Text

7_W3-2018-11 Final Outlines
	ML18333A346
Person / Time
Site:	Waterford
Issue date:	11/14/2018
From:	Vincent Gaddy; Operations Branch IV
To:	; Entergy Operations
References
	Download: ML18333A346 (64)
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ES-401 PWR Examination Outline (RO)

Form ES-401-2 Rev 1 Facility:

Waterford 3 Date of Exam: November 14, 2018 Tier Group RO K/A Category Points SRO-Only Points K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

Total A2 G*

Total

1.

Emergency &

Abnormal Plant Evolutions 1

2 3

3 4

4 2

18 2

1 2

2 2

1 1

9 Tier Totals 3

5 5

6 5

3 27

2.

Plant Systems 1

3 3

2 3

2 2

3 2

3 3

2 28 2

1 1

0 1

1 1

1 2

0 10 Tier Totals 4

4 2

4 3

3 4

5 2

38

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 3

2 2

3 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 These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000007 (EPE 7; BW E02&E10; CE E02)

Reactor Trip, Stabilization, Recovery / 1 X

EK1 Knowledge of the operational implications of the following concepts as they apply to the (Reactor Trip Recovery)

EK1.2 Normal, abnormal and emergency operating procedures associated with (Reactor Trip Recovery) 3.0 1

000008 (APE 8) Pressurizer Vapor Space Accident / 3 X

AK2 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

AK2.02 Sensors and detectors 2.7*

2 000009 (EPE 9) Small Break LOCA / 3 X

EK3 Knowledge of the reasons for the following responses as the apply to the small break LOCA:

EK3.11 Dangers associated with inadequate core cooling 4.4 3

000011 (EPE 11) Large Break LOCA / 3 X

EA1 Ability to operate and monitor the following as they apply to a Large Break LOCA:

EA1.09 Core flood tank initiation 4.3 4

000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 Not Sampled 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X

AA2 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup:

AA2.01 Whether charging line leak exists (New K/A) 3.2 5

000025 (APE 25) Loss of Residual Heat Removal System / 4 X 2.1.20 Ability to interpret and execute procedure steps.

4.6 6

000026 (APE 26) Loss of Component Cooling Water / 8 X

AA2 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water:

AA2.02 The cause of possible CCW loss 2.9 7

000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X

AA1 Ability to operate and / or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions:

AA1.03 Pressure control when on a steam bubble 3.6 8

000029 (EPE 29) Anticipated Transient Without Scram / 1 X

EK3 Knowledge of the reasons for the following responses as the apply to the ATWS:

EK3.11 Initiating emergency boration 4.2 9

000038 (EPE 38) Steam Generator Tube Rupture / 3 X

EK1 Knowledge of the operational implications of the following concepts as they apply to the SGTR:

EK1.01 Use of steam tables 3.1 10 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 X

EK2 Knowledge of the interrelations between the (Excess Steam Demand) and the following:

EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

3.3 11

ES-401 3

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

EK2 Knowledge of the interrelations between the (Loss of Feedwater) and the following:

EK2.2 Facility*s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

3.5 12 000055 (EPE 55) Station Blackout / 6 X

EK3 Knowledge of the reasons for the following responses as the apply to the Station Blackout:

EK3.02 Actions contained in EOP for loss of offsite and onsite power 4.3 13 000056 (APE 56) Loss of Offsite Power / 6 X

AA1 Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power:

AA1.37 Instrument Air (New K/A) 3.4 14 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X

AA2 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

AA2.19 The plant automatic actions that will occur on the loss of vital ac electrical instrument bus (New KA) 4.0 15 000058 (APE 58) Loss of DC Power / 6 X

2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

4.5 16 000062 (APE 62) Loss of Nuclear Service Water / 4 X

AA2 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:

AA2.04 The normal values and upper limits for the temperatures of the components cooled by the SWS. (New K/A) 2.5 17 000065 (APE 65) Loss of Instrument Air / 8 X

AA1 Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air:

AA1.05 RPS 3.3*

18 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 Not Sampled (W E04) LOCA Outside Containment / 3 Not Applicable (W E11) Loss of Emergency Coolant Recirculation / 4 Not Applicable (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 Not Applicable K/A Category Totals:

2 3 3 4 4 2 Group Point Total:

18

ES-401 4

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 X

AA2 Ability to determine and interpret the following as they apply to the Dropped Control Rod:

AA2.03 Dropped rod, using in-core/ex-core instrumentation, in-core or loop temperature measurements 3.6 20 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 X

AA1 Ability to operate and / or monitor the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

AA1.02 Level trip bypass 3.0 21 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 X 2.4.11 Knowledge of abnormal condition procedures. (New K/A) 4.0 19 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X

AK3 Knowledge of the reasons for the following responses as they apply to the Accidental Liquid Radwaste Release:

AK3.04 Actions contained in EOP for accidental liquid radioactive-waste release 3.8 22 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms / 7 000067 (APE 67) Plant Fire On Site / 8 X

AK1 Knowledge of the operational implications of the following concepts as they apply to Plant Fire on Site:

AK1.02 Fire fighting 3.1 23 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 X

AK2 Knowledge of the interrelations between the Loss of Containment Integrity and the following:

AK2.03 Personnel access hatch and emergency access hatch 2.8*

24 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling / 4

ES-401 5

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000076 (APE 76) High Reactor Coolant Activity/9 X

AK2 Knowledge of the interrelations between the High Reactor Coolant Activity and the following:

AK2.01 Process radiation monitors 2.6 25 000078 (APE 78*) RCS Leak / 3 Not Applicable?

(W E01 & E02) Rediagnosis & SI Termination/ 3 Not Applicable (W E13) Steam Generator Overpressure / 4 Not Applicable (W E15) Containment Flooding / 5 Not Applicable (W E16) High Containment Radiation /9 Not Applicable (BW A01) Plant Runback / 1 Not Applicable (BW A02 & A03) Loss of NNI-X/Y / 7 Not Applicable (BW A04) Turbine Trip / 4 Not Applicable (BW A05) Emergency Diesel Actuation / 6 Not Applicable (BW A07) Flooding / 8 Not Applicable (BW E03) Inadequate Subcooling Margin / 4 Not Applicable (BW E08; W E03) LOCA Cooldown Depressurization / 4 Not Applicable (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 X

AK3 Knowledge of the reasons for the following responses as they apply to the (Natural Circulation Operations)

AK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.

3.4 26 (BW E13 & E14) EOP Rules and Enclosures Not Applicable (CE A11**; W E08) RCS Overcooling Pressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery X

EA1 Ability to operate and / or monitor the following as they apply to the (Functional Recovery)

EA1.3 Desired operating results during abnormal and emergency situations.

3.6 27 (CE E13*) Loss of Forced Circulation /

LOOP/Blackout / 4 Not Applicable?

K/A Category Point Totals:

1 2 2 2 1 1 Group Point Total:

9

ES-401 6

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump X

K1 Knowledge of the physical connections and/or cause-effect relationships between the RCPS and the following systems:

K1.12 CCWS 3.0 28 004 (SF1; SF2 CVCS) Chemical and Volume Control X

X K2 Knowledge of bus power supplies to the following:

K2.01 Boric acid makeup pumps K6 Knowledge of the effect of a loss or malfunction on the following CVCS components:

K6.09 Purpose of VCT divert valve (New K/A) 2.9 2.8 29 30 005 (SF4P RHR) Residual Heat Removal X

X K3 Knowledge of the effect that a loss or malfunction of the RHRS will have on the following:

K3.05 ECCS 2.1.28 - Knowledge of the purpose and function of major system components and controls.

3.7*

4.1 31 32 006 (SF2; SF3 ECCS) Emergency Core Cooling X

K4 Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following:

K4.11 Reset of SIS 3.9 33 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X

K5 Knowledge of the operational implications of the following concepts as the apply to PRTS:

K5.02 Method of forming a steam bubble in the PZR 3.1 34 008 (SF8 CCW) Component Cooling Water X

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

A1.01 CCW flow rate A4 Ability to manually operate and/or monitor in the control room:

A4.08 CCW pump control switch 2.8 3.1*

35 36 010 (SF3 PZR PCS) Pressurizer Pressure Control X

K6 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:

K6.01 Pressure detection Systems (New K/A) 2.7 37 012 (SF7 RPS) Reactor Protection X

X K5 Knowledge of the operational implications of the following concepts as the apply to the RPS:

K5.01 DNB A2 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Faulty bistable operation (New K/A) 3.3*

3.1 38 39

ES-401 7

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 013 (SF2 ESFAS) Engineered Safety Features Actuation X

A3 Ability to monitor automatic operation of the ESFAS including:

A3.02 Operation of actuated equipment 4.1 40 022 (SF5 CCS) Containment Cooling X

A4 Ability to manually operate and/or monitor in the control room:

A4.03 Dampers in the CCS 3.2*

41 025 (SF5 ICE) Ice Condenser Not Applicable 026 (SF5 CSS) Containment Spray X

X A3 Ability to monitor automatic operation of the CSS, including:

A3.02 Verification that cooling water is supplied to the containment spray heat exchanger 2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of operation. (New K/A) 3.9*

4.3 42 43 039 (SF4S MSS) Main and Reheat Steam X

K1 Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems:

K1.02 Atmospheric relief dump valves 3.3 44 059 (SF4S MFW) Main Feedwater X

K3 Knowledge of the effect that a loss or malfunction of the MFW will have on the following:

K3.04 RCS 3.6 45 061 (SF4S AFW)

Auxiliary/Emergency Feedwater X

X K2 Knowledge of bus power supplies to the following:

K2.02 AFW electric drive pump (New K/A)

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.07 Air or MOV failure 3.7*

3.4 46 47 062 (SF6 ED AC) AC Electrical Distribution X

X K4 Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following:

K4.02 Circuit breaker automatic trips A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:

A1.01 Significance of D/G load limits 2.5 3.4 48 49 063 (SF6 ED DC) DC Electrical Distribution X

K2 Knowledge of bus power supplies to the following:

K2.01 Major DC loads 2.9*

50

ES-401 8

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 064 (SF6 EDG) Emergency Diesel Generator X

A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including:

A1.08 Maintaining minimum load on ED/G (to prevent reverse power) 3.1 51 073 (SF7 PRM) Process Radiation Monitoring X

K1 Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems:

K1.01 Those systems served by PRMs 3.6 52 076 (SF4S SW) Service Water X

K4 Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following:

K4.06 Service water train separation 2.8 53 078 (SF8 IAS) Instrument Air X

A3 Ability to monitor automatic operation of the IAS, including:

A3.01 Air pressure 3.1 54 103 (SF5 CNT) Containment X

A4 Ability to manually operate and/or monitor in the control room:

A4.09 Containment vacuum system (New K/A) 3.1 55 053 (SF1; SF4P ICS*) Integrated Control Not Applicable?

