ML17179A006

2017-06-DRAFT Outlines

ML17179A006
Person / Time
Site:	Comanche Peak
Issue date:	06/19/2017
From:	Vincent Gaddy Operations Branch IV
To:	Vistra Energy
References
50-445/OL-17, 50-446/OL-17
Download: ML17179A006 (57)

Text

ES-401 PWR Examination Outline Form ES-401-2 Page 1 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 Facility: CPNPP Date of Exam: June 19, 2017 Tier Group RO K/A Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total

1. Emergency & Abnormal Plant Evolutions 1 3 3 3 N/A 3 3 N/A 3 18 6 2 1 1 1 2 2 2 9 4 Tier Totals 4 4 4 5 5 5 27 10

2. Plant Systems 1 3 3 2 4 1 2 2 2 3 3 3 28 5 2 2 1 1 1 1 0 1 1 0 1 1 10 3 Tier Totals 5 4 3 5 2 2 3 3 3 4 4 38 8 3. Generic Knowledge and Abilities Categories 1 3 2 3 3 2 4 2 10 1 2 3 4 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category). 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points. 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements. 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution. 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively. 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories. 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As. 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams. 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-2 Page 2 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emer g enc y and Abnormal Plant Evolutions -Tier 1/Grou p 1 (RO/ SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s) IR # 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 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) EA2.22 Charging flow trend recorder 3.0 39 (1) 000011 Large Break LOCA / 3 000015/17 RCP 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) AK1.04 Basic steady state thermodynamic relationship between RCS loops and S/Gs resulting from unbalanced RCS flow 2.9 40 (2) 000022 Loss of Rx Coolant Makeup / 2 x 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (CFR: 41.10 / 43.5 / 45.12) 3.9 41 (3) 000025 Loss of RHR System / 4 x Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: (CFR 41.7 / 45.7) AK2.01 RHR heat exchangers 2.9 42 (4) 000026 Loss of Component Cooling Water / 8 x Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: (CFR: 43.5 / 45.13) AA2.04 The normal values and upper limits for the temperatures of the components cooled by CCW.

2.5 43 (5) 000027 Pressurizer Pressure Control System Malfunction / 3 x Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: (CFR 41.5,41.10 /

45.6 / 45.13) AK3.03 Actions contained in EOP for PZR PCS malfunction 3.7 44 (6) 000029 ATWS / 1 x Knowledge of the interrelations between the and the following an ATWS: (CFR 41.7 / 45.7) EK2.06 Breakers, relays, and disconnects 2.9 45 (7) 000038 Steam Gen. Tube Rupture / 3 x Knowledge of the operational implications of the following concepts as they apply to the SGTR:

(CFR 41.8 / 41.10 / 45.3) EK1.03 Natural circulation 3.9 46 (8)

ES-401 3 Form ES-401-2 Page 3 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emer g enc y and Abnormal Plant Evolutions -Tier 1/Grou p 1 (RO/ SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s) IR # 000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture - Excessive Heat Transfer / 4 x Ability to operate and / or monitor the following as they apply to the Steam Line Rupture: (CFR 41.7 /

45.5 / 45.6) AA1.04 Isolation of all steam lines from header 4.3 47 (9) 000054 (CE/E06) Loss of Main Feedwater / 4 x Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): (CFR 41.8 / 41.10 / 45.3) AK1.02 Effects of feedwater introduction on dry S/G 3.6 48 (10) 000055 Station Blackout / 6 x Knowledge of the reasons for the following responses as they apply to the Station Blackout: (CFR 41.5 / 41.10 / 45.6 / 45.13) EK3.02 Actions contained in EOP for loss of offsite and onsite power 4.3 49 (11) 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 x Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

(CFR: 43.5 / 45.13) AA2.16 Normal and abnormal Pzr level for various modes of plant operation.

3.0 50 (12) 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc 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 ) AK3.04 Effect on the nuclear service water discharge flow header of a loss of CCW.

3.5 51 (13) 000065 Loss of Instrument Air / 8 x 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) 4.0 52 (14) W/E04 LOCA Outside Containment / 3 x Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

(CFR: 41.7 / 45.7) EK2.2 Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

3.8 53 (15) W/E11 Loss of Emergency Coolant Recirc. / 4 x 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR: 41.10 / 43.5 / 45.13) 3.8 54 (16)

ES-401 4 Form ES-401-2 Page 4 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emer g enc y and Abnormal Plant Evolutions -Tier 1/Grou p 1 (RO/ SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s) IR #

BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 x Ability to operate and / or monitor the following as they apply to the (Loss of Secondary Heat Sink)

(CFR: 41.7 / 45.5 / 45.6) EA1.2 Operating behavior characteristics of the facility.

3.7 55 (17) 000077 Generator Voltage and Electric Grid Disturbances / 6 x Ability to operate and/or monitor the following as they apply to Generator Voltage and Electric Grid Disturbances: (CFR: 41.5 and 41.10 / 45.5 / 45.7 /

45.8 ) AA1.03 Voltage regulator controls 3.8 56 (18) K/A Category Totals: 3 333 3 3 Group Point Total: 18

ES-401 5 Form ES-401-2 Page 5 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 x Knowledge of the operational implications of the following concepts as they apply to Emergency Boration: (CFR 41.8 / 41.10 /

45.3) AK1.02 Relationship between boron addition and reactor power 3.6 57 (19) 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 x 2.2.36 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) 3.1 58 (20) 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 x Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: (CFR 41.7 / 45.5 /

45.6) AA1.07 CVCS letdown flow indicator 3.1 59 (21) 000051 Loss of Condenser Vacuum / 4 x Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum: (CFR: 43.5 / 45.13) AA2.02 Conditions requiring reactor and/or turbine trip 3.9 60 (22) 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inadequate Core Cooling / 4 x Ability to determine and interpret the following as they apply to the (Degraded Core Cooling) (CFR: 43.5 / 45.13) EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

3.4 61 (23) 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 x Ability to operate and / or monitor the following as they apply to the (Steam Generator Overpressure) (CFR: 41.7 / 45.5 / 45.6) EA1.3 Desired operating results during abnormal and emergency situations.

3.1 62 (24)

ES-401 6 Form ES-401-2 Page 6 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # W/E15 Containment Flooding / 5 x Knowledge of the reasons for the following responses as they apply to the (Containment Flooding) (CFR: 41.5 / 41.10 / 45.6 / 45.13) EK3.3 Manipulation of controls required to obtain desired operating results during abnormal and emergency situations.

2.9 63 (25) 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 Dies el Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 x Knowledge of the interrelations between the (LOCA Cooldown and Depressurization) 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.

3.6 64 (26) BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 x 2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7 / 45.8) 4.5 65 (27) BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

1112 22 Group Point Total: 9

ES-401 7 Form ES-401-2 Page 7 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 1 (RO/ SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 003 Reactor Coolant Pump x Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) K4.03 Adequate lubrication of the RCP.

2.5 1 (28) 004 Chemical and Volume Control x Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) K4.11 Temperature/pressure control in letdown line: prevent boiling, lifting reliefs, hydraulic shock, piping damage, and burst.

3.1 2 (29) 005 Residual Heat Removal x Knowledge of the physical connections and/or cause/effect relationships between the RHRS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.01 CCWS 3.2 3 (30) 006 Emergency Core Cooling x Knowledge of bus power supplies to the following: (CFR: 41.7) K2.04 ESFAS-operated valves 3.6 4 (31) 007 Pressurizer Relief/Quench Tank x Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: (CFR: 41.7 / 45.6) K3.01 Containment 3.3 5 (32) 008 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) A1.04 Surge tank level 3.1 6 (33) 010 Pressurizer Pressure Control x Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) A4.01 PZR spray valve 3.7 7 (34) 012 Reactor Protection x Ability to monitor automatic operation of the RPS, including: (CFR: 41.7 / 45.5)

A3.02 Bistables 3.6 8 (35) 013 Engineered Safety Features Actuation x Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: (CFR: 41.7 / 45.6) K3.02 RCS 4.3 9 (36) 022 Containment Cooling x 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 3.7 10 (37)

ES-401 8 Form ES-401-2 Page 8 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 1 (RO/ SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 025 Ice Condenser 026 Containment Spray x 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 / 45.13) 3.8 11 (38) 039 Main and Reheat Steam x Knowledge of the operational implications of the following concepts as the apply to the MRSS: (CFR: 41.5 /

45.7) K5.08 Effect of steam removal on reactivity 3.6 12 (39) 059 Main Feedwater x Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; 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) A2.12 Failure of feedwater regulating valves 3.1 13 (40) 061 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) K6.01 Controllers and positioners 2.5 14 (41) 062 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) A2.12 Restoration of power to a system with a fault on it 3.2 15 (42) 063 DC Electrical Distribution x Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) A4.02 Battery voltage indicator 2.8 16 (43) 064 Emergency Diesel Generator x Ability to monitor automatic operation of the ED/G system, including: (CFR: 41.7

/ 45.5) A3.07 Load Sequencing 3.6* 17 (44)

ES-401 9 Form ES-401-2 Page 9 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 1 (RO/ SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 073 Process Radiation Monitoring x Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) K4.01 Release termination when radiation exceeds setpoint 4.0 18 (45) 076 Service Water x Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

K4.02 Automatic start features associated with SWS pump controls 2.9 19 (46) 078 Instrument Air x Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.02 Service air 2.7 20 (47) 103 Containment x Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: (CFR: 41.5 / 45.5) A1.01 Containment pressure, temperature, and humidity 3.7 21 (48) 005 Residual Heat Removal x Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 / 45.7) K6.03 RHR heat exchanger 2.5 22 (49) 007 Pressurizer Relief/Quench Tank x 2.1.28 Knowledge of the purpose and function of major system components and controls. (CFR: 41.7) 4.1 23 (50) 012 Reactor Protection x Knowledge of bus power supplies to the following: (CFR: 41.7) K2.01 RPS channels, components, and interconnections 3.3 24 (51) 039 Main and Reheat Steam x Knowledge of the physical connections and/or cause/effect relationships between the MRSS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.08 MFW 2.7 25 (52) 062 AC Electrical Distribution x Ability to monitor automatic operation of the ac distribution system, including: (CFR: 41.7 / 45.5) A3.04 Operation of inverter (e.g., precharging synchronizing light, static transfer) 2.7 26 (53)

ES-401 10 Form ES-401-2 Page 10 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 1 (RO/ SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 063 DC Electrical Distribution x Knowledge of bus power supplies to the following: (CFR: 41.7) K2.01 Major DC loads 2.9 27 (54) 076 Service Water x Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) A4.01 SWS pumps 2.9 28 (55) K/A Category Point Totals: 3 3 2 4122233 3 Group Point Total: 28

ES-401 11 Form ES-401-2 Page 11 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 2 (RO/ SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 001 Control Rod Drive x Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CRDS controls including: (CFR: 41.5 / 45.5) A1.02 T-ref.

