ML030230761

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Draft - Outlines
ML030230761
Person / Time
Site: Ginna Constellation icon.png
Issue date: 02/11/2002
From: Watts R
Rochester Gas & Electric Corp
To: Conte R
NRC/RGN-I/DRS/OSB
Conte R
References
50-244/01-301, ES-401, ES-401-3 50-244/01-301
Download: ML030230761 (55)
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Text

ES-401 PWR SRO Examination Outline Printed: 10/05/2001 Facility: R.E. Ginna Nuclear Power Plant Form ES-401-3 Exam Date: 02/08/2002 Exam Level: SRO I - r -

K/A Category Points Tier Group Point Total K1 K2 K3 I VAt I T<J Al A2 A3 A4 G 1 4 4 4 4 4 4 24

1. 2 2 3 3 3 2 3 16 Emergency

& 3 0 1 0 0 1 3 Abnormal Plant Tier Evolutions Totals 6 8 7 7 7 8 43 1 1 2 2 2 2 1 2 2 2 11 2 19 2.

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

3. Generic Knowledge And Abilities 4 5 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

Facility:

i 1R.*. Ginna Nuclear Power Plant PWR SRO Examin( i Outline Printed: 10/05/200(

1"* .... lr*L-q Ad'kl *'J ES - 401 Emerge

.y and Abnormal Piant Evolutions - Iier 1 I Group I Form L--quI-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 001 Continuous Rod Withdrawal / 1 AK3.02 Tech-Spec limits on rod operability 003 Dropped Control Rod / 1 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

003 Dropped Control Rod / 1 AK2.05 Control rod drive power supplies and logic circuits 005 Inoperable/Stuck Control Rod / 1 AA2.01 Stuck or inoperable rod from in-core and ex-core NIS, in-core or loop temperature measurements 011 Large Break LOCA / 3 EK2.02 Pumps 015 Reactor Coolant Pump (RCP) Malfunctions / 4 AK2.07 RCP seals 029 Anticipated Transient Without Scram (ATWS) / 1 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

040 Steam Line Rupture / 4 AA2.05 When ESFAS systems may be secured 040 Steam Line Rupture / 4 AA 1.01 Manual and automatic ESFAS initiation 057 Loss of Vital AC Electrical Instrument Bus / 6 AA2.03 RPS panel alarm annunciators and trip indicators 057 Loss of Vital AC Electrical Instrument Bus / 6 AA1.01 Manual inverter swapping I

Facility: (

R.*. Ginna Nuclear Power Plant PWR SRO Examin4 n Outline Printed: 10/05/200(

l .. rL'* dfl I ES - 401 E cy and Abnormal Plant EvOlUtiOnS - Iier I / Group 1 rorm ES-'.ui-.3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 068 Control Room Evacuation /8 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

068 Control Room Evacuation / 8 AK3.02 System response to turbine trip 069 Loss of Containment Integrity / 5 AK3.01 Guidance contained in EOP for loss of containment integrity E01 Rediagnosis / 3 2.4.6 Knowledge symptom based EOP mitigation strategies.

E01 Rediagnosis / 3 EK1.1 Components, capacity, and function of emergency systems E02 SI Termination /3 EK3.1 Facility operating characteristics during transient Similar to E02 EA1.2 conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics E02 SI Termination / 3 EA1.2 Operating behavior characteristics of the facility Similar to E02 EK3.1 E04 LOCA Outside Containment /3 EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments E07 Saturated Core Cooling / 4 EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the Saturated Core Cooling E08 Pressurized Thermal Shock / 4 EA 1.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 2

PWR SRO Examii. n Outline Printed: 10/05/240(

Facility: R.;. Ginna Nuclear Power Plant ES - 401 Emergency and Abnormal P'lant Evolutions - Tier I / Group I rForm S-4,0u-.

E/APE # E/APE Name / Safety Function KA KA Topic Comment E09 Natural Circulation Operations / 4 EK1.1 Components, capacity, and function of emergency Similar to E09 EK2.1 systems E09 Natural Circulation Operations / 4 EK2.1 Components, and functions of control and safety Similar to E09 EK1. 1 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E14 High Containment Pressure / 5 EK1.2 Normal, abnormal and emergency operating 1procedures associated with High Containment Pressure 3

PWR SRO Exami4 n Outline Printed: 10/05/20d0 Facility: R.*. Ginna Nuclear Power Plant ES - 401 cy'merge and Abnormal Piant Evolutions - I ier I / Group 2 _ _ _-__ rForm E/APE # E/APE Name / Safety Function KA KA Topic Comment 007 Reactor Trip / 1 EK2.02 Breakers, relays and disconnects 008 Pressurizer (PZR) Vapor Space Accident (Relief 2.1.30 Ability to locate and operate components, Valve Stuck Open) / 3 including local controls.

009 Small Break LOCA /3 EK1.01 Natural circulation and cooling, including reflux boiling 009 Small Break LOCA / 3 EK2.03 S/Gs 027 Pressurizer Pressure Control (PZR PCS) AA2.18 Operable control channel Malfunction / 3 027 Pressurizer Pressure Control (PZR PCS) AK2.03 Controllers and positioners Malfunction / 3 033 Loss of Intermediate Range Nuclear Instrumentation AAl.03 Manual restoration of power

/7 037 Steam Generator (S/G) Tube Leak / 3 2.2.22 Knowledge of limiting conditions for operations and safety limits.

037 Steam Generator (S/G) Tube Leak / 3 AK3.08 Criteria for securing RCP 038 Steam Generator Tube Rupture (SGTR) / 3 EKI.04 Reflux boiling 060 Accidental Gaseous Radwaste Release / 9 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

I

Facility: (R.r_.. Ginna Nuclear Power Plant PWR SRO Examir( n Outline Printed: 10/05/20d(

ES - 401 Emerge cy and Abnormal Plant Evolutions - Tier 1 / GIroup 2 Form ES-4u1-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 065 Loss of Instrument Air / 8 AAl.02 Components served by instrument air to minimize drain on system E03 LOCA Cooldown and Depressurization / 4 EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments Eli Loss of Emergency Coolant Recirculation /4 EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations E16 High Containment Radiation / 9 EK3.2 Normal, abnormal and emergency operating procedures associated with High Containment Radiation E 16 High Containment Radiation / 9 EA1.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 2

PWR SRO Examin( i Outline Printed: 10/05/200i Facility: R._. Ginna Nuclear Power Plant ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 3 Form ES-401-3 E/APE # E/APE Name / Safety Function KA KA Topic Comment 028 Pressurizer (PZR) Level Control Malfunction / 2 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

028 Pressurizer (PZR) Level Control Malfunction / 2 AK2.03 Controllers and positioners 036 Fuel Handling Incidents / 8 AA2.01 ARM system indications I

PWR SRO( imination Outline Printed: 10/05/t N

Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Svstems - Tier 2 / Groun 1 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 001 Control Rod Drive System / I K1.04 RCS 003 Reactor Coolant Pump System K3.04 RPS (RCPS) / 4 003 Reactor Coolant Pump System A1.05 RCS flow (RCPS) / 4 004 Chemical and Volume Control KS. 14 Reduction process of gas concentration in RCS:

System (CVCS) / 1 vent-accumulated non-condensable gases from PZR bubble space, depressurized during cooldown or by alternately heating and cooling (spray) within allowed pressure band (drive more gas out of solution) 004 Chemical and Volume Control A1.05 S/G pressure and level System (CVCS) / 1 013 Engineered Safety Features K2.01 ESFAS/safeguards equipment control Actuation System (ESFAS) / 2 013 Engineered Safety Features K6.01 Sensors and detectors Actuation System (ESFAS) / 2 014 Rod Position Indication System K3.02 Plant computer (RPIS) / 1 017 In-Core Temperature Monitor A2.01 Thermocouple open and short circuits (ITM) System / 7 022 Containment Cooling System 2.2.22 Knowledge of limiting conditions for operations (CCS) / 5 and safety limits.

