ML20106E466

2020-02-DRAFT Outlines

ML20106E466
Person / Time
Site:	Diablo Canyon
Issue date:	03/06/2020
From:	Greg Werner Operations Branch IV
To:	Pacific Gas & Electric Co
References
Download: ML20106E466 (48)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Diablo Canyon Date of Exam: March 6, 2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 2 2 4 4 3 3 18 6 Emergency and Abnormal Plant 2 2 2 1 N/A 1 1 N/A 2 9 4 Evolutions Tier Totals 4 4 5 5 4 5 27 10 1 3 2 3 3 2 2 3 3 3 2 2 28 5 2.

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

• These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

• • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

Rev. 11

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

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

000007 (EPE 7) Reactor Trip, Stabilization, Ability to operate and monitor the following as they Recovery / 1 X apply to a reactor trip: (CFR 41.7 / 45.5 / 45.6) 4.2 40 EA1.03 RCS pressure and temperature 000008 (APE 8) Pressurizer Vapor Space 2.4.4 Ability to recognize abnormal indications for Accident / 3 system operating parameters that are entry-level X 4.5 42 conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 Knowledge of the operational implications of the following concepts as they apply to the Large Break X LOCA: (CFR 41.8 / 41.10 / 45.3) 4.1 55 EK1.01 Natural circulation and cooling, including reflux boiling 000015 (APE 15) Reactor Coolant Pump Ability to operate and / or monitor the following as Malfunctions / 4 they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 41.7 / 45.5 /

X 45.6) 3.7* 52 AA1.03 Reactor trip alarms, switches, and indicators 000022 (APE 22) Loss of Reactor Coolant Knowledge of the reasons for the following Makeup / 2 responses as they apply to the Loss of Reactor X Coolant Makeup: (CFR 41.5, 41.10 / 45.6 / 45.13) 3.2 45 AK3.06 RCP thermal barrier cooling 000025 (APE 25) Loss of Residual Heat Knowledge of the interrelations between the Loss of Removal System / 4 Residual Heat Removal System and the following:

X (CFR 41.7 / 45.7) 2.9 44 AK2.01 RHR heat exchangers 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Ability to operate and / or monitor the following as Control System Malfunction / 3 they apply to the Pressurizer Pressure Control X Malfunctions: (CFR 41.7 / 45.5 / 45.6) 4.0 46 AA1.01 PZR heaters, sprays, and PORVs 000029 (EPE 29) Anticipated Transient Without Scram / 1 000038 (EPE 38) Steam Generator Tube Ability to determine or interpret the following as they Rupture / 3 X apply to a SGTR: (CFR 43.5 / 45.13) 3.9* 53 EA2.12 Status of MSIV activating system 000040 (APE 40) Steam Line Rupture Ability to determine and interpret the following as Excessive Heat Transfer / 4 they apply to the Steam Line Rupture: (CFR: 43.5 /

X 45.13) 4.2 41 AA2.01 Occurrence and location of a steam line rupture from pressure and flow indications 000054 (APE 54) Loss of Main Feedwater

/4 000055 (EPE 55) Station Blackout / 6 Knowledge of the reasons for the following responses as the apply to the Station Blackout:

X (CFR 41.5 / 41.10 / 45.6 / 45.13) 4.3 39 EK3.02 Actions contained in EOP for loss of offsite and onsite power 000056 (APE 56) Loss of Offsite Power / 6 2.4.46 Ability to verify that the alarms are consistent X with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 4.2 54 45.12)

Rev. 11

ES-401 3 Form ES-401-2 000057 (APE 57) Loss of Vital AC Knowledge of the reasons for the following Instrument Bus / 6 responses as they apply to the Loss of Vital AC X Instrument Bus: (CFR 41.5,41.10 / 45.6 / 45.13) 4.1 51 AK3.01 Actions contained in EOP for loss of vital ac electrical instrument bus 000058 (APE 58) Loss of DC Power / 6 Knowledge of the operational implications of the following concepts as they apply to Loss of DC X Power: (CFR 41.8 / 41.10 / 45.3) 2.8 43 AK1.01 Battery charger equipment and instrumentation 000062 (APE 62) Loss of Nuclear Service Knowledge of the reasons for the following Water / 4 responses as they apply to the Loss of Nuclear X Service Water: (CFR 41.4, 41.8 / 45.7) 4.0 49 AK3.03 Guidance actions contained in EOP for Loss of nuclear service water 000065 (APE 65) Loss of Instrument Air / 8 Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air: (CFR 41.7 /

X 45.5 / 45.6) 2.9 56 AA1.03 Restoration of systems served by instrument air when pressure is regained 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment / 3 Knowledge of the interrelations between the (LOCA Outside Containment) and the following: (CFR: 41.7

/ 45.7)

X EK2.2 Facility*s heat removal systems, including 3.8 50 primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

(W E11) Loss of Emergency Coolant 2.2.22 Knowledge of limiting conditions for Recirculation / 4 X operations and safety limits. (CFR: 41.5 / 43.2 / 4.0 48 45.2)

(W E05) Inadequate Heat TransferLoss of Ability to determine and interpret the following as Secondary Heat Sink / 4 they apply to the (Loss of Secondary Heat Sink)

(CFR: 43.5 / 45.13)

X 3.7 47 EA2.2 Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments K/A Category Totals: 2 2 4 4 3 3 Group Point Total: 18 Rev. 11

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

000001 (APE 1) Continuous Rod Withdrawal / 1 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

X (CFR: 43.5 / 45.13) 4.4 62 AA2.05 Uncontrolled rod withdrawal, from available indications 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 Knowledge of the reasons for the following responses as they apply to Emergency Boration: (CFR X 41.5, 41.10 / 45.6 / 45.13) 4.2 60 AK3.02 Actions contained in EOP for emergency boration 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Knowledge of the reasons for the Instrumentation / 7 following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR X 41.5,41.10 / 45.6 / 45.13) 3.2 59 AK3.01 Termination of startup following loss of intermediate range instrumentation 000036 (APE 36) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 2.4.45 Ability to prioritize and interpret the significance of each X 4.1 57 annunciator or alarm. (CFR:

41.10 / 43.5 / 45.3 / 45.12) 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the X following: (CFR 41.7 / 45.7) 2.7 65 AK2.02 Auxiliary building ventilation system 000061 (APE 61) Area Radiation Monitoring System Alarms 2.4.8 Knowledge of how abnormal

/7 operating procedures are used in X 3.8 63 conjunction with EOPs. (CFR:

41.10 / 43.5 / 45.13) 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74) Inadequate Core Cooling / 4 Knowledge of the operational implications of the following concepts as they apply to the X Inadequate Core Cooling: (CFR 2.8 64 41.8 / 41.10 / 45.3)

EK1.07 Definition of saturated steam 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 Rev. 11

ES-401 5 Form ES-401-2 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 Ability to operate and / or monitor the following as they apply to the (Steam Generator Overpressure)

X (CFR: 41.7 / 45.5 / 45.6) 3.0 61 EA1.2 Operating behavior characteristics of the facility (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 Knowledge of the operational implications of the following concepts as they apply to the (High Containment Radiation)

X (CFR: 41.8 / 41.10, 45.3) 3.0 58 EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the (High Containment Radiation)

(BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals: 2 1 2 1 1 2 Group Point Total: 9 Rev. 11

ES-401 6 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

003 (SF4P RCP) Reactor Coolant Knowledge of the effect that a loss or Pump malfunction of the RCPS will have on the X following: (CFR: 41.7 / 45.6) 3.5 1 K3.02 S/G 004 (SF1; SF2 CVCS) Chemical and Knowledge of the effect of a loss or Volume Control malfunction on the following CVCS X components: (CFR: 41.7 / 45.7) 2.7 15 K6.29 Reason for excess letdown and its relationship to CCWS 005 (SF4P RHR) Residual Heat Ability to predict and/or monitor changes in Removal parameters (to prevent exceeding design limits) associated with operating the RHRS X controls including: (CFR: 41.5 / 45.5) 2.5 16 A1.03 Closed cooling water flow rate and temperature 006 (SF2; SF3 ECCS) Emergency Knowledge of bus power supplies to the Core Cooling X following: (CFR: 41.7) 3.6 22 K2.04 ESFAS-operated valves 007 (SF5 PRTS) Pressurizer Ability to (a) predict the impacts of the Relief/Quench Tank following malfunctions or operations on the PRTS; and (b) based on those predictions, use X procedures to correct, control, or mitigate the 2.6 28 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.06 Bubble formation in PZR 008 (SF8 CCW) Component Cooling Ability to monitor automatic operation of the Water CCWS, including: (CFR: 41.7 / 45.5)