K/A Category Point Totals:

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

28

ES-401 9

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication X

K1 Knowledge of the physical connections and/or cause-effect relationships between the RPIS and the following systems:

K1.01 CRDS 3.2*

56 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (New K/A)

A2.01 Detector failure 3.0*

60 017 (SF7 ITM) In-Core Temperature Monitor X

K5 Knowledge of the operational implications of the following concepts as they apply to the ITM system:

K5.01 Temperature at which cladding and fuel melt (New K/A) 3.1 57 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X

A4 Ability to operate and/or monitor in the control room: HRPS controls A4.01 HRPS controls 4.0 58 029 (SF8 CPS) Containment Purge Not Applicable?

033 (SF8 SFPCS) Spent Fuel Pool Cooling X

K4 Knowledge of design feature(s) and/or interlock(s) which provide for the following:

K4.03 Anti-siphon devices 2.6 59 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator X

K6 Knowledge of the effect of a loss or malfunction on the following will have on the S/GS:

K6.01 MSIVs (New K/A) 3.2 61 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X

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

A1.02 Steam pressure 3.1 62

ES-401 10 Form ES-401-2 Rev 1 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 2 (RO)

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 045 (SF 4S MTG) Main Turbine Generator X

A3 Ability to monitor automatic operation of the MT/G system, including:

A3.05 Electrohydraulic control 2.6 63 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring X

A4 Ability to manually operate and/or monitor in the control room:

A4.02 Major Components (New K/A) 2.5 65 075 (SF8 CW) Circulating Water X

K2 Knowledge of bus power supplies to the following:

K2.03 Emergency/essential SWS pumps 2.6*

64 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation Not Applicable?

K/A Category Point Totals:

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

10

ES-401 Generic Knowledge and Abilities Outline (Tier 3 - RO)

Form ES-401-3 Rev 1 Facility:

Waterford 3 Date of Exam:

November 14, 2018 Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements.

3.8 66 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

4.3 67 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. (New K/A) 3.3 68 Subtotal 3

2.

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

3.9 69 2.2.38 Knowledge of conditions and limitations in the facility license.

3.6 70 Subtotal 2

3.

Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

3.2 71 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

3.2 72 Subtotal 2

4.

Emergency Procedures /

Plan 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

3.8 73 2.4.14 Knowledge of general guidelines for EOP usage.

3.8 74 2.4.32 Knowledge of operator response to loss of all annunciators.

3.6 75 Subtotal 3

Tier 3 Point Total 10

ES-401 Record of Rejected K/As Form ES-401-4 Tier /

Group Randomly Selected K/A Reason for Rejection 1/1 Q# 5 022 AA2.01 022 A2.03 rejected. W3 is not equipped with Charging line flow control valves or controllers in its Reactor Coolant Makeup (Charging) system. Randomly selected another K/A from 022 AA2.

1/1 Q# 14 056 AA1.37 056 AA1.33 rejected. W3 is not equipped with a PORV valve.

Randomly selected another K/A from 056 AA1.

1/1 Q# 15 057 AA2.19 057 AA2.06 rejected. Following a review of W3 annunciators for a loss of vital bus and auto swap to the alternate power supply, determined that an RO knowledge question with plausible distractors could not be constructed. Randomly selected another K/A from 057 AA2 1/1 Q# 17 062 AA2.04 062 AA2.03 rejected. There is no procedural guidance in W3 procedures for performing valve lineups to restart the ACCW system while bypassing portion of the system causing the abnormal condition. Randomly selected another K/A from 062 A2.

1/2 Q# 19 037 2.4.11 001 2.4.6 rejected. The actions for continuous rod motion are not located in a W3 EOP. Actions are located in a W3 AOP. Could not develop an RO question that would not conflict with the continuous rod withdrawal question presented on the 2015 NRC RO Exam Q#

19.

2/1 Q# 30 004 K6.09 004 K6.13 rejected. This K/A matched an event in 2018 NRC Exam simulator scenario #4. (inadvertent dilution due to a failed PMU batch controller). To eliminate overlap, randomly selected another K/A from 004 K6.

2/1 Q# 37 010 K6.01 010 K6.03 rejected. W3 has no guidance for a simultaneous loss or malfunction of both the PZR spray and heaters. The spray and heaters control systems are independent of one another. Randomly selected another K/A for 010 K6.

2/1 Q# 39 012 A2.01 012 A2.04 rejected. W3 does not have guidance for erratic power supply to the RPS (only for a loss of AC or DC power). Randomly selected another K/A for 012 A2.

2/1 Q# 43 026 2.1.23 026 2.1.25 rejected. Could not identify a table, graph or curve pertaining to Containment Spray in a normal, abnormal or emergency operating procedure. Randomly selected another K/A for Containment Spray in the 2.1 section of generic K/As.

2/1 Q# 46 061 K2.02 061 K2.03 rejected. Waterford 3 does not have an AFW diesel driven pump. Randomly selected another K/A for 061 K2.

2/1 Q# 55 103 A4.09 103 A4.04 rejected. W3 is not equipped with containment isolation phase A and phase B resets. Randomly selected another K/A for 103 A4.

2/2 Q# 57 017 K5.01 017 K5.02 rejected. Could not develop a Question that would not duplicate a question on the 2017 or 2018 RO Exam (i.e. determine saturated or superheated using a steam table). Randomly selected another K/A for 017 K5.

ES-401 Record of Rejected K/As Form ES-401-4 2/2 Q# 58 028 A4.01 028 K3.01 rejected. Waterford 3 has no guidance on the loss of a Hydrogen Recombiner nor its effect on hydrogen concentration in containment. The hydrogen recombiners have been removed from TS/TRM and Emergency Operating procedures at W3. Randomly selected another K/A for the Hydrogen Recombiner system.

2/2 Q# 60 016 A2.01 034 A2.03 rejected. Could not develop a question that the RO could predict the impact on the Fuel Handling System due to a fuel mispostion. Randomly selected another Tier 2/Group 2 system for A2.

2/2 Q#61 035 K6.01 035 K6.02 rejected. Could not create a question on a failure of an ADV that did not duplicate a question on neither the 2015, 2017 or 2018 NRC Exam. Randomly selected another K6 K/A for 035 Steam Generators.

2/2 Q# 65 072 A4.02 079 A4.01 rejected. A question on Station Air Crossover valve (SA-125) would create the fourth question on this exam for IA/SA set-points. This would be an oversampling of the IA system and directly conflicted with RO question 54. Randomly selected another Tier 2/

Group 2 system for A4.

3 Q# 68 2.1.4 2.1.39 rejected. The only question that could be developed using W3 references that involved conservative decision practices is when an RO has the authority to trip the reactor. There is a question on this subject in the previous two exams (2015 RO Exam Q# 66).

The allotted 4 questions from the previous two exams have already been used.

ES-401 PWR Examination Outline (SRO)

Form ES-401-2 Rev 4 Facility:

Waterford 3 Date of Exam: November 14, 2018 Tier Group RO K/A Category Points SRO-Only Points K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

Total A2 G*

Total

1.

Emergency &

Abnormal Plant Evolutions 1

N/A N/A 18 3

3 6

2 9

2 2

4 Tier Totals 27 5

5 10

2.

Plant Systems 1

28 2

2 4

2 10 1

2 1

4 Tier Totals 38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

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 These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2

Form ES-401-2 Rev 4 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000007 (EPE 7; BW E02&E10; CE E02)

Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 EA2 Ability to determine and interpret the following as they apply to the Large Break LOCA:

EA2.01 Actions to be taken based on RCS temperature and pressure-saturated and superheated (new K/A) 4.7 80 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X

AA2 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions:

AA2.01 Cause of RCP failure (new K/A) 3.0 76 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X

2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

4.4 77 000029 (EPE 29) Anticipated Transient Without Scram / 1 000038 (EPE 38) Steam Generator Tube Rupture / 3 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

EA2. Ability to determine and interpret the following as they apply to the (Loss of Feedwater)

EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

4.2 78 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 X 2.4.18 Knowledge of the specific bases for EOPs.

4.0 79 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 000062 Loss of Nuclear Service Water/4 X

ES-401 3

Form ES-401-2 Rev 4 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X

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

4.1 81 (W E04) LOCA Outside Containment / 3 Not Applicable (W E11) Loss of Emergency Coolant Recirculation / 4 Not Applicable (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 Not Applicable K/A Category Totals:

3 3 Group Point Total:

6

ES-401 4

Form ES-401-2 Rev 4 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 (SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod /

1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 X

AA2 Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions:

AA2.11 Leak in PZR 3.6 83 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak /

3 X

2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (new K/A) 4.2 82 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms / 7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling / 4 000076 (APE 76) High Reactor Coolant Activity/9 000078 (APE 78*) RCS Leak / 3 Not Applicable (W E01 & E02) Rediagnosis & SI Termination/ 3 Not Applicable (W E13) Steam Generator Overpressure / 4 Not Applicable (W E15) Containment Flooding / 5 Not Applicable (W E16) High Containment Radiation /9 Not Applicable

ES-401 5

Form ES-401-2 Rev 4 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 (SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR (BW A01) Plant Runback / 1 Not Applicable (BW A02 & A03) Loss of NNI-X/Y / 7 Not Applicable (BW A04) Turbine Trip / 4 Not Applicable (BW A05) Emergency Diesel Actuation / 6 Not Applicable (BW A07) Flooding / 8 Not Applicable (BW E03) Inadequate Subcooling Margin / 4 Not Applicable (BW E08; W E03) LOCA Cooldown Depressurization / 4 Not Applicable (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures Not Applicable (CE A11**; W E08) RCS Overcooling Pressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 X 2.2.40 Ability to apply Technical Specifications for a system.

