ML20106E466
| ML20106E466 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/06/2020 |
| From: | Greg Werner Operations Branch IV |
| To: | Pacific Gas & Electric Co |
| References | |
| Download: ML20106E466 (48) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Rev. 11 Facility: Diablo Canyon Date of Exam: March 6, 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
2 2
4 N/A 4
3 N/A 3
18 6
2 2
2 1
1 1
2 9
4 Tier Totals 4
4 5
5 4
5 27 10
- 2.
Plant Systems 1
3 2
3 3
2 2
3 3
3 2
2 28 5
2 1
0 1
1 1
1 1
1 1
1 1
10 3
Tier Totals 4
2 4
4 3
3 4
4 4
3 3
38 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
Note: 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000007 (EPE 7) Reactor Trip, Stabilization, Recovery / 1 X
Ability to operate and monitor the following as they apply to a reactor trip: (CFR 41.7 / 45.5 / 45.6)
EA1.03 RCS pressure and temperature 4.2 40 000008 (APE 8) Pressurizer Vapor Space Accident / 3 X
2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 4.5 42 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 X
Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA: (CFR 41.8 / 41.10 / 45.3)
EK1.01 Natural circulation and cooling, including reflux boiling 4.1 55 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X
Ability to operate and / or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 41.7 / 45.5 /
45.6)
AA1.03 Reactor trip alarms, switches, and indicators 3.7*
52 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X
Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: (CFR 41.5, 41.10 / 45.6 / 45.13)
AK3.06 RCP thermal barrier cooling 3.2 45 000025 (APE 25) Loss of Residual Heat Removal System / 4 X
Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following:
(CFR 41.7 / 45.7)
AK2.01 RHR heat exchangers 2.9 44 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X
Ability to operate and / or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: (CFR 41.7 / 45.5 / 45.6)
AA1.01 PZR heaters, sprays, and PORVs 4.0 46 000029 (EPE 29) Anticipated Transient Without Scram / 1 000038 (EPE 38) Steam Generator Tube Rupture / 3 X
Ability to determine or interpret the following as they apply to a SGTR: (CFR 43.5 / 45.13)
EA2.12 Status of MSIV activating system 3.9*
53 000040 (APE 40) Steam Line Rupture Excessive Heat Transfer / 4 X
Ability to determine and interpret the following as they apply to the Steam Line Rupture: (CFR: 43.5 /
45.13)
AA2.01 Occurrence and location of a steam line rupture from pressure and flow indications 4.2 41 000054 (APE 54) Loss of Main Feedwater
/4 000055 (EPE 55) Station Blackout / 6 X
Knowledge of the reasons for the following responses as the apply to the Station Blackout:
(CFR 41.5 / 41.10 / 45.6 / 45.13)
EK3.02 Actions contained in EOP for loss of offsite and onsite power 4.3 39 000056 (APE 56) Loss of Offsite Power / 6 X
2.4.46 Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 /
45.12) 4.2 54
ES-401 3
Form ES-401-2 Rev. 11 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X
Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: (CFR 41.5,41.10 / 45.6 / 45.13)
AK3.01 Actions contained in EOP for loss of vital ac electrical instrument bus 4.1 51 000058 (APE 58) Loss of DC Power / 6 X
Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: (CFR 41.8 / 41.10 / 45.3)
AK1.01 Battery charger equipment and instrumentation 2.8 43 000062 (APE 62) Loss of Nuclear Service Water / 4 X
Knowledge of the reasons for the following responses 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 Loss of nuclear service water 4.0 49 000065 (APE 65) Loss of Instrument Air / 8 X
Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air: (CFR 41.7 /
45.5 / 45.6)
AA1.03 Restoration of systems served by instrument air when pressure is regained 2.9 56 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment / 3 X
Knowledge of the interrelations between the (LOCA Outside Containment) and the following: (CFR: 41.7
/ 45.7)
EK2.2 Facility*s 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.
3.8 50 (W E11) Loss of Emergency Coolant Recirculation / 4 X
2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 /
45.2) 4.0 48 (W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 X
Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink)
(CFR: 43.5 / 45.13)
EA2.2 Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments 3.7 47 K/A Category Totals:
2 2
4 4
3 3
Group Point Total:
18
ES-401 4
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal / 1 X
Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:
(CFR: 43.5 / 45.13)
AA2.05 Uncontrolled rod withdrawal, from available indications 4.4 62 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 X
Knowledge of the reasons for the following responses as they apply to Emergency Boration: (CFR 41.5, 41.10 / 45.6 / 45.13)
AK3.02 Actions contained in EOP for emergency boration 4.2 60 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 X
Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.5,41.10 / 45.6 / 45.13)
AK3.01 Termination of startup following loss of intermediate range instrumentation 3.2 59 000036 (APE 36) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X
2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR:
41.10 / 43.5 / 45.3 / 45.12) 4.1 57 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 X
Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the following: (CFR 41.7 / 45.7)
AK2.02 Auxiliary building ventilation system 2.7 65 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 X
2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR:
41.10 / 43.5 / 45.13) 3.8 63 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74) Inadequate Core Cooling / 4 X
Knowledge of the operational implications of the following concepts as they apply to the Inadequate Core Cooling: (CFR 41.8 / 41.10 / 45.3)
EK1.07 Definition of saturated steam 2.8 64 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3
ES-401 5
Form ES-401-2 Rev. 11 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 X
Ability to operate and / or monitor the following as they apply to the (Steam Generator Overpressure)
(CFR: 41.7 / 45.5 / 45.6)
EA1.2 Operating behavior characteristics of the facility 3.0 61 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 X
Knowledge of the operational implications of the following concepts as they apply to the (High Containment Radiation)
(CFR: 41.8 / 41.10, 45.3)
EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the (High Containment Radiation) 3.0 58 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:
2 1
2 1
1 2
Group Point Total:
9
ES-401 6
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump X
Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: (CFR: 41.7 / 45.6)
K3.02 S/G 3.5 1
004 (SF1; SF2 CVCS) Chemical and Volume Control X
Knowledge of the effect of a loss or malfunction on the following CVCS components: (CFR: 41.7 / 45.7)
K6.29 Reason for excess letdown and its relationship to CCWS 2.7 15 005 (SF4P RHR) Residual Heat Removal X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: (CFR: 41.5 / 45.5)
A1.03 Closed cooling water flow rate and temperature 2.5 16 006 (SF2; SF3 ECCS) Emergency Core Cooling X
Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.04 ESFAS-operated valves 3.6 22 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X
Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
A2.06 Bubble formation in PZR 2.6 28 008 (SF8 CCW) Component Cooling Water X
Ability to monitor automatic operation of the CCWS, including: (CFR: 41.7 / 45.5)
A3.02 Operation of the CCW pumps, including interlocks and the CCW booster pump 3.2 7
010 (SF3 PZR PCS) Pressurizer Pressure Control X
Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: (CFR: 41.5 / 45.7)
K5.01 Determination of condition of fluid in PZR, using steam tables 3.5 25 012 (SF7 RPS) Reactor Protection X
Knowledge of the operational implications of the following concepts as the apply to the RPS: (CFR: 41.5 / 45.7)
K5.02 Power density 3.3*
24 013 (SF2 ESFAS) Engineered Safety Features Actuation X
Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: (CFR: 41.7 / 45.6)
K3.03 Containment 4.3 23 022 (SF5 CCS) Containment Cooling X
Knowledge of the physical connections and/or cause-effect relationships between the CCS and the following systems: (CFR: 41.2 to 41.9 /
45.7 to 45.8)
K1.01 SWS/cooling system 3.5 21 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray X
Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.04 Reduction of temperature and pressure in containment after a LOCA by condensing steam, to reduce radiological hazard, and protect equipment from corrosion damage (spray) 3.7 3
ES-401 7
Form ES-401-2 Rev. 11 039 (SF4S MSS) Main and Reheat Steam X
Knowledge of the effect that a loss or malfunction of the MRSS will have on the following: (CFR: 41.7 / 45.6)
K3.03 AFW pumps 3.2*
12 059 (SF4S MFW) Main Feedwater X
Ability to (a) predict the impacts 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: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
A2.07 Tripping of MFW pump turbine 3.0*
4 061 (SF4S AFW)
Auxiliary/Emergency Feedwater X
Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: (CFR: 41.7 / 45.7)
K6.02 Pumps 2.6 6
062 (SF6 ED AC) AC Electrical Distribution X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR:
41.5 / 45.5)
A1.01 Significance of D/G load limits 3.4 27 063 (SF6 ED DC) DC Electrical Distribution X
Knowledge of the physical connections and/or cause-effect relationships between the DC electrical system and the following systems:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.02 AC electrical system 2.7 10 064 (SF6 EDG) Emergency Diesel Generator X
Ability to monitor automatic operation of the ED/G system, including: (CFR: 41.7 / 45.5)
A3.01 Automatic start of compressor and ED/G 4.1 14 073 (SF7 PRM) Process Radiation Monitoring X
2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1
/ 45.13) 3.6 19 076 (SF4S SW) Service Water X
Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: (CFR: 41.2 to 41.9 /
45.7 to 45.8)
K1.05 D/G 3.8*
26 078 (SF8 IAS) Instrument Air X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.01 Pressure gauges 3.1 20 103 (SF5 CNT) Containment X
Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations (CFR: 41.5 / 43.5 /
45.3 / 45.13)
A2 03 Phase A and B isolation 3.5*
17 053 (SF1; SF4P ICS*) Integrated Control 006 (SF2; SF3 ECCS) Emergency Core Cooling X
2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.9 9
008 (SF8 CCW) Component Cooling Water X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including: (CFR: 41.5 / 45.5)
A1.02 CCW temperature 2.9 18
ES-401 8
Form ES-401-2 Rev. 11 010 (SF3 PZR PCS) Pressurizer Pressure Control X
Ability to monitor automatic operation of the PZR PCS, including: (CFR: 41.7 / 45.5)
A3.01 PRT temperature and pressure during PORV testing 3.0 5
026 (SF5 CSS) Containment Spray X
Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.01 Containment spray pumps 3.4*
11 076 (SF4S SW) Service Water X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.04 Emergency heat loads 3.5*
8 078 (SF8 IAS) Instrument Air X
Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.03 Securing of SAS upon loss of cooling water 3.1*
2 103 (SF5 CNT) Containment X
Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
K4.06 Containment isolation system 3.1 13 K/A Category Point Totals:
2 2
3 3
2 2
3 3
3 3
2 Group Point Total:
28
ES-401 9
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication X
Knowledge of the operational implications of the following concepts as they apply to the RPIS: (CFR: 41.5 / 45.7)
K5.01 Reasons for differences between RPIS and step counter 2.7 37 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor X
Knowledge of the effect that a loss or malfunction of the ITM system will have on the following: (CFR: 41.7 / 45.6)
K3.01 Natural circulation indications 3.5*
38 027 (SF5 CIRS) Containment Iodine Removal X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.03 CIRS fans 3.3*
34 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X
2.1.19 Ability to use plant computers to evaluate system or component status. (CFR:
41.10 / 45.12) 3.9 35 029 (SF8 CPS) Containment Purge X
Knowledge of design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.02 Negative pressure in containment 2.9 29 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fuel Handling System controls including: (CFR:
A1.02 Water level in the refueling canal 2.9 31 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X
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: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
A2.02 Steam valve stuck open 3.6 32 045 (SF 4S MTG) Main Turbine Generator X
Knowledge of the physical connections and/or cause-effect relationships between the MT/G system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.06 RCS, during steam valve test 2.6 36 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste X
Ability to monitor automatic operation of the Liquid Radwaste System including: (CFR: 41.7 A3.02 Automatic isolation 3.6 33 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring
ES-401 10 Form ES-401-2 Rev. 11 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection X
Knowledge of the effect of a loss or malfunction on the Fire Protection System following will have on the: (CFR: 41.7 / 45.7)
K6.04 Fire, smoke, and heat detectors 2.6 30 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:
1 0
1 1
1 1
1 1
1 1
1 Group Point Total:
10
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev. 11 Facility:
Date of Exam:
Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.
2.1.28 Knowledge of the purpose and function of major system components and controls. (CFR: 41.7) 4.1 73 2.1.
2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR: 41.1 / 43.6
/ 45.6) 4.3 74 2.1.
2.1.44 Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation. (CFR: 41.10 / 43.7 / 45.12) 3.9 66 Subtotal 3
- 2. Equipment Control 2.2.
2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR: 41.6 / 41.7 / 45.2) 4.6 68 2.2.
2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) 3.4 67 2.2.
2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12) 4.2 70 Subtotal 3
- 3. Radiation Control 2.3.
2.3.11 Ability to control radiation releases. (CFR: 41.11 /
43.4 / 45.10) 3.8 75 2.3.
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 areas, aligning filters, etc. (CFR: 41.12 / 45.9 / 45.10) 3.2 71 Subtotal 2
- 4. Emergency Procedures/Plan 2.4.
2.4.17 Knowledge of EOP terms and definitions. (CFR:
41.10 / 45.13) 3.9 69 2.4.
2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.
(CFR: 41.10 / 43.5 / 45.13) 3.8 72 Subtotal 2
Tier 3 Point Total 10
ES-401 PWR Examination Outline Form ES-401-2 Rev. 11 Facility: Diablo Canyon Date of Exam: March 6, 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
N/A N/A 18 3
3 6
2 9
2 2
4 Tier Totals 27 5
5 10
- 2.
Plant Systems 1
28 3
2 5
2 10 1
2 3
Tier Totals 38 4
4 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 2
1 2
Note: 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 X
Ability to determine or interpret the following as they apply to a small break LOCA: (CFR 43.5 / 45.13)
EA2.34 Conditions for throttling or stopping HPI 4.2 88 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 X 2.4.18 Knowledge of the specific bases for EOPs.
(CFR: 41.10 / 43.1 / 45.13) 4.0 89 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient Without Scram / 1 X
2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 / 43.5 / 45.13) 4.2 87 000038 (EPE 38) Steam Generator Tube Rupture / 3 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X
Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW):
(CFR: 43.5 / 45.13)
AA2.08 Steam flow-feed trend recorder 3.3*
86 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X
2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR:
41.7 / 43.5 / 45.12) 4.6 85
ES-401 3
Form ES-401-2 Rev. 11 (W E04) LOCA Outside Containment / 3 X
Ability to determine and interpret the following as they apply to the (LOCA Outside Containment)
(CFR: 43.5 / 45.13)
EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.
4.2 84 (W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals:
3 3
Group Point Total:
6
ES-401 4
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 X
2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10
/ 43.5 / 45.13) 4.7 93 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 X
2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 /
45.7 / 45.8) 4.6 91 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 (W E10) Natural Circulation with Steam Void in Vessel with/without RVLIS/4 X
Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS)
(CFR: 43.5 / 45.13)
EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
3.9 90 (BW E13 & E14) EOP Rules and Enclosures
ES-401 5
Form ES-401-2 Rev. 11 (W E08) RCS OvercoolingPressurized Thermal Shock / 4 X
EA2. Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock) (CFR: 43.5 / 45.13)
EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
4.2 92 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals:
2 2
Group Point Total:
4
ES-401 6
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control X
Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/ 43/5 / 45/3 / 45/5)
A2.27 Improper RWST boron concentration 4.2 76 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 013 (SF2 ESFAS) Engineered Safety Features Actuation X
Ability to (a) predict the impacts 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; (CFR: 41.5 / 43.5 /
45.3 / 45.13)
A2.01 LOCA 4.8 80 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam X
2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 /
45.12 / 45.13) 4.7 77 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)
Auxiliary/Emergency Feedwater X
Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
A2.07 Air or MOV failure 3.5 78 062 (SF6 ED AC) AC Electrical Distribution
ES-401 7
Form ES-401-2 Rev. 11 063 (SF6 ED DC) DC Electrical Distribution X
2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7
/ 43.2) 4.2 79 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals:
3 2 Group Point Total:
5
ES-401 8
Form ES-401-2 Rev. 11 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant X 2.4.3 Ability to identify post-accident instrumentation. (CFR: 41.6 / 45.4) 3.9 83 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator X
Ability to (a) predict the impacts of the following malfunctions or operations on the S/GS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)
A2.03 Pressure/level transmitter failure 3.6 82 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air X 2.1.32 Ability to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 / 45.12) 4.0 81 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:
1 2 Group Point Total:
3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev. 11 Facility:
Date of Exam:
Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.
2.1.34 Knowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12) 3.5 94 2.1.
2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.
(CFR: 41.10 / 43.6 / 45.6) 4.3 97 Subtotal 2
- 2. Equipment Control 2.2.
2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12) 4.6 98 2.2.
2.2.43 Knowledge of the process used to track inoperable alarms. (CFR: 41.10 / 43.5 / 45.13) 3.3 95 Subtotal 2
- 3. Radiation Control 2.3.
2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR:
41.12 / 43.4 / 45.9 / 45.10) 3.8 99 Subtotal 2
- 4. Emergency Procedures/Plan 2.4.
2.4.29 Knowledge of the emergency plan. (CFR: 41.10 /
43.5 / 45.11) 4.4 96 2.4.
2.4.40 Knowledge of SRO responsibilities in emergency plan implementation. (CFR: 41.10 / 43.5 / 45.11) 4.5 100 Subtotal 2
Tier 3 Point Total 7
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group Randomly Selected K/A Reason for Rejection None at the Draft Level. NRC developed outline.
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Diablo Can~on Date of Examination:
02/24/2020 Examination Level: RO
~ SRO D Operating Test Number:
L181 Administrative Topic (see Note)
Type Describe activity to be performed Code*
Conduct of Operations Apply Overtime Limit Restrictions (NRCL 181-A1)
M,R 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc (2.9)
(NRCADM061-COO-R02)
Conduct of Operations Estimate Decay Heat and Heat Removal Rate (NRCL 181-A2)
M, R 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc.
