ML20267A205

08 Draft Outlines

ML20267A205
Person / Time
Site:	Waterford
Issue date:	08/26/2020
From:	Greg Werner Operations Branch IV
To:	Entergy Operations
References
Download: ML20267A205 (66)

Text

ES-401 PWR Examination Outline Form ES-401-2 Rev. 11 Facility: Waterford 3 Date of Exam: August 26, 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and 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 3

2 5

2 10 2

1 3

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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it 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 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 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. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 SRO E/APE # / Name / Safety Function K1 K2 K3 A1 A2 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 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X

AA2 Ability to determine and/or interpret the following as they apply to Loss of Reactor Coolant Makeup: (CFR: 43.5 / 45.13)

AA2.02 Charging pump problems 3.7 76 000025 (APE 25) Loss of Residual Heat Removal System / 4 X

2.2.25 Knowledge of the bases in TS for limiting conditions for operations and safety limits. (SRO Only) (CFR: 43.2) 4.2 77 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X

AA2 Ability to determine and/or interpret the following as they apply to a Pressurizer Pressure Control System Malfunction: (CFR:

43.5 / 45.13)

AA2.02 Normal values for RCS pressure 3.9 78 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 000055 (EPE 55) Station Blackout / 6 X

2.1.6 Ability to manage the control room crew during plant transients. (CFR: 41.5 /

43.5/ 45.12/ 45.13) 4.8 81 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X 2.4.17 Knowledge of EOP terms and definitions. (CFR: 41.10 / 45.13) 4.3 79 000058 (APE 58) Loss of DC Power / 6 X

AA2 Ability to determine and/or interpret the following as they apply to Loss of DC Power:

DC loads lost; impact on ability to operate and monitor plant systems (43.5 / 45.13) 3.9 80 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8

ES-401 3

Form ES-401-2 Rev. 11 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (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. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 SRO E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 X

2.1.28 Knowledge of the purpose and function of major system components and controls (CFR: 41.7) 4.1 82 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 Pressurizer Level Control Malfunctions: (CFR:

43.5 / 45.13)

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 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X

2.3.6 Ability to approve release permits. (CFR: 41.13 /

43.4 / 45.10) 3.8 84 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 (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 CooldownDepressurization / 4 Not Applicable (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not Applicable (BW E13 & E14) EOP Rules and Enclosures Not Applicable (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2

ES-401 5

Form ES-401-2 Rev. 11 (CE E09) Functional Recovery X

EA2. Ability to determine and interpret the following as they apply to the (Functional Recovery) (CFR: 43.5 /

45.13)

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 X

EA2 Ability to determine and/or interpret the following as they apply to Loss of Forced Circulation and/or LOOP and/or a Blackout:

(CFR: 41.10 / 43.5 / 45.13)

EA2.7 RCS subcooling 3.9 K/A Category Point Totals:

2 2

Group Point Total:

4

ES-401 6

Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 SRO System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control X

A2 Ability to (a) predict the impacts of the following malfunctions on the Chemical and Volume Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 to 41.7 / 43.5 /

45.3 / 45.5)

A2.18 High VCT level 3.1 86 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water X 2.1.9 Ability to direct personnel inside the control room. (CFR: 41.10 / 45.5 /

45.12/ 45.13) 4.5 87 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser Not Applicable 026 (SF5 CSS) Containment Spray X

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

41.5 / 43.5 / 45.3 / 45.13)

A2.07 Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit 3.9 88 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)

Auxiliary/Emergency Feedwater

ES-401 7

Form ES-401-2 Rev. 11 062 (SF6 ED AC) AC Electrical Distribution X 2.2.21 Knowledge of pre-and post-maintenance operability requirements.

(CFR: 41.10 / 43.2) 4.1 89 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel Generator X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

A2.15 Water buildup in cylinders 3.1 90 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:

3 2 Group Point Total:

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

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 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 X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

A2.04 Loss of one, two, or three charging pumps 3.7 91 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor

ES-401 8

Form ES-401-2 Rev. 11 027 (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 X

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

43.5/ 45.13) 4.5 92 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste X

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

A2.04 Failure of automatic isolation 3.3 93 071 (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 K/A Category Point Totals:

2 1 Group Point Total:

3

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

Form ES-401-3 Rev. 11 Facility:

Date of Exam:

Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7) 4.1 94 2.1.39 Knowledge of conservative decision making practices.

(CFR: 41.10 / 43.5 / 45.12) 4.3 95 Subtotal 2

2. Equipment Control 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 /

43.2 / 45.13) 4.1 96 2.2.43 Knowledge of the process used to track inoperable alarms. (CFR: 41.10 / 43.5 / 45.13) 3.3 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. (CFR: 41.12 / 43.4 / 45.10) 3.8 98 Subtotal 1

4. Emergency Procedures/Plan 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR:

41.10 / 43.2 / 45.6) 4.7 99 2.4.42 Knowledge of emergency response facilities. (CFR:

41.10 / 45.11) 3.8 100 Subtotal 2

Tier 3 Point Total 7

ES-401 PWR Examination Outline (RO)

Form ES-401-2 Rev. 11 Facility: Waterford 3 Date of Exam: August 26, 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 2

3 N/A 4

4 N/A 2

18 6

2 1

2 2

1 1

2 9

4 Tier Totals 4

4 5

5 5

4 27 10

2.

Plant Systems 1

2 2

2 5

2 2

4 2

3 2

2 28 5

2 1

0 2

1 1

0 1

1 1

1 1

10 3

Tier Totals 3

2 4

6 3

2 5

3 4

3 3

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

2 3

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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it 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 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 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. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 RO E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

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

Reactor Trip, Stabilization, Recovery / 1 X

Knowledge of the operational implications of the following concepts as they apply to the (Reactor Trip Recovery)(CFR: 41.8 / 41.10, 45.3)

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: (CFR 41.7 /

45.7)

AK2.03 Controllers and positioners 2.5 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: (CFR 41.5 / 41.10 / 45.6 /

45.13)

EK3.16 Containment temperature, pressure, humidity and level limits 3.8 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:

(CFR 41.7 / 45.5 / 45.6)

EA1.16 Balancing of HPI loop flows 3.5 4

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 (Loss of RC Flow): (CFR 43.5/ 45.13)

AA2.01 Cause of RCP failure 3.0 5

000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X

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, or alignment of filters. (CFR: 41.12 / 45.9 /

45.10) 3.2 6

000025 (APE 25) Loss of Residual Heat Removal System / 4 Not Sampled 000026 (APE 26) Loss of Component Cooling Water / 8 X

AA2 Ability to determine and/or interpret the following as they apply to Loss of Component Cooling Water: (CFR: 41.10 /

43.5 / 45.13)

AA2.06 The length of time after the loss of CCW flow to a component before that component may be damaged 2.8 7

000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 Not Sampled

ES-401 3

Form ES-401-2 Rev. 11 000029 (EPE 29) Anticipated Transient Without Scram / 1 X

EA1 Ability to operate and/or monitor the following as they apply to an Anticipated Transient Without Scram: (CFR: 41.7 / 45.5 /

45.6)

EA1.12 M/G set power supply and reactor trip breakers 4.1 8

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

EK3 Knowledge of the reasons for the following responses as the apply to the SGTR: (CFR 41.5 / 41.10 / 45.6 / 45.13)

EK3.09 Criteria for securing/throttling ECCS 4.1 9

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: CFR: 41.7 / 45.7)

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

3.3 10 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

EK1. Knowledge of the operational implications of the following concepts as they apply to the (Loss of Feedwater) (CFR:

41.8 / 41.10 / 45.3)

EK1.2 Normal, abnormal and emergency operating procedures associated with (Loss of Feedwater).

3.2 11 000055 (EPE 55) Station Blackout / 6 X

X EK1 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout :

(CFR 41.8 / 41.10 / 45.3)

EK1.01 Effect of battery discharge rates on capacity 3.3 12 000056 (APE 56) Loss of Offsite Power / 6 X

AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: (CFR 41.5,41.10 /

45.6 / 45.13)

AK3.02 Actions contained in EOPs for loss of offsite power 4.4 13 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X

AA1 Ability to operate and/or monitor the following as they apply to Loss of Vital AC Electrical Instrument Bus: (CFR: 41.7 / 45.5 /

45.6)

AA1.06 Manual control of components for which automatic control is lost 3.5 14 000058 (APE 58) Loss of DC Power / 6 X

AA2 Ability to determine and/or interpret the following as they apply to Loss of DC Power:

(CFR: 43.5 / 45.13)

AA2.02 125V dc bus voltage, low/critical low, alarm 3.4 15

ES-401 4

Form ES-401-2 Rev. 11 000062 (APE 62) Loss of Nuclear Service Water / 4 X

2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination the transmission system operator. (CFR: 41.10 /

43.5 / 45.13) 2.6 16 000065 (APE 65) Loss of Instrument Air / 8 X

AA2 Ability to determine and/or interpret the following as they apply to Loss of Instrument Air: (CFR: 41.10 / 43.5 / 45.13)

AA2.01 Cause and effect of low-pressure instrument air alarm 2.9 17 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X

AA2 Ability to determine and/or interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: (CFR: 41.5 /

43.5 / 45.5 / 45.7 / 45.8)

AA2.04 VARs outside the capability curve 3.6 18 (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 2

3 4

4 2

Group Point Total:

18

ES-401 5

Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 RO E/APE # / Name / Safety Function K1 K2 K3 A1 A2 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/or interpret the following as they apply to a Dropped Control Rod: (CFR: 43.5 /

45.13)

AA2.03 Dropped rod, using in-core/ex-core instrumentation, in-core or loop temperature measurements 3.6 19 000005 (APE 5) Inoperable/Stuck Control Rod / 1 X