4.7 84 (CE E09) Functional Recovery EA2 Ability to determine and interpret the following as they apply to the (Functional Recovery)

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

4.4 85 (CE E13*) Loss of Forced Circulation /

LOOP/Blackout / 4 Not Applicable K/A Category Point Totals:

0 0 0 0 2 2 Group Point Total:

4

ES-401 6

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.13 Inadvertent SIS actuation 4.2 86 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection X

2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

4.7 87 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser Not Applicable 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)

Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.12 Restoration of power to a system with a fault on it 3.6 88

ES-401 7

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air X 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

4.2 89 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control Not Applicable K/A Category Point Totals:

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

4

ES-401 8

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04 Effects on axial flux density of control rod alignment and sequencing, xenon production and decay, and boron vs. control rod reactivity changes 3.8 90 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment X

A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fuel Handling System controls including:

A1.02 Water level in the refueling canal (new K/A) 3.7 91 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator X 2.1.9 Ability to direct personnel activities inside the control room. (new K/A) 4.5 93 055 (SF4S CARS) Condenser Air Removal

ES-401 9

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 056 (SF4S CDS) Condensate X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04 Loss of condensate pumps 2.8 92 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation Applicable (Facility Licensee has list of K/As)

K/A Category Point Totals:

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

4

ES-401 Generic Knowledge and Abilities Outline (Tier 3 - SRO)

Form ES-401-3 Rev 4 Facility:

Waterford 3 Date of Exam:

November 14, 2018 Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.2 Knowledge of operator responsibilities during all modes of plant operation.

4.4 94 2.1.36 Knowledge of procedures and limitations involved in core alterations.

4.1 95 Subtotal 2

2.

Equipment Control 2.2.7 Knowledge of the process for conducting special or infrequent tests.

3.6 96 2.2.37 Ability to determine operability and/or availability of safety related equipment.

4.6 97 Subtotal 2

3.

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

3.8 98 Subtotal 1

4.

Emergency Procedures /

Plan 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

4.4 99 2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations 4.4 100 Subtotal 2

Tier 3 Point Total 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier /

Group Randomly Selected K/A Reason for Rejection 1/1 SRO 76 015/017 AA2.01 008 AA2.12 rejected. Could not develop an SRO question pertaining to a Vapor Space Accident. W3 does not specify between a vapor space accident and LOCA within OP-902-002. Randomly selected another system in Tier 1/Group 1 for K/A AA2.

1/1 SRO80 011 EA2.01 062 AA2.02 rejected. Could not develop an A2 SRO level question on possible cause of nuclear service water (ACCW) loss. The possible causes for loss of ACCW is pump trip or ACC-126 failure and these would conflict with the previous two exams. Randomly selected another APE/EPE from Tier 1/Group 1 for K/A EA2.

1/2 SRO 82 037 2.2.25 005 2.2.25 rejected. Nothing in the basis for an inoperable control rod met a level of importance that could be developed into an SRO question. Could not develop an SRO question for a stuck rod without including a question on a less than one hour action (RO level). Randomly selected another APE/EPE from Tier 1/Group 2 while retaining 2.2.25.

2/2 SRO 91 034 A1.02 034 A1.01 The load limit setpoints for the refueling machines could not be used because this would be minutia and is not a number that SRO are expected to remember. Randomly selected another K/A from 034 FHS for A1.

2/2 SRO 93 045 2.1.9 086 2.1.31 rejected. Could not develop an SRO question on Fire Protection that did not involve fire brigade manning. This type of question would have overlapped a 2017 SRO question. Could not make an SRO question on the ability to locate control room switches and indications. This K/A will always fall in the realm of an RO question. Randomly selected a system from Tier 2/Group 2.

Randomly selected a K/A from section 2.1 of the Generic portion of the catalog.

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Waterford 3 Date of Examination:

11/5/2018 Examination Level: RO SRO Operating Test Number:

1 Administrative Topic (see Note)

Type Code*

Describe activity to be performed A1 Conduct of Operations K/A Importance: 4.6 D,R 2.1.18 Ability to interpret and execute procedure steps.

Complete a calculation for determining the amount of pure water that may be added to the Refuel Cavity without dilution to below shutdown margin requirements in accordance with OP-010-006, Outage Operations, section 9.24, Refueling Cavity Boron Concentration.

A2 Conduct of Operations K/A Importance: 4.3 D,R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Calculate required and available Feedwater to place the plant on Shutdown Cooling and determine the maximum time requirement to be on Shutdown Cooling.

A3 Equipment Control K/A Importance: 3.9 M,R 2.2.14, Knowledge of the process for controlling equipment configuration or status.

Determine the appropriate CEA subgroup selections by performing OP-004-015, Reactor Power Cutback, 1.1, Manual CEA Subgroup Selection.

A4 Radiation Control K/A Importance: 3.2 P,M,R 2.3.4, Knowledge of radiation exposure limits under normal and emergency conditions.

Calculate stay time to perform a tagout verification. Room dose rate and operators yearly dose provided (Modified From 2015 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 and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect IURPEDQNIRU52VIRU652VDQG52UHWDNHV

1HZRU0RGLILHGIURPEDQN

3UHYLRXVH[DPVUDQGRPO\\VHOHFWHG

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Waterford 3 Date of Examination:

11/5/2018 Examination Level: RO SRO Operating Test Number:

1 Administrative Topic (see Note)

Type Code*

Describe activity to be performed A5 Conduct of Operations K/A Importance: 4.2 N,R 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, etc.

Direct actions to take for plant personnel during a Toxic Chemical Emergency in accordance with EP-004-010, Toxic Chemical Contingency procedure.

A6 Conduct of Operations K/A Importance: 4.4 D,R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Review and approve a completed calculation for required and available Feedwater to place the plant on Shutdown Cooling and review recommendation for the maximum time requirement to be on Shutdown Cooling.

A7 Equipment Control K/A Importance: 4.7 D,R 2.2.40, Ability to apply Technical Specifications for a System.

Review and approve calculation for determining leak rate of RC Loop 2 Hot Leg Injection Inside Containment Check, SI-510B.

A8 Radiation Control K/A Importance: 4.3 P,D,R 2.3.11, Ability to Control Radiation Releases.

Terminate release after review of OP-903-001, Technical Specification Surveillance Logs, Attachment 11.6, Containment Purge Cumulative Hours Calculation.

(From 2015 NRC Exam)

A9 Emergency Plan K/A Importance: 4.6 N,R 2.4.41, Knowledge of the emergency action level thresholds and classifications.

Determine appropriate Emergency Plan action level in accordance with EP-001-001, Recognition and Classification of Emergency Conditions.

NOTE: All items (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 and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom

'LUHFWIURPEDQNIRU52VIRU652VDQG52UHWDNHV

1HZRU0RGLILHGIURPEDQN

3UHYLRXVH[DPVUDQGRPO\\VHOHFWHG

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function S1 004 Chemical and Volume Control Secure Emergency Boration in accordance with OP-901-103, Emergency Boration.

A4.07 Boration/Dilution RO-3.9, SRO-3.7 May be Performed with S2 D, L, S 1

S2 005 Shutdown Cooling System Secure Shutdown Cooling Train B in accordance with OP-009-005, Shutdown Cooling.

A4.01 Controls and for RHR Pumps RO - 3.6, SRO - 3.4 May be Performed with S1 D,L, S 4P S3 076 Service Water System Perform a manual start of the Auxiliary Component Cooling Water System in accordance with OP-002-001, Auxiliary Component Cooling Water.

A4.01 SWS Pumps RO - 2.9, SRO - 2.9 May be Performed with S4 D, S 4S S4 068 Liquid Radwaste System (2017 NRC Exam)

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.

A4.03 Stoppage of releases if limits exceeded RO-3.9, SRO-3.8 May be Performed with S3 A, D, P, S 9

S5 006 Emergency Core Cooling System Establish simultaneous hot and cold leg injection in accordance with OP-902-002, Loss of Coolant Accident Recovery A4.07 ECCS Pumps and valves RO-4.4, SRO 4.4 D, L, EN, S 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function S6 026 Containment Spray System Respond to an inadvertent CSAS in accordance with OP-901-504, Inadvertent ESFAS actuation. CCW flow cannot be restored due to CC-710 unable to be re-opened. Reactor Coolant Pumps must be secured within 3 minutes of losing CCW flow.

A4.01 RO-4.5, SRO-4.3 A, N, EN, S 5

S7 064 Emergency Diesel Generators (2015 NRC Exam)

Restore Power to Safety Bus 3B in accordance with OP-902-000, Standard Post Trip Actions.

Fault: EDG B voltage is low out of the band preventing auto closure of the EDG B output breaker. (WF3 OE)

A4.02 Adjustment of exciter voltage (using voltage control switch)

RO - 3.3, SRO - 3.4 A, D, P,S 6

S8 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 013 Engineered Safety Features Actuation System (ESFAS)

Actuate a Recirculation Actuation Signal (RAS) manually in accordance with OP-902-009 Appendix 34, RAS Manual Actuation. The RAS will not actuate using the manual pushbuttons and will actuate by opening the breakers.

A4.03 ESFAS Initiation RO-4.5, SRO-4.7 A, E, L, N 2

P2 006 Emergency Core Cooling System (2017 NRC Exam)

Isolate RWSP from Purification in accordance with OP-902-009, EOP Standard Appendices, App. 40.

Alt Path: FS-423, RWSP Suction Isolation is unable to be closed.

