ML17137A190

2017-03 Final Outlines

ML17137A190
Person / Time
Site:	Waterford
Issue date:	04/05/2017
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML17137A190 (71)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (RO Exam) Date of Exam: March 27, 2017 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 3 3 3 3 3 3 18 6 Emergency &

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

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

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 7 Categories 3 3 2 2 10 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced 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. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

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

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

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

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

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

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

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

G* Generic K/As

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

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

1 2 3 1 2 CE/E02, EA1.2: Ability to operate and 000007 (BW/E02&E10; CE/E02) Reactor X / or monitor the following as they 3.3 1 Trip - Stabilization - Recovery / 1 apply to the (Reactor Trip Recovery):

Operating behavior characteristics of the facility (CFR: 41.7 / 45.5 / 45.6)

AK2.01: Knowledge of the 000008 Pressurizer Vapor Space X interrelations between the Pressurizer 2.7 2 Accident / 3 Vapor Space Accident and the following: Valves (CFR 41.7 / 45.7)

EK3.04: Knowledge of the reasons for 000009 Small Break LOCA / 3 X the following responses as the apply 4.1 3 to the small break LOCA: Starting additional Charging Pumps (CFR 41.5 /

41.10 / 45.6 / 45.13)

EA2.03: Ability to determine or 000011 Large Break LOCA / 3 X interpret the following as they apply 3.7 4 to a Large Break LOCA: Consequences of managing LOCA with loss of CCW (CFR 43.5 / 45.13)

AK2.07: Knowledge of the 000015/17 RCP Malfunctions / 4 X interrelations between the Reactor 2.9 5 Coolant Pump Malfunctions (Loss of RC Flow) and the following: RCP seals (CFR 41.7 / 45.7)

AK1.03: Knowledge of the operational 000022 Loss of Rx Coolant Makeup / 2 X implications of the following concepts 3.0 6 as they apply to Loss of Reactor Coolant Makeup: Relationship between charging flow and PZR level (CFR 41.8 / 41.10 / 45.3) 2.1.32: Ability to explain and apply 000025 Loss of RHR System / 4 X system limits and precautions. (CFR: 3.8 7 41.7).

AA2.01: Ability to determine and 000026 Loss of Component Cooling X interpret the following as they apply 2.9 16 Water / 8 to the Loss of Component Cooling Water: Location of a leak in the CCW System. (CFR: 43.5 / 45.13) 2.2.44: Ability to interpret control 000027 Pressurizer Pressure Control X room indications to verify the status 4.2 8 System Malfunction / 3 and operation of a system, and understand how operator actions and directives affect plant and system conditions (CFR: 41.5 / 43.5 / 45.12)

EA1.12: Ability to operate and 000029 ATWS / 1 X monitor the following as they apply to 4.1 9 a ATWS: M/G set power supply and reactor trip breakers (CFR 41.7 / 45.5

/ 45.6)

EK3.02: Knowledge of the reasons for 000038 Steam Gen. Tube Rupture / 3 X the following responses as the apply 4.4 10 to the SGTR: Prevention of secondary PORV cycling (CFR 41.5 / 41.10 / 45.6

/ 45.13)

CE/E05, EK1.2: Knowledge of the 000040 (BW/E05; CE/E05; W/E12) X operational implications of the 3.2 11 Steam Line Rupture - Excessive Heat following concepts as they apply to Transfer / 4 the (Excess Steam Demand): Normal, abnormal and emergency operating procedures associated with (Excess Steam Demand)(CFR: 41.8 / 41.10 /

45.3)

CE/E06, EK1.1: Knowledge of the 000054 (CE/E06) Loss of Main X operational implications of the 3.2 12 Feedwater / 4 following concepts as they apply to the (Loss of Feedwater): Components, capacity, and function of emergency systems. (CFR: 41.8 / 41.10 / 45.3)

EA2.03: Ability to determine or 000055 Station Blackout / 6 X interpret the following as they apply 3.9 13 to a Station Blackout: Actions necessary to restore power (CFR 43.5 /

45.13) 2.4.18: Knowledge of the specific 000056 Loss of Off-site Power / 6 X bases of the EOPs (CFR: 41.10 / 43.1 / 3.3 14 45.13) 000057 Loss of Vital AC Inst. Bus / 6 AK3.01: Knowledge of the reasons for 000058 Loss of DC Power / 6 X the following responses as they apply 3.4 15 to the Loss of DC Power: Use of dc control power by D/Gs (CFR 41.5,41.10

/ 45.6 / 45.1) 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 X AA1.01: Ability to operate and / or 2.7 17 monitor the following as they apply to the Loss of Instrument Air: Remote manual loaders W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 AK2.07: Knowledge of the 000077 Generator Voltage and Electric X interrelations between Generator 3.6 18 Grid Disturbances / 6 Voltage and Electric Grid Disturbances and the following: Turbine/Generator Control (CFR: 41.4, 41.5, 41.7, 41.10

/ 45.8)

K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18/

6

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

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

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 X AK2.01: Knowledge of the 2.7 20 interrelations between Emergency Boration and the following:

Valves (CFR 41.7 / 45.7) 000028 Pressurizer Level Malfunction / 2 X AA2.12: Ability to determine and 3.1 21 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) 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 X AK1.02: Knowledge of the 3.5 23 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) 000051 Loss of Condenser Vacuum / 4 X AK3.01: Knowledge of the reasons 2.8 22 for the following responses as they apply to the Loss of Condenser Vacuum: Loss of Steam Dump capability upon loss of condenser vacuum. (CFR 41.5,41.10

/ 45.6 / 45.13) 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 X 2.4.31: Knowledge of annunciator 4.2 24 alarms, indications, or response procedures (CFR: 41.10 / 45.3) 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 AK2.03 Knowledge of the 000068 (BW/A06) Control Room Evac. / 8 X interrelations between the 2.9 25 Control Room Evacuation and the following: Controllers and positioners (CFR 41.7 / 45.7) 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 X EA1.01: Ability to operate and 4.2 26 monitor the following as they apply to a Inadequate Core Cooling: RCS Water Inventory (CFR 41.7 / 45.5 / 45.6) 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4

BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 X AA2.2: Ability to determine and 3.0 27 interpret the following as they apply to the (RCS Overcooling):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments (CFR: 43.5

/ 45.13)

AK1.1: Knowledge of the CE/A16 Excess RCS Leakage / 2 X operational implications of the 3.2 19 following concepts as they apply to the (Excess RCS Leakage):

Components, capacity, and function of emergency systems (CFR 41.8 / 41.10 / 45.3)

CE/E09 Functional Recovery K/A Category Point Totals: 2 2 1 1 2 1 Group Point Total: 9/4

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K6.14: Knowledge of the 003 Reactor Coolant Pump X effect of a loss or 2.6 28 malfunction on the following will have on the RCPS: Starting requirements (CFR: 41.7 / 45/5)

A4.07: Ability to manually 004 Chemical and Volume X operate and/or monitor in the 3.9 29 Control control room:

Boration/dilution (CFR: 41/7

/ 45.5 to 45.8)

X K5.30: Knowledge of the 3.8 51 operational implications of the following concepts as they apply to the CVCS:

Relationship between temperature and pressure in CVCS components during solid plant operation A4.02: Ability to manually 005 Residual Heat Removal X X operate and/or monitor in the 3.4 30 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. 4.6 31 (CFR: 41.10 / 43.5 / 45.12)

K4.17: Knowledge of ECCS 006 Emergency Core Cooling X design feature(s) and/or 3.8 32 interlock(s) which provide for the following: Safety Injection valve interlocks (CFR: 41.7)

A2.03: Ability to (a) predict 007 Pressurizer Relief/Quench X the impacts of the following 3.6 33 Tank 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)

K1.05: Knowledge of the 008 Component Cooling Water X X physical connections and/or 3.0 34 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 3.6 35 CCWS, including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal (CFR: 41.7 / 45.5)

K6.01: Knowledge of the 010 Pressurizer Pressure Control X X effect of a loss or 2.7 36 malfunction of the following will have on the PZR PCS:

Pressure detection systems (CFR: 41.7 / 45.7)

A1.01: Ability to predict 2.8 37 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)

K1.02: Knowledge of the 012 Reactor Protection X physical connections and/or 3.4 38 cause effect relationships between the RPS and the following systems: 125V dc system (CFR:

41.2 to 41.9 / 45.7 to 45.8)

K2.01: Knowledge of bus power 013 Engineered Safety Features X X supplies to the following: 3.6 39 Actuation 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 4.3 40 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)

K3.02: Knowledge of the 022 Containment Cooling X effect that a loss or 3.0 41 malfunction of the CCS will have on the following: Containment instrumentation readings (CFR: 41.7 / 45.6) 025 Ice Condenser K2.01: Knowledge of bus power 026 Containment Spray X supplies to the following: 3.4 42 Containment spray pumps (CFR:

41.7)

K5.08: Knowledge of the 039 Main and Reheat Steam X operational implications of 3.6 43 the following concepts as the apply to the MRSS:

Effect of steam removal on reactivity (CFR: 441.5 /

45.7)

K4.18: Knowledge of MFW 059 Main Feedwater X design feature(s) and/or 2.8 44 interlock(s) which provide for the following: Automatic feedwater reduction on plant trip (CFR: 41.7)

K2.02: Knowledge of bus power 061 Auxiliary/Emergency X X supplies to the following: 3.7 46 Feedwater AFW electric drive pumps(CFR:

41.7)

A3.03: Ability to monitor automatic operation of the AFW, including: AFW S/G level 3.9 47 control on automatic start (CFR: 41.7 / 45.5)

A3.05: Ability to monitor 062 AC Electrical Distribution X automatic operation of the ac 3.5 48 distribution system, including: Safety-related indicators and controls (CFR:

41.7 / 45.5)

A4.02: Ability to manually 063 DC Electrical Distribution X operate and/or monitor in the 2.8 49 control room: Battery voltage indicator (CFR: 41.7 / 45.5 to 45.8)

K4.10: Knowledge of ED/G 064 Emergency Diesel Generator X system design feature(s) 3.5 50 and/or interlock(s) which provide for the following:

Automatic load sequencer:

blackout (CFR: 41.7) 073 Process Radiation Monitoring X 2.4.4: Ability to recognize 4.5 52 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)

K1.19: Knowledge of the 076 Service Water X physical connections and/or 3.6 53 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)

X A1.02: Ability to predict 2.6 45 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 K3.01: Knowledge of the 078 Instrument Air X effect that a loss or 3.1 54 malfunction of the IAS will have on the following: Containment air system (CFR: 41.7 / 45.6) 2.2.44: Ability to interpret 103 Containment X control room indications to 4.2 55 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)

K/A Category Point Totals: 3 3 2 3 2 2 2 2 3 3 3 Group Point Total: 28/5

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive X K5.06: Knowledge of the 3.8 56 following operational implications as they apply to the CRDS: Effects of control rod motion on axial offset(CFR:

41.5/45.7)

K6.02: Knowledge of the effect 002 Reactor Coolant X or a loss or malfunction on the 3.6 57 following RCS components: RCP(CFR: 41.7 /

