ML19050A489
| ML19050A489 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/25/2019 |
| From: | Greg Werner Operations Branch IV |
| To: | Pacific Gas & Electric Co |
| References | |
| Download: ML19050A489 (47) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Rev. 11 Facility: Diablo Canyon Date of Exam: January 25, 2019 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
3 3
3 N/A 3
3 N/A 3
18 6
2 2
2 1
1 1
2 9
4 Tier Totals 5
5 4
4 4
5 27 10
- 2.
Plant Systems 1
2 2
3 3
2 3
3 3
2 2
2 28 5
2 1
1 1
1 0
1 1
1 1
1 1
10 3
Tier Totals 3
3 4
4 2
4 4
4 3
3 3
38 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
3 3
2 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 X
Knowledge of the reasons for the following as the apply to a reactor trip: (CFR 41.5 /41.10 / 45.6 /
45.13)
EK3.01 Actions contained in EOP for reactor trip 4.0 1
000008 (APE 8) Pressurizer Vapor Space Accident / 3 X
Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: (CFR 41.7 / 45.7)
AK2.01 Valves 2.7*
2 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 X
Knowledge of the reasons for the following responses as the apply to the Large Break LOCA:
(CFR 41.5 / 41.10 / 45.6 / 45.13)
EK3.09 Maintaining D/Gs available to provide standby power 4.2 3
000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X
Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup:
(CFR 41.7 / 45.5 / 45.6)
AA1.02 CVCS charging low flow alarm, sensor, and indicator 3.0 4
000025 (APE 25) Loss of Residual Heat Removal System / 4 X
Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: (CFR: 43.5 / 45.13)
AA2.04 Location and isolability of leaks 3.3*
5 000026 (APE 26) Loss of Component Cooling Water / 8 X 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10 /
43.5 / 45.13) 3.8 6
000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 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 7
000029 (EPE 29) Anticipated Transient Without Scram / 1 X
Knowledge of the operational implications of the following concepts as they apply to the ATWS:
(CFR 41.8 / 41.10 / 45.3)
EK1.03 Effects of boron on reactivity 3.6 8
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 000055 (EPE 55) Station Blackout / 6 Replaced KA X
2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR:
41.10 / 43.2 / 45.13) 2.2.37 - Ability to determine operability and/or availability of safety related equipment.
3.1 3.6 9
ES-401 3
Form ES-401-2 Rev. 11 000056 (APE 56) Loss of Offsite Power / 6 X
Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: CFR 41.8 / 41.10 / 45.3)
AK1.03 Definition of subcooling: use of steam tables to determine it 3.1*
10 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 11 000058 (APE 58) Loss of DC Power / 6 X
Ability to operate and / or monitor the following as they apply to the Loss of DC Power: (CFR 41.7 /
45.5 / 45.6)
AA1.03 Vital and battery bus components 3.1 12 000062 (APE 62) Loss of Nuclear Service Water / 4 X
Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:
(CFR: 43.5 / 45.13)
AA2.04 The normal values and upper limits for the temperatures of the components cooled by SWS 2.5 13 000065 (APE 65) Loss of Instrument Air / 8 Replaced KA X
Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: (CFR 41.5,41.10 / 45.6 / 45.13)
AA1.04 Emergency air compressor AA1.05 - RPS 3.5*
3.3 14 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X
Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)
AA2.08 Criteria to trip the turbine or reactor 4.3 15 (W E04) LOCA Outside Containment / 3 X
Knowledge of the interrelations between the (LOCA Outside Containment) and the following: (CFR: 41.7
/ 45.7)
EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
3.5 16 (W E11) Loss of Emergency Coolant Recirculation / 4 X
Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following: (CFR: 41.7 / 45.7)
EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
3.6 17 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 X
Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink) (CFR: 41.8 / 41.10, 45.3)
EK1.1 Components, capacity, and function of emergency systems.
3.8 18 K/A Category Totals:
3 3
3 3
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 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 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
Knowledge of the reasons for the following responses as they apply to the Fuel Handling Incidents:
(CFR 41.5,41.10 / 45.6 / 45.13)
AK3.03 Guidance contained in EOP for fuel handling incident 3.7 19 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 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) 4.5 20 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 Replaced KA E14 Loss of Containment Integrity X
Knowledge of the operational implications of the following concepts as they apply to Loss of Containment Integrity: (CFR 41.8
/ 41.10 / 45.3)
AK1.01 Effect of pressure on leak rate EK 1.1 Components, capacity, and function of emergency systems.
