ML16336A575
ML16336A575 | |
Person / Time | |
---|---|
Site: | Diablo Canyon ![]() |
Issue date: | 10/20/2016 |
From: | Vincent Gaddy Operations Branch IV |
To: | |
References | |
Download: ML16336A575 (42) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Facility: Date of Exam:
RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total
- 1. 1 3 3 3 3 3 3 18 6 Emergency &
Abnormal 2 1 2 2 N/A 1 2 N/A 1 9 4 Plant Evolutions Tier Totals 4 5 5 4 5 4 27 10 1 2 3 2 2 3 2 3 3 2 3 3 28 5 2.
Plant 2 1 1 1 1 0 1 1 1 1 1 1 10 3 Systems Tier Totals 3 4 3 3 3 3 4 4 3 4 4 38 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As
ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000007 Reactor Trip - Stabilization - X EK1.05 Knowledge of the operational implications 3.3 39 Recovery / 1 of decay power as a function of time as they apply to the reactor trip. (CFR 41.8 / 41.10 / 45.3) 000008 Pressurizer Vapor Space X AA1.02 Ability to operate and / or HPI pump to 4.1 40 Accident / 3 control PZR level/pressure as they apply to the Pressurizer Vapor Space Accident. (CFR 41.7 /
45.5 / 45.6) 000009 Small Break LOCA / 3 X EA2.38 Ability to determine or interpret the 3.9 41 Existence of head bubble as they apply to a small break LOCA: (CFR 43.5 / 45.13) 000011 Large Break LOCA / 3 X EA2.13 Ability to determine or interpret the 3.7* 42 difference between overcooling and LOCA indications as they apply to a Large Break LOCA.
(CFR 43.5 / 45.13) 000015/17 RCP Malfunctions / 4 X AK3.01 Knowledge of the reason for the potential 2.5* 43 damage from high winding and/or bearing temperatures as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow).
(CFR 41.5, 41.10/45.6/45.13) 000022 Loss of Rx Coolant Makeup / 2 X AK1.03 Knowledge of the operational implications 3.0 44 of the relationship between charging flow and PZR level as they apply to Loss of Reactor Coolant Makeup. (CFR 41.8 / 41.10 / 45.3) 000025 Loss of RHR System / 4 X AA1.23 Ability to operate and / or monitor RHR 2.8 45 heat exchangers as they apply to the Loss of Residual Heat Removal System: (CFR 41.7 / 45.5
/ 45.6) 000026 Loss of Component Cooling X G2.2.39 Knowledge of less than or equal to one 46 Water / 8 hour Technical Specification action statements for systems. (CFR: 41.7 / 41.10 / 43.2 / 45.13)
G2.4.11 - Knowledge of abnormal condition 4.0 procedures. (IR 4.0) 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 X EK2.06 Knowledge of the interrelations between 2.9* 47 breakers, relays, and disconnects following an ATWS. (CFR 41.7 / 45.7) 000038 Steam Gen. Tube Rupture / 3 X EK1.02 Knowledge of the operational implications 3.2 48 of leak rate vs. pressure drop as they apply to the SGTR. (CFR 41.8 / 41.10 / 45.3) 000040 (W/E12) Steam Line Rupture - X G2.4.49 Ability to perform without reference to 4.6 49 Excessive Heat Transfer / 4 procedures those actions that require immediate operation of system components and controls.
(CFR: 41.10 / 43.2 / 45.6)
G2.4.2 - Knowledge of system set points, 4.5 interlocks and automatic actions associated with EOP entry conditions.
000054 Loss of Main Feedwater / 4 X AK3.03 Knowledge of the reasons for manual 3.8 50 control of AFW flow control valves as they apply to the Loss of Main Feedwater (MFW). (CFR 41.5,41.10 / 45.6 / 45.13) 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 X AA2.09 Ability to determine and interpret the 2.7 51 operational status of reactor building cooling unit as they apply to the Loss of Offsite Power:
Replaced with AA2.21 - Ability to determine and interpret the following as they apply to the Loss of 3.6 Offsite Power: EDG voltage and frequency indicators 000057 Loss of Vital AC Inst. Bus / 6 X G2.4.3 Ability to identify post-accident 3.7 52 instrumentation. (CFR: 41.6 / 45.4) 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 X AK3.04 Knowledge of the reasons for the effect 3.5 53 on Nuclear Service water discharge flow header of a loss of CCW as they apply to the Loss of Nuclear Service Water. (CFR 41.4, 41.8 / 45.7 )
AK3.03 - Guidance actions contained in EOP for 4.0 Loss of nuclear service water (ASW) 000065 Loss of Instrument Air / 8 X AA1.05 Ability to operate and / or monitor the 3.3* 54 following as they apply to the Loss of Instrument Air. (CFR 41.7 / 45.5 / 45.6) RPS W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 W/E05 Inadequate Heat Transfer - Loss X EK2.1 Knowledge of the interrelations between 3.7 55 of Secondary Heat Sink / 4 the (Loss of Secondary Heat Sink) and Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. (CFR: 41.7 / 45.7) 000077 Generator Voltage and Electric X AK2.05 Knowledge of the interrelations between 3.1 56 Grid Disturbances / 6 Generator Voltage and Electric Grid Disturbances and Pumps. (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18/6
ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 X AK3.02 Knowledge of the reasons for the 2.9 57 relationship between PZR pressure increase and reactor makeup/letdown imbalance as they apply to the Pressurizer level Control Malfunctions. (CFR 41.5, 41.10/45.6/45.13) 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 Fuel Handling Accident / 8 X AA2.01 AK2.01Knowledge of the 2.9 58 interrelations between the Fuel Handling Incidents and fuel handling equipment.
(CFR 41.7 / 45.7)
Ability to determine and interpret the following as they apply to the Fuel 3.2 Handling Incidents: ARM system indications 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 X AA1.02 Ability to operate and / or monitor 3.3 59 the ARM system as they apply to the Accidental Liquid Radwaste Release.
NOTE: ARM should be PRM as rad monitor for liquid radwaste is a process, not area monitor at DCPP (CFR 41.7 /
45.5 / 45.6) 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 Control Room Evac. / 8 000069 Loss of CTMT Integrity / 5 X G2.1.7 Ability to evaluate plant 4.4 60 performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5
/ 45.12 / 45.13) 000074 Inad. Core Cooling / 4 X EK3.08 Knowledge of the reasons for the 4.1 61 Securing RCPs as they apply to the Inadequate Core Cooling. (CFR 41.5 /
41.10 / 45.6 / 45.13) 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 X EK1.2 Knowledge of the operational 3.4 62 implications of the Normal, abnormal and emergency operating procedures associated with (SI Termination).as they apply to the (SI Termination) (CFR: 41.8 /
41.10, 45.3)
W/E03 LOCA Cooldown - Depress. / 4 X EK2.1 EK2.2 - Knowledge of the 3.7 63 interrelations between the (LOCA Cooldown and Depressurization) and the facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility..
(CFR: 41.7 / 45.7) (KA statement/importance for EK2.2 not EK2.1)
W/E13 Steam Generator Over-pressure / 4 X EA2.1 Ability to determine and interpret the 2.9 64 facility conditions and selection of appropriate procedures during abnormal and emergency operations as they apply to the (Steam Generator Overpressure)
(CFR: 43.5 / 45.13)
W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 X EK2.1 Knowledge of the interrelations 3.0 65 between the (High Containment Radiation) and the Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. (CFR: 41.7 / 45.7)
K/A Category Point Totals: 1 2 2 1 2 1 Group Point Total: 9/4
ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump X A1.08 Ability to predict and/or monitor 2.5 1 changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including Seal water temperature.
