ML093200655
| ML093200655 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/16/2009 |
| From: | Division of Reactor Safety III |
| To: | |
| References | |
| SO-2009-10 | |
| Download: ML093200655 (32) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Facility: SONGS 2 & 3 NRC Date of Exam: 10/19/09 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4 *
- 1. Emergency 1 4 3 2 2 5 2 18 3 3 6
& Abnormal 2 0 1 3 3 1 1 9 2 2 4 Plant Evolutions Tier Totals 4 4 5 5 6 3 27 5 5 10 1 3 3 2 2 2 1 3 4 1 3 4 28 2 3 5
- 2. Plant Systems 2 1 0 1 0 1 2 1 2 1 1 0 10 0 1 2 3 Tier Totals 4 3 3 2 3 3 4 6 2 4 4 38 3 5 8
- 3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 Categories 10 7 3 3 2 2 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems / evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems / evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
NUREG-1021, Revision 9.1 Page 1 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 E/APE # / Name Safety Function K1 K2 K3 A1 A2 G Number K/A Topic(s) Imp. Q#
Ability to determine or interpret the following as 040 / Steam Line Rupture - Excessive Heat X AA2.05 they apply to the Steam Line Rupture: When 4.5 76 Transfer / 4 ESFAS systems may be secured Ability to determine and interpret the following as they apply to the Reactor Trip Recovery:
CE / E02 / Reactor Trip - Stabilization - Recovery X EA2.2 Adherence to appropriate procedures and 4.0 77
/1 operation within the limitations in the facility*s license and amendments.
Conduct of Operations: Ability to interpret and 015/17 RCP Malfunctions / 4 X 2.1.20 4.6 78 execute procedure steps Equipment Control: Ability to determine 009 / Small Break LOCA / 3 X 2.2.37 operability and/or availability of safety related 4.6 79 equipment Ability to determine the following as they apply 025 Loss of RHR System / 4 X AA2.07 to the Loss of Residual Heat Removal System: 3.7 80 Pump cavitation Equipment Control: Ability to recognize system 011 Large Break LOCA / 3 X 2.2.42 parameters that are entry-level conditions for 4.6 81 Technical Specifications Knowledge of the interrelations between a 007 Reactor Trip - Stabilization - Recovery / 1 X EK2.03 reactor trip and the following: Reactor trip status 3.5 39 panel Knowledge of the operational implications of the following concepts as they apply to a 008 / Pressurizer Vapor Space Accident / 3 X AK1.02 3.1 40 Pressurizer Vapor Space Accident: Change in leak rate with change in pressure Knowledge of the interrelations between the 009 / Small Break LOCA / 3 X EK2.03 3.0 41 small break LOCA and the following: SGs Ability to operate and/or monitor the following as they apply to a large break LOCA: Manual 011 / Large Break LOCA / 3 X EA1.05 4.3 42 and/or automatic transfer of suction of charging pump to borated source Ability to determine and interpret the following as they apply to the Reactor Coolant Pump 015 / 17 / RCP Malfunctions / 4 X AA2.01 3.0 43 Malfunctions (Loss of RC Flow): Cause of RCP failure NUREG-1021, Revision 9.1 Page 2 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 E/APE # / Name Safety Function K1 K2 K3 A1 A2 G Number K/A Topic(s) Imp. Q#
Ability to determine and interpret the following as they apply to the Loss of Residual Heat 025 / Loss of RHR System / 4 X AA2.05 3.1 44 Removal System: Limitations on LPI flow and temperature rates of change Knowledge of the reasons for the following responses as they apply to the Loss of 026 / Loss of Component Cooling Water / 8 X AK3.02 Component Cooling Water: The automatic 3.6 45 actions (alignments) within the CCWS resulting from the actuation of ESFAS Knowledge of the operational implications of the 027 / Pressurizer Pressure Control System following concepts as they apply to the X AK1.02 2.8 46 Malfunction / 3 Pressurizer Pressure Control Malfunctions:
Expansion of liquids as temperature increases Ability to determine or interpret the following as 029 / ATWS / 1 X EA2.07 they apply to a ATWS: Reactor trip breaker 4.2 47 indicating lights Knowledge of the operational implications of the 038 / Steam Generator Tube Rupture / 3 X EK1.01 following concepts as they apply to the SGTR: 3.1 48 Use of steam tables Equipment Control: Ability to interpret control room indications to verify the status and 055 / Station Blackout / 6 X 2.2.44 operation of a system, and understand how 4.2 49 operator actions and directives affect plant and system conditions Ability to determine and interpret the following as 056 / Loss of Offsite Power / 6 X AA2.20 they apply to the Loss of Offsite Power: AFW 3.9 50 flow indicator Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC 057 / Loss of Vital AC Instrument Bus / 6 X AK3.01 4.1 51 Instrument Bus: Actions contained in EOP for loss of vital AC electrical instrument bus Knowledge of the operational implications of the following concepts as they apply to Loss of DC 058 / Loss of DC Power / 6 X AK1.01 2.8 52 power: Battery charger equipment and instrumentation Emergency Procedures/Plan: Knowledge of 062 / Loss of Nuclear Service Water / 4 X 2.4.35 local auxiliary operator tasks during an 3.8 53 emergency and the resultant operational effects NUREG-1021, Revision 9.1 Page 3 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 E/APE # / Name Safety Function K1 K2 K3 A1 A2 G Number K/A Topic(s) Imp. Q#
Ability to determine and interpret the following as 065 / Loss of Instrument Air / 8 X AA2.03 they apply to the Loss of the Instrument Air: 2.6 54 Location and isolation of leaks Ability to operate and/or monitor the following as 077 / Generator Voltage and Electric Grid X AA1.05 they apply to Generator Voltage and Electric 3.9 55 Disturbances / 6 Grid Disturbances: Engineered safety features Knowledge of the interrelations between the Excess Steam Demand and the following:
Facility*s heat removal systems, including CE / E05 / Steam Line Rupture - Excessive Heat X EK2.2 primary coolant, emergency coolant, the decay 3.7 56 Transfer / 4 heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
K/A Category Point Totals: 4 3 2 2 5/3 2/3 Group Point Total: 18 / 6 NUREG-1021, Revision 9.1 Page 4 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 E/APE # / Name Safety Function K1 K2 K3 A1 A2 G Number K/A Topic(s) Imp. Q#
Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear 032 / Loss of Source Range NI / 7 X AA2.07 3.4 82 Instrumentation: Maximum allowable channel disagreement Emergency Procedures / Plan: Knowledge of the 068 / Control Room Evacuation / 8 X 2.4.41 emergency action level thresholds and 4.6 83 classifications Ability to determine and interpret the following as 067 / Plant Fire on Site / 8 X AA2.03 3.5 84 they apply to the Plant Fire on Site: Fire alarm Equipment Control: Ability to recognize system CE / E09 / Functional Recovery X 2.2.42 parameters that are entry level conditions for 4.6 85 Technical Specifications Ability to operate and/or monitor the following as 060 / Accidental Gaseous Radwaste Release / 9 X AA1.02 they apply to the Accidental Gaseous Radwaste: 2.9 57 Ventilation system Knowledge of the reasons for the following responses as they apply to the Loss of 051 / Loss of Condenser Vacuum / 4 X AK3.01 2.8 58 Condenser Vacuum: Loss of steam dump capability upon loss of condenser vacuum Conduct of Operations: Knowledge of system 032 / Loss of Source Range NI / 7 X 2.1.27 3.9 59 purpose and/or function Knowledge of the reasons for the following responses as they apply to the Steam Generator 037 / Steam Generator Tube Leak / 3 X AK3.06 3.6 60 Tube Leak: Normal operating precautions to preclude or minimize SGTR Knowledge of the interrelations between the 059 / Accidental Liquid Radwaste Release / 9 X AK2.01 accidental liquid Radwaste release and the 2.7 61 following: Radioactive liquid monitors Ability to determine and interpret the following as 067 / Plant Fire on Site / 8 X AA2.05 they apply to the Plant Fire on Site: Ventilation 3.2 62 alignment necessary to secure affected area NUREG-1021, Revision 9.1 Page 5 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 E/APE # / Name Safety Function K1 K2 K3 A1 A2 G Number K/A Topic(s) Imp. Q#