2.4.6 Knowledge of EOP mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) 3.7 20 000024 (APE 24) Emergency Boration / 1 X

AK2 Knowledge of the relationship between the Emergency Boration and the following systems or components: (CFR: 41.7 /

45.7)

AK2.04 Pumps 2.6 21 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 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 X

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

41.5 / 43.5 / 45.12) 4.2 22 000037 (APE 37) Steam Generator Tube Leak / 3 X

AK3 Knowledge of the reasons for the following responses as they apply to a Steam Generator Tube Leak:

(CFR: 41.5 / 41.7 /41.10 /

45.6 / 45.13)

AK3.07 Actions contained in EOP for S/G tube leak 4.2 23 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9

ES-401 6

Form ES-401-2 Rev. 11 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 X

AK1. Knowledge of the operational implications of the following concepts as they apply to Accidental Gaseous Radwaste Release:

(CFR 41.8 / 41.10 / 45.3)

AK1.04 Calculation of offsite doses due to a release from the power plant 2.5 24 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 X

AA1 Ability to operate and/or monitor the following as they apply to Control Room Evacuation: (CFR: 41.7 / 45.5

/ 45.6)

AA1.12 Auxiliary Shutdown panel controls and indicators 4.4 25 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 (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 CooldownDepressurization / 4 Not Applicable (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not Applicable (BW E13 & E14) EOP Rules and Enclosures Not Applicable (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4

ES-401 7

Form ES-401-2 Rev. 11 (CE A16) Excess RCS Leakage / 2 X

AK2 Knowledge of the relationship between the Excess RCS Leakage and the following systems or components: (CFR: 41.7 /

45.7)

AK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

3.2 26 (CE E09) Functional Recovery X

EK3 Knowledge of the reasons for the following responses as they apply to Functional Recovery: (CFR:

41.5 / 41.10, 45.6, 45.13)

EK3.1 Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

3.5 27 (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:

1 2

2 1

1 2

Group Point Total:

9

ES-401 8

Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 RO System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump X

K1 Knowledge of the physical connections and/or cause and effect relationships between the Reactor Coolant Pump System and the following systems: (CFR: 41.2 to 41.9 / 45.7 / 45.8)

K1.01 RCP lube oil 2.6 28 004 (SF1; SF2 CVCS) Chemical and Volume Control X

K2 Knowledge of bus power supplies to the following: (CFR: 41.7)

K2.06 Control instrumentation 2.6 29 005 (SF4P RHR) Residual Heat Removal X

X K3 Knowledge of the effect that a loss or malfunction of the Residual Heat Removal System will have on the following systems or system parameters:

(CFR: 41.7 / 45.6)

K3.01 RCS K6 Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 /

45.7)

K6.03 RHR heat exchanger 3.9 2.5 30 31 006 (SF2; SF3 ECCS) Emergency Core Cooling X

K4 Knowledge of Emergency Core Cooling System design feature(s) and/or interlock(s), which provide for the following: (CFR: 41.7 / 41.8)

K4.17 Safety Injection valve interlocks 3.8 32 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X

K5 Knowledge of the operational implications of the following concepts as the apply to PRTS: (CFR: 41.5 / 45.7)

K5.02 Method of forming a steam bubble in the PZR 3.1 33 008 (SF8 CCW) Component Cooling Water X

X K4 Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

K4.02 Operation of the surge tank, including the associated valves and controls 2.9 34 010 (SF3 PZR PCS) Pressurizer Pressure Control X

A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: (CFR: 41.5 / 45.5)

A1.06 RCS heatup and cooldown effect on pressure 3.1 35

ES-401 9

Form ES-401-2 Rev. 11 012 (SF7 RPS) Reactor Protection X

X 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: (CFR:

41.5 / 43.5 / 45.3 / 45.5)

A2.01 Faulty bistable operation K4 Knowledge of Reactor Protection System design feature(s) and/or interlock(s), which provide for the following: (CFR: 41.7)

K4.02 Automatic reactor trip when RPS setpoints are exceeded for each RPS function; functional basis for each 3.1 3.9 36 37 013 (SF2 ESFAS) Engineered Safety Features Actuation X

X A3 Ability to monitor automatic operation of the ESFAS including: (CFR:

41.7 / 45.5)

A3.01 Input channels and logic 2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5/45.13 3.7 4.0 38 55 022 (SF5 CCS) Containment Cooling X

A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8)

A4.02 CCS pumps 3.6 39 025 (SF5 ICE) Ice Condenser Not Applicable 026 (SF5 CSS) Containment Spray X 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

(CFR: 41.10 / 43.5 / 45.3 / 45.12) 4.2 40 039 (SF4S MSS) Main and Reheat Steam X

X A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8)

A4.04 Emergency feedwater pump turbines A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including:

(CFR: 41.5 / 45.5)

A1.05 RCS T-ave 3.8 3.2 41 42 059 (SF4S MFW) Main Feedwater X

A3 Ability to monitor automatic features of the Main Feedwater System, including: (CFR: 41.7 / 45.5)

A3.04 Turbine driven feed pump 2.5 43

ES-401 10 Form ES-401-2 Rev. 11 061 (SF4S AFW)

Auxiliary/Emergency Feedwater X

X 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: (CFR:

41.5 / 43.5 / 45.3 / 45.13)

A2.05 Automatic control malfunction K5 Knowledge of the operational implications of the following concepts as the apply to the AFW: (CFR: 41.5 /

45.7)

K5.05 Feed line voiding and water hammer 3.1 2.7 44 45 062 (SF6 ED AC) AC Electrical Distribution X

A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 41.5 / 45.5)

A1.03 Effect on instrumentation and controls of switching power supplies 2.5 46 063 (SF6 ED DC) DC Electrical Distribution X

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

41.7)

K4.02 Breaker interlocks, permissives, bypasses and cross-ties 2.9 47 064 (SF6 EDG) Emergency Diesel Generator X X K6 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: (CFR: 41.7 / 45.7)

K6.07 Air receivers 2.7 48 073 (SF7 PRM) Process Radiation Monitoring X

X K4 Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

K4.01 Release termination when radiation exceeds setpoint A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including: (CFR: 41.5 / 45.7)

A1.01 Radiation levels 4.0 3.2 49 50

ES-401 11 Form ES-401-2 Rev. 11 076 (SF4S SW) Service Water X

K3 Knowledge of the effect that a loss or malfunction of the Service Water System will have on the following systems or system parameters: (CFR: 41.7 / 45.6)

K3.02 Secondary closed cooling water 2.5 51 078 (SF8 IAS) Instrument Air X

X X K2 Knowledge of bus power supplies to the following: (CFR: 41.7)

K2.01 Instrument Air Compressor A3 Ability to monitor automatic operation of the IAS, including: (CFR:

41.7 / 45.5)

A3.01 Air pressure 2.7 3.1 52 53 103 (SF5 CNT) Containment X

K1 Knowledge of the physical connections and/or cause and effect relationships between the Containment System and the following systems: (CFR:

41.9 / 45.7 / 45.8)

K1.01 CCS 3.6 54 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:

2 2

2 5

2 2

4 2

3 2

2 Group Point Total:

28 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 RO System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant X

K3 Knowledge of the effect that a loss or malfunction of the Reactor Coolant System will have on the following systems or system parameters: (CFR:

41.7)

K3.03 Containment 4.2 56 011 (SF2 PZR LCS) Pressurizer Level Control X

K1 Knowledge of the physical connections and/or cause and effect relationships between the Fire Protection System and the following systems: (CFR:

41.4 / 41.7 / 41.8 / 45.7 / 45.8)

K1.05 Reactor regulating system 3.4 65 014 (SF1 RPI) Rod Position Indication

ES-401 12 Form ES-401-2 Rev. 11 015 (SF7 NI) Nuclear Instrumentation X

A3 Ability to monitor automatic operation of the NIS, including: (CFR:

41.7 / 45.5)

A3.03 Verification of proper functioning/operability 3.9 62 016 (SF7 NNI) Nonnuclear Instrumentation X

K3 Knowledge of the effect that a loss or malfunction of the Nonnuclear Instrumentation System will have on the following systems or system parameters:

(CFR: 41.7 / 45.6)

K3.04 MFW system 2.6 57 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X

X K2 Knowledge of bus power supplies to the following: (CFR: 41.7)

K2.01 Hydrogen recombiners 2.5 58 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator X

K5 Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Steam Generator System: (CFR: 41.5 / 45.7)

K5.01 Effect of secondary parameters, pressure, and temperature on reactivity 3.4 59 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X

K6 Knowledge of the effect of a loss or malfunction on the following will have on the SDS: (CFR: 41.7 / 45.7)

K6.03 Controller and positioners, including ICS, S/G, CRDS 2.7 60 045 (SF 4S MTG) Main Turbine Generator X

A1 Ability to predict and/or monitor changes in parameters associated with operation of the Main Turbine Generator System, including: (CFR: 41.5 / 45.5)

A1.06 Expected response of secondary plant parameters following T/G trip 3.3 61 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate X 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: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

A2.04 Loss of condensate pumps 2.5 63

ES-401 13 Form ES-401-2 Rev. 11 068 (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 X

2.4.12 Knowledge of operating crew responsibilities during emergency and abnormal operations.

4.0 64 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:

1 1

2 0

1 0

1 1

1 1

1 Group Point Total:

10

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

Form ES-401-3 Rev. 11 Facility: Waterford 3 Date of Exam: August 26, 2020 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps. (CFR:

41.10 / 43.5 / 45.12) 4.6 66 2.1.41 2.1.41 Knowledge of the refueling process.