X 3.2 7 A3.02 Operation of the CCW pumps, including interlocks and the CCW booster pump 010 (SF3 PZR PCS) Pressurizer Knowledge of the operational implications of Pressure Control the following concepts as the apply to the PZR X PCS: (CFR: 41.5 / 45.7) 3.5 25 K5.01 Determination of condition of fluid in PZR, using steam tables 012 (SF7 RPS) Reactor Protection Knowledge of the operational implications of the following concepts as the apply to the X RPS: (CFR: 41.5 / 45.7) 3.3* 24 K5.02 Power density 013 (SF2 ESFAS) Engineered Knowledge of the effect that a loss or Safety Features Actuation malfunction of the ESFAS will have on the X following: (CFR: 41.7 / 45.6) 4.3 23 K3.03 Containment 022 (SF5 CCS) Containment Cooling Knowledge of the physical connections and/or cause-effect relationships between the CCS X and the following systems: (CFR: 41.2 to 41.9 / 3.5 21 45.7 to 45.8)

K1.01 SWS/cooling system 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

X K4.04 Reduction of temperature and pressure 3.7 3 in containment after a LOCA by condensing steam, to reduce radiological hazard, and protect equipment from corrosion damage (spray)

Rev. 11

ES-401 7 Form ES-401-2 039 (SF4S MSS) Main and Reheat Knowledge of the effect that a loss or Steam malfunction of the MRSS will have on the X following: (CFR: 41.7 / 45.6) 3.2* 12 K3.03 AFW pumps 059 (SF4S MFW) Main Feedwater Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use X procedures to correct, control, or mitigate the 3.0* 4 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.07 Tripping of MFW pump turbine 061 (SF4S AFW) Knowledge of the effect of a loss or Auxiliary/Emergency Feedwater malfunction of the following will have on the X AFW components: (CFR: 41.7 / 45.7) 2.6 6 K6.02 Pumps 062 (SF6 ED AC) AC Electrical Ability to predict and/or monitor changes in Distribution parameters (to prevent exceeding design limits) associated with operating the ac X distribution system controls including: (CFR: 3.4 27 41.5 / 45.5)

A1.01 Significance of D/G load limits 063 (SF6 ED DC) DC Electrical Knowledge of the physical connections and/or Distribution cause-effect relationships between the DC X electrical system and the following systems: 2.7 10 (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.02 AC electrical system 064 (SF6 EDG) Emergency Diesel Ability to monitor automatic operation of the Generator X ED/G system, including: (CFR: 41.7 / 45.5) 4.1 14 A3.01 Automatic start of compressor and ED/G 073 (SF7 PRM) Process Radiation 2.2.38 Knowledge of conditions and limitations Monitoring X in the facility license. (CFR: 41.7 / 41.10 / 43.1 3.6 19

/ 45.13) 076 (SF4S SW) Service Water Knowledge of the physical connections and/or cause-effect relationships between the SWS X and the following systems: (CFR: 41.2 to 41.9 / 3.8* 26 45.7 to 45.8)

K1.05 D/G 078 (SF8 IAS) Instrument Air Ability to manually operate and/or monitor in X the control room: (CFR: 41.7 / 45.5 to 45.8) 3.1 20 A4.01 Pressure gauges 103 (SF5 CNT) Containment Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, X or mitigate the consequences of those 3.5* 17 malfunctions or operations (CFR: 41.5 / 43.5 /

45.3 / 45.13)

A2 03 Phase A and B isolation 053 (SF1; SF4P ICS*) Integrated Control 006 (SF2; SF3 ECCS) Emergency 2.2.39 Knowledge of less than or equal to one Core Cooling X hour Technical Specification action statements 3.9 9 for systems. (CFR: 41.7 / 41.10 / 43.2 / 45.13) 008 (SF8 CCW) Component Cooling Ability to predict and/or monitor changes in Water parameters (to prevent exceeding design X limits) associated with operating the CCWS 2.9 18 controls including: (CFR: 41.5 / 45.5)

A1.02 CCW temperature Rev. 11

ES-401 8 Form ES-401-2 010 (SF3 PZR PCS) Pressurizer Ability to monitor automatic operation of the Pressure Control PZR PCS, including: (CFR: 41.7 / 45.5)

X 3.0 5 A3.01 PRT temperature and pressure during PORV testing 026 (SF5 CSS) Containment Spray Knowledge of bus power supplies to the X following: (CFR: 41.7) 3.4* 11 K2.01 Containment spray pumps 076 (SF4S SW) Service Water Ability to manually operate and/or monitor in X the control room: (CFR: 41.7 / 45.5 to 45.8) 3.5* 8 A4.04 Emergency heat loads 078 (SF8 IAS) Instrument Air Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following:

X (CFR: 41.7) 3.1* 2 K4.03 Securing of SAS upon loss of cooling water 103 (SF5 CNT) Containment Knowledge of containment system design feature(s) and/or interlock(s) which provide for X the following: (CFR: 41.7) 3.1 13 K4.06 Containment isolation system K/A Category Point Totals: 2 2 3 3 2 2 3 3 3 3 2 Group Point Total: 28 Rev. 11

ES-401 9 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Knowledge of the operational implications of Indication the following concepts as they apply to the X RPIS: (CFR: 41.5 / 45.7) 2.7 37 K5.01 Reasons for differences between RPIS and step counter 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Knowledge of the effect that a loss or Monitor malfunction of the ITM system will have on the X following: (CFR: 41.7 / 45.6) 3.5* 38 K3.01 Natural circulation indications 027 (SF5 CIRS) Containment Iodine Ability to manually operate and/or monitor in Removal X the control room: (CFR: 41.7 / 45.5 to 45.8) 3.3* 34 A4.03 CIRS fans 028 (SF5 HRPS) Hydrogen 2.1.19 Ability to use plant computers to Recombiner and Purge Control X evaluate system or component status. (CFR: 3.9 35 41.10 / 45.12) 029 (SF8 CPS) Containment Purge Knowledge of design feature(s) and/or interlock(s) which provide for the following:

X (CFR: 41.7) 2.9 29 K4.02 Negative pressure in containment 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Ability to predict and/or monitor changes in Equipment parameters (to prevent exceeding design X limits) associated with operating the Fuel 2.9 31 Handling System controls including: (CFR:

A1.02 Water level in the refueling canal 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Ability to (a) predict the impacts of the following Dump/Turbine Bypass Control malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the X consequences of those malfunctions or 3.6 32 operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.02 Steam valve stuck open 045 (SF 4S MTG) Main Turbine Knowledge of the physical connections and/or Generator cause-effect relationships between the MT/G X system and the following systems: (CFR: 41.2 2.6 36 to 41.9 / 45.7 to 45.8)

K1.06 RCS, during steam valve test 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste Ability to monitor automatic operation of the X Liquid Radwaste System including: (CFR: 41.7 3.6 33 A3.02 Automatic isolation 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring Rev. 11

ES-401 10 Form ES-401-2 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection Knowledge of the effect of a loss or malfunction on the Fire Protection System X following will have on the: (CFR: 41.7 / 45.7) 2.6 30 K6.04 Fire, smoke, and heat detectors 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10 Rev. 11

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 #

2.1. 2.1.28 Knowledge of the purpose and function of major 4.1 73 system components and controls. (CFR: 41.7) 2.1. 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR: 41.1 / 43.6 4.3 74

/ 45.6)

1. Conduct of 2.1. 2.1.44 Knowledge of RO duties in the control room during Operations fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage 3.9 66 facility, systems operated from the control room in support of fueling operations, and supporting instrumentation. (CFR: 41.10 / 43.7 / 45.12)

Subtotal 3 2.2. 2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and 4.6 68 designated power levels. (CFR: 41.6 / 41.7 / 45.2) 2.2. 2.2.40 Ability to apply Technical Specifications for a 3.4 67

2. Equipment system. (CFR: 41.10 / 43.2 / 43.5 / 45.3)

Control 2.2. 2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand 4.2 70 how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)

Subtotal 3 2.3. 2.3.11 Ability to control radiation releases. (CFR: 41.11 /

3.8 75 43.4 / 45.10) 2.3. 2.3.12 Knowledge of radiological safety principles

3. Radiation pertaining to licensed operator duties, such as Control containment entry requirements, fuel handling 3.2 71 responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 / 45.10)

Subtotal 2 2.4. 2.4.17 Knowledge of EOP terms and definitions. (CFR:

3.9 69 41.10 / 45.13)

4. Emergency 2.4. 2.4.35 Knowledge of local auxiliary operator tasks during Procedures/Plan an emergency and the resultant operational effects. 3.8 72 (CFR: 41.10 / 43.5 / 45.13)

Subtotal 2 Tier 3 Point Total 10 Rev. 11

ES-401 PWR Examination Outline Form ES-401-2 Facility: Diablo Canyon Date of Exam: March 6, 2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 18 3 3 6 Emergency and Abnormal Plant 2 N/A N/A 9 2 2 4 Evolutions Tier Totals 27 5 5 10 1 28 3 2 5 2.