(CFR: 41.5 / 45.5)
G2.2.4 Ability to explain the variations 003 Reactor Coolant Pump X in control board/control room l layouts, 3.6 systems, instrumentation, and procedural actions between units at a facility. (CFR: 41.6 / 41.7 / 41.10 /
45.1 / 45.13)
G2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the 3.1 2 status of limiting conditions for operations. (CFR: 41.10 / 43.2 /
45.13) 004 Chemical and Volume X A1.09 Ability to predict and/or monitor 3.6 3 Control changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including RCS pressure and temperature. (CFR: 41.5 / 45.5) 005 Residual Heat Removal X K5.05 Knowledge of the operational 2.7* 4 implications of the plant response during "solid plant": pressure change due to the relative incompressibility of water as they apply the RHRS. (CFR:
41.5 / 45.7) 006 Emergency Core Cooling X K4.14 Knowledge of ECCS design 3.9 5 feature(s) and/or interlock(s) which provide for Cross-Connection of HPI/LPI/SIP. (CFR: 41.7) 007 Pressurizer Relief/Quench X K5.02 Knowledge of the operational 3.1 6 Tank implications of the method of forming a steam bubble in the PZR as they apply to PRTS. (CFR: 41.5 / 45.7) 008 Component Cooling Water X G2.2.25 Knowledge of the bases in 3.2 7 Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2)
G2.1.28 Knowledge of the purpose 4.1 and function of major system components and controls. (CFR: 41.7) 008 Component Cooling Water X K3.03 Knowledge of the effect that a 4.1 8 loss or malfunction of the CCWS will have on the RCP. (CFR 41.7) 010 Pressurizer Pressure Control X A4.03 Ability to manually operate 4.0 9 and/or monitor PORV and block valves in the control room. (CFR: 41.7
/ 45.5 to 45.8)
012 Reactor Protection X K2.01 Knowledge of bus power 3.3 10 supplies to the RPS channels, components, and interconnections.
(CFR: 41.7) 012 Reactor Protection X K6.03 Knowledge of the effect of a 3.1 11 loss or malfunction of the trip logic circuits will have on the RPS. (CFR:
41.7 / 45/7) 013 Engineered Safety Features X K2.01 Knowledge of bus power 3.6* 12 Actuation supplies to ESFAS/safeguards equipment control. (CFR 41.7) 013 Engineered Safety Features X K5.02 Knowledge of the operational 2.9 13 Actuation implications of safety system logic and reliability as they apply to the ESFAS.
(CFR: 41.5 / 45.7) 022 Containment Cooling X A3.01 Ability to monitor automatic 4.1 14 operation of the CCS, including initiation of safeguards mode of operation. (CFR: 41.7 / 45.5) 026 Containment Spray X K1.01 Knowledge of the physical 4.2 15 connections and/or cause effect relationships between the CSS and ECCS. (CFR: 41.2 to 41.9 / 45.7 to 45.8) 039 Main and Reheat Steam X A3.02 Ability to monitor automatic 3.1 16 operation of the MRSS, including isolation of the MRSS. (CFR: 41.5 /
45.5) 039 Main and Reheat Steam X K4.06 Knowledge of MRSS design 3.3 17 feature(s) and/or interlock(s) which prevent reverse steam flow on steam line break. (CFR: 41.7) 059 Main Feedwater X K3.02 Knowledge of the effect that a 3.6 18 loss or malfunction of the MFW will have on the AFW system. (CFR: 41.7
/ 45.6) 061 Auxiliary/Emergency X K6.02 Knowledge of the effect of a 2.6 19 Feedwater loss or malfunction of Pumps will have on the AFW components. (CFR: 41.7 /
45.7) 062 AC Electrical Distribution X A1.01 Ability to predict and/or monitor 3.4 20 changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including the significance of D/G load limits. (CFR:
41.5 / 45.5) 062 AC Electrical Distribution X A4.03 Ability to manually operate 2.8 21 and/or monitor synchro scope, including an understanding of running and incoming voltages in the control room. (CFR: 41.7 / 45.5 / to 45.8)
063 DC Electrical Distribution X G2.4.1 Knowledge of EOP entry 4.6 22 conditions and immediate action steps. (CFR: 41.10 / 43.5 / 45.13)
G2.4.6Knowledge of EOP mitigation 3.7 strategies. (CFR: 41.10 / 43.5 /
45.13) 064 Emergency Diesel Generator X K1.02 Knowledge of the physical 3.1 23 connections and/or cause effect relationships between the ED/G system and the D/G cooling water system. (CFR: 41.2 to 41.9 / 45.7 to 45.8) 073 Process Radiation Monitoring X A2.01 Ability to (a) predict the impacts 2.5 24 of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply. (CFR: 41.5 / 43.5
/ 45.3 / 45.13) 076 Service Water X K2.08 Knowledge of bus power 3.1* 25 supplies to ESF-actuated MOVs.
(CFR: 41.7)
K2.01 Knowledge of bus power 2.7 supplies to Service Water.
078 Instrument Air X A4.01 Ability to manually operate 3.1 26 and/or monitor pressure guages in the control room. (CFR: 41.7 / 45.5 to 45.8) 103 Containment X A2.03 Ability to (a) predict the impacts 3.5* 27 of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation. (CFR: 41.5 / 43.5 /
45.3 / 45.13) 103 Containment X A2.05 Ability to (a) predict the impacts 2.9 28 of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Emergency Containment entry. (CFR:
41.5 / 43.5 / 45.3 / 45.13)
K/A Category Point Totals: 2 3 2 2 3 2 3 3 2 3 3 Group Point Total: 28/5
ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant X K6.06 Knowledge of the effect or a loss 2.5 29 or malfunction on Sensors and Detectors.
(CFR: 41.7 / 45.7) 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation X K2.01 Knowledge of bus power supplies 3.3 30 to NIS channels, components, and interconnections. (CFR: 41.7) 016 Non-Nuclear Instrumentation X K5.01 Knowledge of the operational 2.7 31 implication of separation of control and protection circuits as they apply to the NNIS. (CFR: 41.5 / 45.7)
K3.01 -Knowledge of the effect that a 3.4*
loss or malfunction of the NNIS will have on the following: RCS 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X A1.02 Knowledge of the physical 2.5 32 connections and/or cause effect relationships between the Spent Fuel Pool Cooling System and RHRS. (CFR:
41.2 to 41.9 / 45.7 to 45.8)
A1.01 Ability to predict and/or monitor changes in parameters (to prevent 2.7 exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Spent fuel pool water level 034 Fuel Handling Equipment X A4.02 Ability to manually operate and/or 3.5 33 monitor neutron levels in the control room. (CFR: 41.7 / 45.5 to 45.8) 035 Steam Generator X A3.01 Ability to monitor automatic 4.0 34 operation of the S/G including S/G water level control. (CFR: 41.7 / 45.5) 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste
071 Waste Gas Disposal X G2.1.30 Ability to locate and operate 4.4 35 components, including local controls.
(CFR: 41.7 / 45.7) 072 Area Radiation Monitoring X K4.01 Knowledge of ARM system design 3.3* 36 feature(s) and/or interlock(s) which provide for containment ventilation isolation. (CFR: 41.7) 075 Circulating Water 079 Station Air X K1.01 Knowledge of the physical 3.0 37 connections and/or cause effect relationships between the SAS and IAS.