3.1 29 (56) 002 Reactor Coolant x Knowledge of the physical connections and/or cause-effect relationships between the RCS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.12 NIS 3.5* 30 (57) 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation x Knowledge of bus power supplies to the following: (CFR: 41.7) K2.01 NIS channels, components, and interconnections 3.3 31 (58) 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor x Knowledge of the effect that a loss or malfunction of the ITM system will have on the following: (CFR: 41.7 / 45.6) K3.01 Natural circulation indications 3.5 32 (59) 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge x Knowledge of design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

K4.03 Automatic Purge isolation 3.2* 33 (60) 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment x Knowledge of the physical connections and/or cause-effect relationships between the Fuel Handling System and the following systems: (CFR: 41.7) K1.04 NIS 2.6 34 (61)

ES-401 12 Form ES-401-2 Page 12 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 2 (RO/ SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 035 Steam Generator x Ability to (a) predict the impacts of the following malfunctions or operations on the GS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5) A2.03 Pressure/level transmitter failure 3.4 35 (62) 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator x Knowledge of the operational implications of the following concepts as the apply to the MT/B System: (CFR: 41.5 / 45.7) K5.18 Purpose of low-power reactor trips (limited to 25% power) 2.7 36 (63) 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste x Ability to manually operate and/or monitor in the control room (CFR:41.7 / 45.5 to 45.8) A4.04 Automatic isolation 3.8 38 (65) 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water x 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7) 3.9 37 (64) 079 Station Air 086 Fire Protection K/A Category Point Totals: 2 1 1 1101 1011 Group Point Total: 10

ES-401 Generic Knowledge and Abilities Outline (Tier

3) Form ES-401-3 Page 13 of 13 CPNPP 2017 NRC RO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 Facility: CPNPP Date of Exam: June 19, 2017 Category K/A # Topic RO SRO-Only IR # IR #

1. Conduct of Operations 2.1. 2.1.3 Knowledge of shift or short-term relief turnover practices. (CFR: 41.10 / 45.13) 3.7 66 2.1. 2.1.21 Ability to verify the controlled procedure copy. (CFR: 41.10 / 45.10 / 45.13) 3.5* 67 2.1. 2.1.42 Knowledge of new and spent fuel movement procedures. (CFR: 41.10 / 43.7 / 45.13) 2.5 68 Subtotal 3 2. Equipment Control 2.2. 2.2.3 (multi-unit license) Knowledge of the design, procedural, and operational differences between units. (CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12) 3.8 69 2.2. 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 3.7 70 2.2. 2.2.7 Knowledge of the process for conducting special or infrequent tests. (CFR: 41.10 / 43.3 / 45.13) 2.9 71 Subtotal 3 3. Radiation Control 2.3. 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 / 41.12 / 43.4 / 45.9) 2.9 72 2.3. 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 / 45.10) 3.2 73 Subtotal 2 4. Emergency Procedures /

Plan 2.4. 2.4.8 Knowledge of general operating crew responsibilities during emergency operations. (CFR: 41.10 / 45.12) 4.0 74 2.4. 2.4.19 Knowledge of EOP layout, symbols, and icons. (CFR: 41.10 / 45.13) 3.4 75 Subtotal 2 Tier 3 Point Total 10 ES-401 PWR Examination Outline Form ES-401-2 Page 1 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 Facility: CPNPP Date of Exam: June 19, 2017 Tier Group RO K/A Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total

1. Emergency & Abnormal Plant Evolutions 1 N/A N/A 18 2 4 6 2 9 2 2 4 Tier Totals 27 4 6 10

2. Plant Systems 1 28 2 3 5 2 10 0 2 1 3 Tier Totals 38 4 4 8 3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 2 2 3 1 4 2 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category). 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points. 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements. 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution. 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively. 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories. 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As. 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams. 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-2 Page 2 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emer g enc y and Abnormal Plant Evolutions -Tier 1/Grou p 1 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s) IR # 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 x Ability to determine or interpret the following as they apply to a Large Break LOCA: (CFR 43.5 / 45.13) EA2.01Actions to be taken, based on RCS temperature and pressure - saturated and superheated 4.7 76 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 x 2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13) 4.5 77 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 x Ability to determine or interpret the following as they apply to a ATWS: (CFR 43.5 / 45.13) EA2.01 Reactor Nuclear Instrumentation.

4.7 78 000038 Steam Gen. Tube Rupture / 3 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture - Excessive Heat Transfer / 4 x 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13) 4.7 81 000054 (CE/E06) Loss of Main Feedwater / 4 000055 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) 4.0 79 000056 Loss of Off-site Power / 6 x 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12) 4.3 80 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4

ES-401 3 Form ES-401-2 Page 3 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emer g enc y and Abnormal Plant Evolutions -Tier 1/Grou p 1 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s) IR # 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 0 000 2 4 Group Point Total: 6

ES-401 4 Form ES-401-2 Page 4 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO /

SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2G* K/A Topic(s) IR # 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 x Ability to determine and interpret the following as they apply to the Inoperable /

Stuck Control Rod: (CFR: 43.5 / 45.13) AA2.03 Required actions if more than one rod is stuck or inoperable.

4.4 82 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 x Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: (CFR: 43.5 / 45.13) AA2.07 Seal water flow indicator for RCP 2.9 84 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 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 83 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 x 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13) 4.7 85 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7

ES-401 5 Form ES-401-2 Page 5 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO /

SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2G* K/A Topic(s) IR # BW/A04 Turbine Trip / 4 BW/A05 Emergency Dies el Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

0000 22 Group Point Total: 4

ES-401 6 Form ES-401-2 Page 6 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 1 (RO /

SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s) IR # 003 Reactor Coolant Pump 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) 4.0 86 004 Chemical and Volume Control x Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those

predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5) A2.15 High or Low Pzr Level.

3.7 87 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features Actuation 022 Containment Cooling x 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or transmission system operator. (CFR: 41.10 / 43.5 / 45.11) 4.1 88 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam x Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on 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 Increasing steam demand, its relationship to increases in reactor power 3.6 89 059 Main Feedwater 061 Auxiliary/Emergency Feedwater x 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2) 4.7 90 ES-401 7 Form ES-401-2 Page 7 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 1 (RO /

SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s) IR #

062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals: 0 0 0 0000200 3 Group Point Total: 5

ES-401 8 Form ES-401-2 Page 8 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 2 (RO /

SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4G* K/A Topic(s) IR # 001 Control Rod Drive x Ability to (a) predict the impacts of the following malfunction or operations on the CRDS- 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) A2.14 Urgent failure alarm, including rod-out-of-sequence and motion-inhibit

alarms. 3.9 91 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling x Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling 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) A2.03 Abnormal spent fuel pool water level or loss of water level 3.5 92 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate

ES-401 9 Form ES-401-2 Page 9 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 2 (RO /

SRO)System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4G* K/A Topic(s) IR # 068 Liquid Radwaste x 2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13) 4.5 93 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 0 0 0 0000 2001 Group Point Total: 3

ES-401 Generic Knowledge and Abilities Outline (Tier

3) Form ES-401-3 Page 10 of 10 CPNPP 2017 NRC SRO ES-401-2 WRITTEN EXAM OUTLINE REV. 0 Facility: CPNPP Date of Exam: June19, 2017Category K/A # Topic RO SRO-Only IR # IR#1. Conduct of Operations 2.1. 2.1.6 Ability to manage the control room crew during plant transients. (CFR: 41.10 / 43.5 / 45.12 / 45.13) 4.8 94 2.1. 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12) 3.9 95 Subtotal 2 2. Equipment Control 2.2. 2.2.21 Knowledge of pre- and post-maintenance operability requirements. (CFR: 41.10 / 43.2) 4.1 96 2.2. 2.2.20 Knowledge of the process for managing troubleshooting activities. (CFR: 41.10 / 43.5 / 45.13) 3.8 97 Subtotal 2 3. Radiation Control 2.3. 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 / 45.10) 3.7 98 Subtotal 1 4. Emergency Procedures /

Plan 2.4. 2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13) 4.2 99 2.4. 2.4.29 Knowledge of the emergency plan. (CFR: 41.10 / 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 4 CPNPP 2017 NRC RO & SRO ES-401-4 RECORD OF REJECTED K/As REV. 0 Tier / Group Randomly Selected K/A Reason for Rejection RO EXAM 2/1 003 K4.02 Question 1 Could not write a question about RCP interlocks that prevent a cold water accident because there are none.

Replaced K/A 003 K4.02 with K/A 003 K4.03. 2/1 004 K4.10 Question 2 Unable to write an operational ly valid question on the design features and/or interlocks which provide for the

minimum temperature require ments of the CVCS system to prevent boron crystallizati on as the only system that performs this function is the heat trace system which is not operated by Reactor Operators. All attempts resulted

in LOD = 5 questions. Repl aced K/A 004 K4.10 with K/A 004 K4.11 2/1 005 K1.08 Question 3 Could not write a question about RHR and service water because the RHR HXs are cooled by CCW. Replaced

K/A 005 K1.08 with K/A 005 K1.01. 2/1 026 G2.4.2 Question 11 Could not write a discrim inating question on CS setpoints/interlocks and automatic actions that were also Entry Conditions for the EOP.

Replaced K/A 026 G2.4.2 with K/A 026 G2.4.20. 2/1 064 A3.09 Question 17 Could not write a question on EDG automatic transfer switch because it is not utilized at Comanche Peak.

Replaced K/A 064 A3.09 with K/A 064 A3.07. 2/2 001 A1.10 Question 29 Could not write a question about CCW cooling of CRDS because they are air cooled. Replaced K/A 001 A1.10

with K/A 001 A1.02. 2/2 029 K4.02 Question 33 Could not write a question about Containment Purge interlocks/design features that mitigate/alleviate negative pressure in containment because there are none.

Replaced K/A 029 K4.