EPE 011 EK3.12 Actions contained in the EOP for emergency LOCA (large break)

RO-4.4, SRO-4.6 A, D, E, L, P, R

3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function P3 041 Steam Dump System and Turbine Bypass Control Operate Atmospheric Dump Valve A locally in accordance with OP-902-009, Attachment 22-A, Local Pneumatic Operation of ADVs.

A4.06 Atmospheric relief valve controllers RO-2.9, SRO-3.16 D, E, L 4S 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, and in-plant systems and functions may overlap those tested in the control room.

Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 5 0

8

3

FRQWUROURRPV\\VWHP 2

6

2

UDQGRPO\\VHOHFWHG 3

1 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function S1 004 Chemical and Volume Control Secure Emergency Boration in accordance with OP-901-103, Emergency Boration.

A4.07 Boration/Dilution RO-3.9, SRO-3.7 May be Performed with S2 D, L, S 1

S2 005 Shutdown Cooling System Secure Shutdown Cooling Train B in accordance with OP-009-005, Shutdown Cooling.

A4.01 Controls and for RHR Pumps RO - 3.6, SRO - 3.4 May be Performed with S1 D,L, S 4P S3 076 Service Water System Perform a manual start of the Auxiliary Component Cooling Water System in accordance with OP-002-001, Auxiliary Component Cooling Water.

A4.01 SWS Pumps RO - 2.9, SRO - 2.9 May be Performed with S4 D, S 4S S4 068 Liquid Radwaste System (2017 NRC Exam)

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.

A4.03 Stoppage of releases if limits exceeded RO-3.9, SRO-3.8 May be Performed with S3 A, D, P, S 9

S5 006 Emergency Core Cooling System Establish simultaneous hot and cold leg injection in accordance with OP-902-002, Loss of Coolant Accident Recovery A4.07 ECCS Pumps and valves RO-4.4, SRO 4.4 D, L, EN, S 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function S6 026 Containment Spray System Respond to an inadvertent CSAS in accordance with OP-901-504, Inadvertent ESFAS actuation. CCW flow cannot be restored due to CC-710 unable to be re-opened. Reactor Coolant Pumps must be secured within 3 minutes of losing CCW flow.

A4.01 RO-4.5, SRO-4.3 A, N, EN, S 5

S7 064 Emergency Diesel Generators (2015 Exam)

Restore Power to Safety Bus 3B in accordance with OP-902-000, Standard Post Trip Actions.

Fault: EDG B voltage is low out of the band preventing auto closure of the EDG B output breaker. (WF3 OE)

A4.02 Adjustment of exciter voltage (using voltage control switch)

RO - 3.3, SRO - 3.4 A, D, P,S 6

S8 008 Component Cooling Water System Transfer the 3AB Bus to the A train and Start Component Cooling Water Pump AB as the second CCW Pump.in accordance with OP-901-311, Loss of Train B Safety Bus.

A4.01 CCW indications and Controls RO-3.3, SRO-3.1 D, S 8

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 013 Engineered Safety Features Actuation System (ESFAS)

Actuate a Recirculation Actuation Signal (RAS) manually in accordance with OP-902-009 Appendix 34, RAS Manual Actuation. The RAS will not actuate using the manual pushbuttons and will actuate by opening the breakers.

A4.03 ESFAS Initiation RO-4.5, SRO-4.7 A, E, L, N 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function P2 006 Emergency Core Cooling System (2017 NRC Exam)

Isolate RWSP from Purification in accordance with OP-902-009, EOP Standard Appendices, App. 40.

Alt Path: FS-423, RWSP Suction Isolation is unable to be closed.

EPE 011 EK3.12 Actions contained in the EOP for emergency LOCA (large break)

RO-4.4, SRO-4.6 A, D, E, L, P, R

3 P3 041 Steam Dump System and Turbine Bypass Control Operate Atmospheric Dump Valve A locally in accordance with OP-902-009, Attachment 22-A, Local Pneumatic Operation of ADVs.

A4.06 Atmospheric relief valve controllers RO-2.9, SRO-3.16 D, E, L 4S 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, and in-plant systems and functions may overlap those tested in the control room.

Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 5 0

9

3

FRQWUROURRPV\\VWHP 2

6

2

UDQGRPO\\VHOHFWHG 3

1 8

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function S1 004 Chemical and Volume Control (2014 Exam)

Secure Emergency Boration in accordance with OP-901-103, Emergency Boration.

A4.07 Boration/Dilution RO-3.9, SRO-3.7 May be Performed with S2 D, L, S 1

S2 005 Shutdown Cooling System (2017 Audit Exam)

Secure Shutdown Cooling Train B in accordance with OP-009-005, Shutdown Cooling.

A4.01 Controls and for RHR Pumps RO - 3.6, SRO - 3.4 May be Performed with S1 D,L, S 4P S3 S4 S5 S6 026 Containment Spray System Respond to an inadvertent CSAS in accordance with OP-901-504, Inadvertent ESFAS actuation. CCW flow cannot be restored due to CC-710 unable to be re-opened. Reactor Coolant Pumps must be secured within 3 minutes of losing CCW flow.

A4.01 RO-4.5, SRO-4.3 A, N, EN, S 5

S7 S8 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Waterford 3 Date of Examination:

11/5/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function P1 013 Engineered Safety Features Actuation System (ESFAS)

Actuate a Recirculation Actuation Signal (RAS) manually in accordance with OP-902-009 Appendix 34, RAS Manual Actuation. The RAS will not actuate using the manual pushbuttons and will actuate by opening the breakers.

A4.03 ESFAS Initiation RO-4.5, SRO-4.7 A, E, L, N 2

P2 006 Emergency Core Cooling System (2017 Exam)

Isolate RWSP from Purification in accordance with OP-902-009, EOP Standard Appendices, App. 40.

Alt Path: FS-423, RWSP Suction Isolation is unable to be closed.

EPE 011 EK3.12 Actions contained in the EOP for emergency LOCA (large break)

RO-4.4, SRO-4.6 A, D, E, L, P, R

3 P3 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, and in-plant systems and functions may overlap those tested in the control room.

Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 3 0

3

2

FRQWUROURRPV\\VWHP 1

4

2

UDQGRPO\\VHOHFWHG 1

1 3

Appendix D Scenario Outline Form ES-D-1 2018 NRC Exam Scenario 1 D-1 Rev 3 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%. EDG A (TS 3.8.1.1b and d) and LPSI Pump A (TS 3.5.2a) are out of service.

Critical Tasks:

(1) Establish Required Safety Injection Flow (2) Isolate Most Affected Steam Generator (3) Prevent Lifting Affected Steam Generator Safety Valves Event No.

Malf.

No.

Event Type*

Event Description 1

SG11C I-BOP I-SRO TS-SRO Steam Generator #2 Narrow Range level Safety Channel C fails low (SG-ILT-1123C).

2 CV01B C-ATC C-SRO TS-SRO Charging Pump B Trip.

3 RX14A I-ATC I-SRO PZR pressure control channel failure low (RC-IPIC-0100X).

4 SG01B C-ATC C-BOP C-SRO TS-SRO Steam Generator #2 Tube Leak (~5 gpm).

5 DI-01A07A2S01-1 C-BOP C-SRO During power reduction, the main turbine will revert to manual control.

6 SG01B M-All Steam Generator Tube Rupture requiring a Reactor Trip and initiation of a Safety Injection. (Critical Task 2, Isolate most affected Steam Generator. Critical Task 3, Prevent Lifting Affected Steam Generator Safety Valves 7

SI01A N/A HPSI Pump A trips upon start.

8 SI02B C-BOP C-SRO HPSI Pump B fails to autostart. (Critical Task 1, Establish Required Safety Injection Flow)

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

Appendix D Scenario Outline Form ES-D-1 2018 NRC Exam Scenario 1 D-1 Rev 3 Scenario Quantitative Attributes

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

2. Abnormal events (2-4) [Events 2,3,4 credited]

3

3. Major transients (1-2) 1

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

5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 0

6. Preidentified critical tasks (> 2) 3

NRC Scenario 1 - Narrative 2018 NRC Exam Scenario 1 D-1 Rev 3 The crew assumes the shift at 100% power with instructions to maintain 100% power. Emergency Diesel Generator A and Low Pressure Safety Injection Pump A is out of service.

Event 1: After taking the shift, Steam Generator #2 Narrow Range level channel SG-ILT-1123C will fail low on Safety Channel C. The crew will enter TS 3.3.1 action 2 (Functional Unit 8 - Steam Generator Level - Low), TS 3.3.2 action 19 (Functional Units 7b. Steam Generator Level - Low and 3+LJKDQG

7c. Steam Generator Level - Low and No S/G Pressure - Low Trip), TRM 3.3.1 action 1 (Functional Unit 1 - Steam Generator Level - High) and TRM 3.3.2 and bypass bistables 8, 10, and 20 on CP-10 for Channel C in accordance with OP-009-007, Plant protection System.

Event 2: After Technical Specifications are addressed and Channel C bistables are bypassed, Charging Pump B trips on overcurrent. The SRO will implement OP-901-112, Charging or Letdown Malfunction, Section E1, 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 and comply with the actions of TS 3.8.1.1.d. 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.

Event 3: After Technical Specifications have been addressed, Pressurizer Pressure Control Channel RC-IPIC-0100X will fail low. The crew will observe pressurizer pressure alarms and that all PZR heaters are energized. The SRO will implement OP-901-120, Pressurizer Pressure Control Malfunction, Section E1, Pressurizer Pressure Control Channel Instrument Failure. The SRO should direct the ATC to align the alternate pressurizer pressure channel and verify correct Pressurizer pressure control response.

Event 4: After the crew has established control over the Pressurizer Pressure Control System, a Steam Generator Tube leak (~5 gpm) will occur on #2 Steam Generator. The SRO will enter OP-901-202, Steam Generator Tube Leakage or High Activity and attempt to quantify the leakage. Once leakage is determined to exceed 75 gallons per day, the SRO will enter OP-901-212, Rapid Plant Power Reduction and commence a plant shutdown. The SRO should enter Technical Specification 3.4.5.2 condition c, action a to be in HSB in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and CSD in 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The SRO should enter EN-OP-200, Transient Response Rules. 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.

