ML030800318

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Initial Exam - 11/2002 - Final Outlines
ML030800318
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 07/12/2002
From: Gody A
Operations Branch IV
To: Rueger G
Pacific Gas & Electric Co
References
50-275/02-301, 50-323/02-301 50-275/02-301, 50-323/02-301
Download: ML030800318 (43)
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Text

ES-401 P WR SRO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant Form ES-401-3 Exam Date: 10/21/2002 Exam Level: SRO p-p K/A Category Points Tier Group Point Total KI K2 K3 K4 K5 I K6 I Al I G 1 4 4 4 4 4 4 24

1. 3 3 2 2 I* 3 16 Emergency

& 3 1 0 0 0 1 3 Abnormal Plant Tier Evolutions Totals 8 7 6 6 8 8 43 I 1 2 2 2 2 I 2 2 2 1 2 19 2.

2 1 2 2 2 1 1 1 2 2 1 2 17 Plant Systems 3 0 1 0 0 0 1 0 0 1 0 1 4 Tier Totals 2 5 4 4 3 3 3 4 5 2 5 40 Cat I Cat 2 Cat 3 Cat 4

3. Generic Knowledge And Abilities 5 4 4 4 17 Note: 1. Ensure that at least two topics from every K/A category are sampled within each teir (i.e., the "Tier Totals" in each
2. Actual point totals must match those specified in the table.
3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless
4. Systems/evolutions within each group are identified on the associated outline.
5. The shaded areas are not applicable to the category/tier.
6. The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be
7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. K/As below 2.5 should be 1

PWR SRO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 001 Continuous Rod Withdrawal / 1 AA1.04 Operating switch for emergency boration QI motor-operated valve 005 Inoperable/Stuck Control Rod / 1 AA1.05 RPI Q6 017 Reactor Coolant Pump (RCP) Malfunctions (Loss AAI.02 RCP oil reservoir level and alarm indicators Q19 of RC Flow) / 4 024 Emergency Boration / 1 2.2.30 Knowledge of RO duties in the control room Q104 during fuel handling such as alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

024 Emergency Boration / 1 AK3.02 Actions contained in EOP for emergency boration Q23 026 Loss of Component Cooling Water (CCW) / 8 AA1.02 Loads on the CCWS in the control room Q24 029 Anticipated Transient Without Scram (ATWS) / 1 EK2.06 Breakers, relays, and disconnects Q27 040 Steam Line Rupture / 4 AA2.03 Difference between steam line rupture and LOCA Q107 051 Loss of Condenser Vacuum / 4 AA2.01 Cause for low vacuum condition Q 108 051 Loss of Condenser Vacuum / 4 2.2.12 Knowledge of surveillance procedures. Q109 1

PWR SRO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-4U1-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 057 Loss of Vital AC Electrical Instrument Bus / 6 AA2.20 Interlocks in effect on loss of ac vital electrical Q111 instrument bus that must be bypassed to restore normal equipment operation 059 Accidental Liquid Radwaste Release / 9 AK1.05 The calculation of offsite doses due to a release Q41 from the power plant 059 Accidental Liquid Radwaste Release / 9 AK3.02 Implementation of E-plan Q42 062 Loss of Nuclear Service Water / 4 AK3.03 Guidance actions contained in EOP for Loss of Q47 nuclear service water 067 Plant Fire on Site / 9 AK1.01 Fire classifications, by type Q50 068 Control Room Evacuation / 8 AA2.01 S/G level Q113 068 Control Room Evacuation / 8 AK3.02 System response to turbine trip Q51 074 Inadequate Core Cooling / 4 EK2.09 Controllers and positioners Q57 076 High Reactor Coolant Activity / 9 AK2.01 Process radiation monitors Q58 E01 Rediagnosis / 3 EK2.1 Components, and functions of control and safety Q60 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 2

PWR SRO Examination Outline Printed: 06/10/2002

.Facility: Diablo Canyon Power Plant ES - 401 Emergen cy anu A normal Piant Evolutions - iler I / iGroup I r orm ES-qui-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment E02 SI Termination /3 EKI.2 Normal, abnormal and emergency operating Q61 procedures associated with SI Termination E07 Saturated Core Cooling / 4 2.4.18 Knowledge of the specific bases for EOPs. Q1 17 E08 Pressurized Thermal Shock / 4 EKI.2 Normal, abnormal and emergency operating Q64 procedures associated with Pressurized Thermal Shock E12 Uncontrolled Depressurization of all Steam 2.2.34 Knowledge of the process for determining the Q118 I Generators / 4 1 internal and external effects on core reactivity.

3

PWR SRO Examination Outline Printed: 06/10/2002

.Facility: Diablo Canyon Power Plant ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / Group 2 Form ES-4U1-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 008 Pressurizer (PZR) Vapor Space Accident (Relief AA2.05 PORV isolation (block valve switches and Q 102 Valve Stuck Open) / 3 indicators) 008 Pressurizer (PZR) Vapor Space Accident (Relief AA1.03 Turbine bypass in manual control to maintain Q8 Valve Stuck Open) /3 header pressure 009 Small Break LOCA /3 EK2.03 S/Gs Q10 027 Pressurizer Pressure Control (PZR PCS) AK1.01 Definition of saturation temperature Q25 Malfunction / 3 027 Pressurizer Pressure Control (PZR PCS) AK2.03 Controllers and positioners Q26 Malfunction / 3 038 Steam Generator Tube Rupture (SGTR) / 3 2.4.15 Knowledge of communications procedures Q105 associated with EOP implementation.

054 Loss of Main Feedwater (MFW) / 4 AK1.01 MFW line break depressurizes the S/G (similar to Q36 a steam line break) 054 Loss of Main Feedwater (MFW) / 4 AK3.03 Manual control of AFW flow control valves Q37 058 Loss of DC Power / 6 AK1.01 Battery charger equipment and instrumentation Q40 061 Area Radiation Monitoring (ARM) System Alarms / AA1.01 Automatic actuation Q45 7

065 Loss of Instrument Air / 8 AA2.07 Whether backup nitrogen supply is controlling Q1 12 1_ 1valve position I

PWR SRO Examination Outline Printed: 06/10/2002

.Facility: Diablo Canyon Power Plant ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / Group 2 Form ES-4U1-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment E03 LOCA Cooldown and Depressurization /4 EK3.2 Normal, abnormal and emergency operating Q62 procedures associated with LOCA Cooldown and Depressurization E05 Loss of Secondary Heat Sink /4 2.4.21 Knowledge of the parameters and logic used to Q63 assess the status of safety functions including: 1.

Reactivity control; 2. Core cooling and heat removal; 3. Reactor coolant system integrity; 4.

Containment conditions; 5. Radioactivity release control.