Plant 2 10 1 2 3 Systems Tier Totals 38 4 4 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

• These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

• • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

Rev. 11

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

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

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

Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 Ability to determine or interpret the following as they X apply to a small break LOCA: (CFR 43.5 / 45.13) 4.2 88 EA2.34 Conditions for throttling or stopping HPI 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component X 2.4.18 Knowledge of the specific bases for EOPs. 89 4.0 Cooling Water / 8 (CFR: 41.10 / 43.1 / 45.13) 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient 2.4.9 Knowledge of low power/shutdown Without Scram / 1 X implications in accident (e.g., loss of coolant 87 4.2 accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 / 43.5 / 45.13) 000038 (EPE 38) Steam Generator Tube Rupture / 3 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Ability to determine and interpret the following as Feedwater /4 they apply to the Loss of Main Feedwater (MFW):

X 3.3* 86 (CFR: 43.5 / 45.13)

AA2.08 Steam flow-feed trend recorder 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and 2.4.21 Knowledge of the parameters and logic used Electric Grid Disturbances / 6 to assess the status of safety functions, such as X reactivity control, core cooling and heat removal, 85 4.6 reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR:

41.7 / 43.5 / 45.12)

Rev. 11

ES-401 3 Form ES-401-2 (W E04) LOCA Outside Containment / 3 Ability to determine and interpret the following as they apply to the (LOCA Outside Containment)

(CFR: 43.5 / 45.13)

X 4.2 84 EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

(W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals: 3 3 Group Point Total: 6 Rev. 11

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 2.4.6 Knowledge of EOP X mitigation strategies. (CFR: 41.10 4.7 93

/ 43.5 / 45.13) 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 2.4.2 Knowledge of system set points, interlocks and automatic X actions associated with EOP 4.6 91 entry conditions. (CFR: 41.7 /

45.7 / 45.8) 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms

/7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /

4 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 (W E10) Natural Circulation with Steam Void in Vessel Ability to determine and interpret with/without RVLIS/4 the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS)

X 3.9 90 (CFR: 43.5 / 45.13)

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(BW E13 & E14) EOP Rules and Enclosures Rev. 11

ES-401 5 Form ES-401-2 (W E08) RCS OvercoolingPressurized Thermal Shock / 4 EA2. Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock) (CFR: 43.5 / 45.13)

X 4.2 92 EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals: 2 2 Group Point Total: 4 Rev. 11

ES-401 6 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

003 (SF4P RCP) Reactor Coolant Pump 004 (SF1; SF2 CVCS) Chemical and Ability to (a) predict the impacts of the Volume Control following malfunctions or operations on the CVCS; and (b) based on those predictions, X use procedures to correct, control, or mitigate 4.2 76 the consequences of those malfunctions or operations: (CFR: 41.5/ 43/5 / 45/3 / 45/5)

A2.27 Improper RWST boron concentration 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 013 (SF2 ESFAS) Engineered Ability to (a) predict the impacts of the Safety Features Actuation following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, X or mitigate the consequences of those 4.8 80 malfunctions or operations; (CFR: 41.5 / 43.5 /

45.3 / 45.13)

A2.01 LOCA 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat 2.1.7 Ability to evaluate plant performance and Steam make operational judgments based on X operating characteristics, reactor behavior, and 4.7 77 instrument interpretation. (CFR: 41.5 / 43.5 /

45.12 / 45.13) 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW) Ability to (a) predict the impacts of the Auxiliary/Emergency Feedwater following malfunctions or operations on the AFW; and (b) based on those predictions, use X procedures to correct, control, or mitigate the 3.5 78 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.07 Air or MOV failure 062 (SF6 ED AC) AC Electrical Distribution Rev. 11

ES-401 7 Form ES-401-2 063 (SF6 ED DC) DC Electrical 2.2.25 Knowledge of the bases in Technical Distribution Specifications for limiting conditions for X 4.2 79 operations and safety limits. (CFR: 41.5 / 41.7

/ 43.2) 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals: 3 2 Group Point Total: 5 Rev. 11

ES-401 8 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor 2.4.3 Ability to identify post-accident X 3.9 83 Coolant instrumentation. (CFR: 41.6 / 45.4) 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator Ability to (a) predict the impacts of the following malfunctions or operations on the S/GS; and (b) based on those predictions, use procedures X to correct, control, or mitigate the 3.6 82 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

A2.03 Pressure/level transmitter failure 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 2.1.32 Ability to explain and apply system limits X 4.0 81 and precautions. (CFR: 41.10 / 43.2 / 45.12) 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals: 1 2 Group Point Total: 3 Rev. 11

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 #

2.1. 2.1.34 Knowledge of primary and secondary plant 3.5 94 chemistry limits. (CFR: 41.10 / 43.5 / 45.12)

1. Conduct of 2.1. 2.1.43 Ability to use procedures to determine the effects Operations on reactivity of plant changes, such as reactor coolant 4.3 97 system temperature, secondary plant, fuel depletion, etc.

(CFR: 41.10 / 43.6 / 45.6)

Subtotal 2 2.2. 2.2.37 Ability to determine operability and/or availability of 4.6 98 safety related equipment. (CFR: 41.7 / 43.5 / 45.12)

2. Equipment Control 2.2. 2.2.43 Knowledge of the process used to track inoperable 3.3 95 alarms. (CFR: 41.10 / 43.5 / 45.13)

Subtotal 2 2.3. 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry

3. Radiation 3.8 99 requirements, fuel handling responsibilities, access to Control locked high-radiation areas, aligning filters, etc. (CFR:

41.12 / 43.4 / 45.9 / 45.10)

Subtotal 2 2.4. 2.4.29 Knowledge of the emergency plan. (CFR: 41.10 /

4.4 96 43.5 / 45.11)

4. Emergency Procedures/Plan 2.4. 2.4.40 Knowledge of SRO responsibilities in emergency 4.5 100 plan implementation. (CFR: 41.10 / 43.5 / 45.11)

Subtotal 2 Tier 3 Point Total 7 Rev. 11

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A None at the Draft Level. NRC developed outline.

Rev. 11

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Diablo Can~on Date of Examination: 02/24/2020 Examination Level: RO ~ SRO D Operating Test Number: L181 Administrative Topic (see Note) Type Describe activity to be performed Code*

Conduct of Operations Apply Overtime Limit Restrictions (NRCL 181-A1) M,R 2.1 .5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc (2.9)

(NRCADM061-COO-R02)

Conduct of Operations Estimate Decay Heat and Heat Removal Rate (NRCL 181-A2) M, R 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc.

(3 .9)

(Bank: LJC-014)

Equipment Control Perform STP I-IC (NRCL 181-A3) N, R 2.2.37 Ability to determine Operability and/or availability of safety related equipment.

(3 .6)

Radiation Control Determine Primary to Secondary Leakrate N, R (NRCL 181-A4) 2.3.1 1 Ability to control radiation releases. (3.8)

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 (2: 1)

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

Rev 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Diablo Cany:on Date of Examination : 02/24/2020 Examination Level : RO D SRO ~ Operating Test Number: L181 Administrative Topic (see Note) Type Describe activity to be performed Code*

Conduct of Operations Apply Overtime Limit Restrictions (NRCL 181-A5) M, R 2.1 .5 Ability to use procedures related to shift staffing , such as minimum crew complement, overtime limitations, etc (3.9)

(NRCADM061-COO-SR01)

Conduct of Operations Review Estimated Decay Heat and Heat Removal Rate (NRCL 181 -A6) M, R 2.1 .25 Ability to interpret reference materials such as graphs, curves , tables, etc.

(4.2)

(Bank: LJC-014)

Equipment Control Determine 230 kV Operability (NRCL 181-A?) N, R 2.2.37 Ability to determine Operability and/or availability of safety related equipment.

(4.6)

Radiation Control Determine Primary to Secondary Leakrate and Required Action (NRCL 181-A8) N, R 2.3.11 Ability to control radiation releases. (4.3)

Emergency Plan Review Emergency Notification for Steam Generator Tube Rupture (NRCL 181-A9) N, R 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation.