2.6 3.3 21 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 000076 (APE 76) High Reactor Coolant Activity / 9 X
Knowledge of the interrelations between the High Reactor Coolant Activity and the following:
(CFR 41.7 / 45.7)
AK2.01 Process radiation monitors 2.6 22 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 X
Knowledge of the operational implications of the following concepts as they apply to the (Reactor Trip or Safety Injection/Rediagnosis) (CFR: 41.8
/ 41.10 / 45.3)
EK1.1 Components, capacity, and function of emergency systems.
3.1 23
ES-401 5
Form ES-401-2 Rev. 11 (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)
EA1.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
3.1 24 (W E15) Containment Flooding / 5 X
2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 /
43.5 / 45.12) 4.2 25 (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 X
Knowledge of the interrelations between the (LOCA Cooldown and Depressurization) and the following: (CFR: 41.7 / 45.7)
EK2.2 Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
3.7 26 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 X
Ability to determine and interpret the following as they apply to the (Natural Circulation Operations)
(CFR: 43.5 / 45.13)
EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
3.1 27 (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 2
1 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 28 004 (SF1; SF2 CVCS) Chemical and Volume Control X
Ability to monitor automatic operation of the CVCS, including: (CFR: 41.7 / 45.5)
A3.16 Interpretation of emergency borate valve position indicating lights 3.8 29 005 (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 30 006 (SF2; SF3 ECCS) Emergency Core Cooling X
Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.09 Valve positioning on safety injection signal 3.9 31 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X
Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: (CFR: 41.7 / 45.6)
K3.01 Containment 3.3 32 008 (SF8 CCW) Component Cooling Water X
Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: (CFR: 41.2 to 41.9 /
45.7 to 45.9)
K1.05 Sources of makeup water 3.0 33 010 (SF3 PZR PCS) Pressurizer Pressure Control Replaced KA X
Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.06 CVCS K1.02 ESFAS 2.9 3.9 34 012 (SF7 RPS) Reactor Protection X
Ability to monitor automatic operation of the RPS, including: (CFR: 41.7 / 45.5)
A3.07 Trip breakers 4.0 35 013 (SF2 ESFAS) Engineered Safety Features Actuation X
Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS: (CFR: 41.7 / 45.5 to 45.8)
K6.01 Sensors and detectors 2.7*
36 022 (SF5 CCS) Containment Cooling 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) 3.8 37 025 (SF5 ICE) Ice Condenser
ES-401 7
Form ES-401-2 Rev. 11 026 (SF5 CSS) Containment Spray Replaced KA X
Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.08 Automatic swapover to containment sump suction for recirculation phase after LOCA (RWST low-low level alarm)
K4.01 - Source of water for CSS, including recirculation phase after LOCA 4.1*
4.2 38 039 (SF4S MSS) Main and Reheat Steam X
Knowledge of the operational implications of the following concepts as the apply to the MRSS: (CFR: 441.5 / 45.7)
K5.05 Bases for RCS cooldown limits 2.7 39 059 (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)
A2.12 Failure of feedwater regulating valves 3.1*
40 061 (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)
K6.01 Controllers and positioners 2.5 41 062 (SF6 ED AC) AC Electrical Distribution X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 / to 45.8)
A4.01 All breakers (including available switchyard) 3.3 42 063 (SF6 ED DC) DC Electrical Distribution X
Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.01 Major DC loads 2.9*
43 064 (SF6 EDG) Emergency Diesel Generator X
Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G 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)
A2.09 Synchronization of the ED/G with other electric power supplies 3.1 44 073 (SF7 PRM) Process Radiation Monitoring X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.03 Check source for operability demonstration 3.1 45 076 (SF4S SW) Service Water X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: (CFR: 41.5 / 45.5)
A1.02 Reactor and turbine building closed cooling water temperatures 2.6*
46 078 (SF8 IAS) Instrument Air X
2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.
