ML21299A025

From kanterella
Jump to navigation Jump to search
CP-2021-08-FINAL Outlines
ML21299A025
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 09/23/2021
From: Heather Gepford
Operations Branch IV
To:
Vistra Energy
References
Download: ML21299A025 (39)
v • d • e


Text

ES-401 1

Form ES-401-2 Rev. 5 Facility: Comanche Peak Date of Exam: August 19, 2021 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 3

3 N/A 3

3 N/A 3

18 6

2 1

2 2

2 1

1 9

4 Tier Totals 4

5 5

5 4

4 27 10

2.

Plant Systems 1

2 2

3 3

3 3

3 3

2 2

2 28 5

2 1

0 1

1 1

1 1

1 1

1 1

10 3

Tier Totals 3

2 4

4 4

4 4

4 3

3 3

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 3

2 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. 5 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 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 X

Ability to determine or interpret the following as they apply to a small break LOCA: CFR 43.5 / 45.13)

EA2.36 Difference between overcooling and LOCA indications 4.2 39 000011 (EPE 11) Large Break LOCA / 3 X

Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA: CFR 41.8 / 41.10 / 45.3)

EK1.01 Natural circulation and cooling, including reflux boiling 4.1 40 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X

Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow): CFR 41.8 /

41.10 / 45.3)

AK1.01 Natural circulation in a nuclear reactor power plant 4.4 41 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 X

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) 4.5 42 000026 (APE 26) Loss of Component Cooling Water / 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 43 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X

Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: CFR 41.7 / 45.7)

AK2.03 Controllers and positioners 2.6 44 000029 (EPE 29) Anticipated Transient Without Scram / 1 X

Knowledge of the operational implications of the following concepts as they apply to the ATWS: CFR 41.8 / 41.10 / 45.3)

EK1.03 Effects of boron on reactivity 3.6 45 000038 (EPE 38) Steam Generator Tube Rupture / 3 X

Ability to operate and monitor the following as they apply to a SGTR: (CFR 41.7 / 45.5 / 45.6)

EA1.16 S/G atmospheric relief valve and secondary PORV controllers and indicators 4.4 46 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer Uncontrolled Depressurization of all Steam Generators / 4 X

Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: (CFR: 41.8 / 41.10 / 45.3) 000040.AK1.02 Leak rate versus pressure change Ability to determine and interpret the following as they apply to the (Uncontrolled Depressurization of all Steam Generators) (CFR: 43.5 / 45.13)

WE12.EA2.2 Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments 3.2 40 3.4 47

ES-401 3

Form ES-401-2 Rev. 5 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW):

(CFR 41.7 / 45.5 / 45.6)

AA1.02 Manual startup of electric and steam-driven AFW pumps 4.4 48 000055 (EPE 55) Station Blackout / 6 X

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 49 000056 (APE 56) Loss of Offsite Power / 6 X

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 EOP for loss of offsite power 4.4 50 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 X

Ability to determine and interpret the following as they apply to the Loss of DC Power: (CFR: 43.5 /

45.13)

AA2.03 DC loads lost; impact on ability to operate and monitor plant systems 3.5 51 000062 (APE 62) Loss of Nuclear Service Water / 4 X

Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: (CFR 41.4, 41.8 / 45.7 )

AK3.01 The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the nuclear service water coolers AK3.03 Guidance actions contained in EOP for Loss of nuclear service water 3.2*

4.0 52 000065 (APE 65) Loss of Instrument Air / 8 X

Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: (CFR 41.5,41.10 / 45.6 / 45.13)

AK3.08 Actions contained in EOP for loss of instrument air 3.7 53 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X

Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

AK2.06 Reactor power 3.9 54 (W E04) LOCA Outside Containment / 3 (W E11) Loss of Emergency Coolant Recirculation / 4 X

Ability to operate and / or monitor the following as they apply to the (Loss of Emergency Coolant Recirculation) (CFR: 41.7 / 45.5 / 45.6)

EA1.3 Desired operating results during abnormal and emergency situations 3.7 55 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 X

Knowledge of the interrelations between the (Loss of Secondary Heat Sink) 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 3.7 56 K/A Category Totals:

3 3

3 3

3 3

Group Point Total:

18

ES-401 4

Form ES-401-2 Rev. 5 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 X

Knowledge of the interrelations between the Continuous Rod Withdrawal and the following:

(CFR 41.7 / 45.7)

AK2.07 Boric acid pump running lights AK2.06 T-ave./ref. deviation meter 2.8 3.0*

57 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 X

Ability to operate and / or monitor the following as they apply to Emergency Boration: (CFR 41.7 /

45.5 / 45.6)

AA1.26 Boric acid storage tank 3.3 58 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 000037 (APE 37) Steam Generator Tube Leak / 3 X

Knowledge of the reasons for the following responses as they apply to the Steam Generator Tube Leak: (CFR 41.5,41.10 / 45.6 /

45.13)

AK3.02 Reset and check of Condensate air ejector exhaust monitor AK3.05 Actions contained in procedures for radiation monitoring, RCS water inventory balance, S/G tube failure, and plant shutdown 3.2 3.5 59 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X

Ability to operate and / or monitor the following as they apply to the Accidental Liquid Radwaste Release: (CFR 41.7 / 45.5 / 45.6)

AA1.03 Flow rate controller AA1.01 Radioactive-liquid monitor 3.0*

3.5 60 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 X

2.4.3 Ability to identify post-accident instrumentation. (CFR:

41.6 / 45.4) 2.4.34 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) 3.7 4.2 61 000069 (APE 69; W E14) Loss of Containment Integrity -

High Containment Pressure / 5

ES-401 5

Form ES-401-2 Rev. 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling -

Degraded Core Cooling - Saturated Core Cooling / 4 000076 (APE 76) High Reactor Coolant Activity / 9 X

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

(CFR 41.7 / 45.7)

AK2.01 Process radiation monitors 2.6 62 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis - SI Termination / 3 X

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

EK1.2 Normal, abnormal and emergency operating procedures associated with (SI Termination) 3.4 63 (W E13) Steam Generator Overpressure / 4 X

Knowledge of the reasons for the following responses as they apply to the (Steam Generator Overpressure) (CFR: 41.5 /

41.10, 45.6, 45.13)

EK3.2 Normal, abnormal and emergency operating procedures associated with (Steam Generator Overpressure) 2.9 64 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation Operations - Natural Circulation with Steam Void in Vessel with/without RVLIS /4 X

Ability to determine and interpret the following as they apply to the (Natural Circulation Operations)

(CFR: 43.5 / 45.13)

EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments 3.4 65 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:

1 2

2 2

1 1

Group Point Total:

9

ES-401 6

Form ES-401-2 Rev. 5 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

2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.9 1

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

Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: (CFR: 41.7/45/6)

K3.04 RCPS 3.7 2

005 (SF4P RHR) Residual Heat Removal X

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

A1.07 Determination of test acceptability by comparison of recorded valve response times with Tech-Spec requirements A1.02 RHR flow rate 2.5 3.3 3