E05 Loss of Secondary Heat Sink /4 2.1.25 Ability to obtain and interpret station reference Q116 materials such as graphs, monographs, and tables which contain performance data.

Eli Loss of Emergency Coolant Recirculation / 4 EK2.2 Facility's heat removal systems, including primary Q65 coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Eli Loss of Emergency Coolant Recirculation /4 EA2.1 Facility conditions and selection of appropriate Q66 procedures during abnormal and emergency operations 2

PWR SRO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / Group 3 Form ES-401-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 056 Loss of Offsite Power / 6 AA2.42 Occurrence of a reactor trip Q1 10 E13 Steam Generator Overpressure / 4 2.4.35 Knowledge of local auxiliary operator tasks during Q119 emergency operations including system geography and system implications.

E15 Containment Flooding / 5 EK1.3 Annunciators and conditions indicating signals, Q68 I and remedial actions associated with the Containment Flooding 1

PWR SRO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 001 Control Rod Drive System / 1 K5.02 Definitions of differential rod worth and integral Q2 rod worth; their applications 004 Chemical and Volume Control K2.06 Control instrumentation Q4 System (CVCS) / 1 004 Chemical and Volume Control K3.08 RCP seal injection Q5 System (CVCS) / 1 013 Engineered Safety Features K2.01 ESFAS/safeguards equipment control Q 15 Actuation System (ESFAS) / 2 013 Engineered Safety Features A3.02 Operation of actuated equipment Q 16 Actuation System (ESFAS) / 2 014 Rod Position Indication System A4.04 Re-zeroing of rod position prior to startup Q17 (RPIS) / 1 017 In-Core Temperature Monitor K6.01 Sensors and detectors Q20 (ITM) System / 7 022 Containment Cooling System A1.02 Containment pressure Q21 (CCS) / 5 022 Containment Cooling System A2.05 Major leak in CCS Q22 (CCS) / 5 056 Condensate System / 4 A2.04 Loss of condensate pumps Q39 059 Main Feedwater (MFW) System / K3.04 RCS Q43 4

059 Main Feedwater (MFW) System / A1.03 Power level restrictions for operation of MFW Q44 4 1aumps and valves I

PWR SRO Examination Outline Printed: 06/10/2002

'Facility: Diablo Canyon Power Plant ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 061 Auxiliary / Emergency Feedwater A3.01 AFW startup and flows Q46 (AFW) System / 4 063 D.C. Electrical Distribution KI.03 Battery charger and battery Q49 System / 6 068 Liquid Radwaste System (LRS) / 9 2.2.24 Ability to analyze the affect of maintenance Q1 14 activities on LCO status.

068 Liquid Radwaste System (LRS) / 9 K4.01 Safety and environmental precautions for handling Q52 hot, acidic, and radioactive liquids 071 Waste Gas Disposal System 2.4.33 Knowledge of the process used track inoperable Ql 15 (WGDS) / 9 alarms.

072 Area Radiation Monitoring (ARM) K4.02 Fuel building isolation Q53 System / 7 072 Area Radiation Monitoring (ARM) K5.02 Radiation intensity changes with source distance Q54 System / 7 2

PWR SRO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Plant Systems - Tier 2 / Group 2 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 006 Emergency Core Cooling System K3.02 Fuel Q7 (ECCS) / 2 010 Pressurizer Pressure Control K5.01 Determination of condition of fluid in PZR, using Q11 System (PZR PCS) / 3 steam tables 011 Pressurizer Level Control System K6.04 Operation of PZR level controllers Q12 (PZR LCS) / 2 011 Pressurizer Level Control System A2.03 Loss of PZR level Q13 (PZR LCS) / 2 012 Reactor Protection System / 7 2.2.8 Knowledge of the process for determining if the Q103 proposed change, test, or experiment involves an unreviewed safety question.

012 Reactor Protection System / 7 K2.01 RPS channels, components, and interconnections Q14 016 Non-Nuclear Instrumentation A2.02 Loss of power supply Q18 System (NNIS) / 7 029 Containment Purge System (CPS) / A3.01 CPS isolation Q28 8

034 Fuel Handling Equipment System A4.02 Neutron levels Q30 (FHES) / 8 035 Steam Generator System (S/GS) / K4.06 S/G pressure Q31 4

035 Steam Generator System (S/GS) / A3.01 S/G water level control Q32 4

039 Main and Reheat Steam System 2.4.11 Knowledge of abnormal condition procedures. Q 106 (MRSS) / 4 _1 1 _

1

PWR SRO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Plant Svstems - Tier 2 / Groun 2 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 055 Condenser Air Removal System KI.06 PRM system Q38 (CARS) / 4 062 A.C. Electrical Distribution System K2.01 Major system loads Q48

/6 073 Process Radiation Monitoring A1.01 Radiation levels Q56 (PRM) System / 7 073 Process Radiation Monitoring K4.01 Release termination when radiation exceeds Q55 (PRM) System / 7 setpoint 103 Containment System / 5 K3.01 Loss of containment integrity under shutdown Q59 I_ conditions 2

PWR SRO Examination Outline Printed: 06/10/2002 "Facility: Diablo Canyon Power Plant ES - 401 Plant Systems - Tier 2 / Grout 3 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 007 Pressurizer Relief Tank/Quench 2.4.28 Knowledge of procedures relating to emergency Q101 Tank System (PRTS) / 5 response to sabotage.

008 Component Cooling Water System K2.02 CCW pump, including emergency backup Q9 (CCWS) / 8 041 Steam Dump System (SDS) and K6.03 Controller and positioners, including ICS, S/G, Q33 Turbine Bypass Control / 4 CRDS 041 Steam Dump System (SDS) and A3.02 RCS pressure, RCS temperature, and reactor Q34 I Turbine Bypass Control / 4 12ower II I

Generic Knowledge and Abilities Outline (Tier 3) Printed: 06/10/2002 PWR SRO Examination Outline Form ES-401-5 Facility: Diablo Canyon Power Plant Generic Category KA KA Topic Comment Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements. Q70 2.1.3 Knowledge of shift turnover practices. Q71 2.1.13 Knowledge of facility requirements for controlling vital / controlled Q 120 access.

2.1.32 Ability to explain and apply all system limits and precautions. Q69 2.1.33 Ability to recognize indications for system operating parameters Q121 which are entry-level conditions for technical specifications.

Category Total: 5 Equipment Control 2.2.4 (multi-unit) Ability to explain the variations in control board layouts, Q3 systems, instrumentation and procedural actions between units at a facility.

2.2.14 Knowledge of the process for making configuration changes. Q123 2.2.31 Knowledge of procedures and limitations involved in initial core Q122 loading.

