ML22083A061

Wat 2022-02 Draft Outlines

ML22083A061
Person / Time
Site:	Waterford
Issue date:	03/09/2022
From:	Heather Gepford NRC/RGN-IV/DORS/OB
To:	Entergy Operations
References
Download: ML22083A061 (40)

Text

Page 1 of 15 Facility: Waterford K/A Catalog Rev. 2 Rev. 1 Date of Exam: 02/01/2022 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

2 1

4 4

4 3

18 3

3 6

2 1

1 2

2 1

2 9

2 2

4 Tier Totals 3

2 6

6 5

5 27 5

5 10

2.

Plant Systems 1

2 3

3 2

2 3

2 2

4 3

2 28 3

2 5

2 1

0 1

1 2

0 1

1 0

2 1

10 0

2 1

3 Tier Totals 3

3 4

3 4

3 3

3 4

5 3

38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 3

2 2

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.

Page 2 of 15 ES-401 PWR Examination Outline (Waterford)

Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)

Item #

E/APE # / Name /

Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR Q#

1 000007 (EPE 7; BW E02 & E10; CE E02) Reactor Trip, Stabilization, Recovery / 1 X

(000007EK2.03) Knowledge of the interrelations between (EPE

7) REACTOR TRIP, STABILIZATION, RECOVERY / 1 and the following (CFR: 41.7 / 45.7): Reactor trip status panel 3.5 11 2

000008 (APE 8)

Pressurizer Vapor Space Accident / 3 X

(000008AK3.02) Knowledge of the reasons for the following responses as they apply to the (APE 8) PRESSURIZER VAPOR SPACE ACCIDENT / 3 (CFR 41.5,41.10 / 45.6 /

45.13): Why PORV or code safety exit temperature is below RCS or PZR temperature 3.6 31 3

000009 (EPE 9)

Small Break LOCA /

3 X

(000009EA1.09) Ability to operate and / or monitor the following as they apply to (EPE 9) SMALL BREAK LOCA / 3 (CFR: 41.7 / 45.5 / 45.6): RCP 3.6 12 4

000009 (EPE 9)

Small Break LOCA

/ 3 X

(000009EA2.01) Ability to determine and interpret the following as they apply to (EPE 9) SMALL BREAK LOCA / 3 (CFR: 43.5 / 45.13): Actions to be taken, based on RCS temp and press, saturated and superheated 4.8 81 5

000011 (EPE 11)

Large Break LOCA /

3 X

(000011EA2.06) Ability to determine and interpret the following as they apply to (EPE 11) LARGE BREAK LOCA / 3 (CFR:

43.5 / 45.13): That fan is in slow speed and dampers are in accident mode during LOCA 3.7 32 6

000015 (APE 15)

Reactor Coolant Pump Malfunctions

/ 4 X (000015 (APE 15) Reactor Coolant Pump Malfunctions / 4)

(G2.4.11) Knowledge of abnormal condition procedures.

(CFR: 41.10) 4.0 33 7

000022 (APE 22)

Loss of Reactor Coolant Makeup / 2 X

(000022AA2.03) Ability to determine and interpret the following as they apply to the (APE 22) LOSS OF REACTOR COOLANT MAKEUP / 2 (CFR: 43.5 / 45.13): Failures of flow control valve or controller 3.1 13 8

000025 (APE 25)

Loss of Residual Heat Removal System / 4 X

(000025AK3.01) Knowledge of the reasons for the following responses as they apply to the (APE 25) LOSS OF RESIDUAL HEAT REMOVAL SYSTEM / 4 (CFR 41.5,41.10 / 45.6 /

45.13): Shift to alternate flowpath 3.1 34 9

000025 (APE 25)

Loss of Residual Heat Removal System / 4 X (000025 (APE 25) Loss of Residual Heat Removal System /

4) (G2.2.40) Ability to apply Tech Specs for a system.

(CFR: 43.2 / 43.5) 4.7 82 10 000026 (APE 26)

Loss of Component Cooling Water / 8 X

(000026AK3.01) Knowledge of the reasons for the following responses as they apply to the (APE 26) LOSS OF COMPONENT COOLING WATER / 8 (CFR 41.5,41.10 / 45.6 /

45.13): The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the CCWS coolers 3.2 14 11 000027 (APE 27)

Pressurizer Pressure Control System Malfunction

/ 3 X

(000027AA2.11) Ability to determine and interpret the following as they apply to the (APE 27) PRESSURIZER PRESSURE CONTROL SYSTEM MALFUNCTION / 3 (CFR: 43.5 / 45.13):

RCS pressure 4.0 35 12 000027 (APE 27)

Pressurizer Pressure Control System Malfunction / 3 X (000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3) (G2.2.36) Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR:

41.10 / 43.2 / 45.13) 4.2 83 13 000029 (EPE 29)

Anticipated Transient Without Scram / 1 X

(000029EK1.01) Knowledge of the operational implications of the following concepts as they apply to (EPE 29)

ANTICIPATED TRANSIENT WITHOUT SCRAM / 1 (CFR:

41.8 / 41.10 / 45.3): Reactor nucleonics and thermo-hydraulics behavior 2.8 15

Page 3 of 15 14 000038 (EPE 38)

Steam Generator Tube Rupture / 3 X

(000038EA1.19) Ability to operate and / or monitor the following as they apply to (EPE 38) STEAM GENERATOR TUBE RUPTURE / 3 (CFR: 41.7 / 45.5 / 45.6): MFW System status indicator 3.4 36 15 000038 (EPE 38)

Steam Generator Tube Rupture / 3 X

(000038EA2.08) Ability to determine and interpret the following as they apply to (EPE 38) STEAM GENERATOR TUBE RUPTURE / 3 (CFR: 43.5 / 45.13): Viable alternatives for placing plant in safe condition when condenser is not available 4.4 84 16 000040 (APE 40; BW E05; CE E05; W E12) Steam Line Rupture -

Excessive Heat Transfer / 4 X

(000040AK1.04) Knowledge of the operational implications of the following concepts as they apply to (APE 40) STEAM LINE RUPTURE - EXCESSIVE HEAT TRANSFER / 4 (CFR 41.8 /

41.10 / 45.3): Nil ductility temperature 3.2 16 17 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

(000054AA1.02) Ability to operate and / or monitor the following as they apply to the (APE 54) LOSS OF MAIN FEEDWATER /4 (CFR 41.7 / 45.5 / 45.6): Manual startup of electric and steam-driven AFW pumps 4.4 37 18 000055 (EPE 55)

Station Blackout /

6 X (000055 (EPE 55) Station Blackout / 6) (G2.1.7) Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 43.5) 4.7 85 19 000056 (APE 56)

Loss of Offsite Power / 6 X (000056 (APE 56) Loss of Offsite Power / 6) (G2.4.8)

Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR: 41.10 / 43.5 / 45.13) 3.8 17 20 000057 (APE 57)

Loss of Vital AC Instrument Bus / 6 X (000057 (APE 57) Loss of Vital AC Instrument Bus / 6)

(G2.4.4) 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) 4.5 38 21 000058 (APE 58)

Loss of DC Power /

6 X

(000058AA1.01) Ability to operate and / or monitor the following as they apply to the (APE 58) LOSS OF DC POWER

/ 6 (CFR 41.7 / 45.5 / 45.6): Cross-tie of the affected dc bus with the alternate supply 3.4 18 22 000062 (APE 62)

Loss of Nuclear Service Water / 4 X

(000062AA2.03) Ability to determine and interpret the following as they apply to the (APE 62) LOSS OF NUCLEAR SERVICE WATER / 4 (CFR: 43.5 / 45.13): The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition 2.9 86 23 000065 (APE 65)

Loss of Instrument Air / 8 X

(000065AA2.08) Ability to determine and interpret the following as they apply to the (APE 65) LOSS OF INSTRUMENT AIR / 8 (CFR: 43.5 / 45.13): Failure modes of air-operated equipment 2.9 39 24 000077 (APE 77)

Generator Voltage and Electric Grid Disturbances / 6 X

(000077AK3.01) Knowledge of the reasons for the following responses as they apply to the (APE 77) GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES / 6 (CFR 41.5,41.10 / 45.6 / 45.13): Reactor and turbine trip criteria.