45.7) 011 Pressurizer Level Control 014 Rod Position Indication K4.07: Knowledge of NIS design 015 Nuclear Instrumentation X feature(s) and/or interlock(s) 3.7 58 provide for the following:

Permissives (CFR: 41.7) 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor X A4.02: Ability to manually 3.8 59 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) 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X A1.01: Ability to predict and/or 2.7 60 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) 034 Fuel Handling Equipment 035 Steam Generator A3.02: Ability to monitor 041 Steam Dump/Turbine Bypass X automatic operation of the SDS, 3.3 61 Control including: RCS Pressure, RCS Temperature, and reactor power (CFR: 41.7 / 45.5) 045 Main Turbine Generator X A2.08: Ability to (a) predict 2.8 62 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)

K3.01: Knowledge of the effect 055 Condenser Air Removal X that a loss or malfunction of 2.5 63 the CARS will have on the following: Main condenser (CFR:

41.7 / 45.6) 056 Condensate 068 Liquid Radwaste K1.04: Knowledge of the physical 071 Waste Gas Disposal X connections and/or cause-effect 2.7 64 relationships between the Waste Gas Disposal System and the following systems: Station ventilation (CFR: 41.2 to 41.9 /

45.7 to 45.8) 072 Area Radiation Monitoring 2.1.30: Ability to locate and 075 Circulating Water X operate components, including 4.4 65 local controls. (CFR: 41.7 /

45.7) 079 Station Air 086 Fire Protection K/A Category Point Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10/

3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 (RO) Date of Exam: September 14, 2015 Category K/A # Topic RO SRO-Only IR # IR #

Knowledge of shift or short-term relief turnover practices.

2.1.3 3.7 66 (CFR: 41.10 / 45.13)

Ability to use plant computers to evaluate system or 2.1.19 3.9 67 component status. (CFR: 41.10 / 45.12)

Knowledge of RO duties in the control room during fuel 2.1.44 3.9 68

1. handling, such as responding to alarms from the fuel Conduct of handling area, communication with the fuel storage Operations facility, systems operated from the control room in support of fueling operations, and supporting instrumentation. (CFR: 41.10 / 43.7 / 45.12) 2.1.

Subtotal Knowledge of the process used to track inoperable 2.2.43 3.0 69 alarms. (CFR: 41.10 / 43.3 / 45.13)

Ability to determine Technical Specification Mode of

2. 2.2.35 3.6 70 Operation. (CFR: 41.7 / 41.10 / 43.2 / 45.13)

Equipment Control 2.2.39 Knowledge of less than or equal to one hour Technical 3.9 71 Specification action statements for systems.(CFR: 41.7 /

41.10 / 43.2 / 45.13)

Subtotal Knowledge of radiological safety procedures pertaining to 2.3.13 3.4 72 licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 /

3. 45.10)

Radiation Knowledge of radiation monitoring systems, such as fixed 2.3.15 2.9 73 Control radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR:

41.12 / 43.4 / 45.9) 2.3.

Subtotal Knowledge of how abnormal operating procedures are 2.4.8 3.8 74 used in conjunction with EOPs.(CFR: 41.10 / 43.5 /

45.13)

4. Knowledge of fire protection procedures.

Emergency 2.4.25 3.3 75 (CFR: 41.10 / 43.5 / 45.13)

Procedures /

Plan 2.4.

2.4.

Subtotal Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/1 009 EK3.04 009 EK 3.17 rejected because suitable distractors were not available for knowledge of the reason for a containment RO 3 isolation. All the distractors we came up with were not plausible.

1/1 025 2.1.32 025 2.1.27 rejected because it was not possible to create a question that asked about the purpose of an offnormal RO 7 procedure and include credible distractors.

1/1 056 2.4.18 056 2.4.45 rejected because we could not develop a question with plausible distractors for a loss of offsite power that would RO 14 require a prioritization of annunciators (first part of the K/A) 1/1 026 AA2.01 062 AA2.02 The only possible loss of ACCW is a pump trip.

The only ACCW pump trips is overcurrent and undervoltage.

RO 16 A question was developed on the ACCW pump failure to start on the EDG sequencer but discovered it overlapped a question on one of the previous two exams. Could not develop a question that did not involve the EDG sequencer that would have plausible distractors.

1/1 065 AA1.01 065 AA1.04 is related to an Emergency Air Compressor for a Loss of Instrument Air. W3 does not have an Emergency Air RO17 Compressor.

1/1 077 AK2.07 077 AK2.02 Could not link breakers, relays with generator voltage and grid disturbance that could be supported with RO 18 adequate technical reference.

1/2 CE/A16 AK1.1 0003 AK1.19 rejected because we could not identify plausible distractors when creating a question. There is no references RO 19 in OPS procedures that discussed differential rod worth during a dropped CEA.

1/2 051 AK3.01 032 AK3.01 rejected because we could not develop a loss of source range question that did not duplicate existing RO RO 22 questions on this exam (RO68) or the previous two exams.

1/2 074 EA1.01 074 EA1.13 and the K/A for question number 59 were both asking about subcooling monitors. Rejected the K/A for RO RO 26 26 because we could not develop independent questions.

2/1 003 K6.14 003 K6.04 rejected because we could not develop a question where a fault of a CIV would affect a RCP. CBO flow has two RO 28 CIVs but isolating them has no effect due to flow being directed to another location, and the CCW valves to the RCPs fail in the open (no effect) position.

2/1 012 K1.02 012 K1.04 rejected. The rod position indicating system does not have a direct physical connection with the W3 RPS RO 38 system. Could not develop a question that effectively matched the K/A.

2/1 076 A1.02 059 A1.03 rejected because a question could not be developed that did not require a reference and a direct look RO 45 up.

2/1 004 K5.30 073 K5.01 rejected because we could only find questions for this K/A that pertained to radiation theory. We could not apply RO 51 this theory to the Process Radiation Monitors and have adequate technical references.

2/1 076 K1.19 076 K1.05 rejected because the ACCW system does not have a physical connection with the EDG system.

RO 53 2/2 041 A3.02 041 A3.05 rejected. Two independent Main Steam crossover pressures would have to fail such that the SDS is affected RO 61 (permissive and demand). The question developed for this K/A was not plausible.

3/2 2.2.43 2.2.6 rejected. The W3 procedure had mostly steps for managing the database and minutia steps for procedure RO 69 changes. The comment made documenting a problem with plausible distractors and cuing. Could not develop new distractors so a new K/A was randomly selected.

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (SRO) Date of Exam: March 27, 2017 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 18 3 3 6 Emergency &

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

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced 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. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

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

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

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

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

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

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

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

G* Generic K/As

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

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

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 X EA2.14: Ability to determine and interpret the 4.4 76 following as they apply to a small break LOCA:

Actions to be taken if PTS limits are violated (CFR 43.5) 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 2.2.22: Knowledge of limiting conditions for 000022 Loss of Rx Coolant Makeup / 2 X operations and safety limits.(CFR: 41.5 / 43.2 / 4.7 81 45.2) 000025 Loss of RHR System / 4 X 2.4.9: Knowledge of low power/shutdown 4.2 77 implications in accident (LOCA or loss of heat removal) mitigation strategies (CFR 41.10, 43.5, 45.13) 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 000038 Steam Gen. Tube Rupture / 3 AA2.01 Ability to determine and interpret the 000040 (BW/E05; CE/E05; W/E12) X following as they apply to the Steam Line Rupture: 4.7 78 Steam Line Rupture - Excessive Heat Occurrence and location of a steam line rupture Transfer / 4 from pressure and flow indications (CFR 43.5,45.13) 000054 (CE/E06) Loss of Main Feedwater / 4 2.4.6: Knowledge of EOP mitigation strategies.

000055 Station Blackout / 6 X (CFR: 41.10 / 43.5 / 45.13) 4.7 79 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 X AA2.03: Ability to determine and interpret the following as they apply to the Loss of DC Power:

DC loads lost; impact on ability to operate and 3.9 80 monitor plant systems(CFR: 43.5) 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4

BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 3 Group Point Total: 18/

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ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

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

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 X 3.9 82 AA2.05: Amount of boron to add to achieve required shutdown margin (CFR 43.5) 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 X 2.3.12 Knowledge of radiological safety 3.7 84 principles pertaining to licensed-operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high rad areas, aligning filters, etc. (CFR 43.7) 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 X AA2.17: Ability to determine and interpret 4.3 85 the following as they apply to plant fire on site: systems that may be affected by the fire (CFR 43.5) 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery X 2.2.37: Ability to determine operability and/or 4.6 83 availability of safety related equipment.

(CFR 43.5)

K/A Category Point Totals: 2 2 Group Point Total: 9/4 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump A2.15: Ability to (a) predict the impacts 004 Chemical and Volume X of the following malfunctions or 3.7 86 Control operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High or low PZR level (CFR 43.5) 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water X 2.2.40: Ability to apply Technical 4.7 87 specifications for a system.(CFR: 43.2 /

43.5) 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features X A2.06: Ability to (a) predict the impacts 4.0 88 Actuation of the following malfunctions or operations on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent ESFAS Actuation (CFR 43.5) 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency X A2.04: Ability to (a) predict the impacts 3.8 90 Feedwater of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: pump failure or improper operation (CFR 43.5) 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator X 2.2.25 Knowledge of the bases in 4.2 89 Technical Specifications for limiting conditions for operation and safety limits (CFR 43.2) 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air

103 Containment K/A Category Point Totals: 3 2 Group Point Total: 28/

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication X 2.4.11 Knowledge of abnormal condition 4.2 91 procedures.(CFR: 41.10 / 43.5 / 45.13) 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge X A2.03: Ability to (a) predict the impacts of 3.1 92 the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Startup operations and the associated required valve lineups. (CFR 43.5) 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air X 2.4.8 Knowledge of how abnormal 4.5 93 operating procedures are used in conjunction with EOPs (CFR 43.5) 086 Fire Protection

K/A Category Point Totals: 1 2 Group Point Total: 10/

3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 (SRO) Date of Exam: September 14, 2015 Category K/A # Topic RO SRO-Only IR # IR #

Knowledge of individual licensed operator responsibilities 2.1.4 3.8 94 related to shift staffing, such as medical requirements, no solo operation, maintenance of active license status,

1. 10CFR55, etc. (CFR 43.2)

Conduct of Ability to identify and interpret diverse indications to 2.1.45 4.3 95 Operations validate the response of another indication (CFR 43.5) 2.1.

Subtotal Knowledge of tagging and clearance procedures 2.2.13 4.3 96 2.

2.2.12 Knowledge of surveillance procedures (CFR 43.5) 4.1 97 Equipment Control 2.2.

Subtotal Ability to control releases (CFR 43.4) 2.3.11 4.3 98 3.

2.3.

Radiation Control 2.3.

Subtotal Knowledge of the parameters and logic used to assess 2.4.21 4.6 99 the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant integrity, containment conditions, radioactivity release control, etc.

