESCRIPTIONTRIGGER DELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTION 1FW06AMOTOR DRIVEN EFW PMP A TRIP 1 00:00:00 00:00:00ACTIVEMOTOR DRIVEN EFW PMP A TRIPS DURING OPERABILITY CHECK 2RC21ARCS HOT LEG 1 CONTROL TT 111X FAILS (0-100%)

2 00:00:00 00:00:00 0%HOT LEG 1 TEMPERATURE FAILS LOW 3RC08CRCP 2A LOWER SEAL FAILURE (0-100%)

3 00:00:00 00:00:00 100%RCP 2A LOWER SEAL FAILS 4 SG04EMS LINE IPT-1013A FAIL (0-100%)

4 00:00:00 00:00:00 0%SG 1 PRESSURE INSTRUMENT SG-IPT-1013A FAILS LOW 5FW35BLP FW HEATER 5B TUBE LEAK (100% = 10% OF TUBES) 5 00:00:00 00:00:30 15%FW HTR 5B TUBE LEAK FROM CONDENSATE TO HEATER SHELL, RAPID DOWN POWER TO < 72% POWER 6RC09CRCP 2A MIDDLE SEAL FAILURE (0-100%)

6 00:00:00 00:00:00 100%RCP 2A MIDDLE SEAL FAILS 7RC11A1CODE SAFETY RC-317A FAIL OPEN 7 00:00:00 00:00:00ACTIVEPRESSURIZER CODE SAFETY, RC-317A, FAILS OPEN 8 SI02BHPSI PUMP B FAILS TO AUTO STARTN/A 00:00:00 00:00:00ACTIVEHPSI PUMP B FAILS TO AUTO START 9 SI19A SI01AHPSI PUMP A DEGRADATION (Triggered when PZR Press <1684 psia)HPSI PUMP A TRIPPED (Triggered when PZR Press <1684 psia)8 (AUTO)8 (AUTO)00:02:00 00:03:00 00:01:00 00:00:00 100%ACTIVEHPSI PUMP A DEGRADES AND TRIPSN/A EGR26EDG A LOCAL ANNUN ACK 20 00:00:00 00:00:00 ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGEN/A EGR27EDG B LOCAL ANNUN ACK 21 00:00:00 00:00:00 ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGE NRC Scenario 2- 8 -2017 NRC Exam Scenario 2 D-1 Rev 1 EVENT KEYDESCRIPTIONTRIGGER DELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTIONN/A SIR32LPSI PUMP A (will not show up in summary tab)N/A 00:00:00 00:00:00RKOUTLPSI PUMP A BREAKER NRC Scenario 2- 9 -2017 NRC Exam Scenario 2 D-1 Rev 1REFERENCESEventProcedures 1OP-903-046, Emergency Feed Pump Operability Check, Rev. 318Technical Specification 3.7.1.2 2OP-901-110, Pressurizer Level Control Malfunction, Rev. 9OP-901-501, PMC or Core Operating Limits Supervisory System Malfunction, Rev. 15 3OP-901-130, Reactor Coolant Pump Malfunction, Rev. 11 4OP-009-007, Plant Protection System, Rev. 17OP-903-013, Monthly Channel Checks, Rev. 18Technical Specification 3.3.1Technical Specification 3.3.2Technical Specification 3.3.3.5Technical Specification 3.3.3.6 5OP-901-221, Secondary System Transient, Rev. 4OP-901-212, Rapid Plant Power Reduction, Rev. 8OP-002-005, Chemical and Volume Control, Rev. 56OP-004-004, Control Element Drive, Rev. 23 6OP-901-130, Reactor Coolant Pump Malfunction, Rev. 11OP-902-000, Standard Post Trip Actions, Rev. 16OP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart 7OP-902-002, Loss of Coolant Accident Recovery Procedure, Rev. 20OP-902-009, Standard Appendices, Rev. 315, Appendix 2, FiguresOP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart8 & 9 OP-902-002, Loss of Coolant Accident Recovery Procedure, Rev. 20GEN EN-OP-115, Conduct of Operations, Rev. 17EN-OP-115-08, Annunciator Response, Rev. 4EN-OP-200, Plant Transient Response Rules, Rev. 3OI-038-000, EOP Operations Expectations / Guidance, Rev. 14 Appendix DScenario OutlineForm ES-D 1 -2017 NRC Exam Scenario 3 D-1 Rev 1Facility:WaterfordScenario No.:

3Op Test No.:

1Examiners:Operators:Initial Conditions: ~ 1% Reactor Power; 1 st SGFP in service; AB Buses are aligned to Train B. Charging Pumps A and B running. No major equipment out of service.Turnover: Protected Train is B, Secure AFW pump, Raise power to 5-10% using CEAs.EventNo.Malf.No.EventType*EventDescription 1N/AR - ATCN - BOP N - SROSecure the Auxiliary Feedwater Pump and raise powerto 5-10% using CEAs in accordance with OP-010-003,Plant Startup.

2RX08AI - ATCI - SROPressurizer Level Controller, RC-ILIC-0110, fails offrequiring implementation of OP-901-110, PressurizerLevel Control Malfunction (E3).