Event 5: During the power reduction and turbine MW is equal to 1180 MW, the Main Turbine controls will shift from automatic to manual mode. The SRO will direct the BOP to maintain the turbine load reduction by controlling the turbine in manual mode in accordance with OP-901-212, Rapid Plant Power Reduction or OP-005-007, Main Turbine and Generator. The BOP will be manually changing governor valve position instead of controlling at a certain megawatt per minute rate.

Event 6: Once the BOP has established control of the turbine load reduction in manual, the Steam Generator Tube Leak will get worse and will turn into a Steam Generator Tube Rupture. The ATC will note that Pressurizer level is lowering and all available Charging pumps (A and AB) will be operating.

When it is determined that the Charging system cannot maintain RCS inventory, the SRO should direct the ATC to perform a manual Reactor Trip and to initiate a manual Safety Injection (SIAS) and Containment Isolation (CIAS).

Event 7/8: After the crew initiates SIAS, the A HPSI pump will degrade and trip upon starting and the B High Pressure Safety Injection (HPSI) Pump will fail to autostart. The BOP should manually start the B HPSI Pump (CRITICAL TASK 1). The SRO will enter OP-902-000, Standard Post Trip Actions and diagnose to OP-902-007, Steam Generator Tube Rupture Recovery.

The SRO will enter OP-902-007 and will perform a rapid plant cooldown to less than 520°F and isolate Steam Generator #2 (CRITICAL TASK 2). The crew should commence a depressurization of the RCS in an attempt to minimize the differential pressure and RCS leakage across the SG to prevent lifting a Main Steam Safety Valve or Atmospheric Dump Valve in automatic (CRITICAL TASK 3).

NRC Scenario 1 - Narrative 2018 NRC Exam Scenario 1 D-1 Rev 3 The scenario can be terminated after the crew has isolated Steam Generator 2 and commenced an RCS depressurization or at the lead examiners discretion.

NRC Scenario 1 - Critical Task Determination 2018 NRC Exam Scenario 1 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: Establish Required Safety Injection Flow This task is satisfied by starting the B High Pressure Safety Injection Pump such that Safety Injection flow is greater than or equal to the minimum required per the flow delivery curve in OP-902-009, Appendix 2-E prior to exiting the step (step 7) to establish adequate HPSI flow in OP-902-007.

This task becomes applicable following the Safety Injection Actuation Signal.

(OP-902-007, 7.a.)

Based on Emergency Operating Procedure Required flow. Failure to take action to maintain the core covered demonstrates a significant performance deficiency regarding the maintenance of the RCS inventory Control Safety Function and jeopardizes the Fuel clad fission product barrier. OPS management Standard documented in TM-OP-100-

03.

(TM-OP-100-03, CT-16)

HPSI Flow on CP-8 less than required HPSI Pump running light extinguished Procedurally driven from OP-902-007 step 7 Starts HPSI Pump B using control switch Adequate HPSI Flow on CP-8 indicated HPSI Pump running light illuminated CT-2: Isolate Most Affected Steam Generator This task is satisfied by manually closing Main Steam Isolation valve #2, Main Feedwater Isolation Valve #2, and MS-401B after RCS THOT has reduced below 520°F and prior to exiting the step (Step 17) to isolate the most affected Steam generator in OP-902-007, Steam Generator Tube Rupture Recovery. This task becomes applicable once the SGTR has commenced. (OP-902-007, 17)

RCS THOT < 520°F is based on Emergency Operating Procedure standard. Isolating the SG at a temperature higher than this value may complicate the response later in the scenario by not allowing a full depressurization of the RCS to below the lowest Main Steam Safety Valve setpoint. This higher RCS pressure will allow the RCS to continue to flow into the affected SG ultimately raising the SG level and pressure to the safety valve setpoint and cause a release that could have been prevented. OPS management Standard documented in TM-OP-100-

03.

(TM-OP-100-03, CT-14)

Procedurally driven from OP-902-007 step 17 Closes MS-124B (MSIV

2) using control switch Closes FW-184B (MFIV

2) using control switch Closes MS-401B using control switch Closed valve indication for MS-124B, FW-184B, &

MS-401B

NRC Scenario 1 - Critical Task Determination 2018 NRC Exam Scenario 1 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-3: Prevent Lifting Affected SG Safety Valves This task is satisfied by performing a cooldown to THOT less than 520°F using the Steam Bypass Control System prior to closing the affected MSIV and commencing RCS depressurization towards <930 psia prior to lifting an Atmospheric Dump Valve or Main Steam Safety Valve on the affected SG.

This task becomes applicable once the SGTR has commenced. The ADV set point is 990 psig and the MSSV set point is 1070 psig. (OP-902-007, 11, 12)

RCS THOT < 520°F and RCS pressure

< 930 psia are based on Emergency Operating Procedure Standard. RCS temperature must be reduced below the temperature corresponding to the saturation pressure for the ADV/MSSV set point prior to closing the affected SG MSIV. RCS must be depressurized to below the lift setpoint of the MSSVs to prevent overfilling the SG and lifting the MSSVs. Depressurization does not necessarily need to be completed, but must be commenced to receive credit for the task. OPS management Standard documented in TM-OP-100-

03.

(TM-OP-100-03, CT-21)

RCS THOT > 520°F RCS pressure > 930 psia Procedurally driven from OP-902-007 step 11 and step 12 Opens Steam Bypass Control valve Opens either Main Spray valve(s) or Auxiliary Spray valve(s)

RCS THOT < 520°F RCS pressure lowering Critical Task (NUREG-1021, Rev. 11 Appendix D) x 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.

x Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.

NRC Scenario 1 2018 NRC Exam Scenario 1 D-1 Rev 3 REFERENCES Event Procedures 1

OP-009-007, Plant Protection System, Rev. 19 Technical Specification 3.3.1, 3.3.2 Technical Requirements Manual 3.3.1 2

OP-901-112, Charging or Letdown Malfunction, Rev. 6 Technical Specification 3.1.2.4 Technical Specification 3.8.1.1 3

OP-901-120, Pressurizer Pressure Control Malfunction, Rev. 302 4

OP-901-202, Steam Generator Tube Leakage or High Activity, Rev. 15 OP-901-212, Rapid Plant Power Reduction, Rev. 9 Technical Specification 3.4.5.2 5

OP-901-212, Rapid Plant Power Reduction, Rev. 9 OP-005-007, Main Turbine and Generator, Rev. 306 6

OP-901-202, Steam Generator Tube Leakage or High Activity, Rev. 15 OP-902-000, Standard Post Trip Actions, Rev. 16 OP-902-007, Steam Generator Tube Rupture Recovery Procedure, Rev. 17 OP-902-009, Standard Appendices, Rev. 317 7

OP-902-007, Steam Generator Tube Rupture Recovery Procedure, Rev. 17 GEN EN-OP-115, Conduct of Operations, Rev. 24 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 4 OI-038-000, EOP Operations Expectations / Guidance, Rev. 16 TM-OP-100-03, Simulator Training, Rev. 14

NRC Scenario 1 2018 NRC Exam Scenario 1 D-1 Rev 3 SCENARIO SETUP A. Reset Simulator to IC-20. 100% power MOC.

B. Open and run schedule file located in 2018 NRC LOI Exam folder Scenario #1.sch.

C. Verify Scenario Malfunctions, Remotes, and Overrides are loaded, as listed in the Scenario Timeline.

D. Verify EDG A is removed from service as follows:

1. Place Danger Tag on EDG A C/S

2. Place Danger Tag on EDG A Output Breaker E. Verify LPSI Pump A is removed from service as follows:

1. Place LPSI Pump A C/S in OFF

2. Place Danger Tag on LPSI Pump A C/S F. Ensure Protected Train B sign is placed in SM office window.

G. Verify EOOS is 8.7 Yellow with EDG A and LPSI pump A out of service.

H. Protected Equipment covers on running SFP pump and as listed on the protected equipment posting sheets for EDG A and LPSI Pump A.

I.

Complete the simulator setup checklist.

J. 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, EOOS, EOI Library and the Brief database.

K. Capture pertinent simulator data by performing Scenario Based Testing using guidance contained in WSIM-DIR-006 for each scenario that is run for the exam. The intent of this step is to ensure adequate data is captured for reproduction of events independently following scenario completion.

NRC Scenario 1 2018 NRC Exam Scenario 1 D-1 Rev 3 SIMULATOR BOOTH INSTRUCTIONS Event 1 SG level instrument, SG-ILT-1123C, Fails Low

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

2. If directed as NAO to report SG levels on LCP-43, wait 3 minutes and report actual levels on SG

1 and SG #2 using Extreme View. (No failed indicators on LCP-43)

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 Charging Pump B Trip

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 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 3 PZR pressure control channel, RC-IPIC-0100X, 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.

Event 4 Steam Generator 2 Tube Leakage, Rapid Plant Power Reduction

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

2. If called as Chemistry and Radiation Protection to carry out the actions of UNT-005-032, acknowledge the request.

3. If called as Chemistry to sample both Steam generators for Activity, acknowledge the request.

NRC Scenario 1 2018 NRC Exam Scenario 1 D-1 Rev 3 Event 5 Main Turbine Shifts to Manual

1. At 1180 MW indicated on DEH panel, initiate Event Trigger 5.

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

Event 6 Steam Generator Tube Rupture, SG #2, HPSI Pump A Trip

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

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 HPSI Pump A 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 HPSI Pump A.

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 called as NAO to isolate the RWSP from purification, wait 5 minutes, and initiate Event Trigger 16 (SIR38 set to OFF) and report the following to the CR:

a. Steps 1 and 2 of Appendix 40 is complete (Closed FS-423 and FS-404).

b. Ask if the CR wants to continue on with the steps in Appendix 40 (Turn off the RWSP purification pump).

c. If directed to stop the purification pump, wait 2 minutes and report that it is complete.

6. If directed to investigate the operation of EDG B, wait 3 minutes and initiate Event Trigger 26 (EGR27 set to ACKN). Report that the B EDG is operating properly unloaded with no abnormal noises or vibrations.