022 Containment Cooling System A4.04 Valves in the CCS (CCS) / 5 I

PWR SRT imination Outline Printed: 10/05/(

Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Systems - Tier 2 / Grout 1 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 026 Containment Spray System (CSS) / 2.4.30 Knowledge of which events related to system 5 operations/status should be reported to outside agencies.

026 Containment Spray System (CSS) / K4.05 Prevention of material from clogging nozzles 5 during recirculation 059 Main Feedwater (MFW) System / K4.16 Automatic trips for MFW pumps 4

059 Main Feedwater (MFW) System / A2.05 Rupture in MFW suction or discharge line 4

063 D.C. Electrical Distribution K2.01 Major DC loads System / 6 068 Liquid Radwaste System (LRS) / 9 A3.02 Automatic isolation 072 Area Radiation Monitoring (ARM) K5.02 Radiation intensity changes with source distance System / 7 072 Area Radiation Monitoring (ARM) A3.01 Changes in ventilation alignment System /7 2

PWR SR imination Outline Printed: 10/05/(

Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Systems - Tier 2 / Group 2 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 002 Reactor Coolant System (RCS) / 2 K5.09 Relationship of pressure and temperature for water at saturation and subcooling conditions 002 Reactor Coolant System (RCS) / 2 K6.03 Reactor vessel level indication 010 Pressurizer Pressure Control 2.1.14 Knowledge of system status criteria which require System (PZR PCS) / 3 the notification of plant personnel.

010 Pressurizer Pressure Control K2.01 PZR heaters System (PZR PCS) / 3 012 Reactor Protection System / 7 A2.05 Faulty or erratic operation of detectors and function generators 029 Containment Purge System (CPS) / A 1.02 Radiation levels 8

033 Spent Fuel Pool Cooling System A1.02 Radiation monitoring systems (SFPCS) / 8 035 Steam Generator System (S/GS) / K5.01 Effect of secondary parameters, pressure, and 4 temperature on reactivity 039 Main and Reheat Steam System K1.04 RCS temperature monitoring and control (MRSS) / 4 039 Main and Reheat Steam System 2.1.32 Ability to explain and apply all system limits and Was 039 2.4.6 replaced with 039 2.1.32 since there (MRSS) / 4 precautions. are no EOP assosiated with this system (poker chip method) 073 Process Radiation Monitoring K3.01 Radioactive effluent releases (PRM) System / 7 075 Circulating Water System / 8 A4.01 Emergency/essential SWS pumps 1

PWR SRO imination Outline Printed: 10/05/1 Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Systems - Tier 2 / Group 2 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 079 Station Air System (SAS) / 8 K4.01 Cross-connect with LAS 079 Station Air System (SAS) / 8 A2.01 Cross-connection with LAS 086 Fire Protection System (FPS) / 8 KI.02 Raw service water 086 Fire Protection System (FPS) / 8 A3.01 Starting mechanisms of fire water pumps 103 Containment System / 5 K3.01 Loss of containment integrity under shutdown conditions 2

PWR SRO mination Outline Printed: 10/05/17 Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Systems - Tier 2 / Groun 3 Form ES-401-3 Sys/Ev # System / Evolution Name KA KA Topic Comment 005 Residual Heat Removal System K6.03 RHR heat exchanger (RHRS) / 4 041 Steam Dump System (SDS) and A3.03 Steam flow Turbine Bypass Control / 4 041 Steam Dump System (SDS) and 2.4.31 Knowledge of annunciators alarms and Was 041 2.4.49 replaced with 041 2.4.31 since Turbine Bypass Control / 4 indications, and use of the response instructions. there are no immediate actions assosiated with this system (poker chip method) 076 Service Water System (SWS) / 4 K2.01 Service water I

( Generic Knowledge ( Abilities Outline (Tier 3) Printed: 10/05/20t, PWR SRO Examination Outline Form ES-401-5 Facility: R.E. Ginna Nuclear Power Plant Generic Category KA KA Topic Comment Conduct of Operations 2.1.6 Ability to supervise and assume a management role during plant transients and upset conditions.

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

2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

2.1.34 Ability to maintain primary and secondary plant chemistry within allowable limits.

Category Total: 4 Equipment Control 2.2.23 Ability to track limiting conditions for operations.

2.2.26 Knowledge of refueling administrative requirements.

2.2.32 Knowledge of the effects of alterations on core configuration.

2.2.33 Knowledge of control rod programming.

2.2.34 Knowledge of the process for determining the internal and external effects on core reactivity.

Category Total: 5 Radiation Control 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.

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

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

2.3.11 Ability to control radiation releases.

Category Total: 4 I

r Generic Knowledge( Abilities Outline (Tier 3) Printed: 10/05/20(

PWR SRO Examination Outline Form ES-401-5 Facility: R.E. Ginna Nuclear Power Plant Generic Category KA KA Topic Comment Emergency Procedures/Plan 2.4.1 Knowledge of EOP entry conditions and immediate action steps.

2.4.7 Knowledge of event based EOP mitigation strategies.

2.4.23 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

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

Category Total: 4 Generic Total: 17 2

ES-401 PWR RO Examination Outline Printed: 10/05/2001 Facility: R.E. Ginna Nuclear Power Plant Form ES-401-4 Exam Date: 02/08/2002 Exam Level: RO K/A Category Points Tier Group KI K2 K3 K4 K5 K6 Al A2 A3 A4 G Point Total 1 3 3 2 3 3 2 16

1. ____ ________

Emergency 2 3 4 5 3 1 1 17 Abnormal Plant 3 0 1 0 1 1 0 3 Evolutions Totals 6 8 7 7 5 3 36 Tier 1 2 2 2 2 2 2 2 2 2 2 3 23 2.

2 2 2 2 2 2 2 2 2 1 2 1 20 Plant Systems 3 1 1 1 1 0 1 1 0 1 1 0 8 Tier Totals 5 5 5 5 4 5 5 4 4 5 4 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

PWR RO Examina( Outline Printed: 10/05/200(

Facility: R.n. Ginna Nuclear Power Plant ES - 401 Emerge cy and Abnormal Plant Evolunons - Tier 1I Group I rorm Eý1-4LU-4 E/APE # E/APE Name / Safety Function KA KA Topic Comment 015 Reactor Coolant Pump (RCP) Malfunctions /4 AK2.07 RCP seals 026 Loss of Component Cooling Water (CCW) / 8 AA2.04 The normal values and upper limits for the temperatures of the components cooled by CCW 027 Pressurizer Pressure Control (PZR PCS) AK2.03 Controllers and positioners Malfunction / 3 040 Steam Line Rupture / 4 AA 1.01 Manual and automatic ESFAS initiation 040 Steam Line Rupture / 4 AA2.03 Difference between steam line rupture and LOCA 057 Loss of Vital AC Electrical Instrument Bus /6 AA 1.01 Manual inverter swapping 067 Plant Fire on Site / 9 AA2. 10 Time limit of long-term-breathing air system for control room 068 Control Room Evacuation / 8 AK3.02 System response to turbine trip 068 Control Room Evacuation / 8 2.1.2 Knowledge of operator responsibilities during all modes of plant operation.