(3.9)
(Bank: LJC-014)
Equipment Control Perform STP I-IC (NRCL 181-A3)
N, R 2.2.37 Ability to determine Operability and/or availability of safety related equipment.
(3.6)
Radiation Control N, R Determine Primary to Secondary Leakrate (NRCL 181-A4) 2.3.1 1 Ability to control radiation releases. (3.8)
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(:;; 3 for ROs; :;; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (2: 1)
(P)revious 2 exams(:;; 1, randomly selected)
Rev 0
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Diablo Cany:on Date of Examination:
02/24/2020 Examination Level: RO D SRO
~
Operating Test Number:
L181 Administrative Topic (see Note)
Type Describe activity to be performed Code*
Conduct of Operations Apply Overtime Limit Restrictions (NRCL 181-A5)
M, R 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc (3.9)
(NRCADM061-COO-SR01)
Conduct of Operations Review Estimated Decay Heat and Heat (NRCL 181 -A6)
M, R Removal Rate 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc.
(4.2)
(Bank: LJC-014)
Equipment Control Determine 230 kV Operability (NRCL 181-A?)
N, R 2.2.37 Ability to determine Operability and/or availability of safety related equipment.
(4.6)
Radiation Control Determine Primary to Secondary Leakrate (NRCL 181-A8)
N, R and Required Action 2.3.11 Ability to control radiation releases. (4.3)
Emergency Plan Review Emergency Notification for Steam (NRCL 181-A9)
N, R Generator Tube Rupture 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation.
(4.5)
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:5 3 for ROs; :5 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (2: 1)
(P)revious 2 exams (:5 1, randomly selected)
Rev O
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Diablo Can1'on Date of Examination:
02/24/2020 Exam Level: RO D SRO-I D SRO-U
~
Operating Test Number:
L181 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. (S1) (001.A2.11) Dropped Rods During Rod Misalignment Verification A,M,S 1
(Modified LJC-066)
- b. (S2) (013.A1.02) Resp to Changing Plant Params During Rx Trip Resp A,N,EN,L,S 2
C.
- d.
- e.
- f.
- g.
- h.
In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i.
(P1) (010.A2.02) Transfer Pzr Heater Grp 23 to Backup Pwr (LJP-029A)
D 3
- j.
(P2) (062.A2.11) Transfer the TSC to Vital Power(LJP-058A)
A,D,E,L 6
- k. (P3) (013.A4.01) Verfiy Slowdown Isolation Outside Containment (Bank E,D,R 8
LJP-096)
All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank
- 5 9/:5 8/:5 4 (E)mergency or abnormal in-plant
?: 1/?: 1/?: 1 (EN)gineered safety feature
?: 1/?: 1/?: 1 (control room system)
(L)ow-Power/Shutdown
- 1/?
- 1/?: 1 (N)ew or (M)odified from bank including 1 (A)
?: 2/?: 2/?: 1 (P)revious 2 exams
- 5 3/::. 3/::. 2 (randomly selected)
(R)CA
?: 1/?: 1/?: 1 (S)imulator Rev 0
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Diablo Canz'on Date of Examination:
02/24/2020 Exam Level:
~ SRO-I D SRO-U D Operating Test Number:
L181 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. (S1) (001.A2.11) Dropped Rods During Rod Misalignment Verification A,M,S 1
(Modified LJC-066)
- b. (S2) (013.A 1.02) Resp to Changing Plant Params During Rx Trip Resp A,N,EN,L,S 2
D,L,S 3
- d. (S4P) (011.EA 1.11) Transfer to Cold Leg Recirc (Bank LJC-27 A)
A,D,L,S 4P
- e. (S4S) (059.A2.11) Monitor Plant Response During Full Load Rejection A,N,S 4S
- f. (S5) (E14.E1.2) Manually Initiate Containment Spray (Bank LJC-010)
D,E,L,S 5
- g. (S6) (064.A4.06) Crosstie Vital Bus G to H (LJC-032)
D,E,L,S 6
- h. (S8) (068.AA2.17) Fire in 480V Bus G Switchgear Room A,C,N,S 8
In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i.
(P1) (010.A2.02) Transfer Pzr Heater Grp 23 to Backup Pwr (LJP-029A)
D 3
- j. (P2) (062.A2.11) Transfer the TSC to Vital Power(LJP-058A)
A,D,E,L 6
- k. (P3) (013.A4.01) Verfiy Blowdown Isolation Outside Containment (Bank E,D,R 8
LJP-096)
All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank
- 9/$ 8/$ 4 (E)mergency or abnormal in-plant 21/~1/.
- :1 (EN)gineered safety feature 2 1/~ 1/.:: 1 (control room system)
(L)ow-Power/Shutdown
.:: 1/2 1/.:: 1 (N)ew or (M)odified from bank including 1 (A)
.:: 2/.:: 2/21 (P)revious 2 exams s 3/s 3/S 2 (randomly selected)
(R)CA 21/.::1/.::1 (S)imulator Rev 0
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Diablo Canton Date of Examination:
02/24/2020 Exam Level: RO D SRO-I
[Z]
SRO-U D Operating Test Number:
L181 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. (S1) (001.A2.11) Dropped Rods During Rod Misalignment Verification A,M,S 1
(Modified LJC-066)
- b. (S2) (013.A 1.02) Resp to Changing Plant Params During Rx Trip Resp A,N,EN,L,S 2
- c. (S3) (E04.EA1.1) Isolate LOCA Outside Containment (Bank LJC-118)
D,L,S 3
- d. (S4P) (011.EA 1.11) Transfer to Cold Leg Recirc (Bank LJC-27 A)
A,D,L,S 4P
- e. (S4S) (059.A2.11) Monitor Plant Response During Full Load Rejection A,N,S 4S
- f. (S5) (E14.E1.2) Manually Initiate Containment Spray (Bank LJC-010)
D,E,L,S 5
- g.
- h. (S8) (068.M2.17) Fire in 480V Bus G Switchgear Room A,C,N,S 8
In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i.
(P1) (010.A2.02) Transfer Pzr Heater Grp 23 to Backup Pwr (LJP-029A)
D 3
- j. (P2) (062.A2.11) Transfer the TSC to Vital Power(LJP-058A)
A,D,E,L 6
- k. (P3) (013.A4.01) Verfiy Slowdown Isolation Outside Containment (Bank E,D,R 8
LJP-096)
All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank s 9/s 8/s 4 (E)mergency or abnormal in-plant
~1,~11~1 (EN)gineered safety feature
~ 1/;:: 1/~ 1 (control room system)
(L)ow-Power/Shutdown
~ 1 /;?: 1/~ 1 (N)ew or (M)odified from bank including 1 (A)
~ 2/~ 2/~ 1 (P)revious 2 exams s 3/s 3/s 2 (randomly selected)
(R)CA
~1,~11~1 (S)imulator Rev 0
ES-301 Transient and Event Checklist Form ES-301-5 Group 1 (U1, I1, R1, R2)
Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios S1 S2 S3 S4 T
O T
A L
M I
N I
M U
M(*)
CREW POSITION CREW POSITION CREW POSITION CREW POSITION S
R O
A T
C B
O P
S R
O A
T C
B O
P S
R O
A T
C B
O P
S R
O A
T C
B O
P R
I U
RO SRO-I SRO-U RX 1
1 0
NOR 1
1 1
I/C 1,2,3,5,7 5 4
4 2
MAJ 4
1 2
2 1
TS 1,2 2
0 2
2 RO SRO-I1 SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,3,4,6 1,2,3,4,6, 7,8 1,3,5,7 15 4
4 2
MAJ 5
5 4
3 2
2 1
TS 3,4 2,4 4
0 2
2 RO1 SRO-I SRO-U RX 0
1 1
0 NOR 0
1 1
1 I/C 2,3,4,5,6 1,4,7,8 1,2,3,5,6, 7
15 4
4 2
MAJ 5
5 4
3 2
2 1
TS 0
0 2
2 RO2 SRO-I SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,4 2,3,4,6,7, 8
8 4
4 2
MAJ 5
5 2
2 2
1 TS 0
0 2
2 Instructions:
- 1.
Check the applicant level and enter the operating test number and Form ES D 1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at the controls (ATC) and balance of plant (BOP) positions. Instant SROs (SRO I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one for one basis.
- 3.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right hand columns.
- 4.
For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO I applicants in either the ATC or BOP position to best evaluate the SRO I in manipulating plant controls.