006 (SF2; SF3 ECCS) Emergency Core Cooling X

Knowledge of the physical connections and/or cause-effect relationships between the ECCS and the following systems: (CFR: 41.2 to 41.9

/ 45.7 to 45.8)

K1.02 ESFAS 4.3 4

007 (SF5 PRTS) Pressurizer Relief/Quench Tank X

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 5

008 (SF8 CCW) Component Cooling Water X

Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, 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)

A2.03 High/low CCW temperature 3.0 6

010 (SF3 PZR PCS) Pressurizer Pressure Control X

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

K5.01 Determination of condition of fluid in PZR, using steam tables 3.5 7

012 (SF7 RPS) Reactor Protection X

Knowledge of the effect of a loss or malfunction of the following will have on the RPS: (CFR: 41.7 / 45/7)

K6.06 Sensors and detectors 2.7*

8 013 (SF2 ESFAS) Engineered Safety Features Actuation X

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

K2.01 ESFAS/safeguards equipment control 3.6*

9 022 (SF5 CCS) Containment Cooling X

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.05 Containment readings of temperature, pressure, and humidity system 3.8 10 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray X

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

A1.03 Containment sump level A1.06 Containment spray pump cooling 3.5 2.7 11

ES-401 7

Form ES-401-2 Rev. 5 039 (SF4S MSS) Main and Reheat Steam X

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

(CFR: 41.7)

K4.06 Prevent reverse steam flow on steam line break K4.05 Automatic isolation of steam line 3.3 3.7 12 059 (SF4S MFW) Main Feedwater X

Ability to manually operate and monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.11 Recovery from automatic feedwater isolation 3.1 13 061 (SF4S AFW)

Auxiliary/Emergency Feedwater X

Ability to monitor automatic operation of the AFW, including: (CFR: 41.7 / 45.5)

A3.04 Automatic AFW isolation A3.03 AFW S/G level control on automatic start 4.1 3.9 14 062 (SF6 ED AC) AC Electrical Distribution X

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

45.3 / 45.13)

A2.08 Consequences of exceeding voltage limitations 2.7 15 063 (SF6 ED DC) DC Electrical Distribution X

2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1

/ 45.13) 3.6 16 064 (SF6 EDG) Emergency Diesel Generator X

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 17 073 (SF7 PRM) Process Radiation Monitoring X

Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.01 Those systems served by PRMs 3.6 18 076 (SF4S SW) Service Water X

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

(CFR: 41/7)

K4.06 Service water train separation 2.8 19 078 (SF8 IAS) Instrument Air X

Ability to monitor automatic operation of the IAS, including: (CFR: 41.7 / 45.5)

A3.01 Air pressure 3.1 20 103 (SF5 CNT) Containment X

Knowledge of the effect that a loss or malfunction of the containment system will have on the following: (CFR: 41.7 / 45.6)

K3.03 Loss of containment integrity under refueling operations 3.7 21 053 (SF1; SF4P ICS*) Integrated Control 004 (SF1; SF2 CVCS) Chemical and Volume Control X

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

K2.02 Makeup pumps 2.9 22 008 (SF8 CCW) Component Cooling Water X

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

A1.01 CCW flow rate 2.8 23

ES-401 8

Form ES-401-2 Rev. 5 012 (SF7 RPS) Reactor Protection X

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

K5.01 DNB 3.3*

24 026 (SF5 CSS) Containment Spray X

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

(CFR: 41.7)

K4.07 Adequate level in containment sump for suction (interlock)

K4.09 Prevention of path for escape of radioactivity from containment to the outside (interlock on RWST isolation after swapover).

3.8*

3.7*

25 039 (SF4S MSS) Main and Reheat Steam X

Knowledge of the effect that a loss or malfunction of the MRSS will have on the following: (CFR: 41.7 / 45.6)

K3.03 AFW pumps 3.2*

26 061 (SF4S AFW)

Auxiliary/Emergency Feedwater X

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

K6.02 Pumps 2.6 27 064 (SF6 EDG) Emergency Diesel Generator X

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)

A2.18 Consequences of premature opening of breaker under load A2.03 Parallel operation of ED/Gs 2.6*

3.1 28 K/A Category Point Totals:

2 2

3 3

3 3

3 3

2 2

2 Group Point Total:

28

ES-401 9

Form ES-401-2 Rev. 5 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 X

Ability to manually operate and/or monitor in the control room: (CFR: 41.7/45.5 to 45.8)

A4.06 Control rod drive disconnect/connect 2.7 33 002 (SF2; SF4P RCS) Reactor Coolant X

Knowledge of the effect or a loss or malfunction on the following RCS components:

(CFR: 41.7 / 45.7)

K6.07 Pumps 2.5 29 011 (SF2 PZR LCS) Pressurizer Level Control X

Knowledge of the physical connections and/or cause-effect relationships between the PZR LCS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.04 RPS 3.8 30 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation X

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

K5.04 Factors affecting accuracy and reliability of calorimetric calibrations 2.6 31 016 (SF7 NNI) Nonnuclear Instrumentation X

Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: (CFR: 41.7 / 45.6)

K3.02 PZR LCS 3.4*

32 017 (SF7 ITM) In-Core Temperature Monitor X

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

(CFR: 41.7)

K4.01 Input to subcooling monitors 3.4 38 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge X

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.04 Containment evacuation signal 3.5 33 033 (SF8 SFPCS) Spent Fuel Pool Cooling X

2.1.28 Knowledge of the purpose and function of major system components and controls.

(CFR: 41.7) 4.1 34 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X

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

A1.02 Steam pressure 3.1 35 045 (SF 4S MTG) Main Turbine Generator X

Ability to monitor automatic operation of the MT/G system, including: (CFR: 41/7 / 45.5)

A3.08 Determination from throttle and governor indicators of turbine trip: several indications, including CRDS trip alarm A3.07 Turbine stop/governor valve closure on turbine trip 3.3*

3.5 36 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste

ES-401 10 Form ES-401-2 Rev. 5 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water X

Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water 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)

A2.01 Loss of intake structure A2.02 Loss of circulating water pumps 3.0*

2.5 37 079 (SF8 SAS**) Station Air X

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

(CFR: 41.7)

K4.01 Cross-connect with IAS 2.9 38 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:

1 0

1 1

1 1

1 1

1 1

1 Group Point Total:

10

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

Form ES-401-3 Rev. 5 Facility: Comanche Peak Date of Exam: August 19, 2021 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR:

41.10 / 43.5 / 45.2 / 45.6) 4.3 66 2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. (CFR: 41.10 / 45.1 /

45.12) 4.1 67 2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. (CFR: 41.10 / 43.6 / 45.6) 4.1 68 Subtotal 3

2. Equipment Control 2.2.4 (multi-unit license) Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility. (CFR: 41.6 / 41.7 / 41.10 / 45.1 / 45.13) 3.6 69 2.2.14 Knowledge of the process for controlling equipment configuration or status. (CFR: 41.10 / 43.3 / 45.13) 3.9 70 2.2.35 Ability to determine Technical Specification Mode of Operation. (CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.6 71 Subtotal 3
3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10) 3.2 72 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.12 / 43.4 / 45.9) 2.9 73 Subtotal 2