Event 7 HPSI Pump B Fails to Autostart

1. No communications should occur for this event.

At the end of the scenario, before resetting, end data collection and save the file as 2018 Scenario 1-(start-end time).tid. Export to.csv file. Save the file into the folder for the appropriate crew.

NRC Scenario 1 2018 NRC Exam Scenario 1 D-1 Rev 3 SCENARIO TIMELINE KEY DESCRIPTION TRIGGER DELAY HH:MM:SS RAMP HH:MM:SS FINAL SI02B HPSI PUMP B FAILS TO AUTO START N/A 00:00:00 00:00:00 ACTIVE EG10A DG A OVERSPEED TRIP N/A 00:00:00 00:00:00 ACTIVE SI01A HPSI PUMP A TRIPPED N/A 00:00:00 00:00:00 ACTIVE SG11C FAIL SG2 NR LVL XMTR SG-ILT-1123C (0-100%)

1 00:00:00 00:00:00 0

CV01B CHARGING PUMP B TRIPPED 2

00:00:00 00:00:00 ACTIVE RX14A PZR PRESSURE CNTL CHL 100X FAIL (0-100%)(1500-2500 PSIA) 3 00:00:00 00:00:00 0

SG01B SG2 TUBE LEAK (100% = 3200 GPM) 4 00:00:00 00:00:00 0.17 SG01B SG2 TUBE LEAK (100% = 3200 GPM) 6 00:00:00 00:00:00 9.5 EGR29A EMERGENCY DIESEL GENERATOR A OUTPUT BREAKER N/A 00:00:00 00:00:00 RACKOUT SIR32 LPSI PUMP A N/A 00:00:00 00:00:00 RACKOUT SIR38 RWSP PURIFICATION PMP (FS-MPMP-0004) 16 00:00:00 00:00:00 OFF EGR27 EDG B LOCAL ANNUN ACK 26 00:00:00 00:00:00 ACKN

NRC Scenario 1 2018 NRC Exam Scenario 1 D-1 Rev 3 KEY DESCRIPTION TRIGGER DELAY HH:MM:SS RAMP HH:MM:SS FINAL DI-01A07A2S01-1 DEH TURBINE CONTROL TURBINE MANUAL 5

00:00:00 00:00:00 PUSH

Appendix D Scenario Outline Form ES-D-1 2018 NRC Exam Scenario 2 D-1 Rev 3 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; Maintain 100%. EDG A (TS 3.8.1.1b and d) and LPSI Pump A (3.5.2a) are out of service.

Critical Tasks:

(1) Establish Reactivity Control (2) Energize at least 1 vital AC bus Event No.

Malf.

No.

Event Type*

Event Description 1

RC22F2 I-BOP I-SRO TS-SRO RCS Wide Range Pressure Transmitter Safety Channel B fails low (RC-IPT-0102B).

2 CV05B2 C-ATC C-SRO Letdown Backpressure Control Valve, CVC-123B, fails closed.

3 RD02A11 C-ATC C-BOP C-SRO TS-SRO CEA 11 drops into the core.

4 MS09A I-BOP I-SRO During the down power, Steam Generator #1 steam flow transmitter fails low (FW-IFR-1011).

5 RD02A57 C-ATC C-SRO CEA 57 drops into the core resulting in an automatic Reactor Trip.

6 RD11A40 RD11A65 C-ATC 2 CEAs (40 & 60) mechanically stuck. (Critical Task 1, Establish Reactivity Control) 7 ED01A ED01B ED01C ED01D M-All A loss of offsite power occurs.

8 EG08B C-BOP C-SRO EDG B does not auto start (Critical Task 2, Energize at least 1 vital AC bus).

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Quantitative Attributes

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

2. Abnormal events (2-4) [Events 2,3,4 credited]

3

3. Major transients (1-2) 1

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

5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 0

6. Preidentified critical tasks (> 2) 2

NRC Scenario 2 - Narrative 2018 NRC Exam Scenario 2 D-1 Rev 3 The crew assumes the shift at 100% power with instructions to maintain 100% power. Emergency Diesel Generator A and Low Pressure Safety Injection Pump A is out of service.

Event 1: After taking the shift, RCS Wide Range Pressure Transmitter RC-IPT-0102B fails low. The crew will enter TS 3.3.1 action 2 (Functional Unit 5), TS 3.3.2 action 13 (Functional Units 1C and 3C) and bypass bistable 6 on CP-10 for Channel B in accordance with OP-009-007, Plant protection System.

Event 2: After Technical Specifications are addressed and Channel B bistables bypassed, the in service letdown backpressure control valve, CVC-123B will fail closed. The SRO will implement OP-901-112, Charging or Letdown Malfunction, Section E2, Letdown Malfunction. The SRO should direct the ATC and NAO to place the standby Backpressure control valve in service. The SRO may implement EN-OP-200, Transient Response Rules.

Event 3: After the standby Letdown backpressure Control Valve has been placed in service, 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 E1, CEA Misalignment Greater than 7 inches. The SRO will direct the BOP to adjust turbine load to match TAVG to TREF initially and then perform a rapid plant down power 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 and TS 3.2.3 action a.,

and b.. The SRO should implement EN-OP-200, Transient Response Rules.

Event 4: After the crew has commenced the rapid plant down power, Steam Generator #1 Steam flow instrument FW-IFR-1011 fails low. The Feedwater Control System will respond by lowering 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.

Event 5/6: After the crew has established control over SGWL for #1 SG, CEA 57 will drop into the core resulting in an automatic trip. The crew will respond using OP-902-000, Standard Post Trip Actions. The ATC should notice that 2 CEAs (40 & 65) have remained stuck out on the reactor trip and Emergency Boration is warranted (CRITICAL TASK 1).

Event 7/8: Once the crew has established Emergency Boration, a loss of offsite power will occur. During the loss of offsite power the B Emergency Diesel Generator will fail to auto start. The BOP should manually start the B EDG using the control switch to prevent the crew from entering the station blackout procedure (CRITICAL TASK 2). The SRO will diagnose and enter OP-902-003, Loss of Offsite Power /

Loss of Forced Circulation to stabilize the plant and to protect the Main Condenser. If Emergency Boration was previously started using BAM Pump A or BAM Pump B (A powered), then Emergency Boration will have to be re-initiated using gravity feed valves (B powered).

The scenario can be terminated after the crew has performed actions in OP-902-003 or at the lead examiners discretion.

NRC Scenario 2 - Critical Task Determination 2018 NRC Exam Scenario 2 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: Establish Reactivity Control This task is satisfied by manually initiating emergency boration using the gravity feed flowpath prior to entering OP-902-003, Loss of Offsite Power /

Loss of Forced Circulation. This is accomplished by opening BAM-113A or BAM-113B and closing CVC-183. This task becomes applicable following the Reactor Trip and the Loss of Offsite Power.

Based on Emergency Operating Procedure Required actions for Reactivity Control. Failure to initiate emergency boration would result in a condition that is not allowed by the facility license as analysis assumes that all CEAs are fully inserted during a reactor trip with the exception of the most reactive rod. OPS management Standard documented in TM-OP-100-

03.

(TM-OP-100-03, CT-1)

Rod bottom lights extinguished CEAs 40 & 60 CEA indicates withdrawn on CEAC Procedurally driven from OP-902-000 step 1.c.1 Opens BAM-113A or BAM-113B using control switch Closes CVC-183 using control switch Charging flow > 40 gpm on CP-4 BAM-113A or BAM-113B open indicating light illuminated CVC-183 closed indicating light illuminated CT-2: Energize at least 1 vital AC Bus This task is satisfied by manually starting Emergency Diesel generator B using the control switch prior to performing actions in OP-902-005, Station Blackout Recovery. This task becomes applicable once the Loss of Offsite power occurs.

Failure to energize at least one emergency bus will result in the plant remaining in a configuration that will not support protection if a subsequent event would occur. This lowers the mitigative capability of the plant.

Once the crew transitions to and begins taking actions in an inappropriate procedure without taking action to establish power to at least 1 safety bus when one is available demonstrates a significant performance deficiency regarding protecting critical safety functions.

OPS management Standard documented in TM-OP-100-03.

(TM-OP-100-03, CT-3)

Control room lighting reduces EDG B start light extinguished Bus 3B is de-energized Procedurally driven from OP-902-000 step 3.a.1.1)

Starts EDG B using the control switch Control room lighting brightens EDG B start light illuminated Bus 3B is energized EDG B sequencer is timing out Critical Task (NUREG-1021, Rev. 11 Appendix D) x 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.

x Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.

NRC Scenario 2 2018 NRC Exam Scenario 2 D-1 Rev 3 REFERENCES Event Procedures 1

OP-009-007, Plant Protection System, Rev. 19 Technical Specification 3.3.1, 3.3.2 2

OP-901-112, Charging or Letdown Malfunction, Rev. 6 3

OP-901-102, CEA or CEDMCS Malfunction, Rev. 304 OP-901-212, Rapid Plant Power Reduction, Rev. 9 OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction, Rev. 16 Technical Specification 3.1.3.1 Technical Specification 3.2.3 4

OP-901-201, Steam Generator Level Control Malfunction, Rev. 6 5

OP-902-000, Standard Post Trip Actions, Rev. 16 6

OP-902-003, Loss of Offsite Power / Loss of Forced Circulation, Rev. 10 OP-902-009, Standard Appendices, Rev. 317 7

OP-902-000, Standard Post Trip Actions, Rev. 16 GEN EN-OP-115, Conduct of Operations, Rev. 24 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 4 OI-038-000, EOP Operations Expectations / Guidance, Rev. 16 TM-OP-100-03, Simulator Training, Rev. 14

NRC Scenario 2 2018 NRC Exam Scenario 2 D-1 Rev 3 SCENARIO SETUP A. Reset Simulator to IC-20. 100% power MOC.