02 with K/A 029 K4.03 2/2 034 K6.02 Question 34 Could not write a question on the effect of loss or

malfunction of rad monitoring systems on the fuel handling systems because there are no interlocks/trips

associated between the two.

There are no other K6 category items with a value >

2.5 therefore, a K1 item was selected. Replaced K/A 034 K6.02 with K/A 034

K1.04.

ES-401 Record of Rejected K/As Form ES-401-4 Page 2 of 4 CPNPP 2017 NRC RO & SRO ES-401-4 RECORD OF REJECTED K/As REV. 0 Tier / Group Randomly Selected K/A Reason for Rejection 2/2 079 K4.01 Question 38 Could not write a question about a cross tie valve between station (service) and Instrument Air because at CP the systems do not have a cross tie valve.

Furthermore, all other K/As for system 079 with a value >

2.5 have to do with cross connecting the two systems and/or overlap with Q 20. Therefore, replaced K/A 079 K4.01 with K/A 068 A4.04. 1/1 026 AA2.06 Question 43 Could not write a question about time to component

damage due to a loss of component cooling due to lack

of data to support a specific time for equipment damage.

Replaced K/A 026 AA2.06 with K/A 026 AA2.04. 1/1 027 AK3.02 Question 44 Could not write a discrimi nating question on the reason for verifying alternate transmitters prior to shifting flow chart indication that had a LO D > 1. Replaced K/A 027 AK3.02 with K/A 027 AK3.03. 1/1 057 AA2.20 Question 50 Could not write a question about bypassing interlocks to

re-energize Vital AC buses.

There are various interlocks associated with the system but during a loss of power to vital AC buses, there is no guidance to BYPASS these

interlocks. Replaced K/

A 057 AA2.20 with K/A 057 AA2.16. 1/1 062 AK3.01 Question 51 Could not write a questi on about automatic opening

/closing of valve on SSW HX because they do

automatically reposition upon a loss of SSW. Replaced K/A 062 AK3.01 with K/A 062 AK3.04.

3 2.2.4 Question 70 Could not write a question to this K/A without referencing plant specific information and still be generic in nature.

Replaced K/A G2.2.4 with K/A G2.2.12.

3 2.2.20 Question 71 Unable to write an RO level question to knowledge of the process for managing troublesho oting activities as ROs at CPNPP do not manage troubleshooting activities.

Replaced K/A G2.2.20 with K/A G2.2.7 3 2.4.8 Question 74 Could not write a question generic in nature on how ABNs are used in conjunction with EOPs without referencing specific procedur es. Replaced K/A G2.4.8 with K/A G2.4.12.

ES-401 Record of Rejected K/As Form ES-401-4 Page 3 of 4 CPNPP 2017 NRC RO & SRO ES-401-4 RECORD OF REJECTED K/As REV. 0 Tier / Group Randomly Selected K/A Reason for Rejection SRO Exam 1/1 011 EA2.12 Question 76 Could not write a question with regard to reflux boiling during a Large Break LOCA as there is no reflux boiling during a Large Break LOCA. Replaced K/A 011 EA2.12 with K/A 011 EA2.01. 1/1 029 EA2.04 Question 78 Could not write a discrimi nating question at the SRO level for CCP operating indication during an ATWS.

Replaced K/A 029 EA2.04 with K/A 029 EA2.01. 1/1 065 AA2.05 Question 81 Could not write an SRO level question on when to

commence a plant shutdown for Loss of Instrument Air.

Unable to write SRO level question on Loss of Instrument Air. Resampled to another Tier 1 Group 1 category.

Replaced K/A 065 AA2.05 with 040 G2.1.7 1/2 005 AA2.01 Question 82 Could not write a discrimi nating question with plausible distractors for determining a stuck control rod from RCS temperature or NIs. Replaced K/A 005 AA2.01 with K/A

005 AA2.03. 1/2 060 G2.4.18 Question 83 Could not write a question on EOP bases for a gaseous

release because the EOP doesn't address accidental

gaseous releases. Replac ed K/A 060 G2.4.18 with K/A 060 G2.2.25. 1/2 068 AA2.07 Question 84 Could not write a question to the SRO level for this KA.

All KA's in the AA2 category for this KA are RO knowledge. Replaced K/

A 068 AA2.07 with K/A 0028 AA2.07. 1/2 BW/A02 & A03 G2.1.25 Question 85 BW/A02&A03 is a B&W K/A and is not applicable to

Comanche Peak. Replaced K/A BW/A02 & A03 2.1.25 with K/A W/E15 Containment Flooding G2.1.7. 2/1 003 G2.2.3 Question 86 Could not write a question asking about the RCP differences between units because they are the same

pump and use the same procedures. Replaced K/A 003

G2.2.3 with K/A 003 G2.4.35. 2/1 004 A2.18 Question 87 Could not write a discrimi nating question at the SRO level for high VCT level. Replaced K/A 004 A2.18 with

K/A 004 A2.15.

ES-401 Record of Rejected K/As Form ES-401-4 Page 4 of 4 CPNPP 2017 NRC RO & SRO ES-401-4 RECORD OF REJECTED K/As REV. 0 Tier / Group Randomly Selected K/A Reason for Rejection 2/2 001 A2.04 Question 91 Could not write a question about axial power shaping rods because Comanche Peak does not use them.

Replaced K/A 001 A2.04 with K/A 001 A2.14. 2/2 068 G2.4.3 Question 93 Could not write a question on Po st Accident Inst for the Liquid Radwaste system because there is none.

Replaced K/A 068 G2.4.3 with K/A 068 G2.2.38.

3 2.1.7 Question 95 Could not write a question based on plant specific information and still be generic in nature. Replaced K/A

G2.1.7 with K/A G2.1.5.

3 2.2.44 Question 97 Could not write a question about interpreting control room indications to underst and operator actions without being Unit/System specific.

Replaced K/A G2.2.44 with K/A G2.2.20. 3 2.3.4 Question 98 Could not write a discrim inating SRO question on radiation exposure limits t hat did not overlap with Question 83. Replaced K/A G2.3.4 with K/A G2.3.12. 3 2.4.21 Question 100 Could not write a discrim inating SRO question about assessing safety functions without being Unit/System

specific and using actual parameters. Replaced K/A G2.4.21 with K/A G2.4.29.

ES-301 Administrative Topics Outline Form ES-301-1Task Summary Page 1 of 2 CPNPP 2017 NRC RO ES-301-1 ADMINISTRATIVE TOPICS OUTLINE REV. 0 Facility: CPNPP Units 1 and 2 Date of Examination:

June 2017 Examination Level: RO SRO Operating Test Number:

NRC Administrative Topic (See Note)

Type Code* Describe activity to be performed Conduct of Operations (RA1) M,R 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (3.9). JPM: Determine Time to Boil (RO1413M

). Conduct of Operations (RA2) M,R 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperatures, high pressure, caustic chlorine, oxygen, and hydrogen. (3.4) JPM: Determine Electrical Safe Work Practice Requirements. (BA1110)

Equipment Control (RA3) M,R 2.2.12 Knowledge of surveillance procedures.

(3.7) JPM: Control Room AC Surveillance / Tech Specs (RO4108A)

Radiation Control (RA4) M,R 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(3.2) JPM: Determine Escorted Radiation Worker Allowable Dose (BA1402)

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-301-1Task Summary Page 2 of 2 CPNPP 2017 NRC RO ES-301-1 ADMINISTRATIVE TOPICS OUTLINE REV. 0 RA1 The applicant will calculate/determine time to saturation, approximate heat up rate, time to core uncovery, and contai nment closure time following a Loss of Residual Heat Removal System per ABN-104, RHR System Malfunction. The critical steps are to determine Time to Saturation, approximate Heat-up Rate, Time to Core Uncovery, and Containment Cl osure Time. This is a modified bank JPM. (K/A 2.1.25 - IR 3.9) RA2 The applicant is presented with a ta sk to determine the Personnel Protective Equipment and Safety Boundaries for racki ng of the XCICE1, ventilation chiller X-01 compressor motor breaker from connect to disconnect in accordance with STA-124, Electrical Safe Work Practices. The critical steps will be to identify the Hazard/Risk Category, Clothing require ments and Boundaries. This is a modified bank JPM. (K/A 2.1.26 - IR 3.4)

RA3 The applicant will complete the Control Room Air Conditioning System surveillance per OPT-116, CR AC SYSTEM. Record and complete all data on

OPT-116-1, CR AC System Data Sheet. The critical steps are to determine if the surveillance is sat or unsat and determi ne a Technical Specification entry is required and inform the Unit Supervisor. This is a modified bank JPM.

(K/A 2.2.12 - IR 3.7)

RA4 The applicant will utilize STA-655, Exposure Monitoring Program, and STA-656, Radiation Work Control, and determine if either of two escorted radiation workers (pump experts) can perform a designated task with or without shielding. The critical steps are to determine the radiation workers will exceed their administrative limits without shielding, however they can perform the designated task with shielding. This is a modified bank JPM. (K/A 2.3.4 - IR 3.2)

ES-301 Administrative Topics Outline Form ES-301-1Task Summary Page 1 of 2 CPNPP 2017 NRC SRO ES-301-1 ADMINISTRATIVE TOPICS OUTLINE REV. 0 Facility: CPNPP Units 1 and 2 Date of Examination:

June 2017 Examination Level: RO SRO Operating Test Number:

NRC Administrative Topic (See Note)

Type Code* Describe activity to be performed Conduct of Operations (SA1) D,R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (4.4) JPM: Perform Reactor Coolant System Pressure/Temperature Verification and Evaluate Technical Specifications. (SO1005) Conduct of Operations (SA2) M,R 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, A no-solo@ operation, maintenance of active license status, 10CFR55, etc. (3.8) JPM: Determine Active / Inactive Status Off Shift License Personnel (SO1004)

Equipment Control (SA3) M,R 2.2.12 Knowledge of surveillance procedures.

(4.1) JPM: Control Room AC Surveillance / Tech Specs. (SO1202M)

Radiation Control (SA4) D,R 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(3.7) JPM: Determine Radiation Levels and Reporting Requirements. (SO1112B)

Emergency Procedures/Plan (SA5) M,R 2.4.41 Knowledge of emergency action level thresholds and classifications. (4.6) JPM: Classify an Emergency Plan Event and based on updated conditions determine if a upgrade is required. (SO1136M) 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-301-1Task Summary Page 2 of 2 CPNPP 2017 NRC SRO ES-301-1 ADMINISTRATIVE TOPICS OUTLINE REV. 0 SA1 The applicant will perform reactor coolant system pressure/temperature verification in accordance with ABN-905A , Loss of Control Room Habitability, Attachment 7, RCS Pressure / Temperature Verification and Evaluate Technical Specifications.