4. Emergency Procedures/Plan 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 /

43.5 / 45.13) 3.7 74 2.4.17 Knowledge of EOP terms and definitions. (CFR: 41.10 /

45.13) 3.9 75 Subtotal 2

Tier 3 Point Total 10

ES-401 1

Form ES-401-2 Rev. 5 Facility: Comanche Peak Date of Exam: August 19, 2021 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 1

1 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. 5 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 X

Ability to determine or interpret the following as they apply to a reactor trip: (CFR 41.7 / 45.5 / 45.6)

EA2.04 If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP 4.6 84 000008 (APE 8) Pressurizer Vapor Space Accident / 3 X

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

45.11) 2.4.21 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) 4.6 4.6 85 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

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

(CFR 43.5/ 45.13)

AA2.02 Charging pump problems 3.7 86 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 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 Uncontrolled Depressurization of all Steam Generators / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X

2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7 / 41.10 / 43.2 / 43.3 /

45.3) 4.6 87 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6

ES-401 3

Form ES-401-2 Rev. 5 (W E04) LOCA Outside Containment / 3 X

Ability to determine and interpret the following as they apply to the (LOCA Outside Containment)

(CFR: 43.5 / 45.13)

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations 4.3 88 (W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 X 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) 4.7 89 K/A Category Totals:

3 3

Group Point Total:

6

ES-401 4

Form ES-401-2 Rev. 5 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 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 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 X

Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR: 43.5 /

45.13)

AA2.09 Effect of improper HV setting 2.9 90 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 X 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10 / 43.2 /

45.6) 4.4 91 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms

/ 7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity -

High Containment Pressure / 5 X

Ability to determine and interpret the following as they apply to the (High Containment Pressure)

(CFR: 43.5 / 45.13)

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations 3.8 92 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling -

Degraded Core Cooling - Saturated Core Cooling / 4 X 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5 / 45.13) 4.3 93 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis - SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8

ES-401 5

Form ES-401-2 Rev. 5 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation Operations - Natural Circulation with Steam Void in Vessel with/without RVLIS /4 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:

2 2

Group Point Total:

4

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

Form ES-401-3 Rev. 5 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 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling X

Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 45.5)

A2.07 Loss of heat tracing A2.13 Inadvertent SIS actuation 3.1 4.2 76 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 013 (SF2 ESFAS) Engineered Safety Features Actuation X

2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7

/ 43.2) 4.2 77 022 (SF5 CCS) Containment Cooling X 2.4.18 Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13) 4.0 78 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)

Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution 063 (SF6 ED DC) DC Electrical Distribution X

Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; 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)

A2.02 Loss of ventilation during battery charging 3.1 79 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water

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

Form ES-401-3 Rev. 5 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment X

Ability to (a) predict the impacts of the following malfunctions or operations on the containment 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)

A2.05 Emergency containment entry 3.9 80 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:

3 2 Group Point Total:

5

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

Form ES-401-3 Rev. 5 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (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 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment X

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

(CFR: 41.7)

K4.03 Overload protection 3.3 81 035 (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 X

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 2.8*

82 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection X

2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

(CFR: 41.10 / 45.12) 2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 /45.12) 4.3 4.6 83 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:

1 1

1 Group Point Total:

3

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

Form ES-401-3 Rev. 5 Facility: Comanche Peak Date of Exam: August 19, 2021 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.35 Knowledge of the fuel-handling responsibilities of SROs.

(CFR: 41.10 / 43.7) 3.9 94 2.1.36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7) 4.1 95 Subtotal 2

2. Equipment Control 2.2.11 Knowledge of the process for controlling temporary design changes. (CFR: 41.10 / 43.3 / 45.13) 3.3 96 2.2.43 2.2.12 Knowledge of the process used to track inoperable alarms. (CFR: 41.10 / 43.5 / 45.13)

Knowledge of surveillance procedures. (CFR: 41.10 /

45.13) 3.3 4.1 97 Subtotal 2

3. Radiation Control 2.3.13 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) 3.8 98 Subtotal 1

4. Emergency Procedures/Plan 2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. (CFR: 41.7 /

41.10 / 43.5 / 45.12) 4.4 99 2.4.44 Knowledge of emergency plan protective action recommendations. (CFR: 41.10 / 41.12 / 43.5 / 45.11) 4.4 100 Subtotal 2

Tier 3 Point Total 7

ES-401 Record of Rejected K/As Form ES-401-4 Page 1 of 3 CPNPP 2021-08 NRC RO/SRO ES-401-4 Record of Rejected K/As Rev. 5 Tier /

Group Randomly Selected K/A Reason for Rejection RO EXAM 2/1 005.A1.07 Question 3 Did not have appropriate reference material to generate a closed reference question that matches K/A.

Replaced original K/A with new K/A 005.A1.02 2/1 026.A1.03 Question 11 CPNPP does not use containment sump level as a requirement to operate CS, but implies adequate level based on RWST level. Replaced original K/A with new K/A 026.A1.06 2/1 039.K4.06 Question 12 CPNPP does not have any means of preventing reverse flow in main steam lines other than isolation of faulted SG. Replaced original K/A with new K/A 039.K4.05 2/1 061.A3.04 Question 14 CPNPP does not have auto isolation of AFW, but relies on limiting flow rather than isolating. Replaced original K/A with new K/A 061.A3.03 2/1 026.K4.07 Question 25 CPNPP does not use containment sump level interlocks to operate CS, but implies adequate level based on RWST level. Essentially same K/A as original K/A for Question 11. Replaced original K/A with new K/A 026.K4.09 2/1 064.A2.18 Question 28 Did not have appropriate reference material to generate a closed reference question that matches K/A.

Replaced original K/A with new K/A 064.A2.03 2/2 029.A4.04 Question 33 Did not have appropriate reference material to generate a closed reference question that matches K/A.

Randomly reselected from Tier 2/Group 2 K/A Category K4 due to no 029.A4 K/As being available to generate a question at the RO level. Replaced original K/A with new K/A 001.A4.004 2/2 045.A3.08 Question 36 CPNPP does not have a CRDS alarm associated with turbine trip. Replaced original K/A with new K/A 045.A3.07 2/2 075.A2.01 Question 37 CPNPP does not have procedures that respond to a loss of the CW intake structure. Replaced original K/A with new K/A 075.A2.02

ES-401 Record of Rejected K/As Form ES-401-4 Page 2 of 3 CPNPP 2021-08 NRC RO/SRO ES-401-4 Record of Rejected K/As Rev. 5 Tier /

Group Randomly Selected K/A Reason for Rejection RO EXAM 2/2 079.K4.01 Question 38 CPNPP does not have any cross-ties between IA and SA systems. Randomly reselected from Tier 2/Group 2 K/A Category K4 due to no 079.K4 K/As being available to generate a question at the RO level.