B. Open and run schedule file located in 2018 NRC LOI Exam folder Scenario #2.sch.

C. Verify Scenario Malfunctions, Remotes, and Overrides are loaded, as listed in the Scenario Timeline.

D. Verify EDG A is removed from service as follows:

1. Place Danger Tag on EDG A C/S

2. Place Danger Tag on EDG A Output Breaker E. Verify LPSI Pump A is removed from service as follows:

1. Place LPSI Pump A C/S in OFF

2. Place Danger Tag on LPSI Pump A C/S F. Ensure Protected Train B sign is placed in SM office window.

G. Verify EOOS is 8.7 Yellow with EDG A and LPSI pump A out of service.

H. Protected Equipment covers on running SFP pump and as listed on the protected equipment posting sheets for EDG A and LPSI Pump A.

I.

Complete the simulator setup checklist.

J. 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, EOOS, EOI Library and the Brief database.

K. Capture pertinent simulator data by performing Scenario Based Testing using guidance contained in WSIM-DIR-006 for each scenario that is run for the exam. The intent of this step is to ensure adequate data is captured for reproduction of events independently following scenario completion.

NRC Scenario 2 2018 NRC Exam Scenario 2 D-1 Rev 3 SIMULATOR BOOTH INSTRUCTIONS Event 1 RCS Pressure instrument, RC-IPT-0102B, Fails Low

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

2. If directed as NAO to report RCS pressure on LCP-43, wait 3 minutes and report that RCS wide range pressure transmitter RC-IPI-0102-B1 indicates failed low, report actual pressure on channel A using Extreme View.

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 Backpressure Control Valve CVC-123B, Fails Closed

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 valve.

3. When called to open CVC-125A, wait 1 minute and report that CVC-125A is open. (CVC-121A and CVC-125A are combined in one remote function)

4. When called to open CVC-121A, initiate Event Trigger 12 (CVR03 set to 100% ramped over 1 minute) and when valve indicates open in the director summary, report that valve CVC-121A is open (CVC-121A and CVC-125A are combined in one remote function).

5. When called to close CVC-121B, initiate Event Trigger 22 (CVR04 set to 0% ramped over 1 minute) and when valve indicates closed in the director summary, report that valve CVC-121B is closed (CVC-121B and CVC-125B are combined in one remote function).

6. When called to close CVC-125B, wait 1 minute and report that CVC-125B is closed. (CVC-121B and CVC-125B are combined in one remote function)

Event 3 CEA 11 drops into the core

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 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.

6. If requested to monitor Polisher Vessel D/P and remove as necessary, acknowledge the report.

Event 4 Steam Generator 1 Steam flow instrument FW-IFR-1011 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.

Event 5 CEA 57 drops into the core, Reactor Trip

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

NRC Scenario 2 2018 NRC Exam Scenario 2 D-1 Rev 3 Event 6 CEAs 40 & 60 mechanically stuck

1. No communications for this event Event 7 Loss of Offsite Power

1. On Lead Examiner's cue, initiate Event Trigger 7.

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

3. If called as NAO to perform Appendix 33-B, acknowledge direction, wait 3 minutes and open and run schedule file located in Remote Operator Actions (CAEP) OP-902-009 Appdx 33B.sch and follow prompts on the schedule file and with a copy of OP-902-009, Appendix 33-B available for reference.

4. If called as Gretna TOC about the status of offsite power, Report that an investigation is in progress and you will not have an estimate as to when power can be restored to WF3 until at least 30 minutes from now.

Event 8 Emergency Diesel Generator B Failure to Auto start

1. If directed to investigate the operation of EDG B, wait 3 minutes and initiate Event Trigger 17 (EGR27 set to ACKN). Report that EDG B is operating properly loaded with no abnormal noises or vibrations.

At the end of the scenario, before resetting, end data collection and save the file as 2018 Scenario 2-(start-end time).tid. Export to.csv file. Save the file into the folder for the appropriate crew.

NRC Scenario 2 2018 NRC Exam Scenario 2 D-1 Rev 3 SCENARIO TIMELINE KEY DESCRIPTION TRIGGER DELAY HH:MM:SS RAMP HH:MM:SS FINAL EG08B FAILURE OF DG B TO AUTOSTART N/A 00:00:00 00:00:00 ACTIVE EG10A DG A OVERSPEED TRIP N/A 00:00:00 00:00:00 ACTIVE RD11A40 CEA 40 MECHANICALLY STUCK N/A 00:00:00 00:00:00 ACTIVE RD11A65 CEA 65 MECHANICALLY STUCK N/A 00:00:00 00:00:00 ACTIVE RC22F2 PZR CPC SAFETY RC-IPT-0102B FAILS LO 1

00:00:00 00:00:00 ACTIVE CV05B2 LTDN BP REG VLV CVC-123B FAILS CLOSED 2

00:00:00 00:00:00 ACTIVE RD02A11 DROPPED CEA 11 3

00:00:00 00:00:00 ACTIVE MS09A SG 1 STEAM FLOW TRANSMIT FAIL (0-100% OF RANGE) 4 00:00:00 00:00:00 0

RD02A57 DROPPED CEA 57 5

00:00:00 00:00:00 ACTIVE ED01A FEEDER BREAKER 7172 TRIP IN SWITCHYARD 7

00:00:00 00:00:00 ACTIVE ED01B FEEDER BREAKER 7176 TRIP IN SWITCHYARD 7

00:00:00 00:00:00 ACTIVE ED01C FEEDER BREAKER 7182 TRIP IN SWITCHYARD 7

00:00:00 00:00:00 ACTIVE

NRC Scenario 2 2018 NRC Exam Scenario 2 D-1 Rev 3 KEY DESCRIPTION TRIGGER DELAY HH:MM:SS RAMP HH:MM:SS FINAL ED01D FEEDER BREAKER 7186 TRIP IN SWITCHYARD 7

00:00:00 00:00:00 ACTIVE SIR32 LPSI PUMP A N/A 00:00:00 00:00:00 RACKOUT EGR29A EMERGENCY DIESEL GENERATOR A OUTPUT BREAKER N/A 00:00:00 00:00:00 RACKOUT CVR03 CVC-121A/125A LTDN BPCV A ISOL (0-100%)

12 00:00:00 00:01:00 100 CVR04 CVC-121B/125B LTDN BPCV B ISOL (0-100%)

22 00:00:00 00:01:00 0

EGR27 EDG B LOCAL ANNUN ACK 17 00:00:00 00:00:00 ACKN

Appendix D Scenario Outline Form ES-D-1 2018 NRC Exam Scenario 3 D-1 Rev 3 Facility:

Waterford 3 Scenario No.:

3 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%. EDG A (TS 3.8.1.1b and d) and LPSI Pump A (TS 3.5.2a) are out of service.

Critical Tasks:

(1) Trip any RCP exceeding operating limits (2) Establish Required Safety Injection Flow Event No.

Malf.

No.

Event Type*

Event Description 1

CV12A1 I-ATC I-SRO Volume Control Tank level transmitter CVC-ILT-0227 failure high.

2 NI01A I-BOP I-SRO TS-SRO Excore Nuclear Instrument ENI-IJI-001-A fails low.

3 CC01A C-BOP C-SRO TS-SRO Component Cooling Water Pump A Trips on overcurrent.

4 FW21A FW21AA C-ATC C-BOP C-SRO A lowering of Main Condenser vacuum.

5 DI-02A05A1S02-1 DI-02A05A1S04-1 DI-02A05A1S07-1 C-ATC C-SRO Group P CEAs will insert into the core automatically and uncontrollably.

6 RC23A M-All Loss of Coolant Accident will occur. (Critical Task 1, Trip any RCP exceeding operating limits).

7 SI01E C-BOP C-SRO Low Pressure Safety Injection Pump B trips on overcurrent.

(Critical Task 2, Establish Required Safety Injection flow).

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Quantitative Attributes

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

2. Abnormal events (2-4) [Events 1,3,4 credited]

3

3. Major transients (1-2) 1

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

5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 1

6. Preidentified critical tasks (> 2) 2

NRC Scenario 3 - Narrative 2018 NRC Exam Scenario 3 D-1 Rev 3 The crew assumes the shift at 100% power with instructions to maintain 100% power. Emergency Diesel Generator A and Low Pressure Safety Injection Pump A is out of service.

Event 1: 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.

Event 2: After the crew addresses the VCT instrument malfunction, ENI Channel A Linear Power Instrument (ENI-IJI-0001-A), (upper channel), fails low. The SRO should review and enter Technical Specification 3.3.1 action 2 (Functional Unit 2 - Linear Power Level - High, Functional Unit 9 - Local Power Density - High, Functional Unit 10 - DNBR - Low) and bypass Hi Linear Power, Hi LPD, Lo DNBR (1, 3 & 4) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in accordance with OP-009-007, Plant Protection System.

Event 3: After Technical Specifications are addressed and Channel A bistables bypassed, Component Cooling Water Pump A 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 A. 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 not be exited because the AB electrical bus is aligned to the B side.

Event 4: After the standby Component Cooling Water Pump has been placed in service, a leak in the Main Condenser develops and Main Condenser vacuum begins to drop (approx. 2 minutes until first alarm and 4 minutes until <25 Hg vacuum). 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.

Event 5: After the crew has commenced the rapid plant down power, and the ATC commences ASI control, Group P CEAs will automatically and uncontrollably begin to insert into the core. The ATC will recognize this and the SRO will direct a manual Reactor Trip in accordance with OP-901-102, CEA or CEDMCS Malfunction. The SRO will direct the crew to perform the Standard post Trip Actions in accordance with OP-902-000, Standard Post Trip Actions.

Event 6: After the reactor trip, an RCS leak occurs on RCS Cold Leg 1A that progresses to a Large Break Loss of Coolant Accident. The ATC should manually stop all RCPs following a loss of subcooling or loss of CCW to the RCPs (CRITICAL TASK 1). The crew should re-diagnose to OP-902-002, Loss of Coolant Accident Recovery Procedure.