(CFR: 41.2 to 41.9 / 45.7 to 45.8) 086 Fire Protection X A2.04 Ability to (a) predict the impacts of 3.3 38 failure to actuate the FPS when required, resulting in fire damage on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of failure to actuate the FPS when required, resulting in fire damage: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
K/A Category Point Totals: 1 1 0 1 1 1 1 1 1 1 1 Group Point Total: 10/3
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Date of Exam:
Category K/A # Topic RO SRO-Only IR # IR #
Knowledge of administrative requirements for temporary 2.1.15 2.7 66 management directives, such as standing orders, night orders, Operations memos, etc. (CFR: 41.10 / 45.12) 2.1.32 Ability to explain and apply system limits and precautions. 3.8 67
- 1. (CFR: 41.10 / 43.2 / 45.12)
Conduct of 2.1.45 Ability to identify and interpret diverse indications to 4.3 68 Operations validate the response of another indication. (CFR: 41.7 /
43.5 / 45.4)
Subtotal 2.2.42 Ability to recognize system parameters that are entry- 3.9 69 level conditions for Technical Specifications. (CFR: 41.7 /
41.10 / 43.2 / 43.3 / 45.3) 2.2.43 3.0 Knowledge of the process used to track inoperable alarms.
- 2. 2.2.44 Ability to interpret control room indications to verify the 4.2 70 Equipment status and operation of a system, and understand how Control operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)
Subtotal 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 3.8 71 45.10)
Knowledge of radiation monitoring systems, such as fixed 2.3.15 2.9 72 radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR:
- 3. 41.12 / 43.4 / 45.9)
Radiation Control Subtotal 2.4.4 Ability to recognize abnormal indications for system 4.5 73 operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR:
41.10 / 43.2 / 45.6) 4.
Emergency 2.4.32 Knowledge of operator response to loss of all 3.6 74 Procedures / annunciators. (CFR: 41.10 / 43.5 / 45.13)
Plan 2.4.37 Knowledge of the lines of authority during implementation 3.0 75 of the emergency plan. (CFR: 41.10 / 45.13)
Subtotal
Tier 3 Point Total 10 7 ES-401 PWR Examination Outline Form ES-401-2 Facility: Date of Exam:
RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total
- 1. 1 18 3 3 6 Emergency &
Abnormal 2 N/A N/A 9 2 2 4 Plant Evolutions Tier Totals 27 5 5 10 1 28 3 2 5 2.
Plant 2 10 0 2 1 3 Systems Tier Totals 38 5 3 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 2 1 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As DCPP Rev 0A, 07/28/2016
ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000007 Reactor Trip - Stabilization -
Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 X AA2.10 Ability to determine and interpret when to 3.7 76 secure RCPs on loss of cooling or seal injection as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow). (CFR 43.5 /
45.13) 3.8 AA2.11 -when to jog RCPs during ICC 000022 Loss of Rx Coolant Makeup / 2 X AA2.01 Ability to determine and interpret whether 3.8 77 charging line leak exists as they apply to the Loss of Reactor Coolant Makeup. (CFR 43.5/ 45.13) 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 X G2.4.41 Knowledge of the emergency action level 4.6 78 thresholds and classifications. (CFR: 41.10 / 43.5
/ 45.11) 000038 Steam Gen. Tube Rupture / 3 X EA2.03 Ability to determine or interpret which S/G 4.6 79 is ruptured as they apply to a SGTR. (CFR 43.5 /
45.13) 000040 Steam Line Rupture - Excessive Heat Transfer / 4 000054 Loss of Main Feedwater / 4 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant X G2.4.2 Knowledge of system set points, 4.6 80 Recirc. / 4 interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7 / 45.8)
W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 DCPP Rev 0A, 07/28/2016
000077 Generator Voltage and Electric X G2.4.30 Knowledge of events related to system 4.1 81 Grid Disturbances / 6 operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)
K/A Category Totals: 3 3 Group Point Total: 18/6 DCPP Rev 0A, 07/28/2016
ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 X G2.4.47 Ability to diagnose and recognize 4.2 82 trends in an accurate and timely manner utilizing the appropriate control room reference material.l (CFR: 41.10 / 43.5 /
45.12) 000024 Emergency Boration / 1 X AA2.04 Ability to determine and interpret 4.2 83 the availability of BWST as they apply to the Emergency Boration. (CFR: 43.5 /
45.13) 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 X AA2.12 Ability to determine and interpret 3.1* 84 the maximum allowable channel disagreement as they apply to the Loss of Intermediate Range Nuclear instrumentation. (CFR: 43.5 / 45.13) 000036 Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 Control Room Evac. / 8 000069 Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 X G2.4.31 Knowledge of annunciator alarms, 4.1 85 indications, or response procedures. (CFR:
41.10 / 45.3)
W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 K/A Category Point Totals: 2 2 Group Point Total: 9/4 DCPP Rev 0A, 07/28/2016
ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump 004 Chemical and Volume X G2.1.20 Ability to interpret and 4.6 86 Control execute procedure steps. (CFR: 41.10
/ 43.5 / 45.12) 005 Residual Heat Removal 006 Emergency Core Cooling X (Replacement for 078 A2.01) 90 A2.13,Ability to (a) predict the impacts 4.2 of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Inadvertent SIS actuation 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features X A2.04 Ability to (a) predict the impacts 4.2 87 Actuation of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Loss of instrument bus (CFR: 41.5 / 43.5 /
45.3 / 45.13) 022 Containment Cooling 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater X A2.03 Ability to (a) predict the impacts 3.1* 88 of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Overfeeding event (CFR: 41.5 / 43.5 /
45.3 / 45.13) 061 Auxiliary/Emergency X G2.1.25 Ability to interpret reference 4.2 89 Feedwater materials, such as graphs, curves, tables, etc. (CFR: 41.10 / 43.5 / 45.12) 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring DCPP Rev 0A, 07/28/2016
076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals: 3 2 Group Point Total: 28/5 DCPP Rev 0A, 07/28/2016
ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge X G2.4.8 Knowledge of how abnormal 4.5 91 operating procedures are used in conjunction with EOPs. (CFR: 41.10 /
43.5 / 45.13) 4.2 G2.2.25 - Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass A2.02 - Ability to (a) predict the impacts 3.9 Control of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations: Steam valve stuck open 045 Main Turbine Generator X A2.13 Ability to (a) predict the impacts of 2.5* 92 the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Opening of the steam dumps at low pressure (CFR: 41.5 / 43.5 / 45.3 / 45.5) 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring DCPP Rev 0A, 07/28/2016
075 Circulating Water X A2.02 Ability to (a) predict the impacts of 2.7 93 the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of circulating water pumps (CFR: 41.5 / 43.5 / 45.3 / 45.13) 079 Station Air 086 Fire Protection K/A Category Point Totals: 2 1 Group Point Total: 10/3 DCPP Rev 0A, 07/28/2016
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Date of Exam:
Category K/A # Topic RO SRO-Only IR # IR #
Ability to interpret reference materials, such as graphs, 2.1.25 4.2 curves, tables, etc. (CFR: 41.10 / 43.5 / 45.12) 2.1.41 Knowledge of the refueling process. (CFR: 41.2 / 41.10 / 3.7 95 43.6 / 45.13)
- 1. 2.1.40 Knowledge of refueling administrative requirements. 3.9 94 Conduct of Operations 2.2.23 Ability to track Technical Specification limiting conditions 4.6 96 for operations. (CFR: 41.10 / 43.2 / 45.13) 2.2.37 Ability to determine operability and/or availability of safety 4.6 97 related equipment. (CFR: 41.7 / 43.5 / 45.12) 2.