The critical steps are to calculate saturation temperatures, RCS subcooling margin, and RCS cooldown rate as well as identify the correct Technical Specification LCO Condition, Required Actions, and Completion Time.

This is a bank JPM. (K/A 2.1.23 - IR 4.4)

SA2 The applicant is presented with a detailed record (in table form) of watch standing and other activities performed by 4 individual Reacto r Operators over a period of two quarters. The applicant will be required to analyze the work records of these four operators, and apply the guidance of ODA-315, Licensed Operator Maintenance Tra cking, to evaluate and determine if the RO license status is active or inactive for each of the three operators. The critical steps are to determine that the RO licenses for tw o of the four operat ors are NOT active.

This is a modified bank JPM.

(K/A 2.1.4 - IR 3.8) SA3 The applicant will complete the Control Room Air Conditioning System surveillance per OPT-116, CR AC SYSTEM. Record and complete all data on

OPT-116-1, CR AC System Data Sheet. The critical steps are to determine if the surveillance is sat or unsat and the correct Technical Specification LCO Condition, Required Action, and Completion Time. This is a modified bank JPM. (K/A 2.2.12 - IR 4.1)

SA4 The applicant will perform calculations to determine the dose that will be received to two NEOs while performing maintenance, utilizing STA-657, ALARA Job

Planning/Debriefing and determine reporting requirements for an overexposure event per STA-501, Nonroutine reporting. The critical steps will be to calculate the dose received when performing main tenance in the fuel building for two different operators under different conditions and determined the correct oral and written Reporting Requirements for an overexposure event. This is a bank JPM. (K/A 2.3.12 - IR 3.7) 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. One of the critical steps will be the

determination of the correct classificati on. The second critical step will be to determine if an upgrade or follow up message is required, and the latest time the message may be sent. This is a modified bank JPM. (K/A 2.4.41 - IR 4.6)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 1 of 4 CPNPP 2017 NRC RO & SRO ES-301-2 SYSTEM JPM OUTLINE REV. 0 Facility:

CPNPP 1 & 2 Date of Examination:

June 2017 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 A,M,S 1 S-2 004 - Chemical and Volume Control System (RO1305) Isolate Leakage and Establish Excess Letdown A,D,S 2 S-3 011 - LOCA Emergency Core Cooling System (RO1507N)

Transfer From Hot-Leg Recirculation back to Cold leg Recirculation using EOP-1.4A Attachment 2 L,N,S 3 S-4 061 - Auxiliary Feedwater System (RO3516B)

Respond to Inadvertent Start of Turbine Driven Auxiliary Feedwater Pump (RO ONLY)

A,D,EN,S 4S S-5 026 - Containment Spray System (RO1702) Verify Containment Spray Not Required A,EN,L,M,S 5 S-6 064 - Emergency Diesel Generator System (RO4302M) Load Emergency Diesel Generator A,M,S 6 S-7 016 - Non Nuclear Instrumentation System (RO1827A) Respond to Turbine Impulse Pressure Malfunction D,S 7 S-8 086 - Fire Protection System (RO4406C) Respond to a Fire In the Control Room D,L,S 8 In-Plant Systems

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

P-1 004 - Chemical and Volume Control System (AO5403C) Perform Emergency Boration, Boric Acid Gravity Flow Valve Lineup D,E,R 1 P-2 055 - Loss of Offsite and Onsite Power (Station Blackout) (RO4217F) Alignment of PRZR Heaters with APGs supplying Safeguards Bus (Unit 1 Train B)

E,L,M,R 6 P-3 068 - Control Room Evacuation (AO6413B)

Respond to a Fire in the Control Room or Cable Spreading Room, NEO #2 Actions (2015 NRC Exam) A,D,E,L,R,P8

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 2 of 4 CPNPP 2017 NRC RO & SRO ES-301-2 SYSTEM JPM OUTLINE REV. 0

@ 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 / 4-6 / 2-3

< 9 / < 8 / < 4 > 1 / > 1 / > 1 - / - / > 1(control room system) > 1 / > 1 / > 1 > 2 / > 2 / > 1 < 3 < 3 / < 2 (randomly selected) > 1 / > 1 / > 1 NRC JPM Examination Summary Description S-1 The applicants are given a copy of OPT-106A, Control Rod Exercise, and will perform a rod exercise on Control Bank D rods. This is an Alternate Path JPM as

when restoring control rods to the initial position, one control rod will drop, and 1 second later another rod will dr op. The Critical Steps are to select Control Bank D on the Rod Bank Select switch, move C ontrol Bank D >10 and <13 steps using the Control Rod Motion Control switch and t hen back, trip the reactor per ABN-712, Rod Control System Malfuncti on, and manually trip the Turb ine as it failed to trip automatically. This is a modified bank JPM. This JPM is under the Control Rod Drive System - Reactivity Control Safety Function. (K/A 001.A4

.03 - 4.0 / 3.7) (15 Min) S-2 The applicants will Utilize ABN-103, Excessive Reactor Coolant Leakage, and determine an RCS leak is in progress loca ted in the Chemical and Volume Control System on the Letdown line. This is an Al ternate Path JPM as the applicant must determine RCS Makeup intervals ar e NOT normal and isolate Letdown and Charging per Step 2.3.8 RNO. The applicant will then place Excess letdown in service in accordance with SOP-103A, Chem ical and Volume Control System, as directed by ABN-103. The Critical Steps are to isolate Letdown, charge to RCP Seals only, close the RCS Loop 4 Charging Valve, close ONE of the Charging Pump to RCS Isolation valves, and place Excess Letdown in service flushing to the RCDT. This is a bank JPM. This JPM is under the Chemical and Volume Control System - Reactor Coolant Inventory Control Safety Function. (K/A 004.A2.07 - IR 3.4 / 3.7) (13 Min)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 3 of 4 CPNPP 2017 NRC RO & SRO ES-301-2 SYSTEM JPM OUTLINE REV. 0 S-3 The applicants will transfer from Hot Leg Recirculation to Cold Leg Recirculation using EOS-1.4A, Transfer to Hot Leg Recircul ation, Attachment 2, Transfer to Cold Leg Recirculation from Hot Leg Recirculat ion, with a LOCA in progress. The Critical Steps are to close the Train A RHR Pump 1 Crosstie valve, open the RHR to Cold Leg 1 and 2 Injection Isolation valve, close the Train B RHR Pump 1 Crosstie valve, open the RHR to Cold Leg 3 and 4 Injection Isolation valve, close the RHR to Hot Leg 2 and 3 Injection Isolation valve, stop SI Pump 1, close the SI to Hot Leg 2 and 3 Injection Isolation valve, open the SI to Cold Leg 1 and 4 Injection Isolation valve, re-s tart SI Pump 1, stop SI Pump 2, close the SI to Hot Leg 1 and 4 Injection Isolation valve, and re-start SI Pump 2. This is a New JPM.

This JPM is under the Emergency Core Cooling System - Reactor Pressure Control Safety Function. (K/A 011.

EA1.11 - IR 4.2 / 4.2) (12 Min) S-4 With the unit at 100% power and Cont rol Rods in manual due to a previous Nuclear Instrument malfunction, the applic ant will respond to an inadvertent start of the TDAFW pump. This is an Alternate Path JPM as the applicant will attempt to isolate the TDAFW pump steam supply valve but it will fail to close. The applicant will verify the TDAFWP may be taken out of service and trip the pump. The Critical Steps are to place the Control Rods in Auto due to their previous failure, initiate a 50 MW Turbine Load Reduction, and trip the TDAFWP. This is a direct from bank JPM. This JPM is under the Auxiliary Feedwater System - Heat Removal From Reactor Core, Secondary Safety Function. Th is JPM is for the RO s ONLY. (K/A 061.A2.07 - 3.4 / 3.5) (9 Min) S-5 The applicants will perform the actions of EOP-0.0A, Reactor Trip or Safety Injection, Step 7, Verify Containment Spray Not Required. This is an Alternate Path JPM as containment pressure will have raised greater than 18 psig and several valves will have failed to operate on Phase B of Containment Isolation, also 2 Containment Spray pum ps will fail to auto start. The Critical Steps are to actuate Phase B from CB-07, open the Train A Chem Add Tank Disch valve, open the Train A Heat Exchanger Outlet valve, and secure all RCPs. This is a Modified bank JPM. This JPM is under the Cont ainment Spray System - Containment Integrity Safety Function. This is a PRA significant action.

(K/A 103.A2.03 - IR 3.

5 / 3.8) (8 min) S-6 With OPT-214A, Diesel Generator Operability Test in progress and following a fast start of Diesel Generator 1-01, the applicant is to continue with the surveillance and load DG 1-01 onto the Safeguards bus.

This is an Alternate Path JPM as when the DG is loaded to approximately 2.2 MW and the operator is adjusting VARS, unit 1 reactor will trip. This will cause the operator to follow the test termination guidance in Attachment 10.7 to open the DG output breaker and adjust voltage and frequency. The Critical Steps ar e to turn on the Synchroscope, close the DG Output Breaker, load the DG to 2.2 - 2.5 MW, adjust KVARS to maintain reactive load on the DG between 0 - 500 KVAR, and open the DG Output Breaker when the Reactor trips. This is a modi fied bank JPM. This JPM is under the Emergency Diesel Generator System -

Electrical Safety Function.

(K/A 064.A4.06 - IR 3.9 / 3.9) (15 min)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 4 of 4 CPNPP 2017 NRC RO & SRO ES-301-2 SYSTEM JPM OUTLINE REV. 0 S-7 The applicant will respond to a Turbine Impulse Pressure Instrument Malfunction per ABN-709, Steam Line, Steam Header, Turbine 1 st Stage Pressure, and Feed Header Pressure Instrument Malfunction. T he Critical Steps ar e to place Control Rods in manual, disable Steam Dumps, place Steam Dumps in Steam Pressure Mode in Auto, and transfer Turbine Impulse pressure to an Operable channel.