(4 .5)

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 (:5 3 for ROs ; :5 4 for SROs and RO retakes)

(N)ew or (M)odified from bank (2: 1)

(P)revious 2 exams (:5 1, randomly selected)

Rev O

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Diablo Can1'on Date of Examination : 02/24/2020 Exam Level : RO D SRO-I D SRO-U ~ Operating Test Number: L181 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) (001 .A2 .11) Dropped Rods During Rod Misalignment Verification A,M,S 1 (Modified LJC-066)

b. (S2) (013.A1 .02) Resp to Changing Plant Params During Rx Trip Resp A,N,EN,L,S 2 C.

d.

e.

f.

g.

h.

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. (P1) (010.A2.02) Transfer Pzr Heater Grp 23 to Backup Pwr (LJP-029A) D 3

j. (P2) (062.A2 .11) Transfer the TSC to Vital Power(LJP-058A) A,D ,E,L 6

k. (P3) (013.A4.01) Verfiy Slowdown Isolation Outside Containment (Bank E,D,R 8 LJP-096)

• 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 /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank :5 9/:5 8/:5 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 :5 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 Canz'on Date of Examination : 02/24/2020 Exam Level: RO ~ SRO-I D SRO-U D Operating Test Number: L181 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) (001.A2.11) Dropped Rods During Rod Misalignment Verification A,M,S 1 (Modified LJC-066)

b. (S2) (013.A 1.02) Resp to Changing Plant Params During Rx Trip Resp A,N,EN,L,S 2

c. (S3) (E04 .EA 1.1) Isolate LOCA Outside Containment (Bank LJC-118) D,L,S 3

d. (S4P) (011.EA 1.11) Transfer to Cold Leg Recirc (Bank LJC-27 A) A,D,L,S 4P

e. (S4S) (059 .A2.11) Monitor Plant Response During Full Load Rejection A,N,S 4S

f. (S5) (E14.E1 .2) Manually Initiate Containment Spray (Bank LJC-010) D,E,L,S 5

g. (S6) (064.A4.06) Crosstie Vital Bus G to H (LJC-032) D,E,L,S 6

h. (S8) (068.AA2.17) Fire in 480V Bus G Switchgear Room A,C,N,S 8 In-Plant Systems:* 3 for RO , 3 for SRO-I , and 3 or 2 for SRO-U

i. (P1) (010 .A2.02) Transfer Pzr Heater Grp 23 to Backup Pwr (LJP-029A) D 3

j. (P2) (062.A2 .11) Transfer the TSC to Vital Power(LJP-058A) A,D ,E,L 6

k. (P3) (013.A4.01) Verfiy Blowdown Isolation Outside Containment (Bank E,D,R 8 LJP-096)

• 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 /SR0-1/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 21/~1/.::1 (EN)gineered safety feature 2 1/~ 1/.:: 1 (control room system)

(L)ow-Power/Shutdown .:: 1/2 1/.:: 1 (N)ew or (M)odified from bank including 1(A) .:: 2/.:: 2/21 (P)revious 2 exams s 3/s 3/S 2 (randomly selected)

(R)CA 21/.::1/.::1 (S)imulator Rev 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Diablo Canton Date of Examination : 02/24/2020 Exam Level : RO D SRO-I [Z] SRO-U D Operating Test Number: L181 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) (001 .A2.11) Dropped Rods During Rod Misalignment Verification A,M,S 1 (Modified LJC-066)

b. (S2) (013.A 1.02) Resp to Changing Plant Params During Rx Trip Resp A,N ,EN,L,S 2

c. (S3) (E04.EA1.1) Isolate LOCA Outside Containment (Bank LJC-118) D,L,S 3

d. (S4P) (011.EA 1.11) Transfer to Cold Leg Recirc (Bank LJC-27A) A,D,L,S 4P

e. (S4S) (059.A2.11) Monitor Plant Response During Full Load Rejection A,N,S 4S

f. (S5) (E14 .E1.2) Manually Initiate Containment Spray (Bank LJC-010) D,E,L,S 5 g.

h. (S8) (068 .M2.17) Fire in 480V Bus G Switchgear Room A,C,N,S 8 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. (P1) (010.A2.02) Transfer Pzr Heater Grp 23 to Backup Pwr (LJP-029A) D 3 j . (P2) (062.A2 .11) Transfer the TSC to Vital Power(LJP-058A) A,D,E,L 6

k. (P3) (013.A4.01) Verfiy Slowdown Isolation Outside Containment (Bank E,D,R 8 LJP-096)

• 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 /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank s 9/s 8/s 4 (E)mergency or abnormal in-plant ~1,~11~1 (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 s 3/s 3/s 2 (randomly selected)

(R)CA ~1,~11~1 (S)imulator Rev 0

ES-301 Transient and Event Checklist Form ES-301-5 Group 1 (U1, I1, R1, R2)

Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A E Scenarios P V S1 S2 S3 S4 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 RO NOR 1 1 1 SRO-I I/C 1,2,3,5,7 5 4 4 2 SRO-U MAJ 4 1 2 2 1 TS 1,2 2 0 2 2 RX 1 1 1 1 0 RO NOR 0 1 1 1 SRO-I1 1,2,3,4,6, I/C 2,3,4,6 7,8 1,3,5,7 15 4 4 2 SRO-U MAJ 5 5 4 3 2 2 1 TS 3,4 2,4 4 0 2 2 RX 0 1 1 0 RO1 NOR 0 1 1 1 SRO-I 1,2,3,5,6, I/C 2,3,4,5,6 1,4,7,8 7 15 4 4 2 SRO-U MAJ 5 5 4 3 2 2 1 TS 0 0 2 2 RX 1 1 1 1 0 RO2 NOR 0 1 1 1 SRO-I 2,3,4,6,7, I/C 2,4 8 8 4 4 2 SRO-U MAJ 5 5 2 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 (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 0

ES-301 Transient and Event Checklist Form ES-301-5 Group 2 (I2, I3, R3, R4)

Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A E Scenarios P V S1 S2 S3 S4 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 1 1 0 RO NOR 0 1 1 1 SRO-I2 I/C 2,3,4,6 1,2,3,5,7 2,3,4,7 13 4 4 2 SRO-U MAJ 5 4 5,6 4 2 2 1 TS 3,4 1,2 4 0 2 2 RX 0 1 1 0 RO NOR 0 1 1 1 SRO-I3 1,2,3,4,6, I/C 7,8 1,3,5,7 1,2,3,4,7 16 4 4 2 SRO-U MAJ 5 4 5,6 4 2 2 1 TS 2,4 1,3 4 0 2 2 RX 1 1 1 1 0 RO3 NOR 0 1 1 1 SRO-I 2,3,4,6,7, I/C 2,4 8

1,3,4,7 12 4 4 2 SRO-U MAJ 5 5 5,6 4 2 2 1 TS 0 0 2 2 RX 0 1 1 0 RO4 NOR 0 1 1 1 SRO-I 1,2,3,5,6, I/C 2,3,4,5,6 1,4,7,8 7 15 4 4 2 SRO-U MAJ 5 5 4 3 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 (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 0

ES-301 Transient and Event Checklist Form ES-301-5 Group 3 (I4, I5, R5, R6)

Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A E Scenarios P V S1 S2 S3 S4 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 1 1 0 RO NOR 0 1 1 1 SRO-I4 I/C 2,3,4,6 1,2,3,5,7 2,3,4,7 13 4 4 2 SRO-U MAJ 5 4 5,6 4 2 2 1 TS 3,4 1,2 4 0 2 2 RX 0 1 1 0 RO NOR 0 1 1 1 SRO-I5 1,2,3,4,6, I/C 7,8 1,3,5,7 1,2,3,4,7 16 4 4 2 SRO-U MAJ 5 4 5,6 4 2 2 1 TS 2,4 1,3 4 0 2 2 RX 1 1 1 1 0 RO5 NOR 0 1 1 1 SRO-I 2,3,4,6,7, I/C 2,4 8

1,3,4,7 12 4 4 2 SRO-U MAJ 5 5 5,6 4 2 2 1 TS 0 0 2 2 RX 0 1 1 0 RO6 NOR 0 1 1 1 SRO-I 1,2,3,5,6, I/C 2,3,4,5,6 1,4,7,8 7 15 4 4 2 SRO-U MAJ 5 5 4 3 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 (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.