Replaced original K/A with new K/A 017.K4.01 1/1 000011.EK1.01 Question 40 Essentially same K/A as original K/A for Question 41 addressing natural circulation. Randomly reselected from Tier 1/Group 1 K/A Category EK1 due to no 000011.EK1 K/As being available to generate a question at the RO level. Replaced original K/A with new K/A 000007.EK1.05.

Per NRC, replaced new selected K/A with randomly selected K/A from 000040.AK1 - selected 000040.AK1.02.

1/1 000062.AK3.01 Question 52 CPNPP does not have automatic isolation valves on the SW system. Replaced original K/A with new K/A 000062.AK3.03 1/2 000001.AK2.07 Question 57 Did not have appropriate reference material to generate a question that matches K/A. Replaced original K/A with new K/A 000001.AK2.06 1/2 000037.AK3.02 Question 59 Did not have appropriate reference material to generate a question that matches K/A. Replaced original K/A with new K/A 000037.AK3.05 1/2 000059.AA1.03 Question 60 CPNPP does not have flow controllers associated with radiation releases or Rad Monitors. Replaced original K/A with new K/A 000059.AA1.01 1/2 000068.G.2.4.3 Question 61 CPNPP does not have any post-accident instrumentation outside of the Control Room. Unable to write a question to test Control Room Evacuation combined with Post-Accident Instrumentation.

Replaced original K/A with new K/A 000068.G.2.4.34

ES-401 Record of Rejected K/As Form ES-401-4 Page 3 of 3 CPNPP 2021-08 NRC RO/SRO ES-401-4 Record of Rejected K/As Rev. 5 Tier /

Group Randomly Selected K/A Reason for Rejection SRO Exam 2/1 006.A2.07 Question 76 CPNPP does not have any procedures addressing a loss of heat tracing. Replaced original K/A with new K/A 006.A2.13 2/1 086.G.2.1.31 Question 83 Did not have appropriate reference material to generate a question that matches K/A. Replaced original K/A with new K/A 086.G.2.1.20 2/2 000008.G.2.4.41 Question 85 Did not have appropriate reference material to generate a question that matches K/A. Replaced original K/A with new K/A 000008.G.2.4.21 3

2.2.43 Question 97 Did not have appropriate reference material to generate a question that matches K/A. Replaced original K/A with new K/A 2.2.12

ES-301 Administrative Topics Outline Form ES-301-1 Task Summary Page 1 of 2 CPNPP 2021 NRC RO ES-301-1 Administrative Topics Outline Rev. 1 Facility:

CPNPP Units 1 and 2 Date of Examination: August 2021 Examination Level: RO SRO Operating Test Number: NRC Administrative Topic (See Note)

Type Code*

Describe activity to be performed Conduct of Operations (RA1)

R, M 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (3.9)

JPM:

Determine RCS Dilution Requirements and Change in Reactivity (RO1009)

Conduct of Operations (RA2)

R, M 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (3.4)

JPM:

Determine Electrical Safe Work Practice Requirements (BA1110)

Equipment Control (RA3)

R, M 2.2.13 Knowledge of tagging and clearance procedures. (4.1)

JPM:

Determine Clearance Isolation Requirements (RO5005)

Radiation Control (RA4)

R, M 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. (3.5)

JPM:

Determine RWP requirements (RWT056)

Emergency Procedures/Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria:

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

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

(P)revious 2 exams (< 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Task Summary Page 2 of 2 CPNPP 2021 NRC RO ES-301-1 Administrative Topics Outline Rev. 1 RA1 The applicant is presented with a set of RCS conditions including time after shutdown, core burn-up, current boron concentration, and estimated critical boron concentration. The applicant will calculate the amount of dilution required to achieve the estimated critical boron concentration provided. The applicant will the determine the change in reactivity that will be achieved based on diluting the RCS to the given estimated critical boron concentration. The critical steps include determination of RCS dilution (in gallons) required to achieve the estimated critical boron concentration and calculation of the change in reactivity that will be achieved based on diluting to the estimated critical boron concentration. This is a modified bank JPM. (K/A 2.1.25 - IR 3.9)

RA2 The applicant is presented with a task to determine the Hazard Category, Personnel Protective Equipment requirements, Tool requirements, and Safety Boundary requirements for installing grounding straps on the Main Generator Isophase busbar in accordance with STA-124, Electrical Safe Work Practices.

The critical steps will be to identify the Hazard/Risk Category, Clothing requirements, Hearing Protection requirements, Glove requirements, Boundaries and Maximum Time the Flash Suit/Hood can be worn continuously before an Air Blower is required. This is a modified bank JPM. (K/A 2.1.26 - IR 3.4)

RA3 The applicant is presented a scenario in which CCP 1-01 is required to be isolated, vented, and drained for a maintenance clearance. The applicant will identify the correct mechanical and electrical isolation components as well as the correct positions and document on a manual clearance form. The critical steps are to identify the correct components and position of each isolation on the clearance. This is a modified bank JPM. (K/A 2.2.13 - IR 4.1).

RA4 The applicant will be required to refer to the latest Radiation Work Permit associated with Used Fuel Outage 7 (UFO 7). The applicant will be assigned a specific task on the RWP and will be required to determine the following information: if entry into a High Radiation Area is allowed, additional requirements for entry into a posted neutron dosimetry required area, the minimum protective clothing requirements for a firewatch to enter a posted contamination area of < 10K DPM/cm2, and expected total dose received if work on the task takes 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to complete. The critical steps will be to correctly determine the information above. This is a modified bank JPM.

(K/A 2.3.7 - IR 3.5)

ES-301 Administrative Topics Outline Form ES-301-1 Task Summary Page 1 of 2 CPNPP 2021 NRC SRO ES-301-1 Administrative Topics Outline Rev. 2 Facility:

CPNPP Units 1 and 2 Date of Examination: August 2021 Examination Level: RO SRO Operating Test Number: NRC Administrative Topic (See Note)

Type Code*

Describe activity to be performed Conduct of Operations (SA1)

R, D, P 2.1.42 Knowledge of new and spent fuel movement procedures. (3.4)

JPM:

Determine Close Contact Fuel Assembly Movement (FH1305)

Conduct of Operations (SA2)

R, M 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (4.2)

JPM:

Determine Loss of RHR Time Limitations and Adequate Hot Leg Vent Path (SO1002)

Equipment Control (SA3)

R, N 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (4.6)

JPM:

Determine Technical Specification (SO1005)

Radiation Control (SA4)

R, M 2.3.4 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (3.8)

JPM:

Determine Personnel Contamination Requirements and Reporting Requirements (SO1112)

Emergency Procedures/Plan (SA5)

R, N 2.4.41 Knowledge of the emergency action level thresholds and classifications. (4.6)

JPM:

Classify an Emergency Plan Event (SO1136)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria:

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

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

(P)revious 2 exams (< 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Task Summary Page 2 of 2 CPNPP 2021 NRC SRO ES-301-1 Administrative Topics Outline Rev. 2 SA1 The applicant will determine the correct close contact movement for identified fuel assemblies in accordance with RFO-302, Handling of Fuel Assemblies. Three separate fuel moves are required to be analyzed for close contact. The critical steps are to describe the correct close contact fuel movement for each assembly.