4. (CFR 43.4)

Emergency Knowledge of events related to system operation/status 2.4.30 4.1 100 Procedures / that must be reported to internal organizations or external Plan agencies, such as the State, the NRC, or the transmission system operator (CFR 43.5) 2.4.

Subtotal Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/1 040 AA2.01 038 EA2.14 rejected. Could not make an SRO question from the original K/A that did not conflict with the K/A for RO10.

SRO 3 They were too much alike and one of the K/As had to be (Q78) rejected.

1/1 058 AA2.03 058 AA2.02 rejected. Could not develop an SRO only question associated with the determination and interpretation SRO 5 of a low DC voltage.

(Q80) 1/1 022 2.2.22 077 2.2.22 rejected. There is a 2015 RO question (RO18) written on voltage and Electrical Grid Disturbance basis SRO 6 information. Could not devise an SRO question independent (Q81) of that one.

1/2 CE/E09 033 2.2.37 rejected. The intermediate range detectors at W3 are the log channel instruments. There is a failed log channel SRO 8 on the audit simulator exam (scenario 1). Was not able to (Q83) create a question that would not create overlap with the audit exam.

2/1 004 A2.15 004 A2.05 rejected. RCP seal failures do not have an effect on CVCS other than CBO going to the VCT. Could not SRO 11 develop a question based solely on rising CBO temperature (Q86) or flow rate.

2/1 013 A2.06 013 A2.02 rejected. This K/A was an exact duplicate of the K/A for RO40.

SRO 13 (Q88) 2/1 064 2.2.25 062 2.2.25 rejected. The basis section for AC distribution (TRM 3.8.1.1) was used as an RO question on the 2015 RO SRO 14 exam (RO18). Could not develop a question that did not (Q89) duplicate the RO question.

2/1 061 A2.04 103 A2.03 rejected. W3 does not utilize a Phase A and Phase B Isolation system.

SRO15 (Q90) 2/2 029 A2.03 017 A2.02 rejected. Could not locate enough procedural guidance on the effects that core damage would have on SRO 17 Core Exit Thermocouples to develop an SRO test question.

(Q92) 3/2 2.2.13 2.2.5 rejected. The steps in the W3 procedure for making design changes to the facility are cumbersome and designing SRO21 a specific question on these detailed steps resulted in (Q96) questions that are considered minutia.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Waterford 3 Date of Examination: Mar 27, 2017 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A1 2.1.18, Ability to make accurate, clear, and concise logs, records, status boards, and reports.

Conduct of Operations D, R Complete OP-004-005, Core Operating Limits Supervisory System Operation, Attachment 11.6, Calculation of K/A Importance: 3.6 Charging and Letdown Parameters.

A2 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Conduct of Operations N,R Determine times to boil and core uncovery in accordance with OP-901-131, Shutdown Cooling Malfunction.

K/A Importance: 4.3 A3 2.2.12, Knowledge of Surveillance Procedures.

Equipment Control Perform Keff Calculation in accordance with OP-903-090, N,R Shutdown Margin, Section 7.5, Keff Calculation.

K/A Importance: 3.7 A4 2.3.11, Ability to control radiation releases.

Radiation Control Evaluate Meteorological conditions for gaseous release P,D,R from the Gaseous Waste Management System in accordance with OP-007-003, Gaseous Waste K/A Importance: 3.8 Management.

(From 2014 NRC Exam)

Emergency Plan Not Selected NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are 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) 2017 NRC Revision 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Waterford 3 Date of Examination: Mar 27, 2017 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A5 2.1.18, Ability to make accurate, clear, and concise logs, Conduct of Operations records, status boards, and reports.

D,R Review and approve completed OP-903-117, Emergency Diesel Generator Fuel Oil Transfer Pump Operability Check, K/A Importance: 3.8 Attachment 10.1, Fuel Oil Transfer Pump A IST Data.

A6 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Conduct of Operations N,R Determine time to boil and identify containment closure requirements in accordance with OP-901-131, Shutdown K/A Importance: 4.4 Cooling Malfunction.

A7 2.2.12, Knowledge of Surveillance Procedures Equipment Control Review Keff Calculation in accordance with OP-903-090, N,R Shutdown Margin, Section 7.5, Keff Calculation. Applicant K/A Importance: 4.1 determines Keff does not meet Tech Spec 3.1.2.9 requirements and identifies required corrective actions.

A8 2.3.14, Knowledge of radiation or contamination hazards Radiation Control that may arise during normal, abnormal, or emergency conditions or activities.

Calculate dose and assign non-licensed operators to vent K/A Importance: 3.8 P,M,R Safety Injection piping in Safeguards Room A. Given dose rate with and without shielding installed, time to install shielding, and job completion time using 1 team or using 2 teams, determine proper job assignment.

(Modified from 2014 NRC Exam)

A9 2.4.41, Knowledge of the emergency action level thresholds Emergency Plan and classifications.

N,R Determine appropriate Emergency Plan action level in accordance with EP-001-001, Recognition and K/A Importance: 4.6 Classification of Emergency Conditions.

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are 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) 2017 NRC Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: Mar 27, 2017 Exam Level RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1 001 Control Rod Drive System Place Reactor Cutback (RXC) in service and perform Immediate Operator Actions following a Cutback with unanalyzed rod configuration.

Alt. Path: Feedwater pump will trip resulting in a RXC. During the A,D,S 1 cutback, an incorrect CEA will drop.

GEN 2.4.49 The ability to perform without reference to procedures those actions that require immediate operation of system components and controls. RO - 4.6, SRO - 4.4 S2 006 Emergency Core Cooling System Reduce RCS pressure and use High Pressure Safety Injection Pumps to restore Pressurizer level in accordance with OP-901-112, Charging D,L,S 3 or Letdown Malfunction.

A1.18 PZR level and pressure RO - 4.0, SRO - 4.3 S3 003 Reactor Coolant Pump System Perform 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 of A,D,L,P,S 4P remaining Reactor Coolant Pumps. (W3 OE)

A2.02 Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP RO - 3.7, SRO - 3.9 S4 061 Emergency Feedwater System Reset EFW Pump AB after Overspeed Trip in accordance with OP-009-003, Emergency Feedwater (Control Room actions) EN,L,N,S 4S GEN EPE 074 EA1.07 AFW System RO - 4.2, SRO - 4.3 S5 022 Containment Cooling System Perform OP-903-037, Containment Cooling Fans Operability Verification D,S 5 A4.01 CCS Fans RO - 3.6, SRO - 3.6 S6 064 Emergency Diesel Generator (ED/G) System Parallel Emergency Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator.

Alt. Path: After EDG A load is raised, EDG A load will rise without A,D,S 6 manipulation requiring a trip of EDG A.

A4.06 Manual start, loading, and stopping of the ED/G RO - 3.9, SRO - 3.9 1 2017 NRC Revision 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S7 012 Reactor Protection System Reset High Containment Pressure ESFAS trip in accordance with OP-902-009, EOP Standard Appendices, Att. 5-D. EN,L,N,S 7 A4.04 Bistable, trips, reset and test switches RO - 3.3, SRO - 3.3 S8 068 Liquid Radwaste System Discharge Waste Condensate Tank A to the Circulating water System in accordance with OP-007-004, Liquid Waste Management System.

Fault: Upon initiation of flow, LWM flow controller output fails high, raising flow beyond what is permitted by the release permit. (2014 A,D,P,S 9 NRC Exam)

A4.03 Stoppage of releases if limits RO - 3.9, SRO - 3.8 exceeded 2 2017 NRC Revision 2

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

P1 076 Service Water System (ACCW)

Transfer EFW Pump Suctions to Wet Cooling Tower after Condensate Storage Pool Depletion using EOP OP-902-009, Standard Appendices, D,E,L,R 4S Attachment 10 (Top 10 PSA Action)

K1.20 AFW RO - 3.4, SRO - 3.4 P2 064 Electrical Diesel Generators Reset EDG A following an overspeed trip with a LOOP in accordance with OP-009-002, Emergency Diesel Generator, Section 8.8. D,E,L,R 6 EPE 055 EA1.06 Restoration of power with one ED/G RO - 4.1, SRO - 4.5 P3 006 Emergency Core Cooling System Isolate RWSP from Purification in accordance with OP-902-009, EOP Standard Appendices, Att. 40.

Alt. Path: FS-423, RWSP Suction Isolation is unable to be closed A,E,L,N,R 2 EPE 011 EK3.12 Actions contained in EOP for 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 different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 8 (E)mergency or abnormal in-plant 1/ 1/ 1 3 (EN)gineered safety feature 1 / 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) 2 (R)CA 1/ 1/ 1 3 (S)imulator 8 3 2017 NRC Revision 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: Mar 27, 2017 Exam Level RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1 001 Control Rod Drive System Place Reactor Cutback (RXC) in service and perform Immediate Operator Actions following a Cutback with unanalyzed rod configuration.

Alt. Path: Feedwater pump will trip resulting in a RXC. During the A,D,S 1 cutback, an incorrect CEA will drop.

GEN 2.4.49 The ability to perform without reference to procedures those actions that require immediate operation of system components and controls. RO - 4.6, SRO - 4.4 S2 006 Emergency Core Cooling System Reduce RCS pressure and use High Pressure Safety Injection Pumps to restore Pressurizer level in accordance with OP-901-112, Charging D,L,S 3 or Letdown Malfunction.

A1.18 PZR level and pressure RO - 4.0, SRO - 4.3 S3 003 Reactor Coolant Pump System Perform 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 of A,D,L,P,S 4P remaining Reactor Coolant Pumps.

A2.02 Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP RO - 3.7, SRO - 3.9 S4 061 Emergency Feedwater System Reset EFW Pump AB after Overspeed Trip in accordance with OP-009-003, Emergency Feedwater (Control Room actions) EN,L,N,S 4S GEN EPE 074 EA1.07 AFW System RO - 4.2, SRO - 4.3 S5 S6 064 Emergency Diesel Generator (ED/G) System Parallel Emergency Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator.

Alt. Path: After EDG A load is raised, EDG A load will rise without A,D,S 6 manipulation requiring a trip of EDG A.

A4.06 Manual start, loading, and stopping of the ED/G RO - 3.9, SRO - 3.9 S7 012 Reactor Protection System Reset High Containment Pressure ESFAS trip in accordance with OP-902-009, EOP Standard Appendices, Att. 5-D. EN,L,N,S 7 A4.04 Bistable, trips, reset and test switches RO - 3.3, SRO - 3.3 4 2017 NRC Revision 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S8 068 Liquid Radwaste System Discharge Waste Condensate Tank A to the Circulating water System in accordance with OP-007-004, Liquid Waste Management System.