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ES-301 Transient and Event Checklist Form ES-301-5 Group 4 (I6, I7, R7)

Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A E Scenarios P V S1 S2 S3 S4 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 1 1 0 RO NOR 0 1 1 1 SRO-I6 I/C 2,3,4,6 1,4,7,8 1,2,3,5,7 13 4 4 2 SRO-U MAJ 5 5 4 3 2 2 1 TS 3,4 1,2 4 0 2 2 RX 0 1 1 0 RO NOR 0 1 1 1 SRO-I7 1,2,3,4,6, I/C 2,3,4,5,6 7,8 1,3,5,7 16 4 4 2 SRO-U MAJ 5 5 4 3 2 2 1 TS 2,4 2 0 2 2 RX 1 1 1 1 0 RO7 NOR 0 1 1 1 SRO-I 2,3,4,6,7, 1,2,3,5,6, I/C 2,4 8 7 14 4 4 2 SRO-U MAJ 5 5 4 3 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.

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ES-301 Transient and Event Checklist Form ES-301-5 Group x (In, In, Rn, Rn)

Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A E Scenarios P V S5 (Spare) N-Day N-Day N-Day 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 RO NOR 1 1 1 SRO-I 1,2,3,4,8, 1,3,4,7,8, I/C 9 2,4,8,9 9

4 4 2 SRO-U MAJ 5,6 5,6 5,6 2 2 1 TS 1,3 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.

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Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 1 Op-Test No: L181 NRC Examiners: Operators:

Initial Conditions: 2% with MFW in service, aligned to Start-Up Power. MOL with CFCU 1-1 OOS.

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 2% to 8% OP L-3, sec 6.28.

SRO) 2 VLV_CVC22_2 .5 delay=0 ramp=15 I (ALL) Regen Hx Isolation Valve, LCV-459, fails to mid-position (OP AP-18).

3 H_V1_034M_1, XMT_VEN6_3, TS, C CFCU 1-2 high stator/bearing temperature due to low CCW XMT_VEN7_3, XMT_VEN8_3 (BOP, flow (AR PK01-21, TS 3.6.6.C).

SRO) 4 RLY_PPL63_2 OPEN TS, I SSPS relay actuation causes inadvertent start of TDAFW pump (ALL) and blowdown sample isolation valves to close (AR PK04-03, RLY_PPL59_2 OPEN OP D-1:III, OP1.DC10; TS 3.7.5.B).

5 MAL_MSS4 1.57E+07 ramp=30 M (ALL) MSLB outside containment.

6 VLV_MSS7_2, VLV_MSS8_2, C (ALL) All MSIVs fail open; No manual close for FCV-42.

VLV_MSS9_2, VLV_MSS10_2 1 7 MAL_PPL3B BOTH C (BOP) Safety Injection, Train B fails to actuate.

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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) 6

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

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

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

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

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

7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S1CT-1) Shutdown TD AFW pump prior to Steam Carryover into the steam lines can result in damage

• Tech Spec 3.3.2 Generator Overfill (S/G wide range greater than to downstream piping and valves, placing the Basis 100%) by either: secondary heat sink at risk. High steam generator Documentation

• Closing the LCVs to the individual S/Gs level can also result in reactivity excursions due to excessive cooldown of the primary system.

OR

• Closing steam supply valves FCV-37 and FCV-38 to leads 1 and 2 respectively OR

• Directing FCV-95 closed in the field (S1CT-2) Stop uncontrolled RCS cooldown before a An event or series of events which leads to a

• Background severe challenge to Integrity Safety Function relatively rapid and severe reactor vessel Information for develops (magenta path on F-0.4 RCS Integrity) as downcomer cooldown can result in a thermal WOG Emergency follows: shock to the vessel wall that may lead to a small Response

• Verify FCV-42 closed (S/G 1-2 steamline flaw, which may already exist in the vessel wall, Guideline isolation). growing into a larger crack. The growth or

• Close/verify closed S/G Blowdown isolation extension of such a flaw may lead, in some cases valves FCV-761, FCV-154, and FCV-248. (where propagation is not stopped within the wall),

• Verify all steam dumps closed. to a loss of vessel integrity

• Isolate feed flow to S/G 1-2 by closing/verifying closed LCV-107 and LCV-111.

(Note: LCV-107 is critical only when TDAFW pump is running or capable of an autostart).

• Isolate steam flow from S/G 1-2 by closing/verifying closed FCV-37.

• Throttling Feed flow to S/Gs 1-1, 1-3, and 1-4 while maintaining the minimum heatsink requirements (435 gpm until S/G NR level is greater than 15% in one non-faulted S/G).

(S1CT-3) Terminate SI prior to rupture of PRT by Failure to terminate ECCS flow when SI termination

• Westinghouse closing 8801A/B and/or 8803A/B. criteria are met results in overfill of the Pressurizer Owners Group and the eventual rupture of the PRT. This WCAP-17711-NP (Note: CT is met by closing either 8801A/B OR 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.

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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. Regen Hx Isolation Valve, LCV-459, drifts to mid-position causing letdown orifice valve 8149C to close.

Shift Foreman enters OP AP-18, Letdown Line Failure. Excess Letdown is established per OP B-1A:IV CVCS - Excess Letdown - Place In Service and Remove From Service.

3. CFCU 1-2 has a loss of CCW flow due to debris migration causing stator and motor bearing temperatures to rise rapidly and bring in annunciator alarm PK01-21, Contmt Fan Clr. Reactor operators identify low flow indications on vertical boards and rapidly rising stator/bearing temperatures using plant process computer trends. The crew secures the CFCU to prevent motor damage and contacts maintenance/engineering for assistance. Shift Foreman enters TS 3.6.6 Condition C, one required CFCU system inoperable such that a minimum of two CFCUs remain OPERABLE (7 day).

4. SSPS relay actuation results in Turbine Driven AFW (TDAFW) Pump Steam Supply Isolation Valve, FCV-95, failing open and isolation of half of the blowdown sample valves inside and outside containment. S/G levels rise and RCS temperature lowers. FCV-95 cannot be closed and the crew must isolate the TDAFW Pump by closing the LCVs to the individual S/Gs or by closing steam supply valves FCV-37 and FCV-38 from leads 2 and 3 respectively, or by directing FCV-95 manually closed in the field (S1CT-1) Shutdown TD AFW pump prior to Steam Generator Overfill. Shift Foreman implements TS 3.7.5.B, AFW System for one AFW train inoperable (72 hrs).

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. Train B of Safety Injection fails to actuate, requiring the crew to perform numerous manual alignments and pump starts as part of Appendix E. Appendix E. ESF Auto Actions, Secondary and Auxiliaries Status.

7. All four main steam isolation valves fail open. Steam leads 1, 3, and 4 may be closed from the control room, but lead 2 (FCV-42) requires field action. The crew transitions to EOP E-2, Faulted Steam Generator Isolation to isolate S/G 1-2 and dispatches an operator to locally close FCV-42 as part of the critical task to stop the uncontrolled cooldown (S1CT-2) Stop uncontrolled cooldown before a severe challenge (magenta path ) develops on F-0.4 RCS Integrity.

8. The crew transitions to EOP E-1.1, SI Termination where they complete the final critical task of the scenario (S1CT-3) Terminate SI prior to rupture of PRT.

The scenario is terminated once the final critical task is complete.

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Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 2 Op-Test No: L181 NRC Examiners: Operators:

Initial Conditions: 75% Power, MOL with AFW 1-2 cleared for a bearing oil leak Turnover: At 75% power for SCCW HX Clearance Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 PZ04PRZ_PRH11_.. I (ATC, Pressurizer heater control problem causes RCS pressure to RCARRL_IPC455JXTVSP 0 SRO) slowly lower (AR PK05-16, OP AP-13).

2 XMT_PZR24_3 ramp=1 TS, I PT-474, Pressurizer Pressure Transmitter, Fails Low ( OP AP-5, (BOP, TS 3.3.1.E, M, 3.3.2.D, 3.4.11,.B).

SRO) 3 XMT_CVC2_3 ramp=75 I(BOP, PT-135 Fails High causing letdown pressure control valve to SRO) go full open (AR PK04-21).

4 MAL_RCS4H 30.0 TS, C 30 gpm SGTL on loop 4; plant shutdown required (OP AP-3, (ALL) AP-25, TS 3.4.13.B).

5 MAL_RCS4H 300.0 ramp=60 M (ALL) Tube leak grows to 300 gpm rupture during ramp offline.

6 RC41SW_52VE3_CSTA_SWIT 1 C (BOP, RCPs trip off shortly after transferring to startup power.

cd='H_V5_195B_1'... SRO) 7 MAL_EPS4D_2 DIFFERENTIAL C (ALL) 4kV Bus G differential trip on transfer to startup power.

cd='h_v4_221r_1' 8 VLV_PZR5_1 1 cd='H_V2_014M_1 C (ALL) PORV PCV-456 fails closed.

LT 1915' delay=8 Aux spray valves 8145 and 8148 fail closed.