2.8 67 Subtotal 2

2. Equipment Control 2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, lineups, or tag-outs. (Reference Potential) (CFR: 41.10 / 43.3 / 45.13) 3.9 68 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination the transmission system operator. (CFR: 41.10 / 43.5 /

45.13) 2.6 69 2.2.35 Ability to determine TS for mode of operation. (CFR:

41.7 / 41.10 / 43.2 / 45.13) 3.6 70 Subtotal 3

3. Radiation Control 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4

/ 45.10) 3.8 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, or alignment of filters.

(CFR: 41.12 / 45.9 / 45.10) 3.2 72 Subtotal 2

4. Emergency Procedures/Plan 2.4.1 Knowledge of EOP entry conditions and immediate action steps. (CFR: 41.10 / 43.5 / 45.13) 4.6 73 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. (CFR:

41.10 / 43.5 / 45.13) 3.8 74 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR:

41.10 / 43.5 / 45.13) 3.8 75 Subtotal 3

Tier 3 Point Total 10

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /

Group Randomly Selected K/A Reason for Rejection 1/1 000007 (EPE 7; BW E02&E10; CE E02)

Reactor Trip, Stabilization, Recovery / 1 RO question 1 EK1.02 is not listed in rev 2 of K/A catalog. Randomly resampled and selected EK1.2 for procedures.

1/1 000008 (APE 8)

Pressurizer Vapor Space Accident / 3 RO question 2 AK2.09 is not listed in rev 2 of K/A catalog. Randomly resampled and selected AK2.03 Controllers and positioners 1/1 000015 (APE 15)

Reactor Coolant Pump Malfunctions

/ 4 RO question 5 AA2.14 is not listed in rev 2 of K/A catalog. Randomly resampled and selected AA2.01 Cause of RCP failure.

1/1 1/1 000029 (EPE 29)

Anticipated Transient Without Scram / 1 RO question 8 0029 EA1.12 RO question 8 EA1.17 not listed in rev 2 of K/A catalog. Randomly resampled and selected EA1.03 Charging pump suction valves from VCT operating switch.

029 EA1.03 rejected. Could not find a relationship between an ATWS and the Charging Pump suction from the VCT operating switch for W3. Randomly selected another EA1 item for 029.

Randomly selected 029 EA1.12.

1/1 000038 (EPE 38)

Steam Generator Tube Rupture / 3 RO question 9 EK3.11 not listed in rev 2 of K/A catalog. Randomly resampled and selected EK3.09 Criteria for securing/throttling ECCS 1/1 000040 (APE 40; BW E05; CE E05; W E12) Steam Line Rupture Excessive Heat Transfer / 4 RO question 10 0040 CE05 RO question 10 AK2.17 is a Westinghouse APE K/A. Resampled and replaced with CE05 EK2.2.

CE05 EK2.2 rejected. Could not develop a question referring to emergency coolant and heat removal systems that did not overlap a question presented in the last 2 exams. See 2018 RO11, 2017 RO11, 2017 RO27, 2017 RO33 and RO54 of the 2020 exam. Randomly selected another K/A from CE05 EK2.2 field.

The only other number in this field is CE05 EK2.1.

1/1 000054 (APE 54; CE E06) Loss of Main Feedwater /4 RO question 11 AK1.04 is a Westinghouse APE K/A. Resampled and replaced with CE06 EK1.2.

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 1/1 000055 (EPE 55)

Station Blackout / 6 RO question 12 EK2.09 is not listed in rev 2 of the K/A catalog. No EK 2 K/As have an IR of 2.5 or higher. Randomly resampled and selected EK1.01.

1/1 062 2.2.39 RO question 16 062 2.2.17 rejected. K/A 2.2.17 was already selected for RO69 of this exam. A question is already written for RO69 and finding another question in the same risk assessment procedure (OI-037-000) on a RO level was not possible. Randomly selected another generic K/A from section 2.0. Randomly selected 2.2.39.

1/2 000003 (APE 3)

Dropped Control Rod / 1 RO question 19 AA2.11 is not listed in rev 2 of the K/A catalog. Randomly resampled and selected AA2.03.

1/2 005 2.4.6 question 20 005 2.4.26 rejected. There is no tie at W3 for a stuck/inoperable control to knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage. Randomly selected another 2.4 generic item for 005.

Randomly selected 005 2.4.6.

1/2 000024 (APE 24)

Emergency Boration / 1 RO question 21 AK2.07 is not listed in rev 2 of the K/A catalog. Randomly resampled and selected AK2.04 Pumps.

1/2 036 2.2.44 RO question 22 036 2.2.3 rejected. This K/A is specifically for a multi-unit license. Waterford 3 Is not a multi-unit license. Randomly selected another 2.2 generic item for 036. Randomly selected 036 2.2.44.

1/2 1/2 000068 (APE 68; BW A06) Control Room Evacuation /

8 RO question 25 068 AA1.12 RO question 25 AA1.33 is not listed in rev 2 of the K/A catalog. Randomly resampled and selected AA1.14 Reactor Trip Breakers and Switches 068 AA1.14 rejected. No direction is provided in the W3 control room evacuation procedure, OP-901-502, for operation of reactor trip breakers or switches outside of tripping the reactor before leaving the control room. This is only one step and an RO question could not be developed from it. Randomly selected another AA1 item for 068. Randomly selected 068 AA1.12.

1/2 (CE A16) Excess RCS Leakage / 2 RO question 26 AK2.8 is not listed in rev 2 of the K/A catalog. Randomly resampled and selected AK2.1 1/2 (CE E09)

Functional Recovery RO question 27 EK3.30 is not listed in rev 2 of the K/A catalog. Randomly resampled and selected EK3.1

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 2/1 005 (SF4P RHR)

Residual Heat Removal RO question 30 K3.02 was less than 2.5 IR on rev 2 of K/A catalog. Randomly selected new K/A K3.01 RCS.

2/1 005 (SF4P RHR)

Residual Heat Removal RO question 31 K6.11 was less than 2.5 IR on rev 2 of K/A catalog. Randomly resampled and replaced with K/A K6.03 RHR heat exchanger.

2/1 008 (SF8 CCW)

Component Cooling Water RO question 34 None of the K6 topics were 2.5 or higher IR on rev 2 of k/a catalog. Randomly resampled and replaced with K4.02 Operation of the surge tank, including the associated valves and controls.

2/1 010 (SF3 PZR PCS) Pressurizer Pressure Control RO question 35 A1.10 PZR liquid temperature not listed in rev 2. Randomly resampled and replaced with A1.06 RCS heatup and cooldown effect on pressure.

2/1 039 (SF4S MSS)

Main and Reheat Steam RO question 41 A4.05 MSR startup has an IR of 1.8 on rev 2, randomly resampled and replaced with A4.04 Emergency feedwater pump turbines 2/1 2/1 039 (SF4S MSS)

Main and Reheat Steam RO question 42 039 A1.05 RO Question 42 A1.07 only has IR of 2.4 on rev 2, randomly resampled and replaced with A1.06 Main steam pressure 039 A1.06 Main Steam Pressure rejected. Could not develop an RO question to predict or monitor Main Steam pressure due to overlap with this exam and the previous two exams. See RO61 on this exam and 2017 RO43. Randomly selected another K/A from system 039 in the A1 field. Randomly selected 039 A1.05.

2/1 059 (SF4S MFW)

Main Feedwater RO question 43 A3.09 MFW pump trips is not listed in rev 2. Randomly resampled and replaced with A3.04 Turbine driven feed pump.

2/1 061 (SF4S AFW)

Auxiliary/Emergenc y Feedwater RO question 44 A3.03 Automatic AFW S/G level control should have been selected from an A2 K/A. Replaced with A2.05 Automatic control malfunction 2/1 061 (SF4S AFW)

Auxiliary/Emergenc y Feedwater RO question 45 K5.07 Back leakage through discharge check valves is not listed in rev 2; randomly resampled and replaced with K5.05 Feed line voiding and water hammer.

2/1 062 (SF6 ED AC)

AC Electrical Distribution RO question 46 A1.07 Inverter outputs has IR rating of 2.4 in rev 2; randomly resampled and replaced with A1.03 Effect on instrumentation and controls of switching power supplies.

2/1 063 (SF6 ED DC)

DC Electrical Distribution RO question 47 All of the K6 topics have less than 2.5 IR in rev 2, randomly resampled and replaced with K/A K4.02 Breaker interlocks, permissives, bypasses and cross-ties.

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 2/1 064 (SF6 EDG)

Emergency Diesel Generator RO question 48 All the K5 topics are less than 2.5 IR in rev 2, randomly resampled and replaced with K/A K6.07 Air receivers.

2/1 076 K3.02 RO question 51 076 K3.09 has a K/A rating of 1.9. Must be 2.5 or higher.

Resampled K3 K/As for system 076. Randomly selected 076 A3.02 Secondary closed cooling water.

2/1 078 K2.01 question 52 078 K2.02 rejected. Waterford 3 does not have an Emergency Air Compressor. Waterford 3 has 2 IA compressors and 3 SA compressors. Randomly selected another K/A for system 078 in the K2 section.

2/1 078 (SF8 IAS)

Instrument Air RO question 53 A2.02 is not listed in rev 2, no other A2 K/As are 2.5 or higher in IRs. Randomly resampled and replaced with A3.01 Air Pressure.

2/1 103 (SF5 CNT)

Containment RO question 54 K1.10 CSS is not listed in rev 2. Randomly resampled and replaced with K1.01 CCS.

2/1 013 2.4.11 RO question 55 053 2.4.23 rejected. Waterford 3 does not have an Integrated Control System. Randomly selected another Tier 2/ Group 1 system that did not already have two questions selected to it.

All other systems were used at least once. Randomly selected another 2.4 Generic Item. Randomly selected 013 2.4.11.

2/2 002 (SF2; SF4P RCS) Reactor Coolant RO question 56 K3.04 RMS is not listed in rev 2. Randomly resampled and replaced with K3.03 Containment.