ES-301, Page 26 of 27 Rev 0
ES-301 Transient and Event Checklist Form ES-301-5 Group 2 (I2, I3, R3, R4)
Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios S1 S2 S3 S4 T
O T
A L
M I
N I
M U
M(*)
CREW POSITION CREW POSITION CREW POSITION CREW POSITION S
R O
A T
C B
O P
S R
O A
T C
B O
P S
R O
A T
C B
O P
S R
O A
T C
B O
P R
I U
RO SRO-I2 SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,3,4,6 1,2,3,5,7 2,3,4,7 13 4
4 2
MAJ 5
4 5,6 4
2 2
1 TS 3,4 1,2 4
0 2
2 RO SRO-I3 SRO-U RX 0
1 1
0 NOR 0
1 1
1 I/C 1,2,3,4,6, 7,8 1,3,5,7 1,2,3,4,7 16 4
4 2
MAJ 5
4 5,6 4
2 2
1 TS 2,4 1,3 4
0 2
2 RO3 SRO-I SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,4 2,3,4,6,7, 8
1,3,4,7 12 4
4 2
MAJ 5
5 5,6 4
2 2
1 TS 0
0 2
2 RO4 SRO-I SRO-U RX 0
1 1
0 NOR 0
1 1
1 I/C 2,3,4,5,6 1,4,7,8 1,2,3,5,6, 7
15 4
4 2
MAJ 5
5 4
3 2
2 1
TS 0
0 2
2 Instructions:
- 1.
Check the applicant level and enter the operating test number and Form ES D 1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at the controls (ATC) and balance of plant (BOP) positions. Instant SROs (SRO I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one for one basis.
- 3.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right hand columns.
- 4.
For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO I applicants in either the ATC or BOP position to best evaluate the SRO I in manipulating plant controls.
ES-301, Page 26 of 27 Rev 0
ES-301 Transient and Event Checklist Form ES-301-5 Group 3 (I4, I5, R5, R6)
Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios S1 S2 S3 S4 T
O T
A L
M I
N I
M U
M(*)
CREW POSITION CREW POSITION CREW POSITION CREW POSITION S
R O
A T
C B
O P
S R
O A
T C
B O
P S
R O
A T
C B
O P
S R
O A
T C
B O
P R
I U
RO SRO-I4 SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,3,4,6 1,2,3,5,7 2,3,4,7 13 4
4 2
MAJ 5
4 5,6 4
2 2
1 TS 3,4 1,2 4
0 2
2 RO SRO-I5 SRO-U RX 0
1 1
0 NOR 0
1 1
1 I/C 1,2,3,4,6, 7,8 1,3,5,7 1,2,3,4,7 16 4
4 2
MAJ 5
4 5,6 4
2 2
1 TS 2,4 1,3 4
0 2
2 RO5 SRO-I SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,4 2,3,4,6,7, 8
1,3,4,7 12 4
4 2
MAJ 5
5 5,6 4
2 2
1 TS 0
0 2
2 RO6 SRO-I SRO-U RX 0
1 1
0 NOR 0
1 1
1 I/C 2,3,4,5,6 1,4,7,8 1,2,3,5,6, 7
15 4
4 2
MAJ 5
5 4
3 2
2 1
TS 0
0 2
2 Instructions:
- 1.
Check the applicant level and enter the operating test number and Form ES D 1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at the controls (ATC) and balance of plant (BOP) positions. Instant SROs (SRO I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one for one basis.
- 3.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right hand columns.
- 4.
For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO I applicants in either the ATC or BOP position to best evaluate the SRO I in manipulating plant controls.
ES-301, Page 26 of 27 Rev 0
ES-301 Transient and Event Checklist Form ES-301-5 Group 4 (I6, I7, R7)
Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios S1 S2 S3 S4 T
O T
A L
M I
N I
M U
M(*)
CREW POSITION CREW POSITION CREW POSITION CREW POSITION S
R O
A T
C B
O P
S R
O A
T C
B O
P S
R O
A T
C B
O P
S R
O A
T C
B O
P R
I U
RO SRO-I6 SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,3,4,6 1,4,7,8 1,2,3,5,7 13 4
4 2
MAJ 5
5 4
3 2
2 1
TS 3,4 1,2 4
0 2
2 RO SRO-I7 SRO-U RX 0
1 1
0 NOR 0
1 1
1 I/C 2,3,4,5,6 1,2,3,4,6, 7,8 1,3,5,7 16 4
4 2
MAJ 5
5 4
3 2
2 1
TS 2,4 2
0 2
2 RO7 SRO-I SRO-U RX 1
1 1
1 0
NOR 0
1 1
1 I/C 2,4 2,3,4,6,7, 8
1,2,3,5,6, 7
14 4
4 2
MAJ 5
5 4
3 2
2 1
TS 0
2 2
RO SRO-I SRO-U RX 1
1 0
NOR 1
1 1
I/C 4
4 2
MAJ 2
2 1
TS 0
2 2
Instructions:
- 1.
Check the applicant level and enter the operating test number and Form ES D 1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at the controls (ATC) and balance of plant (BOP) positions. Instant SROs (SRO I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one for one basis.
- 3.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right hand columns.
- 4.
For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO I applicants in either the ATC or BOP position to best evaluate the SRO I in manipulating plant controls.
ES-301, Page 26 of 27 Rev 0
ES-301 Transient and Event Checklist Form ES-301-5 Group x (In, In, Rn, Rn)
Facility: Diablo Canyon Date of Exam: Feb 24, 2020 Operating Test Number: L181 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios S5 (Spare)
N-Day N-Day N-Day T
O T
A L
M I
N I
M U
M(*)
CREW POSITION CREW POSITION CREW POSITION CREW POSITION S
R O
A T
C B
O P
S R
O A
T C
B O
P S
R O
A T
C B
O P
S R
O A
T C
B O
P R
I U
RO SRO-I SRO-U RX 1
1 0
NOR 1
1 1
I/C 1,2,3,4,8, 9
2,4,8,9 1,3,4,7,8, 9
4 4
2 MAJ 5,6 5,6 5,6 2
2 1
TS 1,3 0
2 2
RO SRO-I SRO-U RX 1
1 0
NOR 1
1 1
I/C 4
4 2
MAJ 2
2 1
TS 0
2 2
RO SRO-I SRO-U RX 1
1 0
NOR 1
1 1
I/C 4
4 2
MAJ 2
2 1
TS 0
2 2
RO SRO-I SRO-U RX 1
1 0
NOR 1
1 1
I/C 4
4 2
MAJ 2
2 1
TS 0
2 2
Instructions:
- 1.
Check the applicant level and enter the operating test number and Form ES D 1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at the controls (ATC) and balance of plant (BOP) positions. Instant SROs (SRO I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2.
Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one for one basis.
- 3.
Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right hand columns.
- 4.
For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO I applicants in either the ATC or BOP position to best evaluate the SRO I in manipulating plant controls.
ES-301, Page 26 of 27 Rev 0
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-01 r0.docx Page 1 of 3 Rev 0 Facility:
Diablo Canyon (PWR)
Scenario No:
1 Op-Test No:
L181 NRC Examiners:
Operators:
Initial Conditions:
2% with MFW in service, aligned to Start-Up Power. MOL with CFCU 1-1 OOS.
Turnover:
In OP L-3, performing step 6.28, raising power to 8%.
Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 N/A R (ATC, SRO)
Raise reactor power from 2% to 8% OP L-3, sec 6.28.
2 VLV_CVC22_2.5 delay=0 ramp=15 I (ALL)
Regen Hx Isolation Valve, LCV-459, fails to mid-position (OP AP-18).
3 H_V1_034M_1, XMT_VEN6_3, XMT_VEN7_3, XMT_VEN8_3 TS, C (BOP, SRO)
CFCU 1-2 high stator/bearing temperature due to low CCW flow (AR PK01-21, TS 3.6.6.C).
4 RLY_PPL63_2 OPEN RLY_PPL59_2 OPEN TS, I (ALL)
SSPS relay actuation causes inadvertent start of TDAFW pump and blowdown sample isolation valves to close (AR PK04-03, OP D-1:III, OP1.DC10; TS 3.7.5.B).
5 MAL_MSS4 1.57E+07 ramp=30 M (ALL) MSLB outside containment.
6 VLV_MSS7_2, VLV_MSS8_2, VLV_MSS9_2, VLV_MSS10_2 1 C (ALL)
All MSIVs fail open; No manual close for FCV-42.
7 MAL_PPL3B BOTH C (BOP)
Safety Injection, Train B fails to actuate.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-01 r0.docx Page 2 of 3 Rev 0 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4)
Actual Attributes
- 1. Total malfunctions (5-8) (Events 2,3,4,5,6,7) 6
- 2. Malfunctions after EOP entry (1-2) (Events 6,7) 2
- 3. Abnormal events (1-4) (Events 2,3,4) 3
- 4. Major transients (1-2) (Event 5) 1
- 5. EOPs entered/requiring substantive actions (1-2) (E-2, E-1.1) 2
- 6. EOP contingencies requiring substantive actions (0-2) 0
- 7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S1CT-1) Shutdown TD AFW pump prior to Steam Generator Overfill (S/G wide range greater than 100%) by either:
Closing the LCVs to the individual S/Gs OR Closing steam supply valves FCV-37 and FCV-38 to leads 1 and 2 respectively OR Directing FCV-95 closed in the field Carryover into the steam lines can result in damage to downstream piping and valves, placing the secondary heat sink at risk. High steam generator level can also result in reactivity excursions due to excessive cooldown of the primary system.
- Tech Spec 3.3.2 Basis Documentation (S1CT-2) Stop uncontrolled RCS cooldown before a severe challenge to Integrity Safety Function develops (magenta path on F-0.4 RCS Integrity) as follows:
- Verify FCV-42 closed (S/G 1-2 steamline isolation).