3.9 40 (W E04) LOCA Outside Containment / 3 Not Applicable to CE (W E11) Loss of Emergency Coolant Recirculation / 4 Not Applicable to CE (BW E04; W E05)

Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 Not Applicable to CE K/A Category Totals:

2 1

4 4

4 / 3 3 / 3 Group Point Total:

18 / 6

Page 4 of 15 ES-401 PWR Examination Outline (Waterford)

Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

Item #

E/APE # / Name /

Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR Q#

25 000001 (APE 1)

Continuous Rod Withdrawal / 1 X

(000001AK2.05) Knowledge of the interrelations between the (APE 1) CONTINUOUS ROD WITHDRAWAL / 1 and the following (CFR 41.7 / 45.7): Rod motion lights 2.9 41 26 000003 (APE 3)

Dropped Control Rod / 1 X

(000003AK3.10) Knowledge of the reasons for the following responses as they apply to the (APE 3) DROPPED CONTROL ROD / 1 (CFR 41.5,41.10 / 45.6 / 45.13): RIL and PDIL 3.2 42 27 000024 (APE 24)

Emergency Boration / 1 X (000024 (APE 24) Emergency Boration / 1) (G2.2.37) Ability to determine operability and/or availability of safety related equipment (CFR: 43.5 / 45.7) 4.6 87 28 000028 (APE 28)

Pressurizer (PZR)

Level Control Malfunction / 2 X

(000028AK1.01) Knowledge of the operational implications of the following concepts as they apply to (APE 28)

PRESSURIZER (PZR) LEVEL CONTROL MALFUNCTION / 2 (CFR 41.8 / 41.10 / 45.3): PZR reference leak abnormalities 2.8 43 000032 (APE 32)

Loss of Source Range Nuclear Instrumentation / 7 Not Selected 29 000033 (APE 33)

Loss of Intermediate Range Nuclear Instrumentation / 7 X

(000033AA1.02) Ability to operate and / or monitor the following as they apply to the (APE 33) LOSS OF INTERMEDIATE RANGE NUCLEAR INSTRUMENTATION / 7 (CFR 41.7 / 45.5 / 45.6): Level trip bypass 3.0 44 30 000036 (APE 36; BW/A08) Fuel-Handling Incidents /

8 X

(000036AA2.03) Ability to determine and interpret the following as they apply to the Fuel handling accidents (CFR: 43.5 / 45.13): Magnitude of potential radioactive release 4.2 88 31 000037 (APE 37)

Steam Generator Tube Leak / 3 X

(000037AK3.08) Knowledge of the reasons for the following responses as they apply to the (APE 37) STEAM GENERATOR TUBE LEAK / 3 (CFR 41.5,41.10 / 45.6 /

45.13): Criteria for securing RCP 4.1 45 000051 (APE 51)

Loss of Condenser Vacuum / 4 Not Selected 000059 (APE 59)

Accidental Liquid Radwaste Release /

9 Not Selected 000060 (APE 60)

Accidental Gaseous Radwaste Release /

9 Not Selected 000061 (APE 61)

Area Radiation Monitoring System Alarms / 7 Not Selected 32 000067 (APE 67)

Plant Fire On Site /

8 X

(000067AA1.07) Ability to operate and / or monitor the following as they apply to the (APE 67) PLANT FIRE ON SITE

/ 8 (CFR 41.7 / 45.5 / 45.6): Fire alarm reset panel 2.9 46 33 000068 (APE 68; BW A06) Control Room Evacuation /

8 X

(000068AA2.03) Ability to determine and interpret the following as they apply to the (APE 68) CONTROL ROOM EVACUATION / 8 (CFR: 43.5 / 45.13): T-hot, T-cold, and in-core temperatures 4.2 89

Page 5 of 15 34 000069 (APE 69; W E14) Loss of Containment Integrity / 5 X (000069 (APE 69; W E14) Loss of Containment Integrity / 5)

(G2.4.21) Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12) 4.0 47 35 000074 (EPE 74; W E06 & E07)

Inadequate Core Cooling / 4 X

(000074EA2.08) Ability to determine and interpret the following as they apply to (EPE 74) INADEQUATE CORE COOLING / 4 (CFR: 43.5 / 45.13): The effect of turbine bypass valve operation on RCS temperature and pressure 3.8 48 000076 (APE 76)

High Reactor Coolant Activity / 9 Not Selected 000078 (APE 78*)

RCS Leak / 3 Not Selected (W E01 & E02)

Rediagnosis & SI Termination / 3 Not Applicable to CE (W E13) Steam Generator Overpressure / 4 Not Applicable to CE (W E15)

Containment Flooding / 5 Not Applicable to CE (W E16) High Containment Radiation /9 Not Applicable to CE (BW A01) Plant Runback / 1 Not Applicable to CE (BW A02 & A03)

Loss of NNI-X/Y/7 Not Applicable to CE (BW A04) Turbine Trip / 4 Not Applicable to CE (BW A05)

Emergency Diesel Actuation / 6 Not Applicable to CE (BW A07) Flooding /

8 Not Applicable to CE (BW E03)

Inadequate Subcooling Margin /

4 Not Applicable to CE (BW E08; W E03)

LOCA Cooldown -

Depressurization / 4 Not Applicable to CE 36 (BW E09; CE A13**; W E09 &

E10) Natural Circulation/4 X ((BW E09; CE A13**; W E09 & E10) Natural Circulation/4)

(G2.1.20) Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12) 4.6 49 (BW E13 & E14)

EOP Rules and Enclosures Not Applicable to CE (CE A11**; W E08)

RCS Overcooling -

Pressurized Thermal Shock / 4 Not Selected 37 (CE A16)

Excess RCS Leakage / 2 X (CE A16 Excess RCS leakage / 2) (G2.2.25) Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits (43.2) 4.2 90

Page 6 of 15 (CE E09)

Functional Recovery Not Selected (CE E13*) Loss of Forced Circulation /

LOOP / Blackout / 4 Not Selected K/A Category Totals:

1 1

2 2

1 / 2 2 / 2 Group Point Total:

9 / 4

Page 7 of 15 ES-401 PWR Examination Outline (Waterford)

Form ES-401-2 Plant Systems Tier 2/Group 1 (RO/SRO)

Item #

System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

38 003 (SF4P RCP)

Reactor Coolant Pump System X

(003K6.04) Knowledge of the of the effect of a loss or malfunction on the following will have on the (SF4P RCP) REACTOR COOLANT PUMP SYSTEM (CFR: 41.7 / 45.7):

Containment isolation valves affecting RCP operation 2.8 1

39 004 (SF1; SF2 CVCS) Chemical and Volume Control System X (004 (SF1; SF2 CVCS) CHEMICAL AND VOLUME CONTROL SYSTEM) (G2.4.18) Knowledge of the specific bases for EOPs. (CFR:

41.10 / 43.1 / 45.13) 3.3 2

40 004 (SF1; SF2 CVCS) Chemical and Volume Control System X

(004K6.05) Knowledge of the of the effect of a loss or malfunction on the following will have on the (SF1; SF2 CVCS) CHEMICAL AND VOLUME CONTROL SYSTEM (CFR: 41.7 /

45.7): Sensors and detectors 2.5 9

41 005 (SF4P RHR)

Residual Heat Removal System X

(005K3.06) Knowledge of the effect that a loss or malfunction of the (SF4P RHR) RESIDUAL HEAT REMOVAL SYSTEM will have on the following (CFR: 41.7 / 45.6):

CSS 3.1 3

42 006 (SF2; SF3 ECCS)

Emergency Core Cooling System X

(006K2.04) Knowledge of the bus power supplies to the following (SF2; SF3 ECCS) EMERGENCY CORE COOLING SYSTEM (CFR:

41.7): ESFAS-operated valves 3.6 4

43 006 (SF2; SF3 ECCS)

Emergency Core Cooling System X

(006A2.02) Ability to (a) predict the impacts of the following on the (SF2; SF3 ECCS)

EMERGENCY CORE COOLING SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR: 41.5 /43.5/

45.3/45.13): Loss of flow path 4.3 76

Page 8 of 15 44 007 (SF5 PRTS)

Pressurizer Relief

/ Quench Tank System X

(007K5.02) Knowledge of the operational implications of the following concepts as they apply to the (SF5 PRTS) PRESSURIZER RELIEF/QUENCH TANK SYSTEM (CFR: 41.5 / 45.7): Method of forming a steam bubble in the PZR 3.1 5

45 007 (SF5 PRTS)

Pressurizer Relief / Quench Tank System X

(007A2.02) Ability to (a) predict the impacts of the following on the (SF5 PRTS) PRESSURIZER RELIEF/QUENCH TANK SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR: 41.5 /43.5/

45.3/45.13): Abnormal pressure in the PRT 3.2 77 46 008 (SF8 CCW)

Component Cooling Water System X

(008A3.08) Ability to monitor automatic operations of the (SF8 CCW) COMPONENT COOLING WATER SYSTEM including (CFR:

41.7 / 45.5): Automatic actions associated with the CCWS that occur as a result of a safety injection signal 3.6 6

47 010 (SF3 PZR PCS)

Pressurizer Pressure Control System X

(010A4.03) (SF3 PZR PCS)

PRESSURIZER PRESSURE CONTROL SYSTEM Ability to manually operate and/or monitor in the control room (CFR: 41.7 / 45.5 to 45.8): PORV and block valves 4.0 7

48 010 (SF3 PZR PCS)

Pressurizer Pressure Control System X (010 (SF3 PZR PCS)

PRESSURIZER PRESSURE CONTROL SYSTEM) (G2.4.41)

Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5

/ 45.11) 4.6 78 49 012 (SF7 RPS)

Reactor Protection System X

(012A1.01) Ability to predict and/or monitor changes in parameters associated with operating the (SF7 RPS) REACTOR PROTECTION SYSTEM controls including (CFR:

41.5 / 45.5): Trip setpoint adjustment 2.9 8

50 012 (SF7 RPS)

Reactor Protection System X

(012A4.01) (SF7 RPS) REACTOR PROTECTION SYSTEM Ability to manually operate and/or monitor in the control room (CFR: 41.7 / 45.5 to 45.8): Manual trip button 4.5 19

Page 9 of 15 51 013 (SF2 ESFAS)

Engineered Safety Features Actuation System X

(013K6.01) Knowledge of the of the effect of a loss or malfunction on the following will have on the (SF2 ESFAS) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (CFR: 41.7 / 45.7): Sensors and detectors 2.7 10 52 013 (SF2 ESFAS)

Engineered Safety Features Actuation System X

(013 G2.2.38) Engineered Safety Features Actuation System Knowledge of conditions and limits in the facility license (CFR:

43.1) 4.5 79 53 022 (SF5 CCS)

Containment Cooling System X (022 (SF5 CCS) CONTAINMENT COOLING SYSTEM) (G2.4.34)

Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (CFR:

41.10 / 43.5 / 45.13) 4.2 20 025 (SF5 ICE) Ice Condenser System Not Applicable to WAT-3 54 026 (SF5 CSS)

Containment Spray System X

(026A2.08) Ability to (a) predict the impacts of the following on the (SF5 CSS) CONTAINMENT SPRAY SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR: 41.5 /43.5/

45.3/45.13): Safe securing of containment spray when it can be done) 3.2 21 55 026 (SF5 CSS)

Containment Spray System X

(026A2.06) Ability to (a) predict the impacts of the following on the (SF5 CSS) CONTAINMENT SPRAY SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR:

41.5 /43.5/ 45.3/45.13): Safe Securing of Containment Spray (when it can be done) 3.2 80 56 039 (SF4S MSS)

Main and Reheat Steam System X

(039A1.06) Ability to predict and/or monitor changes in parameters associated with operating the (SF4S MSS) MAIN AND REHEAT STEAM SYSTEM controls including (CFR:

41.5 / 45.5): Main steam pressure 3.0 22 053 (SF1; SF4P ICS*) Int Control System Not Applicable to CE 57 059 (SF4S MFW)

Main Feedwater System X

(059A4.11) (SF4S MFW) MAIN FEEDWATER SYSTEM Ability to manually operate and/or monitor in the control room (CFR: 41.7 / 45.5 to 45.8): Recovery from automatic feedwater isolation 3.1 23

Page 10 of 15 58 061 (SF4S AFW)

Auxiliary /

Emergency Feedwater System X

(061K4.06) Knowledge of (SF4S AFW) AUXILIARY / EMERGENCY FEEDWATER SYSTEM design feature(s) and or interlock(s) which provide for the following (CFR:

41.7): AFW startup permissives 4.0 24 59 062 (SF6 ED AC)

AC Electrical Distribution System X

(062K2.01) (SF6 ED AC) AC ELECTRICAL DISTRIBUTION SYSTEM Knowledge of electrical power supplies to the following (CFR: 41.7): Major system loads 3.3 50 60 062 (SF6 ED AC)

AC Electrical Distribution System X

(062K3.01) Knowledge of the effect that a loss or malfunction of the (SF6 ED AC) AC ELECTRICAL DISTRIBUTION SYSTEM will have on the following (CFR: 41.7 / 45.6):

Major system loads 3.5 25 61 063 (SF6 ED DC)

DC Electrical Distribution System X

(063K2.01) (SF6 ED DC) DC ELECTRICAL DISTRIBUTION SYSTEM Knowledge of electrical power supplies to the following (CFR: 41.7): Major DC loads 2.9 51 62 064 (SF6 EDG)

Emergency Diesel Generator System X

(064A3.13) Ability to monitor automatic operations of the (SF6 EDG) EMERGENCY DIESEL GENERATOR SYSTEM including (CFR: 41.7 / 45.5): Rpm controller/megawatt load control (breaker-open/ breaker-closed effects) 3.0 26 63 073 (SF7 PRM)

Process radiation Monitoring System X

(073K3.01) Knowledge of the effect that a loss or malfunction of the (SF7 PRM) PROCESS RADIATION MONITORING SYSTEM will have on the following (CFR: 41.7 / 45.6):

Radioactive effluent releases 3.6 52 64 073 (SF7 PRM)

Process radiation Monitoring System X

(073K5.01) Knowledge of the operational implications of the following concepts as they apply to the (SF7 PRM) PROCESS RADIATION MONITORING SYSTEM (CFR: 41.5 / 45.7):

Radiation theory, including sources, types, units, and effects 2.5 27 65 076 (SF4S SW)

Service Water System X

(076A2.02) Ability to (a) predict the impacts of the following on the (SF4S SW) SERVICE WATER SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR: 41.5 /43.5/

45.3/45.13): Service water header pressure 2.7 53

Page 11 of 15 66 076 (SF4S SW)

Service Water System X

(076K1.19) Knowledge of the physical connections and/or cause-effect relationships between (SF4S SW) SERVICE WATER SYSTEM and the following (CFR: 41.2 to 41.9

/ 45.7 to 45.8): SWS emergency heat loads 3.6 28 67 078 (SF8 IAS)

Instrument Air System X

(078A3.01) Ability to monitor automatic operations of the (SF8 IAS) INSTRUMENT AIR SYSTEM including (CFR: 41.7 / 45.5): Air pressure 3.1 54 68 078 (SF8 IAS)

Instrument Air System X

(078K1.05) Knowledge of the physical connections and/or cause-effect relationships between (SF8 IAS) INSTRUMENT AIR SYSTEM and the following (CFR: 41.2 to 41.9

/ 45.7 to 45.8): MSIV air 3.4 29 69 103 (SF5 CNT)

Containment System X

(103A3.01) Ability to monitor automatic operations of the (SF5 CNT) CONTAINMENT SYSTEM including (CFR: 41.7 / 45.5):