VLV_CVC17_1 1 VLV_CVC27_1 1

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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 1,2,3,4,5,6,7,8) 8

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

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-3, ECA-3.3) 2

6. EOP contingencies requiring substantive actions (0-2) (ECA-3.3) 1

7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S2CT-1) Isolate the ruptured steam SG inventory increase leads to water release

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

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

• SGTR UFSAR 15.4.3

• WCAP-17711-NP Isolate feedwater by ensuring closed:

LCV-109 (TDAFW Level Control Valve)

LCV-113 (MDAFW Level Control Valve)

(S2CT-2) Perform RCS cooldown at Transition to contingency procedures to address

• W Margin to Overfill maximum rate to CETC target temperature inadequate subcooling or Pressurized Thermal (CN-CRA-05-53 Rev1) specified in E-3, step 6, using steam dumps Shock conditions results in delaying RCS

• SGTR UFSAR 15.4.3 such that RCS subcooled margin still exists depressurization and SI termination. This delay

• WCAP-17711-NP following the cooldown. allows excess inventory in the ruptured S/G to continue to increase, with the potential of For 40% steam dumps, maximum rate limit is 120 challenging SG overpressure components or psi/min (PPC value). Above this, main steam line causing an overfill condition to occur.

isolation will occur. Operator should attempt highest rate possible without getting main steam line isolation (not critical). If steam line isolation occurs, maximum rate cooldown requires 10% steam dumps on intact S/Gs to be at least 90% open.

(S2CT-3) Restore RCS pressure control by Failure to establish RCS pressure control results in

• W Margin to Overfill (CN-restarting RCP 1-2 and returning from EOP safety injection termination occurring at a CRA-05-53 Rev1) contingency procedure (ECA-3.3) to normal significantly higher RCS pressure. Avoidable

• SGTR UFSAR 15.4.3 EOP flow path (E-3) primary to secondary leakage into a ruptured

• WCAP-17711-NP steam generator complicates results in the unnecessary continuation of a degraded condition beyond that 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 #2

1. The firing rate for Pressurizer Proportional Heaters fail to minimum and RCS pressure begins to lower slowly. Backup heater relay pc455jx fails to actuate when RCS pressure reaches 2210 psig which also results in a failure of annunciator AR PK05-16, PZR PRESSURE HI/LO activation. Pressurizer pressure control HC-455K output signal indicates that backup heaters should be on and proportional heaters should be operating at the maximum firing rate. When RCS pressure reaches 2185 psig, AR PK05-16, PZR PRESSURE HI/LO alarms based on Pressurizer pressure going below the PORV interlock setpoint. The crew follows AR PK05-16 guidance to manually turn on all backup pressurizer heaters and follow up with the actions of OP AP-13, Malfunction of Reactor Pressure Control System, to restore pressure to normal using manual control. The crew may diagnose the failure prior to the annunciator response activating and enter OP AP-13 directly which also directs turning on all backup heaters. If RCS pressure falls below 2175 psig, the Shift Foreman will enter Tech Spec 3.4.1 for RCS departure from nucleate boiling limits (2 hr).

2. PT-474, Pressurizer Pressure Transmitter, Fails low bringing in multiple Annunciator Alarms. There is no transient associated with this failure, but the failure has significant Operational implications due to its input function as part of various Reactor Protection logic schemes. When failed low, PT-474s interlock function prevents Pressurizer PORVs PCV-455C and PCV-474 from opening on a valid high pressure signal; only PCV-456 will still function. The Shift Foreman may elect to enter any of the associated Annunciator Response alarms, but in all cases, will be directed to OP AP-5, Malfunction of Eagle-21 Protection or Control Channel, which provides information regarding indications, controls, and a listing of the associated Tech Specs:

• TS 3.3.1.E, PC-474C High Press Trip & TC 441C OT Delta T Trip (72 hrs).

• TS 3.3.1.M, PC 474A Low Press Trip (72 hrs).

• TS 3.3.2.D, PC 474D Low Press S.I. (72 hrs).

• TS 3.4.11, PC 474B PORV Press Interlock o PCV-474 (non-class I), 3.4.11.B1 & B2 to close & remove power from associated block valve (1 hr) o PCV-455C (class I), 3.4.11.B1 & B2 to close & remove power from associated block valve (1 hr);

3.4.11.B3 to return to OPERABLE status (72 hrs).

3. PT-135, Transmitter for Letdown Pressure Control Valve, fails High causing letdown pressure control valve to go full open and letdown flow to rise. AR PK04-21, LETDOWN PRESS / FLOW TEMP comes into alarm for Letdown Heat Exchanger Outlet Pressure High as a result of the failed transmitter, while actual letdown pressure lowers to approximately 90 psig as a result of full open control valve response.

Letdown flow increases approximately 8 gpm above normal, resulting in a charging/letdown mismatch.

Procedural guidance in AR PK04-21 directs crew to take manual control of PCV-135. Crew performs diagnostic brief to determine nature of the malfunction as well as actions required to restore letdown pressure back to normal band.

Steam Generator 1-4 develops a 30 gpm tube leak as indicated by rising counts on various radiation 4.

monitors. The crew enters OP AP-3, Steam Generator Tube Failure. Shift Foreman determines TS 3.4.13.B, RCS Operational Leakage applies and enters OP AP-25, Rapid Load Reduction or Shutdow for the ramp off-line.

5. During the ramp the tube leak develops into a 300 gpm rupture. The crew determines the leak is substantial in size based on a rapid drop in pressurizer level. The Shift Foreman directs a reactor trip and safety injection and the crew enters EOP E-0, Reactor Trip or Safety Injection.

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6. On the transfer to start up power, 4 kV bus G experiences a differential fault. The crew also notes all four RCPs trip off on the transfer, but power to the pumps remains available.

7. The crew transitions to EOP E-3, Steam Generator Tube Rupture, based on RM-74 and rising S/G 1-4 level, where they address the following critical tasks:

• (S2CT-1) Isolate the ruptured steam generator from the intact steam generators prior to commencing cooldown.

• (S2CT-2) Perform RCS cooldown at maximum rate to CETC target temperature.

8. The crew discovers they are unable to depressurize the RCS based on a combination of equipment failures:

• With no running RCPs, normal pressurizer spray is ineffective.

• PCV-456 has failed closed and will not opened.

• Block valve 8000B, associated with PCV-455C was closed during an earlier event and cannot be opened due loss of bus G.

• Block valve 8000A, associated with PCV-474 was also closed during the earlier event and will not open.

• Aux Spray valves 8145 and 8148 are failed closed as well.

9. The crew transitions to EOP ECA-3.3, SGTR Without Pressurizer Pressure Control, where they complete critical task (S2CT-3) Restore RCS pressure control by restarting RCP 1-2. Following procedural guidance, the crew determines normal spray is available and returns to EOP E-3, Steam Generator Tube Rupture.

The scenario is terminated once the crew has started RCP 1-2 and transitioned back into E-3.

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Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 3 Op-Test No: L181 NRC Examiners: Operators:

Initial Conditions: 100% Power, MOL with AFW 1-2 cleared Turnover:

Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 PMP_CVC3_2 OVERLOAD_DEV_FAIL TS, C CCP 1-3 OC Trip (OP AP-17, ECG 8.1.A).

(ALL) 2 AS01ASW_ASP11_MTFSEIZUR 1 TS, C ASW Pp 1-1 Seizes; Pp 1-2 can not be started (OP AP-10, TS (BOP, 3.0.3).

SRO) 3 MAL_CWS3A 80 C (ALL) High DP on Intake Screens; ramp required (AR PK13-01, OP MAL_CWS3B 75 AP-7, OP AP-25).

MAL_CWS1A 0.15 ramp=120 MAL_CWS1B 0.15 ramp=150 4 PMP_CWS1_2 M Both Circ Water pumps trip off during ramp, requiring OVERLOAD_DEV_FAIL cd='smss lt (ALL) Reactor Trip (OP AP-7).

1140' PMP_CWS2_2 OVERLOAD_DEV_FAIL cd='smss lt 1080' 5 MAL_EPS4C_2 DIFFERENTIAL C (ALL) Vital 4kV Bus F differential trip.

cd='fnispr lt 5' delay=30 6 VLV_PZR4_2 0.3 cd='jpplp4' del C (BOP) Pressurizer PORV PCV-455C fails slightly open on trip ay=60 requiring manual isolation by associated block valve 7 MAL_AFW1 1 cd='fnispr lt 5' C (ALL) Turbine driven AFW pump overspeed trip.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L181 NRC ES-D-1-03 r0.docx Page 1 of 3 Rev 0

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 1,2,3,4,5,6,7) 7

2. Malfunctions after EOP entry (1-2) (Events 5,6,7) 3

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

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

5. EOPs entered/requiring substantive actions (1-2) (E-0.1, FR-H.1) 2

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 (S3CT-1) Close the block MOV upstream of The open PORV and block valve constitute the

• Westinghouse Owners the stuck open PORV prior rupture of the degradation of a fission product barrier. Closing the Group WCAP-17711-NP PRT. block valve is essential to safety since failure to do so results in the unnecessary continuation of the degraded condition.

