ML18064A354

DC 2018-01 Final Outlines

ML18064A354
Person / Time
Site:	Diablo Canyon
Issue date:	02/09/2018
From:	Vincent Gaddy Operations Branch IV
To:	Pacific Gas & Electric Co
References
Download: ML18064A354 (50)

Text

Form ES-401-2 Facility: DCPP Date of Exam: 2018-01 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 3 3 3 3 3 3 18 6 Emergency &

Abnormal 2 2 2 2 N/A 1 1 N/A 1 9 4 Plant Evolutions Tier Totals 5 5 5 4 4 4 27 10 1 3 2 2 4 2 2 2 3 3 3 2 28 5 2.

Plant 2 2 0 1 2 1 0 1 1 0 1 1 10 3 Systems Tier Totals 5 2 3 6 3 2 3 4 3 4 3 38 8

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 7 Categories 3 3 2 2 10 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

[Outline developed using Rev. 2 Supp. 1 of NUREG-1122, K/A Catalog for Nuclear Power Plant Operators: PWR, the latest revision of the K/A catalog available at the time of outline generation (2/16/2017), per NUREG 1021 ES-401 D.1.b.]

Rev 1

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

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 2.4.3 Ability to identify post-accident 000008 Pressurizer Vapor Space X instrumentation. 3.7 1/39 Accident / 3 (CFR: 41.6 / 45.4) 2.4.21 Knowledge of the parameters and 000009 Small Break LOCA / 3 X logic used to assess the status of safety 4.0 2/40 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) 000011 Large Break LOCA / 3 AK3.07 Knowledge of the reasons for the 000015/17 RCP Malfunctions / 4 X following responses as they apply to the 4.1 3/41 Reactor Coolant Pump Malfunctions (Loss of RC Flow): Ensuring that S/G levels are controlled properly for natural circulation enhancement.

(CFR 41.5, 41.10 / 45.6 / 45.13)

AA1.08 Ability to operate and / or monitor the 000022 Loss of Rx Coolant Makeup / 2 X following as they apply to the Loss of Reactor 3.4 4/42 Coolant Makeup: VCT level (CFR 41.7 / 45.5 / 45.6)

AK1.01 Knowledge of the operational 000025 Loss of RHR System / 4 X implications of the following concepts as 3.9 5/43 they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation (CFR 41.8 / 41.10 / 45.3)

AA2.04 Ability to determine and interpret the 000026 Loss of Component Cooling X following as they apply to the Loss of 2.5 6/44 Water / 8 Component Cooling Water: The normal values and upper limits for the temperatures of the components cooled by CCW (CFR: 43.5 / 45.13)

AK2.03 Knowledge of the interrelations 000027 Pressurizer Pressure Control X between the Pressurizer Pressure Control 2.6 7/45 System Malfunction / 3 Malfunctions and the following: Controllers and positioners (CFR 41.7 / 45.7)

EA1.12 Ability to operate and monitor the 000029 ATWS / 1 X following as they apply to a ATWS: M/G set 4.1 8/46 power supply and reactor trip breakers (CFR 41.7 / 45.5 / 45.6)

EK1.01 Knowledge of the operational 000038 Steam Gen. Tube Rupture / 3 X implications of the following concepts as 3.1 9/47 they apply to the SGTR: Use of steam tables (CFR 41.8 / 41.10 / 45.3)

Rev 1

AK2.02 Knowledge of the interrelations 000040 (BW/E05; CE/E05; W/E12) X between the Steam Line Rupture and the 2.6 10/48 Steam Line Rupture - Excessive Heat Transfer / 4 following: Sensors and detectors (CFR 41.7 / 45.7)

AK3.04 Knowledge of the reasons for the 000054 (CE/E06) Loss of Main X following responses as they apply to the Loss 4.4 11/49 Feedwater / 4 of Main Feedwater (MFW): Actions contained in EOPs for loss of MFW (CFR 41.5,41.10 / 45.6 / 45.13) 000055 Station Blackout / 6 AA2.45 Ability to determine and interpret the 000056 Loss of Off-site Power / 6 X following as they apply to the Loss of Offsite 3.6 12/50 Power: Indicators to assess status of ESF breakers (tripped/not-tripped) and validity of alarms (false/not-false)

(CFR: 43.5 / 45.13)

AA1.06 Ability to operate and / or monitor the 000057 Loss of Vital AC Inst. Bus / 6 X following as they apply to the Loss of Vital 3.5 13/51 AC Instrument Bus: Manual control of components for which automatic control is lost (CFR 41.7 / 45.5 / 45.6) 2.2.22 Knowledge of limiting conditions for 000058 Loss of DC Power / 6 X 4.0 14/52 operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2)

Replaced with KA - G2.2.37 Ability to determine operability and/or availability of 3.6 safety related equipment.

AK3.01 Knowledge of the reasons for the 000062 Loss of Nuclear Svc Water / 4 X following responses as they apply to the Loss 3.2 15/53 of Nuclear Service Water: The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the nuclear service water coolers (CFR 41.4, 41.8 / 45.7 )

Replaced with AK3.03 Knowledge of the 4.0 reasons for the following responses as they apply to the Loss of Nuclear Service Water:

Guidance actions contained in EOP for Loss of nuclear service water AA2.08 Ability to determine and interpret the 000065 Loss of Instrument Air / 8 X following as they apply to the Loss of 2.9 16/54 Instrument Air: Failure modes of air-operated equipment.

(CFR: 43.5 / 45.13)

W/E04 L OCA Outside Containment / 3 EK1.3 Knowledge of the operational W/E11 Loss of Emergency Coolant X implications of the following concepts as they 3.6 17/55 Recirc. / 4 apply to the (Loss of Emergency Coolant Recirculation) Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Emergency Coolant Recirculation).

(CFR: 41.8 / 41.10 / 45.3)

BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 Rev 1

AK2.07 Knowledge of the interrelations 000077 Generator Voltage and Electric X between Generator Voltage and Electric Grid 3.6 18/56 Grid Disturbances / 6 Disturbances and the following: Turbine /

generator control.

(CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18 Rev 1

ES-401 3 Form ES-401-2 od E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 AK1.18 Knowledge of the operational X 3.4 19/57 DCPP Bank - P-40331 implications of the following concepts as they apply to Continuous Rod Withdrawal: Fuel temperature coefficient.

(CFR 41.8 / 41.10 / 45.3) 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 AK2.02 Knowledge of the interrelations X 2.5 20/58 between the Inoperable / Stuck Control Rod and the following: Breakers, relays, disconnects, and control room switches (CFR 41.7 / 45.7) 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 AK3.01 Knowledge of the reasons for X 3.2 21/59 the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: Startup termination on source-range loss (CFR 41.5,41.10 / 45.6 / 45.13) 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 2.1.32 Ability to explain and apply X 4.0 22/60 system limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 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 AK2.03 Knowledge of the interrelations X 2.8 23/61 between the Loss of Containment Integrity and the following: Personnel access hatch and emergency access hatch (CFR 41.7 / 45.7) 000074 (W/E06&E07) Inad. Core Cooling / 4 EA2.03 Ability to determine or interpret X 3.8 24/62 the following as they apply to Inadequate Core Cooling: Availability of turbine bypass valves for cooldown (CFR 43.5 / 45.13) 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 Rev 1

W/E13 Steam Generator Over-pressure / 4 EA1.1 Ability to operate and / or 25/63 X 3.1 monitor the following as they apply to the (Steam Generator Overpressure)

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

(CFR: 41.7 / 45.5 / 45.6)

W/E15 Containment Flooding / 5 EK3.2 Knowledge of the reasons for X 2.8 26/64 the following responses as they apply to Containment Flooding: Normal, abnormal and emergency operating procedures associated with Containment Flooding.

(CFR: 41.5 / 41.10, 45.6, 45.13)

W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA 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 EK1.3 Knowledge of the operational X 3.5 27/65 implications of the following concepts as they apply to Pressurized Thermal Shock: Annunciators and conditions indicating signals, and remedial actions associated with Pressurized Thermal Shock.

(CFR: 41.8 / 41.10, 45.3)

CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 2 2 2 1 1 1 Group Point Total: 9 Rev 1

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K4.04 Knowledge of RCPS design 003 Reactor Coolant Pump X 2.8 28/1 feature(s) and/or interlock(s) which provide for the following: Adequate cooling of RCP motor and seals (CFR: 41.7)

K5.14 Knowledge of the 004 Chemical and Volume X 2.5 29/2 operational implications of the Control following concepts as they apply to the CVCS: Reduction process of gas concentration in RCS: vent accumulated non-condensable gases from PZR bubble space, depressurized during cooldown or by alternately heating and cooling (spray) within allowed pressure band (drive more gas out of solution)

(CFR: 41.5/45.7) 2.4.4 Ability to recognize abnormal 005 Residual Heat Removal X 4.5 30/3 indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

(CFR: 41.10 / 43.2 / 45.6)

K1.10 Knowledge of the physical 005 Residual Heat Removal X 3.2 31/4 connections and/or cause effect relationships between the RHRS and the following systems:

Containment Spray System (CSS)

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

A4.11 Ability to manually operate 006 Emergency Core Cooling X 4.2 32/5 and/or monitor in the control room:

Overpressure protection system.