Knowledge of the reasons for the following responses as they apply to the Natural Circulation Operations: RO or SRO function within the control room team as appropriate to CE / A13 / Natural Circulation / 4 X AK3.4 3.1 63 the assigned position, in such a way that procedures are adhered to and the limitations in the facility license and amendments are not violated Ability to operate and monitor the following as they apply to the Excess RCS Leakage:
Components, and functions of control and safety CE / A16 / Excess RCS Leakage / 2 X AA1.1 3.4 64 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Ability to operate and / or monitor the following 068 / Control Room Evacuation / 8 X AA1.28 as they apply to the Control Room Evacuation: 3.8 65 PZR level control and pressure control K/A Category Point Totals: 0 1 3 3 1/2 1/2 Group Point Total: 9/4 NUREG-1021, Revision 9.1 Page 6 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Number K/A Topics Imp. Q#
Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; 010 / Pressurizer Pressure and (b) based on those predictions, use X A2.01 3.6 86 Control procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater failures Equipment Control: Ability to interpret control room indications to verify the status and 059 / Main Feedwater X 2.2.44 operation of a system, and understand how 4.4 87 operator actions affect plant and system conditions 007 / Pressurizer Equipment Control: Ability to apply Technical X 2.2.40 4.7 88 Relief/Quench Tank Specifications for a system Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and 013 / Engineered Safety (b) based on those predictions, use procedures X A2.06 4.0 89 Features Actuation to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent ESFAS actuation Emergency Procedures/Plan: Knowledge of the parameters and logic used to assess safety functions, such as reactivity control, core 076 / Service Water X 2.4.21 4.6 90 cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
Emergency Procedures/Plan: Knowledge of 003 / Reactor Coolant Pump X 2.4.11 4.0 1 abnormal condition procedures Knowledge of the physical connections and/or 003 / Reactor Coolant Pump X K1.10 cause-effect relationships between the RCPs 3.0 2 and the following systems: RCS Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and 004 / Chemical and Volume (b) based on those predictions, use procedures X A2.25 3.8 3 Control to correct, control, or mitigate the consequences of those malfunctions or operations: Uncontrolled boration or dilution NUREG-1021, Revision 9.1 Page 7 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Number K/A Topics Imp. Q#
Knowledge of the operational implications of 005 / Residual Heat Removal X K5.03 the following concepts as they apply to the 2.9 4 RHRS: Reactivity effects of RHR fill water Knowledge of bus power supplies to the 005 / Residual Heat Removal X K2.01 3.0 5 following: RHR pumps Knowledge of the operational implications of the following concepts as they apply to ECCS:
006 / Emergency Core X K5.07 Expected temperature levels in various 2.7 6 Cooling locations of the RCS due to various plant conditions Knowledge of the PRTS feature(s) and/or 007 / Pressurizer Relief /
X K4.01 interlock(s) which provide for the following: 2.6 7 Quench Tank Quench tank cooling Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS; and 008 / Component Cooling (b) based on those predictions, use procedures X A2.04 3.3 8 Water to correct, control, or mitigate the consequences of those malfunctions or operations: PRMS alarm Emergency Procedures/Plan: Ability to verify 008 / Component Cooling X 2.4.46 that the alarms are consistent with plant 4.2 9 Water conditions Knowledge of the effect that a loss or 010 / Pressurizer Pressure X K3.02 malfunction of the PZR PCS will have on the 4.0 10 Control following: RPS Knowledge of bus power supplies to the 012 / Reactor Protection X K2.01 following: RPS channels, components, and 3.3 11 interconnections Knowledge of the effect of a loss or 013 / Engineered Safety X K6.01 malfunction on the following will have on the 2.7 12 Features Actuation ESFAS: Sensors and detectors Ability to predict and/or monitor changes in parameters (to prevent exceeding design 022 / Containment Cooling X A1.04 3.2 13 limits) associated with operating the CCS controls including: Cooling water flow Ability to manually operate and/or monitor in 026 / Containment Spray X A4.01 4.5 14 the control room: CSS controls NUREG-1021, Revision 9.1 Page 8 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Number K/A Topics Imp. Q#
Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to 026 / Containment Spray X A2.01 correct, control, or mitigate the consequences 2.7 15 of those malfunctions or operations: Reflux boiling pressure spike when first going on recirculation Ability to manually operate and/or monitor in 039 / Main and Reheat Steam X A4.01 2.9 16 the control room: Main steam supply valves Conduct of Operations: Knowledge of system 059 / Main Feedwater X 2.1.27 3.9 17 purpose and/or function Ability to predict and/or monitor changes in 061 / Auxiliary/Emergency parameters (to prevent exceeding design X A1.02 3.3 18 Feedwater limits) associated with operating the AFW controls including: Steam generator pressure Knowledge of the physical connections and/or 061 / Auxiliary/Emergency cause-effect relationships between the AFW X K1.07 3.6 19 Feedwater system and the following systems: Emergency water source Knowledge of the effect that a loss or 062 / AC Electrical malfunction of the AC distribution system will X K3.02 4.1 20 Distribution have on the following: Emergency diesel generator Knowledge of DC electrical system feature(s) 063 / DC Electrical and/or interlock(s) which provide for the X K4.02 2.9 21 Distribution following: Breaker interlocks, permissives, bypasses and cross-ties 064 / Emergency Diesel Knowledge of the bus power supplies to the X K2.02 2.8 22 Generator following: Fuel oil pumps Ability to (a) predict the impacts of the following malfunctions or operations on the EDG system; and (b) based on those predictions, 064 / Emergency Diesel use procedures to correct, control, or mitigate X A2.08 2.7 23 Generator the consequences of those malfunctions or operations: Consequences of opening closing breaker between buses (VARS, out of phase, voltage)
NUREG-1021, Revision 9.1 Page 9 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Number K/A Topics Imp. Q#
Ability to predict and/or monitor changes in 073 / Process Radiation parameters (to prevent exceeding design X A1.01 3.2 24 Monitoring limits) associated with operating the PRM system controls including: Radiation levels Ability to manually operate and/or monitor in 076 / Service Water X A4.02 2.6 25 the control room: SWS valves Conduct of Operations: Ability to interpret 076 / Service Water X 2.1.25 reference materials, such as graphs, curves, 3.9 26 tables, etc.