2/2 016 (SF7 NNI)

Nonnuclear Instrumentation RO question 57 K3.05 Condensate is less than 2.5 IR in rev 2. Randomly resampled and replaced with K3.04 MFW system.

2/2 028 (SF5 HRPS)

Hydrogen Recombiner and Purge Control RO question 58 No K4 K/As are listed in rev 2. Randomly resampled and replaced with K2.01 Hydrogen recombiners.

2/2 035 (SF 4P SG)

Steam Generator RO question 59 K5.06 S/G tube leakage detection not listed in rev 2. Randomly resampled and replaced with K5.01 Effect of secondary parameters, pressure, and temperature on reactivity.

2/2 041 (SF4S SDS)

Steam Dump/Turbine Bypass Control RO question 60 K6.13 IAS not listed in rev 2. Randomly resampled and replaced with K6.03 Controller and positioners, including ICS, S/G, CRDS.

2/2 045 (SF 4S MTG)

Main Turbine Generator RO question 61 A1.07 Lights and Alarms is not listed in rev 2. Randomly resampled and replaced with A1.06 Expected response of secondary plant parameters following T/G trip.

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 2/2 055 (SF4S CARS)

Condenser Air Removal RO question 62 015 A3.03 RO question 62 A2.01 Loss of circulating/cooling water system is less than 2.5 IR in rev 2. All A2 K/As are less than 2.5 IR. Randomly resampled to K4 - all K4s less than 2.5 IR in rev 2. Randomly resampled again to K6 - all K6s less than 2.5 IR in rev 2. Randomly resampled again to A3: A3 Ability to monitor automatic operation of the CARS, including: A3.03 Automatic diversion of CARS exhaust.

055 A3.03 was rejected. Waterford 3 does not have an automatic diversion of CARS exhaust. There were no other K/As in the A3 field for CARS to randomly select that were greater than 2.5. Selected another system in Tier 2/Group 2.

Randomly selected 015 Nuclear Instrument System. Randomly selected a K/A in the A3 field for system 015. Randomly selected A3.03 Verification of proper functioning/operability.

2/2 056 (SF4S CDS)

Condensate RO question 63 A3.07 has an IR of 1.7 in rev 2. All A3 K/As have IRs less than 2.5. Randomly resampled to A2.04 Loss of condensate pumps.

2/2 011 K1.05 question 65 086 K1.03 rejected. There is not enough written guidance in W3 procedures that reference fire protection and its association with the EFW system that would allow for an RO question to be written. Randomly selected a different Tier 2 Group 2 system (011 PLC) and then randomly selected a K/A number from the K1 field. Randomly selected 011 K1.05.

1/1 00022 AA2.02 question 76 (SRO1) 00022 AA2.03 rejected. W3 does not have any flow control valves or controllers in the charging system (only solenoid operated isolation valves). Randomly selected another K/A for system 00022 in the AA2 section. Randomly selected 00022 AA2.02.

1/1 00027 AA2.02 question 78 (SRO3) 00027 AA2.01 rejected. This K/A is very similar to SRO question 16. Also, the W3 Pressurizer Pressure Control System has no features or failures that will result in an increase in Pressurizer level. Randomly selected another K/A for system 00027 in the AA2 section. Randomly selected 00027 AA2.02.

1/2 000065 (APE 65)

Loss of Instrument Air / 8 SRO question 80 058 AA2.03 SRO Question 80 AA2.09 not listed in rev 2. Randomly resampled and replaced with AA2.01 Cause and effect of low-pressure instrument air alarm.

065 AA2.01 rejected. 065 A2.01 is already a selected K/A for the 2020 RO Exam. See RO17 on the 2020 Exam. There is also an abundance of low instrument air K/As on the previous two exams and it was not possible to create a new question here that did not overlap in part any of those questions. Randomly selected another system in Tier 1/Group 1 and another K/A in the A2 field. (Randomly selected AA2.02 which is already used on the RO exam) So, randomly selected again and selected 058 AA2.01.

1/2 00055 2.1.6 SRO Question 81 077 2.1.15 rejected. Could not develop a question that did not overlap RO18 on the 2020 Exam or SRO81 on the 2018 exam.

Randomly selected another Tier 1/Group 1 system and another Generic K/A in the 2.1 field. Randomly selected 075 G2.1.6.

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 1/2 028 AA2.11 SRO question 83 033 AA2.07 rejected. Intermediate nuclear instrumentation at Waterford 3 is called the control channels. These channels only go to Reactor Regulating System and Steam Bypass Control. The failure of this channel is not covered in any Waterford 3 procedure. Randomly selected another system from Tier 1 Group 2 and another K/A for that system in the A2 field. Randomly selected 028 A2.11.

1/2 059 2.3.6 SRO question 84 059 2.3.14 rejected. There is no guidance in OP-007-004, Liquid Waste Management, OP-901-412, Liquid Waste Discharge High Radiation or any other W3 operations procedure pertaining to radiation or contamination hazards while discharging a Waste Condensate Tank or Boric Acid Tank. Randomly selected another section 3 Generic K/A while for 059 Accidental Liquid Rad Waste Release. Randomly selected 059 2.3.6.

1/2 (CE E09)

Functional Recovery / 4 SRO question 85 This K/A not listed in rev 2. Randomly resampled to CE E09 Functional Recovery, K/A EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

2/1 004 (SF1; SF2 CVCS) SRO question 86 004 A2.28 rejected. Could not develop an SRO question for depressurizing the RCS when it is hot. This would occur when a pressurizer spray valve fails open are depressurizing the RCS following a SGTR. Neither are related to the CVCS System in a way that a question could be developed. Randomly selected another K/A for CVC system in the A2 field. Randomly selected A2.18 High VCT level.

2/1 2/1 007 (SF5 PRTS)

Pressurizer Relief/Quench Tank SRO question 87 008 (SF8 CCW)

Component Cooling Water SRO question 87 Generic Knowledge K/a 2.1.47 is not listed in rev 2 of the K/A catalog. Randomly resampled and replaced with 2.1.44.

007 2.1.44 rejected. Could not develop an SRO question that relates PRTS to RO duties during Fuel Handling. The PRTS has no interrelation with any fuel handling activities at W3.

Randomly selected another system in Tier 2/ Group 1 that was not previously used. Then randomly selected another Generic knowledge K/A in group 1. Randomly selected 008 2.1.9.

2/2 033 (SF8 SFPCS)

Spent Fuel Pool Cooling SRO question 92 029 2.4.39 rejected. Could not develop an SRO question relating to RO duties in the Emergency Plan procedures and Containment Purge. RO duties in the Emergency Plan procedures are minimal and there is no duties associated with Containment Purge in E-Plan procedures. Randomly selected another Tier2/Group 2 system not previously used and then randomly selected another Generic Knowledge K/A in group 4 not previously used. Randomly selected 033 2.4.8.

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11

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

Waterford 3 Date of Examination:

08/17/2020 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 N,R 2.1.20 Ability to interpret and execute procedure steps.

Complete a calculation for projecting the final RCS boron concentration due to draining a Steam Generator to the RCS in accordance with OP-902-009 Appendix 39, SG Backflow Log.

A2 Conduct of Operations K/A Importance: 3.9 M,R 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, etc.

Determine Spent Fuel Pool (SFP) level by alternate monitoring and calculate time for level to reach the top of the fuel assemblies upon a loss of all cooling to the SFP per OP-901-513, SFP Cooling Malfunction.

A3 Equipment Control K/A Importance: 3.7 P,R 2.2.12, Knowledge of Surveillance Procedures.

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

A4 Radiation Control K/A Importance: 3.2 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. The stay time will be based on the W3 administrative yearly dose limit (2000 mr/year).

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 from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

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

Waterford 3 Date of Examination:

8/17/2020 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: 4.6 D, R 2.1.20, Ability to interpret and execute procedure steps Review a completed Containment Pressure calculation in accordance with OP-903-001, Technical Specification Surveillance Logs, Attachment 11.5, Containment Pressure Calculation.

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.

Perform SM/CRS review of OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction, Attachments 1, 2 and 3 following a PMC failure.

A7 Equipment Control K/A Importance: 4.1 P, R 2.2.12, Knowledge of Surveillance Procedures Review Keff Calculation in accordance with OP-903-090, Shutdown Margin, Section 7.5, Keff Calculation.

Applicant determines Keff does not meet Tech Spec 3.1.2.9 requirements and identifies required corrective actions.

A8 Radiation Control K/A Importance: 3.7 M, R 2.3.4, Knowledge of radiation exposure limits under normal or emergency conditions.

Authorize Emergency Exposure as the Emergency Director in accordance with EP-002-030, Emergency Radiation Exposure Guidelines and Controls.

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).

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

• Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

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

Waterford 3 Date of Examination:

8/17/2020 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 Align Charging to HPSI Header A in accordance with OP-901-112, Charging or Letdown Malfunction A4.08 Charging RO-3.8, SRO-3.4 D, S 1

S3 005 Shutdown Cooling System Place Shutdown Cooling Train A to service in accordance with OP-009-005, Shutdown Cooling.

Fault: When 4100 gpm SDC flow is verified through SI-135, an Inadvertent RAS occurs that will secure the running LPSI pumps. The applicant will be required to restart LPSI Pump A.

005 A4.01 Controls and indication for RHR Pumps RO - 3.6, SRO - 3.4 A, EN, N, L, S

4P S8 034 Fuel Handling Equipment Place the FHB Emergency Filtration Unit in service in accordance with OP-002-009, Fuel Handling Building HVAC 034 A4.01 Radiation levels RO - 3.3, SRO - 3.7 D, S 8

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:

8/17/2020 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.