(CFR: 41.7 / 45.5 to 45.8)

A3.01 Ability to monitor automatic 007 Pressurizer Relief/Quench X 2.7 33/6 operation of the PRTS, including:

Tank Components which discharge to the PRT (CFR: 41.7 / 45.5)

K3.01 Knowledge of the effect that 008 Component Cooling Water X 3.4 34/7 a loss or malfunction of the CCWS will have on the following: Loads cooled by CCWS K2.02 Knowledge of bus power 008 Component Cooling Water X supplies to the following: CCW 3.0 35/8 pump, including emergency backup.

(CFR: 41.7)

K6.01 Knowledge of the effect of a 010 Pressurizer Pressure Control X 2.7 36/9 loss or malfunction of the following will have on the PZR PCS:

Pressure detection systems (CFR: 41.7 / 45.7)

Rev 1

K5.02 Knowledge of the 012 Reactor Protection X 3.1 37/10 operational implications of the following concepts as the apply to the RPS: Power density (CFR: 41.5 / 45.7)`

A1.01 Ability to predict and/or 012 Reactor Protection X 2.9 38/11 monitor Changes in parameters (to prevent exceeding design limits) associated with operating the RPS controls including: Trip setpoint adjustment (CFR: 41.5 / 45.5) 013 Engineered Safety Features X 2.4.9 Knowledge of low power / 3.8 39/12 Actuation shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13)

A2.05 Ability to (a) predict the 013 Engineered Safety Features X 3.7 40/13 impacts of the following Actuation malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Loss of dc control power (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A3.01 Ability to monitor automatic 022 Containment Cooling X 4.1 41/14 operation of the CCS, including:

Initiation of safeguards mode of operation (CFR: 41.7 / 45.5) 025 Ice Condenser A4.05 Ability to manually operate 026 Containment Spray X 3.5 42/15 and/or monitor in the control room:

Containment spray reset switches (CFR: 41.7 / 45.5 to 45.8)

K3.05 Knowledge of the effect that 039 Main and Reheat Steam X 3.6 43/16 a loss or malfunction of the MRSS will have on the following: RCS (CFR: 41.7 / 45.6)

K4.02 Knowledge of MFW design 059 Main Feedwater X 3.3 44/17 feature(s) and/or interlock(s) which New -lower provide for the following: Automatic turbine/reactor trip runback.

(CFR: 41.7)

A1.04 Ability to predict and/or 061 Auxiliary/Emergency X 3.9 45/18 monitor changes in parameters Feedwater (to prevent exceeding design limits) associated with operating the AFW controls including: AFW source tank level.

(CFR: 41.5 / 45.5)

K1.01 Knowledge of the physical 061 Auxiliary/Emergency X 4.1 46/19 connections and/or cause-effect Feedwater relationships between the AFW and the following systems: S/G system (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Rev 1

A2.03 Ability to (a) predict the 062 AC Electrical Distribution X 2.9 47/20 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:

Consequences of improper sequencing when transferring to or from an inverter.

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

K4.03 Knowledge of ac distribution 062 AC Electrical Distribution X 2.8 48/21 system design feature(s) and/or interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers (CFR: 41.7)

A2.01 Ability to (a) predict the 063 DC Electrical Distribution X 2.5 49/22 impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Grounds (CFR: 41.5 / 43.5 / 45.3 / 45.13)

K4.01 Knowledge of ED/G system 064 Emergency Diesel Generator X 3.8 50/23 design feature(s) and/or interlock(s) which provide for the following: Trips while loading the ED/G (frequency, voltage, speed)

(CFR: 41.7)

K6.07 Knowledge of the effect of a 064 Emergency Diesel Generator X 2.7 51/24 loss or malfunction of the following will have on the ED/G system: Air receivers (CFR: 41.7 / 45.7)

K1.01 Knowledge of the physical 073 Process Radiation Monitoring X 3.6 52/25 connections and/or cause-effect relationships between the PRM system and the following systems:

Those systems served by PRMs (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K2.08 Knowledge of bus power 076 Service Water X 3.1 53/26 supplies to the following: ESF-actuated MOVs 2.7 Replaced with KA K2.01 (CFR: 41.7)

A3.01 Ability to monitor automatic 078 Instrument Air X 3.1 54/27 operation of the IAS, including: Air pressure (CFR: 41.7 / 45.5)

A4.04 Ability to manually operate 103 Containment X 3.5 55/28 and/or monitor in the control room:

Phase A and phase B resets.

(CFR: 41.7 / 45.5 to 45.8)

K/A Category Point Totals: 3 2 2 4 2 2 2 3 3 3 2 Group Point Total: 28 Rev 1

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive K1.07 Knowledge of the physical 002 Reactor Coolant X 3.5 56/29 connections and/or cause-effect relationships between the RCS and the following systems: Reactor vessel level indication system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 011 Pressurizer Level Control K4.05 Knowledge of RPIS design 014 Rod Position Indication X 3.1 57/30 feature(s) and/or interlock(s) which provide for the following: Rod hold interlocks (CFR: 41.5 / 45.7)

A1.05 Ability to predict and/or 015 Nuclear Instrumentation X 3.7 58/31 monitor changes in parameters to prevent exceeding design limits) associated with operating the NIS controls including: Imbalance (axial shape)

(CFR: 41.5 . 45.5) 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge A2.02 Ability to (a) predict the 033 Spent Fuel Pool Cooling X 2.7 59/32 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: Loss of SFPCS (CFR: 41.5 / 43.5 / 45.3 / 45.13) 034 Fuel Handling Equipment K5.01 Knowledge of operational 035 Steam Generator X 3.4 60/33 implications of the following concepts as they apply to the S/GS: Effect of secondary parameters, pressure, and temperature on reactivity.

(CFR: 41.5 / 45.7)

A4.04 Ability to manually operate 041 Steam Dump/Turbine Bypass X 2.7 61/34 and/or monitor in the control room:

Control Pressure mode (CFR: 41.7 / 45.5 to 45.8)

Rev 1

K3.01 Knowledge of the effect that a 045 Main Turbine Generator X 2.9 62/35 loss or malfunction of the MT/G system will have on the following:

Remainder of the plant.

(CFR: 41.7 / 45.6) 055 Condenser Air Removal K1.03 Knowledge of the physical 056 Condensate X 2.6 63/36 connections and/or cause-effect relationships between the Condensate System and the following systems: MFW (CFR: 41.2 to 41.9 / 45.7 to 45.8) 068 Liquid Radwaste 071 Waste Gas Disposal 2.2.44 Ability to interpret control 072 Area Radiation Monitoring X 4.2 64/37 room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12) 075 Circulating Water 079 Station Air K4.06 Knowledge of design 086 Fire Protection X 3.0 65/38 feature(s) and/or interlock(s) which provide for the following: CO2 (CFR: 41.7)

K/A Category Point Totals: 2 0 1 2 1 0 1 1 0 1 1 Group Point Total: 10 Rev 1

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Date of Exam:

Category K/A # Topic RO SRO-Only IR # IR #

Knowledge of shift or short-term relief turnover practices.

2.1.3 3.7 66 (CFR: 41.10 / 45.13)

Knowledge of RO duties in the control room during fuel 2.1.44 3.9 67 handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in

1. support of fueling operations, and supporting Conduct of instrumentation.

Operations (CFR: 41.10 / 43.7 / 45.12) 2.1.34 Knowledge of primary and secondary plant 2.1.34 2.7 68 chemistry limits.

(CFR: 41.10 / 43.5 / 45.12)

Subtotal 3 Knowledge of the process used to track inoperable 2.2.43 3.0 69 alarms.

(CFR: 41.10 / 43.5 / 45.13)

Knowledge of tagging and clearance procedures.

2. 2.2.13 4.1 70 (CFR: 41.10 / 45.13)

Equipment Control Knowledge of the design, procedural, and operational 2.2.3 3.8 71 differences between units.

(CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12)

Subtotal 3 Knowledge of radiation monitoring systems, such as fixed 2.3.15 2.9 72 radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.12 / 43.4 / 45.9) 3.

Radiation Knowledge of radiation exposure limits under normal or Control 2.3.4 3.2 73 emergency conditions.