This is a bank JPM. This JPM is under the Non-Nuclear Instrumentation System -

Instrumentation Safety Function. (K/A 016.A2.01 - IR 3.0

/ 3.1) (10 min) S-8 The applicant will perform the actions for a Control Room fire in accordance with ABN-803A, Respond to a Fire in the C ontrol Room or Cable Spreading Room, Attachment 1, Reactor Operator Actions to Achieve Hot Shutdown. The Critical Steps are to trip the Reactor, trip the TDAFWP, isolat e Main Steam lines, isolate Letdown, open CCP suctions from the RWST, place CCP 1 in pull-out, secure all RCPs, place both RHR pumps in pull-out, and close the RWST to RHR pump 1

and 2 suction valves. This is a bank JPM.

This JPM is under the Fire Protection System - Plant Service Systems Safety Function.

(K/A 068.AK3.12 - IR 4.

1 / 4.5) (10 Min)

P-1 The applicant will perform actions to align boric acid gravity flow per ABN-107, Emergency Boration, Attachment 6, Bori c Acid Gravity Flow Valve Lineup. The Critical Steps include positio ning the appropriate valves to align gravity flow from Boric Acid Tank X-01 to Unit 1. This is a Time Critical JPM per STI-214.01, Control of Timed Operator Actions. This is a bank JPM. This JPM is under the Chemical and Volume Control System - Reacti vity Control Safety Function. (K/A 004.A4.18 - IR 4.3 / 4.1) (15 min) P-2 The applicant will utilize ECA-0.1A, Loss of All AC Power Recovery Without SI Required, Attachment 2, Alignment of PRZR Heaters With APGs Supplying AC Safeguards Bus to locally energize the co rrect number of PRZR Heater Groups based on current APG loading. The Critical Steps are to locally de-energize the PRZR Heater Groups, determine the maximu m number of heater groups that may be energized based on current APG loading (calculation), and re-energize the correct number of PRZR Heater Groups. Th is is a modified from bank JPM. This JPM is under the Loss of Offsite and Onsite Power (Station Blackout) System -

Electrical Safety Function (K/A 055.

EA2.03 - IR 3.9 / 4.7) (15 min)

P-3 During a Control Room evacuation due to a fire, the applicant is required to take action to control plant parameters from out side the control room.

Actions will be performed using ABN-803A, Response to a Fire in the Control Room or Cable Spreading Room, Attachment 4, Nuclear Equipment Oper ator No. 2 Actions to Achieve Hot Shutdown. The critical st eps include starting the Safety Chiller, Isolating RHR from the RWST and controlli ng AFW flow to the Steam Generators. This is a PRA significant action. This is a direct from bank JPM. This JPM is under the Control Room Evacuation System - Pl ant Service Systems Safety Function.

This is a Previous JPM from the 2015 NRC exam.

(K/A 068.AA1.26 - IR 3.

6 / 3.8) (10 min)

ES-301 Transient and Event Checklist Form ES-301-5 Page 1 of 2 CPNPP 2017 NRC ES-301-5 TRANSIENT AND EVENT CHECKLIST REV. 0 Facility: CPNPP 1 and 2 Date of Exam: 06/12/17 Operating Test No.: NRC A P P L I C A N T E V E N T T Y P E SCENARIOS CPNPP #1 CPNPP #2 CPNPP #3 (SPARE) CPNPP #4 T O T A L MINIMUM(*)

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 SRO-U1 RX - 0 1 1 0 NOR - 0 1 1 1 I/C 1,2,3,4,6,7 6 4 4 2 MAJ 5,8 2 2 2 1 TS 1,2,3, 4 4 0 2 2 SRO-I1 RX - 4 - 1 1 1 0 NOR - - - 0 1 1 1 I/C 1,2,3,4,5,8 1,3,7 2,4,5, 6 13 4 4 2 MAJ 6,7 5,8 7,8 6 2 2 1 TS 1,4 - 3,4 4 0 2 2 SRO-I2 RX - - 6 1 1 1 0 NOR - - - 0 1 1 1 I/C 2,3,5 1,2,3,4,6,7 2,4,9 12 4 4 2 MAJ 6,7 5,8 7,8 6 2 2 1 TS - 1,2,3, 4 - 4 0 2 2 SRO-I3 RX - 4 - 1 1 1 0 NOR - - - 0 1 1 1 I/C 1,2,3,4,5,8 1,3,7 2,4,5, 6 13 4 4 2 MAJ 6,7 5,8 7,8 6 2 2 1 TS 1,4 - 3,4 4 0 2 2 ES-301 Transient and Event Checklist Form ES-301-5 Page 2 of 2 CPNPP 2017 NRC ES-301-5 TRANSIENT AND EVENT CHECKLIST REV. 0 Facility: CPNPP 1 and 2 Date of Exam: 06/12/17 Operating Test No.: NRC A P P L I C A N T E V E N T T Y P E SCENARIOS CPNPP #1 CPNPP #2 CPNPP #3 (SPARE) CPNPP #4 T O T A L MINIMUM(*)

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 RO1 RX 5 - - 1 1 1 0 NOR - - 1 1 1 1 1 I/C 1,4,8 1,2,4, 6,9 4,5,6 11 4 4 2 MAJ 6,7 5,8 7,8 6 2 2 1 TS - - - 0 0 2 2 RO2 RX 5 - - 1 1 1 0 NOR - - 1 1 1 1 1 I/C 1,4,8 1,2,4, 6,9 4,5,6 11 4 4 2 MAJ 6,7 5,8 7,8 6 2 2 1 TS - - - 0 0 2 2 RO3 RX - 6 1 1 1 0 NOR - - 0 1 1 1 I/C 2,3,5 2,4,9 6 4 4 2 MAJ 6,7 7,8 4 2 2 1 TS - - 0 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 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 Instant SRO 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 instrument or component malfunctions on a 1-for-1 basis.

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

Appendix D Scenario Outline Form ES-D-1 Page 1 of 46 CPNPP NRC 2017 Scenario 1 Facility: CPNPP 1 & 2 Scenario No.: 1 Op Test No.: June 2017 NRC Examiners: Operators: Initial Conditions: 100% power MOL - RCS Boron is 924 ppm (by sample). MDAFW Pump 1-02 is out of service for an oil change. Turnover: Maintain steady state power conditions Critical Tasks: CT Ensure Control Rods inserting 48 Steps / Minute During Reactor Trip Failure Prior to Exiting FRS-0.1A, Response to Nuclear Power Generation / ATWT. CT Initiate Emergency Boration During Anticipated Transient Without Trip Prior to Exiting FRS-0.1A, Response to Nuclear Power Generation / ATWT. CT Identify and Isolate the Ruptured Steam Generator Prior to Commencing an Operator Induced Cooldown per EOP-3.0A, Steam Generator Tube Rupture.

Event No. Malf. No. Event Type* Event Description 1 RX08A I (RO,SRO) TS (SRO) PT-455 Pressurizer Pressure Channel (PT-455) fails high 2 RX02G C (BOP, SRO) SG 1-04 Steam Flow (FI-543A) Fails Low 3 CH10 C (BOP, SRO) CRDM Vent Fan #1 trips 4 RP05A I (RO, SRO) TS (SRO) NR Cold Leg TI (TE-411B) fails high 5 RC03C R (RO) C (BOP ,SRO)

Loss of B MFP 6 RC19C M (RO,BOP,SRO)

Loss of A MFP. Reactor fails to trip. Reactor trip breakers fail to open. Bus Breaker CS-1B4-1 Fails to Open 7 SG02D M (RO,BOP,SRO)

SG 1-04 Tube Rupture (2 tubes) 8 MS08D C, (RO,SRO) SG 1-04 MSIV Fails to Close * (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) 5 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) 3 Critical tasks (2-3)

Scenario Event Description NRC Scenario 1 Page 2 of 46 CPNPP NRC 2017 Scenario 1 SCENARIO 1

SUMMARY

Event 1 The first event is a failure of Pressurizer Pressure Channel PT-455 high. The crew will enter ABN-705, Pressurizer Pressure Malfunction, Section 2.0, Pressurizer Pressure Instrument Malfunction. The associated PORV will open and the operator will close the PORV, its associated Block Valve, and place 1-PK-455A, Master Pressurizer Pressure Controller in manual and control PZR pressure. The SRO will refer to Technical Specifications.

Event 2 1-FI-543A, SG 1-04 STM FLO, Selected Steam Flow transmitter fails Low. The crew will enter ABN-707, Section 2.0, Steam Flow Instrument Malfunction. The operators will take manual control of the affected FRV and master feed pump speed control. The alternate channel will be selected for control and the system will be returned back to automatic control.

Event 3 The operating CRDM vent fan trips. The crew will refer to 1-ALB-3A, Window 1.6, CRDM SHROUD EXH TEMP HI, and ensure that at least one CRDM v ent 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.

Event 4 Failure of Cold Leg Loop 1 NR Temperature Transmitter (TE-411B). It will fail high (630°F). The Reactor Operator will take action per ABN-704, Tc/N-16 Instrumentation Malfunction, Section 2.0. This event requires taking manual control of rods, since the Tc failure results in a higher Tave and rods will be inserting in automatic. The SRO will refer to Technical Specifications for this malfunction.

Event 5 Event 5 is the precursor to the major event and involves a trip of the main feed pump with a turbine runback (rod control is still in manual from the previous event). Operators will take action per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 2.0, and ramp the unit down. The second feed pump will trip 3 minutes after the first.

Event 6,7,8 After the loss of the 2 nd MFP a reactor trip is warranted and an attempt will be made to manually trip the Reactor via the Normal Trip Switches and by de-energizing both buses supplying the Control Rod Drive Mechanism Motor Generators. Operators will enter FRS-0.1A, Response To Nuclear Power Generation/ATWT. Operators will be required to drive control rods inward until the reactor trip breakers are opened locally. After the reactor is shutdown a tube rupture will occur on SG 1-04. Operators will exit FRS-0.1A; perform the actions of EOP-0.0A, Reactor Trip or Safety Injection, and transition to EOP-3.0A, Steam Generator Tube Rupture. A failure of the 1-04 MSIV to close will complicate the

event. Terminating Criteria Scenario will be terminated when the operators have completed an RCS cooldown, and an RCS depressurization has begun, or at the Examiner's discretion.