Fault: Upon initiation of flow, LWM flow controller output fails high, raising flow beyond what is permitted by the release permit. (2014 A,D,P,S 9 NRC Exam)

A4.03 Stoppage of releases if limits RO - 3.9, SRO - 3.8 exceeded In-Plant Systems * (3 for RO; (3 for SRO-I); (3 or 2 for SRO-U)

P1 076 Service Water System (ACCW)

Transfer EFW Pump Suctions to Wet Cooling Tower after Condensate Storage Pool Depletion using EOP OP-902-009, Standard Appendices, D,E,L,R 4S Attachment 10 (Top 10 PSA Action)

K1.20 AFW RO - 3.4, SRO - 3.4 P2 064 Electrical Diesel Generators Reset EDG A following an overspeed trip with a LOOP in accordance with OP-009-002, Emergency Diesel Generator, Section 8.8. D,E,L,R 6 EPE 055 EA1.06 Restoration of power with one ED/G RO - 4.1, SRO - 4.5 P3 006 Emergency Core Cooling System Isolate RWSP from Purification in accordance with OP-902-009, EOP Standard Appendices, Att. 40.

Alt. Path: FS-423, RWSP Suction Isolation is unable to be closed A,E,L,N,R 2 EPE 011 EK3.12 Actions contained in EOP for 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 different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 7 (E)mergency or abnormal in-plant 1/ 1/ 1 3 (EN)gineered safety feature 1 / 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) 2 (R)CA 1/ 1/ 1 3 (S)imulator 7 5 2017 NRC Revision 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: Mar 27, 2017 Exam Level RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1 001 Control Rod Drive System Place Reactor Cutback (RXC) in service and perform Immediate Operator Actions following a Cutback with unanalyzed rod configuration.

Alt. Path: Feedwater pump will trip resulting in a RXC. During the A,D,S 1 cutback, an incorrect CEA will drop.

GEN 2.4.49 The ability to perform without reference to procedures those actions that require immediate operation of system components and controls. RO - 4.6, SRO - 4.4 S2 S3 S4 061 Emergency Feedwater System Reset EFW Pump AB after Overspeed Trip in accordance with OP-009-003, Emergency Feedwater (Control Room actions) EN,L,N,S 4S GEN EPE 074 EA1.07 AFW System RO - 4.2, SRO - 4.3 S5 S6 S7 012 Reactor Protection System Reset High Containment Pressure ESFAS trip in accordance with OP-902-009, EOP Standard Appendices, Att. 5-D. EN,L,N,S 7 A4.04 Bistable, trips, reset and test switches RO - 3.3, SRO - 3.3 S8 6 2017 NRC Revision 2

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

P1 P2 064 Electrical Diesel Generators Reset EDG A following an overspeed trip with a LOOP in accordance with OP-009-002, Emergency Diesel Generator, Section 8.8. D,E,L,R 6 EPE 055 EA1.06 Restoration of power with one ED/G RO - 4.1, SRO - 4.5 P3 006 Emergency Core Cooling System Isolate RWSP from Purification in accordance with OP-902-009, EOP Standard Appendices, Att. 40.

Alt. Path: FS-423, RWSP Suction Isolation is unable to be closed A,E,L,N,R 2 EPE 011 EK3.12 Actions contained in EOP for 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 different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 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 feature 1 / 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 7 2017 NRC Revision 2

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

Event Malf. Event Event No. No. Type* Description VCT level instrument CVC-ILT-0227 Fails high diverting letdown to the Boron Management system.

I - ATC OP-901-113, Volume Control Tank Makeup Control 1 CV12A1 I - SRO Malfunction I - BOP Safety Channel C RCS Cold Leg instrument RC-ITI-I - SRO 0102CC (Loop T112C) fails high requiring TS 3.3.1 2 RC19C TS - SRO entry and bypassing affected bistables.

C - ATC Charging Pump B trips on overcurrent requiring C - SRO implementation of OP-901-112, Charging or Letdown 3 CV01B TS - SRO malfunction. (TS 3.1.2.4)

Steam Generator 2 Level Control Transmitter, SG-ILT-1106, fails low requiring implementation of I - BOP OP-901-201, Steam Generator Level Control 4 SG05B I - SRO Malfunction and manual control of SG level.

Lowering Main Condenser vacuum requiring R- ATC implementation of OP-901-220, Loss of Condenser FW21A N-BOP Vacuum and a plant power reduction in accordance 5 FW21AA N-SRO with OP-901-212, Rapid Plant Power Reduction.

RC03C RP01A&B RCP 2A sustains a locked rotor and an automatic RP02A,B,C,&D reactor trip does not occur. Manual action (trip 32A DI-02A06S02-1 and 32B Supply Bkrs) is needed to trip the reactor (CT 6 DI-02A06S03-1 M - All 1, manually trip the reactor)

Main Steam Line Break outside Containment, SG 1, OP-902-004, Excess Steam Demand Recovery. (CT 7 MS13A M - All 2, stabilize RCS temperature and pressure)

I - ATC I - BOP Relay K202A fails, CVC-401, CVC-109, IA-909, and 8 RP08C I - SRO FP-601A fail to close automatically

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 2017 NRC Exam Scenario 1 D-1 Rev 4

Scenario Event Description NRC Scenario 1 The crew assumes the shift at 100% power with instructions to maintain 100% power. High Pressure Safety Injection (HPSI) pump A is out of service.

After taking the shift, Volume Control Tank (VCT) level instrument CVC-ILT-0227 fails high resulting in valve CVC-169 diverting letdown to the Boron Management System. The SRO should enter into procedure OP-901-113, Volume Control Tank Makeup Control Malfunction, and direct the ATC to place valve CVC-169 to the VCT position.

After the crew addresses the VCT instrument malfunction, RCS Cold Leg instrument RC-ITI-0102CC on CP-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 trips on 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 a suction 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 review and enter Technical Specification 3.1.2.4. Technical Specification 3.1.2.4 may be exited after aligning Charging Pump AB to replace Charging Pump B. The SRO may implement EN-OP-200, Transient Response 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 Generator Level Control Malfunction and implement Attachment 1, General Actions. Manual action by the BOP to control 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 Main Condenser vacuum begins to drop. The SRO will enter OP-901-220, Loss of Condenser Vacuum. Main Condenser vacuum will drop below 25 inches, requiring a rapid plant power reduction. The SRO will enter OP-901-212, Rapid Plant Power Reduction and should implement EN-OP-200, Transient Response Rules. Vacuum will drop below 25 inches but remain above 20 inches, the procedure trigger for tripping the Reactor. For the power reduction, the ATC will perform direct boration to the RCS as well as ASI control with CEAs and Pressurizer boron equalization. The BOP will manipulate the controls to reduce Main Turbine load.

After the reactivity manipulation is satisfied, Reactor Coolant Pump 2A rotor seizes and the RCP breaker trips. The Reactor Protection System fails to open the required Reactor Trip Breakers and an ATWS condition exists. The ATC should recognize that an automatic protection system has failed to occur and attempt to manually trip the reactor. Manual trip buttons on CP-2 (including Diverse Rx Trip) will not trip the reactor. The crew must manually trip 32A and 32B supply breakers to successfully trip the reactor (CRITICAL TASK 1). 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 steam demand event will occur on SG 1 outside containment upstream of the MSIV. The SRO will direct the ATC/BOP to initiate Safety Injection, Containment Isolation and Main Steam Isolation. The SRO will direct action to establish RCS Temperature and Pressure control using SG 2 (CRITICAL TASK 2) when CET temperature and PZR pressure begins to rise indicating a blown dry SG. Relay K202A will fail to actuate resulting in CVC-109, Letdown Outside Containment Isolation, CVC-401, Controlled Bleedoff Outside Containment Isolation, IA-909, Instrument Air Containment Isolation and FP-601A, Fire Water A Containment Isolation, valves not going to their required positions. The ATC and BOP will take action to close these valves. The crew should diagnose to OP-902-004, Excess Steam Demand Recovery and isolate Steam Generator 1.

The scenario can be terminated after the crew has isolated Steam Generator 1 or at the lead examiners discretion.

2017 NRC Exam Scenario 1 D-1 Rev 4

NRC Scenario 1 Critical Task Number Description Basis Establish Reactivity Control Failure to trip the reactor when an automatic PPS signal has failed to actuate can lead to This task is satisfied by manually degradation of fission product barriers. OPS tripping the reactor using the manual Management Expectation of 1 minute is 1 pushbuttons, Diverse Reactor Trip, or determined to be a reasonable time limit to de-energizing bus 32A and 32B within 1 identify and take action for satisfactory minute of RCP 2A tripping. This task performance.

becomes applicable following the RCP trip. (OP-902-000, 1.a.1) (TM-OP-100-03, CT-1)

Establish RCS Pressure and Temperature Control This task is satisfied by manually feeding An ESDE will result in a rapid cooldown and and steaming the unaffected Steam depressurization of the RCS. After the Steam Generator to stabilize RCS temperature Generator dries out, RCS temperature and and pressure prior to exiting the step to pressure will begin to recover. Operator action is required to stabilize RCS pressure and stabilize RCS temperature in OP-902-2 temperature to prevent a situation that may 004, Excess Steam Demand Recovery cause pressurized thermal shock which could and take action to achieve and maintain jeopardize the RCS integrity. A large D/P less than 1600 PSID across the affected across the Steam Generator tubes will make a Steam Generator. This task becomes subsequent SGTR more likely.

applicable once CET temperature and PZR pressure begins to rise following (TM-OP-100-03, CT-7) the ESDE. (OP-902-004, step 18 or OP-902-009, App. 13)

• 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 follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Scenario Quantitative Attributes

1. Malfunctions after EOP entry (1-2) 1

2. Abnormal events (2-4) 4

3. Major transients (1-2) 2

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

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

6. EOP based Critical tasks (2-3) 2 2017 NRC Exam Scenario 1 D-1 Rev 4

NRC Scenario 1 SCENARIO SETUP A. 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. Insert SIR29 (HPSI pump A breaker) to RKOUT (should be saved in IC)

2. 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.

2017 NRC Exam Scenario 1 D-1 Rev 4

NRC Scenario 1 SIMULATOR BOOTH INSTRUCTIONS Event 1 VCT level instrument, CVC-ILT-0227, Fails High

1. On Lead Examiner's cue, initiate Event Trigger 1. (Ensure the designed applicant is the ATC.)

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 Safety Channel C RCS Cold Leg Temp, RC-ITI-0112CC (RC-ITI-0102CC) fails high

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

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If sent to LCP-43, wait 3 minutes and report all Cold Leg temperatures on LCP-43 read approximately 545F.

Event 3 Charging Pump B Trip

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

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and 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 are tripped on all 3 phases for Charging Pump B.

4. If called as NAO to investigate the pump, wait 3 minutes and report that there are some indications 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 report the following for directed pump:

a. Suction and discharge valves are open

b. Proper oil level exists

c. Motor vents unobstructed

d. All personnel clear of the pump

6. If directed to check a started Charging pump for proper operation following start, wait 1 minute and report the following:

a. Suction and discharge valves are open

b. Proper oil pressure and seal water flow exist

c. No abnormal vibrations or noises present Event 4 Steam Generator 2 Level Control Transmitter, SG-ILT-1106, fails low

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

2017 NRC Exam Scenario 1 D-1 Rev 4

NRC Scenario 1 Event 5 Main Condenser Leak, Rapid Power Reduction

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

2. If called as TGB watch report all Air Evacuation Pumps look normal, Vacuum pump separators are 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 is preventing 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 Building to assist.