- Close/verify closed S/G Blowdown isolation valves FCV-761, FCV-154, and FCV-248.
- Verify all steam dumps closed.
- Isolate feed flow to S/G 1-2 by closing/verifying closed LCV-107 and LCV-111.
(Note: LCV-107 is critical only when TDAFW pump is running or capable of an autostart).
- Isolate steam flow from S/G 1-2 by closing/verifying closed FCV-37.
- Throttling Feed flow to S/Gs 1-1, 1-3, and 1-4 while maintaining the minimum heatsink requirements (435 gpm until S/G NR level is greater than 15% in one non-faulted S/G).
An event or series of events which leads to a relatively rapid and severe reactor vessel downcomer cooldown can result in a thermal shock to the vessel wall that may lead to a small flaw, which may already exist in the vessel wall, growing into a larger crack. The growth or extension of such a flaw may lead, in some cases (where propagation is not stopped within the wall),
to a loss of vessel integrity
- Background Information for WOG Emergency
Response
Guideline (S1CT-3) Terminate SI prior to rupture of PRT by closing 8801A/B and/or 8803A/B.
(Note: CT is met by closing either 8801A/B OR 8803A/B.)
Failure to terminate ECCS flow when SI termination criteria are met results in overfill of the Pressurizer and the eventual rupture of the PRT. This constitutes the avoidable degradation of the RCS as a fission product barrier.
- Westinghouse Owners Group WCAP-17711-NP Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.
L181 NRC ES-D-1-01 r0.docx Page 3 of 3 Rev 0 SCENARIO
SUMMARY
- NRC #1
- 1.
Control rods are used to raise power from 2% to 8% OP L-3, Secondary Plant Startup, step 6.28. ATC operator complies with 1 step pull and wait procedural requirement while monitoring relevant controls and diverse indicators. Shift Foreman provides reactivity oversight.
- 2.
Regen Hx Isolation Valve, LCV-459, drifts to mid-position causing letdown orifice valve 8149C to close.
Shift Foreman enters OP AP-18, Letdown Line Failure. Excess Letdown is established per OP B-1A:IV CVCS - Excess Letdown - Place In Service and Remove From Service.
- 3.
CFCU 1-2 has a loss of CCW flow due to debris migration causing stator and motor bearing temperatures to rise rapidly and bring in annunciator alarm PK01-21, Contmt Fan Clr. Reactor operators identify low flow indications on vertical boards and rapidly rising stator/bearing temperatures using plant process computer trends. The crew secures the CFCU to prevent motor damage and contacts maintenance/engineering for assistance. Shift Foreman enters TS 3.6.6 Condition C, one required CFCU system inoperable such that a minimum of two CFCUs remain OPERABLE (7 day).
- 4.
SSPS relay actuation results in Turbine Driven AFW (TDAFW) Pump Steam Supply Isolation Valve, FCV-95, failing open and isolation of half of the blowdown sample valves inside and outside containment. S/G levels rise and RCS temperature lowers. FCV-95 cannot be closed and the crew must isolate the TDAFW Pump by closing the LCVs to the individual S/Gs or by closing steam supply valves FCV-37 and FCV-38 from leads 2 and 3 respectively, or by directing FCV-95 manually closed in the field (S1CT-1) Shutdown TD AFW pump prior to Steam Generator Overfill. Shift Foreman implements TS 3.7.5.B, AFW System for one AFW train inoperable (72 hrs).
- 5.
A main steamline break develops downstream of the Main Steam Isolation Valves, outside containment.
The crew identifies the need to isolate the Main Steam Isolation Valves and perform a safety injection (SI) based on pressurizer pressure and level lowering rapidly. Shift Foreman directs a reactor trip and SI and enters EOP E-0, Reactor Trip or Safety Injection.
- 6.
Train B of Safety Injection fails to actuate, requiring the crew to perform numerous manual alignments and pump starts as part of Appendix E. Appendix E. ESF Auto Actions, Secondary and Auxiliaries Status.
- 7.
All four main steam isolation valves fail open. Steam leads 1, 3, and 4 may be closed from the control room, but lead 2 (FCV-42) requires field action. The crew transitions to EOP E-2, Faulted Steam Generator Isolation to isolate S/G 1-2 and dispatches an operator to locally close FCV-42 as part of the critical task to stop the uncontrolled cooldown (S1CT-2) Stop uncontrolled cooldown before a severe challenge (magenta path ) develops on F-0.4 RCS Integrity.
- 8.
The crew transitions to EOP E-1.1, SI Termination where they complete the final critical task of the scenario (S1CT-3) Terminate SI prior to rupture of PRT.
The scenario is terminated once the final critical task is complete.
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-02 r0.docx Page 1 of 4 Rev 0 Facility:
Diablo Canyon (PWR)
Scenario No:
2 Op-Test No:
L181 NRC Examiners:
Operators:
Initial Conditions:
75% Power, MOL with AFW 1-2 cleared for a bearing oil leak Turnover:
At 75% power for SCCW HX Clearance Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 PZ04PRZ_PRH11_..
RCARRL_IPC455JXTVSP 0 I (ATC, SRO)
Pressurizer heater control problem causes RCS pressure to slowly lower (AR PK05-16, OP AP-13).
2 XMT_PZR24_3 ramp=1 TS, I (BOP, SRO)
PT-474, Pressurizer Pressure Transmitter, Fails Low ( OP AP-5, TS 3.3.1.E, M, 3.3.2.D, 3.4.11,.B).
3 XMT_CVC2_3 ramp=75 I(BOP, SRO)
PT-135 Fails High causing letdown pressure control valve to go full open (AR PK04-21).
4 MAL_RCS4H 30.0 TS, C (ALL) 30 gpm SGTL on loop 4; plant shutdown required (OP AP-3, AP-25, TS 3.4.13.B).
5 MAL_RCS4H 300.0 ramp=60 M (ALL) Tube leak grows to 300 gpm rupture during ramp offline.
6 RC41SW_52VE3_CSTA_SWIT 1 cd='H_V5_195B_1'...
RCPs trip off shortly after transferring to startup power.
7 MAL_EPS4D_2 DIFFERENTIAL cd='h_v4_221r_1' C (ALL) 4kV Bus G differential trip on transfer to startup power.
8 VLV_PZR5_1 1 cd='H_V2_014M_1 LT 1915' delay=8 VLV_CVC17_1 1 VLV_CVC27_1 1 C (ALL)
PORV PCV-456 fails closed.
Aux spray valves 8145 and 8148 fail closed.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-02 r0.docx Page 2 of 4 Rev 0 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4)
Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,2,3,4,5,6,7,8) 8
- 2. Malfunctions after EOP entry (1-2) (Events 6,7,8) 3
- 3. Abnormal events (1-4) (Events 1,2,3,4) 4
- 4. Major transients (1-2) (Event 5) 1
- 5. EOPs entered/requiring substantive actions (1-2) (E-3, ECA-3.3) 2
- 6. EOP contingencies requiring substantive actions (0-2) (ECA-3.3) 1
- 7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S2CT-1) Isolate the ruptured steam generator from the intact steam generators prior to commencing cooldown of the RCS in step 9.c (40% steam dumps) or 10.b (10%
steam dump) by completing the following:
Isolate feedwater by ensuring closed:
LCV-109 (TDAFW Level Control Valve)
LCV-113 (MDAFW Level Control Valve)
SG inventory increase leads to water release through the S/G PORV or safety valve(s) or to SG overfill, which would seriously compromise the SG as a fission-product barrier and complicate mitigation.
W Margin to Overfill (CN-CRA-05-53 Rev1)
W Offsite Doses (CN-CRA-05-54)
SGTR UFSAR 15.4.3 WCAP-17711-NP (S2CT-2) Perform RCS cooldown at maximum rate to CETC target temperature specified in E-3, step 6, using steam dumps such that RCS subcooled margin still exists following the cooldown.
For 40% steam dumps, maximum rate limit is 120 psi/min (PPC value). Above this, main steam line isolation will occur. Operator should attempt highest rate possible without getting main steam line isolation (not critical). If steam line isolation occurs, maximum rate cooldown requires 10% steam dumps on intact S/Gs to be at least 90% open.
Transition to contingency procedures to address inadequate subcooling or Pressurized Thermal Shock conditions results in delaying RCS depressurization and SI termination. This delay allows excess inventory in the ruptured S/G to continue to increase, with the potential of challenging SG overpressure components or causing an overfill condition to occur.
W Margin to Overfill (CN-CRA-05-53 Rev1)
SGTR UFSAR 15.4.3 WCAP-17711-NP (S2CT-3) Restore RCS pressure control by restarting RCP 1-2 and returning from EOP contingency procedure (ECA-3.3) to normal EOP flow path (E-3)
Failure to establish RCS pressure control results in safety injection termination occurring at a significantly higher RCS pressure. Avoidable primary to secondary leakage into a ruptured steam generator complicates results in the unnecessary continuation of a degraded condition beyond that irreparably introduced by the scenario.