Scenario Event Description NRC Scenario 1 Page 3 of 46 CPNPP NRC 2017 Scenario 1 Risk Significance: Failure of risk important system prior to trip: Pressurizer Pressure Channel Fails high Main Feed Pump B Trips Risk significant core damage sequence: Main Feed Pump A Trips; ATWT Risk significant operator actions: Isolation of Ruptured Steam Generator complicated by MSIV failure to close

Scenario Event Description NRC Scenario 1 Page 4 of 46 CPNPP NRC 2017 Scenario 1 Critical Task Determination Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback Ensure Control Rods inserting 48 Steps / Minute During Reactor Trip Failure Prior to

Exiting FRS-0.1A, Response to Nuclear Power Generation / ATWT The safeguards systems that protect the plant during accidents are designed assuming that only decay heat and pump heat are being added to the RCS. DRPI lights indicating rods are withdrawn after both reactor trip switches have been turned, two red indicating lights lit for both reactor trip breakers after the reactor trip switches have been turned, power range detectors showing power greater than 5%. Procedurally driven from FRS-0.1A Observance of the RO verifying control rods are inserting 48 Steps / Minute in auto and when speed slows then rods are placed in manual and driven in DRPI indicating lights moving in the inward direction, rod speed indicator showing rod speed during the transient. After reactor trip breakers opened two green lights for the reactor trip breakers Initiate Emergency Boration During Anticipated Transient Without Trip Prior to Exiting FRS-0.1A, Response to Nuclear Power Generation / ATWT After control rod trip and rod insertion functions, boration is the next most direct manner of adding negative reactivity to the core. This must be started early in FRS-0.1A to add negative reactivity to the core to aid in shutting down the reaction in the core. DRPI lights indicating rods are withdrawn after both reactor trip switches have been turned, two red indicating lights lit for both reactor trip breakers after the reactor trip switches have been turned, power range detectors showing power greater than 5%. Procedurally driven from FRS-0.1A Observance of the red running lights for the boric acid pump and the CCP with charging flow as indicated on the meter greater than 30 gpm. Boric acid pump and Charging pump running with discharge flow of greater than 30 gpm Identify and Isolate the Ruptured Steam Generator Prior to Commencing an Operator Induced Cooldown per EOP-3.0A, Steam Generator Tube Rupture. Take one or more actions that would prevent a challenge to plant safety. STI-214.01, TCA-1.9; FSAR 15.6.3.1.1; WCAP-16871-P, Section 6.4; DBD-ME-027. (NOT TCA due to additional failure) Procedurally driven from EOP-3.0A, to identify and isolate a ruptured SG. Indications include MSL Radiation alarms

and SG level.

The operator will not be able to close the MSIV, so all other MSIVs must be closed.

The operator will ensure the FW isolation valves are closed, and reduce AFW flow to SG 1-04. SG pressure increasing, AFW flow reduced to zero and valve position indications.

Scenario Event Description NRC Scenario 1 Page 5 of 46 CPNPP NRC 2017 Scenario 1 SIMULATOR OPERATOR INSTRUCTIONS for SIMULATOR SETUP Initialize to IC18 and LOAD 2017 NRC Scenario 1.

EVENT TYPE MALF # DESCRIPTION DEMAND VALUE INITIATING PARAMETER SETUP IRF FWR021 MDAFWP 1-02 Breaker Racked Out f:0 K0 6 IMF RP15E Reactor Trip Breakers Jammed Closed f:1 K0 IOR DIED1B41 Bus Breaker CS-1B4-1 Fails to Open f:3 K0 8 IMF MS08D SG (1-04) MSIV Fails to Close f:1 K0 1 IMF RX08A Pressurizer Pressure Channel (PT-455) fails high f:2500 K1 2 IMF RX02G SG 1-04 Steam Flow (FI-543A) fails low f:0 K2 3 IMF CH10 CRDM Vent Fan #1 trips f:1 K3 4 IMF RP05A NR Cold Leg TI (TE-411B) fails high f:630 K4 5 IMF FW03B Main Feedwater Pump B trip f:1 K5 6 IMF FW03A Main Feedwater Pump A trip f:1 K5 +180 IMF RP15E Reactor Fails to trip -Reactor trip breakers jammed closed f:1 K0 IOR DIED1B41 Bus Breakers CS-1B4-1 Fails to open f:3 K0 6 IRF RPR112 Locally open Reactor Trip Breaker Train A f:2 K10 IRF RPR113 Locally open Reactor Trip Breaker Train B f:2 K10 7 IMF SG02D SG (1-04) Tube Rupture (2 tubes) f:2 (1) 8 IMF MS08D SG (1-04) MSIV fails to close f:1 K0 (1) {LORPRTBAL_1.Value} IMF SG02D f:2 r:60 Tube rupture will be set to actuate upon the RTB lights changing from red to green (60 second ramp)

Page 1 of 40 CPNPP NRC 2017 Scenario 2 Appendix D Scenario Event Description Form ES-D-1 Facility: CPNPP 1 & 2 Scenario No.: 2 Examiners: Operators: Op Test No.: June 2017 NRC Initial Conditions: 100% power MOL - RCS Boron is 924 ppm (by sample). MDAFWP 1-02 is out of service for scheduled maintenance. Turnover: Maintain steady-state power conditions. Severe weather has been reported in the area, Unit 2 has entered ABN-907, Acts of Nature.

Pressurizer Steam Space Sample is in progress by Chemistry. Critical Tasks: CT Isolate Reactor Coolant System Leakage Paths in accordance with ECA-0.0A, Loss of All AC Power prior to exiting ECA-0.0A. CT Restore Power to Bus 1EA1 in accordance with ECA-0.0A, Loss of All AC Power, prior to exit from ECA-0.0A. CT Manually Start Safety Injection Pump 1-01 in accordance with EOP-0.0A, Attachment 2 or EOP-1.0A, Attachment 1A prior to exiting EOP-1.0A, Loss of Reactor or Secondary Coolant. Event No. Malf. No. Event Type* Event Description 1 ED07A C (RO, BOP, SRO)

TS (SRO) Loss of Inverter (IV1PC1) 2 SW01B C (BOP, SRO) TS (SRO) SSW Pump 1-02 trips 3 CV16A I (RO, SRO) TS (SRO) VCT Level Channel LT-112 Fails Low 4 FW14B TC09I RD15A R (RO) C (BOP, SRO) TS (SRO) Heater Drain Pump 1-02 Trip Automatic Turbine Runback Failure Rods fail to control in automatic 5 ED01 M (RO, BOP, SRO)Loss of All AC Power Due to Loss of Offsite Power 6 EG15A C (BOP, SRO) Emergency Diesel Generator 1-01 fails to start Emergency Diesel Generator 1-02 in pull-out due to SSW pump trip 7 OVRD C (RO, SRO) Pressurizer Steam Space Sample Valves (1/1-4165A & 1/1-4176A) fail to auto close. Manual closure required.

8 RC19A M (RO, BOP, SRO)LOCA on DG 1-01 Emergency Start 9 SI04C C (BOP) SI pump 1-01 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) 3 Critical tasks (2-3)

Page 2 of 40 CPNPP NRC 2017 Scenario 2 Scenario Event DescriptionNRC Scenario 2SCENARIO 2

SUMMARY

The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. Severe weather has been reported in the area. MDAFWP 1-02 is tagged out for scheduled maintenance. A Pressurizer Steam Space sample is in progress. Event 1 The first event is a loss of Inverter IV1PC1, 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 Specifications and determine that TS 3.8.9 is applicable during the loss and exited upon power restoration. Event 2 The next event is a trip of Station Service Water Pump 1-02. The crew will enter ABN-501, Section 2.0, Station Service Water Pump Trip. Various equipment controls, as directed by ABN-501, are placed in PULL OUT to prevent starting with no cooling water available. The SRO will refer to Technical Specifications. Event 3 VCT level channel LT-112 will fail low. This will result in an automatic makeup. The RO will respond in accordance with the ALM and stop the auto makeup. The crew will refer to ABN-105, Chemical and Volume Control System Malfunction to place the makeup system into manual alignment until automatic control is restored. Event 4 The next event is a trip of a Heater Drain Pump with an automatic turbine runback failure. The crew responds per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section, 4.0. When it is determined that automatic plant response has not activated, control rods are placed/verified in auto and a manual Turbine Runback will be initiated. The control rods will fail to operate in auto, and must be manually controlled by the RO. The crew will stabilize load at 700 MWe. During this event, control rod position may drop below the Rod Insertion Limit (RIL) and when informed, the SRO will refer to Technical Specifications. Events 5, 6, 7 The major event is a Loss of Offsite Power with a failure of the 1-01 Diesel Generator to automatically start. Operators will perform an emergency start of the 1-01 DG 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. Event 8 A LOCA will occur (delayed by 120 seconds) when the 1-01 DG is emergency started. SI pump1-01 fails to auto-start from the SI sequencer; it is a critical task to manually start the only available SI pump. Depending on plant conditions entry into FRZ-0.1, Response To High Containment Pressure and FRP-0.1, Response to Imminent Pressurized Thermal Shock Condition may be required, however the actions of these procedures will not be substantive. Termination Criteria This scenario is terminated when operators have performed the action of EOP-1.0, Loss of Reactor or Secondary Coolant, and transition to EOS-1.3 A, Transfer to Cold Leg Recirculation, or at the Lead Examiners' discretion.

Page 3 of 40 CPNPP NRC 2017 Scenario 2 Scenario Event DescriptionNRC Scenario 2 Risk Significance: Failure of risk important system prior to trip: Loss of Inverter IV1PC1 Loss of SSWP 1-02 / DG 1-02Turbine Runback Failure Risk significant core damage sequence:LOCA Failure of SI pump 1-01 Risk significant operator actions: Isolate RCS Leakage Paths Restore Safeguards Bus Manual Start of SI pump 1-01

Page 4 of 40 CPNPP NRC 2017 Scenario 2 Scenario Event DescriptionNRC Scenario 2 Critical Task DeterminationCritical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback Isolate Reactor Coolant System Leakage Paths in accordance with ECA-0.0A, Loss of All AC Power prior to exiting ECA-0.0A.

Take one or more actions that would prevent a challenge to plant safety. Procedural direction at ECA-0.0A Step 3 to minimize RCS inventory loss. Valve position indication and letdown flow.

The operator will manually close the Letdown Isolation Valves and Primary Sample Isolation Valves. Valve position will change and letdown flow will lower to zero.