2.2.32 Knowledge of the effects of alterations on core configuration. Q124 Category Total: 4 Radiation Control 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control Q35 requirements.

2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are Q126 outside the control room (e.g., waste disposal and handling systems).

2.3.4 Knowledge of radiation exposure limits and contamination control, Q29 including permissible levels in excess of those authorized.

2.3.10 Ability to perform procedures to reduce excessive levels of radiation Q125 and guard against personnel exposure.

Category Total: 4 1

Generic Knowledge and Abilities Outline (Tier 3) Printed: 06/10/2002 PWR SRO Examination Outline Form ES-401-5 Facility: Diablo Canyon Power Plant Generic Category KA KA Topic Comment Emergency Procedures/Plan 2.4.4 Ability to recognize abnormal indications for system operating Q127 parameters which are entry-level conditions for emergency and abnormal operating procedures.

2.4.27 Knowledge of fire in the plant procedure. Q129 2.4.32 Knowledge of operator response to loss of all annunciators. Q128 2.4.45 Ability to prioritize and interpret the significance of each Q67 annunciator or alarm.

Category Total: 4 Generic Total: 17 2

ES-401 PWR RO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant Form ES-401-4 Exam Date: 10/21/2002 Exam Level: RO I - P K/A Category Points Tier Group KI K2 K3 K4 IK K6 Al A2 A3 A4 G Point Total 1 3 3 3 3 2 2 16

1. ____

Emergency 2 4 4 3 3 2 1 17 Abnormal Plant 3 1 1 1 0 0 0 3 Evolutions Totals 8 8 7 6 4 3 36 Tier 1 3 2 2 2 2 2 2 2 2 2 2 23 2.

2 2 2 2 2 1 2 2 2 2 2 1 20 Plant Systems 3 1 1 1 1 0 1 0 0 1 2 0 8 Tier Totals 6 5 5 5 3 5 4 4 5 6 3 51 Cat I Cat 2 Cat 3 Cat 4

3. Generic Knowledge And Abilities 3 3 3 4 13 Note: 1. Ensure that at least two topics from every K/A category are sampled within each teir (i.e., the "Tier Totals" in each
2. Actual point totals must match those specified in the table.
3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless
4. Systems/evolutions within each group are identified on the associated outline.
5. The shaded areas are not applicable to the category /tier.
6. The generic K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be
7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. K/As below 2.5 should be 1

PWR RO Examination Outline Printed: 06/10/2002 jFacility: Diablo Canyon Power Plant ES - 401J1merge cy and Abnormal Plant Evolutions - Tier I / Group 1 Form E5-'ut-4 E/APE # E/APE Name / Safety Function KA KA Topic Comment 005 Inoperable/Stuck Control Rod / I AA1.05 RPI Q6 015 Reactor Coolant Pump (RCP) Malfunctions / 4 AA2.01 Cause of RCP failure Q80 017 Reactor Coolant Pump (RCP) Malfunctions (Loss AA1.02 RCP oil reservoir level and alarm indicators Q19 of RC Flow) / 4 024 Emergency Boration / 1 AK3.02 Actions contained in EOP for emergency boration Q23 024 Emergency Boration / 1 2.1.25 Ability to obtain and interpret station reference Q83 materials such as graphs, monographs, and tables which contain performance data.

026 Loss of Component Cooling Water (CCW) / 8 AA1.02 Loads on the CCWS in the control room Q24 027 Pressurizer Pressure Control (PZR PCS) AK1.01 Definition of saturation temperature Q25 Malfunction / 3 027 Pressurizer Pressure Control (PZR PCS) AK2.03 Controllers and positioners Q26 Malfunction / 3 055 Loss of Offsite and Onsite Power (Station Blackout) 2.2.27 Knowledge of the refueling process. Q85

/6 062 Loss of Nuclear Service Water / 4 AK3.03 Guidance actions contained in EOP for Loss of Q47 nuclear service water 067 Plant Fire on Site / 9 AK1.01 Fire classifications, by type Q50 I

PWR RO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant A * ... .. . . O .. . .. .

ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-4 E/APE # E/APE Name / Safety Function KA KA Topic Comment 068 Control Room Evacuation / 8 AK3.02 System response to turbine trip Q51 074 Inadequate Core Cooling / 4 EK2.09 Controllers and positioners Q57 076 High Reactor Coolant Activity /9 AK2.01 Process radiation monitors Q58 E07 Saturated Core Cooling / 4 EA2.2 Adherence to appropriate procedures and Q93 operation within the limitations in the facility's license and amendments E08 Pressurized Thermal Shock / 4 EKI.2 Normal, abnormal and emergency operating Q64 procedures associated with Pressurized Thermal Shock 2

PWR RO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / tGroup 2 Form hN-4IJl-4 E/APE # E/APE Name / Safety Function KA KA Topic Comment 001 Continuous Rod Withdrawal / I AA1.04 Operating switch for emergency boration Qi motor-operated valve 008 Pressurizer (PZR) Vapor Space Accident (Relief AA1.03 Turbine bypass in manual control to maintain Q8 Valve Stuck Open) /3 header pressure 009 Small Break LOCA /3 EK2.03 S/Gs Q10 029 Anticipated Transient Without Scram (ATWS) / 1 EK2.06 Breakers, relays, and disconnects Q27 054 Loss of Main Feedwater (MFW) / 4 AKI.01 MFW line break depressurizes the S/G (similar to Q36 a steam line break) 054 Loss of Main Feedwater (MFW) / 4 AK3.03 Manual control of AFW flow control valves Q37 058 Loss of DC Power / 6 AKI.01 Battery charger equipment and instrumentation Q40 059 Accidental Liquid Radwaste Release / 9 AKI.05 The calculation of offsite doses due to a release Q41 from the power plant 059 Accidental Liquid Radwaste Release / 9 AK3.02 Implementation of E-plan Q42 061 Area Radiation Monitoring (ARM) System Alarms / AA1.01 Automatic actuation Q45 17 1

PWR RO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / Group 2 Form S-4ui-,4 E/APE # E/APE Name / Safety Function KA KA Topic Comment E01 Rediagnosis / 3 EK2.1 Components, and functions of control and safety Q60 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E02 SI Termination / 3 EK1.2 Normal, abnormal and emergency operating Q61 procedures associated with SI Termination E03 LOCA Cooldown and Depressurization /4 EK3.2 Normal, abnormal and emergency operating Q62 procedures associated with LOCA Cooldown and Depressurization E04 LOCA Outside Containment / 3 EA2.2 Adherence to appropriate procedures and Q92 operation within the limitations in the facility's license and amendments E05 Loss of Secondary Heat Sink / 4 2.4.21 Knowledge of the parameters and logic used to Q63 assess the status of safety functions including: 1.