Equipment Control 2.3.12 Knowledge of radiological safety principles pertaining to 3.7 98 licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR:
41.12 / 45.9 / 45.10)
Knowledge of radiation or contamination hazards that 2.3.14 3.8 99
- 3. may arise during normal, abnormal, or emergency Radiation conditions or activities. (CFR: 41.12 / 43.4 / 45.10)
Control
- 4. 2.4.43 Knowledge of emergency communications systems and 3.8 100 Emergency techniques. (CFR: 41.10 / 45.13)
Procedures /
Plan Subtotal Tier 3 Point Total 10 7 DCPP Rev 0A, 07/28/2016
ES-401 Record of Rejected K/As Form ES-401-4 Diablo Canyon Exam 10/2016 SRO KA's REJECTED Tier / Randomly Selected Reason for Rejection Group K/A SRO T1G1 015/017 AA2.10 This KA is very similar to questions in the RO section.
Of the available KA's in this APE, selected AA2.11, When to jog RCPs during ICC (IR 3.8)
SRO T2G1 078A2.01 Unable to write to SRO level. Randomly replaced with:
006 A2.13, Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent SIS actuation (IR 4.2)
SRO T3G1 G2.1.25 Not able to write SRO "generic" question for KA Randomly replaced with G2.1.40, Knowledge of refueling administrative requirements. (IR 3.9)
SRO T2G2 029 G2.4.8 No abnormal procedures for containment purge.
Randomly replaced with G2.2.25, Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (IR 4.2)
DCPP Rev 0A - 07/28/2016
SRO T2G2 045 A2.13 Unable to write to proper level, low operational validity for SRO.
Shifted KA to 041, Steam Dumps, not tested on either examination and randomly selected A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations: Steam valve stuck open (IR 3.9)
DCPP Rev 0A - 07/28/2016
RO KA's REJECTED RO T1G1 APE 026 G2.2.39 No less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> LCO for selected system.
G2.4.11 - Knowledge of abnormal condition procedures (4.0)
RO T1G1 APE040 G2.4.49 There are no immediate actions for a steam line rupture.
Randomly replaced with G.2.4.2- Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (IR 4.5)
RO T1G1 APE 056 AA2.09 Rejected due to already sampled with KA 022 A3.01. Both require knowledge of CFCU operation.
Randomly replaced with AA2.21 Ability to determine and interpret the following as they apply to the Loss of Offsite Power: EDG voltage and frequency indicators (IR 3.6)
RO T1G1 APE 062 AK3.04 Unable to write question to address KA. Replaced with AK3.03 Guidance actions contained in EOP for Loss of nuclear service water (ASW) (IR 4.0)
RO T1G2 E03 EK2.1 Apparent typo - write up and importance align with EK2.2.
Question written to EK2.2 RO T1G2 APE036 AA2.01 Apparent typo - write up and importance align with AK2.01.
To maintain outline balance, wrote question to AA2.01.
Updated ES-401-2 to reflect wording and importance for AA2.01.
DCPP Rev 0A - 07/28/2016
RO T2G1 003 G2.2.4 There are no unit differences for RCP control board instrumentation, etc.
Randomly replaced with 003 G2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
(IR 3.1)
RO T2G1 008 G2.2.25 Tech Spec Bases is not RO Knowledge.
Randomly replaced with 008 G2.1.28 Knowledge of the purpose and function of major system components and controls. (IR 4.1)
RO T2G1 063 G2.4.1 No immediate action steps for DC electrical distribution.
Randomly replaced with 063 G2.4.6 Knowledge of EOP mitigation strategies. (IR 3.7)
RO T2G1 076 K2.08 No ESF actuated MOVs.
Randomly replaced with 076 K2.01 Knowledge of bus power supplies to Service Water. (IR 2.7)
RO T2G2 033 A1.02 No tie between SFP and RHR Randomly replaced with 033 A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Spent fuel pool water level (IR 2.7)
DCPP Rev 0A - 07/28/2016
RO T2G2 016 K5.01 KA is too similar to T2G1 KA013 K5.02. As this is the only K5 for 016, therefore, randomly selected K3.01 as replacement. This results in a more balanced distribution of the RO Systems Tiers. The original distribution was 3/4/2/3/3/4 (2 K3's and 4 K5's). Now there is 3 K3's and 3 K5's and an overall distribution of 3/4/3/3/3/3.
016 K3.01 - Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: RCS (IR 3.4)
RO T3G2 2.2.42 KA, Ability to recognize system parameters that are entry-level conditions for Technical Specifications is not a "generic" KA for tier 3.
Randomly replaced with 2.2.43,Knowledge of the process used to track inoperable alarms. (IR 3.0)
DCPP Rev 0A - 07/28/2016
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Diablo Canyon Date of Examination: 10/14/2016 Examination Level: RO SRO Operating Test Number: L161 Administrative Topic Type Describe activity to be performed (See Note) Code*
Evaluate Shift Staffing Assignments Conduct of Operations 2.1.5 Ability to use procedures related to shift staffing, such M, R as minimum crew complement, overtime limitations, etc.
(NRCL161-A5) (3.9)
(modified from NRCL081LJA_SROA1)
Evaluate Fire Zone Operability Conduct of Operations 2.1.25 Ability to interpret reference materials, such as graphs, N, R curves, tables, etc.
(NRCL161-A6) (4.2)
Evaluate Valve Stroke Surveillance Test Equipment Control 2.2.12 Knowledge of surveillance procedures.
N, R (4.1)
(NRCL161-A7)
Authorize Emergency Exposure Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or N, R emergency conditions.
(NRCL161-A8) (3.7)
Classify Hostile Action Emergency Procedures/Plan 2.4.41 Knowledge of the emergency action level thresholds N, R and classifications.
(NRCL161-A9) (4.6)
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (> 1)
(P)revious 2 exams (< 1; randomly selected)
ES 301, Page 22 of 27 Rev 0 Rev 0: rev follows initial submittal
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Diablo Canyon Date of Examination: 10/14/2016 Exam Level: RO SRO-I SRO-U Operating Test Number: L161 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)
Safety System / JPM Title Type Code*
Function
- a. (LJC-S1) (001.A2.11) Respond to Unexpected Rod Motion during Routine A,E,M,S 1 Dilution
- b. (LJC-S2) (013.A4.01) SSPS Main Steam Line Actuation Failure A,E,EN,L,N,S 2
- d. (LJC-S4) (E05.EA1.1) Initiate Feed and Bleed for a Loss of Heat Sink (LJC-116) D,E,L,S 4S
- e. (LJC-S5) (E14.EA1.1) Initiate Containment Spray Manually (LJC-010) D,E,L,S 5
- f. (LJC-S6) (064.A4.06) Crosstie of Vital Bus G to H (LJC-032) A,D,E,L,S 6 g.