This is a direct from bank JPM previously used on the 2018 NRC exam.

(K/A 2.1.42 - IR 3.4)

SA2 The applicant will calculate the time after shutdown and utilize ABN-104, Residual Heat Removal System Malfunction and IPO-010B, Reactor Coolant System Reduced Inventory Operations to determine the Core Time to Saturation, approximate Heat Up Rate, Time to Uncovery, Containment Closure Times, and if an Adequate Hot Leg Vent Path exists. This is a modified from bank JPM.

(K/A 2.1.25 - IR 4.2)

SA3 The applicant will be provided a set of conditions and will be required to determine any Technical Specification Limiting Conditions for Operation that may be impacted by the given conditions, including the associated CONDITION, REQUIRED ACTION, and COMPLETION TIME for each applicable specification.

The critical steps will include identifying the correct Technical Specification LCO along with the correct condition, required action, and completion time. This is a new JPM (K/A 2.2.42 - IR 4.6)

SA4 The applicant will determine the minimum Personnel Contamination Event (PCE) classification for a contamination event as well as the correct methods for removal of a Discrete Radioactive Particle (DRP) from clothing and skin. The applicant will then determine written and oral Reporting Requirements for an overexposure event per STA-501, Nonroutine reporting. The critical steps will be to determine the correct PCE level classification as well as the correct methods for removal of a DRP from skin and clothing and determine the correct oral and written Reporting Requirements for an overexposure event. This is a modified bank JPM.

(K/A 2.3.14 - IR 3.8)

SA5 The applicant will determine the appropriate Emergency Plan Classification in accordance with EPP-201, Assessment of Emergency Action Levels, Emergency Classification, and Plan Activation. The critical steps will be to determine the correct classification within the notification time. This is a new JPM.

(K/A 2.4.41 - IR 4.6)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 1 of 5 CPNPP 2021 NRC RO & SRO ES-301-2 Systems JPM Outline Rev. 3 Facility: CPNPP 1 & 2 Date of Examination: August 2021 Exam Level: RO SRO(I) SRO (U)

Operating Test Number: NRC Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code*

Safety Function S-1 001 - Control Rod Drive System (RO1008)

Perform Control Rod Exercises (RO ONLY)

D, S 1

S-2 004 - Chemical and Volume Control System (RO1335)

Emergency Boration from the RWST A, D, S 2

S-3 061 - Auxiliary/Emergency Feedwater System (RO3505)

Align Station Service Water to Feed Steam Generator from AFW A, EN, L, N, S 4P S-4 035 - Steam Generator System (RO3005)

Control RCS Temperature during Reactor Trip Response A, L, N, S 4S S-5 064 - Emergency Diesel Generator System (RO4302)

Loss of Both 6.9 KV Safeguard Busses A, EN, L, M, S 6

S-6 015 - Nuclear Instrumentation System (RO1818)

Respond to a Source Range Channel Energizing at Power N, S 7

S-7 086 - Component Cooling Water System (RO3603)

Rotate Component Cooling Water Pumps A, D, S 8

S-8 071 - Waste Gas Disposal System (RO4001)

Establish Sample Flow to South Vent Stack Wide Range Gas Monitor N, S 9

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P-1 013 - Emergency Safety Feature Actuation System (AO5422)

Restore SCW Surge Tank Level to Maintain Cooling to ESF Pump Rooms E, L, N, R 2

P-2 056 - Condensate System (AO3529)

Secondary System Isolation following Steam Generator Tube Rupture E, L, N 4S P-3 086 - Fire Protection System (AO4405)

Perform Actions for a Fire in Containment A, E, M, R 8

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 2 of 5 CPNPP 2021 NRC RO & SRO ES-301-2 Systems JPM Outline Rev. 3 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (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 (6) / 4-6 (6) / 2-3 (3)

< 9 (3) / < 8 (2) / < 4 (0)

> 1 (3) / > 1 (3) / > 1 (2)

> 1 (2) / > 1 (2) / > 1 (2) (control room system)

> 1 (5) / > 1 (5) / > 1 (3)

> 2 (8) / > 2 (8) / > 1 (5)

< 3 (0) / < 3 (0) / < 2 (0) (randomly selected)

> 1 (2) / > 1 (2) / > 1 (2)

(8) /

(7) / (3)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 3 of 5 CPNPP 2021 NRC RO & SRO ES-301-2 Systems JPM Outline Rev. 3 NRC JPM Examination Summary Description S-1 The applicant will perform Control Rod Exercises for Control Bank D rods per OPT-106A, Control Rods Exercise. The critical steps include selecting the proper bank to be tested (Bank D), withdrawing the bank to the desired position, returning the bank to the pre-test position, placing the control rods in Manual to verify conditions to return to Auto are met, and then placing the control rods in Auto. This is a direct from bank JPM. This JPM is under the Reactivity Control Safety Function. This will be an RO only JPM. (K/A 001 A4.03 - IR 4.0 / 3.7)

S-2 The applicant will be required to initiate Emergency Boration per ABN-107, Emergency Boration, for 2 stuck control rods following a reactor trip. The applicant will initially attempt to emergency borate per Attachment 1, Emergency Boration through Emergency Borate Valve 1-8104, however, this flowpath will be unavailable (after the running BAT pump trips.) The applicant will then be required to use attachment 4, Transfer of Charging Pump Suction to the RWST, (the only attachment that does not require the use of BAT pumps). The critical steps include opening one of the RWST to Charging Pump Suction Valves, closing both of the VCT to Charging Pump Suction Valves, and opening one of the CCP Miniflow Valves. This is a direct from bank JPM. This JPM is under the Reactor Coolant System Inventory Control Safety Function. (K/A 004 A2.14 - IR 3.8 / 3.9)

S-3 Following a tornado causing a loss of offsite power and damage to the Unit 1 CST, the applicant will be required to align the Station Service Water System to Supply the Auxiliary Feedwater System and feed a Steam Generator due to excessively low Condensate Storage Tank Level. The actions are per ABN-305, Auxiliary Feedwater System Malfunction. The applicant will be directed to supply SG 1-01 from either MDAFWP 1-01 or the TDAFWP, however the SSW to AFW pump suction valve will fail to open on the selected pump requiring the applicant to select the alternate pump. The critical steps include closing all AFW isolation valves to the Steam Generators not being used as Heat Sink, opening the SSW to AFW pump suction valves, opening the selected AFW pump suction valve, and starting the selected AFW pump. This is a new JPM. This JPM is under the Heat Removal from Reactor Core - Primary Systems Safety Function.