MLB indication for closed valve position. Restore Power to Bus 1EA1 in accordance with ECA-0.0A, Loss of All AC Power, prior to exit from ECA-0.0A.

Recognize a failure or an incorrect automatic actuation of an ESF system or component resulting in degraded ECCS capacity.

Procedural direction at ECA-0.0A Step 5 to restore power via EDG 1-01 to Safeguard Bus

1EA1. Bus voltage indication and EDG parameters.

The operator will manually perform an emergency start on EDG 1-01 using the handswitch on CB-11. Indication of DG running and loading via bus voltage and frequency. Manually Start Safety Injection Pump 1-01 in accordance with EOP-0.0A or EOP-1.0A, Attachment 1A prior to exiting EOP-1.0A, Loss of Reactor or Secondary Coolant.

Recognize a failure or an incorrect automatic actuation of an ESF system or component. Procedural direction in EOP-0.0A, Attachment 2 to verify SI Pumps running. Also procedural direction in EOP-1.0A, Attachment 1A to manually start ECCS pumps as necessary to maintain PRZR level. SI Pump 1-02 on this case has no power, therefore SI Pump 1-01 must be manually started to provide makeup flow to the RCS as this is a LOCA of such size where SIPs are required.

The operator will start SI Pump 1-01 using the handswitch on CB-02. Indication pump start including light

indication, flow and discharge pressure on CB-02.

Page 5 of 40 CPNPP NRC 2017 Scenario 2 Scenario Event DescriptionNRC Scenario 2 SIMULATOR OPERATOR INSTRUCTIONS for SIMULATOR SETUP INITIALIZE to IC-18 and LOAD 2017 NRC Scenario 2.

EVENT TYPE MALF # DESCRIPTION DEMAND VALUE INITIATING PARAMETER SETUP IRF FWR021 MDAFWP 1-02 Breaker Racked Out f:0 K0 4 IMF TC09I Automatic Turbine Runback Failure f:1 K0 6IMF EG15A DG 1-01 Fails to Auto Start f:1 K0 7IOR LOANMLB 1 A 2_1 PSS Valve MLB Lights 1-4165A f:1 K0 IOR LOANMLB 1B2_1 PSS Valve MLB Light 1-4176A f:1 K0 9 IMF SI04C SI 1-01 pump fails to start on sequencer f:1 K0 1 IMF ED07A Loss of Inverter (IV1PC1) f:1 K1 1 IRF EDR01 Transfer 1PC1 to alternate power f:0 K10 2 IMF SW01B Loss of SSW Pump 1-02 f:1 K2 3 IMF CV16A VCT Level Channel LT-112 Fails Low f:0 K3 4 IMF FW14B Heater Drain Pump 1-02 Trip f:1 K4 4 IMF RD15A Rods fail to move in Auto f:1 K4 4 IMF TC09I Automatic Turbine Runback Failure f:1 K0 5 IMF ED01 Loss of Offsite Power f:1 K5 6 IMF EG15A Diesel Generator 1-01 Fails to Auto Start f:1 K0 7 IMF OVRD PRZR Steam Space Sample Valves (1/1-4165A & 1/1-4176A) Failure f:1 K0 8 IMF RC19A LOCA linked to DG Emergency Start {DIEG1DG1E.Value=4} f:5000 +120 9 IMF SI04C SI 1-01 pump fails to start on sequencer f:1 K0 Appendix D Scenario Outline Form ES-D-1 Page 1 of 40 CPNPP NRC 2017 Scenario 3 Facility: CPNPP 1 & 2 Scenario No.: 3 Op Test No.: June 2017 NRC Examiners: Operators: Initial Conditions: 1 x 10

-8 amps following a refueling outage. MDAFWPs are maintaining Steam Generator Water Levels 60-75%. Steam dumps are in Steam Pressure mode. Boron is 1669 ppm (by sample). Turnover: Raise power to 3% per IPO-002A, Plant Startup From Hot Standby, Section 5.4 Critical Tasks: CT 1 - Initiate a MSLI or Manually close MSLI valves prior to exiting EOP-0.0A, Reactor Trip or Safety Injection, or EOP-2.0, Faulted Steam Generator Isolation.

CT 2 - Trip reactor coolant pumps 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)

N (BOP, SRO) Raise power to 2% to 3%

2 TP06A TP07B C (BOP, SRO) Turbine Plant Cooling Water Pump 1 Trip Turbine Plant Cooling Water Pump 2 Failure to Auto-Start 3 OVRD C (RO, SRO) Letdown HX Outlet flow controller Failure TK-130 fails low 4 RX08B RX16B C (RO, SRO) TS (SRO)) PT-456 PZR Pressure Transmitter fails high, PORV PCV-456 fails 25% open 5 FW24B C (BOP, SRO) TS (SRO) AFW Pump 1-02 trips, manual start of TDAFW Pump required 6 RD09B6 RD04B6 RD04F6 M (ALL) Seismic event, Ejected rod, SBLOCA @ 200 gpm, Stuck rod 7 MS02 M (ALL) Main Steam line leak downstream of the MSIVs (MSLI does not occur automatically) * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 7 Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 2 Major transients (1-2) 2 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

Scenario Event Description NRC Scenario 3 Page 2 of 40 CPNPP NRC 2017 Scenario 3 SCENARIO 3

SUMMARY

Event 1 In accordance with turnover instructions, the crew begins raising power to 2% to 3%, per IPO-002A, Plant Startup from Hot Standby, Section 5.4, Increasing Reactor Power to Approximately 2% Following Reactor Startup and Establishing Main Feedwater Flow to the SGs.

Event 2 When the lead examiner is satisfied with the power increase (stable between 2-3%) a trip of the running TPCW Pump will occur. The standby pump will fail to automatically start and manual operator action will be required to start the standby pump. Crew response will be per ABN-306, Turbine Plant Cooling Water System Malfunction, Section 3.0. The crew will start the standby pump and verify other parameters for the system.

Event 3 The next event is a failure of the Letdown Heat Exchanger Outlet Flow Controller, TK-130. The controller output will fail to zero demand and cause TCV-4646, LTDN HX OUT TEMP CTRL valve to close. This will result in Letdown Heat Exchanger High temperature alarms and Letdown flow will fail to divert to the VCT on high temperature. The crew will respond per the ALM, take manual control of

TK-130 and raise demand to establish Letdown Heat Exchanger Outlet temperature to approximately 95°F, and manually divert letdown flow to the VCT.

Event 4 Pressurizer Pressure channel PT-456 will fail high. PORV PCV-456 will open and when closed will stick at 25% open. The crew will enter ABN-705, Section 2.0, Pressurizer Pressure Instrument Malfunction.

The primary action is to close the PORV block valve. The SRO will refer to Technical Specifications.

Event 5 After the crew has control of RCS pressure, the Motor Driven Auxiliary Feedwater Pump (MDAFWP) 1-02 will trip. The crew will enter ABN-305, Auxiliary Feedwater System Malfunction. The crew will manually start the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) and feed Steam Generators 1-03 and 1-04 with the TDAFWP. The SRO will refer to Technical Specifications.

Event 6 A seismic event occurs; this is a precursor for upcoming events. The crew will enter ABN-907, Acts of Nature, Section 2.0, Earthquake. 120 seconds after the seismic annunciators have come in Control Rod B6 will partially eject from the core (SBLOCA) and Control Rod F6 will stick at 228 steps on the reactor trip. The reactor will trip and the crew will enter EOP-0.0A, Reactor Trip or Safety Injection.

Event 7 Main steam line leak in the turbine building (downstream of the MSIVs,) as a result of the seismic event) requiring the MSIVs to be manually closed as they will fail to close automatically.

Terminating Criteria Scenario will be terminated when the operators have transitioned to EOS-1.2, Post LOCA Cooldown and Depressurization, or at the Lead Examiner's discretion.

Scenario Event Description NRC Scenario 3 Page 3 of 40 CPNPP NRC 2017 Scenario 3 Risk Significance: Failure of risk significant systems prior to trip:

PORV stuck open MDAFW Pump trips Risk significant core damage sequence: Rod Ejection then Small Break LOCA Main Steam Line Break

Risk significant operator actions: Emergency Boration for two stuck control rods Manual Main Steam line Isolation

Scenario Event Description NRC Scenario 3 Page 4 of 40 CPNPP NRC 2017 Scenario 3 Critical Task Determination Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback Manually isolate

Main Steam lines prior to exiting EOP-0.0A, Reactor Trip or Safety

Injection, or EOP-2.0A, Faulted Steam Generator Isolation. Take one or more actions that would prevent a challenge to plant safety. SG pressure along with RCS pressure and temperature falling.

The operator will manually close the MSIVs from CB-07. All MSIV valve light indications will change from Red lit to Green lit and steam flow will go to zero for SGs. Trip reactor coolant pumps 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.

Take one or more actions that would prevent a challenge to plant safety. FSAR II.K.3.5; WCAP-9584; WOG ERG Generic Issue for RCP Trip / Restart. Procedurally driven from EOP-0.0A and EOP-1.0A Foldout pages.

Availability of Subcooling indication both on meters and computer.

The operator will secure ALL RCPs using the handswitches on CB-05. Indication of pump stop including light indication, flow and motor current.

Scenario Event Description NRC Scenario 3 Page 5 of 40 CPNPP NRC 2017 Scenario 3 SIMULATOR OPERATOR INSTRUCTIONS for SIMULATOR SETUP INITIALIZE to IC 8 and LOAD 2017 NRC Scenario 3.