(CFR: 41.12 / 43.4 / 45.10)

Subtotal 2 Knowledge of fire protection procedures.

2.4.25 3.3 74 (CFR: 41.10 / 43.5 / 45.13) 4.

Emergency Knowledge of EOP layout, symbols, and icons.

Procedures / 2.4.19 3.4 75 (CFR: 41.10 / 45.13)

Plan Subtotal 2 Tier 3 Point Total 10 7 Rev 1

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A RO -T2G1 076 K2.08 In the Service Water (Aux Saltwater) system there are no "ESF actuated MOVs".

Replaced with the other available K2 KA for 076. K2.01 - Service Water (2.7)

RO-T1G1 APE 058 G2.2.22 Unable to write RO question to "apply LCO" for loss of DC.

Replaced with another G2 .2, 2.2.37 (3.6)

RO-T1G1 APE 062 AK3.01 There are not automatic valves in Diablo Canyon's Aux Saltwater System (equivalent to Nuclear Service Water).

Replaced with AK3.03 (4.0)

Rev 1

Form ES-401-2 Facility: DCPP Date of Exam: 2018-01 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 18 3 3 6 Emergency &

Abnormal 2 N/A N/A 9 2 2 4 Plant Evolutions Tier Totals 27 5 5 10 1 28 2 3 5 2.

Plant 2 10 3 0 3 Systems Tier Totals 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 Categories 2 2 1 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 Rev 1

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

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 2.1.7 Ability to evaluate plant performance 000009 Small Break LOCA / 3 X and make operational judgments based on 4.7 76 operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 2.4.41 Knowledge of the emergency action 000022 Loss of Rx Coolant Makeup / 2 X level thresholds and classifications. 4.6 77 (CFR: 41.10 / 43.5 / 45.11) (#76 - L091)

Replace with KA 2.4.4 Ability to recognize 4.7 abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 EA2.08 Ability to determine or interpret the 000029 ATWS / 1 X following as they apply to a ATWS: Rod bank 3.5 78 step counters and RPI.

(CFR 43.5 / 45.13) 000038 Steam Gen. Tube Rupture / 3 000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture - Excessive Heat Transfer / 4 2.1.23 Ability to perform specific system and 000054 (CE/E06) Loss of Main X integrated plant procedures during all modes 4.4 79 Feedwater / 4 of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6) 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 AA2.03 Ability to determine and interpret the 000058 Loss of DC Power / 6 X following as they apply to the Loss of DC 3.9 80 Power: DC loads lost; impact on ability to operate and monitor plant systems.

(CFR: 43.5 / 45.13)

Rev 1

000062 Loss of Nuclear Svc Water / 4 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 AA2.05 Ability to determine and interpret the 000077 Generator Voltage and Electric X following as they apply to Generator Voltage 3.8 81 Grid Disturbances / 6 and Electric Grid Disturbances: Operational New - higher status of offsite circuit.

(CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)

K/A Category Totals: 3 3 Group Point Total: 6 Rev 1

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2

• 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 AA2.04 Ability to determine and interpret X 3.2 82 the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Satisfactory overlap between source-range, intermediate-range and power-range instrumentation.

(CFR: 43.5 / 45.13) 000036 (BW/A08) Fuel Handling Accident / 8 AA2.03 Ability to determine and interpret X 4.2 83 the following as they apply to the Fuel Handling Incidents: Magnitude of potential radioactive release (CFR: 43.5 /

45.13).

000037 Steam Generator Tube Leak / 3 2.2.22 Knowledge of limiting conditions X 4.7 84 for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2) 000051 Loss of Condenser Vacuum / 4 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) 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 W/E16 High Containment Radiation / 9 2.4.2 Knowledge of system set points, X 4.6 85 interlocks and automatic actions associated with EOP entry conditions.

(CFR: 41.7 / 45.7 / 45.8)

BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 Rev 1

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: 2 2 Group Point Total: 4 Rev 1

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump A2.26 Ability to (a) predict the 004 Chemical and Volume X 3.0 86 impacts of the following Control 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: Low VCT pressure.

(CFR: 41.5/ 43/5 / 45/3 / 45/5) 005 Residual Heat Removal 006 Emergency Core Cooling A2.04 Ability to (a) predict the 007 Pressurizer Relief/Quench X 2.9 87 impacts of the following Tank malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Overpressurization of the waste gas vent header.

(CFR: 41.5 / 43.5 / 45.3 / 45.13) 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features Actuation 2.4.41 Knowledge of the 022 Containment Cooling X 4.6 88 emergency action level thresholds and classifications.

(CFR: 41.10 / 43.5 / 45.11) 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator Rev 1

2.1.32 Ability to explain and apply 073 Process Radiation Monitoring X 4.0 89 system limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 2.2.25 Knowledge of the bases in 076 Service Water X 4.2 90 Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) 078 Instrument Air 103 Containment K/A Category Point Totals: 2 3 Group Point Total: 5 Rev 1

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation A2.01 Ability to (a) predict the impacts 016 Non-Nuclear Instrumentation X 3.1 91 of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure.

(CFR: 41.5 / 43.5 / 45.3 / 45.5) 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment A2.05 Ability to (a) predict the impacts 035 Steam Generator X 3.4 92 of the following malfunctions or operations on the S/G; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Unbalanced flows to the S/Gs.

(CFR: 41.5 / 43.5 / 45.3 / 45.5) 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate Rev 1

A2.02 Ability to (a) predict the impacts 068 Liquid Radwaste X 2.8 93 of the following malfunctions or operations on the Liquid Radwaste System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Lack of tank recirculation prior to release.

A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences 3.3 of those malfunctions or operations:

Failure or automatic isolation (CFR: 41.5 / 43.5 / 45.3 / 45.13) 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 3 0 Group Point Total: 3 Rev 1

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Date of Exam:

Category K/A # Topic RO SRO-Only IR # IR #

Knowledge of industrial safety procedures (such as 2.1.26 3.6 94 rotating equipment, electrical, high temperature, high

1. pressure, caustic, chlorine, oxygen and hydrogen).

Conduct of (CFR: 41.10 / 45.12)

Operations Knowledge of the fuel-handling responsibilities of SROs.

2.1.35 3.9 95 (CFR: 41.10 / 43.7)

Subtotal 2 Knowledge of the process for making changes to 2.2.6 3.6 96 procedures.

(CFR: 41.10 / 43.3 / 45.13)

2. Knowledge of the process for managing maintenance 2.2.17 3.8 97 Equipment activities during power operations, such as risk Control assessments, work prioritization, and coordination with the transmission system operator.(CFR: 41.10 / 43.5 /

45.13)

Subtotal 2

3. Ability to approve release permits. (CFR: 41.13 / 43.4 /

2.3.6 3.8 98 Radiation 45.10)

Control Subtotal 1 Knowledge of the organization of the operating 2.4.5 4.3 99 procedures network for normal, abnormal, and

4. emergency evolutions.

Emergency (CFR: 41.10 / 43.5 / 45.13)

Procedures / Knowledge of the lines of authority during implementation 2.4.37 4.1 100 Plan of the emergency plan.

(CFR: 41.10 / 45.13)

Subtotal 2 Tier 3 Point Total 10 7 Rev 1

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A T1/G1 APE 022 G2.4.41 KA is covered by SRO Admin JPM.

Randomly selected G2.4.4 (IR 4.7)

T2/G2 068 A2.02 Not able to write relevant SRO level question. Selected A2.04 as replacement, IR 3.3 Rev 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Diablo Canyon Date of Examination: 01/19/2018 Examination Level: RO SRO Operating Test Number: L162 Administrative Topic (see Note) Type Describe activity to be performed Code*

Determine Affected Indicators Due To Malfunction of Eagle 21 Protection or Conduct of Operations N, R Control Channel (NRCL162-A1) 2.1.7 Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.

(4.4)

Calculate Rod Position Alignment and Rod Insertion Limits Conduct of Operations N, R 2.1.23 Ability to perform specific system and (NRCL162-A2) integrated plant procedures during all modes of plant operation.

(4.3)

Calculate Axial Flux Difference Equipment Control 2.2.42 Ability to recognize system parameters that M, R are entry-level conditions for Technical (NRCL162-A3) Specifications. (3.9) (modified from L061C)

Calculate Maximum Stay Time Radiation Control 2.3.4 Knowledge of radiation exposure limits M, R under normal or emergency conditions.

(NRCL162-A4) (3.2)

(modified from L111 NRC)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

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

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

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

Rev 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Diablo Canyon Date of Examination: 01/19/2018 Examination Level: RO SRO Operating Test Number: L162 Administrative Topic (see Note) Type Describe activity to be performed Code*

Review AP-5 Bistable Trip Authorization Form Conduct of Operations N, R 2.1.7 Ability to evaluate plant performance and (NRCL162-A5) make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.