Event 6 RCP 2A Locked Rotor, Auto Reactor Trip & Manual Pushbuttons (Ch. A, B, & DRT) at CP-2 Fail

1. On Lead Examiner's cue, initiate Event Trigger 6.

2. No expected communications for this event.

Event 7 Main Steam Line Break outside Containment, SG 1

1. On Lead Examiner's cue, initiate Event Trigger 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 are operating 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 of steam is issuing from the west MSIV area.

Event 8 Relay K202A fails, CVC-401, CVC-109, IA-909, and FP-601A fail to close automatically

1. No communications should occur for this event.

At the end of the scenario, before resetting, end data collection and save the file as 2017 Scenario 1-(start-end time).tid. Export to .csv file. Save the file into the folder for the appropriate crew.

2017 NRC Exam Scenario 1 D-1 Rev 4

NRC Scenario 1 SCENARIO TIMELINE DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 1 CV12A1 VCT LEVEL XMTR CVC-ILIC-0227 FAILS HI 1 00:00:00 00:00:00 ACTIVE VCT LEVEL TRANSMITTER FAILS HIGH 2 RC19C RCS COLD LEG 1A SAFETY TT 0112C FAILS (0-100%) 2 00:00:00 00:00:00 100 SAFETY CHANNEL C RCS COLD LEG TEMPERATURE (Indicator RC-ITI-0102CC) 3 CV01B CHARGING PUMP B TRIPPED 3 00:00:00 00:00:00 ACTIVE CHARGING PUMP B TRIP 4 SG05B SG LEVEL ILT-1106 FAIL (0-100%) 4 00:00:00 00:00:10 0 STEAM GENERATOR 2 LEVEL CONTROL TRANSMITTER, SG-ILT-1106, FAILS LOW 5 FW21A CONDENSER A AIR INLEAK (100%=100% OF VAC BKR) 5 00:00:00 00:03:00 20 FW21AA CONDENSER A AIR INLEAK VACUUM SETPOINT 5 00:00:00 00:03:00 23.3 MAIN CONDENSER LEAK, RAPID POWER REDUCTION 6 RC03C RCP RC-MPMP-0002A SHAFT SEIZURE 6 00:00:00 N/A ACTIVE RP01A RPS MANUAL PUSHBUTTON CH A N/A 00:00:00 N/A ACTIVE RP01B RPS MANUAL PUSHBUTTON CH B N/A 00:00:00 N/A ACTIVE RP02A RPS CH A AUTO TRIP FAILURE N/A 00:00:00 N/A ACTIVE RP02B RPS CH B AUTO TRIP FAILURE N/A 00:00:00 N/A ACTIVE RP02C RPS CH C AUTO TRIP FAILURE N/A 00:00:00 N/A ACTIVE RP02D RPS CH D AUTO TRIP FAILURE N/A 00:00:00 N/A ACTIVE DI-02A06S02-1 DRT 1 OF 2 PB N/A 00:00:00 N/A OFF DI-02A06S03-1 DRT 2 OF 2 PB N/A 00:00:00 N/A OFF RCP 2A LOCKED ROTOR, AUTO REACTOR TRIP & MANUAL PUSHBUTTONS (CH. A, B, & DRT) AT CP-2 FAIL 7 MS13A MS A BREAK OUTSIDE CNTMT BEFORE MSIV (0-100%) 7 00:00:00 00:00:00 8%

MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT, SG 1 2017 NRC Exam Scenario 1 D-1 Rev 4

NRC Scenario 1 DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 8 RP08C RELAY K202 FAILED, CIAS TRAIN A (CVC/IA/FP) N/A 00:00:00 00:00:00 ACTIVE CVC-401, CVC-109, IA-909, AND FP-601A FAIL TO CLOSE AUTOMATICALLY N/A EGR26 EDG A LOCAL ANNUN ACK 20 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE N/A EGR27 EDG B LOCAL ANNUN ACK 21 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE N/A SIR29 HPSI PUMP A N/A 00:00:00 00:00:00 RKOUT HPSI PUMP A BREAKER 2017 NRC Exam Scenario 1 D-1 Rev 4

NRC Scenario 1 REFERENCES Event Procedures 1 OP-901-113, Volume Control Tank Makeup Control Malfunction, Rev. 302 2 OP-009-007, Plant Protection System, Rev. 17 OP-903-013, Monthly Channel Checks, Rev. 18 Technical Specification 3.3.1 3 OP-901-112, Charging or Letdown Malfunction, Rev. 6 OP-002-005, Chemical Volume Control, Rev. 56 Technical Specification 3.1.2.4 4 OP-901-201, Steam Generator Level Control Malfunction, Rev. 6 5 OP-901-220, Loss of Condenser Vacuum, Rev. 302 OP-002-005, Chemical Volume Control, Rev. 56 OP-004-004, Control Element Drive, Rev. 23 OP-901-212, Rapid Plant Power Reduction, Rev. 8 6 OP-902-000, Standard Post Trip Actions, Rev. 16 7 OP-902-004, Excess Steam Demand Recovery, Rev. 16 OP-902-009, Standard Appendices, Rev. 315, Appendix 2, Figures OP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart 8 OP-902-004, Excess Steam Demand Recovery, Rev. 16 GEN EN-OP-115, Conduct of Operations, Rev. 18 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 3 OI-038-000, EOP Operations Expectations / Guidance, Rev. 14 2017 NRC Exam Scenario 1 D-1 Rev 4

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

Event Malf. Event Event No. No. Type* Description N - BOP N - SRO Manually start EFW Pump A. EFW Pump A fails 1 FW06A TS - SRO during operability check. (TS 3.7.1.2)

Hot Leg 1 Temperature, RC-ITI-0111X, fails low affecting PZR level setpoint. OP-901-110, Pressurizer 2 RC21A I - All Level Control Malfunction (Sect. E2).

C - BOP Reactor Coolant Pump 2A Lower Seal fails.

3 RC08C C - SRO OP-901-130, Reactor Coolant Pump Malfunction.

Steam Generator 1 Pressure Instrument, SG-IPT-1013A, fails low requiring Technical I - BOP Specification entry and bypass of multiple Plant I - SRO Protection System A trip bistables. (TS 3.3.1, 3.3.2, &

4 SG04E TS - SRO 3.3.3.5)

Feedwater Heater 5B tube leak from Condensate to heater shell causing isolation of the Low Pressure R - ATC heater string. OP-901-221, Secondary System N - BOP Transient (Sect. E1) and OP-901-212, Rapid Plant 5 FW35B N - SRO Power Reduction.

Reactor Coolant Pump 2A Middle Seal fails, requiring C - ATC a manual reactor trip, and securing of Reactor Coolant 6 RC09C C - SRO Pump 2A.

Pressurizer Code Safety, RC-317A, fails open. OP-902-002, Loss of Coolant Accident Recovery. All Reactor Coolant Pumps must be secured. (CT 1, Trip 7 RC11A1 M - All RCPs exceeding operating limits) (CT3, Cooldown)

HPSI Pump B fails to AUTO start on the Safety C - BOP Injection Actuation Signal requiring a manual start.

8 SI02B C - SRO (CT 2, Inventory Control)

SI19A High Press Safety Injection (HPSI) Pump A degrades 9 SI01A N/A internally and trips.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 2017 NRC Exam Scenario 2 D-1 Rev 2

Scenario Event Description NRC Scenario 2 The crew assumes the shift at 100% power with instructions to maintain 100% power. Low Pressure Safety 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 trip on overcurrent shortly after it is started. The SRO should declare EFW pump A inoperable and enter Tech Spec 3.7.1.2.d.

After Tech Specs are addressed, Loop 1 Thot instrument, RC-ITI-0111X, fails low. This affects the Reactor Regulating System Tave calculation and the Pressurizer Level Setpoint. The SRO should enter OP-901-110, Pressurizer Level Control Malfunction and implement Section E2, Pressurizer Level Setpoint Malfunction. The crew should take manual control of Pressurizer Level, select the non-faulted Thot instrument (Loop 2) in both Reactor Regulating System cabinets, verify normal setpoint is restored and 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 crew should 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 Water Temperature, Steam Generator 1 Pressure Instrument, SG-IPT-1013A, fails low. The SRO should review and enter Technical Specifications 3.3.1 action 2, 3.3.2 actions 13 and 19 and 3.3.3.5 action a. The SRO will direct the BOP to bypass Steam Generator 1 Pressure Lo, Steam Generator 1 DP, and Steam Generator 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 />, in accordance with OP-009-007, Plant Protection System. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 using OP-903-013, Monthly Channel Checks, and determine that Technical Specification 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 leak occurs in Feedwater Heater 5B, causing Condensate flow to isolate through Low Pressure Feedwater Heaters 5B and 6B. The crew should enter OP-901-221, Secondary System Transient, and implement Section E1, Loss of Feedwater Preheating. This also requires a power reduction in accordance with OP-901-212, Rapid Plant Power Reduction which will prompt a reactivity manipulation. The SRO should implement EN-OP-200, Transient Response Rules.

After the reactivity manipulation is satisfied, Reactor Coolant Pump 2A Middle Seal fails. The crew should trip 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 crew should diagnose to OP-902-002, Loss of Coolant Accident Recovery. The crew should secure an additional 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 vibration or 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. High Pressure Safety Injection (HPSI) pump A will run for about three minutes, degrade internally and trip. The BOP should manually start High Pressure Safety Injection Pump B (CRITICAL TASK 2).

The scenario can be terminated after the crew starts a cooldown (CRITICAL TASK 3) or at the lead examiners discretion.

2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 Critical Task Number Description Basis Trip Any RCP Exceeding Operating Limits The time requirement of 3 minutes is based on This task is satisfied by stopping all the RCP operating limit of 3 minutes without running RCPs within 3 minutes of loss CCW cooling. Continued operation of RCP of Component Cooling Water flow or without CCW or outside of the operating limits 1 prior to completing the step that verifies could lead to a failure of the RCS pressure RCP operating limits. This task boundary at the RCP seal.

becomes applicable after either running RCP Vibration alarms actuate OR (TM-OP-100-03, CT-23; ECS98-001, D.10)

Containment Spray is initiated, whichever occurs first. (OP-902-002, LOCA, 9.b or 9.d.1)

Establish RCS Inventory Control Based on minimum required flow per the flow This task is satisfied by starting High delivery curve in OP-902-009, Appendix 2E.

Pressure Safety Injection Pump B to Failure to take action to establish the minimum establish Reactor Coolant System required Safety Injection flow during a LOCA 2 inventory control before exiting the step would degrade the inventory available to to verify Safety Injection Actuation Signal maintain the fuel covered. Adequate SI flow Actuation. This task becomes applicable ensures RCS Inventory Control and Core Heat following the initiation of a Safety Removal safety functions are satisfied.