- h. (LJC-S7) (068.AA1.23) Perform Control Room Actions Prior to Evacuation (LJC- D,E,S 8 21)
In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. (P1) (040.AA1.03) Close MSIV and Bypass Locally - LJP-212 D,E,EN,L 4S
- j. (P2) (E14.EA1.1) Isolate the Spray Additive Tank - LJP-224 A,E,L,R 5
- k. (P3) (062.A2.11) Transfer the TSC to Vital Power - modified LJP-058A A,M,E,L 6
@ 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 / > 1 / > 1 (control room system)
(L)ow-Power / Shutdown >1 / >1 / >1 (N)ew or (M)odified from bank including 1(A) >2 / >2 / >1 (P)revious 2 exams <3 / < 3 / < 2 (randomly selected)
(R)CA >1 / >1 / >1 (S)imulator ES-301, Page 23 of 27 Rev 2
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Diablo Canyon Date of Examination: 10/14/2016 Exam Level: RO SRO-I SRO-U Operating Test Number: L161 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)
Safety System / JPM Title Type Code*
Function
- a. (LJC-S1) (001.A2.11) Respond to Unexpected Rod Motion during Routine A,E,M,S 1 Dilution
- b. (LJC-S2) (013.A4.01) SSPS Main Steam Line Actuation Failure A,E,EN,L,N,S 2 c.
d.
e.
f.
g.
h.
In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. (P1) (040.AA1.03) Close MSIV and Bypass Locally - LJP-212 D,E,EN,L 4S
- j. (P2) (E14.EA1.1) Isolate the Spray Additive Tank - LJP-224 A,E,L,R 5
- k. (P3) (062.A2.11) Transfer the TSC to Vital Power - modified LJP-058A A,M,E,L 6
@ 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 / > 1 / > 1 (control room system)
(L)ow-Power / Shutdown >1 / >1 / >1 (N)ew or (M)odified from bank including 1(A) >2 / >2 / >1 (P)revious 2 exams <3 / < 3 / < 2 (randomly selected)
(R)CA >1 / >1 / >1 (S)imulator ES-301, Page 23 of 27 Rev 2
Appendix D (rev 10) Scenario Outline Form ES-D-1 Facility: Diablo Canyon (PWR) Scenario No: 1 Op-Test No: L161 NRC Examiners: Operators:
Initial Conditions: 2% with AFW in service, backfeeding from 500 kV, BOL, 1609 ppm boron Turnover: At start of OP L-3, preparing MFPs to place in service.
Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 XMT_CVC19_3 0.0 delay=0 I (ATC, LT-112 Fails Low (auto make-up) (AP-19, AP-5) ramp=120 SRO) 2 DSC_VEN12 BREAKER_OPEN TS only Loss of Power to S-31 (PK15-17; T.S. 3.7.12.B)
(SRO) 3 AS01ASW_ASP11_MTFSEIZUR 1 TS, C ASW Pp 1-1 Seizes; Pp 1-2 SF6 Breaker Pressure Fault (AP-10, (BOP, T.S. 3.0.3)
AS02E03V00_52HG6TF_SF6 2 SRO) 4 XMT_MSS1_3 1215 delay=0 I (ATC, PT-507 Fails High (AP-5) ramp=300 SRO) 5 MAL_RCS3G .75 delay=0 ramp=300 M (ALL) 750 gpm LOCA on Loop 4 Hot Leg due to earthquake 6 MAL_PPL3B BOTH C (BOP) Safety Injection, Train B fails to actuate 7 VLV_SIS1_1 1 C (ATC) 8803 A Fails closed on SI (S1CT-1) 8 MAL_SEI1 0.1500000 ramp=10 C (ALL) RWST drains to less than 4% due to seismic damage (S1CT-2)
ASISRWST 1.53e6 delay=10 ramp=300
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L161 NRC ES-D-1-01 r3.docx Page 1 of 3 Rev 3
Appendix D (rev 10) Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4) Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,3,4,5,6,7,8) 7
- 2. Malfunctions after EOP entry (1-2) (Events 6,7,8) 3
- 3. Abnormal events (1-4) (Events 1,3,4) 3
- 4. Major transients (1-2) (Event 5) 1
- 5. EOPs entered/requiring substantive actions (1-2) (E-1.3) 1
- 6. EOP contingencies requiring substantive actions (0-2) (ECA-1.1) 1
- 7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S1CT-1) Manually align at least one train of FSAR analysis predicates acceptable results on the
- WCAP-17711-NP, CT-2 SIS actuated safeguards before transition assumption that, at the very least, one train of
- WOG Background out of EOP E-0, Reactor Trip or Safety safeguards has actuated and is providing flow to HE0BG_R2 Injection. the core. Failure to manually align the minimum required safeguards equipment results in the persistence of degraded emergency core cooling system capacity.
(S1CT-2) Stop all running ECCS pumps with Damage to the RWST in this scenario results in a
- WCAP-17711-NP, CT-28 suction aligned to the RWST before continuous loss of level and eventual inability to
- WOG Background insufficient RWST level results in ECCS meet the minimum NPSH requirements for the HECA11BG_R2 pump cavitation as indicated by rapid running ECCS pumps. Failure to stop the pumps swings in pump amperage. before cavitation occurs can lead to pump damage sufficient to render the pumps unavailable for use once an alternate make-up supply is aligned to the RCS.
Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.
L161 NRC ES-D-1-01 r3.docx Page 2 of 3 Rev 3
SCENARIO
SUMMARY
- NRC #1
- 1. Volume Control Tank (VCT) level channel LT-112 fails low, causing a continuous (and erroneous) makeup signal. The crew diagnoses the level channel failure by comparing other VCT parameters, and by using OP AP-19, Malfunction of the Reactor Makeup Control System. The makeup system is secured, and makeup is accomplished (if needed) using manual mode (or enabling the auto mode for short periods).
May elect to use OP AP-5, Malfunction of Eagle 21 Protection or Control Channel to take manual control of Makeup Control System.
- 2. Auxiliary Building Supply Fan S-31 loses power and crew responds per AR PK15-17, AUX & FHB VENT PWR FAILURE. Auxiliary Building Ventilation System, ABVS, which was operating in Buildings and Safeguards, swaps to Safeguards only. Crew verifies automatic shutdown of Supply E-1 as well as auto-swap to Safeguards only alignment for the ABVS. Shift Foreman enters TS 3.7.12.B, Auxiliary Building Ventilation System (ABVS) for one ABVS train inoperable.
- 3. ASW Pump 1-1 trips due to a seized shaft. Standby ASW Pump 1-2 fails to start as the result of a fault at the breaker (SF6 pressure fault). The Shift Foreman implements OP AP-10, Loss of Auxiliary Salt Water and cross-ties to the Unit 2 ASW system via the ASW cross-tie valve FCV-601. Shift Foreman enters T.S.
3.0.3 for two trains of ASW inoperable on Unit 1.
- 4. Steam Generator Header Pressure Transmitter, PT-507, fails high over 5 minutes causing actual temperature to lower. Crew identifies malfunction noting increase in steam flow and lowering Tcold, and takes manual control of HC-507. OP AP-5, Malfunction of Eagle 21 Protection or Control Channel is used to address the failure and return primary and secondary plant parameters to normal bands.
- 5. An earthquake occurs, causing a 750 gpm leak to ramp in on loop 4 hot leg. The crew determines the leak is substantial in size based on a rapid drop in pressurizer level. The Shift Foreman directs a reactor trip and safety injection.
- 6. The crew enters EOP E-0, Reactor Trip or Safety Injection. Train B of Safety Injection fails to actuate, requiring the crew to perform numerous manual alignments and pump starts as part of Appendix E.