3RP04B5AO-07A2M11-1I - BOPI - SROTS - SRORWSP Level Instrument, SI-ILI-0305B, fails low andgenerates an RAS trip requiring TS 3.3.2 entry andbypassing the affected trip bistable.

4 CC01BC - BOP C - SROTS - SROComponent Cooling Water Pump B trips requiringentry into OP-901-510, Component Cooling WaterSystem Malfunction (TS 3.7.3 & Cascading).

5RX14A RC14B1C - ATC C - SROSelected Pressurizer Pressure Control Channel (RC-IPR-100X) fails high and Pressurizer Spray Valve RC-301B fails open, requiring entry into OP-901-120,Pressurizer Pressure Control Malfunction and amanual reactor trip to secure selected ReactorCoolant Pumps and stop RCS depressurization.(CT1, RCS Pressure Control) 6ED01 A - DM - AllLoss of Off-site Power, OP-902-003, Loss of OffsitePower/Loss of Forced Circulation Recovery 7EG10AN/AEmergency Diesel Generator A trips on overspeed.

8 ED23BC - BOP C - SROEmergency Diesel Generator B Output Breaker fails toAUTO Close, due to the 3B to 2B Tie Breaker failing toopen on Undervoltage. Crew re-energizes B Safetybus.(CT 2, Energize a Safety Electrical Bus)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event DescriptionNRC Scenario 3- 2 -2017 NRC Exam Scenario 3 D-1 Rev 1The crew assumes the shift at ~ 1% power with instructions to secure the AFW pump and raise power to5-10% to roll the Main Turbine. All requirements have been met to change modes from MODE 2 toMODE 1. The SRO should direct raising power using Control Element Assemblies in accordance with thereactivity plan, OP-010-003, Plant Startup and OP-010-004, Power Operations.After the AFW pump is secured and the reactivity manipulation has been satisfied

,Pressurizer levelcontroller RC-ILIC-0100 fails off. The CRS should enter OP-901-110, Pressurizer Level ControlMalfunction, and implement section E3. This will require the ATC to control Letdown from CP-4. Thereare no Tech Spec consequences of the failure provided the crew restores letdown flow prior to exceeding62.5% level in the pressurizer.After the Pressurizer Level Controller Failure is addressed, RWSP Level instrument,SI-ILI-0305B, fails low and generates an RAS trip on channel B. The ATC operator will review theannunciators for this failure. The CRS should evaluate Tech Specs and enter Tech Spec 3.3.2 anddetermine that the Plant Protection System bistable (18) for Low RWSP Level must be bypassed within 1hour on Channel B. Tech Spec 3.3.3.5 and 3.3.3.6 should be referenced but not entered.After the Low RWSP bistable is bypassed, Component Cooling Water Pump B trips on overcurrent. TheSRO should enter OP-901-510, Component Cooling Water System Malfunction, and direct the start ofComponent Cooling Water Pump AB to replace Component Cooling Water Pump B. The SRO shouldenter Technical Specification 3.7.3 and cascading Technical Specifications per OP-100-014, TechnicalSpecification and Technical Requirements Compliance and comply with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action by performingOP-903-066, Electrical Breaker Alignment Check. Once CCW pump AB is in service Tech Spec 3.7.3 andcascading Tech Specs may be exited.After the SRO has addressed Technical Specifications, the selected Pressurizer Pressure Channel failshigh causing pressurizer spray to initiate. When the crew takes manual control of the Pressurizer SprayController, Pressurizer Spray Valve, RC-301B remains open. The crew should select Spray Valve A.When RC-301B remains open, the crew should determine a reactor trip is required to secure sufficientReactor Coolant Pumps to stop the RCS depressurization(Critical Task 1). The crew will be taking theactions required by OP-901-120, Pressurizer Control Malfunction but may not enter the procedure prior tothe reactor trip due to the pressure dropping in the RCS. The SRO should enter OP-902-000, StandardPost Trip Actions. In order to restore Pressurizer heaters the SRO will have to implement the section inthe offnormal to select the non-faulted pressurizer pressure control channel and the pressurizer levelmust recover above the low level heater cutout setpoint reset (~30%).After the crew has secured sufficient Reactor Coolant Pumps for the Spray valve failure and the crew isperforming Standard Post Trip Actions, a loss of off-site power occurs. Emergency Diesel Generator A willtrip on overspeed. Emergency Diesel Generator B will start but, its output breaker will fail to closeautomatically due to the 3B to 2B Bus Tie Breaker failing to open on undervoltage. The crew mustmanually trip the 3B to 2B Bus Tie Breaker which allows EDG B output breaker to close automatically andre-energize the B Safety Bus(Critical Task 2). If the crew fails to manually trip the 3B to 2B Bus TieBreaker, a station blackout results and EDG B will eventually overheat and fail due to no CCW cooling.The SRO should enter OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery. TheSRO should direct a non-licensed operator to restore power to the Dry Cooling Tower Sump Pumps. TheBOP should take action to protect the Main Condenser from over-pressurization. The scenario can beended after these actions are complete

, or at the lead examiner's discretion.