(S3CT-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.

L181 NRC ES-D-1-03 r0.docx Page 2 of 3 Rev 0

SCENARIO

SUMMARY

- NRC #3

1. Charging Pump CCP 1-3 trips on over current. The crew responds by entering OP AP-17, Loss of Charging to restore normal charging and letdown. Shift Foreman enters ECG 8.1.A - Charging Pump No. 3 Inoperable (establish continuous fire watch immediately; restore to operable status within 7 days).

2. ASW Pump 1-1 trips due to a seized shaft. Standby ASW Pump 1-2 fails to autostart and cannot be started manually. The Shift Foreman implements OP AP-10, Loss of Auxiliary Salt Water and cross-ties to the Unit 2 ASW system via the ASW cross-tie valve FCV-601. Shift Foreman enters T.S. 3.0.3 for two trains of ASW inoperable on Unit 1.

3. Screen differential pressure begins to rise quickly, bringing in AR PK13-01, Bar Racks/Screens. Following annunciator guidance, the crew enters OP AP-7, Degraded Condenser, Section C: Traveling Screen Problem and begins to reduce load to 50% or less per OP AP-25, Rapid Load Reduction.

4. Both Circ Water pumps trip off during ramp, requiring the crew to manually trip the Reactor oper OP AP-

7. The crew enters EOP E-0, Reactor Trip or Safety Injection and performs their immediate actions.

5. On the trip, vital 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.

6. Board operators also identify PCV-455C in mid-position. The valve will not close and must be isolated using the associated block valve 8000B (S3CT-1) Close the block MOV upstream of the stuck open PORV prior rupture of the PRT.

7. The TDAFW pump trips on overspeed. Steam Generator levels are initially high enough to provide an adequate secondary side heat sink and the crew transitions to EOP E-0.1, Reactor Trip Response to stabilize the plant. Steam Generator levels slowly lower below the minimum required level of 15%

narrow range and the crew transitions to EOP FR-H.1, Response to Loss of Secondary Heat Sink. With the condenser unavailable, Condensate is used to restore a secondary side heat sink (S3CT-2) Establish a secondary heat sink.

The scenario is terminated once Critical Task S3CT-2 is complete L181 NRC ES-D-1-03 r0.docx Page 3 of 3 Rev 0

Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 4 Op-Test No: L181 NRC Examiners: Operators:

Initial Conditions: 75% Power, MOL with D/G 1-2 OOS Turnover: At 75% power for SCCW HX Clearance Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 XMT_RMS3_3 1E+06 TS, I RE-3 spurious spike followed by high failure (AR delIA XMT_RMS3_3 2 delay=15 (BOP, PK11-21, ECG 39.3.D).

SRO)

XMT_RMS3_3 1e+06 cd='H_RM_C08B_1' delay=15 2 XMT_CVC19_3 0.0 ramp=120 I (ATC, LT-112 Fails Low (auto make-up) (OP AP-19, AP-5).

SRO) 3 MAL_RCS3B .07 TS, C 70 gpm RCS leak on Loop 2; increased to 90 gpm MAL_RCS3B .09 cd='h_v2_265r_1 or (ALL) on start of 2nd CCP (OP AP-1, TS 3.4.13.A).

h_v2_264r_1' delay=60 ramp=10 4 MAL_SEI1 0.21 ramp=10 C Seismic event causing Full Load Rejection (OP AP-MAL_GEN4_3 TRIP delay=10 cd='jmlsei1' (ALL) 2, AP-25).

LOA_SYD6 OPEN delay=15 cd='jmlsei1' LOA_SYD7 OPEN delay=15 cd='jmlsei1' LOA_SYD8 OPEN delay=15 cd='jmlsei1' 5 MAL_ROD13 TRUE cd='fnispr lt 55' M Auto rod control insertion failure requiring reactor (ALL) trip (OP AP-2).

6 MAL_SYD2 0 cd='fnispr lt 5' delay=2 M Loss of all A/C power.

(ALL)

MAL_EPS4E_2 DIFFERENTIAL cd='h_v4_217r_1' MAL_DEG1C_2 NO_RESET cd='H_V4_224R_1' 7 VLV_AFW7_1 1 C TDAFW Pump fails to autostart - manual start (BOP) required.

DelIA VLV_AFW7_1 2 cd='V3_219S_3'

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L181 NRC ES-D-1-04 r0.docx Page 1 of 3 Rev 0

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 1,2,3,4,5,6,7) 7

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

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

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

5. EOPs entered/requiring substantive actions (1-2) (ECA-0.0, ECA-0.2) 2

6. EOP contingencies requiring substantive actions (0-2) (ECA-0.0, ECA-0.2) 2

7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S4CT-1) Establish a minimum of 435 gpm After the initial RCS cooldown, decay heat

• WCAP-17711-NP, CT-23 AFW flow before S/G reach dryout increases the RCS temperature. Without an

• ECA-0.0 Background condition of 10% WR by starting the adequate secondary heat sink to support natural Document (HECA00BG),

turbine drive AFW pump (AFW Pp 1-1). circulation, the S/Gs could not support any Rev. 3.

significant plant cooldown. Thus the crew would lose the ability to delay the adverse consequences of core uncovery.

(S4CT-2) Energize at least one vital AC bus Failure to restore vital AC power from an available

• WCAP-17711-NP, CT-24 from prior to implementation of FLEX source when available represents an unnecessary

• ECA-0.0 Background strategies (ECA-0.0, step 10 RNO) continuation of a degraded electrical condition and Document (HECA00BG),

associated with entry into Extended Loss of unnecessarily complicates the mitigation strategy Rev. 3.

AC Power Event (ELAP) conditions (S4CT-3) Establish flow from at least one Failure to manually start at least one

• WCAP-17711-NP, CT-7 high head injection pump prior to transition high/intermediate head injection pump under the out of ECA-0.2. postulated conditions constitutes misoperation or incorrect crew performance in which the crew does not prevent degraded core cooling capacity.

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.

L181 NRC ES-D-1-04 r0.docx Page 2 of 3 Rev 0

SCENARIO

SUMMARY

- NRC #4

1. RE-3 (Oily Water Separator Effluent Line Radiation Monitor) spikes high momentarily, bringing in annunciator AR PK11-21, High Radiation. The crew identifies the signal as spurious and performs a reset of RM-3 per PK11-21, section 2.1.4. RE-3A fails high shortly after the reset and cannot be cleared. Shift Foreman enters ECG 39.3.D, Radioactive Liquid Effluent Monitoring Instrumentation (30 days).

2. Volume Control Tank (VCT) level channel LT-112 fails low, causing a continuous (and erroneous) makeup signal. The crew diagnoses the level channel failure by comparing other VCT parameters, and by using OP AP-19, Malfunction of the Reactor Makeup Control System. The makeup system is secured, and makeup is accomplished (if needed) using manual mode (or enabling the auto mode for short periods).

Crew may elect to use OP AP-5, Malfunction of Eagle 21 Protection or Control Channel to take manual control of Makeup Control System.

3. A 70 gpm RCS leak develops, requiring entry in OP AP-1, Excessive Reactor Coolant System Leakage.

Pressure. The leak increases to 90 gpm when a second charging pump is started and letdown must be isolated. VCT level can be maintained at the current leak rate, however, and the crew determines a plant shutdown is required. Shift Foreman enters TS 3.4.13.A, RCS Operational Leakage (4 hrs).

4. A significant seismic event results in a full load rejection on Unit 1. The crew recognizes the condition based on numerous power level alarms and the ensuing secondary side transient. The crew monitors primary and secondary system responses, most notably rod control, steam dumps, and digital feedwater, to ensure all systems respond appropriately in automatic. Shift Foreman implements OP AP-2, Full Load Rejection to stabilize the plant.

5. During the ramp down, automatic rod motion stops and the Shift Foreman directs a Reactor trip.

6. Startup power is lost on the trip followed by a bus differential fault on vital 4kV bus H. Diesel Generator 1-3 trips and cannot be reset. The crew transitions to EOP ECA-0.0, Loss of All Vital AC Power.

7. The turbine driven AFW pump fails to autostart requiring the crew to perform the critical task of manually starting the pump (S4CT-1) Establish a minimum of 435 gpm AFW flow before S/Gs reach dryout conditions.