069 Loss of Containment Integrity / 5 AK3.01 Guidance contained in EOP for loss of containment integrity 069 Loss of Containment Integrity / 5 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

I

PWR RO Examina( Outline Printed: 10/05/200(

Facility: R.-. Ginna Nuclear Power Plant E'... vc Afil A ES - 401 Emergen cy ana Abnormal Plant Evolutions - Iier I I Group I rorun ES-41-*

E/APE # E/APE Name / Safety Function KA KA Topic Comment E07 Saturated Core Cooling / 4 EKI.3 Annunciators and conditions indicating signals, and remedial actions associated with the Saturated Core Cooling E08 Pressurized Thermal Shock / 4 EAL .1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E09 Natural Circulation Operations / 4 EKI.1 Components, capacity, and function of emergency Very similar to E09 EK2.1 systems E09 Natural Circulation Operations / 4 EK2.1 Components, and functions of control and safety Very similar to E09 EK1.2 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features El4 High Containment Pressure / 5 EKI.2 Normal, abnormal and emergency operating I_ I procedures associated with High Containment Pressure 2

PWR RO Examina( Outline Printed: 10/05/200(

Facility: R._. Ginna Nuclear Power Plant ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 Form ES-401-4 E/APE # E/APE Name / Safety Function KA KA Topic Comment 001 Continuous Rod Withdrawal / 1 AK3.W2 Tech-Spec limits on rod operability J 3,C, 3.o1 003 Dropped Control Rod / 1 AK2.05 Control rod drive power supplies and logic circuits 007 Reactor Trip / 1 EK2.02 Breakers, relays and disconnects 008 Pressurizer (PZR) Vapor Space Accident (Relief 2.1.30 Ability to locate and operate components, Valve Stuck Open) /3 including local controls.

009 Small Break LOCA /3 EK1.01 Natural circulation and cooling, including reflux boiling 009 Small Break LOCA /3 EK2.03 S/Gs 011 Large Break LOCA /3 EK2.02 Pumps 029 Anticipated Transient Without Scram (ATWS) / 1 EA2.05 System component valve position indications 033 Loss of Intermediate Range Nuclear Instrumentation AAI.03 Manual restoration of power

/7 037 Steam Generator (S/G) Tube Leak / 3 AK3.08 Criteria for securing RCP I

PWR RO Examin On Outline Printed: 10/05/240 Facility: R.*. Ginna Nuclear Power Plant W*m I*_AIN1 _zl ES - 401 Imerge cy and Abnormal Plantl EoVnluIonlls - I ir I / Group 2 AVI nA' E/APE # E/APE Name / Safety Function KA KA Topic Comment 038 Steam Generator Tube Rupture (SGTR) /3 EKI.04 Reflux boiling E01 Rediagnosis / 3 EKI.1 Components, capacity, and function of emergency systems E02 SI Termination /3 EK3.1 Facility operating characteristics during transient Very similar to E02 EA1.2 conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics E02 SI Termination / 3 EA1.2 Operating behavior characteristics of the facility Very similar to E02 EK3.1 Eli Loss of Emergency Coolant Recirculation /4 EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations El6 High Containment Radiation /9 EK3.2 Normal, abnormal and emergency operating procedures associated with High Containment Radiation E16 High Containment Radiation / 9 EA 1.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 2

Facility: (R.,,. Ginna Nuclear Power Plant PWR RO Examin 4 ,. Outline Printed: 10/05/204 17S-4i0l 1mro'niv 2nd Ahnnrmnl Plnnt EvnIuutinn - Tupr 1 I Grniin 3 iFnrm FS*-4fll-4 E/APE # E/APE Name / Safety Function KA KA Topic Comment 028 Pressurizer (PZR) Level Control Malfunction / 2 AK2.03 Controllers and positioners 065 Loss of Instrument Air / 8 AA1.02 Components served by instrument air to minimize drain on system El5 Containment Flooding / 5 EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility's I license and amendments 1

PWR RO ( mination Outline Printed: 10105 Facility: R.E. Ginna Nuclear 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 / 1 K1.04 RCS 001 Control Rod Drive System / 1 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

003 Reactor Coolant Pump System K3.04 RPS (RCPS) / 4 003 Reactor Coolant Pump System A1.05 RCS flow (RCPS) / 4 004 Chemical and Volume Control K5.14 Reduction process of gas concentration in RCS:

System (CVCS) / I vent-accumulated non-condensable gases from PZR bubble space, depressurized during cooldown or by alternately heating and cooling (spray) within allowed pressure band (drive more gas out of solution) 004 Chemical and Volume Control A1.05 S/G pressure and level System (CVCS) / 1 013 Engineered Safety Features K2.01 ESFAS/safeguards equipment control Actuation System (ESFAS) / 2 013 Engineered Safety Features K6.01 Sensors and detectors Actuation System (ESFAS) / 2 015 Nuclear Instrumentation System / 7 2.1.14 Knowledge of system status criteria which require the notification of plant personnel.

017 In-Core Temperature Monitor A2.01 Thermocouple open and short circuits (ITM) System / 7 017 In-Core Temperature Monitor A4.01 Actual in-core temperatures (ITM) System / 7 1 1 I

PWR RO mination Outline Printed: 10/05/(

(R.E. Ginna Nuclear Power Plant Facility:

ES - 401 Plant S s~terns. Ti-r / 9-2-~n 1 10*__ 1OL':t A/*tt

  • KA KA Topic Comment Sys/Ev # System / Evolution Name 022 Containment Cooling System A4.04 Valves in the CCS (CCS) / 5 022 Containment Cooling System K3.01 Containment equipment subject to damage by high (CCS) / 5 or low temperature, humidity, and pressure 059 Main Feedwater (MFW) System / K4.16 Automatic trips for MFW pumps 4

059 Main Feedwater (MFW) System / A2.05 Rupture in MFW suction or discharge line 4

061 Auxiliary / Emergency Feedwater K2.01 AFW system MOVs (AFW) System / 4 061 Auxiliary / Emergency Feedwater 2.1.23 Ability to perform specific system and integrated (AFW) System / 4 plant procedures during all modes of plant operation.