- 7. Charging Injection Supply Valve, CVCS-1-8803A fails to open on SI as well. The crew must open 8803A or its parallel equivalent, CVCS-1-8803B in order to meet the requirements of S1CT-1, Manually actuate at least one train of SIS actuated safeguards before transition out of EOP E-0.***
- 8. The seismic event damages the RWST, resulting in a large fissure that terminates close to the bottom of the tank. The crew briefly enters E-1, Loss of Reactor or Secondary Coolant prior to transitioning to E-1.3, Transfer to Cold Leg Recirculation when RWST level reaches 33%, which happens quickly due to the leaking RWST. With the Containment Recirc Sump Level less than 92%, the crew is forced into EOP ECA-1.1, Loss of Emergency Coolant Recirculation. The fissure location causes the RWST to continue to drain, requiring the crew to perform the second critical task S1CT2 - Stop ECCS pumps aligned to the RWST before insufficient level results in ECCS pump cavitation.***
The scenario is terminated once the crew has implemented Appendix W, RCS Makeup from VCT.
L161 NRC ES-D-1-01 r3.docx Page 3 of 3 Rev 3
Appendix D (rev 10) Scenario Outline Form ESD1 Facility: Diablo Canyon (PWR) Scenario No: 2 OpTest No: L161 NRC Examiners: Operators:
Initial Conditions: 100% MOL, 878 ppm boron Turnover: TDAFW OOS for repair; Emergent issue on CCP 11 (OOS)
Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 VLV_SIS7_2 0 cd='h_v1_144g_1 C (BOP, SI8923A fails closed during STP V3L10A AND V1_144 S_2' SRO) 2 XMT_CVC4_3 0.0 delay=0 ramp=30 TS, I FT128 Fails low causing high charging flow (OP AP5; OP AP (ATC, 17, T.S. 3.3.4.A)
SRO) 3 MAL_SIS1C 100 delay=0 r amp=180 TS only Accumulator 13 100 gpm leak (PK0210; T.S. 3.5.1.B)
(SRO) 4 MAL_RCS3C .10 delay=0 ramp=180 TS, C 100 gpm RCS leak on Loop 3 (OP AP1, T.S. 3.4.13.A)
(ALL) 5 PMP_CVC2_2 OVERLOAD_DEV_FAIL M (ALL) CCP 12 OC Trip requiring crew to trip/SI plant 6 MAL_RCS3C 11 cd='jpplsia' del ay=0 M (ALL) 4.5 sq in SBLOCA on Loop 3 ramp=120 7 pmp_afw2_1 open delay=0 C (ATC) MDAFW Pp 13 Autostart Failure (Requires manual start; cd='fnispr lt 5' (S2CT1)
(Occurs w/Loss of MDAFW Pp 12 on Bus H Differential) 8 MAL_EPS4E_2 DIFFERENTIAL cd='h_ C (ALL) SIP 12 Lost on Bus H Differential v4_218r_1' CCP 13 cavitates and trips.
CV09CVC_932TASTEM 0 Loss of all high and intermediate head injection; CT to PMP_CVC3_2 OVERLOAD_DEV_FAIL depressurize to inject accumulators (S2CT2).
cd='jpplsia and H_V2_266R_1' delay=120
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor L161 NRC ESD102 r3.docx Page 1 of 3 Rev 3
Appendix D (rev 10) Scenario Outline Form ESD1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES3014) Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,2,4,5,6,7,8) 7
- 2. Malfunctions after EOP entry (12) (Events 7,8) 2
- 3. Abnormal events (2-4) (Events 1,2,4) 3
- 4. Major transients (12) (Event 5,6) 2
- 5. EOPs entered/requiring substantive actions (1-2) (E1) 1
- 6. EOP contingencies requiring substantive actions (FRC.2) 1
- 7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S2CT1) Establish at least 435 gpm AFW Failure to manually establish the minimum WCAP17711NP, CT4 flow to the steam generators prior to required AFW flow rate (when it is possible to do WOG Background exiting EOP E0. so) results in a challenge to the Heat Sink critical HFRH1BG_R2 safety function. In this scenario, adequate S/G level is also required to effectively depressurize the RCS to inject accumulators in the absence of both high and intermediate head injection pumps.
(S2CT2) Depressurize Steam Generators to Failure to depressurize the SGs results in the WCAP17711NP, CT42 inject SI Accumulators to reflood the core avoidable continuation of the degraded of core WOG Background as indicated by RVLIS level returning above cooling condition. Depressurizing the S/Gs provides HFRC2BG_R2 the minimal required level shown below immediate benefit by condensing steam on the before a RED path develops on Core primary side of the Utubes. Once pressure RCS Cooling Critical Safety Function. falls below approximately 625 psig, Accumulators will inject, flooding the core and clearing the RVLIS Dynamic Range magenta path on Core Cooling. Continuing the Indication GREATER THAN: 100oF/hr cooldown after Accumulators have injected ultimately results in RCS pressure lowering RCPs RVLIS RVLIS below RHR shutoff head and the core cooling CSF Running Level Range status returning to normal.
1 14% Dyn 2 20% Dyn 3 30% Dyn 4 44% Dyn Per NUREG1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the postscenario review.
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SCENARIO
SUMMARY
- NRC #2
- 1. Crew performs timed stroke test of SI Pump 11 suction valve 8923A per STP V3L10A, Exercising Valve SI8923A, Safety Injection Pump 1 Suction Valve. The valve strokes closed, but does not respond when the crew attempts to reopen it per the test procedure. The Shift Foreman notes already in Tech Spec for CCP 11 out of service. May contact Maintenance for assistance.
- 2. FT128 (charging flow) fails low, causing actual charging flow to rise. The crew responds per OP AP5, Malfunction of Eagle 21 Protection or Control Channel. FCV128 and HC459D are taken to manual, and charging flow is monitored using alternate indications (RCP seals, Pzr level, VCT level, etc) for the remainder of the scenario. OP AP17, Loss of Charging, Section B (Charging System Equipment Malfunctions), may also be used to respond to the failure. TS 3.3.4.A, Remote Shutdown Systems, is implemented.
- 3. Accumulator 13 develops a 100 gpm leak, bringing in AR PK0210, ACUM LEVEL HILO for level below 60.8%. Shift Foreman enters TS 3.5.1.B, Accumulators, when level falls below 52%.
- 4. A 100 gpm RCS leak on loop 3 ramps in over the next 3 minutes, requiring entry in OP AP1, Excessive Reactor Coolant System Leakage. Pressure and level are stabilized once CCP 12 is started and letdown isolated. VCT level cannot be maintained at the current leak rate, however, and the crew determines a plant shutdown is required. Shift Foreman enters TS 3.4.13.A, RCS Operational Leakage.
- 5. CCP 12 trips and charging is no longer able to keep up with the leak. Shift Foreman directs a Reactor Trip and Safety Injection. The crew enters EOP E0, Reactor Trip or Safety Injection and performs their immediate actions.
- 6. Loop 3 ruptures on the Safety Injection, with a 4.5 inch SBLOCA ramping in over the next 2 minutes.