Fault: The RAS will not actuate using the manual pushbuttons and will actuate by opening the breakers.

(From 2018 NRC Exam)

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

P3 064 Emergency Diesel Generator (ED/G) System; Trip Emergency Diesel Generator B locally.

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

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

RO - 3.9, SRO - 4.2 A, D, E, R 6

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, 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

9/ 8/ 4 4 1/ 1/ 1 2 1/ 1/ 1 (control room system) 1 1/ 1/ 1 2 2/ 2/ 1 1 3/ 3/ 2 (randomly selected) 1 1/ 1/ 1 1 3

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

Waterford 3 Date of Examination:

8/17/2020 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 Align Charging to HPSI Header A in accordance with OP-901-112, Charging or Letdown Malfunction A4.08 Charging RO-3.8, SRO-3.4 D, S 1

S2 006 Emergency Core Cooling System Mitigate the release of radioactivity through RWSP following RAS Actuation in accordance with OP-902-002, Loss of Coolant accident Recovery.

Fault: LPSI pump continues to run after RAS actuates, requiring the applicant to manually stop the running LPSI pump after additional valve manipulations.

011 EA1.12 Long term containment of radioactivity RO-4.1, SRO-4.4 A, D, L, S 2

S3 005 Shutdown Cooling System Place Shutdown Cooling Train A to service in accordance with OP-009-005, Shutdown Cooling.

Fault: When 4100 gpm SDC flow is verified through SI-135, an Inadvertent RAS occurs that will secure the running LPSI pumps. The applicant will be required to restart LPSI Pump A.

005 A4.01 Controls and indications for RHR Pumps RO - 3.6, SRO - 3.4 A, EN, N, L, S

4P

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

Waterford 3 Date of Examination:

8/17/2020 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 0BS4 039 Main and Reheat Steam System; BOP operator immediate operator actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation.

Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

039 A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 A, D, S 4S S5 026 Containment Spray System Reset CSAS in accordance with OP-902-009, Standard Appendices, Section 5 - E.

026 A4.01 CSS Controls RO-4.5, SRO-4.3 D, EN, L, S 5

S6 062 A.C. Electrical Distribution 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.

062 A4.01 All breakers RO-3.3, SRO-3.1 D, P, S 6

S7 S8 034 Fuel Handling Equipment Place the FHB Emergency Filtration Unit in service in accordance with OP-002-009, Fuel Handling Building HVAC A4.01 Radiation levels RO - 3.3, SRO - 3.7 D, S 8

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:

8/17/2020 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.

Fault: The RAS will not actuate using the manual pushbuttons and will actuate by opening the breakers.

(From 2018 NRC Exam)

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

P2 076 Service Water System Align Potable Water to Instrument Air Compressors in accordance with OP-902-009, EOP Standard Appendices, App. 18.

A4.04 Emergency Heat Loads RO-3.5, SRO-3.5 L, N 4S P3 064 Emergency Diesel Generator (ED/G) System; Trip Emergency Diesel Generator B locally.

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

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

RO - 3.9, SRO - 4.2 A, D, E, R 6

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U

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

Waterford 3 Date of Examination:

8/17/2020 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 (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/ 8/ 4 8 1/ 1/ 1 2 1/ 1/ 1 (control room system) 2 1/ 1/ 1 5 2/ 2/ 1 2 3/ 3/ 2 (randomly selected) 2 1/ 1/ 1 1 7

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

Waterford 3 Date of Examination:

08/17/2020 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 Align Charging to HPSI Header A in accordance with OP-901-112, Charging or Letdown Malfunction 004 A4.08 Charging RO-3.8, SRO-3.4 D, S 1

S2 006 Emergency Core Cooling System Mitigate the release of radioactivity through RWSP following RAS Actuation in accordance with OP-902-002, Loss of Coolant accident Recovery.

Fault: LPSI pump continues to run after RAS actuates, requiring the applicant to manually stop the running LPSI pump after additional valve manipulations.

011 EA1.12 Long term containment of radioactivity RO-4.1, SRO-4.4 A, D, L, S 2

S3 005 Shutdown Cooling System Place Shutdown Cooling Train B to service in accordance with OP-009-005, Shutdown Cooling.

Fault: When 4100 gpm SDC flow is verified through SI-135, an Inadvertent RAS occurs that will secure the running LPSI pumps. The applicant will be required to restart LPSI Pump A.

005 A4.01 Controls and indications for RHR Pumps RO - 3.6, SRO - 3.4 A, EN, N, L, S

4P

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

Waterford 3 Date of Examination:

08/17/2020 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 0BS4 039 Main and Reheat Steam System; BOP operator immediate operator actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation.

Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

039 A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 A, D, S 4S S5 026 Containment Spray System Reset CSAS in accordance with OP-902-009, Standard Appendices, Section 5 - E.

026 A4.01 CSS Controls RO - 4.5, SRO - 4.3 D, EN, L, S 5

S6 062 A.C. Electrical Distribution 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.

(From 2018 NRC Exam) 062 A4.01 All breakers RO-3.3, SRO-3.1 D, P, S 6

S7 015 Nuclear Instrumentation System Perform Range Check functional test of startup Channels in accordance with OP-903-101, Startup Channel Functional Test 015 A3.03, Verification of proper functioning/operability RO - 3.9, SRO - 3.9 L, D, S 7

S8 034 Fuel Handling Equipment System Place the FHB Emergency Filtration Unit A in service in accordance with OP-002-009, Fuel Handling Building HVAC 034 A4.01 Radiation levels RO - 3.3, SRO - 3.7 D, S 8

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:

08/17/2020 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.

Fault: The RAS will not actuate using the manual pushbuttons and will actuate by opening the breakers.

(From 2018 NRC Exam)

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

P2 076 Service Water System Align Potable Water to Instrument Air Compressors in accordance with OP-902-009, EOP Standard Appendices, App. 18.

A4.04 Emergency Heat Loads RO-3.5, SRO-3.5 L, N 4S P3 064 Emergency Diesel Generator (ED/G) System; Trip Emergency Diesel Generator B locally.

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

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

RO - 3.9, SRO - 4.2 A, D, E, R 6

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U

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

Waterford 3 Date of Examination:

08/17/2020 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 (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/ 8/ 4 9 1/ 1/ 1 2 1/ 1/ 1 (control room system) 2 1/ 1/ 1 6 2/ 2/ 1 2 3/ 3/ 2 (randomly selected) 2 1/ 1/ 1 1 8

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Waterford 3 Date of Exam:

8/17/2020 Operating Test No.: 1 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO SRO-I SRO-U RX 1

1 1 0 NOR 1

1 1 1 I/C 2,3,4,5 6,8 3,4,6 2,3, 5,8 4

4 2 MAJ 7

7 7

2 2 1 TS 2,4 0

2 2 RO SRO-I SRO-U RX 4

1 1 0 NOR 4

4 1

1 1 I/C 1,2,5 7,8 2,7 1,5,7, 8

4 4 2 MAJ 6

6 6

2 2 1 TS 1,3 0

2 2 RO SRO-I SRO-U RX 4

1 1 0 NOR 1

1 1 I/C 1,2,3 4,7 1,5 2,3,4

,7 4

4 2 MAJ 6

6 6

2 2 1 TS 1,2,3 0

2 2 RO SRO-I SRO-U RX 4

1 1 0 NOR 1

1 1 I/C 1,2,3

,4,5 2,3 1,4,5

,7 4

4 2 MAJ 6

6 6

2 2 1 TS 1,4 0

2 2

ES-301 Transient and Event Checklist Form ES-301-5 Instructions:

1.

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one-for-one basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

Appendix D Scenario Outline Form ES-D-1 2020 NRC Exam Scenario 1 D-1 Rev 0 Facility:

Waterford 3 Scenario No.:

1 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

Mode 2, Reactor Power ~1%. Two Charging Pumps in operation. AB Buses are aligned to Train B.

Turnover:

Protected Train is B. Dilute to 5-10% power.

Critical Tasks:

(1) Establish Reactivity Control (2) Establish Containment Temperature and Pressure Control (3) Trip any RCP exceeding operating limits Event No.

Malf.

No.

Event Type*

Event Description 1

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

2 RC20B I-BOP I-SRO TS-SRO RCS Loop 1 Hot leg transmitter failure high (RC-ITI-102HB).

3 CV35A CVR101 C - ATC C - BOP C - SRO During dilution, PMU counter fails to secure flow.

4 RC15A2 I-ATC I-SRO TS-SRO Pressurizer Level Control Channel Level Transmitter, RC-ILT-0110X, fails low.

5 RC08B C-BOP C-SRO Reactor Coolant Pump 1B Lower Seal fails.

6 RC03B RP02A RP02B RP02C RP02D C-ATC C-SRO RCP 1B Trip with no automatic Reactor Trip (Critical Task 1, Establish Reactivity Control) 7 MS11A M-All Excess Steam Demand Event will occur on #1 Steam Generator inside Containment.

8 RP05A3 RP05B3 RP05C3 RP05D3 C-BOP C-SRO No automatic Containment Spray Actuation Signal (CSAS)

(Critical Task 2, Establish Containment Temperature and Pressure Control) (Critical Task 3, Trip any RCP exceeding Operating limits)

(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 3,4,5 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 2020 NRC Exam Scenario 1 D-1 Rev 0 The crew assumes the shift with the reactor at 1% power following a forced outage. The turnover will include instructions to perform RCS dilution to 5 - 10% power.

Event 1: The reactivity plan will include instructions to dilute in multiple PMU batches. The initial batch will be 100 gallons of PMU. Each subsequent batch will be 50 gallons of PMU. This will allow for an observable power rise without concern for a reactor trip on the PMU failure.

Event 2: After the first 100 gallons of PMU are added, RCS THOT instrument RC-ITI-102HB fails high.