(4.7)

Review Rod Position Alignment and Rod Insertion Limits Conduct of Operations N, R 2.1.23 Ability to perform specific system and (NRCL162-A6) integrated plant procedures during all modes of plant operation.

(4.3)

Verify AFD is within Tech Spec Limits Equipment Control 2.2.42 Ability to recognize system parameters that M, R are entry-level conditions for Technical (NRCL162-A7) Specifications. (4.6) (Modified from L061C)

Approve Liquid Waste Release Permit Radiation Control 2.3.6 Ability to approve release permits. (3.8)

M, R (Modified from L061C)

(NRCL162-A8)

Perform an Emergency Classification Emergency Plan 2.4.41 Emergency Procedures/Plan.

N, R (4.6)

(NRCL162-A9)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

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

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

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

Rev 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Diablo Canyon Date of Examination: 01/19/2018 Exam Level: RO SRO-I SRO-U Operating Test Number: L162 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function

a. (S1) (004.A2.14) Establish Emergency Boration (Bank LJC-063) A,D,S 1

b. (S2) (013.A4.01) Respond to CVI Actuation (Modified from A,EN,M,S 2 NRCL081LJC-S5)

c. (S3) (006.A1.13) Respond to High Accumulator Pressure (Bank LJC-009) D,S 3 (RO Only)

d. (S4) (E03.EA1.1) Start Reactor Coolant Pumps (Bank LJC-044) D,E,L,S 4P

e. (S5) (022.A4.01) Respond to CFCU High Vibration A,N,S 5

f. (S6) (064.A4.01) Transfer Vital 4kV Bus from D/G to Startup (Modified E,L,M,S 6 from Bank LJC-087)

g. (S7) (045.A4.01) Perform Load Trim to Match Tave to Tref N,S 4S

h. (S8) (060.AA1.02) Respond to Gaseous Rad Release A,N,S 9 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. (P1) (004.A2.06) Isolate Dilution Flow Paths (LJP-062) D,E,L 1

j. (P2) (064.A3.06) Perform a Local Start of a Diesel Generator (LJP-038) A,D,E,L 6

k. (P3) (067.AA1.08) Manually Operate the Cardox System (LJP-138A) A,D 8

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

• Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1/ 1/ 1 (control room system)

(L)ow-Power/Shutdown 1/ 1/ 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 (P)revious 2 exams 3/ 3/ 2 (randomly selected)

(R)CA 1/ 1/ 1 (S)imulator Rev 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Diablo Canyon Date of Examination: 01/19/2018 Exam Level: RO SRO-I SRO-U Operating Test Number: L162 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function

a. (S1) (004.A2.14) Establish Emergency Boration (Bank LJC-063) A,D,S 1

b. (S2) (013.A4.01) Respond to CVI Actuation (Modified from A,EN,M,S 2 NRCL081LJC-S5) c.

d. (S4) (E03.EA1.1) Start Reactor Coolant Pumps (Bank LJC-044) D,E,L,S 4P

e. (S5) (022.A4.01) Respond to CFCU High Vibration A,N,S 5

f. (S6) (064.A4.01) Transfer Vital 4kV Bus from D/G to Startup (Modified E,L,M,S 6 from Bank LJC-087)

g. (S7) (045.A4.01) Perform Load Trim to Match Tave to Tref N,S 4S

h. (S8) (060.AA1.02) Respond to Gaseous Rad Release A,N,S 9 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. (P1) (004.A2.06) Isolate Dilution Flow Paths (LJP-062) D,E,L 1

j. (P2) (064.A3.06) Perform a Local Start of a Diesel Generator (LJP-038) A,D,E,L 6

k. (P3) (067.AA1.08) Manually Operate the Cardox System (LJP-138A) A,D 8

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

• Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature 1/ 1/ 1 (control room system)

(L)ow-Power/Shutdown 1/ 1/ 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 (P)revious 2 exams 3/ 3/ 2 (randomly selected)

(R)CA 1/ 1/ 1 (S)imulator Rev 0

ES-301 Transient and Event Checklist Form ES-301-5 Group I (I1, I2, R1, R2)

Facility: DCPP Date of Exam: Jan 19, 2018 Operating Test Number: L162 A E Scenarios P V Day-1 (S4) Day-2 (S1) Day-3 (S2) Day-4 (S5)

P E T M L N CREW CREW CREW CREW O I I T POSITION POSITION POSITION POSITION T N C A I A T S A B S A B S A B S A B L M N Y R T O R T O R T O R T O U T P O C P O C P O C P O C P M(*)

E R I U RX 1 1 0 RO1 NOR 1 1 1 SRO-I I/C 1,3,7 2,3,5,7,8 1,2,4,6,7 13 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 0 0 2 2 RX 1 1 1 1 0 RO2 NOR 1 1 1 SRO-I I/C 1,3,4,7,8 3,4,5 1,2,6 11 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 0 0 2 2 RX 1 1 0 RO NOR 1 1 1 SRO-I1 I/C 1,3,4,7 1,2,4 3,4,6 10 4 4 2 SRO-U MAJ 5,6 5 5 4 2 2 1 TS 1,2 2,3 4 0 2 2 RX 1 1 1 1 0 RO NOR 1 1 1 SRO-I2 I/C 2,3,4,5 2,4,8 1,2,3,4,6 12 4 4 2 SRO-U MAJ 6,9 5 5 4 2 2 1 TS 2,3,5 1,2 5 0 2 2 Instructions:

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

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

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

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

ES-301, Page 26 of 27 Rev 1

ES-301 Transient and Event Checklist Form ES-301-5 Group II (I3, I4, R3, R4)

Facility: DCPP Date of Exam: Jan 19, 2018 Operating Test Number: L162 A E Scenarios P V Day-1 (S4) Day-2 (S1) Day-3 (S2) Day-4 (S5)

P E T M L N CREW CREW CREW CREW O I I T POSITION POSITION POSITION POSITION T N C A I A T S A B S A B S A B S A B L M N Y R T O R T O R T O R T O U T P O C P O C P O C P O C P M(*)

E R I U RX 1 1 0 RO3 NOR 1 1 1 SRO-I I/C 1,3,7 2,3,5,7,8 1,2,4,6,7 13 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 0 0 2 2 RX 1 1 1 1 0 RO4 NOR 1 1 1 SRO-I I/C 1,3,4,7,8 3,4,5 1,2,6 11 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 0 0 2 2 RX 1 1 0 RO NOR 1 1 1 SRO-I2 I/C 1,3,4,7 1,2,4 3,4,6 10 4 4 2 SRO-U MAJ 5,6 5 5 4 2 2 1 TS 1,2 2,3 4 0 2 2 RX 1 1 1 1 0 RO NOR 1 1 1 SRO-I3 I/C 2,3,4,5 2,4,8 1,2,3,4,6 12 4 4 2 SRO-U MAJ 6,9 5 5 4 2 2 1 TS 2,3,5 1,2 5 0 2 2 Instructions:

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

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

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

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

ES-301, Page 26 of 27 Rev 1

ES-301 Transient and Event Checklist Form ES-301-5 Group III (I5, R5, R6)

Facility: DCPP Date of Exam: Jan 19, 2018 Operating Test Number: L162 A E Scenarios P V Day-1 (S4) Day-2 (S1) Day-3 (S2) Day-4 (S5)

P E T M L N CREW CREW CREW CREW O I I T POSITION POSITION POSITION POSITION T N C A I A T S A B S A B S A B S A B L M N Y R T O R T O U T O R T O U T P O C P O C P R C P O C P M(*)

E R I U RX 1 1 0 RO5 NOR 1 1 1 SRO-I I/C 1,3,7 2,3,5,7,8 1,2,4,6,7 13 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 0 2 2 RX 1 1 1 1 0 RO6 NOR 1 1 1 SRO-I I/C 1,3,4,7,8 3,4,5 8 4 4 2 SRO-U MAJ 5,6 6,9 4 2 2 1 TS 0 2 2 RX 1 1 1 1 0 RO NOR 1 1 1 SRO-I5 I/C 1,3,4,7 2,3,4,5 2,4,8 11 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 1,2 2,3,5 5 0 2 2 RX 1 1 0 RO NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 Instructions:

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

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

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

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

ES-301, Page 26 of 27 Rev 1

ES-301 Transient and Event Checklist Form ES-301-5 Group IV (I6, R7, R8)

Facility: DCPP Date of Exam: Jan 19, 2018 Operating Test Number: L162 A E Scenarios P V Day-1 (S4) Day-2 (S1) Day-3 (S2) Day-4 (S5)

P E T M L N CREW CREW CREW CREW O I I T POSITION POSITION POSITION POSITION T N C A I A T S A B S A B S A B S A B L M N Y R T O R T O U T O R T O U T P O C P O C P R C P O C P M(*)