Knowledge of the physical connections and/or 078 / Instrument Air X K1.02 cause-effect relationships between the IAS and 2.7 27 the following systems: Service air Ability to monitor automatic operation of the 103 / Containment X A3.01 containment system, including: Containment 3.9 28 isolation K/A Category Point Totals: 3 3 2 2 2 1 3 4/2 1 3 4/3 Group Point Total: 28 / 5 NUREG-1021, Revision 9.1 Page 10 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Number K/A Topics Imp. Q#
Conduct of Operations: Ability to interpret 045 / Main Turbine Generator X 2.1.20 4.6 91 and execute procedure steps Conduct of Operations: Ability to explain and 071 / Waste Gas Disposal X 2.1.32 4.0 92 apply system limits and precautions 034 / Fuel Handling Ability to monitor automatic operation of the X A3.02 3.1 93 Equipment Fuel Handling System, including: Load limits Knowledge of the operational implications of 002 / Reactor Coolant X K5.01 the following concepts as they apply to the 3.2 29 RCS: Basic heat transfer concepts Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b) based on those predictions, use 015 / Nuclear Instrumentation X A2.05 3.3 30 procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Core void formation Ability to manually operate and/or monitor in 016 / Non-nuclear X A4.01 the control room: NNI channel select 2.9 31 Instrumentation controls Knowledge of the effect that a loss or malfunction of the Containment Purge 029 / Containment Purge X K3.01 2.9 32 System will have on the following:
Containment parameters Knowledge of the effect of a loss or 035 / Steam Generator X K6.03 malfunction on the following will have on the 2.6 33 SGs: Steam generator level detector Ability to (a) predict the impacts of the following malfunctions or operations on the 041 / Steam Dump/Turbine SDS; and (b) based on those predictions, X A2.03 2.8 34 Bypass Control use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of IAS Ability to monitor automatic operation of the 055 / Condenser Air Removal X A3.03 CARS, including: Automatic diversion of 2.5 35 CARS exhaust NUREG-1021, Revision 9.1 Page 11 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 SONGS 2 & 3 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G Number K/A Topics Imp. Q#
Knowledge of the physical connections and/or cause-effect relationships between 056 / Condensate X K1.03 2.6 36 the Condensate System and the following systems: MFW Ability to predict and/or monitor changes in 072 / Area Radiation parameters (to prevent exceeding design X A1.01 3.4 37 Monitoring limits) associated with operating the ARM system controls including: Radiation levels Knowledge of the effect of a loss or malfunction of the following will have on the 086 / Fire Protection X K6.04 2.6 38 Fire Protection System: Fire, smoke, and heat detectors K/A Category Point Totals: 1 0 1 0 1 2 1 2 1/1 1 0 / 2 Group Point Total: 10 / 3 NUREG-1021, Revision 9.1 Page 12 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 Generic Knowledge and Abilities Outline - Tier 3 Form ES-401-3 Facility: SONGS 2& 3 Date of Exam: 10/19/2009 Category K/A # Topic RO SRO-Only IR # IR #
2.1.5 Ability to use procedures related to shift staffing, such as 3.9 94 minimum crew complement, over time limitations, etc.
2.1.23 Ability to perform specific system and integrated plant 4.4 95
- 1. procedures during all modes of plant operation Conduct 2.1.21 Ability to verify the controlled procedure copy 3.5 66 of Operations 2.1.15 Knowledge of administrative requirements for temporary 2.7 67 management directives, such as standing orders, night orders, Operations memos, etc 2.1.31 Ability to locate control room switches, controls, and 4.6 68 indications, and to determine that they correctly reflect the desired plant lineup Subtotal 3 2 2.2.6 Knowledge of the process for making changes to 3.6 96 procedures 2.2.25 Knowledge of the bases in Technical Specifications for 4.2 97 limiting conditions for operations and safety limits 2.
Equipment 2.2.22 Knowledge of limiting conditions for operations and safety 4.0 69 Control limits 2.2.39 Knowledge of less than or equal to one hour Technical 3.9 70 Specification action statements for systems 2.2.41 Ability to obtain and interpret station electrical and 3.5 71 mechanical drawings Subtotal 3 2 2.3.4 Knowledge of radiation exposure limits under normal or 3.7 98 emergency conditions 2.3.12 Knowledge of radiological safety principles pertaining to 3.2 72
- 3. licensed operator duties, such as containment entry Radiation requirements, fuel handling responsibilities, access to Control locked high radiation areas, aligning filters, etc.
2.3.4 Knowledge of radiation exposure limits under normal or 3.2 73 emergency conditions Subtotal 2 1 2.4.6 Knowledge of EOP mitigation strategies 4.7 99 2.4.9 Knowledge of low power/shutdown implications in 4.2 100 accident (e.g., loss of coolant accident or loss of residual
- 4. heat removal) mitigation strategies Emergency Procedures / 2.4.2 Knowledge of system setpoints, interlocks and automatic 4.5 74 Plan actions associated with EOP entry conditions 2.4.39 Knowledge of RO responsibilities in emergency plan 3.9 75 implementation Subtotal 2 2 Tier 3 Point Total 10 7 NUREG-1021, Revision 9.1 Page 13 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A Q #81 - This K/A does not meet the requirements of 10 CFR 55.43 for 1/1 011 G2.1.28 SRO-only questions. Randomly reselected 011 G2.2.42.
Q #25 - This specific K/A does not apply at SONGS. The Turbine Building Closed Cooling Water System is cooled by the Circulating Water System and 2/1 076 A1.02 the Reactor Building Closed Cooling Water System does not exist. Randomly reselected 076 A4.02.
Q #62 - This specific K/A does not apply at SONGS as there is no installed 1/2 067 AA2.10 long-term breathing air system. Randomly reselected 067 AA2.05.
Q #09 - This specific K/A does not apply at SONGS as there is no Post 2/1 008 G 2.4.3 Accident Monitoring Instrumentation associated with the Component Cooling Water System. Randomly reselected G 2.4.46.
Q #79 - Coverage of the Instrument Air System deemed adequate per 1/1 065 G 2.1.25 Questions #27 and #54. Randomly reselected 003 G 2.2.37.