- 7. Bus H is lost on a differential trip during the transfer to Startup and MDAFW Pp 13 fails to Autostart, leading to the critical task of starting MDAFW Pp 13 (S2CT1) Establish at least 435 gpm AFW flow to the steam generators prior to exiting EOP E0.***
- 8. SIP 12 is lost with the loss of bus H. NonECCS Charging Pump CCP 13 begins to cavitate and eventually trips, resulting in a total loss of all high and intermediate head injection. The crew proceeds through E0, noting that RCPs must remaining running when pressure falls below 1300 psig due to a lack of running ECCS CCPs and SIPs. The crew determines the RCS is not intact and transitions to E1, Loss of Reactor or Secondary Coolant. A loss of subcooling and lowering RVLIS level eventually results in a magenta path on the core cooling critical safety function, and the crew transitions to FRC.2, Response to Degraded Core Cooling. Following the guidance of FRC.2, the crew will perform the critical task of temporarily recovering the core:
(S2CT2) Depressurize Steam Generators to inject SI Accumulators to reflood the core before a RED path develops on Core Cooling Critical Safety Function.***
The scenario is terminated once Accumulators have injected enough volume to clear the MAGENTA path on CORE COOLING
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Appendix D (rev 10) Scenario Outline Form ESD1 Facility: Diablo Canyon (PWR) Scenario No: 3 OpTest No: L161 NRC Examiners: Operators:
Initial Conditions: 100% MOL, 878 ppm boron Turnover: OOS Equipment: PT403 Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 CC01CCW_CCP11_MTFSHEAR 1 TS, C CCW Pp 11 Shaft Shear (AR PK0111, AR PK0109, AR PK01 (BOP, 08, AP11; TS 3.7.7.A)
SRO) 2 EECIX5213D5_51TF_ACT 1 C (ATC, Pressurizer Heater Group #1 Over Current Trip (AR PK0519, SRO) OP A4A:I) 3 MAL_PPL7J 1 TS, I Eagle 21 DFP1 Halt in Rack 10 (AP5; T.S. 3.3.1.D,E,M; (BOP, 3.3.2.D, L; 3.4.11 )
SRO) 4 LOA_TUR28 0 C (ALL) Main Turbine Stop Valve #2 (Loop 1) Closes (PK0406, PK 08 12) 5 CNV_MFW6_2 1 delay=0 ramp=30 M (ALL) Loop 4 FW Reg Fails to 100%, P14 (High S/G Level Trip)
Failure (S3CT3) 6 MAL_MFW5D 2e+007 cd='fnispr lt M (ALL) Feedline Header Break Inside Containment on S/G 14 5' delay =0 ramp=10 (S3CT2) 7 MAL_EPS4D_2 DIFFERENTIAL C (ATC) 4kV Bus G Bus Transfer Failure; Isolate feedflow from TDAFW cd='fnispr lt 5' as part of Critical Task (S3CT1, partial) 8 VLV_MFW4_2 1 C (BOP) FCV441 fails open; Isolate feedflow as part of Critical Task (S3CT1, partial) delIA VLV_MFW4_2 2 cd=V3_193S_1
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Appendix D (rev 10) Scenario Outline Form ESD1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES3014) Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,2,3,4,5,6,7,8) 8
- 2. Malfunctions after EOP entry (12) (Events 7,8) 2
- 3. Abnormal events (1-4) (Events 1,2,3,4) 4
- 4. Major transients (12) (Event 5,6) 2
- 5. EOPs entered/requiring substantive actions (1-2) (E2, E1.1) 2
- 6. EOP contingencies requiring substantive actions (0-2) 0
- 7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S3CT1) Manually isolate feedline break Failure to isolate feed flow into containment leads WOG Background before containment wide range sump level to an unnecessary and avoidable severe challenge HFRZ2BG_R2 reaches 94 feet (LI940 & LI941), resulting to the containment integrity safety function as a in a magenta path on the Containment result of flooding.
safety function status tree.
(S3CT2) Terminate ECCS flow before The feedline rupture introduced in this scenario DCPP Design Criteria overfill of the RCS results in a rupturing of results in a Safety Injection due to shrinkage and a Memorandum S7: Reactor the pressurizer relief tank (PRT) as slightly overcooled condition in the RCS. Once Coolant System indicated by a PRT pressure drop and isolated, RCS pressure rises quickly as the result of FSAR Chapter 15, subsequent equalization with wide range ongoing injection flow. Eventually the RCS goes Containment Pressure. solid, with the excess inventory passing water through the Pressure Operated Relief Valves to the PRT. Failure to terminate ECCS flow when it is possible to do so results in a rupture of the PRT and constitutes an avoidable degradation of a fission product barrier.
(S3CT3) Manually trips the reactor before Steam Generator Level above the High High Generic Letter 8128 S/G 14 reaches 92% narrow range. setpoint (P14) normally generates a turbine trip WOG Background signal to protect against high feedwater flow and HFRH3BG_R2 carryover into the steam lines when one out of four S/G has reached a narrow range level greater than 90%. Carryover into the steam lines can result in damage to downstream piping, valves, placing the secondary heat sink at risk.
Per NUREG1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the postscenario review.
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SCENARIO
SUMMARY
- NRC #3
- 1. AR PK0111, CCW Pp 11 Recirc comes into alarm for FCV606, CCW Pump 11 Recirc Valve, open. Crew identifies low pump amps on VB1 and dispatches Nuclear Operator to investigate. Field reports no audible flow sound in spite of indications motor is running. CCW Pump 13 is started manually and CCW Pump 11 shutdown. T.S. 3.7.7.A, Vital Component Cooling Water (CCW) System, is entered for one loop of CCW inoperable.
- 2. Pressurizer Heater Group #1 trips on over current, bringing in AR PK0519, PZR HTRS OC TRIP/FAN FLO LO. Crew places additional backup heater group in service per OP A4A:I, Pressurizer - Make Available, Section 6.6.
- 3. Eagle 21 experiences a Digital Filter Processor (DFP) halt on rack 10. Associated indicators PI456, LI 460A, FI415, FI425, FI435, FI445 (VB2), and PR445, LR459 (CC2) fail asis as well as control channels for PORV 456 (PT456) and Pressurizer Level Control (LT460). Crew responds per OP AP5, Malfunction of Eagle 21 Protection or Control Channel. Shift Foreman reviews Tech Specs, entering:
TS 3.3.2.D, PC 456D Low Press SI (72 hrs)
TS 3.3.1.E, PC456A High Press Trip (72 hrs)
TS 3.3.1.M, PC 456C Low Press Trip (72 hrs)
TS 3.3.1.M, LC 460A High Level Trip (72 hrs)
TS 3.3.1.M, FC415(425,435,445) RCS Loop 1 (2,3,4) Flow (72 hrs)
TS 3.3.2.L, PC456 B, P11 (1 hr)
TS 3.4.11.B1, B2, & B3 PC456 E, to close & remove power from associated block valve (1 hr) and restore to operable (72 hrs)
- 4. Main Turbine Stop Valve #2 (Loop 1) closes, causing a secondary transient, bringing PK 0812, TURB LOAD REJECTION C7A into alarm. Power lowers approximately 10%. Crew may perform a diagnostic brief to identify the cause of the excursion. The crew identifies various indications that the stop valve is fully closed such as deviations in steam flow, temperature, Triconex display, as well as activation of PK0406, PROTECTION CHANNEL ACTIVATED for 1 out of 4 Turb Stm Stop Vlvs Clsd. Power is stabilized per Shift Foremans direction.
- 5. Loop 4 Feedwater Reg Valve, FCV540, fails full open. The crew identifies the malfunction and attempts to take manual control, but is unsuccessful. Shift Foreman directs manual reactor trip before S/G 14 level reaches the auto trip point of 90%. (S3CT3) Manually trips the reactor before S/G 14 reaches 92% narrow range.
- 6. On the trip, the feedline header to S/G 14 fails catastrophically, causing S/G 14 to depressurize into containment.
- 7. Bus G fails to transfer to startup resulting in a loss of power to the TDAFW LCVs and an inability to throttle flow. Feed flow from the pump is isolated during recovery actions as part of critical task S3CT1 (see page 4, below).
- 8. Feedwater Isolation Valve FCV441 fails open, requiring manual isolation at VB3. (Part of critical task S3CT1, see page 4, below).