Injection Actuation Signal. (OP-902-002, (TM-OP-100-03, CT-16; ECS98-001, A.02)

LOCA, step 7)

Commence RCS Cooldown Based on minimizing RCS leakage into This task is satisfied by opening at least containment. Cooling down the RCS will allow one Atmospheric Dump Valve before recovery of sub-cooled margin and subsequent depressurization. The magnitude of the break 3 exiting the step to perform Steam flow is lowered which allows termination of Generator cooldown. This task becomes Safety Injection flow and establishment of long applicable following the initiation of a term cooling by means of Shutdown Cooling.

Safety Injection Actuation Signal. (OP-902-002, LOCA, step 19) (TM-OP-100-03, CT-20)

• 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 follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 Scenario Quantitative Attributes

1. Malfunctions after EOP entry (1-2) 2

2. Abnormal events (2-4) 3

3. Major transients (1-2) 1

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

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

6. EOP based Critical tasks (2-3) 3 2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 SCENARIO SETUP A. 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. Insert SIR32 to RKOUT

2. Place C/S in OFF with a Danger Tag E. Place a copy of OP-903-046, EFW Operability Check, Section 7.1 on the BOP desk. Section 7.1 should 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/Ad. 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 state that 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. Brief Examiners to monitor applicant usage of the business computers (BOP, STA, SM, EC) to ensure that these computers are only used for eB Library, EOI Library and the Brief database.

L. Start Insight, open file Crew Performance.tis.

2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 SIMULATOR BOOTH INSTRUCTIONS Event 1 EFW Pump A trips on overcurrent during operability check

1. Approximately 1 minute after the crew starts EFW Pump A, initiate Event Trigger 1.

2. If Work Week Manager or PME are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If sent to 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 Low

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

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If sent to LCP-43, wait 2 minutes and report Thot loop 1 and 2 read approximately 605F.

Event 3 RCP 2A Lower Seal Fails

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

2. If the Duty Engineering or RCP Engineer is called inform the caller that you will monitor RCP 2A for further degradation.

3. If the Work Week Manager or PMM are called, inform the caller that a work package will be assembled for the next forced outage.

Event 4 Steam Generator Pressure Instrument, SG-IPT-1013A, Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If sent to LCP-43, wait 3 minutes and report that SG-IPI-1013-A1 reads 0 PSIA. Observe other indications of SG pressure using Extreme View on LCP-43 and report actual pressure if asked.

Event 5 Feedwater Heater 5B Tube Leak, Rapid Plant Power Reduction

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

2. If called to verify Low Pressure Heater levels, verify levels using the PMC and report levels to the Control Room.

3. If called to verify position of the Normal and Alternate Control Valves, verify valve positions using the 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 be assembled.

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 asked tell the caller your name is Joe Chemist.

2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 Event 6 RCP 2A Middle Seal Fails

1. After the reactivity manipulation is satisfied and on lead examiner's cue, initiate Event Trigger 6.

2. If the Duty Engineering or RCP Engineer is called inform the caller that you will monitor RCP 2A for further degradation.

3. If the Work Week Manager or PMM are called, inform the caller that a work package will be assembled.

Event 7 Pressurizer Code Safety, RC-317A, Fails Open

1. After the crew secures RCP 2A, initiate Event Trigger 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 2 minutes and manually initiate Event Trigger 20. Wait an additional minute and manually initiate Event Trigger 21 to acknowledge local EDG panels. Report that both A and B EDGs are running properly and unloaded.

Event 8 HPSI Pump B Fails To AUTO Start

1. External communications are not expected for this event.

Event 9 HPSI Pump B Fails To AUTO Start & HPSI Pump A Degrades & Trips

1. Event Trigger 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 be assembled 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 nothing abnormal.

At the end of the scenario, before resetting, end data collection and save the file as 2017 Scenario 2-(start-end time).tid. Export to .csv file. Save the file into the folder for the appropriate crew.

2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 SCENARIO TIMELINE DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 1 FW06A MOTOR DRIVEN EFW PMP A TRIP 1 00:00:00 00:00:00 ACTIVE MOTOR DRIVEN EFW PMP A TRIPS DURING OPERABILITY CHECK 2 RC21A RCS HOT LEG 1 CONTROL TT 111X FAILS (0-100%) 2 00:00:00 00:00:00 0%

HOT LEG 1 TEMPERATURE FAILS LOW 3 RC08C RCP 2A LOWER SEAL FAILURE (0-100%) 3 00:00:00 00:00:00 100%

RCP 2A LOWER SEAL FAILS 4 SG04E MS LINE IPT-1013A FAIL (0-100%) 4 00:00:00 00:00:00 0%

SG 1 PRESSURE INSTRUMENT SG-IPT-1013A FAILS LOW 5 FW35B LP 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 6 RC09C RCP 2A MIDDLE SEAL FAILURE (0-100%) 6 00:00:00 00:00:00 100%

RCP 2A MIDDLE SEAL FAILS 7 RC11A1 CODE SAFETY RC-317A FAIL OPEN 7 00:00:00 00:00:00 ACTIVE PRESSURIZER CODE SAFETY, RC-317A, FAILS OPEN 8 SI02B HPSI PUMP B FAILS TO AUTO START N/A 00:00:00 00:00:00 ACTIVE HPSI PUMP B FAILS TO AUTO START 9 SI19A HPSI PUMP A DEGRADATION (Triggered when PZR Press <1684 psia) 8 (AUTO) 00:02:00 00:01:00 100%

SI01A HPSI PUMP A TRIPPED (Triggered when PZR Press <1684 psia) 8 (AUTO) 00:03:00 00:00:00 ACTIVE HPSI PUMP A DEGRADES AND TRIPS N/A EGR26 EDG A LOCAL ANNUN ACK 20 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE N/A EGR27 EDG B LOCAL ANNUN ACK 21 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE 2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION N/A SIR32 LPSI PUMP A (will not show up in summary tab) N/A 00:00:00 00:00:00 RKOUT LPSI PUMP A BREAKER 2017 NRC Exam Scenario 2 D-1 Rev 2

NRC Scenario 2 REFERENCES Event Procedures 1 OP-903-046, Emergency Feed Pump Operability Check, Rev. 318 Technical Specification 3.7.1.2 2 OP-901-110, Pressurizer Level Control Malfunction, Rev. 9 OP-901-501, PMC or Core Operating Limits Supervisory System Malfunction, Rev. 15 3 OP-901-130, Reactor Coolant Pump Malfunction, Rev. 11 4 OP-009-007, Plant Protection System, Rev. 17 OP-903-013, Monthly Channel Checks, Rev. 18 Technical Specification 3.3.1 Technical Specification 3.3.2 Technical Specification 3.3.3.5 Technical Specification 3.3.3.6 5 OP-901-221, Secondary System Transient, Rev. 5 OP-901-212, Rapid Plant Power Reduction, Rev. 8 OP-002-005, Chemical and Volume Control, Rev. 56 OP-004-004, Control Element Drive, Rev. 23 6 OP-901-130, Reactor Coolant Pump Malfunction, Rev. 11 OP-902-000, Standard Post Trip Actions, Rev. 16 OP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart 7 OP-902-002, Loss of Coolant Accident Recovery Procedure, Rev. 20 OP-902-009, Standard Appendices, Rev. 315, Appendix 2, Figures OP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart 8&9 OP-902-002, Loss of Coolant Accident Recovery Procedure, Rev. 20 GEN EN-OP-115, Conduct of Operations, Rev. 18 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 3 OI-038-000, EOP Operations Expectations / Guidance, Rev. 14 2017 NRC Exam Scenario 2 D-1 Rev 2

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

Initial Conditions: ~ 1% Reactor Power; 1st 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.

Event Malf. Event Event No. No. Type* Description R - ATC Secure the Auxiliary Feedwater Pump and raise power N - BOP to 5-10% using CEAs in accordance with OP-010-003, 1 N/A N - SRO Plant Startup.

Pressurizer Level Controller, RC-ILIC-0110, fails off I - ATC requiring implementation of OP-901-110, Pressurizer 2 RX08A I - SRO Level Control Malfunction (E3).

I - BOP RWSP Level Instrument, SI-ILI-0305B, fails low and RP04B5 I - SRO generates an RAS trip requiring TS 3.3.2 entry and 3 AO-07A2M11-1 TS - SRO bypassing the affected trip bistable.

C - BOP Component Cooling Water Pump B trips requiring C - SRO entry into OP-901-510, Component Cooling Water 4 CC01B TS - SRO System Malfunction (TS 3.7.3 & Cascading).

Selected 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 a M - All manual reactor trip to secure selected Reactor RX14A C - ATC Coolant Pumps and stop RCS depressurization. (CT 5 RC14B1 C - SRO 1, RCS Pressure Control)

Loss of Off-site Power, OP-902-003, Loss of Offsite 6 ED01 A - D M - All Power/Loss of Forced Circulation Recovery 7 EG10A N/A Emergency Diesel Generator A trips on overspeed.

Emergency Diesel Generator B Output Breaker fails to AUTO Close, due to the 3B to 2B Tie Breaker failing to C - BOP open on Undervoltage. Crew re-energizes B Safety 8 ED23B C - SRO bus. (CT 2, Energize a Safety Electrical Bus)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 2017 NRC Exam Scenario 3 D-1 Rev 3

Scenario Event Description NRC Scenario 3 The crew assumes the shift at ~ 1% power with instructions to secure the AFW pump and raise power to 5-10% to roll the Main Turbine. All requirements have been met to change modes from MODE 2 to MODE 1. The SRO should direct raising power using Control Element Assemblies in accordance with the reactivity 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 level controller RC-ILIC-0100 fails off. The CRS should enter OP-901-110, Pressurizer Level Control Malfunction, and implement section E3. This will require the ATC to control Letdown from CP-4. There are no Tech Spec consequences of the failure provided the crew restores letdown flow prior to exceeding 62.5% 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 the annunciators for this failure. The CRS should evaluate Tech Specs and enter Tech Spec 3.3.2 and determine that the Plant Protection System bistable (18) for Low RWSP Level must be bypassed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> on 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. The SRO should enter OP-901-510, Component Cooling Water System Malfunction, and direct the start of Component Cooling Water Pump AB to replace Component Cooling Water Pump B. The SRO should enter Technical Specification 3.7.3 and cascading Technical Specifications per OP-100-014, Technical Specification 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 performing OP-903-066, Electrical Breaker Alignment Check. Once CCW pump AB is in service Tech Spec 3.7.3 and cascading Tech Specs may be exited.