Q #92 - Unable to develop an appropriate SRO Level question using the Incore Temperature Monitoring System with this generic K/A. Plant monitoring 2/2 017 G 2.4.34 outside the Control Room is limited to Thot and Tcold indications on the Emergency Plant Parameters Monitoring (EPPM) Panel. Randomly reselected 071 G 2.4.20.
Q #12 - Unable to develop a psychometrically sound question that 2/1 013 K4.19 discriminates at the appropriate license level. Reselected 013 K6.01 to improve the balance of exam outline.
Q #16 - Unable to develop a psychometrically sound question that 2/1 039 A4.04 discriminates at the appropriate license level. Reselected 039 A4.01 to improve the balance of the exam.
Q #37 - Unable to develop a psychometrically sound question that 2/2 072 G 2.4.6 discriminates at the RO level. Reselected 072 A1.01.
Q #47 - Unable to develop a psychometrically sound question that 1/1 029 G 2.2.38 discriminates at the RO level. Reselected 029 EA2.07.
Q #65 - Unable to develop a psychometrically sound question that 1/2 068 AA2.05 discriminates at the RO level. Reselected 068 AA1.28.
Q #67 - Unable to develop a psychometrically sound question that 3/1 G 2.1.14 discriminates at the RO level. Reselected G 2.1.15.
Q #83 - Unable to develop a psychometrically sound question that 1/2 068 G 2.4.18 discriminates at the SRO level. Reselected 068 G 2.4.41.
Q #92 - Unable to develop a psychometrically sound question that 2/2 071 G 2.4.20 discriminates at the SRO level. Reselected 071 G 2.1.32.
Q #98 - Unable to develop a psychometrically sound question that 3/3 G 2.3.14 discriminates at the SRO level. Reselected G 2.3.4.
NUREG-1021, Revision 9.1 Page 14 of 15 SONGS Oct 2009 NRC Rev 1
ES-401 Record of Rejected K/As Form ES-401-4 Q #87 - Coverage of the Containment Spray System deemed adequate per Questions #14, #15 and JPM S-5, Terminate Containment Spray which 2/1 026 A2.03 includes RNO actions due to failure of Engineered Safety Feature equipment.
Reselected 059 G 2.2.44.
NUREG-1021, Revision 9.1 Page 15 of 15 SONGS Oct 2009 NRC Rev 1
ES-301 Administrative Topics Outline Form ES-301-1 Facility: SONGS 2 & 3 Date of Examination: 10/19/09 Examination Level RO Operating Test Number: NRC Administrative Topic Type Code* Describe Activity to be Performed (see Note) 2.1.23 Ability to perform specific system and integrated plant procedures during all modes Conduct of Operations N, R of plant operation (4.3).
JPM: Perform an RCS Inventory Balance (New).
2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (3.9).
Conduct of Operations M, R JPM: Determine Time to Boil (J213A).
2.2.12 Knowledge of Surveillance Procedures (3.7).
Equipment Control N, R JPM: Perform Core Exit Thermocouple Channel Checks. (New) 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-Radiation Control M, S radiation areas, aligning filters, etc. (3.2).
JPM: Determine Dose for Maintenance Activities (J236A2).
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; for 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
NUREG-1021, Revision 9.1 1 of 2 SONGS Oct 2009 RO ES-301-1 Rev 1
Administrative Topics Outline Task Summary RO A.1.a The candidate will perform a Reactor Coolant System Inventory Balance per SO23-3-3.37, Reactor Coolant System Inventory Balance. The critical steps include correctly documenting parameters and performing calculations within allowable tolerances. This is a new JPM.
RO A.1.b The candidate will calculate Time-to-Boil per SO23-5-1.8.1, Shutdown Nuclear Safety, Attachment 9, Calculation of RCS Time-to-Boil Margin.
The critical steps include correctly interpreting curves within tolerances and performing the final calculation within given tolerances. This is a modified bank JPM.
RO A.2 The candidate will be provided with a set of Core Exit Thermocouple data and will determine if the required OPERABILITY is met using SO23-3-3.35, PAMI / Safe Shutdown Monthly Checks, Attachment 2, Core Exit Thermocouples and Heated Junction Thermocouple System Monthly Channel Checks. The critical steps include identifying any out-of-service thermocouples and correctly determining OPERABILITY of the Core Exit Thermocouple System. This is a new JPM.
RO A.3 The candidate will be required to calculate stay time based on a maintenance activity. The critical steps require determining the optimum total dose using either time, distance or shielding for performing the task.
This is a modified bank JPM.
RO A.4 N/A NUREG-1021, Revision 9.1 2 of 2 SONGS Oct 2009 RO ES-301-1 Rev 1
ES-301 Administrative Topics Outline Form ES-301-1 Facility: SONGS 2 & 3 Date of Examination: 10/19/09 Examination Level SRO Operating Test Number: NRC Administrative Topic Type Code* Describe Activity to be Performed (see Note) 2.1.23 Ability to perform specific system and integrated plant procedures during all modes Conduct of Operations M, R of plant operation (4.4).
JPM: Determine Azimuthal Power Tilt (J250A).
2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (4.2).
Conduct of Operations M, R JPM: Determine Time to Boil (J213A).
2.2.12 Knowledge of Surveillance Procedures (4.1).
Equipment Control N, R JPM: Review Core Exit Thermocouple Channel Check surveillance and verify Technical Specification Compliance. (New) 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling Radiation Control responsibilities, access to locked high-N, R radiation areas, aligning filters, etc. (3.7).
JPM: Determine Containment Access Requirements (New).
2.4.44 Knowledge of emergency plan protective action recommendations. (4.4).
Emergency Plan M, R JPM: Determine Protective Actions (J126A).
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; for 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
NUREG-1021, Revision 9.1 1 of 2 SONGS Oct 2009 SRO ES-301-1 Rev 1
Administrative Topics Outline Task Summary SRO A.1.a The candidate will perform an Azimuthal Power Tilt calculation per SO23-3-3.6, COLSS Out of Service Surveillance, attachment 3, Azimuthal Power Tilt Determination. The critical steps include correctly transposing data, accurately performing all calculations, correctly identifying out of tolerance conditions and identifying required actions. Additionally, a determination of actions for out-of-tolerance Azimuthal Tilt is required.
This is a modified bank JPM.
SRO A.1.b The candidate will calculate Time-to-Boil per SO23-5-1.8.1, Shutdown Nuclear Safety, Attachment 9, Calculation of RCS Time-to-Boil Margin.
The critical steps include correctly interpreting curves within tolerances and performing the final calculation within given tolerances. This is a modified bank JPM.