068 Liquid Radwaste System (LRS) / 9 A3.02 Automatic isolation 068 Liquid Radwaste System (LRS) / 9 K6. 10 Radiation monitors 071 Waste Gas Disposal System K 1.05 Meteorological tower (WGDS) / 9 071 Waste Gas Disposal System K4.06 Sampling and monitoring of waste gas release (WGDS) / 9 tanks 072 Area Radiation Monitoring (ARM) K5.02 Radiation intensity changes with source distance System / 7 072 Area Radiation Monitoring (ARM) A3.01 Changes in ventilation alignment System / 7 2

Printed: 10/05/(

PWR RO - mination Outline Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Systems - Tier 2 / Grouw 2 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 002 Reactor Coolant System (RCS) / 2 K5.09 Relationship of pressure and temperature for water at saturation and subcooling conditions 002 Reactor Coolant System (RCS) / 2 K6.03 Reactor vessel level indication 010 Pressurizer Pressure Control K2.01 PZR heaters System (PZR PCS) / 3 011 Pressurizer Level Control System A4.04 Transfer of PZR LCS from automatic to manual (PZR LCS) / 2 control 011 Pressurizer Level Control System K6.05 Function of PZR level gauges as postaccident (PZR LCS) / 2 monitors 012 Reactor Protection System / 7 A2.05 Faulty or erratic operation of detectors and function generators 014 Rod Position Indication System K3.02 Plant computer (RPIS) / 1 026 Containment Spray System (CSS) / K4.05 Prevention of material from clogging nozzles 5 during recirculation 029 Containment Purge System (CPS) / A1.02 Radiation levels 8

033 Spent Fuel Pool Cooling System A1.02 Radiation monitoring systems (SFPCS) / 8 035 Steam Generator System (S/GS) / K5.01 Effect of secondary parameters, pressure, and 4 temperature on reactivity 039 Main and Reheat Steam System K1.04 RCS temperature monitoring and control (MRSS) / 4 S temperature monitoring and control I

PWR RO1 - mination Outline Printed: 10/05/1 N

Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Systems - Tier 2 / Group 2 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 063 D.C. Electrical Distribution K2.01 Major DC loads System / 6 063 D.C. Electrical Distribution 2.1.32 Ability to explain and apply all system limits and System / 6 precautions.

073 Process Radiation Monitoring K3.01 Radioactive effluent releases (PRM) System / 7 075 Circulating Water System / 8 A4.01 Emergency/essential SWS pumps 079 Station Air System (SAS) / 8 K4.01 Cross-connect with IAS 079 Station Air System (SAS) / 8 A2.01 Cross-connection with IAS 086 Fire Protection System (FPS) / 8 K1.02 Raw service water 086 Fire Protection System (FPS) / 8 A3.01 Starting mechanisms of fire water pumps 2

PWR ROC ,mination Outline Printed: 10105(

Facility: R.E. Ginna Nuclear Power Plant ES - 401 Plant Systems - Tier 2 / Group 3 Form ES-401-4 Sys/Ev # System / Evolution Name KA KA Topic Comment 005 Residual Heat Removal System K6.03 RHR heat exchanger (RHRS) / 4 007 Pressurizer Relief Tank/Quench A4.01 PRT spray supply valve Tank System (PRTS) / 5 007 Pressurizer Relief Tank/Quench A 1.02 Maintaining quench tank pressure Tank System (PRTS) / 5 041 Steam Dump System (SDS) and A3.03 Steam flow Turbine Bypass Control / 4 041 Steam Dump System (SDS) and K4.18 Turbine trip Turbine Bypass Control / 4 076 Service Water System (SWS) /4 K2.01 Service water 076 Service Water System (SWS) /4 K1.16 ESF 103 Containment System / 5 K3.01 Loss of containment integrity under shutdown I conditions I

( Generic Knowledge( Abilities Outline (Tier 3) Printed: 10/05/20(

PWR RO Examination Outline Form ES-401-5 Facility: R.E. Ginna Nuclear Power Plant Generic Category KA KA Topic Comment Conduct of Operations 2.1.9 Ability to direct personnel activities inside the control room.

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

2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

Category Total: 3 Equipment Control 2.2.23 Ability to track limiting conditions for operations.

2.2.30 Knowledge of RO duties in the control room 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.

2.2.34 Knowledge of the process for determining the internal and external effects on core reactivity.

Category Total: 3 Radiation Control 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.

2.3.2 Knowledge of facility ALARA program.

2.3.11 Ability to control radiation releases.

Category Total: 3 I

Generic Knowledge I' Abilities Outline (Tier 3) Printed: 10/05/20ý(

PWR RO Examination Outline Form ES-401-5 Facility: R.E. Ginna Nuclear Power Plant Generic Category KA KA Topic Comment Emergency Procedures/Plan 2.4.10 Knowledge of annunciator response procedures.

2.4.17 Knowledge of EOP terms and definitions.

2.4.23 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

2.4.29 Knowledge of the emergency plan.

Category Total: 4 Generic Total: 13 2

ES-301 Control Room Systems Form ES-301-2 (R8, 51) and Facility Walk-Through Test Outline Facility: Ginna _,1\ Date of Examination: Feb 11, 2002 Exam Level (circle one Qjd)SRO([) / SRO(U) Operating Test No.: 02-01 B.1 Control Room Systems System I JPM Title Type Safety Code* Function

a. 001 Control Rod Drive System D, A, S J001.001 Perform Rod Exercises Per PT-1
b. 004 Chemical and Volume Control System D, S 2 J004.011 Place Excess Letdown in Service
c. 005 Residual Heat Removal System (PRI) M, A, S, L 4 J005.005 Line Up RCDT Pump For Core Cooling
d. 061 Auxiliary/Emergency Feedwater System (SEC) (ESF) D, S, L 4 J061.001 Place the Standby AFW System in Service
e. 062 AC Electrical Distribution D, S 6 J062.024 Transfer 1A Inst. Bus to Maintenance Power
f. 012 Reactor Protection System D, S 7 J012.003 Defeat Failed RCS Temperature Channel
g. 006 Emergency Core Cooling System M, A, S, L 3 J006.006 Transfer ECCS to Cold Leg Recirculation B.2 Facility Walk-Through
a. 004 Chemical and Volume Control System D, R 2 J004.009 Take Local Manual Control or Charging Pump
b. 064 Emergency Diesel Generators (ESF) M, A, L 6 J064.004 Start "A" EDG Locally Per ER-FIRE.1
c. 086 Fire Protection System D, C 8 J086.001 Reconnect Fire System
  • 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

ES-301 Control Room Systems Form ES-301-2 (R8, 51) and Facility Walk-Through Test Outline Facility: Ginna Date of Examination: Feb 11, 2002 Exam Level (circle one): RO / SRO(I) (SR(U) Operating Test No.: 02-01 B.1 Control Room Systems System / JPM Title Type Safety Code* Function

a. 005 Residual Heat Removal System (PRI) M, A, S, L 4 J005.005 Line Up RCDT Pump For Core Cooling
b. 061 Auxiliary/Emergency Feedwater System (SEC) (ESF) D, S, L 4 J061.001 Place the Standby AFW System in Service
c. 012 Reactor Protection System D, S 7 J012.003 Defeat Failed RCS Temperature Channel d.

e.

f.

g.