Containment isolation 3.9 55 70 103 (SF5 CNT)

Containment System X

(103K4.01) Knowledge of (SF5 CNT) CONTAINMENT SYSTEM design feature(s) and or interlock(s) which provide for the following (CFR: 41.7): Vacuum breaker protection 3.0 30 K/A Category Totals:

2 3

3 2

2 3

2 2 / 3 4

3 2 / 2 Group Point Total:

28 / 5

Page 12 of 15 ES-401 PWR Examination Outline (Waterford)

Form ES-401-2 Plant Systems Tier 2/Group 2 (RO/SRO)

Item #

System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

71 001 (SF1 CRDS)

Control Rod Drive System X

(001A2.13) Ability to (a) predict the impacts of the following on the (SF1 CRDS) CONTROL ROD DRIVE SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR: 41.5 /43.5/

45.3/45.13): ATWS 4.4 59 72 002 (SF2; SF4P RCS) Reactor Coolant System X

(002K5.12) Knowledge of the operational implications of the following concepts as they apply to the (SF2; SF4P RCS) REACTOR COOLANT SYSTEM (CFR: 41.5 /

45.7): Relationship of temperature average and loop differential temperature to loop hot-let and cold-leg temperature indications 3.7 60 73 011 (SF2 PZR LCS)

Pressurizer Level Control System X

(011A4.03) (SF2 PZR LCS)

PRESSURIZER LEVEL CONTROL SYSTEM Ability to manually operate and/or monitor in the control room (CFR: 41.7 / 45.5 to 45.8): PZR heaters 3.3 61 74 014 (SF1 RPI) Rod Position Indication System X

(014K5.01) Knowledge of the operational implications of the following concepts as they apply to the (SF1 RPI) ROD POSITION INDICATION SYSTEM (CFR: 41.5 /

45.7): Reasons for differences between RPIS and step counter 2.7 62 75 015 (SF7 NI)

Nuclear Instrumentation System X

(015A2.05) Ability to (a) predict the impacts of the following on the (SF7 NI) NUCLEAR INSTRUMENTATION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR: 41.5 /43.5/

45.3/45.13): Core void formation 3.8 91 016 (SF7 NNI)

Non-Nuclear Instrumentation System Not Selected 017 (SF7 ITM)

In-Core Temperature Monitoring System Not Selected

Page 13 of 15 76 027 (SF5 CIRS)

Containment Iodine Removal System X

(027A4.03) (SF5 CIRS)

CONTAINMENT IODINE REMOVAL SYSTEM Ability to manually operate and/or monitor in the control room (CFR: 41.7 / 45.5 to 45.8): CIRS fans 3.3 66 028 (SF5 HRPS)

Hydrogen Recombiner and Purge Control Not Selected 029 (SF8 CPS)

Containment Purge System Not Selected 77 033 (SF8 SFPCS)

Spent Fuel Pool Cooling System X

(033A1.02) Ability to predict and/or monitor changes in parameters associated with operating the (SF8 SFPCS) SPENT FUEL POOL COOLING SYSTEM controls including (CFR: 41.5 / 45.5):

Radiation monitoring systems 2.8 67 034 (SF8 FHS)

Fuel Handling Equipment System Not Selected 78 035 (SF4P SG)

Steam Generator System X

(035K1.02) Knowledge of the physical connections and/or cause-effect relationships between (SF4P SG) STEAM GENERATOR SYSTEM and the following (CFR:

41.2 to 41.9 / 45.7 to 45.8): MRSS 3.2 68 79 041 (SF4S SDS)

Steam Dump /

Turbine Bypass Control System X

(041K3.04) Knowledge of the effect that a loss or malfunction of the (SF4S SDS) STEAM DUMP/TURBINE BYPASS CONTROL SYSTEM will have on the following (CFR: 41.7 / 45.6):

Reactor power 3.5 73 045 (SF4S MTG)

Main Turbine Generator System Not Selected 050 (SF9 CRV*)

Control Room Ventilation Not selectable KA (r3) 055 (SF4S CARS)

Condenser Air removal Not Selected 80 056 (SF4S CDS)

Condensate System X (056 (SF4S CDS) CONDENSATE SYSTEM) (G2.1.23) Ability to perform specific and integrated plant procedures during all modes of operation. (CFR: 41.10) 4.3 74 068 (SF9 LRS)

Liquid radwaste Not Selected

Page 14 of 15 81 071 (SF9 WGS)

Waste Gas Disposal System X (071 (SF9 WGS) WASTE GAS DISPOSAL SYSTEM) (G2.1.32)

Ability to explain and apply system limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 4.0 92 072 (SF7 ARM)

Area Radiation Monitoring System Not Selected 82 075 (SF8 CW)

Circulating Water System X

(075K4.01) Knowledge of (SF8 CW)

CIRCULATING WATER SYSTEM design feature(s) and or interlock(s) which provide for the following (CFR: 41.7): Heat sink 2.5 75 83 079 (SF8 SAS**)

Station Air System X

(079A2.01) Ability to (a) predict the impacts of the following on the (SF8 SAS**) STATION AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation (CFR: 41.5 /43.5/

45.3/45.13): Cross-connection with IAS 3.2 93 086 (SF8 FPS)

Fire Protection Not Selected K/A Category Totals:

1 0

1 1

2 0

1 1 / 2 0

2 1 / 1 Group Point Total:

10 / 3

Page 15 of 15 ES-401 PWR Examination Outline (Waterford)

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

Facility: Waterford Date of Exam:

02/01/2022 Category K/A #

RO SRO-Only Topic Item #

IR Q#

IR Q#

1. Conduct of Operations G2.1.19 (G2.1.19) Ability to use plant computers to evaluate system or component status. (CFR: 41.10 / 45.12) 84 3.9 56 G2.1.21 (G2.1.21) Ability to verify the controlled procedure copy. (CFR:

41.10 / 45.10 / 45.13) 85 3.5 57 G2.1.8 (G2.1.8) Ability to coordinate personnel activities outside the control room. (CFR: 41.10 / 45.5 / 45.12 / 45.13) 86 3.4 58 G2.1.34 (G2.1.34) Knowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12) 91 3.5 94 G2.1.35 (G2.1.35) Knowledge of the fuel-handling responsibilities of SROs. (CFR: 41.10 / 43.7) 92 3.9 95 Subtotal 3

2

2. Equipment Control G2.2.12 (G2.2.12) Knowledge of surveillance procedures. (CFR: 41.10 /

45.13) 89 3.7 63 G2.2.38 (G2.2.38) Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13) 90 3.6 64 G2.2.43 (G2.2.43) Knowledge of the process used to track inoperable alarms. (CFR: 41.10 / 43.5 / 45.13) 91 3.0 65 G2.2.11 (G2.2.11) Knowledge of the process for controlling temporary design changes. (CFR: 41.10 / 43.3 / 45.13) 96 3.3 96 G2.2.21 (G2.2.21) Knowledge of pre-and post-maintenance operability requirements. (CFR: 41.10 / 43.2) 97 4.1 97 Subtotal 3

2

3. Radiation Control G2.3.13 (G2.3.13) Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 / 45.10) 94 3.4 69 G2.3.7 (G2.3.7) Ability to comply with radiation work permit requirements during normal or abnormal conditions. (CFR:

41.12 / 45.10) 95 3.5 70 G2.3.11 (G2.3.11) Ability to control radiation releases. (CFR: 41.11 /

43.4 / 45.10) 98 4.3 98 Subtotal 2

1

4. Emergency Procedures/Plan G2.4.17 (G2.4.17) Knowledge of EOP terms and definitions. (CFR:

41.10 / 45.13) 97 3.9 71 G2.4.30 (G2.4.30) Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11) 98 2.7 72 G2.4.23 (G2.4.23) Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations. (CFR: 41.10 / 43.5 / 45.13) 99 4.4 99 G2.4.28 (G2.4.28) Knowledge of procedures relating to a security event (non-safeguards information). (CFR: 41.10 / 43.5 /

45.13) 100 4.1 100 Subtotal 2

2 Tier 3 Point Total 10 7

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

Waterford 3 Date of Examination:

01/25/2022 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 EFW Pump operability calculation 2.1.20 Ability to interpret and execute procedure steps.