E R I U RX 1 1 0 RO5 NOR 1 1 1 SRO-I I/C 1,3,7 2,3,5,7,8 1,2,4,6,7 13 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 0 2 2 RX 1 1 1 1 0 RO6 NOR 1 1 1 SRO-I I/C 1,3,4,7,8 3,4,5 8 4 4 2 SRO-U MAJ 5,6 6,9 4 2 2 1 TS 0 2 2 RX 1 1 1 1 0 RO NOR 1 1 1 SRO-I5 I/C 1,3,4,7 2,3,4,5 2,4,8 11 4 4 2 SRO-U MAJ 5,6 6,9 5 5 2 2 1 TS 1,2 2,3,5 5 0 2 2 RX 1 1 0 RO NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 Instructions:

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

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

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

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

ES-301, Page 26 of 27 Rev 1

ES-301 Transient and Event Checklist Form ES-301-5 Spare Facility: DCPP Date of Exam: Jan 19, 2018 Operating Test Number: L162 A E Scenarios P V Spare Day-2 (S1) Day-3 (S2) Day-4 (S5)

P E T M L N CREW CREW CREW CREW O I I T POSITION POSITION POSITION POSITION T N C A I A T S A B S A B S A B S A B L M N Y R T O R T O R T O R T O U T P O C P O C P O C P O C P M(*)

E R I U RX 3 3 3 1 1 0 RO NOR 1 1 1 SRO-I I/C 1,2,4 4,6 1,2,6 4 4 2 SRO-U MAJ 5 5 5 2 2 1 TS 1,4 0 2 2 RX 1 1 0 RO NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 RX 1 1 0 RO NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 RX 1 1 0 RO NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 Instructions:

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

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

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

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

ES-301, Page 26 of 27 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 1 Op-Test No: L162 NRC Examiners: Operators:

Initial Conditions: 3% power with CCP 1-2 In Service (75 gpm letdown); MFP 1-1 supplying S/Gs; on Startup Power; MOL, 1234 ppm boron Turnover: In OP L-3, performing step 6.28, raising power to 8%.

Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 N/A R (ATC, Raise reactor power from 3% to 8% OP L-3, sec 6.28 SRO) 2 XMT_RMS23_3 1E+006 TS, I S/G Blowdown RM-23 fails high. FCV-498/ FCV-499 and half (BOP, of sample valves fail to isolate, but can be manually closed SRO) (ECG 39.3.B)(PK11-17) 3 PMP_CVC2_2 OVERLOAD_DEV_FAIL TS, C Centrifugal Charging Pump 1-2 OC Trip requiring restoration (ALL) of letdown (TS 3.5.2.A) (AP-17) 4 XMT_MSS1_3 15 ramp=300 I (ATC, PT-507, Steam Generator Header Pressure Transmitter, slow SRO) failure low causing Group I dumps to close. (AP-5)

PMP_AFW1_2 OVERLOAD_DEV_FAIL 5 TS, C MFP 1-1 trips. MDAFW pumps start but trip; requires start of PMP_AFW2_2 OVERLOAD_DEV_FAIL (ALL) TDAFW pump. (TS 3.7.5.D)(PK09-12, AP-15)

BST_MFW1_1 1 6 MAL_MSS6A 90 ramp=240 M S/G 1-1 Safety Lifts; reseats 10 seconds after SI (ALL)

MAL_MSS6A 0 delay=10 cd='jpplsia' 7 CVC9CVC_CCP11_MTRSHEAR C CCP 1-1 shaft shear 30 seconds after SI delay=30 cd='jpplsia' (BOP) 8 VLV_PZR4_2 0.3 cd='jpplsia' del C Pressurizer PORV PCV-455C fails slightly open on trip ay=60 (BOP) requiring manual isolation by associated block valve 9 MAL_RCS3B 3.5 cd='V1_240S_1 or M SBLOCA after SI is terminated in E-1.1 V1_241S_1' delay=0 ramp=15 (ALL)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L162 NRC ES-D-1-01 r1.docx Page 1 of 3 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4) Actual Attributes

1. Total malfunctions (5-8) (Events 2,3,4,5,6,7,8,9) 8

2. Malfunctions after EOP entry (1-2) (Event 7,8) 2

3. Abnormal events (1-4) (Events 2,3,4,5) 4

4. Major transients (1-2) (Events 6,9) 2

5. EOPs entered/requiring substantive actions (1-2) (E-1.1) 1

6. EOP contingencies requiring substantive actions (0-2) 0

7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S1CT-1) Close the block MOV upstream of The open PORV and block valve constitute the

• Westinghouse Owners the stuck open PORV prior to performance degradation of a fission product barrier. Closing the Group WCAP-17711-NP of step 8 of EOP E-0, Reactor Trip or Safety block valve is essential to safety since failure to do Injection. so results in the unnecessary continuation of the degraded condition.

(S1CT-2) Reinitate SI before a severe Degraded core cooling is caused by a substantial

• Background Information challenge to the Core Cooling Critical Safety loss of primary coolant. Reinitiation of high for WOG Emergency Function develops (magenta path on F-0.2 pressure safety injection is the most effective Response Guideline Core Cooling). method to restore RCS inventory and core cooling. HFRC2BG Rev 3.

The effectiveness of safety injection in restoring core cooling is determined by the trend in core exit TC temperatures or RVLIS full range when the RCPs are tripped.

Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

L162 NRC ES-D-1-01 r1.docx Page 2 of 3 Rev 1

SCENARIO

SUMMARY

- NRC #1

1. Control rods are used to raise power from 2% to 8% OP L-3, Secondary Plant Startup, step 6.28. ATC operator complies with 1 step pull and wait procedural requirement while monitoring relevant controls and diverse indicators. Shift Foreman provides reactivity oversight.

2. S/G Blowdown RM-23 fails high resulting in only a partial blowdown isolation. The crew responds, manually isolating the unactuated sample isolation valves and realigning blowdown discharge to the Equipment Drain Receiver, following the guidance of AR PK11-17, SG BLOW DOWN HI RAD. Shift Foreman enters ECG 39.3.B, Radioactive Liquid Effluent Monitoring Instrumentation, for Steam Generator Blowdown Tank (RM-23) inoperable.

3. Charging Pump CCP 1-2 trips on over current. The crew responds by entering OP AP-17, Loss of Charging to restore normal charging and letdown. Shift Foreman enters Tech Spec 3.5.2.A, ECCS - Operating, for one ECCS train inoperable.

4. PT-507, Steam Generator Header Pressure Transmitter, slowly fails low causing Group I dumps to close.

Crew diagnoses the failure and takes manual control of HC-507. OP AP-5, Malfunction of Eagle 21 Protection or Control Channel is used to address the failure and return primary and secondary to normal bands.

5. MFP 1-1 trips on high vibration. Both MDAFW pumps start initially, but trip on overcurrent. The crew enters AR PK09-12, Main Feedwater Pump Trip, and follow the guidance of OP AP-15, Loss of Feedwater Flow, Section B: Single Operating MFP Trips, starting the TDAFW pump, tripping the turbine, and inserting rods in manual to reduce power to 2%. Shift Foreman enters Tech Spec 3.7.5.D, AFW System, for two AFW trains inoperable.

6. S/G 1-1 Safety lifts causing uncontrollable depressurization of S/G 1-1. Shift Foreman directs board operators to trip the reactor and initiate Safety Injection once reactor trip has been verified. The crew enters EOP E-0, Reactor Trip or Safety Injection. The safety valve reseats 45 seconds after Safety Injection initiates. The crew throttles AFW to control the cooldown as they work their way towards SI termination.

7. CCP 1-1 fails due to a sheared shaft 30 seconds after SI actuation. The board operator identifies the condition based on low motor amps and flow.

8. Board operators also identify PCV-455C in mid-position. The valve will not close and must be isolated using the associated block valve 8000B (S1CT-1) Close the block MOV upstream of the stuck open PORV before performing step 8 of EOP E-0).

9. Once termination criteria has been met, the crew transitions to EOP E-1.1, SI Termination. A SBLOCA occurs immediately following the shutdown of Safety Injection Pump 1-2. The crew performs the final critical task of reinitiating Safety Injection (S1CT-2) Reinitate SI before a severe challenge to the Core Cooling Critical Safety Function develops (magenta path on F-0.2 Core Cooling).

The scenario is terminated once ECCS pumps have been restarted.

L162 NRC ES-D-1-01 r1.docx Page 3 of 3 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 2 Op-Test No: L162 NRC Examiners: Operators:

Initial Conditions: 50% with CCP 1-3 and MFP 1-1 OOS; CCP 1-2 IS; MOL, 1000 ppm boron Turnover: MFP 1-2 has elevated vibrations. ODM held earlier established action plan with ramping guidelines should conditions degrade.

Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 GGACRL_94BTVSP 1 C, BOP Gen Voltage Regulator fails requiring manual voltage control on the base adjuster (PK14-22).

2 VLV_CVC16_2 .11 delay=15 ramp=3 TS, C CVCS-8152 fails to 90% closed requiring Excess Letdown to (ALL) be placed in service (PK04-21, AP-18) (TS 3.6.3.A).

3 MAL_SEI1 0.12 delay=0 ramp=15 C, TS Large seismic event causes rupture of RWST. (PK06-20, TS ASISRWST 1.6e6 delay=10 ramp=1800 only 3.5.4.B).

(SRO) 4 MAL_MFW2B 2.45 delay=0 ramp=60 C (ALL) Vibrations on MFP 1-2 rise to ODM limit, requiring predesignated Unit 1 shutdown at 6 MW/min. (AP-25).

5 MAL_SEI1 0.2 cd='bsisrwst lt 54.6' M DBA LOCA on aftershock.

delay=0 ramp=10 (ALL)

MAL_RCS1C 100%_DBA cd='bsisrwst lt 54.5' 6 MAL_PPL5A BOTH C ATWS (13D/E Work)

MAL_PPL5B BOTH (BOP) 7 MAL_PPL1A FAILURE_TO_INIT C Phase A Train A fails to actuate.

(BOP) 8 MAL_CNM3 100 cd='rsih8980 lt 0.02' C (ATC) Sump blockage.

delay=60 ramp=15

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L162 NRC ES-D-1-02 r1.docx Page 1 of 4 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4) Actual Attributes

1. Total malfunctions (5-8) (Events 1,2,4,5,6,7,8) 7

2. Malfunctions after EOP entry (1-2) (Events 6,7,8) 3

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

4. Major transients (1-2) (Event 5) 1

5. EOPs entered/requiring substantive actions (1-2) (E-1, E-1.3, ECA-1.3) 3

6. EOP contingencies requiring substantive actions (0-2) (ECA-1.3) 1

7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S2CT-1) Initiate reactor trip prior The safeguards systems that protect the plant during

• Westinghouse Owners to performance of E-0, step 2. accidents are designed assuming that only decay heat and Group WCAP-17711-NP pump heat are being added to the RCS. Failure to manually

• Calc G.2 Rev 5 (08151-trip the reactor causes a extreme challenge to the 2169) subcriticality critical safety function (red path on F-0.1

• OP1.ID2, Time Critical subcriticality) beyond that irreparably introduced by the Operator Actions Rev 8A, postulated conditions. #34.

(S2CT-2) Stop all running ECCS Failure to stop the cavitating pumps leads to damage

• Background Information pumps with suction aligned to the sufficient to render the pumps unavailable for use once an for Westinghouse Owners containment recirc sump by the alternate make-up supply is aligned to the RCS. Group Sump Blockage completion of ECA-1.3, step 5: Guideline, Rev 0.

• CCP 1-1

• CCP 1-2

• SIP 1-1

• SIP 1-2

• RHRP 1-1

• RHRP 1-2 Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

L162 NRC ES-D-1-02 r1.docx Page 2 of 4 Rev 1

SCENARIO

SUMMARY

- NRC #2

1. Generator Voltage Regulator trips due to a loss of sensing voltage. The crew responds by entering AR PK14-22, GENERATOR VLTG REG TRIP and determine the voltage regulator is now operating in manual mode. Annunciator guidance is followed to maintain a lagging power factor.

2. Letdown Hx Inlet Valve, CVCS-8152 fails 90% closed causing letdown to divert to the Pressurizer Relief Tank. AR PK04-21, LETDOWN PRESS / FLOW TEMP comes into alarm, directing the crew to isolate Normal Letdown and place Excess Letdown in service per OP B-1A:IV, CVCS - Excess Letdown - Place In Service and Remove From Service. Alternately, the crew may elect to enter OP AP-18, Letdown Line Failure, which provides equivalent guidance. Shift Foreman enters TS 3.6.3.A - Containment Isolation Valves, for one containment isolation valve inoperable.

3. A 0.12 g seismic event results in a rupture of RWST, causing level to lower rapidly. The crew identifies RWST level lowering by monitoring level indications on VB-2 or by evaluating AR PK06-20, PPC Select which identifies RWST level is below the alarm setpoint. Field operators report a crack in the RWST extending down to approximately the 50% level. The Shift Foreman enters TS 3.5.4.B - Refueling Water Storage Tank (RWST) for borated water volume less than the required minimum of 455,300 gallons

(~94%).

4. Vibrations on MFP 1-2 rise to 2.5 mil which corresponds to a ODM limit, requiring predesignated Unit 1 shutdown at 6 MW/min. A ramp is commenced following the guidance of OP AP-25, Rapid Load Reduction or Shutdown.

5. A 0.20 g seismic aftershock results in 100% DBA LOCA.

6. The crew enters E-0, Reactor Trip or Safety Injection, performing their immediate actions. A reactor trip fails to automatically actuate (ATWS); manual Rx Trip control switches are ineffective as well. Control board operators perform their respective response actions: ATC drives control rods inward and BOP manually opens control rod breakers 13D/E on VB5 (S2CT-1) Initiate reactor trip prior to performance of E-0, step 2).

7. With the reactor tripped, the crew continues on, checking for actuation of emergency safeguards equipment and diagnosing conditions consistent with a large break LOCA (high containment pressure, loss of pressurizer pressure and level, loss of subcooling, high containment sump levels). The crew identifies RCP trip criteria are met, and with Shift Foreman concurrence, trip all four RCPs. Shift Foreman directs the BOP Operator to complete Appendix E, ESF AUTO ACTIONS, SECONDARY AND AUXILIARIES STATUS, and continues on in E-0. Train A of Phase A, Containment Isolation, fails to actuate, requiring board operators to manually align the associated inside containment isolation valves.

(continued on next page)

L162 NRC ES-D-1-02 r1.docx Page 3 of 4 Rev 1

SCENARIO

SUMMARY

- NRC #2

8. The Shift Foreman continues through E-0 diagnostic steps, and transitions to E-1, Loss of Reactor or Secondary Coolant. Functional restoration status trees are checked and crew identifies transition criteria for FR-P.1, Response to Imminent Pressurized Thermal Shock. Conditions will be met for exiting the procedure at the first step.

When RWST level reaches 33%, the crew transitions immediately to E-1.3, Transfer to Cold Leg Recirculation, and performs the required alignment steps. When RWST suction valve SI-8980 is isolated, ECCS recirculation flow is lost due to sump blockage. The crew transitions to ECA-1.3, Sump Blockage either directly, or by way of ECA-1.1, Loss of Emergency Coolant Recirculation, where they secure all running ECCS pumps (S2CT-2, Stop all running ECCS pumps with suction aligned to the containment recirc sump by the completion of ECA-1.3, step 5).

The scenario is terminated once Critical Task S2CT-2 is complete.

L162 NRC ES-D-1-02 r1.docx Page 4 of 4 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 3 Op-Test No: L162 NRC Examiners: Operators:

Initial Conditions: 75% with CFCU 1-5 OOS; MOL, 919 ppm boron Turnover: At 75% power due to grid instability.

Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 EECKSELECT2382371XPWR 0 C, TS Load Tap Changer Auto Control Failure (PK20-04) TS (BOP, 3.8.1.A SRO) 2 XMT_CVC16_3 150 delay=0 ramp=15 I (BOP, TE-130 fails high (PK04-21, AP-5)

SRO) 3 N/A R (ALL) Backdown Order; Shed 150 mw over next 15 minutes (AP-25).

4 DSC_ROD1 cd='smss lt 800' C, TS DRPI loss of normal power requires ramp to be placed (ATC, on hold, rods taken to manual. (AR PK03-21) (TS SRO) 3.1.7.B) 5 MAL_MSS4 720000 delay=0 ramp=60 M (ALL) MSLB outside containment 6 VLV_MSS7_2, VLV_MSS8_2, C (ALL) All MSIVs fail open VLV_MSS9_2, VLV_MSS10_2 1

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L162 NRC ES-D-1-03 r1.docx Page 1 of 3 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4) Actual Attributes

1. Total malfunctions (5-8) (Events 1,2,3,4,5,6) 6

2. Malfunctions after EOP entry (1-2) (Event 6) 1

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

4. Major transients (1-2) (Event 5) 1

5. EOPs entered/requiring substantive actions (1-2) (E-2, ECA-2.1) 2

6. EOP contingencies requiring substantive actions (0-2) (ECA-2.1) 1

7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S3CT-1) Stop uncontrolled cooldown by An event or series of events which leads to a

• Background Information controlling AFW flow before a severe relatively rapid and severe reactor vessel for WOG Emergency challenge to Integrity Safety Function downcomer cooldown can result in a thermal Response Guideline develops (magenta path on F-0.4 RCS shock to the vessel wall that may lead to a small Integrity) flaw, which may already exist in the vessel wall, growing into a larger crack. The growth or extension of such a flaw may lead, in some cases (where propagation is not stopped within the wall),

to a loss of vessel integrity (S3CT-2) Terminate SI prior to rupture of Failure to terminate ECCS flow when SI termination

• Westinghouse Owners PRT by closing 8801A/B and/or 8803A/B. criteria are met results in overfill of the Pressurizer Group WCAP-17711-NP (Note: CT is met by closing either 8801A/B OR and the eventual rupture of the PRT. This 8803A/B.) constitutes the avoidable degradation of the RCS as a fission product barrier.

Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

L162 NRC ES-D-1-03 r1.docx Page 2 of 3 Rev 1

SCENARIO

SUMMARY

- NRC #3

1. Startup Transformer 1-1 Load Tap Changer control power supply fails. Crew responds per AR PK20-04, SU TRANSF 11, 12, OR 21 LOCAL ANNUN, and manually controls transformer voltage. Shift Foreman enters TS 3.8.1.A, AC Sources - Operating for one required offsite circuit inoperable.

2. Letdown heat exchanger temperature element TE-130 fails high, causing actual letdown temperature to lower. AR PK04-21, LETDOWN PRESS / FLO TEMP comes into alarm, directing the crew to take manual control of letdown temperature (TCV-130), to restore temperature to normal range. Alternately, the crew may elect to follow the guidance of OP AP-5, Malfunction of Eagle 21 Protection or Control Channel .

3. Shift Manager reports a confirmed Grid Control Center backdown order due to grid instability. Unit 1 is directed to shed 150 MW within 15 minutes. The Shift Foreman determines an appropriate ramp rate to meet the backdown order requirement (may assign this task to reactor operator) and implements OP AP-25, Rapid Load Reduction or Shutdown. The ATC determines an initial boration based on the Reactivity Handbook and advises the Shift Foreman of his recommendation. The BOP enters the programmed ramp into the turbine control system. The reactivity evolutions are implemented sequentially, with the Shift Foreman providing oversight.

4. DRPI power failure due to normal supply breaker tripping open near the end of the ramp. Ramp is placed on hold, rods are taken to manual, and Tave is matched within 1.5 oF (if required) per AR PK03-21, DRPI FAILURE / ROD BOTTOM. The Shift Foreman enters TS 3.1.7.B - Rod Position Indication for more than one DRPI per group inoperable.

5. A main steamline break develops downstream of the Main Steam Isolation Valves, outside containment.

The crew identifies the need to isolate the Main Steam Isolation Valves and perform a safety injection (SI) based on pressurizer pressure and level lowering rapidly. Shift Foreman directs a reactor trip and SI and enters EOP E-0, Reactor Trip or Safety Injection.

6. All four main steam isolation valves fail open. The crew transitions to EOP E-2, Faulted Steam Generator Isolation, and then to EOP ECA-2.1, Uncontrolled Depressurization of All Steam Generators. The crew performs the critical tasks of stopping the uncontrolled cooldown (S3CT-1) Stop uncontrolled cooldown before a severe challenge (magenta path ) develops on F-0.4 RCS Integrity by minimizing feedflow and then terminating safety injection (S3CT-2) Terminate SI prior to rupture of PRT.

The scenario is terminated once SI is terminated .

L162 NRC ES-D-1-03 r1.docx Page 3 of 3 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 4 Op-Test No: L162 NRC Examiners: Operators:

Initial Conditions: 100% with MAFW Pump 1-3 OOS; MOL, 878 ppm boron Turnover:

Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 MAL_NIS6A 200 delay=0 ramp=420 I, TS NI-41 slow failure HIGH (AP-5; Multiple TS (see (ALL) summary section))

2 XMT_RCS6_3 -376.0 ramp=60 I, TS only PT-403 fails low (PK05-07, 09)(TS 3.3.3.A)

(SRO) 3 MAL_CWS2C 2.3 delay=0 ramp=2 C (ALL) Condenser In-leakage (PK12-05, AP-20 & 25) 4 XMT_CND29_3 282 ramp=240 C (BOP, CBP Set 1-3 high bearing temp when ramp reaches XMT_CND30_3 278 ramp=240 SRO) 1000 MW (PK10-06)

CD04CND_CDP13_MTFSEIZUR 1 cd='(h_v3_225r_1 and (txmtcbmo(3) gt 280))' delay=15 5 MAL_SEI1 0.15 delay=0 ramp=10 M (ALL) FCV-510 fails closed following seismic event.

CNV_MFW3_2 0 delay=0 ramp=60 6 MAL_RCS4F 600 cd='fnispr_2 lt 5' M (ALL) 600 gpm SGTR (S/G 1-2) delay=0 ramp=10 7 MAL_PPL3A BOTH, MAL_PPL3B BOTH C (ALL) SI Actuation Fails (both auto and manual) 8 CNV_RCS1_2 C (BOP) Prz Sprays failed closed / PORV used for CNV_RCS2_2 depressurization fails opened; block valve can not be closed

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L162 NRC ES-D-1-04 r1.docx Page 1 of 4 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4) Actual Attributes

1. Total malfunctions (5-8) (Events 1,3,4,5,6,7,8) 7

2. Malfunctions after EOP entry (1-2) (Events 7,8) 2

3. Abnormal events (1-4) (Events 1,3,4) 3

4. Major transients (1-2) (Event 5,6) 2

5. EOPs entered/requiring substantive actions (1-2) (E-3, ECA-3.1) 2

6. EOP contingencies requiring substantive actions (0-2) (ECA-3.1) 1

7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S4CT-1) Manually trip the reactor before Steam Generator Level below 15% narrow range in

• WOG Backgd HFHR1BG_R3 S/G 1-1 reaches dry out conditions as 1 of 4 loops after a power level dependent time indicated by WR level less than 10%. delay, normally generates a reactor trip signal to protect against a loss of heat sink. For this scenario, power remains above 50%, so the time delay = 0.

Once the S/G has reached dry out conditions, it is no longer capable of RCS heat removal.

Furthermore, the S/G is susceptible to structural damage as the result of thermal shock once feedwater is re-established from the Auxiliary Feedwater System.

(S4CT-2) Manually align at least one train of FSAR analysis predicates acceptable results on the

• WCAP-17711-NP, CT-2 SIS actuated safeguards before transition assumption that, at the very least, one train of

• WOG Backgrnd HE0BG_R2 out of EOP E-0, Reactor Trip or Safety safeguards has actuated and is providing flow to Injection. the core. Failure to start and manually align the minimum required safeguards equipment results in the persistence of degraded emergency core cooling system capacity.

(S4CT-3) Isolate the ruptured steam SG inventory increase leads to water release

• W Margin to Overfill (CN-generator from the intact steam generators through the S/G PORV or safety valve(s) or to SG CRA-05-53 Rev1) prior to commencing cooldown of the RCS overfill, which would seriously compromise the SG

• W Offsite Doses (CN-CRA-in step 9.c (40% steam dumps) or 10.b (10% as a fission-product barrier and complicate 05-54) steam dump) by completing the following: mitigation.

• SGTR UFSAR 15.4.3 Isolate feedwater by ensuring closed:

• WCAP-17711-NP

• LCV-107 (MDAFW Level Control Valve)

• LCV-111 (TDAFW Level Control Valve)

Isolate steamflow by ensuring closed:

• FCV-42 (S/G 1-2 MSIV)

• FCV-37 (S/G 1-2 supply to TD AFW Pp)

Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

L162 NRC ES-D-1-04 r1.docx Page 2 of 4 Rev 1

SCENARIO

SUMMARY

- NRC #4

1. Power Range Nuclear Instrument NI-41 slowly fails high causing inward rod motion. Crew diagnoses failure, and once motion is deemed unwarranted, takes rods to manual. Failure is addressed per OP AP-5, Malfunction of Eagle 21 Protection or Control Channel, which removes the failed channel from service and directs the Shift Foreman to address Tech Specs 3.3.1.D,E,S,T Reactor Trip System Instrumentation; ECG 37.2 Axial Flux Difference (AFD) monitoring, and ECG 37.3 (Quadrant Power Tilt Ratio Alarms).