- W Margin to Overfill (CN-CRA-05-53 Rev1)
- WCAP-17711-NP Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.
L181 NRC ES-D-1-02 r0.docx Page 3 of 4 Rev 0 SCENARIO
SUMMARY
- NRC #2
- 1.
The firing rate for Pressurizer Proportional Heaters fail to minimum and RCS pressure begins to lower slowly. Backup heater relay pc455jx fails to actuate when RCS pressure reaches 2210 psig which also results in a failure of annunciator AR PK05-16, PZR PRESSURE HI/LO activation. Pressurizer pressure control HC-455K output signal indicates that backup heaters should be on and proportional heaters should be operating at the maximum firing rate. When RCS pressure reaches 2185 psig, AR PK05-16, PZR PRESSURE HI/LO alarms based on Pressurizer pressure going below the PORV interlock setpoint. The crew follows AR PK05-16 guidance to manually turn on all backup pressurizer heaters and follow up with the actions of OP AP-13, Malfunction of Reactor Pressure Control System, to restore pressure to normal using manual control. The crew may diagnose the failure prior to the annunciator response activating and enter OP AP-13 directly which also directs turning on all backup heaters. If RCS pressure falls below 2175 psig, the Shift Foreman will enter Tech Spec 3.4.1 for RCS departure from nucleate boiling limits (2 hr).
- 2.
PT-474, Pressurizer Pressure Transmitter, Fails low bringing in multiple Annunciator Alarms. There is no transient associated with this failure, but the failure has significant Operational implications due to its input function as part of various Reactor Protection logic schemes. When failed low, PT-474s interlock function prevents Pressurizer PORVs PCV-455C and PCV-474 from opening on a valid high pressure signal; only PCV-456 will still function. The Shift Foreman may elect to enter any of the associated Annunciator Response alarms, but in all cases, will be directed to OP AP-5, Malfunction of Eagle-21 Protection or Control Channel, which provides information regarding indications, controls, and a listing of the associated Tech Specs:
TS 3.3.1.E, PC-474C High Press Trip & TC 441C OT Delta T Trip (72 hrs).
TS 3.3.1.M, PC 474A Low Press Trip (72 hrs).
TS 3.3.2.D, PC 474D Low Press S.I. (72 hrs).
TS 3.4.11, PC 474B PORV Press Interlock o PCV-474 (non-class I), 3.4.11.B1 & B2 to close & remove power from associated block valve (1 hr) o PCV-455C (class I), 3.4.11.B1 & B2 to close & remove power from associated block valve (1 hr);
3.4.11.B3 to return to OPERABLE status (72 hrs).
- 3.
PT-135, Transmitter for Letdown Pressure Control Valve, fails High causing letdown pressure control valve to go full open and letdown flow to rise. AR PK04-21, LETDOWN PRESS / FLOW TEMP comes into alarm for Letdown Heat Exchanger Outlet Pressure High as a result of the failed transmitter, while actual letdown pressure lowers to approximately 90 psig as a result of full open control valve response.
Letdown flow increases approximately 8 gpm above normal, resulting in a charging/letdown mismatch.
Procedural guidance in AR PK04-21 directs crew to take manual control of PCV-135. Crew performs diagnostic brief to determine nature of the malfunction as well as actions required to restore letdown pressure back to normal band.
- 4.
Steam Generator 1-4 develops a 30 gpm tube leak as indicated by rising counts on various radiation monitors. The crew enters OP AP-3, Steam Generator Tube Failure. Shift Foreman determines TS 3.4.13.B, RCS Operational Leakage applies and enters OP AP-25, Rapid Load Reduction or Shutdow for the ramp off-line.
- 5.
During the ramp the tube leak develops into a 300 gpm rupture. The crew determines the leak is substantial in size based on a rapid drop in pressurizer level. The Shift Foreman directs a reactor trip and safety injection and the crew enters EOP E-0, Reactor Trip or Safety Injection.
L181 NRC ES-D-1-02 r0.docx Page 4 of 4 Rev 0
- 6.
On the transfer to start up power, 4 kV bus G experiences a differential fault. The crew also notes all four RCPs trip off on the transfer, but power to the pumps remains available.
- 7.
The crew transitions to EOP E-3, Steam Generator Tube Rupture, based on RM-74 and rising S/G 1-4 level, where they address the following critical tasks:
(S2CT-1) Isolate the ruptured steam generator from the intact steam generators prior to commencing cooldown.
(S2CT-2) Perform RCS cooldown at maximum rate to CETC target temperature.
- 8.
The crew discovers they are unable to depressurize the RCS based on a combination of equipment failures:
With no running RCPs, normal pressurizer spray is ineffective.
PCV-456 has failed closed and will not opened.
Block valve 8000B, associated with PCV-455C was closed during an earlier event and cannot be opened due loss of bus G.
Block valve 8000A, associated with PCV-474 was also closed during the earlier event and will not open.
Aux Spray valves 8145 and 8148 are failed closed as well.
- 9.
The crew transitions to EOP ECA-3.3, SGTR Without Pressurizer Pressure Control, where they complete critical task (S2CT-3) Restore RCS pressure control by restarting RCP 1-2. Following procedural guidance, the crew determines normal spray is available and returns to EOP E-3, Steam Generator Tube Rupture.
The scenario is terminated once the crew has started RCP 1-2 and transitioned back into E-3.
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-03 r0.docx Page 1 of 3 Rev 0 Facility:
Diablo Canyon (PWR)
Scenario No:
3 Op-Test No:
L181 NRC Examiners:
Operators:
Initial Conditions:
100% Power, MOL with AFW 1-2 cleared Turnover:
Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 PMP_CVC3_2 OVERLOAD_DEV_FAIL TS, C (ALL)
CCP 1-3 OC Trip (OP AP-17, ECG 8.1.A).
2 AS01ASW_ASP11_MTFSEIZUR 1 TS, C (BOP, SRO)
ASW Pp 1-1 Seizes; Pp 1-2 can not be started (OP AP-10, TS 3.0.3).
3 MAL_CWS3A 80 MAL_CWS3B 75 MAL_CWS1A 0.15 ramp=120 MAL_CWS1B 0.15 ramp=150 C (ALL)
High DP on Intake Screens; ramp required (AR PK13-01, OP AP-7, OP AP-25).
4 PMP_CWS1_2 OVERLOAD_DEV_FAIL cd='smss lt 1140' PMP_CWS2_2 OVERLOAD_DEV_FAIL cd='smss lt 1080' M
(ALL)
Both Circ Water pumps trip off during ramp, requiring Reactor Trip (OP AP-7).
5 MAL_EPS4C_2 DIFFERENTIAL cd='fnispr lt 5' delay=30 C (ALL)
Vital 4kV Bus F differential trip.
6 VLV_PZR4_2 0.3 cd='jpplp4' del ay=60 C (BOP)
Pressurizer PORV PCV-455C fails slightly open on trip requiring manual isolation by associated block valve 7
MAL_AFW1 1 cd='fnispr lt 5' C (ALL)
Turbine driven AFW pump overspeed trip.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-03 r0.docx Page 2 of 3 Rev 0 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4)
Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,2,3,4,5,6,7) 7
- 2. Malfunctions after EOP entry (1-2) (Events 5,6,7) 3
- 3. Abnormal events (1-4) (Events 1,2,3) 3
- 4. Major transients (1-2) (Event 4) 1
- 5. EOPs entered/requiring substantive actions (1-2) (E-0.1, FR-H.1) 2
- 6. EOP contingencies requiring substantive actions (0-2) (FR-H.1) 1
- 7. Critical tasks (2-3)(See description below) 2 Critical Task Justification Reference (S3CT-1) Close the block MOV upstream of the stuck open PORV prior rupture of the PRT.
The open PORV and block valve constitute the degradation of a fission product barrier. Closing the block valve is essential to safety since failure to do so results in the unnecessary continuation of the degraded condition.
- Westinghouse Owners Group WCAP-17711-NP (S3CT-2) Establish a secondary heat sink as indicated by:
WR level rising Core Exit Thermocouple temperatures lowering Prior to reaching bleed and feed criteria which is defined as wide range S/G level in any three S/Gs less than 18% [26%] AND narrow range S/G level in all four S/Gs less than 15% [25%] narrow range.
A loss of all feedwater transient is characterized by a depletion of secondary inventory and eventual degradation of secondary heat transfer capability.
As secondary heat transfer capability degrades, core decay heat generation will increase RCS temperature and pressure causing loss of RCS inventory similar in nature to a small break loss of coolant accident. Failure to restore a secondary heat sink when it is possible to do so constitutes a significant reduction of safety margin beyond that irreparably introduced by the scenario.
- FR-H.1 Background Document (HFRH1BG),
Rev. 3.
Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.
L181 NRC ES-D-1-03 r0.docx Page 3 of 3 Rev 0 SCENARIO
SUMMARY
- NRC #3
- 1.
Charging Pump CCP 1-3 trips on over current. The crew responds by entering OP AP-17, Loss of Charging to restore normal charging and letdown. Shift Foreman enters ECG 8.1.A - Charging Pump No. 3 Inoperable (establish continuous fire watch immediately; restore to operable status within 7 days).