8. Energy Trading informs the crew that 230kV start up power is not available, but 500 kV is expected back shortly. The crew performs actions to isolate RCP seal cooling.

9. Power is restored to vital 4kV buses F and G following the guidance of ECA-0.0, Appendix DD (Backfeed from 500kV Power) (S4CT-2) Energize at least one vital AC bus from prior to implementation of FLEX strategies.

10. The crew transitions to ECA-0.2, Loss of All AC Power with SI Required where they manually start safeguards equipment. The crew performs the final scenario critical task (S4CT-3) Establish flow from at least one high head injection pump.

The scenario is terminated once injection flow is established.

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Appendix D (rev 11) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 5 Op-Test No: L181 NRC Examiners: Operators:

Initial Conditions: 100% Power, MOL with CCP 1-3 cleared Turnover: At 100% power. CCP 1-3 cleared for routine maintenance. CCP 1-1 I/S.

Event Malf Event Type* Event Description No No. (See Summary for Narrative Detail) 1 CC01CCW_CCP11_MTFSHEAR 1 TS, C (BOP, CCW Pp 11 Shaft Shear (AR PK01-11; TS 3.7.7.A).

SRO) 2 MAL_CVC8A 100 ramp=60 C (ATC, SRO) Seal Injection Filter 1-1 plugs causing reduction in charging flow to RCP seals (AR PK04-22).

3 MAL_PPL7J 1 TS, I (BOP, Eagle 21 DFP-1 Halt in Rack 10 (OP AP-5; TS SRO) 3.3.1.E,M; 3.3.2.D, L; 3.4.11 ).

4 PK1421_0829 1 C (ALL) Loss of Main Transformer Cooling (AR PK14-21, AP-25) 5 MAL_SEI1 0.31 delay=0 ramp=15 M (ALL) Large seismic with no automatic or manual reactor MAL_PPL5A; PPL5B BOTH trip (ATWS).

6 MAL_RCS3C 10.0 cd='jmlsei1' M (ALL) SBLOCA following seismic; ramps in over 60 seconds.

delay=10 ramp=60 7 MAL_PPL1A FAILURE_TO_INIT C (BOP) Phase A - Train A and B fail to actuate requiring MAL_PPL1B FAILURE_TO_INIT manual alignment.

8 MAL_SYD2 0 cd='jpplsia' delay=15 C (ALL) Loss of Start-up power 15 seconds after Safety Injection 9 PMP_SIS2_2 OVERLOAD_DEV_FAIL C (ALL) Combination of electrical and mechanical failures cd='h_v4_218r_1' delay=3 result in no high or intermediate injection along with PMP_CVC2_1 AS_IS degraded secondary side heat removal capabilities.

BKR_EPS15 AS_IS BKR_EPS9_1 OVERCURRENT cd='H_V4_225R_1' MAL_AFW1 TRIP cd='h_v3_109m_1 gt 3000'

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L181 NRC ES-D-1-05 r0.docx Page 1 of 4 Rev 0

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 1,2,3,4,5,6,7,8,9) 9

2. Malfunctions after EOP entry (1-2) (Events 7,8,9) 3

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

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

5. EOPs entered/requiring substantive actions (1-2) (E-1, FR-C.1) 2

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

7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S5CT-1) Trip the Reactor by manually de- The safeguards systems that protect the plant

• Westinghouse Owners energizing 480V Buses 13D and 13E within during accidents are designed assuming that only Group WCAP-17711-NP 90 seconds of AR PK04-11, Reactor Trip decay heat and pump heat are being added to the

• Calc G.2 Rev 5 (08151-Initiate coming into alarm. RCS. Failure to manually trip the reactor causes a 2169) challenge to the subcriticality critical safety

• OP1.ID2, Time Critical function beyond that irreparably introduced by the Operator Actions Rev 12, postulated conditions. #34.

(S5CT-2) Manually close containment Failure to perform the critical task leads to an

• WCAP-17711-NP, CT-11 isolation valves such that at least one valve unnecessary release of fission products to the is closed on each Phase A containment auxiliary building, increasing the potential for penetration before transition out of EOP E- release to the environment and reducing

0. accessibility to vital equipment within the auxiliary building (S5CT-3) Depressurize Steam Generators to RCS depressurization allows accumulator injection

• WCAP-17711-NP, CT-39 inject SI Accumulators to re-flood the core and a temporary restoration of Core Cooling.

and clear the Extreme (red path) challenge Continued secondary depressurization allows for to the Core Cooling critical status function. low head injection and a stable source of long term Core Cooling. Failure to depressurize the Steam Generators results in the crew having to rely on the lower priority action of sequentially starting RCPs which constitutes a significant reduction of safety margin beyond that irreparably introduced as part of the scenario's design.

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.

L181 NRC ES-D-1-05 r0.docx Page 2 of 4 Rev 0

SCENARIO

SUMMARY

- NRC #5

1. AR PK01-11, CCW Pp 1-1 Recirc comes into alarm for FCV-606, CCW Pump 1-1 Recirc Valve, open. Crew identifies low pump amps on VB-1 and dispatches Nuclear Operator to investigate. Field reports no audible flow sound in spite of indications motor is running. CCW Pump 1-3 is started manually and CCW Pump 1-1 shutdown. TS 3.7.7.A, Vital Component Cooling Water (CCW) System, is entered for one loop of CCW inoperable (72 hrs).

2. 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.

3. Eagle 21 experiences a Digital Filter Processor (DFP) halt on rack 10. Associated indicators PI-456, LI-460A, FI-415, FI-425, FI-435, FI-445 (VB2), and PR-445, LR-459 (CC2) fail as-is as well as control channels for PORV 456 (PT-456) and Pressurizer Level Control (LT-460). Crew responds per OP AP-5, Malfunction of Eagle 21 Protection or Control Channel. Shift Foreman reviews Tech Specs, entering:

• TS 3.3.2.D, PC 456D Low Press SI (72 hrs).

• TS 3.3.1.E, PC 456A High Press Trip (72 hrs).

• TS 3.3.1.M, PC 456C Low Press Trip (72 hrs).

• TS 3.3.1.M, LC 460A High Level Trip (72 hrs).

• TS 3.3.1.M, FC-415(425,435,445) RCS Loop 1 (2,3,4) Flow (72 hrs).

• TS 3.3.2.L, PC-456 B, P-11 (1 hr).

• TS 3.4.11.B1, B2, & B3 PC-456E, to close & remove power from associated block valve (1 hr) and restore to operable (72 hrs).

4. Crew responds to AR PK14-21, MAIN TRANSF. A nuclear operator is dispatched to investigate local alarms and reports back that NO cooling fans or oil pumps are running on the Main Bank C Transformer.

Shift Foreman enters OP AP-25, Rapid Load Reduction or Shutdown and directs a 50 MW/min power reduction while Maintenance and field Operators attempt to restore transformer cooling.

5. A large earthquake (0.31 g) occurs during the ramp, but the reactor fails to trip automatically. The crew performs the immediate actions of EOP E-0, Reactor Trip or Safety Injection and successfully trips the reactor by opening the breakers for 480 V buses 13D and 13E to de-energize the control rod drive mechanism (CRDM) allowing control rods to fully drop into the core (S5CT-1) Trip the Reactor by manually de-energizing 480V Buses 13D and 13E.

6. A SBLOCA occurs as a result of the earthquake, but both trains of Phase A fail to actuate. The crew performs manual alignment of Phase A containment isolation valves per Appendix E, ESF Auto Actions, Secondary and Auxiliaries Status (S5CT-2) Manually close containment isolation valves such that at least one valve is closed on each Phase A containment penetration.

7. Startup power is lost shortly after Safety Injection initiates and a combination of electrical and mechanical failures result in the loss of both ECCS charging pumps and safety injection pumps.

Secondary heat removal is affected as well. The turbine driven AFW pump trips on overspeed and AFW pump 1-3 has no power due to a loss of 4kV bus F.

(continued)

L181 NRC ES-D-1-05 r0.docx Page 3 of 4 Rev 0

SCENARIO

SUMMARY

- NRC #5

8. The crew proceeds through E-0, transitioning to E-1, Loss of Reactor or Secondary Coolant. A loss of subcooling and lowering RVLIS level eventually results in a magenta path followed quickly by a red path on the core cooling critical safety function. The crew transitions to FR-C.2, Response to Degraded Core Cooling briefly, and then on to FR-C.1, Response to Inadequate Core Cooling.

Following the guidance of FR-C.1, the crew performs the final critical task (S5CT-3) Depressurize Steam Generators to inject SI Accumulators to re-flood the core and clear the Extreme (red path) .

The scenario is terminated once critical task S5CT-3 is complete.

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