NRC Scenario 3- 3 -2017 NRC Exam Scenario 3 D-1 Rev 1Critical TaskNumberDescriptionBasis 1Establish RCS Pressure ControlThis task is satisfied by securingsufficient Reactor Coolant Pumps tostop Reactor Coolant Systemdepressurization prior to loss ofSubcooled Margin. This task becomesapplicable after Pressurizer Spray ValveB, RC-301B, fails open. (OP-901-120, E3 step 3)RCS subcooling is an integral part of adequatepressure control, inventory control, and Coreheat removal. The importance of keeping thefluid surrounding the Core in a subcooled statecarries a high degree of nuclear safetysignificance based on its direct relationship tothese safety functions.(ECS-98-001, S.01) 2Energize at Least One Safety Electrical BusThis task is satisfied by the crew takingaction to energize the B Safety Bus bytripping the 3B-to-2B Bus Tie breakerprior to failure of Emergency DieselGenerator B due to no ComponentCooling Water. This task becomesapplicable after the loss of offsite poweroccurs. (OP-902-000, 3.a.1)Failure to energize at least one emergency buswill result in the plant remaining in aconfiguration that will not support protection ifa subsequent event would occur. This lowersthe capability of the plant to mitigate an event.(TM-OP-100-03, CT-3)

• Critical Task (As defined in NUREG 1021 Appendix D)** Per NUREG-1021, Appendix D, If an operator or the crew significantly deviates from or fails to followprocedures that affect the maintenance of basic safety functions, those actions may form the basis of aCT identified in the post

-scenario review.Scenario Quantitative Attributes1. Malfunctions after EOP entry (1-2)22. Abnormal events (2-4)33. Major transients (1-2)14. EOPs entered/requiring substantive actions (1-2)15. EOP contingencies requiring substantive actions (0-2)06. EOP based Critical tasks (2-3)2 NRC Scenario 3- 4 -2017 NRC Exam Scenario 3 D-1 Rev 1SCENARIO NOTESA. Reset Simulator to IC-163.B. Verify Scenario Malfunctions are loaded, as listed in the Scenario Timeline.C. Verify all EFW Flow Control Valves are in Auto (remove Caution Tags on flow controllers).D. Verify Channel X is selected for PZR pressure control.E. Place a Protect Equipment cover on running SFP pump C/S.F. Ensure Protected Train B sign is placed in SM office window.G. Verify EOOS is 10.0 Green with no equipment out of service.H. Place a copy of OP-010-003, Plant Startup, on the Control Room desk with step 9.4.52.2 (secureAFW) circled and several of the previous steps circle-slashed to show progress. Fill in initials andcircle-slash steps 9.4.53 (adjust Blowdown), 9.4.59 (mode 1 Tech Spec logs) and 9.4.60 (Chemistrycontacted) as complete. Sign step 9.4.61 (SM permission to enter mode 1).I. Complete the simulator setup checklist.J. Establish the following trends:1. C24104 on CP3, CRT 6 (0-10 scale, 1 sec update)2. SG Wide Range levels on CP-35 (15 sec update)K. Start Insight, open file Crew Performance.tis.

NRC Scenario 3- 5 -2017 NRC Exam Scenario 3 D-1 Rev 1SIMULATOR BOOTH INSTRUCTIONSEvent 1 Secure AFW Pump and raise reactor power1. If called as an NAO to standby the AFW pump, acknowledge the communication. Wait 2 minutesand report you are standing by.2. If called as Chemistry to verify SG chemistry is within specification, inform the caller that SGchemistry is satisfactory. If asked for your name, say Joe Chemist.3. If called as an NAO to open or throttle open MS-148, acknowledge the communication. Wait 5minutes, report that you will be slowly opening/throttling MS-148, MS Supply to Gland SealIsolation. Initiate EventTrigger 1. After MS-148 completes ramping, report that MS-148 isopen/throttled open. If you are directed to further throttle open MS-148, simply acknowledge therequest, wait ~30 seconds and report the new throttled position. Repeat as necessary until it isreported that MS-148 is fully open.4. If called as an NAO to transfer Auxiliary Steam from Aux Boiler Steam to Main Steam,acknowledge the communication. Wait 15 minutes, and then report that Auxiliary Steam has beentransferred to Main Steam (no remote necessary).5. If called as an NAO to secure the Portable Auxiliary Boiler, acknowledge the communication. Wait5 minutes, initiate EventTrigger 20 and report that the Portable Aux Boiler is secured..Event 2 Pressurizer Level Controller, RC-ILIC-0110, Output Fails Off1. On Lead Examiner's cue, initiate EventTrigger 2.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.Event 3 RWSP Level Instrument, SI-ILI-0305B, Fails Low & RAS Trip Generated1. On Lead Examiner's cue, initiate EventTrigger 3.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.Event 4 Component Cooling Water Pump B Trips1. On Lead Examiner's cue, initiate EventTrigger 4.2. If called as the watchstander and sent to CCW Pump B, wait 3 minutes, report that the pumplooks normal locally.3. If called as the watchstander and sent to CCW Pump B breaker, wait 3 minutes, report that thebreaker indicates open and that there are various breaker parts on the floor of the cubicle.4. If Work Week Manager or PME are called, inform the caller that a work package will beassembled and a team will be sent to the Control Room.Event 5 Pressurizer Pressure Control Channel, RC-IPT-0100X, Fails High and Pressurizer SprayValve RC-301B Fails Open1. On Lead Examiner's cue, initiate EventTrigger 5.Event 6 Loss of Offsite Power1. On Lead Examiner's cue, initiate EventTrigger 6.2. If the Duty Plant Manager is called, inform the caller that you will make the necessary calls.3. If Chemistry is called to perform samples acknowledge the request.