- 2.
ASW Pump 1-1 trips due to a seized shaft. Standby ASW Pump 1-2 fails to autostart and cannot be started manually. The Shift Foreman implements OP AP-10, Loss of Auxiliary Salt Water and cross-ties to the Unit 2 ASW system via the ASW cross-tie valve FCV-601. Shift Foreman enters T.S. 3.0.3 for two trains of ASW inoperable on Unit 1.
- 3.
Screen differential pressure begins to rise quickly, bringing in AR PK13-01, Bar Racks/Screens. Following annunciator guidance, the crew enters OP AP-7, Degraded Condenser, Section C: Traveling Screen Problem and begins to reduce load to 50% or less per OP AP-25, Rapid Load Reduction.
- 4.
Both Circ Water pumps trip off during ramp, requiring the crew to manually trip the Reactor oper OP AP-
- 7. The crew enters EOP E-0, Reactor Trip or Safety Injection and performs their immediate actions.
- 5.
On the trip, vital 4 kV bus F trips on differential. DRPI loses power, but crew is able to determine the reactor has tripped based on diverse indications (lowering reactor power and reactor trip breakers open). MDAFW Pump 1-3 is also lost due to the bus failure.
- 6.
Board operators also identify PCV-455C in mid-position. The valve will not close and must be isolated using the associated block valve 8000B (S3CT-1) Close the block MOV upstream of the stuck open PORV prior rupture of the PRT.
- 7.
The TDAFW pump trips on overspeed. Steam Generator levels are initially high enough to provide an adequate secondary side heat sink and the crew transitions to EOP E-0.1, Reactor Trip Response to stabilize the plant. Steam Generator levels slowly lower below the minimum required level of 15%
narrow range and the crew transitions to EOP FR-H.1, Response to Loss of Secondary Heat Sink. With the condenser unavailable, Condensate is used to restore a secondary side heat sink (S3CT-2) Establish a secondary heat sink.
The scenario is terminated once Critical Task S3CT-2 is complete
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-04 r0.docx Page 1 of 3 Rev 0 Facility:
Diablo Canyon (PWR)
Scenario No:
4 Op-Test No:
L181 NRC Examiners:
Operators:
Initial Conditions:
75% Power, MOL with D/G 1-2 OOS Turnover:
At 75% power for SCCW HX Clearance Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 XMT_RMS3_3 1E+06 delIA XMT_RMS3_3 2 delay=15 XMT_RMS3_3 1e+06 cd='H_RM_C08B_1' delay=15 TS, I (BOP, SRO)
RE-3 spurious spike followed by high failure (AR PK11-21, ECG 39.3.D).
2 XMT_CVC19_3 0.0 ramp=120 I (ATC, SRO)
LT-112 Fails Low (auto make-up) (OP AP-19, AP-5).
3 MAL_RCS3B.07 MAL_RCS3B.09 cd='h_v2_265r_1 or h_v2_264r_1' delay=60 ramp=10 TS, C (ALL) 70 gpm RCS leak on Loop 2; increased to 90 gpm on start of 2nd CCP (OP AP-1, TS 3.4.13.A).
4 MAL_SEI1 0.21 ramp=10 MAL_GEN4_3 TRIP delay=10 cd='jmlsei1' LOA_SYD6 OPEN delay=15 cd='jmlsei1' LOA_SYD7 OPEN delay=15 cd='jmlsei1' LOA_SYD8 OPEN delay=15 cd='jmlsei1' C
(ALL)
Seismic event causing Full Load Rejection (OP AP-2, AP-25).
5 MAL_ROD13 TRUE cd='fnispr lt 55' M
(ALL)
Auto rod control insertion failure requiring reactor trip (OP AP-2).
6 MAL_SYD2 0 cd='fnispr lt 5' delay=2 MAL_EPS4E_2 DIFFERENTIAL cd='h_v4_217r_1' MAL_DEG1C_2 NO_RESET cd='H_V4_224R_1' M
(ALL)
Loss of all A/C power.
7 VLV_AFW7_1 1 DelIA VLV_AFW7_1 2 cd='V3_219S_3' C
(BOP)
TDAFW Pump fails to autostart - manual start required.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-04 r0.docx Page 2 of 3 Rev 0 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4)
Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,2,3,4,5,6,7) 7
- 2. Malfunctions after EOP entry (1-2) (Event 7) 1
- 3. Abnormal events (1-4) (Events 1,2,3,4) 4
- 4. Major transients (1-2) (Events 5,6) 2
- 5. EOPs entered/requiring substantive actions (1-2) (ECA-0.0, ECA-0.2) 2
- 6. EOP contingencies requiring substantive actions (0-2) (ECA-0.0, ECA-0.2) 2
- 7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S4CT-1) Establish a minimum of 435 gpm AFW flow before S/G reach dryout condition of 10% WR by starting the turbine drive AFW pump (AFW Pp 1-1).
After the initial RCS cooldown, decay heat increases the RCS temperature. Without an adequate secondary heat sink to support natural circulation, the S/Gs could not support any significant plant cooldown. Thus the crew would lose the ability to delay the adverse consequences of core uncovery.
- WCAP-17711-NP, CT-23
- ECA-0.0 Background Document (HECA00BG),
Rev. 3.
(S4CT-2) Energize at least one vital AC bus from prior to implementation of FLEX strategies (ECA-0.0, step 10 RNO) associated with entry into Extended Loss of AC Power Event (ELAP) conditions Failure to restore vital AC power from an available source when available represents an unnecessary continuation of a degraded electrical condition and unnecessarily complicates the mitigation strategy
- WCAP-17711-NP, CT-24
- ECA-0.0 Background Document (HECA00BG),
Rev. 3.
(S4CT-3) Establish flow from at least one high head injection pump prior to transition out of ECA-0.2.
Failure to manually start at least one high/intermediate head injection pump under the postulated conditions constitutes misoperation or incorrect crew performance in which the crew does not prevent degraded core cooling capacity.
- WCAP-17711-NP, CT-7 Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.
L181 NRC ES-D-1-04 r0.docx Page 3 of 3 Rev 0 SCENARIO
SUMMARY
- NRC #4
- 1.
RE-3 (Oily Water Separator Effluent Line Radiation Monitor) spikes high momentarily, bringing in annunciator AR PK11-21, High Radiation. The crew identifies the signal as spurious and performs a reset of RM-3 per PK11-21, section 2.1.4. RE-3A fails high shortly after the reset and cannot be cleared. Shift Foreman enters ECG 39.3.D, Radioactive Liquid Effluent Monitoring Instrumentation (30 days).
- 2.
Volume Control Tank (VCT) level channel LT-112 fails low, causing a continuous (and erroneous) makeup signal. The crew diagnoses the level channel failure by comparing other VCT parameters, and by using OP AP-19, Malfunction of the Reactor Makeup Control System. The makeup system is secured, and makeup is accomplished (if needed) using manual mode (or enabling the auto mode for short periods).
Crew may elect to use OP AP-5, Malfunction of Eagle 21 Protection or Control Channel to take manual control of Makeup Control System.
- 3.
A 70 gpm RCS leak develops, requiring entry in OP AP-1, Excessive Reactor Coolant System Leakage.
Pressure. The leak increases to 90 gpm when a second charging pump is started and letdown must be isolated. VCT level can be maintained at the current leak rate, however, and the crew determines a plant shutdown is required. Shift Foreman enters TS 3.4.13.A, RCS Operational Leakage (4 hrs).
- 4.
A significant seismic event results in a full load rejection on Unit 1. The crew recognizes the condition based on numerous power level alarms and the ensuing secondary side transient. The crew monitors primary and secondary system responses, most notably rod control, steam dumps, and digital feedwater, to ensure all systems respond appropriately in automatic. Shift Foreman implements OP AP-2, Full Load Rejection to stabilize the plant.
- 5.
During the ramp down, automatic rod motion stops and the Shift Foreman directs a Reactor trip.
- 6.
Startup power is lost on the trip followed by a bus differential fault on vital 4kV bus H. Diesel Generator 1-3 trips and cannot be reset. The crew transitions to EOP ECA-0.0, Loss of All Vital AC Power.
- 7.
The turbine driven AFW pump fails to autostart requiring the crew to perform the critical task of manually starting the pump (S4CT-1) Establish a minimum of 435 gpm AFW flow before S/Gs reach dryout conditions.
- 8.
Energy Trading informs the crew that 230kV start up power is not available, but 500 kV is expected back shortly. The crew performs actions to isolate RCP seal cooling.
- 9.
Power is restored to vital 4kV buses F and G following the guidance of ECA-0.0, Appendix DD (Backfeed from 500kV Power) (S4CT-2) Energize at least one vital AC bus from prior to implementation of FLEX strategies.
- 10. The crew transitions to ECA-0.2, Loss of All AC Power with SI Required where they manually start safeguards equipment. The crew performs the final scenario critical task (S4CT-3) Establish flow from at least one high head injection pump.
The scenario is terminated once injection flow is established.