B.2 Facility Walk-Through

a. 004 Chemical and Volume Control System D, R 2 J004.009 Take Local Manual Control or Charging Pump
b. 064 Emergency Diesel Generators (ESF) M, A, L 6 J064.004 Start "A" EDG Locally Per ER-FIRE.1 C.
  • 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

ES-301 Administrative Topics Outline Form ES-301-1 (R8, $1)

Facility: Ginna Date of Examination: Feb 11, 2002 Examination Level (circle one): & SRO Operating Test Number: 02-01 Administrative Describe method of evaluation:

Topic/Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Conduct of JPM: J017.001 Determine RCS Core Exit Subcooling With the Operations PPCS Out of Service K/A 2.1.7 Importance 3.7 Not Applicable Conduct of JPM: J017.001 0-6.13 Daily Performance Logs K/A 2.1.18 Operations Importance 2.9 Not Applicable A.2 Equipment JPM: J343.004 A-52.12, Inoperability of Equipment K/A 2.2.24 Control Importance 2.6 Not Applicable A.3 Radiation Question: Knowledge of Work Stoppage Based on In-Progress Monitoring/ ALARA Review K/A 2.3.10 Importance 2.9 Control Question: Knowledge of Immediate Notification for Radiation Incidents K/A 2.3.1 Importance 2.6 A.4 Emergency JPM: J085.002 Complete NY State Radiological Emergency Procedures/ Data Form Part I (EPIP 1-5, Aft 3A) K/A 2.4.39 Importance 3.3 Plan Not Applicable

ES-301 Form ES-301-1 (R8, S1)

"Facility: Ginna Date of Examination: Feb 11, 2002 Examination Level (circle one): RO / Operating Test Number: 02-01 Administrative Describe method of evaluation:

Topic/Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Conduct of JPM: J001.010 Estimated Critical Rod Position Determination Operations K/A 2.1.23 Importance 4.0 Not Applicable Conduct of JPM: J017.001 Determine RCS Core Exit Subcooling With the Operations PPCS Out of Service K/A 2.1.7 Importance 4.4 Not Applicable A.2 Equipment JPM: Verify Equipment Tagout Boundary K/A 2.2.13 Control Importance 3.8 Not Applicable A.3 Radiation JPM: Approve Liquid Waste Release Form (Inoperable Effluent Monitoring/ Monitor) K/A 2.3.6 Importance 3.1 Control Not Applicable A.4 Emergency JPM: Perform Event Classification K/A 2.4.41 Importance 4.1 Procedures/

Plan Not Applicable

"'*A Lj Appendix D Scenario Outline Form ES-D-'

Facility: Ginna Scenario No.: 1 Op-Test No.: 01-01 Examiners: Lauahlin (Bissett) Operators:

Fish Silk Initial Conditions: Plant is at -48% reactor power, MOL. C, = 824 ppm. Power was reduced 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ago for condenser tube leakage and is ready to go back to full power.

BAST C. - 11.000 ppm. "B" MDAFW DumD and "C" charaing pump are OOS.

Turnover: Plant is at -48% reactor power, MOL. C=824 ppm. Power was reduced 4 hrs ago for condenser tube leakage and is ready to go back to full power. BAST C,=1 1,000 ppm. "C" charqing pump is OOS for excessive leakage, "B" MDAFW pump is OOS for check valve repair.

Event Malf. Event Event No. No. Type* Description 1 N/A N(CRF) Raise power to 100% lAW 0-1.2.

R(HCO 2 NIS07A I(CRF, PR channel N41 fails high, rods insert. (Enter ER-NIS.3, TS HCO) entry) (C-3 3 ROD2A C(CRF, Dropped control rodX (Enter AP-RCC.2 for RCC HCO) malfunction, 0-5.1 for load reduction) (\>,

4 CND07 C(AII) Loss of condenser vacuum-east 1 B, results in turbine/Rx A trip. (Enter AP-TURB.4 and EO) 5 EDS01 M(AII) Loss of offsite power. (Ef-toAP E LC".1) "A" EDG runs on A&B bus 14.

6 GEN04 C(AII) "A" EDG runs on bus 14, "B" EDG fails to auto-start but can B be started manually.

7 GEN04 M(AII) "A" EDG trips, station blackout. (Enter ECA-0.0) Terminate A I when transition to ECA-0.1 t I I I +

  • (N)ormal, (R)eactivity, (I)nrstrument, (C)omponent, (M)ajor (9n ttn ('Y 39 of 40 NUREG-1021, Revision 8, Supplement 1

ý0_2

Appendix D Operator Actions Form ES-D-2 Op-Test No.: __ Scenario No.: 1 Event No.: _1_ Page _1_ of 6_

Event

Description:

-Raise reactor power to 100% lAW )-1.2 Time Position I Applicant's Actions or Behavior CRF Direct start of "B" MFW pump per attachment "MFW Pump B"

,IF'*,,)P*, (Steps 7.0- 13.0) .P "

CO Lineup Service Water to MFW pump B oil cooler Take B MFW out of Pull Stop Verify MFW pump recirculation valve opens (AOV-4148)

Start "B" MFW pump Verify MFW pump discharge pressure and open discharge valve MOV-3976 CRF Direct AO to close B MFW pump discharge bypass valve

ý1-14.q,

/MOV-3976A CO Place HDT Level Controlerl LC-2013A in Auto CRF Direct AO to check MFW pump warmup valves closed HCO Check Delta I cumulative time on PPCS Verify QPTR is <1.02 _.

CRF Verify RP has a leak rate determined from the air ejector sample CO Raise valve position limit to 100%

Raise Setter and start load increase HCO Manually operate control rods/dilute as necessary to control Tave NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: _ Scenario No.: _1_ Event No.: 2 Page _2_ of 6_

Event

Description:

_PR Channel N41 Fails High Time Position Applicant's Actions or Behavior All Identify failed PR channel HCO Place rod control bank selector switch in manual HCO/CO Adiust Tave/ Tref as necessary CRF Address Technical Specifications (ITS3.2.3)

Direct NIS channel 41 to be defeated per "ttachment N-41 Defeat" HCO Verify rod control bank selector switch in manual HCO/CO Place DROPPED ROD MODE switch to bypass and verify following alarms - DROPPED ROD BYPASS is lit:

POWER RANGE ROD DROP BYPASS is lit:

Annunciator E-7 NIS TRIP BYPASS is lit Place T/405E DELTA T DEFEAT switch to LOOP A UNIT 1 Place OVERTEMP TRIP bistable switch to DEFEAT and verify the following - F-23 RCS OTA T CHANNEL ALERT is lit Red bistable status light OTA T LOOP A TC405C is lit Place OVERPOWER TRIP B/S switch to DEFEAT and verify the following: F-32 RCS OPAT CHANNEL ALERT is lit Red B/S status light OPAT LOOP A TC405A is lit Place UPPER SECTION DEFEAT switch to the PRN41 position &

verify the following: Local light for CHANNEL DEFEAT is lit Place LOWER SECTION DEFEAT switch to the PRN41 position &

verify the following: Local ight for CHANNEL DEFEAT is lit Place POWER MISMATCH BYPASS switch to BYPASS PRN41 Place ROD STOP BYPASS switch to BYPASS PRN41 Place COMPARATOR CHANNEL DEFEAT switch to N41 & verify Ifh-fllina COMI'DAPPAT-1 IF=lF=AT liaht is lit NUREG-1021, Revision 8, Supplement 1 40 of 40

dnnnrliv r') l')n*r*tnr Aafinn* Fnrm * .q,-I-)-9 Op-Test No.: __ Scenario No.: _1_ Event No.: _2 (con't)_ Page 3 of 6 Event