NRC Scenario 3- 6 -2017 NRC Exam Scenario 3 D-1 Rev 1Event 7&8 EDG A Trips on Overspeed; 3B-to-2B Bus Tie Breaker Fails to Trip on UV1. If called as an NAO to investigate EDG A, wait 3 minutes; initiate EventTrigger 21 (EGR26) toacknowledge the local alarm panel and report that EDG A is not running, "EMERGENCY STOPor UNIT S/D" and "ENGINE OVERSPEED" alarms are locked in but there is no obvious signs ofdamage. If asked, report Overspeed Butterfly valve is tripped.2. If Work Week Manager or PMM are called, inform the caller that a team will be organized andsent to the field as soon as possible.3. If called as an NAO to check EDG B, wait 2 minutes; initiate EventTrigger 22 (EGR27) toacknowledge the local alarm panel. If EDG B output breaker is closed and CCW pump B isrunning, report EDG B is running and all parameters are normal. If CCW pump B is not running(i.e. EDG B Output breaker is not closed), report EDG B is running and "SERVICE WATER LOWFLOW" alarm is locked in.At the end of the scenario, before resetting, end data collection and save the file as 2017Scenario 3-(start-end time).tid. Export to .csv file. Save the file into the folder for theappropriate crew.

NRC Scenario 3- 7 -2017 NRC Exam Scenario 3 D-1 Rev 1SCENARIO TIMELINEEVENT KEYDESCRIPTIONTRIGGERDELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTION 1MSR09MS-148 MS to GS ISOL VALVE 100:00:0000:01:0012%SECURE AFW PUMP AND RAISE REACTOR POWER 2 RX08APZR LVL CONTROLLER 110 FAILS OFF 200:00:0000:00:00 ACTIVEPRESSURIZER LEVEL CONTROLLER RC-ILIC-0110 OUTPUT FAILS OFF 3 RP04B5AO-07A2M11-1TRIP GENERATED CH B RWSP LVL(RAS)CH B RWSP LEVEL 3 300:00:0000:00:0000:00:0000:00:00ACTIVE 0%RSWP CHANNEL B LEVEL INSTRUMENT SI-ILI-0305B FAILS LOW & RAS TRIP GENERATED 4 CC01BCCW PUMP B TRIP 400:00:0000:00:00 ACTIVECOMPONENT COOLING WATER PUMP B TRIP 5 RX14A RC14B1PZR PRESSURE CNTL CHL 100X FAIL (0-100%)(1500-2500 PSIA)PZR SPRAY VALVE RC-301B FAILS OPEN 5 500:00:0000:00:0000:00:0000:00:00 100ACTIVEPRESSURIZER PRESSURE CONTROL CHANNEL, RC-IPT-0100X, FAILS HIGH AND PRESSURIZER SPRAY VALVE, RC-301B, FAILS OPEN 6 ED01A ED01B ED01C ED01DFEEDER BREAKER 7172 TRIP IN SWITCHYARDFEEDER BREAKER 7176 TRIP IN SWITCHYARDFEEDER BREAKER 7182 TRIP IN SWITCHYARDFEEDER BREAKER 7186 TRIP IN SWITCHYARD 6 6 6 600:00:0000:00:0000:00:0000:00:0000:00:0000:00:0000:00:0000:00:00ACTIVEACTIVEACTIVEACTIVELOSS OF OFFSITE POWER 7EG10ADG A OVERSPEED TRIP 600:00:1000:00:00ACTIVEEDG A TRIPS ON OVERSPEED 8 ED23B3BS TO B2 BUS BREAKER FAILS TO TRIP ON UV 600:00:0000:00:00 ACTIVE3BS TO B2 BUS BREAKER FAILS TO TRIP ON UV NRC Scenario 3- 8 -2017 NRC Exam Scenario 3 D-1 Rev 1EVENT KEYDESCRIPTIONTRIGGERDELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTIONN/AMSR32TEMPORARY AUX BOILER 20N/AN/AOFFLINETEMPORARY AUX BOILER (16 MIN TILL RATED PRESS)N/AEGR26EDG A LOCAL ANNUN ACK 2100:00:0000:00:00ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGEN/AEGR27EDG B LOCAL ANNUN ACK 2200:00:0000:00:00ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGE NRC Scenario 3- 9 -2017 NRC Exam Scenario 3 D-1 Rev 1REFERENCESEventProcedures 1OP-010-003, Plant Startup, Rev. 342OP-003-035, Auxiliary Feedwater, Rev. 305OP-004-004, Control Element Drive, Rev. 23 2OP-901-110, Pressurizer Level Control Malfunction, Rev. 9 3OP-009-007, Plant Protection System, Rev. 17OP-903-013, Monthly Channel Checks, Rev. 18Technical Specification 3.3.2 4OP-901-510, Component Cooling Water Malfunction, Rev. 303Technical Specification 3.7.3 & Cascading 5OP-901-120, Pressurizer Pressure Control Malfunction, Rev. 302OP-902-000, Standard Post Trip Actions, Rev 16OP-902-009, Standard Appendices, Rev. 315, Appendix 1 (Diagnostic Flow Chart),Appendix 2 (Figures) 6OP-902-003, Loss of Offsite Power/Loss of Forced Circ Recovery Procedure, Rev. 107 & 8 OP-902-000, Standard Post Trip Actions, Rev. 16GEN EN-OP-115, Conduct of Operations, Rev. 17EN-OP-115-08, Annunciator Response, Rev. 4EN-OP-200, Plant Transient Response Rules, Rev. 3OP-100-014, TS and TRM Compliance, Rev. 336OI-038-000, EOP Operations Expectations / Guidance, Rev. 14 Appendix DScenario OutlineForm ES-D 1 -2017 NRC Exam Scenario 4 D-1 Rev 1Facility:Waterford 3Scenario No.:

4Op Test No.:

1Examiners:Operators:Initial Conditions: Reactor power is ~90%. AB Buses are aligned to Train B. No major equipment out of service. Heater Drain Pump B is secured. Charging Pumps B (lead) and AB running.Turnover: Protected Train is B; Maintain power while PMI troubleshoots a Heater Drain Pump B annunciator.EventNo.Malf.No.EventType*EventDescription 1FW51ATS - SROCondensate Storage Pool level instrument EFW-ILI-9013A fails low. (TS 3.3.3.5, TS 3.3.3.6) 2CV30A2C - ATC C - SROLetdown Flow Control Valve, CVC-113A, fails closedrequiring entry into OP-901-112, Charging or LetdownMalfunction.

3FW26AI - BOPI - SROSteam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low. OP-901-201, Steam GeneratorLevel Control Malfunction. (TRM 3.3.5) 4 RD02A11R - ATCN - BOP N - SROTS - SROCEA 11 drops into the core requiring a rapid plantdown power in accordance with OP-901-212, RapidPlant Power Reduction. OP-901-102, CEA orCEDMCS Malfunction. (TS 3.1.3.1) 5 RC23A CV02AC - ATC C - SRORCS Cold Leg leak; Charging Pump A fails to auto-start.6 RC23AM - AllThe leak grows into a LOCA requiring implementationof OP-902-000, Standard Post Trip Actions and OP-902-002, Loss of Coolant Accident RecoveryProcedure. Stop RCPs(CT 1, Trip RCPs exceedingoperating limits)

.7 CS04BC - BOP C - SROCS-125B, Containment Spray Header B Isolation, failsto auto-open requiring manual action to open CS-125B(CT 2, Containment Temperature & Pressurecontrol).8MS11BM - AllMain Steam Line 2 Break inside containment requiringentry into OP-902-008, Functional RecoveryProcedure.

9 CS01AC - BOP C - SROContainment Spray Pump A trips requiring action tooverride close CS-125A, Containment Spray Header AIsolation(CT 3, Containment Isolation)