A2 Equipment Control K/A Importance: 3.7 N,R Calculate Pressurizer heater operability 2.2.12, Knowledge of Surveillance Procedures.

A3 Conduct of Operations K/A Importance: 3.9 N,R Use created 1/M plot to estimate criticality 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, etc 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 Times Based on Dose Rates 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)

Rev 0

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

Waterford 3 Date of Examination:

1/25/2022 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 N, R Review EFW pump Operability Calc (TS) 2.1.20, Ability to interpret and execute procedure steps A6 Equipment Control K/A Importance: 4.1 N, R Review PZR HTR Operability (TS) 2.2.12, Knowledge of Surveillance Procedures A7 Conduct of Operations K/A Importance: 4.2 N, R Review 1/M plot and ECP 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, etc.

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 Determine EAL 2.4.41, Knowledge of the emergency action level thresholds and classifications.

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

Waterford 3 Date of Examination:

01/25/2022 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 Adjust Rods for axial flux-Alt path 001 A2.11 Situations requiring Rx trip RO-4.4 N, A, S 1

S2 Establish natural circulation CE A13 AA.1.1 Ability to operate components for Nat Circ RO-3.3 N, L, S 2

S3 Raise SIT pressure A 006 A4.02 Ability to operate ECCS valves (SIT)

RO - 4.1 N, EN, S 3

S4 Start the B RCP-Alt path 003 A4.06 Ability to operate / monitor RCP parameters RO - 2.9 N, A, L, S 4P S5 Place H2 recombiner in service during LOCA-Alt path A4.01 HRPS Controls RO - 4.0 N, A, EN, L, S

5 S6 Energize 1A safety bus from offsite power 064 A4.07 Transfer EDG load to grid operations RO-3.4 (direct from Aug 2011 Exam)

D, S 6

S7 Swap SFP cooling pumps-alt path 033 A2.02, Loss of SFPCS RO - 2.7 N, A, S 8

S8 Discharge Waste Condensate Tank A to CW system 068 A4.03 Stop release if limits exceeded RO - 3.9 (NRC 2017 Exam)

D, A, S 9

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 Locally borate RCS after control room evacuation N, E, L 1

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

Waterford 3 Date of Examination:

01/25/2022 Exam Level: RO SRO-I SRO-U Operating Test Number:

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

Safety Function P2 Reset EFW turb trip 061 A2.04 Pump failure or improper operation RO-3.4 (Used on March 2011 Exam)

Note: Alternate JPMs may be N2 accum surveillance during startup or Liquid radwaste (LR-42A/44 panels)

D, L, R, E 4S P3 Place Battery Charger 1B in service 062 A3.04 Operation of inverter RO - 2.7 N, A 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 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

Alt RCA JPM topics include Swap waste gas, N2 accum SR during startup, Liq Radwaste (local panels LR-42A / LR44), or Boron management LR panel.

Rev 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 NRC JPM Examination Summary Description S1: The plant has just down-powered to 70% power due to a feedwater heater tube leak and the CRS has directed you to balance ASI IAW procedure OP-010-004, Attachment 9.4 using control rods. Once the applicant has started to move control rods, several rods drop into the core, but an automatic trip does not occur (alt path). The applicant must initiate a reactor trip.

S2: The plant has experienced a loss of offsite power and the CRS has directed you to establish natural circulation IAW procedure OP-902-003, Step 17. This will require adjustment of several parameters to get it at the required values IAW step 17.

S3: The A SIT tank pressure has dropped due the low alarm setpoint and must be raised IAW OP-009-008, section 6.1, using N2, to a final value of 630 psig.

S4: The plant is preparing to startup with only the 2A RCP running. The CRS has directed you to start the B RCP IAW procedure OP-001-002. After the start, the applicant is monitoring the 2B RCP parameters and notices alarms for vibrations. The applicant will take actions IAW OP-901-130, section E2, to trip the pump prior to an auto trip.

S5: the plant has experienced a LOCA with containment Hydrogen at 0.7%. The CRS has directed you to place the first hydrogen recombiner in service IAW OP-008-006, section 6.1, starting at 6.1.3. Using time compression, when the first increase to 10kw is completed, power will continue to increase without control (exceeding 75kw) and will require it to be secured.

S6: the CRS has directed you to energize the 1A safety bus from offsite power in accordance with OP-902-009, Standard Appendices. Direct from bank NRC Aug 2011 Exam.

S7: The B SFP cooling pump maintenance was recently completed, and it is ready for a post maintenance run. The CRS has directed you to shift SFP cooling pumps from the A pump to the B pump IAW OP-002-006, section 6.14. Shortly after securing the A pump, the B pump trips, requiring a restart of the A pump due to loss of SFP cooling (Alt Path).

S8: The CRS has directed you to discharge Waste Condensate Tank A to the Circulating Water System in accordance with OP-007-004, Liquid Waste Management System. Upon initiation of flow, LWM flow controller output fails high, raising flow beyond what is permitted by the release permit. (2017 NRC Exam)

P1: The control room was evacuated due to fire and you are at LCP-43 (Remote Shutdown panel) and have completed steps 1-32 of OP-901-502. You are directed to complete steps 33 and 34 to borate the RCS from LCP-43.

P2: Reset overspeed trip of EFW turbine IAW OP-902-005.

P3: The CRS has directed you to place battery charger 1B in service in accordance with OP-006-003, Section 8.3. After alignment, the HI-V Shutdown lamp remains illuminated, which requires both AC and DC isolation breakers to be opened for the 1B battery charger (step 8.3.5).

ES-301 Transient and Event Checklist Form ES-301-5 Rev. 11 Facility: Wat-3 Date of Exam: Feb 21, 2022 Operating Test No.: 1 A

P P

L I

C A

N T

Crew A E

V E

N T

T Y

P E

Scenarios 1

2 4

3 (spare)

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-I1 SRO-U RX 4

1 1

1 0 NOR 1

1 1

1 1 I/C 2,3,4 7,8 3,6 1-4, 6,8,9 14 4

4 2 MAJ 5,6 5,7 5

5 2

2 1 TS 2,3 1,3,4 4

0 2 2 RO SUR SRO-U RX 1

1 0 NOR 1

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

4 2 MAJ 5

2 2 1 TS 0

2 2 RO1 SRO-I SRO-U RX 1

1 1

1 0 NOR 0

4 1

1 1 1 I/C 4,8,9 2,3,8 2,4,9 9

4 4 2 MAJ 5,6 5,7 5

5 2

2 1 TS 0

2 2 RO SRO-I SRO-U RX 1

1 0 NOR 1

1 1 I/C 4

4 2 MAJ 2

2 1 TS 0

2 2 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.

ES-301 Transient and Event Checklist Form ES-301-5 Rev. 11 Facility: Wat-3 Date of Exam: Feb 21, 2022 Operating Test No.: 1 A

P P

L I

C A

N T

Crew B E

V E

N T

T Y

P E

Scenarios 1

2 4

3 (spare)

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-U1 RX 0

0 1

1 0 NOR 1

4 2

1 1 1 I/C 2,3, 7

2,3,4 6, 8 8

4 4 2 MAJ 5,6 5,7 4

2 2 1 TS 1,2 2

0 2 2 RO SRO-I2 SRO-U RX 4

1 1

1 0 NOR 1

1 1

1 1 I/C 2,3,4 7,8 3,6 7

4 4 2 MAJ 5,6 5,7 4

2 2 1 TS 2,3 3

0 2 2 RO2 SRO-I SRO-U RX 1

4 2

1 1 0 NOR 0

2,3,8 3

1 1 1 I/C 4,8,9 5,7 5

4 4 2 MAJ 5,6 4

3 2

2 1 TS 0

2 2 RO SRO-I SRO-U RX 1

1 0 NOR 1

1 1 I/C 4

4 2 MAJ 2

2 1 TS 0

2 2 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.