SRO A.2 The candidate will be provided with a set of Core Exit Thermocouple data and will determine if the required OPERABILITY is met using SO23-3-3.35, PAMI / Safe Shutdown Monthly Checks, Attachment 2, Core Exit Thermocouples and Heated Junction Thermocouple System Monthly Channel Checks. The critical steps include identifying any out-of-service thermocouples, correctly determining OPERABILITY, and recording entry into any required Technical Specification LCOs for the Core Exit Thermocouple System. This is a new JPM.
SRO A.3 The candidate will determine the requirements for Containment access per SO23-3-2.34, Containment Access Control, Inspections and Airlocks Operation. The critical steps include properly identifying all requirements on Attachment 1, Containment Access Requirements. This is a new JPM.
SRO A.4 The candidate will review given plant conditions and offsite dose information and determine required protective actions per SO123-VIII-10.3, Protective Action Recommendations. The critical steps include determining the affected areas and the recommended protective actions. This is a modified bank JPM.
NUREG-1021, Revision 9.1 2 of 2 SONGS Oct 2009 SRO ES-301-1 Rev 1
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Facility: SONGS Units 2 and 3 Date of Examination: 10/19/09 Exam Level: RO SRO(I) SRO (U) Operating Test No.: NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Code* Safety Function S-1 001 - Control Rod Drive System (New) A, N, S 1 Perform Immediate Actions for Control Room Evacuation S-2 004 - Chemical and Volume Control System (J083S) D, S 2 Secure Charging and Letdown S-3 006 - Emergency Core Cooling System (J073S) A, M, EN, S 3 Align Simultaneous Hot Leg and Cold Leg Injection S-4 003 - Reactor Coolant Pump System (J027FS) A, D, L, S 4-P Start a Reactor Coolant Pump S-5 022 - Containment Spray System (J049FS) A, D, EN, S 5 Terminate Containment Spray S-6 064 - Emergency Diesel Generator System (J054S) D, S 6 Restore 1E Bus 2A06 From Cross-Tie Operations C-7 073 - Process Radiation Monitoring System (J120S) (RO only) C, D 7 Reset and Restore Fuel Handling Isolation System C-8 029 - Containment Purge System (J147FS) A, C, D 8 Place Containment Mini-Purge in Service In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P-1 059 - Main Feedwater System (J109) D 4-S Locally Operate Main Feedwater Regulating Valve P-2 004 - Chemical and Volume Control System (New) E, N, R 2 Locally Align Charging Pump Suction to RWST P-3 012 - Reactor Protection System (J021F) E, M, R 7 Locally Open Reactor Trip Breakers (TIME CRITICAL)
NUREG-1021, Revision 9.1 1 of 3 SONGS Oct 2009 NRC ES-301-2 Rev 1
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature - / - / 1 (control room system)
(L)ow Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1/1/1 (S)imulator NRC JPM Examination Summary Description S-1 The candidate will perform the immediate operator actions for a Control Room Evacuation per Abnormal Operating Instruction SO23-13-02, Shutdown from Outside the Control Room. The alternate path is performed when a Reactor Coolant Pump breaker fails to open. This is a new JPM under the Control Rod Drive System - Reactivity Control safety function. This is a PRA significant action.
S-2 The candidate will secure Charging and Letdown as part of an RCS leak investigation per SO23-3-2.1.02, Chemical and Volume Control System Outage Evolutions. This is a bank JPM under the Chemical and Volume Control System -
Reactor Coolant Inventory Control safety function.
S-3 The candidate will align simultaneous Hot Leg and Cold Leg Injection during a Loss of Coolant Accident per SO23-12-11, EOI Supporting Attachments, Attachment 11, Simultaneous Hot / Cold Leg Injection. The alternate path requires the operator to perform actions with a High Pressure Safety Injection Pump out-of-service. This is a modified bank JPM under the Emergency Core Cooling System - Reactor Pressure Control safety function. This is a PRA significant action.
S-4 The candidate will start the fourth Reactor Coolant Pump during a Plant Startup per SO23-3-1.7, Reactor Coolant Pump Operation. The alternate path occurs when Component Cooling Water flow is lost to the RCP and the operator trips the Reactor Coolant Pump per the alarm response procedure. This is a bank JPM NUREG-1021, Revision 9.1 2 of 3 SONGS Oct 2009 NRC ES-301-2 Rev 1
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 under the Reactor Coolant Pump System - Heat Removal from Reactor Core safety function.
S-5 The candidate will be required to terminate Containment Spray per SO23-12-11, EOI Supporting Attachments, Attachment 2, Floating Steps. The alternate path occurs when there is only one Containment Emergency Cooling Unit operating and the actions of the RNO path are required. This is a bank JPM under the Containment Spray System - Containment Integrity safety function.
S-6 The candidate will be required to restore from 1E 4160V unit cross-tie operations per SO23-6-2, Transferring of 4 kV Buses, Section 6.9, Restoring from 1E 4kV Bus 3A06 to 2A06 Cross-Tie Operation. This is a bank JPM under the AC Electrical Distribution System - Electrical safety function.
C-7 The candidate will reset and re-establish normal Fuel Building Ventilation after isolation due to a high radiation signal per SO23-3-2.22, Engineered Safety Features Actuation Systems Operation, Attachment 23, FHIS Reset and Restoration. This is a bank JPM under the Area Radiation Monitoring System -
Instrumentation safety function.
C-8 The candidate will place the Containment Mini-Purge System in operation to support a Containment entry per SO23-1-4.2, Containment Purge and Recirculation Filtration System. The alternate path occurs when a Containment Radiation High alarm is received and the operator isolates mini-purge per the alarm response. This is a bank JPM under the Containment Purge System - Plant Service Systems safety function.
P-1 The candidate will perform the local actions to operate a Main Feedwater Regulating Valve per SO23-9-6, Feedwater Control System Operation, Section 6.4, Local-Manual Operation of Main Feedwater Control Valves. This is a bank JPM under the Main Feedwater System - Secondary System Heat Removal from Reactor Core safety function.
P-2 The candidate will perform the actions to align Charging Pump suction to the Refueling Water Storage Tank per SO23-13-2, Shutdown from Outside the Control Room, Attachments 10 and 11. This is a new JPM under the Chemical and Volume Control System - Reactor Coolant System Inventory Control safety function. This is a PRA significant action.
P-3 The candidate will log into and enter the Radiation Controlled Area and locally open Reactor Trip breakers per SO23-12-1, Standard Post Trip Actions. This is a time critical, modified bank JPM under the Reactor Protection System -
Instrumentation safety function. This is a PRA significant action.
NUREG-1021, Revision 9.1 3 of 3 SONGS Oct 2009 NRC ES-301-2 Rev 1
Appendix D Scenario Outline Form ES-D-1 Facility: SONGS 2 and 3 Scenario No.: 1 Op Test No.: October 2009 NRC Examiners: Operators:
Initial Conditions:
- Train A Component Cooling Water Pump (P-025) in service.
- Train A Low Pressure Safety Injection Pump (P-015) OOS for oil change.
- Train A Emergency Diesel Generator (G-002) OOS for governor repair.