The SRO should review and enter Technical Specification 3.3.1 action 2 and bypass Hi LPD and Lo DNBR bistables (3 & 4) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with OP-009-007, Plant Protection System. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks.

Tech Spec 3.3.3.5 and 3.3.3.6 should not be applicable.

Event 3: During the second dilution, the Primary Water counter will fail to secure dilution. The ATC should attempt to secure Primary Water Flow by operating PMU-144 and CVC-510. Neither of these actions will secure flow. The CRS should enter OP-901-104, Inadvertent Positive Reactivity Addition, and secure Primary Makeup Pump A.

Event 4: After the crew has stopped the inadvertent dilution, pressurizer Level Control Channel Level Transmitter, RC-ILT-0110X, fails low. The SRO should enter OP-901-110, Pressurizer Level Control Malfunction and implement Section E1. The crew should take manual control of the Pressurizer Level Controller and/or operate Charging Pumps to restore Pressurizer level, swap control to the Channel Y level channel, and return the Pressurizer Level Controller back to AUTO. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks. Tech Spec 3.3.3.5 action a should be entered with the action to restore the inoperable channel within 7 days. The SRO should determine that TS 3.3.3.6 requirements are met after verifying QSPDS value for PZR level is within the channel check limit of 5%.

Event 5: Reactor Coolant Pump 1B Lower Seal fails. The crew should enter OP-901-130, Reactor Coolant Pump Malfunction and implement Section E1, Seal Failure. The SRO should direct the BOP to lower Component Cooling Water Temperature by operating Dry Cooling Tower fans or by adjusting Auxiliary Component Cooling Water flow.

Event 6: After the crew is in Section E1 of OP-901-130 AND the BOP has adjusted Component Cooling Water Temperature, RCP 1B will trip and the reactor will not automatically trip. The ATC will manually trip the reactor from the RTGB (CRITICAL TASK 1). The crew will perform Standard Post Trip Actions using OP-902-000, SPTAs and diagnose to OP-902-001, Reactor Trip Recovery.

Event 7: After the crew has completed the Maintenance of Vital Auxiliaries Safety Function in OP-902-000, Standard Post Trip Actions, an Excess Steam Demand Event will occur on Steam generator #1 inside containment. This will result in containment pressure exceeding 17.7 psia (the setpoint for automatic Containment Spray Actuation).

Event 8: The CSAS will not occur and the crew will be required to manually initiate containment spray (CRITICAL TASK 2). The actuation of CSAS will cause CCW to isolate to the RCPs and the crew will need to stop any running RCP prior to exceeding 3 minutes without CCW flow (CRITICAL TASK 3). The crew will diagnose to OP-902-004 and stabilize RCS temperature and pressure using the least affected SG and auxiliary spray valves.

The scenario can be terminated after the crew has stabilized RCS temperature and pressure or at the lead examiners discretion.

NRC Scenario 1 - Critical Task Determination 2020 NRC Exam Scenario 1 D-1 Rev 0 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: Establish Reactivity Control This task is satisfied by manually tripping the Reactor by using the manual trip pushbuttons, Diverse Reactor trip pushbuttons, or by de-energizing busses 32A and 32B within 1 minute of exceeding a Plant Protection System (PPS) limit. This task becomes applicable following the trip of RCP 1B.

Failure to trip the Reactor when an automatic PPS signal has failed to actuate can lead to a degradation of fission product barriers. 1 minute is determined to be a reasonable time limit to identify and take action for satisfactory performance. OPS management standard documented in TM-OP-100-03.

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

RCP off light illuminated Trips and pre-trips on SG lo flow on CP-7 All CEA rod bottom lights extinguished Procedurally driven from OP-902-000 step 1.a.1.1)

Depresses two Reactor Trip pushbuttons on CP-2 or CP-8 Reactor Trip breakers open All CEA rod bottom lights illuminated Reactor power lowering CT-2: Establish Containment Temperature and Pressure Control This task is satisfied by manually initiating CSAS, or manually aligning CS Pumps and valves to establish at least 1 train of Containment Spray prior to exiting OP-902-000, Standard Post Trip Actions. This action satisfies the Containment Temperature and Pressure Control Safety Function in OP-902-000. This task becomes applicable once 17.7 PSIA has been exceeded inside containment following the ESDE.

Failure to take action to establish Containment Temperature and Pressure Control would lead to a degradation of a fission product barrier. OPS management standard documented in TM-OP-100-03.

(TM-OP-100-03, CT-15)

Containment pressure >

17.7 psia CS Pumps stop light illuminated CS-125 indicates closed CS Header flow not indicated Procedurally driven from OP-902-000 step 9.3 Depresses 2 CSAS pushbuttons on CP-7 or Starts CS Pump and open associated CS-125 valve using control switch CSAS annunciators actuated CS Pumps run light illuminated CS-125 indicates open CS Header flow indicated

NRC Scenario 1 - Critical Task Determination 2020 NRC Exam Scenario 1 D-1 Rev 0 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-3: Trip any RCP exceeding operating limits This task is satisfied by manually stopping Reactor Coolant Pumps within 3 minutes of a loss of CCW flow to the RCPs. This task becomes applicable following the actuation of CSAS.

Based on EOP required actions for an Excess Steam Demand Event. The RCPs are stopped to prevent RCP seal damage. 3 minutes is the analyzed time for a RCP to run without CCW cooling to the RCP seal.

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 Critical Task (NUREG-1021, Rev. 11 Appendix D)

• If an operator or the crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

• 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 2020 NRC Exam Scenario 1 D-1 Rev 0 REFERENCES Event Procedures 1

OP-010-003, Plant Startup, Rev. 351 OP-002-005, Chemical and Volume Control, Rev. 64 2

OP-009-007, Plant Protection System, Rev. 20 OP-903-013, Monthly Channel Checks, Rev. 19 Technical Specifications 3.3.1, 3.3.3.5, 3.3.3.6 3

OP-901-104, Inadvertent Positive Reactivity Addition, Rev. 304 4

OP-901-110, Pressurizer Level Control Malfunction, Rev. 11 Technical Specifications 3.3.3.5, 3.3.3.6 5

OP-901-130, Reactor Coolant Pump Malfunction, Rev. 11 6

OP-902-000, Standard Post Trip Actions, Rev. 16 OP-902-009, Standard Appendices, Rev. 319 7/8 OP-902-004, Excess Steam Demand Event Recovery, Rev. 17 OP-902-009, Standard Appendices, Rev. 319 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 OP-100-017, EOP Implementation Guide, Rev 5 TM-OP-100-03, Simulator Training, Rev. 20

Appendix D Scenario Outline Form ES-D-1 2020 NRC Exam Scenario 2 D-1 Rev 1 Facility:

Waterford 3 Scenario No.:

2 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

Reactor power is 100%. AB Buses are aligned to Train B.

Turnover:

Protected Train is B; Maintain 100%. CS Pump A (TS) is out of service.

Critical Tasks:

(1) Trip any RCP exceeding operating limits (2) Critical Task, align LPSI pump to replace CS pump Event No.

Malf.

No.

Event Type*

Event Description 1

DI-18A4s27-1 N - BOP N - SRO TS - SRO During performance of OP-903-052, CVAS Fan A will fail to start.

2 CH08E1 C - BOP C - SRO TS - SRO Plant Protection Sys Ch. D, Cont. Press (CIAS), CB-IPI-6701SMD, fails high.

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

4 FW35B R - ATC C - BOP C - SRO Feedwater Heater 5B tube leak, Rapid Plant Power Reduction 5

TP01A TP08B C - BOP C - SRO Turbine Cooling Water Pump A trips 6

RC23A M - All Loss of Coolant Accident will occur. (Critical Task 1, Trip any RCP exceeding operating limits) 7 RP08C I - ATC I - BOP I - SRO Relay K202A fails, CVC-401, CVC-109, IA-909, and FP-601A fail to close automatically 8

DI-08A04S22-1 CS01B C - BOP C - SRO Containment Spray Pump B trips and cannot be restarted requiring entry into OP-902-008, Functional Recovery and action taken to align Low Pressure Safety Injection pump to provide Containment Spray (Critical Task 2, align LPSI pump to replace CS pump prior to exiting Appendix 28 of OP-902-009)

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

Appendix D Scenario Outline Form ES-D-1 2020 NRC Exam Scenario 2 D-1 Rev 1 Scenario Quantitative Attributes

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

2. Abnormal events (2-4) 3

3. Major transients (1-2) 1

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

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

6. Preidentified critical tasks (> 2) 2

NRC Scenario 2 - Narrative 2020 NRC Exam Scenario 2 D-1 Rev 1 The crew assumes the shift at 100% power with instructions to maintain 100% power. Containment Spray Pump A is out of service.

Event 1: Last shift, it was discovered that OP-903-052, CVAS Operability Test, will exceed its Tech Spec late date this shift. The crew will be directed to start CVAS Train A in accordance with OP-903-052. This surveillance will have the BOP operator secure RAB Normal Supply and Normal Exhaust Fans A and start CVAS Fan A. After securing both normal ventilation fans, CVAS Fan A will fail to start. This will require entering Tech Spec 3.7.7, a 7 day action requirement. RAB Normal Supply and Normal Exhaust Fans A will have to be re-started.

Event 2: After RAB Normal Ventilation is running and Tech Specs have been addressed, CB-IPI-6701SMD, Containment Pressure (CIAS) fails high. The SRO should review Technical Specifications 3.3.1 and 3.3.2. Per Table 3.3-1 under Containment Pressure - High (Functional Unit 6) the SRO should enter Technical Specification 3.3.1 action 2. Per Table 3.3-3 under Functional Units 1b (Safety Injection, Containment Pressure-High), 3b (Containment Isolation, Containment Pressure-High), and 4c (Main Steam Line Isolation, Containment Pressure High) the SRO should enter Tech 3.3.2 action 13. The SRO should direct the BOP to bypass the Containment Pressure High (RPS) and Containment Pressure High (ESF) trip bistables (13&16) in PPS Channel D within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The BOP should bypass the trip bistables in accordance with OP-009-007, Plant Protection System.