.*(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event DescriptionNRC Scenario 4- 2 -2017 NRC Exam Scenario 4 D-1 Rev 1The crew assumes the shift at ~90% power with instructions to start Heater Drain Pump B and continuethe power ascension after PMI resolves a problem with the Low Suction Pressure annunciator on HeaterDrain Pump B.PMI will not resolve the annunciator problem and the crew will maintain ~90% power. Nomajor equipment is out of service.After the crew takes the shift,Condensate Storage Pool level indicator EFW-ILI-9013 A will fail low. TheSRO should use OP-903-013, Monthly channel Checks, and enter Tech Spec 3.3.3.5 and 3.3.3.6.After Technical Specifications are addressed, the in-service letdown flow control valve, CVC-113A, failsclosed. The SRO should enter OP-901-112, Charging or Letdown Malfunction and implement SectionE2, Letdown Malfunction, and place the backup flow control valve, CVC-113B, in-service. The SRO mayimplement EN-OP-200, Transient Response Rules.After the backup letdown flow control valve has been placed in service, Steam Generator #1 Feedwaterflow instrument FW-IFR-1111 fails low. The Feedwater Control System will respond by increasingFeedwater flow to Steam Generator #1. The SRO should direct the BOP to take manual control andmatch Feedwater and Main Steam flow. The SRO should enter OP-901-201, Steam Generator LevelControl Malfunction. Feedwater controls for Steam Generator #1 may remain in manual as a result of thisfailure requiring manual control on a plant down power or reactor trip. The Ultrasonic Flow Meter will failas a result of the instrument failure and require entry into TRM 3.3.5. The SRO may implement EN-OP-200, Transient Response Rules.After the crew has worked through OP-901-201 and level in Steam Generator 1 is between 50% and 70%Narrow Range,CEA 11 (Reg. Group 4) drops into the core. The SRO should enter procedure OP-901-102, CEA or CEDMCS Malfunction and proceed to section E 1, CEA Misalignment Greater than 7 inches.The SRO will direct the BOP to adjust turbine load to match TAVG to TREF initially and then perform a rapidplant downpower in accordance with OP-901-212, Rapid Plant Power Reduction. RCS direct borationmust commence within 15 minutes of the dropped CEA to comply with Technical Specifications and theCOLR. The SRO should enter procedure OP-901-501, PMC or COLSS Malfunction. Actions in OP-901-501 are normally performed by the STA. The SRO should evaluate and enter TS 3.1.3.1 action c. TheSRO should implement EN-OP-200, Transient Response Rules.After the reactivity manipulation has been satisfied, an RCS leak will occur. The RCS leak will ramp into amedium break LOCA. Charging Pump A will fail to auto start requiring a manual start by the ATC. TheSRO may enter OP-901-111, RCS System Leak, but will soon recognize that Pressurizer level is notbeing maintained with available Charging pumps and should direct a manual reactor trip and manualinitiation of Safety Injection and Containment Isolation. The SRO should implement OP-902-000,Standard Post Trip Actions and diagnose to OP-902-002, Loss of Coolant Accident Recovery Procedure.The ATC should stop RCPs exceeding operating limits as RCS pressure lowers or within three minutes ofa Containment Spray actuation(CRITICAL TASK 1). Containment Spray Header B Isolation (CS-125B)will fail to open automatically requiring the BOP to manually open CS-125B(CRITICAL TASK 2)

.After the crew diagnoses to OP-902-002, Main Steam Line 2 breaks inside Containment. ContainmentSpray Pump A will trip on overcurrent. The SRO should go to OP-902-009 Appendix 1, DiagnosticsFlowchart and diagnose to OP-902-008, Functional Recovery OR go directly to OP-902-008 based ontwo events in progress per OP-100-017, Emergency Operating Procedures Implementation Guide. Whenthe SRO performs prioritization Containment Isolation (CI-1) should be the highest priority. The SROshould direct the BOP to override and close CS-125A, Containment Spray Header A Isolation(CRITICALTASK 3).The scenario can be terminated once the crew closes Containment Spray Header A Isolation inaccordance with OP-902-008, Functional Recovery procedure or at the lead examiner's discretion.

NRC Scenario 4- 3 -2017 NRC Exam Scenario 4 D-1 Rev 1Critical TaskNumberDescription Basi s 1Trip Any RCP Exceeding Operating LimitsThis task is satisfied by stopping all runningRCPs within 3 minutes of loss of ComponentCooling Water flow or prior to completing thestep that verifies RCP operating limits. Thistask becomes applicable after either runningRCP Vibration alarms actuate ORContainment Spray is initiated, whicheveroccurs first. (OP-902-002, 9.b or 9.d.1)The time requirement of 3 minutes is based on theRCP operating limit of 3 minutes without CCWcooling. Continued operation of RCP without CCWor outside of the operating limits could lead to afailure of the RCS pressure boundary at the RCPseal.(TM-OP-100-03, CT-23; ECS98-001, D.10) 2Establish Containment Temperature andPressure ControlThis task is satisfied by manually openingCS-125B, Containment Spray Header BIsolation, prior to exceeding containmentdesign pressure of 44 PSIG or prior tocompleting Containment Spray (CS)verification in OP-902-002 or exiting theContainment Temperature and PressureControl Safety Function in OP-902-008. Thistask becomes applicable after CS is initiatedand is critical after CS Pump A trips. (OP-902-002, step 14 or OP-902-008, CTPC-2)The maximum design pressure of the containmentstructure is 44 psig. Failure to take action toestablish containment pressure and temperaturecontrol may result in containment pressureexceeding maximum design and therefore exceeddesign leakage of containment. The operatorsmonitor containment pressure along withContainment Spray and Containment Fan Cooleroperations as verification of adequate containmentheat removal and pressure mitigation.(TM-OP-100-03, CT-15; ECS98-001, P.28) 3Establish Containment IsolationThis task is satisfied by closing CS-125A,Containment Spray Header A Isolation, priorto exiting the Containment Isolation (CI-1)Safety Function in OP-902-008. This taskbecomes applicable after ContainmentSpray (CS) is initiated and CS Pump A trips.(OP-902-00 8 , CI-1, 1.c.1)A Loss of Coolant Accident that has occurred insidecontainment and has not been isolated will result inexcess radioactivity leaving containment and beingreleased to the public.(TM-OP-100-03, CT-9)