Description:

_PRN41 Failure Time Position Applicant's Actions or Behavior HCO/CO Remove 118V 5A AC INSTR POWER fuses & verify the following E-18 POWER RANGE LOSS OF DETECTOR VOLTAGE E-19 POWER RANGE HI RANGE CHANNEL ALERT 108%

E-21 POWER RANGE OVERPOWER ROD STOP 103%

E-27 POWER RANGE LO RANGE CHANNEL ALERT 24%

E-28 POWER RANGE ROD DROP ROD STOP 5%/5 SEC Verify the following red bistable lights (MCB) are lit HI POW RANGE P-10 NC41M HI POW RANGE P-8 NC41 N

_ _ _LO POW RANGE TRIP NC41P HI POW RANGE TRIP NC41 R HI POW RANGE P-9 NC41S Verify various status lighnt on PR N41A drawer are lit Verify following status lights on PRN41 B drawer are extinguished INSTRUMENT POWER ON CHANNEL ON TEST CRF Notify I&C to install jumpers HCO/CO Restore ROD CONTROL back to AUTO Reset dropped rod rod stop signals at OR NIS drawers CRF Check Tech Specs NUREG-1021, Revision 8, Supp5lement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No.: _1 Event No.: _3 Page 4 of 6 Event

Description:

Dropped Control Rod 2A (Enter AP-RCC.2 for RCC malfunction, 0-5.1 for load reduction) Annunciator C-5 Rod Deviation light lit, Annunciator F-29 PPCS or QUADRANT POWER TILT lit.

Time Position Applicant's Actions or Behavior HCO/CRO Place Rod Control Bank Selector Switch to MANUAL Check Dropped Rod Indication - Pwr and Tave decreasing Go to AP-RCC.3 (Dropped Rod Recovery)

CO Place EH control in MANUAL Reduce turbine load as necessary to match Tave and Tref Verify Annunciator G-15 STEAM DUMP ARMED- EXTINGUISHED Check Main Generator Load - GREATER THAN 15 MW Establish Stable Plant Conditions Check REGEN HX Letdown Indications CRF Evaluate Control Rod Operability HCO/CO Go toER-RCC.1 RETRIEVAL OF A DROPPED ROD (are we going to try and retrieve?)

1- 1--

1 1 4 -4 I 4 I I I I t t NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No.: 1 Event No.: 4 Page_5 of 6 Event

Description:

Loss of condenser vacuum- east 1 B resulting in a turbine/Rx trip Time Position Applicant's Actions or Behavior CO Identifies decreasing vacuum, monitors condenser indications CRF Directs entry into AP-TURB.4 LOSS OF CONDENSER VACUUM Dispatches AO to perform local actions CRF Directs Rx Trip and entry into E-0 HCO/CO Performs Immediate Actions of E-0 Verify Rx Trip Verify Turbine Stop Valves Shut Verifv BothTrains of AC Emergency Buses Energized Check if SI is Actuated SI NOT Required - Transition to ES 0.1 Reactor Trip Response Monitor RCS Tave Check SIG Feed Flow Status Verify all rods on bottom

____ I -'. IA- / A Verify AI'AC Buses ENERGIZED BY OFFSITE POWER - NO Perform RNO actions of step 4 of ES-0,1 Verify at Least Two SW Pumps running - NO Start one SW pump per RNO step 5 Verify IA Available Check PZR Level Control - start charging pump(s) per RNO step 7

_ - _: "A" D/G trips - Loss of all AC CRF Directs transition to ECA-0.0 Loss of all AC 1* F

  • 1* F

NUREG-1021, Revision 8, Supplement 1 40 of 40 Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No.: _1_ Event No.:5,6,7 Page 6 of 6 Event

Description:

Loss of all AC Time Position Applicant's Actions or Behavior CRF Directs immediate actions of ECA-0.0 CO Close MSIVs HCO Isolate RCS by closing AOV 200A, B. C, AOV 371, 427& AOV 310 CO Verify adequate TDAFW flow >200 gpm Try to restart a DIG CRF Direct AO to Ioca!ly restart a DIG HCO/CO Pull Stop Equipment Isolate RCP seal injection Place hotwell level control in manual at 50%

Check S/G status - intact CRF Direct manual start of "B" DIG 3)&d W/A6,-17 CO Manually control ARV to stabilize RCS temp Restore SW pumps Verify equipment loaded on available AC emergency buses CRF Direct AO to check battery chargers Direct transition to ECA-0.1 ORF Site Area Classification

Appendix D Scenario Outline Form ES-D-1 i.i Facility: Ginna Scenario No.: 01-02 Op-Test No.:

Examiners: Bissett Operators:

Fish Laughlin Initial Conditions: Plant is at 100% power, BOL, C, 1329, xenon equilibrium. PORV-430 isolated due to high leakage. MOV-516 closed. BAST C, - 11,000 ppm.

Turnover:

Event Malf. Event Event No. No. Type* Description 1 PZR01 C(CRF, PZR spray valve PCV-431A fails open approx 50%. (Enter A HCO) AP-PRZR.1)

NIS8A I (CRF, fuzoor ;nt,tae,, Ml.-n A channe', 35. (Eni

___HCO) ER IN1 . 2) 2A 3 TUR05 C(CRF) Turbine vibration increases. (Enter AP-TURB.3, requires C R(HCO load reduction to stabilize vibration) 4 SGN04 M(AII) ,e-_Ron S/G !A ati7Gd§-gpm. (Enter E-0,-E-.-AO

__ _A ___S____G___--D____ re-5 TUR02 C(CRF, Turbine fails to trip. (Manually trip turbine per E-0)

TUR11 CO)

D b..

,6--7 SIS03B C(CRF, 1B SI pump fails to staFt.

Af HCO)

PZR05 C(CRF, PORV 431 fails open, resulting in SBLOCA. (Enter ECA-3.1, B HCO) ,AP-PZR.1,AP ,RCC.1) Terminate when RCS cool-down is underway.

+ 4

+ 4

ý (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 39 of 40 NUREG-1021, Revision 8, Supplement 1

Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No.: 2 Event No.: 1 Page -1 of 7 Event

Description:

Pzr spray valve PCV-431A fails open (-50%)

Time Position Applicant's Actions or Behavior HCO Identifies stuck open spray valve, RCS pressure decrease CRF Directs entry into AP-PRZR.1 "Abnormal Pressurizer Pressure" HCO Checks Pzr Pressure, Reactor Power, Pzr Heater Status, Pzr Spray valve closed Place controllers in manual (.0% demand Check Pzr Pressure controller 431 K, Demand <50%

Check PORVs closed, Check Pzr safety valves closed Check Aux Spray valve closed Restore Pzr pressure control Check PRT Indications CRF Notify plant supervision and maintenance and reactor Engineering 4 .4.

4 4.