- Fire Computer is OOS.
Turnover: Maintain steady-state power conditions.
Critical Tasks:
- Trip the Reactor following multiple CEA drops.
- Trip any Reactor Coolant Pump not satisfying operating limits (Loss of CCW).
- Establish minimum design Safety Injection flow rate (SIAS component failure).
Event No. Malf. No. Event Type* Event Description 1 CH04D TS (CRS) Containment Wide Range Pressure Transmitter (PT-0352-4) fails
+10 min high.
2 FC05B I (BOP, CRS) Steam Generator (E-088) Main Feedwater Master Controller
+20 min setpoint fails to 50% level on 60 second ramp.
3 RD4403 C (RO, BOP, CRS) Dropped CEA #44.
+30 min TS (CRS) 4 R (RO) Power reduction for dropped CEA.
+45 min N (BOP, CRS) 5 RD0303 C (RO, CRS) Dropped CEA #3. Manual Reactor trip required.
+45 min 6 RC01A M (RO, BOP, CRS) Large Break Loss of Coolant Accident upon Unit trip.
+50 min 7 Bus 2A07 I (BOP) Non-1E 4160 Volt Bus 2A07 auto transfer failure upon Unit trip.
+50 min XFR LP 8 RP01E C (RO) Low Pressure Safety Injection Pump (P-016) fails to auto start.
+55 min
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications SONGS Oct 2009 NRC Sim Scenario #1 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
NRC #1 The crew will assume the watch and maintain steady-state conditions per Operating Instruction (OI)
SO23-5-1.7, Power Operations.
The first event is a Containment Wide Range Pressure Transmitter failure. The crew will respond per Abnormal Operating Instruction (AOI) SO23-13-18, Reactor Protection System Failure/Loss of Vital Bus and Operating Instruction (OI) SO23-3-2.12, Reactor Protective System Operation. The CRS will evaluate Technical Specifications.
When Technical Specification actions are complete, Steam Generator E-088 Master Controller Setpoint fails to 50%. Entry into AOI SO23-13-24, Feedwater Control System Malfunction is required. The CRS will analyze the cause of the failure using Attachment 1 of SO23-13-24. Steam Generator level control is restored by placing the Master Controller in Manual. When Steam Generator level control is achieved the controller will be returned to automatic operation.
When level control is regained, Control Element Assembly #44 will drop into the core. Crew actions are per AOI SO23-13-13, Misaligned or Immovable Control Element Assembly and include a power reduction as required per procedure. The crew will restore RCS Cold Leg temperature per OI SO23-5-1.7, Power Operations and block any further load changes and then continue with a power reduction as required per SO23-13-13. The CRS will evaluate Technical Specifications.
When the crew commences recovery of CEA #44, a second Control Element Assembly will drop into the core necessitating a manual Reactor trip.
When the Reactor is tripped, a Large Break Loss of Coolant Accident will occur. The crew will enter Emergency Operating Instruction (EOI) SO23-12-1, Standard Post Trip Actions and then transition to EOI SO23-12-3, Loss of Coolant Accident. Post trip events include a Non-1E 4160 Volt Bus that fails to transfer as well as a Low Pressure Safety Injection Pump start failure. Both of these failures require actions on the part of the Reactor and Balance of Plant Operators.
The scenario is terminated when conditions for reactor coolant system cooldown is reached.
Risk Significance:
- Risk important components out of service: LPSI Pump (P-015)
Emergency Diesel Generator (G-002)
- Failure of risk important system prior to trip: Dropped Control Element Assembly
- Risk significant core damage sequence: Large Break LOCA
- Risk significant operator actions: Trip Reactor Following Multiple CEA Drop Trip RCPs Due to Loss of CCW Establish Minimum Safety Injection Flowrate SONGS Oct 2009 NRC Sim Scenario #1 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 Facility: SONGS 2 and 3 Scenario No.: 2 Op Test No.: October 2009 NRC Examiners: Operators:
Initial Conditions:
- Train A Component Cooling Water Pump (P-025) in service.
- Train A Low Pressure Safety Injection Pump (P-015) OOS for oil change.
- Train A Emergency Diesel Generator (G-002) OOS for governor repair.
- Fire Computer is OOS.
Turnover: Maintain steady-state power conditions.
Critical Tasks:
- Restore Component Cooling Water flow to the Non-Critical Loop.
- Restore power to at least one 1E 4 kV Bus (Station Blackout).
- Restore Feedwater flow to at least one Steam Generator (Loss of Feedwater).
Event No. Malf. No. Event Type* Event Description 1 ED11 TS (CRS) Loss of Control Room Annunciators.
+10 min 2 FW08B C (BOP, CRS) Main Feedwater Pump Turbine (P-063) loss of oil pressure.
+15 min 3 R (RO) Rapid Power Reduction to 70% for loss of one Main Feed Pump.
+35 min N (BOP, CRS) 4 ED03A C (RO, BOP, CRS) Loss of 1E 4160 Volt Bus 2A04.
+50 min TS (CRS) 5 TU08 M (RO, BOP, CRS) Loss of Offsite Power.
+50 min PG24 6 EG08B M (RO, BOP, CRS) Emergency Diesel Generator (G-003) fails to start.
+50 min Station Blackout.
7 FW25 C (BOP) Turbine Driven Auxiliary Feedwater Pump (P-140) trips on
+55 min overspeed (300 seconds post-trip). Loss of Feedwater.
8 CVCS LP I (RO) Boric Acid Makeup Tank Gravity Feed Valves fail to open during
+ min boration.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications SONGS Oct 2009 NRC Sim Scenario #2 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
NRC #2 The crew will assume the watch and maintain steady-state conditions per Operating Instruction (OI)
SO23-5-1.7, Power Operations.
The first event is a Loss of Control Room Annunciators. The crew will respond per Abnormal Operating Instruction (AOI) SO23-13-22, Loss of Control Room Annunciators. Actions include aligning an alternate power supply to the Control Room Annunciators. The CRS will evaluate Technical Specifications.
When the annunciators are restored, a loss of oil pressure to Main Feedwater Pump P-063 will occur.
The crew will respond per the Annunciator Response Procedures (ARP) and AOI SO23-13-24, Feedwater Control System Malfunction and determine that a Main Feedwater Pump trip is warranted.
This will necessitate entry into AOI SO23-13-28, Rapid Power Reduction in order to low power below the threshold for one (1) Main Feedwater Pump operation.
When plant conditions are stable, a loss of 1E 4160 Volt Bus 2A04 will occur. The crew will enter him the AOI SO23-13-26, Loss of Power to an AC Bus. Crew actions include placing a Charging Pump in service as well as transferring to the Train B Component Cooling Water System. The CRS will evaluate Technical Specifications.