Event 7: When the crew enters OP-902-002, LOCA Recovery Procedure, Low Pressure Safety Injection (LPSI) pump B will trip on overcurrent. The crew should recognize that OP-902-002 safety functions are not met and the SRO should go to OP-902-008, Functional Recovery. When the SRO performs prioritization, Inventory Control (IC-2) should be the highest priority. The SRO should request TSC/Shift Manager permission and direct the BOP to align Containment Spray pump B to replace LPSI pump B and re-establish LPSI flow (CRITICAL TASK 2).

The scenario can be terminated after the crew has re-established low pressure safety injection flow or at the lead examiners discretion.

NRC Scenario 3 - Critical Task Determination 2018 NRC Exam Scenario 3 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: Trip any RCP exceeding operating limits This task is satisfied by manually stopping Reactor Coolant Pumps prior to exiting Standard Post Trip Actions.

This task becomes applicable following a loss of subcooling in the RCS.

Based on EOP required actions for a Loss of Coolant Accident. RCPs are stopped for numerous reasons including potential RCP seal damage and pumping more mass out of the break that can result in uncovering fuel. In this case, the RCPs are stopped to prevent more mass lost out of the RCS which jeopardizes the fuel clad barrier worse than already contained in the scenario. OPS management standard documented in TM-OP-100-03.

(TM-OP-100-03, CT-23)

CCW flow low/lost to RCPs alarms on CP-2 and CP-18 CCW valve status CP-2 CSAS initiated CP-8 Procedurally driven from OP-902-000 step 3.b.1 and 9.3 Stops RCPs using control switch RCP off light illuminated RCP indicated flow lowering CT-2: Establish Required Safety Injection Flow This task is satisfied by manually aligning an available Containment Spray Pump to replace a Low Pressure Safety Injection Pump prior to exiting step 11 of IC-2 Continuing actions located in OP-902-008. This task becomes applicable once the LPSI Pump B Trips.

Based on Emergency Operating Procedure Required flow. Failure to take action to establish required Low Pressure Safety Injection Flow during a loss of coolant accident would degrade the ability to maintain the fuel covered and cooled. OPS management standard documented in TM-OP-100-03.

(TM-OP-100-03, CT-3)

LPSI Pump amber light illuminated Safety Function Status Checklist not met Procedurally driven from OP-902-008 step 11 of IC-2 Cont. Actions Closes CS-125B using control switch Opens SI-125B/412B using control switch Starts CS Pump B using control switch Opens SI-415B using control switch LPSI Header B Flow indicated Critical Task (NUREG-1021, Rev. 11 Appendix D) x 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.

x Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.

NRC Scenario 3 2018 NRC Exam Scenario 3 D-1 Rev 3 REFERENCES Event Procedures 1

OP-901-113, Volume Control Tank Makeup Control Malfunction, Rev. 302 2

OP-009-007, Plant Protection System, Rev. 19 Technical Specifications 3.3.1 3

OP-901-510, Component Cooling Water System Malfunction, Rev. 303 Technical Specification 3.7.3 4

OP-901-220, Loss of Condenser Vacuum, Rev. 303 OP-901-212, Rapid Plant Power Reduction, Rev. 9 5

OP-901-102, CEA or CEDMCS Malfunction, Rev. 304 OP-902-000, Standard Post Trip Actions, Rev. 16 6

OP-902-002, Loss of Coolant Accident Recovery, Rev. 20 OP-902-009, Standard Appendices, Rev. 317 7

OP-902-008, Functional Recovery, Rev. 27 OP-902-009, Standard Appendices, Rev. 317 GEN EN-OP-115, Conduct of Operations, Rev. 24 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 4 OI-038-000, EOP Operations Expectations / Guidance, Rev. 16 TM-OP-100-03, Simulator Training, Rev. 14

NRC Scenario 3 2018 NRC Exam Scenario 3 D-1 Rev 3 SCENARIO SETUP A. Reset Simulator to IC-20. 100% power MOC.

B. Open and run schedule file located in 2018 NRC LOI Exam folder Scenario #3.sch.

C. Verify Scenario Malfunctions, Remotes, and Overrides are loaded, as listed in the Scenario Timeline.

D. Verify EDG A is removed from service as follows:

1. Place Danger Tag on EDG A C/S

2. Place Danger Tag on EDG A Output Breaker E. Verify LPSI Pump A is removed from service as follows:

1. Place LPSI Pump A C/S in OFF

2. Place Danger Tag on LPSI Pump A C/S F. Ensure Protected Train B sign is placed in SM office window.

G. Verify EOOS is 8.7 Yellow with EDG A and LPSI pump A out of service.

H. Protected Equipment covers on running SFP pump and as listed on the protected equipment posting sheets for EDG A and LPSI Pump A.

I.

Complete the simulator setup checklist.

J. 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, EOOS, EOI Library and the Brief database.

K. Capture pertinent simulator data by performing Scenario Based Testing using guidance contained in WSIM-DIR-006 for each scenario that is run for the exam. The intent of this step is to ensure adequate data is captured for reproduction of events independently following scenario completion.

NRC Scenario 3 2018 NRC Exam Scenario 3 D-1 Rev 3 SIMULATOR BOOTH INSTRUCTIONS Event 1 VCT level instrument, CVC-ILT-0227, Fails High

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

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 Channel A Excore NI Safety Channel, ENI-IJI-0001A, upper detector 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 3 minutes and report channel ENI-IJI-0001-A1 appears to be failed downscale. All other power channels read approximately 100%.

Event 3 Component Cooling Water Pump A Trips

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

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

3. If called as the watchstander and sent to CCW Pump A 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 4 Main Condenser Leak, Rapid Power Reduction

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

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 5 Group P CEAs automatically and uncontrollably insert into the core

1. On Lead Examiner's cue (when ASI control is started), initiate Event Trigger 5.

2. There should be no communications for this event.

NRC Scenario 3 2018 NRC Exam Scenario 3 D-1 Rev 3 Event 6 Loss of Coolant Accident

1. After the reactor trips, 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 NAO to isolate the RWSP from purification, wait 5 minutes, and set remote SIR38 to OFF and report the following to the CR:

a. Steps 1 and 2 of Appendix 40 are complete (Closed FS-423 and FS-404).

b. Ask if the CR wants to continue on with the steps in Appendix 40 (Turn off the RWSP purification pump).

c. If directed to stop the purification pump, wait 2 minutes and report that it is complete.

5. If directed to investigate the operation of EDG B, wait 3 minutes and initiate Event Trigger 16 (EGR27 set to ACKN). Report that EDG B is operating properly unloaded with no abnormal noises or vibrations.

Event 7 Low Pressure Safety Injection Pump B Trips / Align CS to replace LPSI

1. After OP-902-002 has been entered and at lead Examiners discretion, 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 called as an NAO to investigate the trip at the breaker, report overcurrent flags on all 3 phases.

5. If called as an NAO to investigate the trip at the pump, report that there is no oil visible on the motor bearing sightglass and oil is dripping from the motor housing.

6. If called as an NAO to rack out LPSI pump B breaker or open the knife switch, wait 2 minutes, initiate Event Trigger 8 (SIR33 set to RACKOUT) and then make report to the Control Room that you have done so.

7. If called as an NAO to place CS-125B override keyswitch in "Override", wait 2 minutes and then initiate Event Trigger 9 (CSR13B set to OVERRIDE) to place CS-125B in override. Make report to the Control Room that you have done so.

8. If called as an NAO to locally close SI-129B, wait 3 minutes and then initiate Event Trigger 23 (SIR50 & SIR51) to close SI-129B. Pull up Extreme View - LP Safety Injection to check position and make report to the Control Room that you have done so.

At the end of the scenario, before resetting, end data collection and save the file as 2018 Scenario 3-(start-end time).tid. Export to.csv file. Save the file into the folder for the appropriate crew.

NRC Scenario 3 2018 NRC Exam Scenario 3 D-1 Rev 3 SCENARIO TIMELINE KEY DESCRIPTION TRIGGER DELAY HH:MM:SS RAMP HH:MM:SS FINAL EG10A DG A OVERSPEED TRIP N/A 00:00:00 00:00:00 ACTIVE CV12A1 VCT LEVEL XMTR CVC-ILIC-0227 FAILS HI 1

00:00:00 00:00:00 ACTIVE NI01A UPPER DETECTOR (A1) SAFETY CHANNEL A FAIL (0-100%)

2 00:00:00 00:00:00 0

CC01A CCW PUMP A TRIP 3

00:00:00 00:00:00 ACTIVE FW21A CONDENSER A AIR INLEAK (100%=100% OF VAC BKR) 4 00:00:00 00:00:00 20 FW21AA CONDENSER A AIR INLEAK VACUUM SETPOINT 4

00:00:00 00:00:00 23.8 RC23A RCS COLD LEG 1A RUPTURE 6

00:00:00 00:02:00 17 SI01E LPSI PUMP B TRIP 7

00:00:00 00:00:00 ACTIVE SIR32 LPSI PUMP A N/A 00:00:00 00:00:00 RACKOUT EGR29A EMERGENCY DIESEL GENERATOR A OUTPUT BREAKER N/A 00:00:00 00:00:00 RACKOUT SIR33 LPSI PUMP B 8

00:00:00 00:00:00 RACKOUT CSR13B CS-125B REMOTE KEY SW TO CLOSE VALVE 9

00:00:00 00:00:00 OVERRIDE

NRC Scenario 3 2018 NRC Exam Scenario 3 D-1 Rev 3 KEY DESCRIPTION TRIGGER DELAY HH:MM:SS RAMP HH:MM:SS FINAL SIR50 SI-129B SDC B FCV CONTROL 23 00:00:00 00:00:00 LOCAL SIR51 SI-129B SDC B FCV POSITION 23 00:00:05 00:00:00 0

EGR27 EDG B LOCAL ANNUN ACK 16 00:00:00 00:00:00 ACKN DI-02A05A1S02-1 CEDMCS GROUP SELECT 5

00:00:00 00:00:00 P

DI-02A05A1S04-1 CEDMCS MODE SELECT 5

00:00:00 00:00:00 MG DI-02A05A1S07-1 CEDMCS WITHDRAW/INSERT 5

00:00:00 00:00:00 INSERT

ML18333A346

Text

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