4 4 4 1 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No.: 2 Event No.: -2_ Page -2 of 7 Event

Description:

Blown fuse on intermediate range A channel 35 Time Position Applicant's Actions or Behavior HCO Identifies IR channel 35 failure CRF Directs entry into ER-NIS.2 "IR Malfunction" HCO Defeats reactor trip and rod stopfunction for IR35 by placing level Bypass position Contacts I&C CRF Refers to TS Section 3.3.1.,iable 3.371-1. Function #3 and #16a Pz, 7 a e2 -. ucto i i t I 4 4 t 4 4 4 I. t 4 I

.4 4 .4

+ 4 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: _ Scenario No.: 2 Event No.: 3 Page 3_ of 7_

Event

Description:

Turbine vibration increase resulting in a load reduction Time Position Applicant's Actions or Behavior CRF Recognize hi turbine vibration, enter AP-TURB.3 "r-27 Verify turbine trip not required CRF/CO Reduce turbine load to stabilize vibrations Continue load reduction until vibrations stabilize, then stop load decrease HCO Stabilize primary systems CRF Direct walkaround inspection of turbine Notify higher supervision Notify Maintenance Mgr.

_ _ I __ ________

i i t 4 4 4 I. t NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No.: 2 Event No.: 4 Page 4 of 7 Event

Description:

SGTR on S/G A at 400qpm Time Position Applicant's Actions or Behavior HCO Identifies prz level decrease and pressure decrease CO Identifies SF/FF mismatch on A S/G HCO Charging pump speed alarm/flow increase Increasing radiation levels on R-15, R-19 and R-31 Start additional charging pumps Close loop B cold leg to reqen Hx AOV-427 CRF Enter AP-SG. 1 Steam Generator Tube Leak Direct Rx Trip if charging pumps running at max speed with Letdown isolated 7-/7O-#-

__ I ____ 1 __________________ .1 t

  • r r 4 1 T

+ 4 NUREG-1021, Revision 8, Supplement 1 40 of 40

Ann*nrliY I'* E*rrr* I:: *_I-LO 11FU% V I I~ aJ

%J.llJIf -L -lC Op-Test No.: _ Scenario No.: 1 Event No.: _5,6- Page 5_ of 7_

Event

Description:

Rx Trip, turbine fails to trip, 1 B SI pump fails to start Time Position Applicant's Actions or Behavior CRF Direct actions of E-0 HCO/CO Verify Rx Trip Verify turbine stop valves closed- NO- MANUALLY TRIP TURBINE Verify AC Emergency Busses Energized Check if SI Actuated Verify SI/RHR pumps running -NO-MANUALLY START B SI PUMP Verify CNMT RECIRC FANS running Verify CNMT Spray NOT required Check if Main Steamlines should be isolated Verify MFW Isolation Verify AFW Pumps Running Verify at least Two SW Pumps running Verify Cl and CVI Check CCW System Status Verify SI and RHR Pump Flow Verify AFW Flow > 200 GPM Verify AFW Valve Alignment Verify SI Pump and RHR Pump Emergency Alignment Check CCW Flow to RCP Thermal Barriers Check PZR PORVs and Spray Valves Monitor RCP Trip Criteria Check if S/G Secondary Side is Intact I Check if S/G Tubes are Intact - NO- Transition to E-3 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.:_ Scenario No.: 2 Event No.: 7 Page 6 of 7 Event

Description:

-Steam Generator Tube Rupture

[Time Position Applicant's Actions or Behavior CRF Direct actions of E-3 Steam Generator Tube Rupture HCO/CO Monitor RCP Trip Criteria Identify Ruptured S/G- 1A S/G Isolate Flow From Ruptured 1A S/G Complete Ruptured S/G Isolation Check Ruptured S/G Level Verify Ruptured S/G Isolated Establish Condenser Steam Dump Pressure Control Reset SI Initiate RCS Cooldown Monitor Intact S/G Levels Check PZR PORVs and Block Valves Reset Cl Monitor AC Busses - Energized by Offsite Power Verify SW Flow Establish IA to CTMT - AOV 5392 FAILS to OPEN Check if RHR Pumps should be stopped Y oor'-,470 Establish Charging Flow Check if RCS Cooldown Should be Stopped Depressurize RCS to minimize break Flow and Refill Pzr via PORV Check RCS Pressure INCREASING - NO - TRANSITION TO ECA-3.1 NUREG-1021, Revision 8, Supplement 1 40 of 40 Appendix D Operator Actions Form ES-D-2

Op-Test No.: _ Scenario No.: _2_ Event No.: 7 Page 7 of 7 Event

Description:

PORV 431 Fails Open During RCS Depressurization Time Position [ Applicant's Actions or Behavior CRF Direct actions of ECA-3.1 HCO/CO Reset SI and CI Verify adequate SW Flow Establish IA to CTMT - NO AC Busses energized by offsite power Monitor CTMT Spray Pumps - STOPPED Check Ruptured 1A S/G Level Stopped RHR Pumps Evaluate Plant Status Establish 75 GPM Charging Flow Check S/G Secondary Side and Intact S/G Levels Initiate RCS Cooldown to Cold Shutdown CRF Classify as Alert NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Scenario Outline Form ES-D-1 Facility: Ginna Scenario No.: 01-03 Op-Test No.:

Examiners: Bissett Operators:

Fish Laughlin Initial Conditions: The plant is at 100% power BOL xenon equilibrium. Boron=1329ppm, BAST conc. = 11,000. Circuit 751 is OOS due to an auto accident, "D"SW pump is OOS due to motor failure.

Turnover:

Event Malf. Event Event No. No. Type* Description 1 PZR2D I(CRF, PZR pressure channel PT-449 fails high. (Enter AP-PRZR.1, HCO) ER-INST.1 to defeat channel) 2 RCS14 C(CRF, "B" RCP #3 seal failure., (Enter AP-FCP.1)

B HCO) 3 RCS2A C(CRF, RCS leak inside containment from loop A hot leg, 15 gpm.

HCO) (Enter AP-RCS.1) i.. _

4 N/A N(CRF) Perform plant shutdown in response to RCS leak. (Enter 0 R(HCO 2.1, 100% to 95%),(A(- -,tAKO_.S) 5 CND8 C(CRF, Condensate header break 20K gpm, complete loss of main CO) feedwater. (Enter E-0, 15< (A- 1w,.1) 6 RPS5A M(AII) ATWS (Enter FR-S.1)

&B 7 TUR2 .. C(CRF, Main turbine fails to automatically trip.

/

I I I 8 -... .

o  ; ion-,. (E.nt, ER CVeC,-tT I I i I Terminate drill when SI termination criteria met in E-1.