(K/A W E05 EA1.1 - IR 4.1 / 4.0)

S-4 Following a Loss of all Thermal Barrier Cooling flow to the RCPs, RCP criteria was met, the unit was tripped and all RCPs were secured. The applicant is directed to perform EOS-0.1A, Reactor Trip Response, beginning at Step 1 - Check RCS Temperature. The alternate path portion of the JPM will require the applicant to increase dumping steam from all Steam Generators via the ARVs and maintain a symmetrical cooldown of the RCS. The Steam Dumps will fail to respond in Steam Pressure Mode in Manual or Automatic control and the applicant must dump steam using the ARVs. The critical steps include manually increasing demand on all four Steam Generator ARV controllers to control RCS Temperature. This is a new JPM. This JPM is under the Heat Removal from Reactor Core - Secondary Systems Safety Function. (K/A EPE E09 EA1.1 - IR 3.5 / 3.5)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 4 of 5 CPNPP 2021 NRC RO & SRO ES-301-2 Systems JPM Outline Rev. 3 S-5 The applicant will respond to a loss of both 6.9 KV Safeguard Buses per ABN-601, Response to a 138/345 KV System Malfunction, Section 7.0, Loss of Both Safeguards Buses - MODE 1, 2, 3, or 4. The alternate path includes attempting to close the Train A or Train B Emergency Diesel Generator Output Breaker. The applicant will determine the first attempted DG Output Breaker will fail to close and energize the alternate AC Safeguards Bus via the alternate DG Output Breaker.

This is a modified from bank JPM. This JPM is under the Electrical Safety Function. This is a PRA significant action. (K/A 064 A4.06 - IR 3.9 / 3.9)

S-6 Following Source Range Instrument N31 Energizing at Power, the applicant is required to perform the actions of ABN-701, Source Range Instrument Malfunction. Critical Steps will include bypassing the N31 Level Trip, placing the N31 High Flux at Shutdown switch in block, and removing the N31 Instrument Power Fuse to de-energize the high voltage. This is a new JPM. This JPM is under the Instrumentation Safety Function. (K/A 015 A2.02 - IR 3.1 / 3.5*)

S-7 The applicant will shift from Train A to Train B Component Cooling Water Pumps per SOP-502A, Component Cooling Water System, Step 5.2.1.1, Starting a Standby CCW Pump During Normal Operation, then Step 5.2.1.2, Placing a CCW Pump in Standby from Dual Pump Operation. The alternate path occurs when the Train B CCW Pump trips shortly after it is started. This is a direct from bank JPM under the Plant Service Systems Safety Function. (K/A 008 A2.01 - IR 3.3 / 3.6)

S-8 The applicant will utilize the PC-11 to establish Sample Flow to the South Vent Stack Wide Range Gas Monitor per SOP-706, Digital Radiation Monitoring System. The critical steps will include selecting the correct radiation monitor on the PC-11 and turning the associated sample pump on. This is a new JPM. This JPM is under the Radioactivity Release Safety Function.

(K/A 071 A4.09 - IR 3.3 / 3.5)

P-1 Actions of ABN-503, Safety Chilled Water System Malfunction are required to maintain Safety Chilled Water Surge Tank level during accident conditions (maintain cooling to ESF pump rooms). The applicant will locally fail open the Safety Chilled Water Surge Tank RMUW Supply Valve in accordance ABN-503 and OWI-206, Guidelines for Operation of Manual and Power Operated Valves, Section 6.3.2.G, Failing/Restoring a Simple Air Operated Valve (without handwheel). Critical Steps include closing the air supply to the filter regulator and opening the blowdown on the filter regulator. This is a new JPM. This JPM is under the Reactor Coolant System Inventory Control Safety Function.

(K/A 013 A3.02 - IR 4.1 / 4.2)

P-2 During Steam Generator Tube Rupture recovery actions, the applicant is required to perform field actions to isolate the condensate polishing demineralizers to minimize the spread of contamination to the secondary systems in accordance with EOP-3.0, Steam Generator Tube Rupture, Attachment 5, Secondary System Isolation. Critical Steps include closing the Condensate Supply Header to Condensate Polishing System Supply Header Isolation valves and isolating the 50 psi Auxiliary Steam Header. This is a new JPM. This JPM is under the Heat Removal from Reactor Core Secondary Systems Safety Function.

(K/A G2.3.14 - IR 3.4 / 3.8)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 5 of 5 CPNPP 2021 NRC RO & SRO ES-301-2 Systems JPM Outline Rev. 3 P-3 The applicant will perform actions for a fire in Containment per ABN-807A/B, Response to a Fire in the Containment Building, Attachment 1, Actions to be Taken by the Nuclear Equipment Operator. Critical Steps include de-energizing various breakers in the plant to preclude spurious valve actuations due to the fire as well as manually, locally closing the RHR Pump 1-01 to CCP Suction Valve which has spuriously opened (Alternate Path). This is a modified bank JPM. This JPM is under the Plant Service Systems Safety Function.

(K/A 067 AA2.17 - IR 3.5 / 4.3)

Appendix D Scenario Outline Form ES-D-1 Page 1 of 40 CPNPP 2021 NRC Simulator Scenario 1 Rev. 1 Facility:

CPNPP 1 & 2 Scenario No.: 1 Op Test No.: August 2021 NRC Examiners:

Operators:

Initial Conditions:

2-3% Power, BOL Turnover: Raise Reactor Power to 6% - 8%, RWST is recirculating with Containment Spray Pump 1-02 (See Turnover Sheet)

Critical Tasks:

CT Manually start Safety Injection Pump 1-01 due to an automatic start failure on Safety Injection, prior to RVLIS 79 above Core Plate Light going DARK.

CT Trip RCPs within 5 minutes upon a Loss of Subcooling per EOP-0.0A, Reactor Trip or Safety Injection or EOP-1.0A, Loss of Reactor or Secondary Coolant Event No. Malf. No.

Event Type*

Event Description 1

R (RO, SRO)

Raise Reactor Power to 6% - 8%.

2 SW01B C (BOP, SRO)

TS (SRO)

SSW Pump 1-02 Trip 3

RX12 I (RO, SRO)

Main Steam Header Transmitter PT-507 Fails Low 4

RX08A RX05A RC12 I (RO, SRO)

TS (SRO)

Pressurizer Common Instrument Line Failure 5

RC13 C (RO, SRO)

TS (SRO) 40 gpm Pressurizer Leak 6

RC12 RC13 RP07A M (RO, BOP, SRO) Spurious Safety Injection Train B, Automatic Safety Injection Train A Failure with a SBLOCA 7

SI04D C (BOP)

Safety Injection Pump 1-01 Auto Start Failure on Safety Injection Signal (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 6

Total malfunctions (5-8) 1 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)

Scenario Event Description NRC Scenario 1 Page 2 of 40 CPNPP 2021 NRC Simulator Scenario 1 Rev. 1 SCENARIO 1

SUMMARY

Event 1 Crew raises Reactor Power to 6%-8% per IPO-003A, Power Operations, Section 5.1, Warmup and Synchronization of the Turbine Generator.