After the SRO has addressed Technical Specifications, the selected Pressurizer Pressure Channel fails high causing pressurizer spray to initiate. When the crew takes manual control of the Pressurizer Spray Controller, 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 at least three Reactor Coolant Pumps to stop the RCS depressurization (Critical Task 1). The crew will be taking the actions required by OP-901-120, Pressurizer Control Malfunction but may not enter the procedure prior to the reactor trip due to the pressure dropping in the RCS. The SRO should enter OP-902-000, Standard Post Trip Actions. In order to restore Pressurizer heaters the SRO will have to implement the section in the offnormal to select the non-faulted pressurizer pressure control channel and the pressurizer level must 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 is performing Standard Post Trip Actions, a loss of off-site power occurs. Emergency Diesel Generator A will trip on overspeed. Emergency Diesel Generator B will start but, its output breaker will fail to close automatically due to the 3B to 2B Bus Tie Breaker failing to open on undervoltage. The crew must manually trip the 3B to 2B Bus Tie Breaker which allows EDG B output breaker to close automatically and re-energize the B Safety Bus (Critical Task 2). If the crew fails to manually trip the 3B to 2B Bus Tie Breaker, 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. The SRO should direct a non-licensed operator to restore power to the Dry Cooling Tower Sump Pumps. The BOP should take action to protect the Main Condenser from over-pressurization. The scenario can be ended after these actions are complete, or at the lead examiners discretion.

2017 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 Critical Task Number Description Basis Establish RCS Pressure Control RCS subcooling is an integral part of adequate pressure control, inventory control, and Core This task is satisfied by securing at least heat removal. The importance of keeping the three Reactor Coolant Pumps to stop fluid surrounding the Core in a subcooled state 1 Reactor Coolant System carries a high degree of nuclear safety depressurization prior to loss of significance based on its direct relationship to Subcooled Margin. This task becomes these safety functions.

applicable after Pressurizer Spray Valve B, RC-301B, fails open. (OP-901-120, (ECS-98-001, S.01)

E3 step 3)

Energize at Least One Safety Electrical Bus This task is satisfied by the crew taking action to energize the B Safety Bus by Failure to energize at least one emergency bus tripping the 3B-to-2B Bus Tie breaker will result in the plant remaining in a configuration that will not support protection if 2 prior to failure of Emergency Diesel a subsequent event would occur. This lowers Generator B due to no Component the capability of the plant to mitigate an event.

Cooling Water which occurs in approximately 20 minutes. This task (TM-OP-100-03, CT-3) becomes applicable after the loss of offsite power occurs. (OP-902-000, 3.a.1)

• 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 follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Scenario Quantitative Attributes

1. Malfunctions after EOP entry (1-2) 2

2. Abnormal events (2-4) 3

3. Major transients (1-2) 2

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

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

6. EOP based Critical tasks (2-3) 2 2017 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 SCENARIO NOTES A. 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 (secure AFW) circled and several of the previous steps circle-slashed to show progress. Fill in initials and circle-slash steps 9.4.53 (adjust Blowdown), 9.4.59 (mode 1 Tech Spec logs) and 9.4.60 (Chemistry contacted) 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.

2017 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 SIMULATOR BOOTH INSTRUCTIONS Event 1 Secure AFW Pump and raise reactor power

1. If called as an NAO to standby the AFW pump, acknowledge the communication. Wait 2 minutes and report you are standing by.

2. If called as Chemistry to verify SG chemistry is within specification, inform the caller that SG chemistry 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 5 minutes, report that you will be slowly opening/throttling MS-148, MS Supply to Gland Seal Isolation. Initiate Event Trigger 1. After MS-148 completes ramping, report that MS-148 is open/throttled open. If you are directed to further throttle open MS-148, simply acknowledge the request, wait ~30 seconds and report the new throttled position. Repeat as necessary until it is reported 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 been transferred to Main Steam (no remote necessary).

5. If called as an NAO to secure the Portable Auxiliary Boiler, acknowledge the communication. Wait 5 minutes, initiate Event Trigger 20 and report that the Portable Aux Boiler is secured..

Event 2 Pressurizer Level Controller, RC-ILIC-0110, Output Fails Off

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

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If called as an NAO, to report to the Letdown valve gallery, acknowledge report. Wait 3 minutes and report you are standing by. Use remotes CVR03 and CVR04 if directed to make Letdown valve alignments.

Event 3 RWSP Level Instrument, SI-ILI-0305B, Fails Low & RAS Trip Generated

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

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 4 Component Cooling Water Pump B Trips

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If called as the watchstander and sent to CCW Pump B, wait 3 minutes, report that the pump looks normal locally.

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

4. If Work Week Manager or PME are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 Pressurizer Pressure Control Channel, RC-IPT-0100X, Fails High and Pressurizer Spray Valve RC-301B Fails Open

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

2017 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 Event 6 Loss of Offsite Power

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

4. If called as an NAO to align power to DCT sump pumps (OP-902-009 App. 20), acknowledge report. Wait 3 minutes, run appropriate schedule file located in Remote Operator Actions (CAEP) and make report once schedule file has timed out.

Event 7&8 EDG A Trips on Overspeed; 3B-to-2B Bus Tie Breaker Fails to Trip on UV

1. If called as an NAO to investigate EDG A, wait 3 minutes; initiate Event Trigger 21 (EGR26) to acknowledge the local alarm panel and report that EDG A is not running, EMERGENCY STOP or UNIT S/D and ENGINE OVERSPEED alarms are locked in but there is no obvious signs of damage. 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 and sent to the field as soon as possible.

3. If called as an NAO to check EDG B, wait 2 minutes; initiate Event Trigger 22 (EGR27) to acknowledge the local alarm panel. If EDG B output breaker is closed and CCW pump B is running, 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 LOW FLOW alarm is locked in.

At the end of the scenario, before resetting, end data collection and save the file as 2017 Scenario 3-(start-end time).tid. Export to .csv file. Save the file into the folder for the appropriate crew.

2017 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 SCENARIO TIMELINE DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 1 MSR09 MS-148 MS to GS ISOL VALVE 1 00:00:00 00:01:00 12%

SECURE AFW PUMP AND RAISE REACTOR POWER 2 RX08A PZR LVL CONTROLLER 110 FAILS OFF 2 00:00:00 00:00:00 ACTIVE PRESSURIZER LEVEL CONTROLLER RC-ILIC-0110 OUTPUT FAILS OFF 3 RP04B5 TRIP GENERATED CH B RWSP LVL(RAS) 3 00:00:00 00:00:00 ACTIVE AO-07A2M11-1 CH B RWSP LEVEL 3 00:00:00 00:00:00 0%

RSWP CHANNEL B LEVEL INSTRUMENT SI-ILI-0305B FAILS LOW & RAS TRIP GENERATED 4 CC01B CCW PUMP B TRIP 4 00:00:00 00:00:00 ACTIVE COMPONENT COOLING WATER PUMP B TRIP 5 RX14A PZR PRESSURE CNTL CHL 100X FAIL (0-100%)(1500-2500 PSIA) 5 00:00:00 00:00:00 100 RC14B1 PZR SPRAY VALVE RC-301B FAILS OPEN 5 00:00:00 00:00:00 ACTIVE PRESSURIZER PRESSURE CONTROL CHANNEL, RC-IPT-0100X, FAILS HIGH AND PRESSURIZER SPRAY VALVE, RC-301B, FAILS OPEN 6 ED01A FEEDER BREAKER 7172 TRIP IN SWITCHYARD 6 00:00:00 00:00:00 ACTIVE ED01B FEEDER BREAKER 7176 TRIP IN SWITCHYARD 6 00:00:00 00:00:00 ACTIVE ED01C FEEDER BREAKER 7182 TRIP IN SWITCHYARD 6 00:00:00 00:00:00 ACTIVE ED01D FEEDER BREAKER 7186 TRIP IN SWITCHYARD 6 00:00:00 00:00:00 ACTIVE LOSS OF OFFSITE POWER 7 EG10A DG A OVERSPEED TRIP 6 00:00:10 00:00:00 ACTIVE EDG A TRIPS ON OVERSPEED 8 ED23B 3BS TO B2 BUS BREAKER FAILS TO TRIP ON UV 6 00:00:00 00:00:00 ACTIVE 3BS TO B2 BUS BREAKER FAILS TO TRIP ON UV 2017 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION N/A MSR32 TEMPORARY AUX BOILER 20 N/A N/A OFFLINE TEMPORARY AUX BOILER (16 MIN TILL RATED PRESS)

N/A EGR26 EDG A LOCAL ANNUN ACK 21 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE N/A EGR27 EDG B LOCAL ANNUN ACK 22 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE 2017 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 REFERENCES Event Procedures 1 OP-010-003, Plant Startup, Rev. 342 OP-003-035, Auxiliary Feedwater, Rev. 305 OP-004-004, Control Element Drive, Rev. 23 2 OP-901-110, Pressurizer Level Control Malfunction, Rev. 9 3 OP-009-007, Plant Protection System, Rev. 17 OP-903-013, Monthly Channel Checks, Rev. 18 Technical Specification 3.3.2 4 OP-901-510, Component Cooling Water Malfunction, Rev. 303 Technical Specification 3.7.3 & Cascading 5 OP-901-120, Pressurizer Pressure Control Malfunction, Rev. 302 OP-902-000, Standard Post Trip Actions, Rev 16 OP-902-009, Standard Appendices, Rev. 315, Appendix 1 (Diagnostic Flow Chart),

Appendix 2 (Figures) 6 OP-902-003, Loss of Offsite Power/Loss of Forced Circ Recovery Procedure, Rev. 10 7&8 OP-902-000, Standard Post Trip Actions, Rev. 16 GEN EN-OP-115, Conduct of Operations, Rev. 18 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 3 OP-100-014, TS and TRM Compliance, Rev. 337 OI-038-000, EOP Operations Expectations / Guidance, Rev. 14 2017 NRC Exam Scenario 3 D-1 Rev 3

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

Event Malf. Event Event No. No. Type* Description Condensate Storage Pool level instrument EFW-ILI-1 FW51A TS - SRO 9013A fails low. (TS 3.3.3.5, TS 3.3.3.6)

Letdown Flow Control Valve, CVC-113A, fails closed C - ATC requiring entry into OP-901-112, Charging or Letdown 2 CV30A2 C - SRO Malfunction.

Steam Generator #1 Feedwater flow instrument FW-I - BOP IFR-1111 fails low. OP-901-201, Steam Generator 3 FW26A I - SRO Level Control Malfunction.

R - ATC CEA 11 drops into the core requiring a rapid plant N - BOP down power in accordance with OP-901-212, Rapid N - SRO Plant Power Reduction. OP-901-102, CEA or 4 RD02A11 TS - SRO CEDMCS Malfunction. (TS 3.1.3.1)

RC23A C - ATC RCS Cold Leg leak; Charging Pump A fails to auto-5 CV02A C - SRO start.

The leak grows into a LOCA requiring implementation of OP-902-000, Standard Post Trip Actions and OP-902-002, Loss of Coolant Accident Recovery Procedure. Stop RCPs (CT 1, Trip RCPs exceeding 6 RC23A M - All operating limits).

CS-125B, Containment Spray Header B Isolation, fails to auto-open requiring manual action to open CS-C - BOP 125B (CT 2, Containment Temperature & Pressure 7 CS04B C - SRO control).

Main Steam Line 2 Break inside containment requiring entry into OP-902-008, Functional Recovery 8 MS11B M - All Procedure.

Containment Spray Pump A trips requiring action to C - BOP override close CS-125A, Containment Spray Header A 9 CS01A C - SRO Isolation (CT 3, Containment Isolation).