Event 3: After the BOP has bypassed bistables at CP-10 and the SRO has addressed Tech Specs, 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 alternate pressurizer pressure channel has been aligned, a tube leak occurs in Feedwater Heater 5B, causing Condensate flow to isolate through Low Pressure Feedwater Heaters 5B and 6B. The crew will enter OP-901-221, Secondary System Transient, Section E1, Loss of Feedwater Preheating. This also requires a power reduction in accordance with OP-901-212, Rapid Plant Power Reduction, which will prompt a reactivity manipulation.

Event 5: The running Turbine Cooling Water Pump A will trip and the standby pump will fail to start automatically. The CRS should direct the BOP to start the standby pump in accordance with OP-901-512, Loss of Turbine Cooling Water and OP-500-005, Control Room Cabinet E.

Event 6: RCS leak occurs on RCS Cold Leg 1A that rapidly progresses to a Large Break Loss of Coolant Accident and reactor trip. 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: Relay K202A will not actuate and CVC-401, CVC-109, IA-909, and FP-601A fail to close automatically. The ATC and BOP should position these valves to ensure Containment Isolation.

Event 8: Once the crew has entered OP-902-002, Containment Spray Pump B will trip and will not be able to be restarted. The crew should determine that Containment isolation and Containment Pressure and Temperature Control Safety Functions are not being met and diagnose into OP-902-008, Functional Recovery. The SRO should prioritize Containment Isolation first due to CS-125B being open and Containment Pressure and Temperature Control second. The crew will perform steps in OP-902-008, Functional Recovery and align the LPSI pump B to replace CS Pump B to establish Containment Temperature and Pressure Control (CRITICAL TASK 2).

The scenario can be terminated after the established Containment Spray flow from the Low Pressure Safety Injection Pump or at the lead examiners discretion.

NRC Scenario 2 - Critical Task Determination 2020 NRC Exam Scenario 2 D-1 Rev 1 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 or a loss of CCW to the RCP seal cooler.

Based on EOP required actions for a Loss of Coolant Accident, the RCPs are stopped for numerous reasons.

The potential for RCP seal damage and pumping more mass out of the break can result in uncovering fuel.

In this case, the RCPs are stopped to prevent more mass being lost out of the RCS which jeopardizes the fuel clad barrier and to prevent RCP seal damage. 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 RCS pressure and temperature on CP-2 CSAS initiated CP-8 Procedurally driven from:

OP-902-000 steps 5.3, 5.4 OP-902-002 steps 8, 9 Stops RCPs using control switch Pump light indication CT-2: Establish Containment Temperature and Pressure Control This task is satisfied by manually aligning an available LPSI pump to replace a CS pump prior to exiting the step to align a LPSI pump to replace a CS pump in Appendix 28.

Failure to take action to establish containment pressure and temperature control may result in containment pressure exceeding its maximum design and therefore exceeding design leakage. This would result in degradation of a fission product barrier.

(TM-OP-100-03, CT-15)

Containment Spray flow on CP_8 CS pump light status on CP-8 Containment pressure OP-902-008, Functional Recovery CPTC actions OP-902-009, Standard Appendices The crew takes actions to manually align an available LPSI pump to replace a CS pump Containment Spray flow indication Critical Task (NUREG-1021, Rev. 11 Appendix D)

• If an operator or the crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

• 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 2020 NRC Exam Scenario 2 D-1 Rev 1 REFERENCES Event Procedures 1

OP-903-052, CVAS Operability Test, Rev 13 Technical Specification 3.7.7 2

OP-009-007, Plant Protection System, Rev. 20 Technical Specification 3.3.1, 3.3.2 3

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

OP-901-221, Secondary System Transient, Rev. 5 OP-901-212, Rapid Plant Power Reduction, Rev. 14 5

OP-901-512, Loss of Turbine Cooling Water, Rev 3 OP-500-005, Control Room Cabinet E, Rev 28 6

OP-902-000, Standard Post Trip Actions, Rev. 16 7

OI-038-000, EOP Operations Expectations/Guidance, Rev 19 8

OP-902-002, Loss of Coolant Accident Recovery, Rev 21 OP-902-008, Function Recovery Procedure, Rev. 30 OP-902-009, Standard Appendices, Rev. 319 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 TM-OP-100-03, Simulator Training, Rev. 21 OP-100-017, Emergency Operating Procedures Implementation Guide, Rev 5

Appendix D Scenario Outline Form ES-D-1 2020 NRC Exam Scenario 3 D-1 Rev 0 0BFacility:

Waterford 3 Scenario No.:

3 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

Reactor power is 100%. AB Buses are aligned to Train B. No equipment OOS Turnover:

Protected Train is B; Maintain 100%.

Critical Tasks:

(1) Trip any RCP exceeding operating limits, (2) Energize at least 1 vital bus Event No.

Malf.

No.

Event Type*

Event Description 1

CV01B C - ATC C - SRO TS - SRO Charging Pump B trips 2

SG04G I - BOP I - SRO TS - SRO Steam Generator 1 Pressure Instrument, SG-IPT-1013C, fails low requiring Technical Specification entry and bypass of multiple Plant Protection System C trip bistables.

3 MS05A1 C - BOP C - SRO TS - SRO Atmospheric Dump Valve on Steam Generator #1 fails open requiring entry into OP-901-221, Secondary System Transient, turbine load reduction and local isolation of the ADV. (TS 3.7.1.7) 4 FW21A FW21AA R-ATC C-BOP C-SRO A lowering of Main Condenser vacuum. Requires a rapid plant downpower IAW OP-901-212. Vacuum continues to lower until a manual or automatic trip.

5 RD11 A07 RD11 A37 RD11 A39 C - ATC 3 Control Element Assemblies fail to insert into the core following the reactor trip, Emergency Boration (Critical Task 1) 6 ED01A ED01B ED01C ED01D M - All Following entry into OP-902-001, a loss of offsite power occurs.

7 EG10B EG13A C-BOP C-SRO EDG B overspeed trip on start EDG A auto voltage regulator fails low, manually raise voltage to allow breaker auto closure (Critical Task 2)

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

Appendix D Scenario Outline Form ES-D-1 2020 NRC Exam Scenario 3 D-1 Rev 0 Scenario Quantitative Attributes

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

2. Abnormal events (2-4) 3

3. Major transients (1-2) 1

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

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

6. Preidentified critical tasks (> 2) 2

NRC Scenario 3 - Narrative 2020 NRC Exam Scenario 3 D-1 Rev 0 The crew assumes the shift at 100% power with instructions to maintain 100% power. No equipment OOS.

Event 1: Charging Pump B trips. Per the Annunciator Response Procedure, 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. The SRO will implement OP-901-112, Charging or Letdown Malfunction, Section E1, Charging Malfunction. If Letdown is isolated, Charging and Letdown will be re-initiated using of OP-901-112. The SRO should review and enter Technical Specification 3.1.2.4.

Technical Specification 3.1.2.4 may be exited after aligning Charging Pump AB to replace Charging Pump B.

Event 2: Steam Generator 1 Pressure Instrument, SG-IPT-1013C, fails low. The SRO should review and enter Technical Specifications 3.3.1 action 2 and 3.3.2 actions 13 and 19. The SRO will direct the BOP to bypass the Steam Generator 1 Pressure Lo, Steam Generator 1 P, and Steam Generator 2 P trip bistables in Plant Protection System Channel C within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with OP-009-007, Plant Protection System. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks, and determine that Technical Specification entry for 3.3.3.5 and 3.3.3.6 is not required.

Event 3: #1 Steam Generator Atmospheric Dump Valve (ADV) fails open and cannot be closed using the CP-8 controller (MS-IPIC-0303-A1). The SRO should enter into procedure OP-901-221, Secondary System Transient and OP-500-012, Control Room Cabinet N, and direct the BOP to reduce Main Turbine load. The crew should direct an NAO to locally isolate or close the ADV. The SRO should enter Technical Specification 3.7.1.7.

Event 4:. A leak in the Main Condenser develops and Main Condenser vacuum begins to drop. The SRO will enter OP-901-220, Loss of Condenser Vacuum. Main Condenser vacuum will drop below 25 inches, requiring a rapid plant power reduction. The SRO will enter OP-901-212, Rapid Plant Power Reduction..

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.

Vacuum will stabilize during the power reduction then the vacuum leak will grow until a manual reactor trip is directed by the CRS or an automatic trip occurs.

Event 5: During the trip 3 CEAs remain fully withdrawn and the ATC will commence emergency boration to the RCS in accordance with OP-902-000, Standard Post Trip Actions (Critical Task 1).

Event 6: After the crew transitions to OP-902-001, Reactor Trip Recovery, a loss of offsite power will occur. 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.

Event 7: Emergency Diesel Generator B will trip immediately on overspeed and Emergency Diesel Generator A will require the BOP to manually raise voltage which will allow its output breaker to automatically close (Critical Task 2)

The scenario can be terminated after the actions are taken to protect the main condensers located in OP-902-003, Loss of Offsite Power / Loss of Forced Circulation

NRC Scenario 3 - Critical Task Determination 2020 NRC Exam Scenario 3 D-1 Rev 0 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 CEA indicates withdrawn on CEAC Procedurally driven from OP-902-000 The crew takes action to manually initiate emergency boration using the gravity drain flowpath.

OP-903-103, Emerg.