  • Event 2 Crew responds per ABN-501, Station Service Water System Malfunction Section 2.0, Station Service Water Pump Trip and ABN-502, Component Cooling Water Systems Malfunctions Section 2.0, CCW Pump Trip. Emergency Diesel 1-02 and Train B equipment are placed in Pull-Out. Technical Specifications 3.7.8 Condition B, 3.8.1 Condition B.
  • Event 3 Crew responds per ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 3.0 Steam Header Pressure Malfunction. Crew takes manual control of Steam Dump Pressure Controller 1-PK-507 controlling RCS Temperature.
  • Event 4 Crew takes manual control of Pressurizer Level and Pressure and responds per ABN-706, Pressurizer Level Instrument Malfunction, Section 2.0 Pressurizer Level Instrument Malfunction and ABN-705, Pressurizer Pressure Malfunction, Section 2.0 Pressurizer Pressure Instrument Malfunction. Both failed channels are bypassed then restored to automatic control. Technical Specifications 3.3.1 Condition E Function 6 and 8b, 3.3.2 Condition D Function 1d, and Condition L Function 8b.
    • Event 5 When Crew defeats failed pressurizer pressure and level channels, the Pressurizer steam space leak will be initiated. Crew responds per ABN-103, Excessive Reactor Coolant Leakage Section 2.0, Excessive Reactor Coolant Leakage. Crew may reduce letdown to 45 gpm to restore Pressurizer Level.

Technical Specification 3.4.13 Condition A

  • Event 6 Crew responds per EOP-0.0A, Reactor Trip or Safety Injection, manually initiates Safety Injection then transitions to EOP-1.0A, Loss of Reactor or Secondary Coolant.

Event 7 Crew starts SIP 1-01 during performance of EOP-0.0A, Attachment 2, Safety Injection Actuation Alignment.

  • - On Lead Examiners Cue
    • - Starts automatically or on Lead Examiners Cue Termination Criteria Scenario will be terminated when the operators transition to EOS-1.2, Post LOCA Cooldown and Depressurization or at the Lead Examiners discretion.

Page 1 of 56 CPNPP 2021 NRC Simulator Scenario 3 Rev. 2 Appendix D Scenario Outline Form ES-D-1 Facility:

CPNPP 1 & 2 Scenario No.: 3 Op Test No.: August 2021 NRC Examiners:

Operators:

Initial Conditions: 100% power MOL - RCS Boron is 771 ppm (by sample). XST1 is out of service to place XST1A in service.

Turnover: Maintain steady-state power conditions. Pressurizer Steam Space Sample is in progress.

Critical Tasks:

CT Restore Power to Bus 1EA2 in accordance with ECA-0.0A, Loss of All AC Power, prior to placing equipment in PULL-OUT per ECA-0.0A, Step 8.

CT Manually start RHR Pump 1-02, in accordance with EOP-0.0A, Reactor Trip or Safety Injection, Attachment 2, Safety Injection Actuation Alignment, OR EOP-1.0A, Loss of Reactor or Secondary Coolant, Attachment 1.A, Foldout for EOP-1.0A, due to an automatic start failure on Safety Injection, prior to completion of EOP-0.0A.

Event No.

Malf. No.

Event Type*

Event Description 1

ED07B C (RO, BOP, SRO)

TS (SRO)

Loss of Inverter (IV1PC2) 2 ED05H C (RO, BOP, SRO)

TS (SRO) 86-1 LOR 6.9KV Safeguards Bus 1EA1 3

LQY-553 C (BOP, SRO)

TS (SRO)

SG 1-03 Level Transmitter LT-553 Oscillations 4

OVRDE C (RO, SRO)

TS (SRO)

Letdown Isolation Valve (HV-8160) fails closed.

5 ED21A ED21B M (RO, BOP, SRO) Loss of 345 KV East and West busses 6

EG15B C (BOP, SRO)

EDG 1-01 out of service due to 86-1 LOR actions EDG 1-02 Auto/Emergency Start failure, Norm Start Required 7

OVRD C (RO, SRO)

Pressurizer Steam Space Sample Valves (1/1-4165A &

1/1-4176A) fail to auto close. Manual closure required 8

RC08A2 M (RO, BOP, SRO) LBLOCA occurs when DG 1-02 Normal Start is Performed 9

RH01D C (BOP)

RHR Pump 1-02 fails to auto-start from sequencer (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 9

Total malfunctions (5-8) 4 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 2 Major transients (1-2) 2 EOPs entered/requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)

Page 2 of 56 CPNPP 2021 NRC Simulator Scenario 3 Rev. 2 Scenario Event Description NRC Scenario 3 SCENARIO 3

SUMMARY

The crew will assume the watch at 100% power per IPO-003A, Power Operations. XST1 is out of service to swap to XST1A. A Pressurizer Steam Space sample is in progress.

  • Event 1 The first event is a loss of Inverter IV1PC2, crew actions are in accordance with ABN-603, Loss of a Protection or Instrument Bus, and include stabilizing the plant, restoring an alternate power source, and verification of instrument restoration. The SRO will refer to Technical Specification LCOs 3.8.7 and 3.8.9 (applicable during the loss and exited upon power restoration).
  • Event 2 The next event is an 86-1 LOR resulting in a loss of 6.9 KV Safeguards Bus 1EA1. The crew will respond per ABN-602, Response to a 6900/480V System Malfunction. Actions include starting Centrifugal Charging Pump 1-02 and placing Emergency Diesel Generator 1-01 (without Station Service Water flow) in Pull-Out. Additionally, the crew will perform actions per ABN-602 to ensure necessary plant equipment is operating and affected equipment is placed in PULL OUT. The SRO will refer to Technical Specification LCOs 3.8.9 and 3.7.5.
  • Event 3 The third event is an oscillation of SG 1-03 feedwater level transmitter LT-530 (controlling channel).

The BOP will diagnose improper control response, place SG 1-03 Flow Control Valve controller 1-FK-530 in manual and control feedwater flow to restore SG 1-03 level to program. The crew will take the actions of ABN-710, Steam Generator Level Instrumentation Malfunction. The SRO will refer to Technical Specification LCOs 3.3.1 and 3.3.2.

  • Event 4 The fourth event is a loss of Letdown due to Letdown Isolation Valve (HV-8160) failing closed.

Actions are per ABN-105, Chemical and Volume Control System Malfunction, and require controlling Charging and Seal Injection flows until Letdown can be restored. The RO will be directed to place Excess Letdown in service. The SRO will refer to Technical Specification LCO 3.6.3.

  • Events 5, 6, 7 The first major event is a loss of the 345 KV East & West busses resulting in a Loss of Offsite Power with Diesel Generator 1-01 previously out of service due to an 86-1 LOR on Safeguards Bus 1EA1.

Diesel Generator 1-02 will fail to start automatically or Manually in Emergency; a Manual Normal start of DG 1-02 will be required in accordance with ECA-0.0A, Loss of All AC Power. The event is complicated by the Pressurizer Steam Space Sample in progress and the valves must be manually closed.

    • Events 8 & 9 A LBLOCA will occur (delayed by 180 seconds) when DG 1-02 is manually (normal) started. RHR Pump1-02 fails to auto-start from the SI sequencer; it is a critical task to manually start the only available RHR Pump. Entries into both FRP-0.1A, Response to Imminent Pressurized Thermal Shock Condition and FRZ-0.1, Response to High Containment Pressure, will be required; however, the actions of these procedures will not be substantive.
  • - On Lead Examiners Cue
    • - Starts automatically

Appendix D Scenario Outline Form ES-D-1 Page 1 of 42 CPNPP 2021 NRC Simulator Scenario 4 Rev. 2 Facility: CPNPP 1 & 2 Scenario No.: 4 Op Test No.: August 2021 NRC Examiners:

Operators:

Initial Conditions:

100% power MOL - RCS Boron is 771 ppm. MDAFW Pump 1-02 is out of service for an oil change.