Facility: Wat-3 Date of Exam: Feb 21, 2021 Operating Test No.: 1

ES-301 Transient and Event Checklist Form ES-301-5 Rev. 11 A

P P

L I

C A

N T

Crew C E

V E

N T

T Y

P E

Scenarios 1

2 4

3 (spare)

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-U2 RX 0

1 1 0 NOR 4

1 1

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

4 4 2 MAJ 5,7 2

2 2 1 TS 1,2 2

0 2 2 RO SRO-I3 SRO-U RX 4

1 1

1 0 NOR 1

1 1

1 1 I/C 2,3,4 7,8 3,6 1-4, 6,8,9 14 4

4 2 MAJ 5,6 5,7 5

5 2

2 1 TS 2,3 1,3,4 4

0 2 2 RO3 SRO-I SRO-U RX 1

4 2

1 1 0 NOR 0

2,3,8 1

4 1

1 1 I/C 4,8,9 5,7 3,6,8 8

4 4 2 MAJ 5,6 4

5 4

2 2 1 TS 0

2 2 RO4 SRO-I SRO-U RX 0

0 1

1 0 NOR 1

1 1

1 1 I/C 2,3, 7

2,4,9 6

4 4 2 MAJ 5,6 5

3 2

2 1 TS 0

2 2 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 2022 NRC Exam Scenario 1 D-1 Rev 3 Facility:

Waterford 3 Scenario No.:

1 Op Test No.:

2022 Examiners:

Operators:

Initial Conditions:

Mode 2, Reactor Power is at POAH. Two Charging Pumps in operation. AB Buses are aligned to Train B. A LPSI pump OOS for repairs. Temp Diesels are not available.

Turnover:

Protected Train is B. Pull Control rods to continue startup to 1%.

Critical Tasks:

(1) Trip RCPs (2) Manually open CS B train spray valve Event No.

Malf.

No.

Event Type*

Event Description 1

N/A R - ATC N - SRO Pull control rods to continue startup 2

RC19C I-BOP I-SRO TS-SRO Safety Channel C RCS Cold Leg instrument RC-ITI- 0102CC (Loop T112C) fails high requiring TS 3.3.1 entry and bypassing affected bistables 3

CC01A C - BOP TS -SRO A CCW pump trips on OC. TS 3.7.3 and cascading 4

RX14A I-ATC I-SRO Selected Pressurizer Pressure Control Channel (RC-IPR-100X) fails high and Both Pressurizer Spray Valves open, will close in manual 5

RC23A M-All RCS leak, ramps into LB LOCA (Critical Task 1, Trip RCPs) 6 ED01A ED01B ED01C ED01D EG10A M-All Loss of Off-Site Power EDG A trips after 10 seconds 7

CS04B C-BOP C-SRO CS B spray valve CS-125B fails to auto open and must be manually opened (CT-2) 8 RP09E C-ATC BAM-113A / CVC-183 fail to reposition (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 2022 NRC Exam Scenario 1 D-1 Rev 3 Scenario Quantitative Attributes

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

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

3

3. Major transients (1-2) 1

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

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

6. Preidentified critical tasks (> 2) 2

NRC Scenario 1 - Narrative 2022 NRC Exam Scenario 1 D-1 Rev 3 The crew assumes the shift with the reactor at POAH following a forced outage. The turnover will include instructions to continue the startup by pulling control rods in accordance with the reactivity plan.

Event 1: The reactivity plan will include instructions to pull control rods. Once enough of a reactivity change is seen by the examiners the next event can be triggered.

Event 2: After the first event is complete, the Safety Channel C RCS Cold Leg instrument RC-ITI-0102CC (Loop T112C) fails high requiring TS 3.3.1 entry and bypassing affected bistables. The SRO should review and enter Technical Specification 3.3.1 Action 2 and bypass 3-HI LOCAL POWER and 4-LOW DNBR bistables 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.

Event 3: Once TS have been entered and bypasses are complete, the A CCW pump trips, requiring a manual start of another pump. The SRO should review and enter 3.7.3 and cascading Technical Specifications and take actions to align the AB CCW pump within 72 hrs per OP-901-510, Component Cooling Water System Malfunction. Tech Spec 3.8.1.1 will be entered for Electrical Breaker Alignment Check and to verify Train A components and EFW AB operability.

Event 4: After event 3 is complete, the selected Pressurizer Pressure Control Channel (RC-IPR-100X) fails high and both spray valves open but will close in manual. The crew should close the spray valves to stop the pressure reduction. The SRO should enter OP-901-120, Pressurizer Pressure Control Malfunction and implement Section E1 Pressurizer Pressure Control Channel Instrument Failure. The crew should take manual control of the Pressurizer Pressure Controller to restore Pressurizer Pressure to within band (if out), swap control to the Channel Y pressure channel, and return the Pressurizer Pressure Controller back to AUTO.

Event 5: A large Break LOCA is inserted on a ramp. This event contains CT-1, Trip RCPs on LBLOCA.

The ATC will manually trip the reactor then trip RCPs within 3 min following loss of CCW Cooling due to containment pressure reaching 17.7 psia (CSAS) or due to low RCS pressure. The crew will perform Standard Post Trip Actions using OP-902-000, SPTAs and diagnose to OP-902-002, Loss of Coolant Accident Recovery.

Event 6: After the reactor is tripped, and RCPs are secured a loss of offsite power occurs. The A EDG energizes the A safety bus for 10 seconds and then trips on overspeed, the A safety bus remains deenergized (dead) for the remainder of the scenario.

Event 7: CS B spray valve fails to auto open and must be manually opened to meet the containment safety function (CT-2, manually open CS-125B) before leaving step 9.3 of the procedure.

Event 8: BAM-113 and CVC-183 fail to reposition following SIAS and should be repositioned. BAM-113 should be manually opened and CVC-183 should be manually closed.

The scenario can be terminated once all CTs are complete (RCPs tripped and CS-125B open), event diagnosed and procedure transition is done, AND event 8 is completed or at the lead examiners discretion.

NRC Scenario 1 - Critical Task Determination 2022 NRC Exam Scenario 1 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: Trip any RCP exceeding operating limits or after 3 min without CCW flow This task is satisfied by manually tripping all 4 Reactor Coolant Pumps within 3 minutes of a loss of CCW flow to the RCPs. This task becomes applicable following the actuation of CSAS.

This step is performed for protection of the RCPs, since CCW, which provides cooling to the RCPs, is isolated upon CSAS actuation.

CCW flow low/lost to RCPs alarms on CP-2 and CP-18 CCW valve status CP-2 CSAS initiated CP-8 Procedurally driven from OP-902-000 step 3.b.1 and 9.3 Stops RCPs using control switch RCP off light illuminated RCP indicated flow lowering CT-2: CS-125B, CS B spray, valve fails to auto open This task is satisfied by manually opening CS-125B. This task becomes applicable following the actuation of CSAS (containment pressure exceeds 17.7 psia) and must be complete prior to leaving step 14 of the LOCA procedure OP-902-002.

Preserves containment building boundary by preventing or minimizing pressure excursions.

Containment pressure >

17.7 psia CS-125 indicates closed CS Header flow not indicated Procedurally driven from OP-902-000 step 9.3 OR OP-902-002 step 14 Opens CS-125 valve using control switch CSAS annunciators actuated CS-125 indicates open CS Header flow indicated 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 2022 NRC Exam Scenario 1 D-1 Rev 3 REFERENCES Event Procedures 1

OP-010-003, Plant Startup 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-901-510, Component Cooling Water Malfunction, Rev 305 OP-903-066, Electrical Breaker Alignment Check OP-100-014, Technical Specifications and Technical Requirements Compliance, Rev 360 Technical Specification 3.7.3 Cascading, 3.8.1.1b, 3.8.1.1 4

OP-901-120, Pressurizer Pressure Control Malfunction, Rev. 303 5/6/7/8/9 OP-902-000, Standard Post Trip Actions OP-902-002, Loss of Coolant Accident Recovery OP-902-009, Standard Appendices, Rev. 319 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 OP-100-017, EOP Implementation Guide, Rev 5

Appendix D Scenario Outline Form ES-D-1 2022 NRC Exam Scenario 2 D-1 Rev 3 0BFacility:

Waterford 3 Scenario No.:

2 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

MOC. Reactor power is 100%. AB Buses are aligned to Train B. Temp Diesels are not available.