2. PT-403, RCS Wide Range Pressure Transmitter, fails low. The crew responds to PK05-07, Subcooling Margin Lo/Lo-Lo and PK05-09, RVLIS Lo Lvl RVLIS/SCMM Trouble, identifying the affected instrumentation. Shift Foreman addresses TS 3.3.3.A, Post Accident Monitoring Instrumentation

3. A saltwater leak develops in the SW quadrant of the condenser, bringing in AR PK12-05, COND PPS DISCH HDR CATION CONDT'Y HI. The crew determines cation conductivity is elevated and the Shift Foreman enters OP AP-20, Condenser Tube Leak, which calls for a 25 MW/min ramp to 50%. The crew immediately implements OP AP-25, Rapid Load Reduction or Shutdown to commence the ramp.

4. Annunciator AR PK10-06, CNDS & CNDS BSTR PPS comes into alarm due to rising bearing temperatures on Condensate Booster Pump Set (CBP) 1-3. Reactor operators identify rapidly rising bearing temperatures using plant process computer trends. The crew manually starts CBP 1-2 and secures CBP 1-3 to prevent motor damage.

(Note: Malfunction is designed to trip CBP 1-3 if crew has not shut the pump down within 15 seconds of bearing temperature reaching 280oF. The Autostart of CBP 1-2 has been disabled and will require a manual start).

5. A 0.15 seismic event results in Main Feed Reg valve FCV-510 failing closed. S/G 1-1 level can not be maintained. S/G 1-1 Low Level trip has been disabled and the crew must manually trip the reactor (S4CT-1) Manually trip the reactor before S/G 1-1 reaches dry out conditions.

6. A 600 gpm tube rupture develops on S/G 1-2 when the reactor trips. The crew enters EOP E-0, Reactor Trip or Safety Injection, and identifies the rupture based on various radiation alarms, rising counts on RM-72, and the inability to maintain RCS pressure and pressurizer level following the trip.

7. Both auto and manual Safety Injection (SI) actuation signals fail and the crew must manually start and align SI actuated equipment (S4CT-2) Manually align at least one train of SIS actuated safeguards before transition out of EOP E-0.

(continued on next page)

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SCENARIO

SUMMARY

- NRC #4

8. The crew transitions to EOP E-3, Steam Generator Tube Rupture, and where they perform the critical task of isolating S/G 1-2 (S4CT-3) Isolate the ruptured steam generator from the intact steam generators prior to commencing cooldown of the RCS.*** Depressurization of the RCS is commenced following the cooldown. Pressurizer spray valves fail to operate and a PORV must be used. When the crew attempts to stop the depressurization, both the PORV and associated block valve fail to operate in the closed direction, and the Shift Foreman transitions to EOP ECA-3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired.

The scenario is terminated once the cooldown to Cold Shutdown in ECA-3.1 has been commenced or verified.

• • • CT / TCOA note: SGTR was evaluated against Time Critical Operator Actions (TCOAs) # 2 (SGTR); initial power level and supporting equipment conditions differ significantly from the conditions used in this scenario. For these reasons, the S/G TCOAs will remain critical (a critical task, per WOG), but TCOA time limits will not be applied to this scenario.

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Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 5 Op-Test No: L162 NRC Examiners: Operators:

Initial Conditions: 100% with AFWP 1-2 OOS; MOL, 878 ppm boron Turnover:

Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 VLV_PZR6_2 0.1 delay=0 ramp=5 C, TS PCV-474 slowly drifts open (AP-13)(TS 3.4.11.B).

(BOP, SRO) 2 PK1823_0132 1 C, TS Ground on ASW Pump 1-1 (PK18-23)(TS 3.7.8.A).

(BOP, SRO) 3 GGAHRL_62GSC3TVSP 0 C (ATC, Partial Stator Water cooling flow/partial runback (PK14-MAL_GEN3 LO_FLOW delay=10 SRO) 19, PK12-12, AP-25).

delIA MAL_GEN3 2 cd='smss lt 925' 4 MAL_CVC8A C (ATC, Seal Injection Filter 1-1 plugs causing reduction in SRO) charging flow to RCP seals (PK04-22).

5 RLY_PPL37 CLOSED(TRUE) M (ALL) Spurious Phase B causes isolation of CCW Header C requiring Reactor Trip and tripping of all four RCPs (PK01-08, AP-11).

6 MAL_AFW1 1 cd='H_V3_109M_1 GT 0.1' C (ALL) Bus F trips on differential on reactor trip causing loss of MAL_MFW2A,B 25 cd='fnispr lt 5.0' DRPI and AFW pump 1-3. Both MFPs Trip and TDAFP MAL_EPS4C_2 DIFFERENTIAL trips on overspeed; (post trip).

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L162 NRC ES-D-1-05 r1.docx Page 1 of 3 Rev 1

Appendix D (rev 11) Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4) Actual Attributes

1. Total malfunctions (5-8) (Events 1,2,3,4,5,6) 6

2. Malfunctions after EOP entry (1-2) (Events 6) 1

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

4. Major transients (1-2) (Event 5) 1

5. EOPs entered/requiring substantive actions (1-2) (FR-H.1) 1

6. EOP contingencies requiring substantive actions (0-2) (FR-H.1) 1

7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S5CT-1) Trip all four Reactor Coolant RCPs are susceptible to catastrophic failure and a

• FSAR Accident Analysis, Pumps (RCPs) as indicated by: loss of reactor coolant flow if left running in the Section 15.2.5 - Partial

• RCP Breaker position = OPEN absence of adequate bearing cooling flow. If the Loss of Forced Ractor

• RCP Amperage lowering reactor is at power at the time of the accident, the Coolant Flow

• RCP thrust bearing temperatures immediate effect of loss of coolant flow is a rapid lowering increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if prior to a partial loss of reactor coolant the reactor is not tripped promptly.

flow due to Reactor Coolant Pump failure.

(S5CT-2) Establish a secondary heat sink as A loss of all feedwater transient is characterized by

• FR-H.1 Background indicated by: a depletion of secondary inventory and eventual Document (HFRH1BG),

• WR level rising degradation of secondary heat transfer capability. Rev. 3.

• Core Exit Thermocouple As secondary heat transfer capability degrades, temperatures lowering core decay heat generation will increase RCS temperature and pressure causing loss of RCS Prior to reaching bleed and feed criteria inventory similar in nature to a small break loss of which is defined as wide range S/G level in coolant accident. Failure to restore a secondary any three S/Gs less than 18% [26%] AND heat sink when it is possible to do so constitutes a narrow range S/G level in all four S/Gs less significant reduction of safety margin beyond that than 15% [25%] narrow range.

irreparably introduced by the scenario.

Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

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SCENARIO

SUMMARY

- NRC #5

1. Pressurizer Pressure Control Valve PCV-474 drifts open and must be isolated using the associated 8000-A block valve. Shift Forman enters TS 3.4.11.B Pressurizer Power Operated Relief Valves (PORVs) - for one PORV inoperable for reasons other than excessive seat leakage.

2. Running ASW Pump 1-1 experiences a ground on 4 kV Bus F. The crew follows the guidance of AR PK18-23, 4KV BUS F GROUND OC ALARM, and shuts down ASW pump 1-1 after starting the 1-2 pump. Shift Forman enters TS 3.7.8.A, Auxiliary Saltwater (ASW) System for one train inoperable.

3. Low Stator Coil Cooling Water flow causes a turbine runback. The crew responds per AR PK14-19, STATOR WTR CLG SYSTEM, and OP AP-25, Rapid Load Reduction or Shutdown. The low flow condition clears quickly (approximately 925 MW), and the crew stabilizes the plant.

4. In-service Seal Injection Filter 1-1 plugs, reducing flow to RCP seals and bringing in AR PK04-22, RCP Seal Inj Fltr Delta-P Hi. Reactor Operators verify CCP seal cooling is still being maintained by CCW and ATC operator throttles RCP seal injection hand control valve, HCV-142, as needed to maintain pressurizer level. Shift Foreman establishes bands for pressurizer level and confirms field operators have been dispatched to swap seal injection filters.

5. A spurious actuation of Train A, Phase B results in the isolation of CCW Header C. The crew responds per AR PK01-08, CCW HEADER C, or alternately, OP AP-11, Section E: Loss of CCW Flow to RCPs, which calls for tripping the reactor and then tripping all four RCPs. (S5CT-1) Trip all four Reactor Coolant Pumps (RCPs) .

6. The crew enters E-0, Reactor Trip or Safety Injection and performs their immediate actions. On the trip, 4 kV bus F trips on differential. DRPI loses power, but crew is able to determine the reactor has tripped based on diverse indications (lowering reactor power and reactor trip breakers open). MDAFW Pump 1-3 is also lost due to the bus failure. Both main feedpumps trip and the TDAFW pump trips on overspeed leading to Loss of Heat Sink condition. The crew transitions to EOP FR-H.1, Response to Loss of Secondary Heat Sink. With the condenser available, Main Feed is used to restore a secondary side heat sink (S5CT-2) Establish a secondary heat sink.

The scenario is terminated once Critical Task S5CT-2 is complete.

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