Reactivity control; 2. Core cooling and heat removal; 3. Reactor coolant system integrity; 4.

Containment conditions; 5. Radioactivity release control.

El1 Loss of Emergency Coolant Recirculation /4 EK2.2 Facility's heat removal systems, including primary Q65 coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Eli Loss of Emergency Coolant Recirculation / 4 EA2.1 Facility conditions and selection of appropriate Q66 procedures during abnormal and emergency operations 2

PWR RO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / Group 3 Form -'-4oi-4 E/APE # E/APE Name / Safety Function KA KA Topic Comment E13 Steam Generator Overpressure /4 EK3.4 RO or SRO function within the control room team Q75 as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated E15 Containment Flooding / 5 EKI.3 Annunciators and conditions indicating signals, Q68 and remedial actions associated with the Containment Flooding E 15 Containment Flooding / 5 EK2.1 Components, and functions of control and safety Q94 I systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 1

PWR RO Examination Outline Printed: 06/10/2002

  • Facility: Diablo Canyon Power Plant ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 001 Control Rod Drive System / I K5.02 Definitions of differential rod worth and integral Q2 rod worth; their applications 001 Control Rod Drive System / I A4.13 Stopping other changes in plant, e.g., turbine, S/G, Q72 SDBCS, boration, before adjusting rods 003 Reactor Coolant Pump System 2.4.47 Ability to diagnose and recognize trends in an Q77 (RCPS) / 4 accurate and timely manner utilizing the appropriate control room reference material.

003 Reactor Coolant Pump System K6.04 Containment isolation valves affecting RCP Q78 (RCPS) / 4 operation 004 Chemical and Volume Control K2.06 Control instrumentation Q4 System (CVCS) / 1 004 Chemical and Volume Control K3.08 RCP seal injection Q5 System (CVCS) / 1 013 Engineered Safety Features K2.01 ESFAS/safeguards equipment control Q15 Actuation System (ESFAS) / 2 013 Engineered Safety Features A3.02 Operation of actuated equipment Q 16 Actuation System (ESFAS) / 2 015 Nuclear Instrumentation System /7 A4.02 NIS indicators Q81 017 In-Core Temperature Monitor K6.01 Sensors and detectors Q20 (ITM) System / 7 017 In-Core Temperature Monitor KI.02 RCS Q82 (ITM) System / 7 022 Containment Cooling System A1.02 Containment pressure Q21 I(CCS) / 5 1 1 I

PWR RO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Plant Svstems - Tier 2 / Groun 1 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 022 Containment Cooling System A2.05 Major leak in CCS Q22 (CCS) / 5 056 Condensate System / 4 A2.04 Loss of condensate pumps Q39 056 Condensate System / 4 KI.03 MFW Q86 059 Main Feedwater (MFW) System / K3.04 RCS Q43 4

059 Main Feedwater (MFW) System / A1.03 Power level restrictions for operation of MFW Q44 4 pumps and valves 061 Auxiliary / Emergency Feedwater Ki.04 RCS Q87 (AFW) System / 4 061 Auxiliary / Emergency Feedwater A3.01 AFW startup and flows Q46 (AFW) System / 4 068 Liquid Radwaste System (LRS) / 9 K4.01 Safety and environmental precautions for handling Q52 hot, acidic, and radioactive liquids 068 Liquid Radwaste System (LRS) / 9 2.1.32 Ability to explain and apply all system limits and Q89 precautions.

072 Area Radiation Monitoring (ARM) K4.02 Fuel building isolation Q53 System / 7 072 Area Radiation Monitoring (ARM) K5.02 Radiation intensity changes with source distance Q54 System / 7 2

PWR RO Examination Outline Printed: 06/10/2002 Facility: Diablo Canyon Power Plant ES - 401 Plant Systems - Tier 2 I Groun 2 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 002 Reactor Coolant System (RCS) / 2 A1.11 Relative level indications in the RWST, the Q76 refueling cavity, the PZR and the reactor vessel during preparation for refueling 006 Emergency Core Cooling System K3.02 Fuel Q7 (ECCS) / 2 010 Pressurizer Pressure Control K5.01 Determination of condition of fluid in PZR, using Q11 System (PZR PCS) / 3 steam tables 011 Pressurizer Level Control System K6.04 Operation of PZR level controllers Q12 (PZR LCS) / 2 011 Pressurizer Level Control System A2.03 Loss of PZR level Q13 (PZR LCS) / 2 012 Reactor Protection System / 7 K2.01 RPS channels, components, and interconnections Q14 012 Reactor Protection System / 7 K3.02 T/G Q79 014 Rod Position Indication System A4.04 Re-zeroing of rod position prior to startup Q17 (RPIS) / 1 016 Non-Nuclear Instrumentation A2.02 Loss of power supply Q18 System (NNIS) / 7 029 Containment Purge System (CPS) / A3.01 CPS isolation Q28 8

029 Containment Purge System (CPS) / A4.01 Containment purge flow rate Q73 8

035 Steam Generator System (S/GS) / K4.06 S/G pressure Q31 4 1 I

PWR RO Examination Outline Printed: 06/10/2002

.Facility: Diablo Canyon Power Plant ES - 401 Plant Svstems - Tier 2 / Grouv 2 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 035 Steam Generator System (S/GS) / A3.01 S/G water level control Q32 4

055 Condenser Air Removal System Ki.06 PRM system Q38 (CARS) / 4 062 A.C. Electrical Distribution System K2.01 Major system loads Q48

/6 063 D.C. Electrical Distribution KI.03 Battery charger and battery Q49 System / 6 064 Emergency Diesel Generator 2.1.32 Ability to explain and apply all system limits and Q88 (ED/G) System / 6 precautions.

073 Process Radiation Monitoring A1.01 Radiation levels Q56 (PRM) System / 7 073 Process Radiation Monitoring K4.01 Release termination when radiation exceeds Q55 (PRM) System / 7 setpoint 086 Fire Protection System (FPS) / 8 K6.04 Fire, smoke, and heat detectors Q91 2

I -Facility: Di.ablo Canyon Power Plant PWR RO Examination Outline Printed: 06/10/2002 ES - 401 Plant Systems - Tier 2 / Group 3 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 008 Component Cooling Water System K2.02 CCW pump, including emergency backup Q9 (CCWS) / 8 034 Fuel Handling Equipment System A4.02 Neutron levels Q30 (FHES) / 8 041 Steam Dump System (SDS) and K6.03 Controller and positioners, including ICS, S/G, Q33 Turbine Bypass Control / 4 CRDS 041 Steam Dump System (SDS) and A3.02 RCS pressure, RCS temperature, and reactor Q34 Turbine Bypass Control / 4 power 045 Main Turbine Generator (MT/G) A4.02 T/G controls, including breakers Q74 System / 4 045 Main Turbine Generator (MT/G) K1.06 RCS, during steam valve test Q84 System / 4 076 Service Water System (SWS) / 4 K4.03 Automatic opening features associated with SWS Q90 isolation valves to CCW heat exchangers 103 Containment System / 5 K3.01 Loss of containment integrity under shutdown Q59 conditions I I

Generic Knowledge and Abilities Outline (Tier 3) Printed: 06/10/2002 PWR RO Examination Outline Form ES-401-5 Facility: Diablo Canyon Power Plant Generic Category KA KA Topic Comment Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements. Q70 2.1.3 Knowledge of shift turnover practices. Q71 2.1.32 Ability to explain and apply all system limits and precautions. Q69 Category Total: 3 Equipment Control 2.2.4 (multi-unit) Ability to explain the variations in control board layouts, Q3 systems, instrumentation and procedural actions between units at a facility.