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-05 r0.docx Page 1 of 4 Rev 0 Facility:
Diablo Canyon (PWR)
Scenario No:
5 Op-Test No:
L181 NRC Examiners:
Operators:
Initial Conditions:
100% Power, MOL with CCP 1-3 cleared Turnover:
At 100% power. CCP 1-3 cleared for routine maintenance. CCP 1-1 I/S.
Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 CC01CCW_CCP11_MTFSHEAR 1 TS, C (BOP, SRO)
CCW Pp 11 Shaft Shear (AR PK01-11; TS 3.7.7.A).
2 MAL_CVC8A 100 ramp=60 C (ATC, SRO)
Seal Injection Filter 1-1 plugs causing reduction in charging flow to RCP seals (AR PK04-22).
3 MAL_PPL7J 1 TS, I (BOP, SRO)
Eagle 21 DFP-1 Halt in Rack 10 (OP AP-5; TS 3.3.1.E,M; 3.3.2.D, L; 3.4.11 ).
4 PK1421_0829 1 C (ALL)
Loss of Main Transformer Cooling (AR PK14-21, AP-
- 25) 5 MAL_SEI1 0.31 delay=0 ramp=15 MAL_PPL5A; PPL5B BOTH M (ALL)
Large seismic with no automatic or manual reactor trip (ATWS).
6 MAL_RCS3C 10.0 cd='jmlsei1' delay=10 ramp=60 M (ALL)
SBLOCA following seismic; ramps in over 60 seconds.
7 MAL_PPL1A FAILURE_TO_INIT MAL_PPL1B FAILURE_TO_INIT C (BOP)
Phase A - Train A and B fail to actuate requiring manual alignment.
8 MAL_SYD2 0 cd='jpplsia' delay=15 C (ALL)
Loss of Start-up power 15 seconds after Safety Injection 9
PMP_SIS2_2 OVERLOAD_DEV_FAIL cd='h_v4_218r_1' delay=3 PMP_CVC2_1 AS_IS BKR_EPS15 AS_IS BKR_EPS9_1 OVERCURRENT cd='H_V4_225R_1' MAL_AFW1 TRIP cd='h_v3_109m_1 gt 3000' C (ALL)
Combination of electrical and mechanical failures result in no high or intermediate injection along with degraded secondary side heat removal capabilities.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L181 NRC ES-D-1-05 r0.docx Page 2 of 4 Rev 0 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4)
Actual Attributes
- 1. Total malfunctions (5-8) (Events 1,2,3,4,5,6,7,8,9) 9
- 2. Malfunctions after EOP entry (1-2) (Events 7,8,9) 3
- 3. Abnormal events (1-4) (Events 1,2,3,4) 4
- 4. Major transients (1-2) (Event 5,6) 2
- 5. EOPs entered/requiring substantive actions (1-2) (E-1, FR-C.1) 2
- 6. EOP contingencies requiring substantive actions (0-2) (FR-C.1) 1
- 7. Critical tasks (2-3)(See description below) 3 Critical Task Justification Reference (S5CT-1) Trip the Reactor by manually de-energizing 480V Buses 13D and 13E within 90 seconds of AR PK04-11, Reactor Trip Initiate coming into alarm.
The safeguards systems that protect the plant during accidents are designed assuming that only decay heat and pump heat are being added to the RCS. Failure to manually trip the reactor causes a challenge to the subcriticality critical safety function beyond that irreparably introduced by the postulated conditions.
- Westinghouse Owners Group WCAP-17711-NP
- Calc G.2 Rev 5 (08151-2169)
- OP1.ID2, Time Critical Operator Actions Rev 12,
- 34.
(S5CT-2) Manually close containment isolation valves such that at least one valve is closed on each Phase A containment penetration before transition out of EOP E-
- 0.
Failure to perform the critical task leads to an unnecessary release of fission products to the auxiliary building, increasing the potential for release to the environment and reducing accessibility to vital equipment within the auxiliary building
- WCAP-17711-NP, CT-11 (S5CT-3) Depressurize Steam Generators to inject SI Accumulators to re-flood the core and clear the Extreme (red path) challenge to the Core Cooling critical status function.
RCS depressurization allows accumulator injection and a temporary restoration of Core Cooling.
Continued secondary depressurization allows for low head injection and a stable source of long term Core Cooling. Failure to depressurize the Steam Generators results in the crew having to rely on the lower priority action of sequentially starting RCPs which constitutes a significant reduction of safety margin beyond that irreparably introduced as part of the scenario's design.
- WCAP-17711-NP, CT-39 Per NUREG-1021, Appendix D, if an operator or crew significantly deviates from or fails to follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.
L181 NRC ES-D-1-05 r0.docx Page 3 of 4 Rev 0 SCENARIO
SUMMARY
- NRC #5
- 1.
AR PK01-11, CCW Pp 1-1 Recirc comes into alarm for FCV-606, CCW Pump 1-1 Recirc Valve, open. Crew identifies low pump amps on VB-1 and dispatches Nuclear Operator to investigate. Field reports no audible flow sound in spite of indications motor is running. CCW Pump 1-3 is started manually and CCW Pump 1-1 shutdown. TS 3.7.7.A, Vital Component Cooling Water (CCW) System, is entered for one loop of CCW inoperable (72 hrs).
- 2.
In-service Seal Injection Filter 1-1 plugs, reducing flow to RCP seals and bringing in AR PK04-22, RCP Seal Inj Fltr Delta-P Hi. Reactor Operators verify CCP seal cooling is still being maintained by CCW and ATC operator throttles RCP seal injection hand control valve, HCV-142, as needed to maintain pressurizer level. Shift Foreman establishes bands for pressurizer level and confirms field operators have been dispatched to swap seal injection filters.
- 3.
Eagle 21 experiences a Digital Filter Processor (DFP) halt on rack 10. Associated indicators PI-456, LI-460A, FI-415, FI-425, FI-435, FI-445 (VB2), and PR-445, LR-459 (CC2) fail as-is as well as control channels for PORV 456 (PT-456) and Pressurizer Level Control (LT-460). Crew responds per OP AP-5, Malfunction of Eagle 21 Protection or Control Channel. Shift Foreman reviews Tech Specs, entering:
TS 3.3.2.D, PC 456D Low Press SI (72 hrs).
TS 3.3.1.E, PC 456A High Press Trip (72 hrs).
TS 3.3.1.M, PC 456C Low Press Trip (72 hrs).
TS 3.3.1.M, LC 460A High Level Trip (72 hrs).
TS 3.3.1.M, FC-415(425,435,445) RCS Loop 1 (2,3,4) Flow (72 hrs).
TS 3.3.2.L, PC-456 B, P-11 (1 hr).
TS 3.4.11.B1, B2, & B3 PC-456E, to close & remove power from associated block valve (1 hr) and restore to operable (72 hrs).
- 4.
Crew responds to AR PK14-21, MAIN TRANSF. A nuclear operator is dispatched to investigate local alarms and reports back that NO cooling fans or oil pumps are running on the Main Bank C Transformer.
Shift Foreman enters OP AP-25, Rapid Load Reduction or Shutdown and directs a 50 MW/min power reduction while Maintenance and field Operators attempt to restore transformer cooling.
- 5.
A large earthquake (0.31 g) occurs during the ramp, but the reactor fails to trip automatically. The crew performs the immediate actions of EOP E-0, Reactor Trip or Safety Injection and successfully trips the reactor by opening the breakers for 480 V buses 13D and 13E to de-energize the control rod drive mechanism (CRDM) allowing control rods to fully drop into the core (S5CT-1) Trip the Reactor by manually de-energizing 480V Buses 13D and 13E.
- 6.
A SBLOCA occurs as a result of the earthquake, but both trains of Phase A fail to actuate. The crew performs manual alignment of Phase A containment isolation valves per Appendix E, ESF Auto Actions, Secondary and Auxiliaries Status (S5CT-2) Manually close containment isolation valves such that at least one valve is closed on each Phase A containment penetration.
- 7.
Startup power is lost shortly after Safety Injection initiates and a combination of electrical and mechanical failures result in the loss of both ECCS charging pumps and safety injection pumps.
Secondary heat removal is affected as well. The turbine driven AFW pump trips on overspeed and AFW pump 1-3 has no power due to a loss of 4kV bus F.
(continued)
L181 NRC ES-D-1-05 r0.docx Page 4 of 4 Rev 0 SCENARIO
SUMMARY
- NRC #5
- 8.
The crew proceeds through E-0, transitioning to E-1, Loss of Reactor or Secondary Coolant. A loss of subcooling and lowering RVLIS level eventually results in a magenta path followed quickly by a red path on the core cooling critical safety function. The crew transitions to FR-C.2, Response to Degraded Core Cooling briefly, and then on to FR-C.1, Response to Inadequate Core Cooling.
Following the guidance of FR-C.1, the crew performs the final critical task (S5CT-3) Depressurize Steam Generators to inject SI Accumulators to re-flood the core and clear the Extreme (red path).
The scenario is terminated once critical task S5CT-3 is complete.