The major event is a Loss of Offsite Power that requires entry into Emergency Operating Instruction (EOI) SO23-12-1, Standard Post Trip Actions. During performance of the Standard Post Trip Actions, the Train B Emergency Diesel Generator fails to start. Additionally, five (5) minutes post-trip the Turbine Driven Auxiliary Feedwater Pump will trip on overspeed. With a Station Blackout and Loss of Feedwater, the CRS will be required to enter Functional Recovery Procedure (FRP) SO23-12-9, Functional Recovery. Boric Acid Makeup Tank Gravity Feed Valves fail to open during boration and the Reactor Operator will be required to manually align the Refueling Water Storage Tank.
With Switchyard power unavailable, restoration of Unit 2 1E 4160 Volt Bus 2A06 will be via a crosstie with Unit 3 1E 4160 Volt Bus 3A06. Once the crosstie is successful, Auxiliary Feedwater Pump P-504 can be started to restore feedwater flow to Steam Generator E-088.
The scenario is terminated when the 1E Bus is reenergized, feedwater flow is restored, and boration via the RWST is commenced.
Risk Significance:
- Risk important components out of service: LPSI Pump (P-015)
Emergency Diesel Generator (G-002)
- Failure of risk important system prior to trip: Loss of 4160 V Bus 2A04
- Risk significant core damage sequence: Station Blackout with Loss of Feedwater
- Risk significant operator actions: Restore Flow to Non-Critical Loop Crosstie Bus 3A06 with Bus 2A06 Align Feedwater Flow to a Steam Generator SONGS Oct 2009 NRC Sim Scenario #2 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 Facility: SONGS 2 and 3 Scenario No.: 3 Op Test No.: October 2009 NRC Examiners: Operators:
Initial Conditions:
- Train A Component Cooling Water Pump (P-025) in service.
- Train A Low Pressure Safety Injection Pump (P-015) OOS for oil change.
- Fire Computer is OOS.
Turnover: Dilution and power ascension in progress at 10% per hour.
Critical Tasks:
- Manually Initiate Main Steam Isolation Signal (Auto Actuation failure).
Event No. Malf. No. Event Type* Event Description 1 R (RO) Dilution and power ascension at 10% per hour.
+15 min N (BOP, CRS) 2 RC09A TS (CRS) Reactor Coolant Pump (P-002) Speed Sensor (SE-0143-1) failure.
+25 min 3 CV16B I (RO, CRS) Volume Control Tank Level Transmitter (LT-0227) fails low.
+35 min 4 ED06D C (RO, BOP, CRS) Overcurrent trip of Feeder Breaker to 1E 480 Volt Bus 2B04.
+50 min TS (CRS) 5 FW LP C (BOP, CRS) Main Feedwater Pump trip.
+55 min RX LP Primary Side Reactor Trip pushbuttons disabled.
6 MS04B M (RO, BOP, CRS) Excess Steam Demand Event downstream of Main Steam Isolation
+60 min Valves (300 second ramp).
7 RD1402 C (RO) Two CEAs fail to insert on the trip. Emergency boration via gravity
+60 min RD1502 feed due to loss of 1E Bus 2B04.
8 MSIS LP I (BOP) Main Steam Isolation Signal fails to actuate; manual actuation
+65 min required.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications SONGS Oct 2009 NRC Sim Scenario #3 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
NRC #3 The crew will assume the watch and resume a dilution and power ascension per Operating Instruction (OI) SO23-5-1.7, Power Operations at 10% per hour.
The next event is a Reactor Coolant Pump speed sensor failure. The crew will respond per the Annunciator Response Procedures (ARP) and place the appropriate Reactor Protection System trips in Bypass per Abnormal Operating Instruction (AOI) SO23-13-18, Reactor Protection System Failure/Loss of Vital Bus and Operating Instruction (OI) SO23-3-2.12, Reactor Protective System Operation. The CRS will evaluate Technical Specifications.
When Technical Specifications have been addressed, a Volume Control Tank (VCT) level transmitter fails low and transfers Charging Pump suction to the Refueling Water Storage Tank. The crew will align LV-0227B, VCT Outlet Valve and refer to Operating Instruction (OI) SO23-3-2.1, CVCS Operations and/or SO23-3-2.2, Makeup Operations.
The next event is the overcurrent trip of the feeder breaker to 1E 480 Volt Bus 2B04. The crew will respond per AOI SO23-13-26, Loss of Power to an AC Bus. Crew actions include restoring Charging flow as required and placing the Train A Emergency Diesel Generator in Maintenance Lockout. The CRS will evaluate Technical Specifications.
During the next event, the running of Main Feedwater Pump will trip. The crew will determine that a Reactor and Turbine trip are required. The Primary Side Reactor Trip pushbuttons are disabled and the BOP will trip the Reactor.
When the Reactor and Turbine are tripped, an Excess Steam Demand Event (ESDE) downstream of the Main Steam Isolation Valves will occur. The crew should determine that an ESDE is occurring and manually trip the Reactor and Turbine. Entry into Emergency Operating Instruction (EOI) SO23-12-1, Standard Post Trip Actions is required. A transition will then be made to EOI SO23-12-5, Excess Steam Demand Event. Two (2) Control Element Assemblies will fail to insert on the trip and an emergency boration using the Gravity Feed Valves is required. Additionally, a manual Main Steam Isolation Signal must be generated.
This scenario is terminated when the Main Steam Isolation Valves are closed and Reactor Coolant System temperature control is restored using the Atmospheric Dump Valves.
Risk Significance:
- Risk important components out of service: Low Pressure Safety Injection Pump (P-015)
- Failure of risk important system prior to trip: Loss of 1E 480 V Bus 2B04
- Risk significant core damage sequence: ESDE with MSIS failure
- Risk significant operator actions: Initiate Emergency Boration Manually Initiate MSIS Establish RCS Temperature Control SONGS Oct 2009 NRC Sim Scenario #3 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 Facility: SONGS 2 and 3 Scenario No.: 4 Op Test No.: October 2009 NRC Examiners: Operators:
Initial Conditions:
- Train B Component Cooling Water Pump (P-026) in service.
- Channel X Pressurizer Level and Pressure Control in service.
- Fire Computer is OOS.
Turnover: Power increase in progress to ~ 2% power.
Critical Tasks:
- Restore Component Cooling Water flow due to Train B leakage.
- Reduce Reactor Coolant System Thot to less than 530°F (SGTR).
- Isolate the most affected Steam Generator (SGTR).
Event No. Malf. No. Event Type* Event Description 1 R (RO) Rod withdrawal and power increase in progress to ~2% power.
+15 min N (BOP, CRS) 2 CVCS LP I (RO, CRS) Letdown Heat Exchanger Outlet Temperature Instrument
+25 min (TI-0224) fails high. TV-0224A and TV-0224B fail to reposition.
3 CC05B C (BOP, CRS) Train B Component Cooling Water Heat Exchanger (E-002) tube
+35 min TS (CRS) leak.