2.2.13 Knowledge of tagging and clearance procedures. Q95 2.2.33 Knowledge of control rod programming. Q96 Category Total: 3 Radiation Control 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control Q35 requirements.

2.3.4 Knowledge of radiation exposure limits and contamination control, Q29 including permissible levels in excess of those authorized.

2.3.11 Ability to control radiation releases. Q97 Category Total: 3 Emergency Procedures/Plan 2.4.14 Knowledge of general guidelines for EOP flowchart use. Q99 2.4.29 Knowledge of the emergency plan. Q98 2.4.45 Ability to prioritize and interpret the significance of each Q67 annunciator or alarm.

2.4.48 Ability to interpret control room indications to verify the status and Q100 operation of system, and understand how operator actions and directives affect plant and system conditions.

Category Total: 4 1

Generic Knowledge and Abilities Outline (Tier 3) Printed: 06/10/2002 PWR RO Examination Outline Form ES-401-5 Facility: Diablo Canyon Power Plant Generic Category KA KA Topic Comment Generic Total: 13 2

ES-401 Record of Rejected K/As Form ES-401-10 Tier / Group Randomly Selected K/A Reason for Rejection RO 1/1 APE:055.2.1.29 (Q85) Generic K/A was inappropriate for particular system or procedure.

RO 1/2; SRO 1/2 APE:E05.2.2.24 (Q63) Generic K/A was inappropriate for particular system or procedure.

SRO 1/1 APE:051.2.4.33 (0109) Generic K/A was inappropriate for particular system or procedure.

SRO 1/1 APE:E12.2.2.29 (0118) Generic K/A was inappropriate for particular system or procedure.

SRO 1/2 APE:038:2.2.26 (Q105) Generic K/A was inappropriate for particular system or procedure.

SRO 1/2 APE:E05.2.2.20 (Q116) Generic K/A was inappropriate for particular system or procedure.

SRO 2/2 SYS:039.2.1.11 (Q106) Generic K/A was inappropriate for particular system or procedure.

RO 2/1 ; SRO 211 SYS: 061 .A3.04 (Q46) Auxiliary feedwater does not automatically isolate at this plant.

RO 1/1 ; SRO 1/1 APE: 062.AK3.04 (Q47) Nuclear Service Water does not exist at this plant. Even when Auxiliary Salt Water is substituted, the K/A is not applicable.

RO 2/2 ; SRO 2/2 SYS: 073.K4.02 (055) Letdown does not isolate due to a process radiation monitor high radiation at this plant.

SRO 1/1 APE: E07.2.2.25 (Q117) Saturated Core Cooling does not have knowledge of bases in T.S. for LCOs and safety limits at this plant.

NUREG-1021, Revision 8, Supplement 1 46 of 46

ES-301 Administrative Topics Outline Form ES-301-1 Facility: DCPP Date of Examination: 10/21/2002 Examination Level (circle one): RO / SRO Operating Test Number: 1 Administrative Describe method of evaluation:

Topic/Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A. 1 Mode ADMNRC - 01, Perform Sealed Valve Checklist (JPM) RO/SRO Requirements Plant ADMNRC - 12SRO, Verify AFD is within Tech Spec Limits (JPM)

Parameters ADMNRC - 2RO, Perform QPTR (JPM)

A.2 Temporary Mods ADMNRC - 3RO, Prepare Main Annunciator Problem Evaluation (JPM)

ADMNRC - 3SRO, Review Main Annunciator Problem Evaluation (JPM)

A.3 Radiation Control ADMNRC- 4, SCA Frisk (JPM) RO/SRO A.4 Emergency Plan Question RO: Responsibilities of Emergency Liaison Coordinator Question RO: Emergency Exposure Limits ADMNRC - 5SRO, Perform offsite Dose Assessment (JPM)

NUREG-1021, Revision 8 21 of 26

ES-301 Administrative Topics Outline Form ES-301-1 Facility: DCPP Date of Examination: 10/21/2002 Examination Level (circle one): RO / SRO Operating Test Number: 2 Administrative Describe method of evaluation:

Topic/Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Plant ADMNRC - 6RO, Calculate SDM (JPM)

Parameters ADMNRC - 6SRO, Verify SDM (JPM)

Fuel Handling ADMNRC - 7RO, Determine SFP Heat Load (JPM)

ADMNRC - 7SRO, Verify SFP Heat Load (JPM)

A.2 Tagging ADMNRC - 8RO, Perform Clearance Review (JPM)

Maintenance ADMNRC - 9SRO, Perform Risk Assessment (JPM)

A.3 Radiation Control ADMNRC - 10, High Radiation Area Entry (JPM) RO/SRO A.4 Emergency Plan Question RO: Notification Times Question RO: OSC Activation and Location ADMNRC - 11 SRO, Perform offsite Dose Assessment (JPM)

NUREG-1021, Revision 8 21 of 26

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 PART B EXAM, TEST I Facility: - DCPP Date of Examination: 10/28/2002 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: 1 B.1 Control Room Systems System / JPM Title Type Safety Code* Function TAB 1 004 - CVCS RO/SROI/SROU D,S,L I Makeup to RWST - NRCLJC - 9 TAB 2 074 - Inadequate Core Cooling RO/SROI D,A,S,L IVA Establish Feed from Condensate System - NRCLJC - 12 TAB 3 006 - ECCS RO/SROI D,A,S,L II Align RHR to Containment Spray- NRCLJC - 3 TAB 4 062 - AC Distribution RO/SROI D,S,L VI Crosstie Vital Bus G to H - NRCLJC - 4 TAB 5 068 - Control Room Evacuation RO/SROI/SROU D,S VIII Control Room Actions Prior to Evacuation - NRCLJC - 5 TAB 6 008 - CCW RO/SROI D,A,S VIII Respond to High Ultimate Heat Sink Temp - NRCLJC - 6 TAB 7 010- PZR Pressure Control RO/SROI/SROU N,A,S,L III Initiate Auxiliary Spray - NRCLJC - 14 B.2 Facility Walk-Through RO/SROI/SROU DA VI TAB 8 064 - Emergency Diesel Generators Local Start of a Diesel Generator - NRCLJP - 15 RO/SROI/SROU MRL IVB TAB 9 040 - Steam Line Rupture Locally Close an MSIV - NRCLJP - 16 TAB 10 061 - Auxiliary Feedwater RO/SROI D,R,L IVB Align Alternate AFW from Fire Water - NRCLJP - 21