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SCENARIO
SUMMARY
- NRC #3 Crew enters EOP E0, Reactor Trip or Safety Injection and performs their immediate actions. E0 diagnostic steps direct the crew to E2, Faulted Steam Generator Isolation to perform the critical task of isolating the feedline break. (S3CT1) Manually isolate feedline break before Containment wide range level reaches 94 feet, resulting in a magenta path on the Containment safety function status tree. ***
Once isolated, the Shift Foreman verifies SI termination criteria has been met and transitions to E1.1, SI Termination, performing the critical task of sequentially reducing ECCS flow and realigning the plant to a preSI configuration. (S3CT2) Terminate ECCS flow before overfill of the RCS results in a rupture of the pressurizer relief tank (PRT). ***
The scenario is terminated once normal charging/letdown is aligned in E1.1, ready to perform step 15.
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Appendix D (rev 10) Scenario Outline Form ESD1 Facility: Diablo Canyon (PWR) Scenario No: 4 OpTest No: L161 NRC Examiners: Operators:
Initial Conditions: 71% MOL, 919 ppm boron Turnover: OOS Equipment: CCP 12 Event Malf Event Event Description No No. Type* (See Summary for Narrative Detail) 1 MAL_DEG2B FAULT TS, C Fail on D/G 12 Manual Start (AR PK1701; TS 3.8.1.B) cd='h_v4_101m_1 gt 680' (BOP, SRO) 2 VLV_CVC22_2 0.2 delay=0 ramp=15 I (ALL) Regen Hx Isolation Valve, LCV459, Fails to midposition (AP 18)
RLY_PPL63 CLOSED (TRUE) 3 TS, I SSPS relay actuation causes inadvertent start of TDAFW (ALL) pump and blowdown sample isolation valves to close (AR PK0403, OP1.DC10; TS 3.7.5.B) 4 PK1421_0829 1 C (ALL) Loss of Main Transformer Cooling (PK1421) 5 MAL_GEN1_3 TRUE M (ALL) Main Transformer Unit Trips causing a Turbine trip; Buses transfer to StartUp Power (PK1401, PK1211, AP25) 6 MAL_PPL5A 3, MAL_PPL5B 3 M (ALL) ATWS; rod control malfunction; CT to add negative reactivity V5_245S_1 0, V5_239S_1 0 (S4CT1).
7 BKR_EPS20 OPEN cd='h_v5_230r_1' C (ALL) 12kV StartUp Feeder Breaker to StartUp Trip, SDR Failure on D/G 11 and 13; Critical Task to start D/G 13 to restore MAL_DEG1C_2 NO_RESET 4kV vital bus (S4CT2).
cd='H_V4_224R_1'
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Appendix D (rev 10) Scenario Outline Form ESD1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES3014) Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,2,3,4,5,6,7) 7
- 2. Malfunctions after EOP entry (12) (Event 7) 1
- 3. Abnormal events (1-4) (Events 1,2,3,4) 4
- 4. Major transients (12) (Events 5,6) 2
- 5. EOPs entered/requiring substantive actions (1-2) (E0.1) 1
- 6. EOP contingencies requiring substantive actions (0-2) (FRS.1, ECA0.0 ) 2
- 7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S4CT1) Insert negative reactivity into the Failure to insert negative reactivity as procedurally WCAP17711NP, CT52 core following per EOP FRS.1 guidance so directed constitutes a failure to provide FRS.1 Background that power is reduced to less than 5% by appropriate reactivity control and represents an Document, Rev. 3.
the completion of step 19. unnecessary and avoidable challenge to the criticality safety function.
(S4CT2) Energize at least one vital AC bus Failure to restore vital AC power when it is WCAP17711NP, CT24 and restore RCP seal cooling prior RCP shut available represents an unnecessary continuation ECA0.0 Background down seal activation which is identifiable of a degraded emergency power condition and Document, Rev. 3.
by seal no. 1 return flow dropping from a presents a potential challenge to the RCS fission DCM No. S7, Rev 29 normal value of greater than 2 gpm to less product barrier due to a loss of cooling to the RCP than 1 gpm. (occurs when seal outlet seals.
temperatures exceed 260oF)
Per NUREG1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the postscenario review.
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SCENARIO
SUMMARY
- NRC #4
- 1. Maintenance requests manual start and loading of D/G 12 per OP J6B:V, Diesel Generators Manual Operation of DG 12, to take thermography readings inside the SED panel due to loose terminations discovered during most recent routine MOW. The diesel fails to start (over crank condition) bringing in AR PK1701, DEISEL 12 FAIL TO START. Shift Foreman enters TS 3.8.1.B, AC Sources - Operating, for one D/G inoperable.
- 2. Regen Hx Isolation Valve, LCV459, drifts to midposition causing letdown orifice valve 8149C to close.
Shift Foreman enters OP AP18, Letdown Line Failure. Excess Letdown is established per OP B1A:IV CVCS Excess Letdown Place In Service and Remove From Service.
- 3. SSPS relay actuation results in Turbine Driven AFW (TDAFW) Pump Steam Supply Isolation Valve, FCV95, failing open and isolation of blowdown sample valves inside and outside containment. S/G levels rise and RCS temperature lowers, causing control rods to step out in response. FCV95 cannot be closed and the crew must isolate the TDAFW Pump by either closing the LCVs to the individual S/Gs or by closing steam supply valves FCV37 and FCV38 to leads 1 and 2 respectively. Shift Foreman implements TS 3.7.5.B, AFW System for one AFW train inoperable.
- 4. Crew responds to AR PK1421, MAIN TRANSF. A nuclear operator is dispatched to investigate local alarms and reports back that NO cooling fans or oil pumps are running on Main Bank C Transformer.
Shift Foreman enters OP AP25, Rapid Load Reduction or Shutdown and directs a 50 MW/min power reduction while Maintenance and field Operators attempt to restore transformer cooling.
- 5. At approximately 60% power, the plant experiences a Main Transformer Unit Trip due to a fault in the main transformer and all buses successfully transfer to Startup power. The Turbine trips as expected, but the reactor trip breakers remain closed. (AR PK1401, UNIT TRIP; AR PK1211 TURBINE TRIP, AR PK0411 REACTOR TRIP INITIATE)
- 6. Reactor power is still greater than 50% and the crew identifies the ATWS condition. EOP FRS.1, Response to Nuclear Power Generation / ATWS, is entered, either directly or from the step 1, response not obtained column of EOP E0, Reactor Trip or Safety Injection. Attempts to trip the reactor from the Control Room are unsuccessful and Autorod motion has failed. The crew performs the critical task of adding negative reactivity by manually driving rods (S4CT1) Insert negative reactivity into the core so that power is less than 5%.*** The crew continues working through FRS.1 until field operators are able to locally open the reactor trip breakers.
- 7. The reactor is verified subcritical and the crew transitions to EOP E0, Reactor Trip or Safety Injection.
EOP E0.1 Reactor Trip Response is entered once the need for a Safety Injection has been ruled out.
Shortly after verifying primary and secondary parameters are stable, Startup Feeder Breaker 52VU12 trips open and cannot be closed. D/G 11 (no reset) and 13 (resettable from Control Room) fail as a result of their associated shutdown relays activating. The crew transitions to EOP ECA0.0, Loss of All Vital AC Power and performs the critical task of starting D/G 13 (S4CT2) Energize at least one vital AC bus and restore RCP seal cooling before RCP shut down seals activate.***
The scenario is terminated once RCP seal cooling has been reestablished.
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