4 RX08 C (BOP, CRS) Steam Bypass Control Valves close. Transfer SBCS Master
+45 min Controller (PIC-8431) to Local-Manual to open valves.
5 SG06B C (RO, CRS) Steam Generator Tube Leak (E-089) at 50 gpm.
+50 min TS (CRS) 6 SG06B M (RO, BOP, CRS) Steam Generator Tube Rupture (E-089) at 300 gpm upon Unit trip.
+50 min 7 CCAS LP I (RO) Containment Cooling Actuation Signal fails to actuate.
+60 min
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Technical Specifications SONGS Oct 2009 NRC Sim Scenario #4 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
NRC #4 The crew will assume the watch with the Reactor critical at ~2x10-3% power. The crew will raise power using CEA withdrawal per Operating Instruction (OI) SO23-5-1.3.1, Plant Startup from Hot Standby to Minimum Load.
The next event is a high failure of a Letdown temperature instrument. The crew will respond per the Annunciator Response Procedures (ARP) and perform actions to isolate the Boronometer and bypass the Ion Exchanger Demineralizers.
When Letdown conditions are normal, a tube leak will develop on the Train B Component Cooling Water Heat Exchanger. The crew will respond per Abnormal Operating Instruction (AOI) SO23-13-7, Loss of Component Cooling Water/Salt Water Cooling. Crew actions include transferring to the Train A Component Cooling Water System as well is attempting to isolate Train B leakage. The CRS will evaluate Technical Specifications.
When CCW flow is restored, the Steam Bypass Control System (SBCS) Valves will close. Crew actions are per OI SO23-3-2.18, Steam Bypass System Operation and include transferring the SBCS Master Controller to Local-Manual control or operating individual SBCS Valves and restoring Reactor Coolant System temperature and Reactor power level to normal.
When plant conditions are stable, a Steam Generator tube leak will occur. The crew will enter AOI SO23-13-4, RCS Leak and take actions to minimize tube leakage. The size of the leak will require an immediate plant trip and at that time the leak will escalate to a rupture. The crew will enter Emergency Operating Instruction (EOI) SO23-12-1, Standard Post Trip Actions and transition to EOI SO23-12-5, Steam Generator Tube Rupture.
Following the Safety Injection Actuation Signal, the Containment Cooling Actuation Signal fails to actuate requiring manual actions by the crew.
The event is terminated when the affected Steam Generator is cooled down and the Reactor Coolant System is depressurized.
Risk Significance:
- Risk important components out of service: None
- Failure of risk important system prior to trip: Loss of Train B Component Cooling Water
- Risk significant operator actions: Restore Non-Critical CCW Loop flow Isolate Ruptured Steam Generator Cooldown and Depressurize RCS SONGS Oct 2009 NRC Sim Scenario #4 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 Facility: SONGS 2 and 3 Scenario No.: 5 Op Test No.: October 2009 NRC Examiners: Operators:
Initial Conditions:
- Train B Component Cooling Water Pump (P-026) in service.
- Channel Y Pressurizer Pressure and Level Control in service.
- Control Element Assembly Calculator #2 in service.
- Vacuum Pump (P-054) OOS for hogging valve repair.
- Fire Computer is OOS.
Turnover: Maintain steady-state plant conditions.
Critical Tasks:
- Manually Trip the Reactor (Reactor Protection System failure).
- Perform High Pressure Safety Injection Throttle/Stop actions (LOCA).
Event No. Malf. No. Event Type* Event Description 1 SC01D C (BOP, CRS) Salt Water Cooling Pump (P-114) trips on seized shaft.
+10 min TS (CRS) 2 RC15B I (RO, CRS) Pressurizer Pressure Control Channel (PT-0100Y) fails low.
+20 min 3 RP18 I (RO, CRS) Control Element Assembly Calculator #2 failure.
+30 min TS (CRS) 4 TU08 M (RO, BOP, CRS) Inadvertent Main Turbine trip.
+35 min 5 RP15 I (RO) Reactor fails to automatically trip.
+35 min 6 TC02E C (BOP) High Pressure Turbine Stop Valve (HV-2200E) fails to close.
+35 min 7 RC18A M (RO, BOP, CRS) Pressurizer Safety Valve (PSV-0200) fails open on the trip and
+45 min reseats after Safety Injection is actuated.
8 FW23 C (BOP) Loss of Condenser vacuum.
+45 min
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Technical Specifications SONGS Oct 2009 NRC Sim Scenario #5 Rev 0.doc
Appendix D Scenario Outline Form ES-D-1 SCENARIO
SUMMARY
NRC #5 The crew will assume the watch and maintain steady-state conditions per Operating Instruction (OI)
SO23-5-1.7, Power Operations.
The first event is a loss of Saltwater Cooling Pump P-114. The crew will respond per Abnormal Operating Instruction (AOI) SO23-13-7, Loss of Component Cooling Water/Saltwater Cooling System by starting Train A Component Cooling Water System. Additionally, the Non-Critical Loop and Letdown Heat Exchanger must also be transferred to Train A. The CRS will evaluate Technical Specifications.
The next event is a low failure of Pressurizer Pressure Channel Y. Operator actions are per AOI SO23-13-27, Pressurizer Pressure and Level Malfunction. The crew will transfer to Channel X and restore Pressurizer pressure and heater functions.
Once plant conditions are stable, Control Element Assembly Calculator (CEAC) #2 will fail. The crew will respond per the Annunciator Response Procedures (ARP) and OI SO23-3-2.13, Core Protection /
Control Element Assembly Calculator Operation and transfer to CEAC #1 to restore Rod Position Indication. The CRS will evaluate Technical Specifications.
The major event is initiated by an inadvertent Main Turbine trip. Upon Main Turbine trip, a Pressurizer Safety Valve will open and remain open until Safety Injection actuates and close when the Safety Injection Actuation Signal is received. The crew will enter Emergency Operating Instruction (EOI)
SO23-12-1, Standard Post Trip Actions. The CRS will determine that entry into EOI SO23-12-2, Reactor Trip Recovery is warranted because Pressurizer pressure and level are rising with no indications of radiation or leakage into Containment. In the event the opening and closing of the Pressurizer Safety Valve is not observed, the crew may enter EOI SO23-12-3, Loss of Coolant Accident.
The inadvertent Main Turbine trip is complicated by the Reactor failing to trip and a High Pressure Turbine Stop Valve remaining open. Additionally, a loss of Condenser vacuum occurs necessitating Reactor Coolant System temperature control using the Atmospheric Dump Valves.
The event is terminated when actions for High Pressure Safety Injection Throttle/Stop are performed per EOI SO23-12-2 or EOI SO23-12-11, EOI Supporting Attachments in order to prevent overfilling of the Pressurizer.
Risk Significance:
- Risk important components out of service: None
- Failure of risk important system prior to trip: Loss of Train B Component Cooling Water
- Risk significant core damage sequence: Inadvertent Turbine trip without Reactor trip