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8 22 of 26

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 PART B EXAM, TEST 2 Facility: - DCPP Date of Examination: 10/28/2002 Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: 2 B.1 Control Room Systems System / JPM Title Type Safety Code* Function TAB 1 006 - ECCS RO/SROI M,A,S III Perform Actions for Trip with SI - NRCLJC - 8 TAB 2 004 - CVCS RO/SROI D,A,S,L Establish Emergency Boration - NRCLJC - I TAB 3 022 - Containment Cooling RO/SROI N,S V Place CFCU Drain Collection In Service - NRCLJC - 10 TAB 4 002 - RCS RO/SROI/SROU D,A,S,L IVA Initiate Bleed and Feed for Loss of Heat Sink - NRCLJC - 22 TAB 5 015- Nuclear Instrumentation RO/SROI/SROU D,S VII Remove PR Channel 42 From Service - NRCLJC - 23 TAB 6 074 - Inadequate Core Cooling RO/SROI/SROU N,A,S,L IVA Actions during FR-C.1 - NRCLJC - 13 TAB 7 064 - Emergency Diesel Generators RO/SROI D,S VI Manual Start DG 12 from Control Room - NRCLJC - 18 B.2 Facility Walk-Through TAB 8 068 - Control Room Evacuation RO/SROI/SROU D VIII Align 480V Buses from HSP - NRCLJP - 19 TAB 9 061 - Auxiliary Feedwater RO/SROI D,R,L IVB Reset TDAFWP - NRCLJP - 20 TAB 10 068 - Liquid Radwaste RO/SROI/SROU DR Isolate Ruptured LHUT - NRCLJP - 17

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8 22 of 26

Appendix D Scenario Outline Form ES-D-1 Facility: DCPP Units 1 & 2 Scenario No.: 1 Op-Test No.: 1 Examiners: Operators:

Objectives: Evaluate the crew's ability to swap condensate booster pump sets Evaluate the crews ability to diagnose and respond to a VCT level control channel failure Evaluate the crew's ability to diagnose and respond to a MFW pump controller problem Evaluate the crew's ability to diagnose and respond to a Turbine Control failure in Auto Evaluate the crew in using EOPs during an ATWS Evaluate the crew in using the EOPs during a loss of 230kV event Evaluate the crew's ability to diagnose and respond to a loss of TDAFW and MDAFW pumps Evaluate the crew in using EOPs during an FRH condition Initial Conditions: 100% power, equilibrium Xe, 1150 ppm, BOL (IC-1) MDAFW pump 1-2 00S last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for bearing inspection, back in service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. PRA OK.

Turnover: Start Standby Condensate Booster Pump set, place set 1-1 in standby.

Time Event Malf. Event Event min No. No. Type* Description 3 1 N,BOP Swap Condensate Booster Pump sets 10 2 N,R, ALL Commence power reduction to 70% (NO report Htr 2 DP oil leak) 20 3 mal tur4, 3 C, BOP Turbine control failure requiring manual ramp 30 4 Xmt cvc19 I, RO VCT Level channel 112 fail high 40 5 Ovr cc3049e C, ALL MFW Pump master controller failure requiring manual control Ovr cc3049h On 6 mal ppl5 I,ALL ATWS MFWP tip Cond on 7 mal syd2 C, ALL Loss of 230kV 13D/E open Cond on 8 prop afw2 M,ALL Loss of All Feedwater (MDAFW and TDAFW Pump failure) 13D/E rmal afwl open

Appendix D Scenario Outline Form ES-D-i1

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

SCENARIO 01 TEST 01 OVERVIEW The Crew will swap condensate booster pump sets, referencing OP C7A:I The Turbine Building NO will report an oil leak on Heater 2 Drip Pump, requiring a power reduction to 70% in preparation for tripping the pump. OP L-4 will be used for the power reduction, providing guidance on boration and setup of the turbine controls. A boration will commence and a controlled power reduction follows.

The Turbine controls will then shift to manual following a fault in the auto circuitry. This produces no alarms, but indications on the turbine control panel will indicate the change as well as the changes in plant parameters when the power reduction stops with boron injection underway. The crew will have to choose between stopping the ramp, or ramping manually to prepare for tripping the Heater 2 Drip Pump.

VCT Level channel 112 fails, giving a high VCT level alarm and diverting letdown to the hold up tanks. The crew should recognize the channel failure and respond per AP-19. Letdown should be restored to the VCT. The ramp may be stopped, but should be recommenced after the crew determines the failure does not impact the ramp.

The Master Feedwater Pump controller fails, requiring the crew to take manual control of both Main feedwater Pumps. The operator may not be able to analyze the problem and take corrective actions quick enough, which will then result in a Reactor Trip signal from low SG levels. If the operator does react and take control of the pumps manually, the crew will be forced to make a decision on continuing a manual ramp with manual feedwater, or trip the unit.

The unit will not trip on an auto trip signal or a manual trip initiation. The crew will be forced to use the RNO of E-0 and open breakers 13D and 13E. This will cause the rods to fall into the core. The crew will continue with E-0 actions.

Upon opening 13DIE, a loss of 230kV will occur. Plant response will lead to a Safety Injection during the implementation of E-0.

Upon opening 13DIE, the TDAFW pump and remaining MDAFW pump will trip and not restart.

The crew should recognize a RED path on Heat Sink, and following transition from E-0 to E-1, enter FR-H.1. With the loss of 230kV, Condensate Booster pumps and MFW pumps are not available, leaving only Bleed and Feed as the method to cool the core. The scenario will end when Bleed and Feed is established.

Appendix D Scenario Outline Form ES-D-1 Facility: DCPP Units 1 &2 Scenario No.: 1 Op-Test No.: 2 Examiners: Operators:

Objectives: Evaluate the crew's ability to increase Accumulator Pressure Evaluate the crews ability to reduce power Evaluate the crew's ability to diagnose and respond to failure in RMUW system Evaluate the crew's ability to diagnose and respond to a PT 505 failure Evaluate the crews ability to diagnose and respond to a failed PZR spray valve controller Evaluate the crew in using EOPs during a Steam Space LOCA Evaluate the crew's ability to diagnose and respond to of failure of the SI signal Initial Conditions: 100% power, equilibrium Xe, BOL 1150 ppm, (IC-1). MDAFW pump 1-2 OOS last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for bearing inspection, back inservice in8 hours. PRA good.