EVENT TYPE MALF # DESCRIPTION DEMAND VALUE INITIATING PARAMETER 2 IMF TP07B Turbine Plant Cooling Water Pump 2 Fail to Auto-Start f:1 K0 7 IMF SS02A1 MSL Isolation Train A Master Relay Failure f:1 K0 7 IMF SS02A2 MSL Isolation Train B Master Relay Failure 2 IMF TP06A Turbine Plant Cooling Water Pump 1 Trip f:1 K2 2 IMF TP07B Turbine Plant Cooling Water Pump 2 Fail to Auto-Start f:1 K0 3 IOR OVRD Letdown HX Outlet Flow Controller Failure (TK-130) Fails Low, with a failure of TCV-129 to divert f:10 OVRD K3 4 IMF RX08B PT-456 PZR Pressure Transmitter fails high f:2500 K4 4 IMF RX16B PORV PCV-456 fails 25% open. f:25 K4 + 4 4 IRF RCR24 PORV Block Valve breaker f:0 K11 5 IMF FW24B AFW Pump 1-02 trips f:1 K5 6 IRF AN2A_02 Seismic Event f:4 K6 AN2A_03 Seismic Event f:4 K6 IMF RD09B6 Ejected Rod B6 f:228 K6 + 120 RD04B6 Stuck Rod B6 (ejected - for indication only) f:228 K6 + 120 RD04F6 Stuck Rod F6 f:228 K6 +120 RC19C SBLOCA f:200 K6 + 120 (1) 7 IMF MS02 Main Steam Line leak downstream of the MSIVs f:1e+006 K6 + 210 7 IMF SS02A1 MSL Isolation Train A Master Relay Failure f:1 K0 7 IMF SS02A2 MSL Isolation Train B Master Relay Failure (1) {DIRPSIA2.Value=1} MMF RC19C f:800 r:60 Modify SBLOCA to 800 gpm on SI Initiation (60 sec ramp)

Appendix D Scenario Outline Form ES-D-1 Page 1 of 28 CPNPP NRC 2017 Scenario 4 Facility: CPNPP 1 & 2 Scenario No.: 4 Op Test No.: June 2017 NRC Examiners: Operators: Initial Conditions: 92% power MOL - RCS boron is 996 ppm. Power reduced for Governor Valve testing. MDAFW Pump 1-02 is out of service for an oil change. (IC-49) Turnover: Maintain 92% power conditions. Place RWST on recirculation using Containment Spray pump 1-01.

Critical Tasks: CT Trip all Reactor Coolant Pumps in accordance with FRH-0.1A, Response To Loss Of Secondary Heat Sink, prior to Initiating Bleed and Feed Cooling.

CT Initiate RCS Feed and Bleed in accordance with FRH-0.1A, Response To Loss Of Secondary Heat Sink, such that the RCS depressurizes sufficiently for Intermediate Head Injection to occur, prior to all Steam Generator Wide Range level lowering to 0%. Event No. Malf. No. Event Type* Event Description 1 - N (BOP) Recirculate the Refueling Water Storage Tank with Containment Spray Pump 1-01. 2 MS13C I (RO, SRO) SG 1-03 Steam Line Pressure Fails High (PT-2327) - ARV Opens 3 CS02A TS (SRO) Containment Spray Pump (1-01) Trip.

4 NI04E I (RO, BOP, SRO) TS (SRO) NI42 Power Range Channel fails high.

5 CH03 C (BOP, SRO) Neutron Detector Well Fan 9 trips on motor overload 6 FW22 R (RO) C (BOP, SRO) Low Pressure Feedwater Heater Bypass Valve (PV-2286) Fails Open. 7 FW20A M (RO, BOP, SRO) 1-01Condensate Pump trips; requiring a manual reactor trip.

8 ED05H FW09A M (RO, BOP, SRO) Loss of 6.9KV Bus 1EA1 (86-1 relay) when Generator Output Breakers Open TDAFW Pump trips on overspeed, Loss of all AFW 9 RX16B C (RO, SRO) PORV 456 fails to open manually or automatically * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 9 Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 5 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) 2 Critical tasks (2-3)

Scenario Event Description NRC Scenario 4 Page 2 of 28 CPNPP NRC 2017 Scenario 4 SCENARIO 4

SUMMARY

Event 1 As directed by the turnover the crew will recirculate the Refueling Water Storage Tank (RWST) using Containment Spray Pump 1-01 per SOP-204A, Containment Spray System, Section 5.1.3, Recirculation through the Recirculation Header.

Event 2 SG ARV PT-2327 fails high opening the ARV. Crew actions are per ABN-709, STM Line, STM HDR and Turbine 1 st Stage Press, Feed HDR Press, Instrument Malfunction. The crew will respond by checking STM Line pressures against set point and determining that it is open below set point and should not be. The RO will place the ARV in manual and close the valve.

Event 3 When conditions are stable, Containment Spray Pump 1-01 will trip. Actions are per ALM-0022A, 1-ALB-2B, Window 1.3 - ANY CSP OVRLD TRIP. The SRO will refer to Technical Specifications.

Event 4 Event 4 is a failure high of NI42 Power Range Channel. The crew will enter ABN-703, Power Range Instrumentation Malfunction. Since the failure is in the high direction, rods will be rapidly inserting. This will require the operator to place rod control to Manual, per Step 1.b of ABN-703. The SRO will refer to Technical Specifications.

Event 5 Event 4 will be a trip of the running Neutron Detector Well Fan #9. This will alarm 2.1 CNTMT FN MASTER TRIP. The ALM will direct the crew to determine which fan has tripped and start the other fan as required using SOP-801A, Containment Ventilation System. The crew will place the tripped fan handswitch in Pull Out or Stop as applicable.

Event 5 The Low Pressure Heater Bypass Valve fails open. Entry into ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 7.0, is required and Rod Control is returned to AUTO and a Manual Turbine Runback to 900 MWe is performed. During this event, Control Rod position may drop below the Rod Insertion Limit (RIL) and when informed, the SRO will refer to Technical Specifications

Event 6 Major Event, Condensate Pump 1-01 trips. Both Main Feedwater Pumps trip and the reactor will be

manually tripped. The crew will enter EOP-0.0A, Reactor Trip or Safety Injection.

Event 7 The crew will experience a loss of Bus 1EA1. This will occur at the same time the Main Generator Breaker opens on the unit trip. With MDAFW Pump 1-02 tagged out, there are no motor driven AFW pumps available. There are no Main Feedwater Pumps or Condensate Pumps available. The only source of feedwater will be the Turbine Driven AFW Pump.

Events 8 & 9 The TDAFW pump will trip on overspeed, leaving no viable source of feedwater and when Heat Sink is lost the crew will transition to FRH-0.1A, Response to Loss of Secondary Heat Sink. The step for checking that both Centrifugal Charging Pumps are available will be answered with a "NO", requiring tripping of all Reactor Coolant Pumps and to initiate bleed and feed. One PORV will fail to open; this will require all reactor vessel and pressurizer head vents to be opened.

Scenario Event Description NRC Scenario 4 Page 3 of 28 CPNPP NRC 2017 Scenario 4 Termination Criteria The scenario will be terminated when bleed and feed is initiated in accordance with FRH-0.1A; or at the discretion of the lead examiner.

Risk Significance:

Failure of risk important system prior to trip: Loss of Main Feedwater Pumps due to Loss of Condensate Pumps Risk significant core damage sequence: Loss of one Safeguards Bus (1EA1) TDAFW Pump trips on overspeed Risk significant operator actions: Restore Pressurizer Pressure Control Manually trip reactor on loss of all feedwater Initiate bleed and feed

Scenario Event Description NRC Scenario 4 Page 4 of 28 CPNPP NRC 2017 Scenario 4 Critical Task Determination Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback Trip all Reactor Coolant Pumps in accordance with FRH-0.1A, Response To Loss Of Secondary Heat Sink, prior to

Initiating Bleed and Feed Cooling. Without a source of water to provide a heat sink on the secondary side of

the SGs, RCPs are tripped to extend the effectiveness if the remaining water inventory in the SGs.

Procedural direction at FRH-0.1A Step 2 RNO a. to immediately stop all RCPs. The operator will manually stop RCPs using the handswitches on CB-05. Control board light and flow indications, along with loss of flow annunciators that the RCPs have stopped. Initiate RCS Bleed and Feed in accordance with

FRH-0.1A, Response To Loss Of Secondary Heat Sink, such that the RCS depressurizes sufficiently for Intermediate-Head Injection to occur, prior to all Steam Generator Wide Range levels

lowering to 0%.

Actuating SI ensures feed path of cool water to the RCS and isolates the containment to confine any RCS releases from the bleed flow. The bleed flow through a PORV/Vent valves will ensure that enough cool water will feed from the ECCS flow path to remove sufficient decay heat.

AFW flow will not be indicated on any AFW flow meter. Also no AFW pumps will be running. A RED path showing on CSFST for heat sink. The need for a heat sink as indicated by RCS temperature and pressure.

Actuated SI, ensured at least one CCP and SI pump is running with flow indicated providing a feed

path for the RCS. PRZR PORV as well as PRZR and Vessel vent valves open providing a bleed path for the RCS. Flow indicated on both a CCP and an SI pump. PRZR PORV open with block valve open. PRZR and Vessel vents open. RCS pressure lowering and CETs will indicate core cooling.

Scenario Event Description 2017 NRC Exam Scenario 4 Page 5 of 28 CPNPP NRC 2017 Scenario 4 SIMULATOR OPERATOR INSTRUCTIONS for SIMULATOR SETUP INITIALIZE to IC 49 and LOAD NRC Scenario 4.

EVENT TYPE MALF # DESCRIPTION DEMAND VALUE INITIATING PARAMETER SETUP IRF FWR021 MDAFWP 1-02 Breaker Racked Out f:0 K0 8 IMF ED05H Bus 1EA1 86-1 lockout. f:1 (1) IMF FW09A TDAFW Pump trips on overspeed. {LORPRTBAL_1.Value=1} IMF FW09A f:1 Rx Trip + 480 9 IMF RX16B PORV 456 fails to open manually or automaticallyf:1 K0 1 - - Recirculate RWST with CSP 1 - 2 IMF MS13C SG ARV PT-2327 Fails High - ARV opens f:1300 K2 3 IMF CS02A Containment Spray Pump 1-01 Trip f:1 K3 4 IMF NI04E NI42 Power Range Channel fails high. f:200 K4 5 IMF CH03 Neutron Detector Well Fan 9 trips on motor overload f:1 K5 6 IMF FW22 Low Pressure Feedwater Heater Bypass Valve (PV-2286) fails open. f:1 K6 7 IMF FW20B Condensate Pump 1-01 trips f:1 K7 8 IMF ED05H Bus 1EA1 86-1 lockout. f:1 (1) IMF FW09A TDAFW Pump trips on overspeed. {LORPRTBAL_1.Value=1} IMF FW09A f:1 Rx Trip + 480 (2) 9 IMF RX16B PORV 456 fails to open manually or automaticallyf:1 K0 (1) {LOEGW3_1.Value=1} IMF ED05H f:1 Inserts ED05H when Gen. Output Bkrs open (2) {LORPRTBAL_1.Value=1} IMF FW09A Trip TDAFWP 480 seconds after Rx Trip