• Critical Task (As defined in NUREG 1021 Appendix D)** Per NUREG-1021, Appendix D, If an operator or the crew significantly deviates from or fails to followprocedures that affect the maintenance of basic safety functions, those actions may form the basis of a CTidentified in the post-scenario review.Scenario Quantitative Attributes1. Malfunctions after EOP entry (1-2)

22. Abnormal events (2-4)

33. Major transients (1-2)

24. EOPs entered/requiring substantive actions (1-2)

15. EOP contingencies requiring substantive actions (0-2)

16. EOP based Critical tasks (2-3) 3 NRC Scenario 4- 4 -2017 NRC Exam Scenario 4 D-1 Rev 1SCENARIO SETUPA. Reset Simulator to IC-164B. Verify Scenario Malfunctions, Remotes, and Overrides are loaded, as listed in the Scenario Timeline.C. Verify reactor power is ~90% with HDPs A and C running and HDP B secured with annunciatorF0802, Htr Drain Pump B Suction Press Lo, locked in.D. Verify all EFW Flow Control Valves are in Auto and caution tags removed.E. Verify CVC-113A (Normal Letdown FCV) is in service.F. Ensure Protected Train B sign is placed in SM office window.G. Place a Protected Equipment cover on running SFP pump C/S.H. Verify EOOS is 10.0 Green with nothing out of service.I. Complete the simulator setup checklist.J. Start Insight, open file Crew Performance.tis.

NRC Scenario 4- 5 -2017 NRC Exam Scenario 4 D-1 Rev 1SIMULATOR BOOTH INSTRUCTIONSEvent 1 Condensate Storage Pool level instrument EFW-ILI-9013 A fails low1. On Lead Examiner's cue, initiate EventTrigger 1.2. If called as an NAO to check the indication at the Remote Shutdown Panel, wait 2 minutes andreport that Condensate Storage Pool Level instrument EFW-ILI-9013 A1 is reading 0%.3. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.Event 2 Letdown Flow Control Valve, CVC-113A, Fails Closed1. On Lead Examiner's cue, initiate EventTrigger 2.2. If Work Week Manager or PMM are called, inform the caller that a work package will beassembled and a team will be sent to the Control Room.3. If called as NAO to place the alternate letdown flow control valve in service, open a copy of OP-901-112 and follow along on step 6 of subsection E2. When directed to slowly open CVC-111B,run Schedule File (CAEP):OP-901-112 Local Operator Actions\Placing Alternate LDFCV inService.sch. Make appropriate reports as the schedule file progresses.Event 3 Steam Generator #1 Feedwater Flow Instrument FW-IFR-1111 Fails Low1. On Lead Examiner's cue, initiate EventTrigger 3.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to the Control Room.Event 4 CEA 11 Falls into the core/Rapid plant power reduction1. On Lead Examiner's cue, initiate EventTrigger 4.2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembledand a team will be sent to CEDMCS Alley.3. If called as RAB and directed to CEDMCs Alley, respond in 3 minutes that you have arrived. Ifasked, report that there is no apparent cause for the dropped CEA.4. If Chemistry is called to sample the RCS for Dose Equivalent Iodine due to the down power,acknowledge and report that samples will be taken 2-6 hours from notification time and if askedtell the caller your name is Joe Chemist.5. If notified as Load Dispatcher (Woodlands) acknowledge the communications and inform thecaller that the grid will remain stable with available backup generation.6. If requested to monitor Polisher Vessel D/P and remove as necessary, acknowledge the report.Event 5 RCS Cold Leg Leak/Charging Pump A fails to auto start1. On Lead Examiner's cue, initiate EventTrigger 5.2. If the Duty Plant Manager is called, inform the caller that you will make the necessary calls.3. If Chemistry is called to perform samples acknowledge the request.

NRC Scenario 4- 6 -2017 NRC Exam Scenario 4 D-1 Rev 1Event 6 RCS Cold Leg Break1. There is no event trigger for this event (event trigger 5 initiates this event).2. If the Duty Plant Manager is called, inform the caller that you will make the necessary calls.3. If Chemistry is called to perform samples acknowledge the request.4. If called as NAO to verify proper operation of unloaded Emergency Diesel Generators, then wait 2minutes and manually initiate EventTrigger 20. Wait an additional minute and manually initiateEventTrigger 21 to acknowledge local EDG panels. Report that both A and B EDGs are runningproperly and unloaded.Event 7 CS-125B Fails to Open Automatically1. There is no event trigger for this event.2. External communications are not expected.Event 8 Main Steam Line 2 Break inside Containment1. On Lead Examiner's cue, initiate EventTrigger 8.2. If the Duty Plant Manager is called, inform the caller that you will make the necessary calls.3. If Chemistry is called to perform samples acknowledge the requestEvent 9 Containment Spray Pump A trips / Override CS-125A, CS Header A Isolation1. There is no trigger for Event 9. CS pump A trip is triggered by Event Trigger 8.2. If called as an NAO to override CS-125A report you are on your way to the Control Room to pickup the key. Have someone role play as NAO and enter the simulator to simulate getting the key.Wait 1 minute, insert remote CSR13A using EventTrigger 22. Make the report to the ControlRoom that you have done so.3. If called as an NAO to investigate the trip of CS pump A breaker, report overcurrent flags on all 3 phases.4. If called as an NAO to investigate CS Pump A, report that the paint on the motor is discolored andthere is a strong odor of burnt insulation, but no fire.At the end of the scenario, before resetting, end data collection and save the file as 2017Scenario 4-(start-end time).tid. Export to .csv file. Save the file into the folder for theappropriate crew.