Boration Required valve and pump indications and adequate charging flow CT-2: Energize at least 1 vital AC bus This task is satisfied by manually adjusting EDG output voltage to the required band for the output breaker to automatically close prior to performing actions in OP-902-005, Station Blackout Recovery.

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.

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

Breaker indication on CP-1 Control Room Lighting OP-902-000, Standard Post Trip Actions The crew takes actions to manually start an available EDG and/or adjusting EDG voltage to within the required band for output breaker closure.

EDG status and output breaker indication Critical Task (NUREG-1021, Rev. 11 Appendix D)

• If an operator or the crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

• 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.

2020 NRC Exam Scenario 3 D-1 Rev 0 REFERENCES Event Procedures 1

OP-500-007, Control Cabinet G, Rev. 023 OP-901-112, Charging or Letdown Malfunction, Rev 8 Technical Specification 3.1.2.4 2

OP-009-007, Plant Protection System, Rev. 20 OP-903-013, Monthly Channel Checks, Rev. 19 Technical Specification 3.3.1 Technical Specification 3.3.2 3

OP-901-221, Secondary System Transient, Rev. 5 OP-500-012, Control Room Cabinet N, Att. 4.61, Rev 29 Technical Specification 3.7.1.7 4

OP-901-220, Loss of Condenser Vacuum, Rev. 304 OP-901-212, Rapid Plant Power Reduction, Rev. 14 5

OP-902-000, Standard Post Trip Actions, Rev. 16 OP-901-103, Emergency Boration, Rev 4 6

OP-902-003, Loss of Offsite Power / Loss of Forced Circulation, Rev. 11 OP-902-009, Standard Appendices, Rev. 319 7

OP-902-000, Standard Post Trip Actions, Rev. 16 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 OP-100-017, EOP Implementation Guide, Rev 5 TM-OP-100-03, Simulator Training, Rev. 20

Appendix D Scenario Outline Form ES-D-1 2020 NRC Exam Scenario 4 D-1 Rev 1 Facility:

Waterford 3 Scenario No.:

4 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%.

Critical Tasks:

(1) Isolate Most Affected Steam Generator (2) Cool and Depressurize RCS to Prevent Lifting Affected SG Safety Valves Event No.

Malf.

No.

Event Type*

Event Description 1

CC02A C - BOP C - SRO TS - SRO Secure Auxiliary Component Cooling Water Pump A following chemical mixing. Component Cooling Water Header A Controller fails to manual.

2 RC36B I - ATC I - SRO Reactor Coolant pressure instrument RC-IPT-9120 B fails high, requiring removal of Diverse Reactor Trip from service.

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

4 SG01B R-ATC C-BOP C-SRO TS-SRO Steam Generator #2 Tube Leak. Requires a rapid plant power reduction IAW OP-901-212.

5 FW26A I - BOP I - SRO Steam Generator #1 Feed Flow Instrument, FW-IFR-1111, fails low requiring implementation of OP-901-201, Steam Generator Level Control Malfunction.

6 SG01B M - All Steam Generator tube leakage leading to Reactor Trip and Safety Injection (Critical Task 1, Isolate Most Affected Steam Generator)

(Critical Task 2, Cool and Depressurize RCS to Prevent Lifting Affected SG Safety Valves 7

EG05 C - BOP C - SRO Main Generator Exciter Field Breaker fails to trip 8

XXXXX C - BOP C - SRO SIAS push buttons on CP-7 fail to actuate Safety Injection signal.

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

Appendix D Scenario Outline Form ES-D-1 2020 NRC Exam Scenario 4 D-1 Rev 1 Scenario Quantitative Attributes

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

2. Abnormal events (2-4) 3

3. Major transients (1-2) 1

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

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

6. Preidentified critical tasks (> 2) 2

NRC Scenario 4 - Narrative 2020 NRC Exam Scenario 4 D-1 Rev 1 The crew assumes the shift at 100% power with instructions to maintain 100% power. No equipment is out of service.

Event 1: Following crew turnover, the crew is directed to secure Auxiliary Component Cooling Water Pump A following basin chemical mixing in accordance with OP-002-001, Auxiliary Component Cooling Water. After ACCW Pump A is off, controller CC-ITIC-7070A for ACCW temperature control will fail in manual. The SRO should declare ACCW Train A inoperable and enter a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action for Tech Spec 3.7.3 as well as cascading Tech Specs. The SRO should address the need to accomplish surveillance OP-903-066, Electrical Breaker Alignment Checks, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with Tech Spec 3.8.1.1.b.

They must also address the need to accomplish the requirements of Tech Spec 3.8.1.1.d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Event 2: When the Tech Spec review is complete, Reactor Coolant pressure instrument RC-IPT-9120 B will fail high. Based on the direction in the annunciator response procedure, the SRO should direct the ATC to remove Diverse Reactor Trip from service using OP-004-021, Anticipated Transient System.

Event 3: After Diverse Reactor Trip has been removed from service, the in-service Letdown flow control valve, CVC-113A, fails closed. The SRO should enter OP-901-112, Charging or Letdown Malfunction and implement Section E2, Letdown Malfunction, and place the backup flow control valve, CVC-113B, in-service. The SRO may implement EN-OP-200, Transient Response Rules.

Event 4:. After the crew has restored Letdown to automatic with CVC-113B in service, Steam Generator 2 develops a tube leak at ~ 600 gpd. The SRO should implement OP-901-202, Steam Generator Tube Leakage or High Activity. The CRS should determine that based on leak indications, the limits of Technical Specification 3.4.5.2 have been exceeded for Primary-to-Secondary Leakage and enter Tech Spec 3.4.5.2 Action a. The SRO should also determine that the current leakage requires implementation of OP-901-212, Rapid Plant Power Reduction. The crew should commence a rapid power reduction to lower power to < 50% power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Event 5: During the power reduction, after Main Turbine load reduction has commenced, Steam generator #1 Feed Flow Instrument FW-IFR-1111 fails low requiring manual control of Feedwater Control and restoring of Steam Generator water level. The SRO should enter OP-901-201, Steam Generator Level Control Malfunction and evaluate and enter Technical Requirements Manual 3.3.5 action a for the Ultrasonic Flowmeter being inoperable. The BOP will control Steam Generator #1 level in manual for the remainder of the power reduction.

Event 6: Once the crew has taken action to stabilize steam generator level and with the power reduction still in progress, 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 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). The SRO should enter OP-902-007, Steam Generator Tube Rupture Recovery Procedure. The crew should perform a rapid RCS cooldown to < 520 °F and isolate Steam Generator #2.

Event 7: When the reactor is tripped, the Main Generator Exciter Field Breaker will fail to open, requiring the BOP operator open the breaker manually.

Event 8: When the CRS directs the actuation of Safety Injection Actuation Signal, the buttons on CP-7 will fail to initiate the signal. The Operators will have to initiate the SIAS from CP-8.

The scenario can be terminated after the crew has completed the Reactor Coolant System rapid cooldown and isolated Steam Generator 2, and commenced an RCS depressurization or at the lead examiners discretion.

NRC Scenario 4 - Critical Task Determination 2020 NRC Exam Scenario 4 D-1 Rev 1 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: 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)

A ruptured SG presents a flowpath for radioactive release. Isolating the SG will terminate the release lowering the dose to plant personnel and the public.

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 would 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 Places MS-116B setpoint to 980 psig and verifies controller in AUTO Verifies MS-124B is closed Verifies FW-184B is closed IF EFAS-2 is NOT initiated, THEN closes EFW Isolation valves EFW-228B and EFW-229B Places EFW Flow Control valves in MAN and THEN closes EFW-224B and EFW-223B Closes MS-401B Close Main Steam Line 2 Drains MS-120B and MS-119B Closes Steam Generator Blowdown Isolation valves BD-103B and BD-102B Checks SG2 East Side Main Steam Safety valves are closed.

Proper indication for listed components

NRC Scenario 4 - Critical Task Determination 2020 NRC Exam Scenario 4 D-1 Rev 1 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-2: Cool and Depressurize RCS to 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.

(OP-902-007, 11, 12)

A ruptured SG presents a flowpath for radioactive release. Cooling and depressurizing the RCS will prevent lifting an Atmospheric Dump Valve or Main Steam Safety Valve and thus prevent an unnecessary release and lowering the dose to plant personnel and the public.

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.

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)

Depressurization does not necessarily need to be completed but must be commenced to receive credit for the task.

RCS THOT <

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

• If an operator or the crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

• 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 4 2020 NRC Exam Scenario 4 D-1 Rev 1 REFERENCES Event Procedures 1

OP-002-001, Auxiliary Component Cooling Water, Rev. 315 Technical Specification 3.7.3 and cascading Tech Specs Technical Requirements Manual 3.7.15 OP-100-014, Technical Specification and Technical Requirements Compliance, Rev. 354 2

OP-500-011, Control Room Cabinet M, Rev. 41 OP-004-021, Anticipated Transient System, Rev. 302 3

OP-901-112, Charging or Letdown Malfunction, Rev. 8 4

OP-901-202, Steam Generator Tube Leakage or High Activity, Rev. 15 OP-901-212, Rapid Plant Power Reduction, Rev. 14 Technical Specification 3.4.5.2 5

OP-901-201, Steam Generator Level Control Malfunction, Rev. 6 Technical Requirements Manual 3.3.5 6

OP-902-000, Standard Post Trip Actions, Rev. 16 OP-902-007, Steam Generator Tube Rupture Recovery Procedure, Rev. 18 OP-902-009, Standard Appendices, Rev. 319 7

OP-902-000, Standard Post Trip Actions, Rev. 16 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 OP-100-017, Emergency Operating Procedure Implementation Guide, Rev 5 TM-OP-100-03, Simulator Training, Rev. 21