I (N)ormal,I I (R)eactivity, I

(I)nstrument, (C)omponent, (M)ajor 39 of 40 NUREG-1021, Revision 8, Supplement 1

Appendix D Operator Actions Form ES-D-2 Op-Test No.: _ Scenario No.: 3 Event No.: ___ Page _1_ of 8 Event

Description:

_PZR pressure channel failure PT-449 fails HI Time Position Applicant's Actions or Behavior CRF PT-449 fails HI, Directs entry into AP-PZR.1Abnormal PZR PRESS HCO/CO Acknowledges ..... 2-.aF,+,,- -, t Checks PZR Press - Refers to ER-INST.1

___-__ Place 431K in MANUAL ( -50%

Refer to Attachment PZR PRESSURE PI-449 YELLOW CHANNEL to defeat failed channel Place P/429A to DEFEAT-1 (PLP PZR PRESS/LVL RACK)

Place T/405F DELTA T DEFEAT switch to LOOP B UNIT 2 (RIL INSERTION LIMIT Rack)

In Y-1 PROTECTION CHANNEL 4 rack Place B/S switches To DEFEAT 'm- F-,=

408 LOOP B OVER TEMP TRIP 449 CHANNEL 4 - LOW PRESS TRIP Place PZR pressure recorder to positinp4-3 (MCB)

Delete 404/408 from the PPCS S..... VV *v*

w D ~~i

+ D7 .. I D oý re I.'tJI E r%J1 n +e nJ CRF Refer to ITS for applicable LCOs Section 3.3.1 Table 3.3.1-1 Functions 5 and 7a Section 3.3.3 Table 3.3.3-1 Functions 1 and 6 Check TRM 3.4.3 ATWS mitigation Notify maintenance and higher supervision NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No 3 Event No.: 2 Page 2 of 8 Event

Description:

_"B" RCP #3 seal failure Time Position Applicant's Actions or Behavior HCO Acknowledges ANN B-i 2RCP STAND PIPE LO LEVEL -4FT...

CRF Directs actions of AP-RCP.1 RCP SEAL MALFUNCTION' -AL HCO/CO Check Total #1 Seal Flow < 8.0 GPM Check RCP Seal Return valve Alignment MOV313 Open, AOV27OA/B Open Check Total #1 Seal Flow Between .8 - 6.0 GPM Check RCP cooling 1--2, A,-15" A, .

Check RCP #2 Seal Indications Check RCP Labyrinth Seal D/Ps > 15" Check RCP #3 Seal Indications ;3//1) -/,

RNO Check CTMT rad monitors R/) R-/2 Monitor RCP Seal Conditions 4 4-9 4 4 4 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Form ES-D-2 Op-Test No.:_ Scenario No.: 3 Event No.: 3 Page 53 of 8 Event

Description:

_RCS leak inside CTMT from A loop hot leg FTime Position Applicant's Actions or Behavior CRF Directs actions of AP-RCS. 1 REACTOR COOLANT LEAK HCO/CO Acknowledges ANN F-14, A-2, E-16. F-4 Check PZR level (Decreasing) RNO actions Start additional charging pumps Check VCT M/U System Check if RCS leakage in CTMT Dispatch AO to Aux Bldg Check for leak to CCW System Check CVCS Conditions Check AUX Bldg radiation levels Check PRT Indications Check S/Gs for Leakage -- A.kA.1t,'

Check Sl Accumulator levels Check RCP Seal Leakoff Flows Check RCDT Leak Rate Check Valve Leakoff Temps Establish Stable Plant Conditions Evaluate RCS Leakage 172 &mA1 RNO - Commence Plant Shutdown(at 1%/min CRF Notify higher supervision NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: _ Scenario No.: 3 Event No.: 4 Page _4_ of 8.

Event

Description:

-Rapid Plant Shutdown due to RCS leak Time Position Applicant's Actions or Behavior CRF Direct actions of AP-TURB.5 RAPID LOAD REDUCTION HCO/CO Initiate load reduction Monitor RCS Tave 5"-*.,

Borate as necessary C,4,

4. ý..

Check IA to CTMT Monitor plant parameters NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: Scenario No.: Event No.: o5 m l5o Page of _8F DPOperato-TCondensate header Event

Description:

Apenime 20K am. breakActioants Loss of MFW FormnsESBehvio Time Position JApplicant's Actions or Behavior ORF Direct actions -AP-FW.1 Partial or complete loss of MFW HCOICO Check MFW requirements Verify MEW pump status

_______Check MEW Pump suction pressure

__________Total loss of MFW - Transition to E-0 4 4.

4. 4.
4. 4.

NUREG-1021, Revision 8, Supplement 1 40 of 40

A .... ,.,,II;.. r'-. SJ. A L *'*w* I*C"* ** *J AppenJix L Operator Actions ruorii io-u-/

Op-Test No.: _ Scenario No.: 3 Event No.: 6 Page _6_ of 8_

Event

Description:

_A'WS-and Failure of Main Turbine to Trip Time Position Applicant's Actions or Behavior CRF Direct actions of E-0 HCO/CO Verify Rx Trip - NO Manually trip the reactor - NO CRF Transition to FR-S.1 HCO/CO Verify Rx Trip - NO RNO- Manually trip reactor Manually insert rods Verify Turbine Stop Valves closed - NO Manually trip turbine Verify AFW flow Initiate Emergency Boration Check PZR PORV status - NO Open PORVs as necessarytIq control pressure Verify CTMT ventilation isolation 7 CRF Dispatch AO to locally trip reactor - YES Transition to E-0 Direct actions of E-0 HCO/CO Verify Rx Trip Verify turbine stop valves closed Verify AC emergency busses Check if SI is actuated Verify SI and RHR pumps running Verify CTMT recirc fans running Verify CTMT spray not actuated NUREG-1021, Revision 8, Supplement 1 40 of 40 Appendix D Operator Actions Form ES-D-2

Op-Test No.: _ Scenario No.: 3 Event No.: _6 con't Page _7_ of 8_

Event

Description:

ATWS and Failure of Main Turbine to Trip

[Time Position Applicant's Actions or Behavior HCO/CO Check if any main steamline should be isolated Verify MFW isolation Verify AFW pumps running Verifv Cl and CVI Check CCW system status Verify SI and RHR flow Verify AFW flow > 200 gpm Verify Sl pump and RHR pump emergency alignment Check CCW flow to RCP Thermal barriers Check if TDAFW pump can be stopped Monitor RCS Tave- stable or trendinq to 547 deqrees Check PZR PORVs and Spray valves Monitor RCP Trip Criteria Check if S/G Secondary side is intact Check if S/G Tubes are intact Check if RCS is intact - NO Transition to E-1 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D, Operator Actions Form ES-D-2 Op-Test No.: _ Scenario No.: 3 Event No.: Page _8_ of 8_

Event

Description:

Loss of Reactor or Secondary Coolant (E-1)

Time Position Applicant's Actions or Behavior CRF Direct actions of E-1 HCO/CO Monitor RCP Trip Criteria Check if S/G secondary side intact Monitor intact S/G levels Monitor if secondary radiation levels are normal Monitor PRZ PORV status Reset SI and CI Verify adequate SW flow Establish IA to CTMT Check normal power to charging pumps Check if charging flow has been established Check if SI should be terminated Monitor if CTMT spray should be stopped Monitor if RHR pumps should be stopped Check RCS and S/G pressures Check if EDGs should be stopped Check if RHR should be throttled Verify CTMT sump2 recirculation capability Evaluate Plant Status NOTE: SHOULD MEET SI TERMINATION CRITERIA PER FOLDOUT PAGE CRITERIA OR STEP 12 OF E-1 CRF Transition to ES-1.1, SI TERMINATION Classify as a Site Area _o

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I /7 asa e2V-Z,:ý) V -A f Lf , LuY'- 06T Ag,4 NUREG-1021, Revision 8, Su6Wment 1 40 of 40 Y I

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