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NRC Scenario 4- 7 -2017 NRC Exam Scenario 4 D-1 Rev 1SCENARIO TIMELINE EVENT KEYDESCRIPTIONTRIGGER DELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTION 1FW51AFAIL CSP LPL XMTR EFW-ILT-9013A (0-100%)

1 00:00:00 00:00:00 0%CONDENSATE STORAGE POOL LEVEL INSTRUMENT EFW-ILI-9013A FAILS LOW 2CV30A2LTDN FLOW CONTROL VALVE CVC-113A FAILS CLOSED 2 00:00:00 00:00:00ACTIVELETDOWN FLOW CONTROL VALVE, CVC-113A, FAILS CLOSED 3FW26AFW FLOW TRANSMITTER 1111 FAIL (0-100% OF RANGE) 3 00:00:00 00:00:00 27%SG1 FEED FLOW INST (FW-IFR-1111) FAILS LOW TO 27%

4RD02A11 DROPPED CEA 11 4 00:00:00 00:00:00ACTIVECEA 11 DROPS INTO THE CORE; RAPID PLANT POWER REDUCTION 5RC23A CV02ARCS COLD LEG 1A RUPTURECHARGING PUMP A FAIL TO AUTOSTART 5N/A 00:00:00 00:00:00 00:08:00 00:00:00 1.5ACTIVERCS COLD LEG LEAK / CHARGING PUMP A FAILS TO AUTO START 6RC23ARCS COLD LEG 1A RUPTURE 5 00:08:00 00:00:00 14RCS COLD LEG BREAK 7 CS04BCS TRAIN B CS-125B FAILS TO AUTO OPENN/A 00:00:00 00:00:00ACTIVECS-125B FAILS TO AUTO OPEN 8 MS11BMS LINE B BREAK INSIDE CNTMT (0-100% = 40 IN) 8 00:00:00 00:00:00 10%MAIN STEAM LINE B BREAK INSIDE CONTAINMENT 9 CS01ALOSS OF CNTMT SPRAY PUMP A 8 00:00:00 00:00:00ACTIVELOSS OF CNTMT SPRAY PUMP AN/A EGR26EDG A LOCAL ANNUN ACK 20 00:00:00 00:00:00 ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGEN/A EGR27EDG B LOCAL ANNUN ACK 21 00:00:00 00:00:00 ACKNLOCAL EDG ANNUNCIATOR ACKNOWLEDGE NRC Scenario 4- 8 -2017 NRC Exam Scenario 4 D-1 Rev 1 EVENT KEYDESCRIPTIONTRIGGER DELAYHH:MM:SSRAMPHH:MM:SSFINALEVENT DESCRIPTIONN/ACSR13ACS-125A REMOTE KEY SW TO CLOSE VALVE 22 00:00:00 00:00:00OVRDCS-125A REMOTE KEY SW TO CLOSE VALVEN/A F_Q12HTR DRAIN PUMP B SUCTION PRESS LON/A 00:00:00 00:00:00FAIL ONANNUNCIATOR MALFUNCTION OVERRIDE NRC Scenario 4- 9 -2017 NRC Exam Scenario 4 D-1 Rev 1REFERENCESEventProcedures 1OP-903-013, Monthly Channel Checks, Rev. 18Tech Spec 3.3.3.5Tech Spec 3.3.3.6 2OP-901-112, Charging or Letdown Malfunction, Rev. 6 3OP-901-201, Steam Generator Level Control Malfunction, Rev. 6TRM 3.3.5 4OP-901-102, CEA or CEDMCS Malfunction, Rev. 304OP-901-212, Rapid Plant Power Reduction, Rev. 8OP-901-501, PMC or COLSS Malfunction, Rev. 15OP-004-004, Control Element Drive, Rev. 23Tech Spec 3.1.3.1 5OP-902-000, Standard Post Trip Actions, Rev. 16OP-902-009, Standard Appendices, Rev. 315, Appendix 2, FiguresOP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart6 & 7 OP-902-002, Loss of Coolant Accident Recovery, Rev. 20OP-902-009, Standard Appendices, Rev. 315, Appendix 2, Figures8 & 9 OP-902-008, Functional Recovery, Rev. 26OP-902-009, Standard Appendices, Rev. 315, Appendix 21-AGEN EN-OP-115, Conduct of Operations, Rev. 17EN-OP-115-08, Annunciator Response, Rev. 4EN-OP-200, Plant Transient Response Rules, Rev. 3OI-038-000, EOP Operations Expectations / Guidance, Rev. 14OP-100-017, Emergency Operating Procedures Implementation Guide, Rev. 5