Turnover: Increase Accumulator 1-1 pressure per OP B-3B:I.

Time Event Malf. Event Event min No. No. Type* Description 3 1 N,RO Increase Accumulator Pressure 10 2 R,All Commence Power Decrease (EPOS request fast ramp to 850 MW)

On 3 Ovr C,RO 43/MU fail to auto borate, manual boration required Boration cc2010c 20 4 xmt TUR2 I,BOP PT 505 failure low 30 5 cnh pzr3 I,ALL PRZ spray valve controller fails open in Auto 40 6 Mal pzrl M,ALL PZR steam space LOCA On SI 7 ppl3a I,ALL Failure of SI to actuate (manual alignment necessary) ppl3b

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

Scenario 01 Test 02 Outline Following a tailboard, the crew will increase pressure inAccumulator 1-1 to normal using OP B-3B:I.

After the Accumulator pressure increase, a call from EPOS will request a fast ramp to < 850 MW. The crew will tailboard the ramp and reactivity needs. A boration will start and a ramp commenced.

The boration will fail, the Makeup deviation alarm will alarm. 43/MU will not work in borate mode and must be used inthe manual mode. The crew will use PK5-11 and AP-19 to determine the problem and use the alternate method to continue the ramp as requested.

After the crew commences manual boration and the ramp is started again, PT-505 will fail low, causing rods to drive in.The RO must recognize an instrument failure and take the rods to manual. Discussion on tripping bistables in 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per ITS 3.3.1-1 should take place.

The PZR spray controller will fail in auto mode next, requiring the RO to take manual control of the spray valves to control pressure. The SFM will use PK5-17 and AP-13 to guide the crews response.

A PZR steam space LOCA takes place over 10 minutes to a final value of 850 gpm. This will require the crew to diagnose the pressure reduction with minimal PZR level change, and to quantify the leak.

After the leak size is sufficient, an SI will be required. The crew should SI before the low pressure setpoint, however an Over Power reactor trip may cause a reactor trip before the crew can respond. The SI signal will fail, requiring a manual SI signal initiation and manually aligning the valves and pumps for injection.

The scenario will terminate after transition to E-1.2 is completed.

Appendix D Scenario Outline Form ES-D-1 Facility: DCPP Units I & 2 Scenario No.: 2 Op-Test No.: 1 Examiners: Operators:

Objectives: Evaluate the crew's ability to swap CCW heat exchangers Evaluate the crews ability to decrease reactor power Evaluate the crews ability to diagnose and respond to a Tc instrument drift Evaluate the crew's ability to diagnose and respond to a loss of non-vital 120 VAC Evaluate the crew's ability to diagnose and respond to an LDTV failure Evaluate the crew in using EOPs during a Seismic event Evaluate the crew's ability to diagnose and respond to a failure of Train A ECCS Equipment Evaluate the crew in using EOPs during a Main Feedline Break Evaluate the crew in using EOPs during a LBLOCA Initial Conditions: 100% power, equilibrium Xe, 1150 ppm BOL (IC-1) DEG 1-1 00S for fuel pump replacement. OOS 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, expected return in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. PRA OK.

Turnover: Swap CCW Heat Exchangers.

Time Event Malf. Event Event min No. No. Type* Description 3 1 N,RO Swap CCW heat exchangers 10 2 N,R,ALL Power decrease to 80% (EPOS: Fire at Midway) 20 3 Xmt rcsl38 I,RO RCS Tc (TE-441) fail high 30 4 Mal eps2a C,RO Loss of non-vital 120 VAC (PY-15) 40 5 Xmt tur22 C,ALL Turbine Governor Valve failure (FCV-142) 50 6 Mal sell Seismic event Cond on 7 Mal mfw5d M,ALL SG 4 MFL Break IC seismic 0 8 Mal ppl3b (3) C,ALL Failure of Tr A ECCS Seismic 9 Mal rcsl (1) M,ALL RCS Loop 1 25% DBA

+ 1min

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

Appendix D Scenario Outline Form ES-D-1 The crew will tailboard swapping the CCW heat exchanger for run time. This will entail swapping the running ASW train, and aligning the CCW heat exchanger per OP E-5:IV. The crew will start ASW pump 1-2, make alignments, and secure ASW pump 1-1 and make associated valve lineups.

EPOS will call requesting a decrease to 900 MW due to a fire at the Midway station. The crew will tailboard the ramp and reactivity change. A crew will borate and start a ramp per OP L-4.

During the ramp, Loop 4 Tc (TE-441) will fail high, causing rods to step in on a false high Tave.

The crew should recognize the failed instrument and place rods in manual. The SFM stop the ramp, and reference OP AP-5 to ensure the plant is stable and for Tech Spec requirements on tripping bistables in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and deselecting that channel from Tave recording and control. Rods should be placed back to auto.

Before bistables are tripped, PY-15, Non-Vital 120 VAC will fail, causing many unrelated alarms.

The crew should let rods control Tave to Tref because MSRs have been lost. The SFM will reference AP-4, and should request PY-15 be placed on backup power, which will restore the bus and clear the most of the alarms associated with the failure. The ramp should be reinstated if stopped.

Following the restoration of PY-15, the Turbine Governor valve, FCV-142, will fail causing a load rejection. The SFM will enter AP-25 and stabilize the plant. The Asset Team will be contacted for repair.

A Seismic event will take place, causing a Main Feed Line Break on SG 1-4 inside containment and a LBLOCA on Loop 1. The MSL Break will mask the LOCA initially. Train A SI will also fail to initiate and will require manual alignment of valves and pumps. The crew will isolate SG 1-4 using E-2, identify the LOCA and transition to E-1 where they will meet conditions to trip the RCPs. The scenario will continue until transition to E-1.3.

Appendix D Scenario Outline Form ES-D-1 Facility: DCPP Units I &2 Scenario No.: 2 Op-Test No.: 2 Examiners: Operators:

Objectives: Evaluate the crew's ability to increase reactor power Evaluate the crew's ability to diagnose and respond to a PZR level channel failure low Evaluate the crew's ability to restore letdown Evaluate the crew's ability to respond to a SGTL Evaluate the crew's ability to diagnose and respond to an SG pressure channel failure Evaluate the crews ability to diagnose and respond to a vacuum leak Evaluate the crews ability to diagnose and respond to a unit trip Evaluate the crew in using EOPs during an Faulted/Ruptured SG Evaluate the crew's ability to diagnose and respond to a failure of Phase A Initial Conditions: 30% power, 235 ppm EOL. (IC-42) MDAFW pump 1-2 00S last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for bearing inspection, back in service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. PRA good.