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 2017 NRC Exam Scenario 4 D-1 Rev 2

Scenario Event Description NRC Scenario 4 The crew assumes the shift at ~90% power with instructions to start Heater Drain Pump B and continue the power ascension after PMI resolves a problem with the Low Suction Pressure annunciator on Heater Drain Pump B. PMI will not resolve the annunciator problem and the crew will maintain ~90% power. No major equipment is out of service.

After the crew takes the shift, Condensate Storage Pool level indicator EFW-ILI-9013 A will fail low. The SRO 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, fails closed. The SRO should enter OP-901-112, Charging or Letdown Malfunction and implement Section E2, Letdown Malfunction, and place the backup flow control valve, CVC-113B, in-service. The SRO may implement EN-OP-200, Transient Response Rules.

After the backup letdown flow control valve has been placed in service, Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low. The Feedwater Control System will respond by increasing Feedwater flow to Steam Generator #1. The SRO should direct the BOP to take manual control and match Feedwater and Main Steam flow. The SRO should enter OP-901-201, Steam Generator Level Control Malfunction. Feedwater controls for Steam Generator #1 may remain in manual as a result of this failure requiring manual control on a plant down power or reactor trip. 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 T AVG to T REF initially and then perform a rapid plant downpower in accordance with OP-901-212, Rapid Plant Power Reduction. RCS direct boration must commence within 15 minutes of the dropped CEA to comply with Technical Specifications and the COLR. 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. The SRO 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 a medium break LOCA. Charging Pump A will fail to auto start requiring a manual start by the ATC. The SRO may enter OP-901-111, RCS System Leak, but will soon recognize that Pressurizer level is not being maintained with available Charging pumps and should direct a manual reactor trip and manual initiation 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 of a 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. Containment Spray Pump A will trip on overcurrent. The SRO should go to OP-902-009 Appendix 1, Diagnostics Flowchart and diagnose to OP-902-008, Functional Recovery OR go directly to OP-902-008 based on two events in progress per OP-100-017, Emergency Operating Procedures Implementation Guide. When the SRO performs prioritization Containment Isolation (CI-1) should be the highest priority. The SRO should direct the BOP to override and close CS-125A, Containment Spray Header A Isolation (CRITICAL TASK 3).

The scenario can be terminated once the crew closes Containment Spray Header A Isolation in accordance with OP-902-008, Functional Recovery procedure or at the lead examiners discretion.

2017 NRC Exam Scenario 4 D-1 Rev 2

NRC Scenario 4 Critical Task Number Description Basis Trip Any RCP Exceeding Operating Limits The time requirement of 3 minutes is based on the This task is satisfied by stopping all running RCP operating limit of 3 minutes without CCW RCPs within 3 minutes of loss of Component cooling. Continued operation of RCP without CCW 1 Cooling Water flow or prior to completing the or outside of the operating limits could lead to a step that verifies RCP operating limits. This failure of the RCS pressure boundary at the RCP task becomes applicable after either running seal.

RCP Vibration alarms actuate OR (TM-OP-100-03, CT-23; ECS98-001, D.10)

Containment Spray is initiated, whichever occurs first. (OP-902-002, 9.b or 9.d.1)

Establish Containment Temperature and Pressure Control The maximum design pressure of the containment This task is satisfied by manually opening structure is 44 psig. Failure to take action to CS-125B, Containment Spray Header B establish containment pressure and temperature Isolation, prior to exceeding containment control may result in containment pressure exceeding maximum design and therefore exceed design pressure of 44 PSIG or prior to design leakage of containment. The operators 2 completing Containment Spray (CS) monitor containment pressure along with verification in OP-902-002 or exiting the Containment Spray and Containment Fan Cooler Containment Temperature and Pressure operations as verification of adequate containment Control Safety Function in OP-902-008. This heat removal and pressure mitigation.

task becomes applicable after CS is initiated and is critical after CS Pump A trips. (OP- (TM-OP-100-03, CT-15; ECS98-001, P.28) 902-002, step 14 or OP-902-008, CTPC-2)

Establish Containment Isolation A Loss of Coolant Accident that has occurred inside This task is satisfied by closing CS-125A, containment and has not been isolated will result in Containment Spray Header A Isolation, prior excess radioactivity leaving containment and being 3 to exiting the Containment Isolation (CI-1) released to the public.

Safety Function in OP-902-008. This task becomes applicable after Containment (TM-OP-100-03, CT-9)

Spray (CS) is initiated and CS Pump A trips.

(OP-902-008, CI-1, 1.c.1)

• 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 follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Scenario Quantitative Attributes

1. Malfunctions after EOP entry (1-2) 2

2. Abnormal events (2-4) 3

3. Major transients (1-2) 2

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

5. EOP contingencies requiring substantive actions (0-2) 1

6. EOP based Critical tasks (2-3) 3 2017 NRC Exam Scenario 4 D-1 Rev 2

NRC Scenario 4 SCENARIO SETUP A. Reset Simulator to IC-164 B. 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 annunciator F0802, 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. Have event 2 schedule file ready.

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.

2017 NRC Exam Scenario 4 D-1 Rev 2

NRC Scenario 4 SIMULATOR BOOTH INSTRUCTIONS Event 1 Condensate Storage Pool level instrument EFW-ILI-9013 A fails low

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

2. If called as an NAO to check the indication at the Remote Shutdown Panel, wait 2 minutes and report that Condensate Storage Pool Level instrument EFW-ILI-9013 A1 is reading 0% and EFW-ILI-9013B1 is approximately 98%.

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

Event 2 Letdown Flow Control Valve, CVC-113A, Fails Closed

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

2. If Work Week Manager or PMM are called, inform the caller that a work package will be assembled 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 in Service.sch. Make appropriate reports as the schedule file progresses.

Event 3 Steam Generator #1 Feedwater Flow Instrument FW-IFR-1111 Fails Low

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

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If called as the TGB watch to investigate a Polisher System Trouble alarm, wait 2 minutes, acknowledge using FWR109 and report resin trap high D/P alarm came in and cleared.

Event 4 CEA 11 Falls into the core/Rapid plant power reduction

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and 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. If asked, 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 asked tell the caller your name is Joe Chemist.

5. If notified as Load Dispatcher (Woodlands) acknowledge the communications and inform the caller 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 start

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

2017 NRC Exam Scenario 4 D-1 Rev 2

NRC Scenario 4 Event 6 RCS Cold Leg Break

1. 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 2 minutes and manually initiate Event Trigger 20. Wait an additional minute and manually initiate Event Trigger 21 to acknowledge local EDG panels. Report that both A and B EDGs are running properly and unloaded.

Event 7 CS-125B Fails to Open Automatically

1. There is no event trigger for this event.

2. External communications are not expected.

Event 8 Main Steam Line 2 Break inside Containment

1. On Lead Examiner's cue, initiate Event Trigger 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 request Event 9 Containment Spray Pump A trips / Override CS-125A, CS Header A Isolation

1. 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 pick up the key. Have someone role play as NAO and enter the simulator to simulate getting the key.

Wait 1 minute, insert remote CSR13A using Event Trigger 22. Make the report to the Control Room that you have done so.

3. If called as an NAO to investigate the trip of CS pump A breaker, wait 2 minutes and report overcurrent flags on all 3 phases.

4. If called as an NAO to investigate CS Pump A, wait 3 minutes and report that the paint on the motor is discolored and there 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 2017 Scenario 4-(start-end time).tid. Export to .csv file. Save the file into the folder for the appropriate crew.

2017 NRC Exam Scenario 4 D-1 Rev 2

NRC Scenario 4 SCENARIO TIMELINE DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 1 FW51A FAIL 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 2 CV30A2 LTDN FLOW CONTROL VALVE CVC-113A FAILS CLOSED 2 00:00:00 00:00:00 ACTIVE LETDOWN FLOW CONTROL VALVE, CVC-113A, FAILS CLOSED 3 FW26A FW 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%

4 RD02A11 DROPPED CEA 11 4 00:00:00 00:00:00 ACTIVE CEA 11 DROPS INTO THE CORE; RAPID PLANT POWER REDUCTION 5 RC23A RCS COLD LEG 1A RUPTURE 5 00:00:00 00:08:00 1.5 CV02A CHARGING PUMP A FAIL TO AUTOSTART N/A 00:00:00 00:00:00 ACTIVE RCS COLD LEG LEAK / CHARGING PUMP A FAILS TO AUTO START 6 RC23A RCS COLD LEG 1A RUPTURE 5 00:08:00 00:00:00 14 RCS COLD LEG BREAK 7 CS04B CS TRAIN B CS-125B FAILS TO AUTO OPEN N/A 00:00:00 00:00:00 ACTIVE CS-125B FAILS TO AUTO OPEN 8 MS11B MS 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 CS01A LOSS OF CNTMT SPRAY PUMP A 8 00:00:00 00:00:00 ACTIVE LOSS OF CNTMT SPRAY PUMP A N/A EGR26 EDG A LOCAL ANNUN ACK 20 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE N/A EGR27 EDG B LOCAL ANNUN ACK 21 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE 2017 NRC Exam Scenario 4 D-1 Rev 2

NRC Scenario 4 DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION N/A CSR13A CS-125A REMOTE KEY SW TO CLOSE VALVE 22 00:00:00 00:00:00 OVRD CS-125A REMOTE KEY SW TO CLOSE VALVE N/A F_Q12 HTR DRAIN PUMP B SUCTION PRESS LO N/A 00:00:00 00:00:00 FAIL ON ANNUNCIATOR MALFUNCTION OVERRIDE 2017 NRC Exam Scenario 4 D-1 Rev 2

NRC Scenario 4 REFERENCES Event Procedures 1 OP-903-013, Monthly Channel Checks, Rev. 18 Tech Spec 3.3.3.5 Tech Spec 3.3.3.6 2 OP-901-112, Charging or Letdown Malfunction, Rev. 6 3 OP-901-201, Steam Generator Level Control Malfunction, Rev. 6 4 OP-901-102, CEA or CEDMCS Malfunction, Rev. 304 OP-901-212, Rapid Plant Power Reduction, Rev. 8 OP-901-501, PMC or COLSS Malfunction, Rev. 15 OP-004-004, Control Element Drive, Rev. 23 Tech Spec 3.1.3.1 5 OP-902-000, Standard Post Trip Actions, Rev. 16 OP-902-009, Standard Appendices, Rev. 315, Appendix 2, Figures OP-902-009, Standard Appendices, Rev. 315, Appendix 1, Diagnostic Flow Chart 6&7 OP-902-002, Loss of Coolant Accident Recovery, Rev. 20 OP-902-009, Standard Appendices, Rev. 315, Appendix 2, Figures 8&9 OP-902-008, Functional Recovery, Rev. 26 OP-902-009, Standard Appendices, Rev. 315, Appendix 21-A GEN EN-OP-115, Conduct of Operations, Rev. 18 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 3 OI-038-000, EOP Operations Expectations / Guidance, Rev. 14 OP-100-017, Emergency Operating Procedures Implementation Guide, Rev. 5 2017 NRC Exam Scenario 4 D-1 Rev 2