Turnover:

Protected Train is B; Maintain 100%.

Critical Tasks:

(1) Trip reactor during ATWS conditions by opening the 32 breakers before exiting step 1 of SPTAs (2) Commence emergency boration before exiting step 1 of SPTAs and within 1 minute of losing two RCPs Event No.

Malf.

No.

Event Type*

Event Description 1

DI-18A3S10-1 = STOP LO-18A3S10-1 = OFF B_M04 = Fail On TS-SRO AH-12A, Control Room Air Handler trips (new)

TS 3.7.6.3a 2

SG11A I - BOP I - SRO TS - SRO Steam Generator #2 Narrow Range level Safety Channel A fails low (SG-ILT-1123A). (TS 3.3.1, 3,3,2, TRM 3.3.1) 3 RC21A I-All Hot Leg 1 Temperature, RC-ITI-0111X, fails low affecting PZR level setpoint.

4 FW35B R-ATC N-BOP N-SRO 5B Feedwater Heater (Low Pressure) tube leak, Rapid Plant Power Reduction (OP Ex April 2021) 5 ED04A M-All Loss of 1A non-safety bus (causes trip of 1A and 2A RCPs), causes reactor trip. ATWS (CT-1, Trip Reactor by de-energizing CEDMs) 6 RD11A30 RD11A28 RD11A40 C-ATC C-SRO Three CEDMs stick out due to bowing (CT-2 emergency boration required) 7 MS03B M-All ESD due to Safety Valve MS-106B Fail to 50%.

8 FW49A1 C - BOP C - SRO Main Feedwater Isolation Valve Steam Generator 1, FW-184A failed open, will close manually (event trigger setup).

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

Appendix D Scenario Outline Form ES-D-1 2022 NRC Exam Scenario 2 D-1 Rev 3 Scenario Quantitative Attributes

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

2. Abnormal events (2-4) 4

3. Major transients (1-2) 2

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

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

6. Preidentified critical tasks (> 2) 2

NRC Scenario 2 - Narrative 2022 NRC Exam Scenario 2 D-1 Rev 3 The crew assumes the shift at 100% power with instructions to maintain 100% power. AB Buses are aligned to Train B. Temp Diesels are not available.

Event 1: Following crew turnover, AH-12A trips and AH-12B starts. CRS declares AH-12A inoperable and enters TS 3.7.6.3a 7 day LCO.

Event 2: SG #2 NR Level channel (1123A) fails low. The SRO should direct the BOP to bypass bistables 8, 10, and 20 on PPS Channel A. The SRO will enter TS 3.3.1 action 2, 3.3.2 action 19, TRM 3.3.1 action 1 and comply with TRM 3.3.2. TS 3.3.3.5 and 3.3.3.6 are evaluated and determined to be not applicable.

At LCP-43, SG-ILI-1123-A1 is indicating failed low. Use Thunder View if asked for other SG levels at LCP-43.

Event 3: Hot Leg 1 Temperature, RC-ITI-0111X, fails low affecting PZR level setpoint. Pressurizer level setpoint will lower which will cause letdown flow to rise with only one charging pump in operation.

Pressurizer level will lower due to this condition. SRO will enter OP-901-110, Pressurizer Level.

Event 4: Once event 3 is complete, 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: Once event 4 is complete, a Loss of 1A non-safety bus (causes trip of 1A and 2A RCPs),

causes reactor trip. ATWS occurs which requires crew to open the 32A and 32B CEDM MG set breakers to insert all control rods per SPTA procedure OP-902-000, step1 CRITICAL TASK (CT-1).

Event 6: Three CEDMs stick out due to bowing and emergency boration is required within 1 minute of the trip of two RCPs and before leaving the step on reactivity control in SPTA procedure CRITICAL TASK (CT-2).

Event 7: AN ESD occurs with MS-106B, Main Steam Line #2 Safety #1, on SG #2 failing 50% open and requires entry into OP-902-004, Excess Steam Demand Recovery Procedure.

Event 8: Main Feedwater Isolation Valve Steam Generator 1, FW-184A fails to AUTO close on MSIS requiring manual closure.

NRC Scenario 2 - Critical Task Determination 2022 NRC Exam Scenario 2 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: Trip Reactor during ATWS event with the two 32 breakers when other methods fail to trip it.

This task is satisfied by manually tripping the Reactor by de-energizing busses 32A and 32B within 1 minute of the two RCPS tripping and before leaving step 1 of SPTA procedure OP-902-000.

The task becomes applicable following the trip of the 1A and 2A RCPs because a PPS limit could be exceeded.

This task is satisfied by manually Opening BOTH of the following breakers for 5 seconds and THEN re-closing: a) SST A32 FEEDER b) SST B32 FEEDER 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 (after the breakers opened)

Procedurally driven from OP-902-000 step 1.a.1.1)

Open indicators for both A32 and B32 feeder breakers Reactor Trip breakers open All CEA rod bottom lights illuminated (except the three that are stuck out)

Reactor power lowering

NRC Scenario 2 - Critical Task Determination 2022 NRC Exam Scenario 2 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-2: Commence emergency boration prior to exiting step 1 of OP-902-000 SPTAs and within 1 minute of tripping two RCPs.

This task is satisfied by commencing Emergency Boration flow by either Boric Acid makeup pumps or gravity feed valves in accordance with OP-902-000, Standard Post Trip Actions step 1, prior to exiting the step to verify Reactivity Control.

This task becomes applicable following the initiation of a Reactor Trip.

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) 3 CEAs stuck out (their respective Rod bottom lights are extinguished)

CEA indicates withdrawn on CEAC Procedurally driven from OP-902-000 step 1.c.1 OP-901-103, Emergency Boration Initiate Emergency Boration 1) using Boric Acid Pump as follows:

a) Place makeup Mode sel switch to MANUAL b) Open Emergency Boration Valve, BAM-133.

c) Start one Boric Acid Pump.

d) Close recirc valve for Boric Acid Pump started:

BAM-126A Boric Acid Makeup Pump Recirc Valve A or BAM-126B OR

2) Initiate Emergency Boration using Gravity Feed as follows:

Open the following Boric Acid Makeup Gravity Feed valves:

a) BAM-113A Boric Acid Makeup Gravity Feed Valve A b) BAM-113B Boric Acid Makeup Gravity Feed Valve B

3) Close VCT Disch Valve, CVC-183.

Charging flow > 40 gpm on CP-4

NRC Scenario 2 - Critical Task Determination 2022 NRC Exam Scenario 2 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback 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 2022 NRC Exam Scenario 2 D-1 Rev 3 REFERENCES Event Procedures 1

OP-500-011, Control Room Cabinet M, Rev 042 (ARP)

TS 3.7.6.3a 2

OP-500-005, Control Room Cabinet K, Rev 20 OP-009-007, Plant Protection System, Rev. 20 OP-901-201, SG level Control Malfunction, Rev 7 Technical Specifications 3.3.1 action 2, 3.3.2 action 19, and TRM 3.3.1 action 1 3

OP-500-008, Control Room Cabinet H, Rev 44 OP-901-110, Pressurizer level Malfunction, Rev 11 4

OP-500-001, Control Room Cabinet A, Rev 27 OP-901-221, Secondary System Transient, Rev. 8 OP-901-212, Rapid Plant Power Reduction, Rev. 16 5

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

OP-902-000, Standard Post Trip Actions, Rev. 16 OP-901-103, Emergency Boration, Rev 004 7

OI-038-000, EOP Operations Expectations/Guidance, Rev 19 GEN EN-OP-115, Conduct of Operations, Rev. 24 EN-OP-115-08, Annunciator Response, Rev. 4 EN-OP-200, Plant Transient Response Rules, Rev. 4 OI-038-000, EOP Operations Expectations / Guidance, Rev. 16 TM-OP-100-03, Simulator Training, Rev. 14 OP-100-017, Emergency Operating Procedures Implementation Guide, Rev 5

Appendix D Scenario Outline Form ES-D-1 2022 NRC Exam Scenario 4 D-1 Rev 3 0BFacility:

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) Energize the 3A safety bus with EDG A. When EDG B trips on overspeed, EDG A fails to energize 3A Safety Bus due to 3A-2A bus tie failing to open on under voltage.