Turnover: Maintain steady-state power conditions.

Critical Tasks:

CT Trip the Reactor and secure all RCPs, due to loss of all Non-Safeguards Loop CCW flow, prior to any RCP tripping on overcurrent per ABN-502, Component Cooling Water System Malfunctions.

CT Manually initiate Train A and/or Train B Safety Injection, due to failure to automatically initiate, prior to exiting EOP-0.0A, Reactor Trip or Safety Injection.

CT Initiate RCS Feed and Bleed in accordance with FRH-0.1A, Response to Loss of Secondary Heat Sink, such that RCS depressurizes sufficiently for Intermediate Head Injection to occur, prior to all SG Wide Range levels lowering to 0%.

Event No.

Malf. No.

Event Type*

Event Description 1

RX05A I (RO, SRO)

TS (SRO)

Pressurizer Level Transmitter (LT-459) fails low.

2 CH10 C (BOP, SRO)

CRDM Vent Fan 1-01 Trips. Requires start of alternate fan.

3 CH21A C (RO, BOP, SRO)

TS (SRO)

Safety Chiller 1-05 trip 4

MS13B C (RO, SRO)

SG 1-02 Steam Pressure Channel (PT-2326) fails high 5

ICM M (RO, BOP, SRO)

Train B CCW Surge Tank Level Transmitter, LT-4501, fails low, Loss of flow to CCW Non-Safeguards Loop, requires Rx trip and stopping RCPs 6

TC07C C (BOP)

Main Turbine fails to trip on Rx trip, Manual pushbutton fails, requires tripping by securing EHC pumps 7

RP07A RP07B C (RO)

Automatic Safety Injection actuation failure (both trains), Manual actuation required from CB-07 8

FW09A M (RO, BOP, SRO) TDAFWP trips, Loss of Heat Sink (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 8

Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

Scenario Event Description NRC Scenario 4 Page 2 of 42 CPNPP 2021 NRC Simulator Scenario 4 Rev. 2 SCENARIO 4

SUMMARY

  • Event 1 -

The first event is a failure low of Pressurizer Level channel. Letdown will isolate due to PRZR LVL CHAN 1, LT-459A failing low. Crew actions are per ABN-706, Pressurizer Level Instrumentation Malfunction Section 2.0 and include manually reducing Charging flow to RCP Seals only, bypassing the failed channel, restoring normal letdown to service, and restoring pressurizer level control to automatic.

The SRO will refer to Technical Specification LCO 3.3.1.

  • Event 2 -

The operating CRDM vent fan trips. The crew will refer to 1-ALB-3A, Window 2.1, CNTMT FN MASTER TRIP, and ensure that at least one CRDM vent fan is in service, and manually start an alternate vent fan, per SOP-801A, Containment Ventilation System. They will use either Section 5.3.1, Control Rod Drive Mechanism Ventilation System Startup, or Section 5.3.3, Alternating Control Rod Drive Mechanism Ventilation Fans, for this evolution. The SRO may refer to the Technical Requirements Manual LCO 13.7.36 (depending on expediency of crew actions).

  • Event 3 -

The next event is a trip of the Train A Safety Chiller (1-05). Crew actions are per ABN-503, Safety Chilled Water System Malfunction and include starting the unaffected train (Train B) Component Cooling Water Pump and Centrifugal Charging Pump. The crew will then shutdown all equipment supplied by the affected train (Train A) and place the equipment in Pull-Out. This equipment includes RHR Pump 1-01, Containment Spray Pumps 1-01 & 1-03, MDAFWP 1-01, SI Pump 1-01, CCW Pump 1-01, and CCP 1-01. The SRO will refer to Technical Specification LCOs 3.7.19 and 3.7.5 (for loss of two trains of AFW).

  • Event 4 -

The next event is a failure high of Steam Line Pressure Transmitter PT-2326 causing SG 1-02 Atmospheric Relief Valve to open. The Reactor Operator will verify steam line pressure is below the lift pressure of 1125 psig and take manual control of 1-PK-2326 and close the ARV. The crew will take the actions of ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st-Stage Pressure and Feed Header Pressure Instrument Malfunction.

  • Event 5 -

The first major event is a failure of the Train B CCW Surge Tank Level Transmitter, LT-4501 low. CCW Safeguards Loop Supply and Return valves automatically close resulting in a loss of CCW cooling flow to the Non-Safeguards loop components. Crew response will be per ABN-502, Section 5.0 and include a verification that the alternate (non-affected) CCW pump cannot be started (due to loss of Safety Chiller 1-05). The crew will then initiate a Reactor Trip and will be required to trip all RCPs. The crew will enter and take the actions of EOP-0.0A, Reactor Trip or Safety Injection and the Reactor Operator will secure all RCPs after performance of Immediate Operator Actions.

    • Events 6 & 7 -

The Reactor Trip will be complicated by the Main Turbine failing to automatically trip or trip from the manual pushbutton at CB-10. The Main Turbine will be tripped when the BOP secures EHC pumps.

The automatic and manual failure of the Main Turbine to trip will cause an RCS cooldown and lowering of SG pressures enough to meet automatic Safety Injection setpoints. Safety Injection will fail to automatically initiate and must be manually initiated from CB-07 by the Reactor Operator.

Scenario Event Description NRC Scenario 4 Page 3 of 42 CPNPP 2021 NRC Simulator Scenario 4 Rev. 2

    • Event 8 -

The second major will be a trip of the TDAFWP on a time delay after Reactor trip, resulting in an immediate or eventual loss of Secondary Heat Sink. The crew will enter and take the actions of FRH-0.1A, Response to Loss of Secondary Heat Sink. Actions include verifying SI has been actuated and both an SI pump and CCP are running, resetting safeguards signals, and establishing an RCS bleed path with both PRZR PORVs open.

  • - On Lead Examiners Cue
    • - Starts automatically or on Lead Examiners Cue Termination Criteria Scenario will be terminated when the crew has established RCS Bleed and Feed cooling in accordance with FRH-0.1A, Response to Loss of Secondary Heat Sink, or at the Lead Examiners discretion.

Risk Significance:

Failure of risk important system prior to trip:

Pressurizer Level Channel fails low Trip of a Safety Chiller Risk significant core damage sequence:

Loss of Heat Sink Risk significant operator actions:

Manually trip the Main Turbine after failure to automatically trip or manually trip with the pushbutton, manually actuate SI after failure to automatically actuate, secure RCPs upon a loss of Non-Safeguards Loop CCW flow, establish RCS Bleed and Feed cooling

Retrieved from "https://kanterella.com/w/index.php?title=ML21299A025&oldid=3806931"