Turnover: Increase power per OP L-4 to 100%.

Time min Event Malf. Event Event No. No. Type* Description 3 1 N,R,ALL Increase power to 50%

10 2 xmt pzr4O I,RO PZR level channel failure low 15 3 N,ALL Restore Letdown 25 4 xmt mss58 C,BOP SG 1 pressure channel PT- 516 fail hi (manually close PCV - 19) 35 5 Mal rcs4a C,ALL SGTL on SG 1-1 (approx. 5 gpm) 45 6 lba cndl C,ALL Vacuum leak / power reduction 50 7 Mal seil Seismic Event (below Rx Trip Selpoint)

Cond on 8 Mal genl C,ALL Main Generator lockout / unit trip Seismic Cond on 9 Mal mss6a M,ALL SG 1-1 MSL fault Seismic Manually 10 Mal rcs4a M,ALL SGTR 1-1 (increase SGTL to 1215 gpm over 5 minutes)

Seismic +

5 min 0 11 Mal ppllb I, RO Failure of Train BPhase A

  • (N)ormal (R)eaclivity (I)nstrument (C)omponent (M)ajor

Scenario Outline The scenario starts at 30% during a startup. The crew will tailboard and commence a ramp to 50% per OP L-4 and dilute as necessary.

During the ramp, PZR Level Channel LT-459 will fail low, giving PZR level and Charging mismatch alarms. The RO will take manual control or charging and maintain seal injection and PZR level in band. The SFM will enter AP-5 and direct LT-459 be removed from input to control and determining per ITS 3.3.1 that bistables must be tripped in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Letdown will then be reestablished per OP B-1A:XII, allowing normal charging and letdown functions in automatic to resume.

SG 1-1 Pressure Channel PT-516 will fail high, causing the atmospheric, PCV-19 to open. There will be no alarms, and only the sound of steam and the indication of a PCV open light will indicate the problem. The BOPCO will have to take manual control of PCV-19 and close the valve. The SFM will respond per AP-5 and ITS 3.3.2 and determine bistables must be tripped in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

A small SGTL will develop on SG 1-1 of approximately 5 gpm. The SJAE Rad alarm (PK1 1-06) will alarm. The BOPCO will also notice RM-15 counts increasing on the chart recorded on VB-I.

The SFM will enter AP-3 and direct the RO to determine the leak rate. He will also determine ITS 3.4.13.d limits of 150 gpd has been exceeded and must start planning for a shutdown.

As the shutdown is planned, a small vacuum leak is initiated. Condensate D02 and conductivity alarms (PK12-04/05) will alarm. The crew will notice vacuum slowly decreasing. The SFM will direct the RO to start a load decrease while entering AP-7. The BOPCO will be directing leak diagnostics outside the control room.

A seismic event will cause the turbine to trip on a lockout, but because the reactor is below P-9, the reactor will stay on line. The SFM must determine that this condition is acceptable and direct the crew to verify normal plant response.

A MSL Break occurs (SG 1-1 safety fails open) following the seismic event, causing a cooldown and SI to occur. The SFM will enter E-0 and E-2 and direct the BOPCO to isolate SG 1-1. The BOPCO will also determine that Phase A train B did NOT occur, and utilizing Attachmnent E, align Phase A manually.

Shortly after the MSL break, a SGTR will develop on SG 1-1. The level increase will be masked from the cooldown and rapid level increases from all AFW pumps running. No rad alarms will occur since these are power dependant on N-16. Once RCS pressure is determined to be too low and SG level response is diagnosed as a SGTR, the SFM will transition to E-3, and direct response from there. He will then transition to ECA-1.3.

The scenario terminates at the transition to ECA-1.3

Appendix D Scenario Outline Form ES-D-1 Facility: DCPP Units 1 & 2 Scenario No.: 3 Op-Test No.: 2 Examiners: Operators:

Objectives: Evaluate the crew's ability to diagnose and respond to a loss of Data A on DRPI Evaluate the crew's ability to diagnose and respond to a Load Transient Bypass Valve failure Evaluate the crew's ability to diagnose and respond to a RCP seal failure Evaluate the crew's ability to shutdown the unit Evaluate the crew's ability to diagnose and respond to a Loss of RWST Evaluate the crew in using EOPs during a SBLOCA Evaluate the crew's ability to diagnose and respond to a loss of Charging Pumps Evaluate the crew in using EOPs during a loss of emergency coolant recirculation Initial Conditions: 100% power, equilibrium xenon, EOL (IC-35). DEG 1-1 O0S for fuel pump replacement. OOS 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, expected return in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. CSP 1-1 OS 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> for scheduled motor work, expected return 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. PRA OK.

Turnover: Maintain Power.

Time Event Malf. Event Event min No. No. Type* Description 3 1 Mal rod8a I,RO Loss of Data A on DRPI 10 2 Mal cndl C,ALL LTB (FCV-230) fail open Cond 3 R,ALL Stabilize Power LTBV 20 4 Mal rcp2a C,RO RCP Seal 2 failure 25 5 R,N, ALL Controlled Shutdown 30 6 Mal sei Seismic event Cond on 7 Mal rcp2a C,RO RCP Seal I failure sei Cond on 8 Loa sis1 C,ALL Loss of RWST sei Cond on 9 prmp cvcl C,ALL Loss of CCP 1 and 2 sei pmp cvc2 Cond on 10 Mal rcs3 M,ALL SBLOCA sei

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

Scenario Outline A Data A failure on DRPI will occur, alarming PK03-21. The SFM will direct DRPI be selected to B train.

The LTB valve, FCV-230, will fail open increasing reactor power above 100% and alarming PK10-07. The crew will have to shed load to maintain power below 100%. Rods will step and boration will be required. OPdT runback may occur. The SFM will enter AP-25 and direct the control room in stabilizing the plant.

RCP 1-1 #2 Seal will fail, causing seal leakoff to #1 to decrease and #2 to increase. PK05-01 will alarm and the SFM will direct the RO/BOPCO to start investigating, including Aux Board RCDT trends while monitoring temperature trends and RCP vibration. The crew should prepare for an orderly shutdown.

A seismic event will cause an RCP I seal 1 leak at 10 gpm requiring a pump trip and closure of the seal leakoff valve, a loss of both CCPs, a SBLOCA of 3000 gpm, and a Loss of RWST. No water will be available for injection. The crew will proceed through E-0, E-1 and transition to ECA-1.1 when Cold Leg Recirculation capability cannot be confirmed. The crew will be challenged to NOT trip the RCPs with no SI pumps available and no subcooling. They will proceed until cooldown is established with dumping steam and a 100°F/hr cooldown rate is established.

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