Manually open 3A-2A bus tie, allowing EDG A output breaker to close and power the bus.

(2) Manually start CCW Pump A (when it does not energize on the sequencer) within 10 minutes of EDG A start in order to prevent overheat of the A EDG.

Event No.

Malf.

No.

Event Type*

Event Description 1

FW05 N - BOP TS - SRO EFW AB pump operability test - EFW AB trips on mechanical overspeed. TS 3.7.1.2 d 2

CV05B2 C - ATC Letdown backpressure control valve, CVC-123B, fails closed 3

NI01H I - BOP TS - SRO Excore Nuclear Instrument ENI-IJI-0001D middle detector fails low. TS 3.3.1, 3.3.3.6 4

RC15A1 I - ATC TS-SRO Pressurizer level transmitter, RC-ILT-110X, Fails Hi.

TS 3.3.3.5a 5

TU06 ED01A-D M - ALL Turbine Trip, Loss Of Offsite Power after 10 seconds, 6

ED23A C - BOP 3AS to A2 Bus Tie Breaker Fails to trip on UV (CT1) 7 EG10B C - None EDG B trips on overspeed after 20 seconds 8

CC23A C - BOP CCW Pump A Fails to Autostart on Sequencer (CT2) 9 CV02A C - ATC Charging Pump A fail to autostart (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 2022 NRC Exam Scenario 4 D-1 Rev 3 Scenario Quantitative Attributes

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

2. Abnormal events (2-4) 2

3. Major transients (1-2) 2

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 2022 NRC Exam Scenario 4 D-1 Rev 3 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 perform OP-903-046 section 5.3, EFW Pump AB Check. EFW AB will trip on mechanical over speed and cannot be reset due to linkage damage. The CRS will enter TS 3.7.1.2 action d.

Event 2: When the Tech Spec review is complete, letdown backpressure control valve, CVC-123B, fails closed. The SRO should enter OP-901-112, Charging or Letdown Malfunction, section E2 which will place the standby backpressure control valve in service.

Event 3: After the standby letdown backpressure control valve is in service, Log Power Channel D will fail low. The CRS should enter TS 3.3.1 functional unit 3 action 2 and TS 3.3.3.6 action 29 and bypass bistables 1-4 on PPS Channel D. Bistable 14 may be bypassed while in Mode 1, not applicable until Mode 2.

Event 4:. After the crew has addressed TS, Pressurizer Level Channel X, RC-ILI-0110X, fails high. The CRS will enter OP-901-110, Pressurizer Level Control Malfunction, section E1, Pressurizer Level Control Channel Malfunction. The crew will swap controlling channel to Channel Y and restore Pressurizer Control back to Auto. The CRS should enter TS 3.3.3.5 action a. TS 3.3.3.6 should be reviewed and determined to not be applicable.

Event 5: After Pressurizer Control is in auto, the Main Turbine will trip followed by a Loss Of Offsite Power. Charging Pump Event 6: The 3A to 2A bus tie breaker will fail to open causing EDG A output breaker failing to close. The crew will take action to open the 3A to 2A bus tie breaker (Critical Task 1) which will allow EDG A to power the 3A Safety Bus.

Event 7: EDG B will trip on overspeed after 20 seconds from event 5.

Event 8: CCW Pump A will fail to load on the sequencer requiring the BOP to manually start CCW Pump A within 10 minutes of output breaker closure (Critical Task 2) to prevent overheating of the only remaining EDG (A) still powering a vital bus and prevent an SBO event.

Event 9: Charging Pump A will fail to auto start. The ATC should recognize that no charging pumps are operating and start Charging Pump A.

The scenario can be terminated after the crew has powered the 3A Safety Bus, verified proper CCW operation, conserved Steam Generator inventory and have discussed actions for restoring Main Feedwater to at least one Steam Generator per OP-902-006, Loss of Feedwater Recovery, or at the lead examiners discretion.

NRC Scenario 4 - Critical Task Determination 2022 NRC Exam Scenario 4 D-1 Rev 3 Critical Task Safety Significance Cueing Measurable Performance Indicator Performance Feedback CT-1: Energize the 3A vital AC Bus with A EDG and prevent entering OP-902-005, Station Blackout Recovery.

Task is applicable when EDG B over-speeds/trips and loss of offsite power occurs.

This task is satisfied by manually opening bus tie breaker 3A-2A, which then allows EDG A to energize the 3 A bus, prior to performing actions in OP-902-005, Station Blackout Recovery.

This task becomes applicable once the Loss of Offsite power occurs.

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

(WTRN-OPS-CRITTASKS, CT-03)

Breaker indication on CP-1 and control room lighting.

OP-902-000, Standard Post Trip Actions The crew takes action to manually energize the required Safety Bus by opening the required 3-2 tie breaker.

EDG status and output breaker indication CT-2: Manually start CCW Pump A (when it does not energize on the sequencer) within 10 minutes of EDG A output breaker being closed in order to prevent overheat of the A EDG.

This task is satisfied by manually starting the A CCW pump when it fails to sequence on the 3A bus.

This task becomes applicable once the Loss of Offsite power occurs.

Failure to establish CCW cooling to an operating and loaded EDG within 10 minutes will overheat the EDG and put the plant in an SBO event and place the plant at increased risk of core damage.

(WTRN-OPS-CRITTASKS, CT-03)

CCW pump indicating lights and CCW flow on CP-8.

OP-902-000, Standard Post Trip Actions The crew takes action to manually start the CCW A pump.

CCW A pump indicating lights and CCW flow.

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 2022 NRC Exam Scenario 4 D-1 Rev 3 REFERENCES Event Procedures 1

OP-903-046, Emergency Feed Pump Operability Check, Rev. 323 Technical Specification 3.7.1.2 OP-100-014, Technical Specification and Technical Requirements Compliance, Rev. 358 2

OP-901-112, Charging or Letdown Malfunction, section E2, Rev. 9 3

Technical Specification 3.3.1 Technical Specification 3.3.3.5 Technical Specification 3.3.3.6 OP-009-007, Plant Protection System, Rev. 20 4

OP-901-110, Pressurizer Level Control Malfunction, Rev. 11 OP-903-013, Monthly Channel Checks, Rev. 21 Technical Specification 3.3.3.5 Technical Specification 3.3.3.6 5

OP-902-000, Standard Post Trip Actions, Rev 17 OP-902-003, Loss of Offsite Power / Loss of Forced Circulation, Rev. 11 OP-902-009, Standard Appendices, Rev. 320 6

OP-902-000, Standard Post Trip Actions, Rev 17 7

None 8

OI-038-000, EOP Operations Expectations / Guidance, Rev. 20 9

EN-OP-115, Conduct of Operations, Rev. 30 GEN EN-OP-115, Conduct of Operations, Rev. 30 EN-OP-115-08, Annunciator Response, Rev. 6 EN-OP-200, Plant Transient Response Rules, Rev. 7 OI-038-000, EOP Operations Expectations / Guidance, Rev. 20 OP-100-017, Emergency Operating Procedure Implementation Guide, Rev 5 EN-TQ-210, Conduct of Simulator Training, Rev. 16 WTRN-OPS-CRITTASK, Waterford 3 Critical Tasks, Rev 0