(CFR: 41.10 / 45.12) 4.6 47
ES-401 8
Form ES-401-2 Rev. 11 103 (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*
48 053 (SF1; SF4P ICS*) Integrated Control 008 (SF8 CCW) Component Cooling Water X
Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following:
(CFR: 41.7)
K4.01 Automatic start of standby pump 3.1 49 012 (SF7 RPS) Reactor Protection X
Knowledge of the effect of a loss or malfunction of the following will have on the RPS: (CFR: 41.7 / 45/7)
K6.03 Trip logic circuits 3.1 50 026 (SF5 CSS) Containment Spray X
Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.02 MOVs 2.7*
51 061 (SF4S AFW)
Auxiliary/Emergency Feedwater X
Knowledge of the operational implications of the following concepts as the apply to the AFW: (CFR: 41.5 / 45.7)
K5.03 Pump head effects when control valve is shut 2.6 52 063 (SF6 ED DC) DC Electrical Distribution X
Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: (CFR: 41.5
/ 45.5)
A1.01 Battery capacity as it is affected by discharge rate 2.5 53 064 (SF6 EDG) Emergency Diesel Generator X
Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following: (CFR: 41.7 / 45.6)
K3.03 ED/G (manual loads) 3.6 54 073 (SF7 PRM) Process Radiation Monitoring Rejected K4.02 - not a DCPP feature.
X Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
K4.02 Letdown isolation on high-RCS activity K4.01 - Release Terminatiion 3.3*
4.0 55 K/A Category Point Totals:
2 2
3 3
2 3
3 3
2 2
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 X
Knowledge of the physical connections and/or cause-effect relationships between the CRDS and the following systems: (CFR: 41.2 to 41.9 /
45.7 to 45.8)
K1.03 CRDM 3.4 56 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control X
Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.01 Charging pumps 3.1 57 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation X
Knowledge of the operational implication of the following concepts as they apply to the NNIS:
(CFR: 41.5 / 45.7)
K5.01 Separation of control and protection circuits 2.7*
58 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
A4.03 Location and operation of hydrogen sampling and analysis of containment atmosphere, including alarms and indications 3.1 59 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment X
Ability to monitor automatic operation of the Fuel Handling System, including: (CFR: 41.7 /
45.5)
A3.02 Load limits A3.03 High Flux at Shutdown 2.5*
2.9 60 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X
Knowledge of the effect that a loss or malfunction of the SDS will have on the following: (CFR: 41.7 / 45.6)
K3.04 Reactor power 3.5 61 045 (SF 4S MTG) Main Turbine Generator X
Ability to (a) predict the impacts of the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)
A2.08 Steam dumps are not cycling properly at low load, or stick open at higher load (isolate and use atmospheric reliefs when necessary)
A2.17 Malfunction of electrohydraulic control 2.8 2.7 62 055 (SF4S CARS) Condenser Air Removal 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 63
ES-401 10 Form ES-401-2 Rev. 11 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring X
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including: (CFR: 41.5 / 45.5)
A1.01 Radiation levels 3.4 64 075 (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 65 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:
1 1
1 1
0 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: Diablo Canyon Date of Exam: January 25, 2019 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.21 Ability to verify the controlled procedure copy.
3.5*
66 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.
3.9 67 2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.
4.1 68 Subtotal
- 2. Equipment Control 2.2.12 Knowledge of surveillance procedures.
3.7 69 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings.
3.5 70 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.
4.2 71 Subtotal
- 3. Radiation Control 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.
3.5 72 2.3.11 Ability to control radiation releases.
3.8 73 Subtotal
- 4. Emergency Procedures/Plan 2.4.32 Knowledge of operator response to loss of all annunciators.
3.6 74 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.
4.2 75 Subtotal Tier 3 Point Total 10 7
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /
Group Randomly Selected K/A Reason for Rejection RO - T2/G1 010 K1.06 Not a testable tie between Pressurizer PCS and CVCS.
Randomly selected K1.02 (IR 3.9) as replacement.
RO T2G1 026 K4.06 No automatic swapover to the containment sump for containment spray. It is manually aligned to be supplied from RHR.
Randomly selected K4.01 (IR 4.2)
RO T2/G2 045 A2.08 Too close to previous KA (041). Randomly selected A2.17 (IR 2.7)
RO T1/G1 EPE 055 G2.2.36 Unable to write question at RO level. Randomly selected 2.2.37 (IR 3.6)
RO T1/G1 APE 065 AA1.04 Not applicable to DCPP, no emergency AC units. Randomly selected AA1.05 (IR 3.3)
RO T1/G2 APE 069 AK1.01 Unable to write a satisfactory question to the KA. E/APE selected 000069 (APE 69; W E14). Because there are no other AK1 statements for APE 069, shifted to E14 and randomly selected EK1.1 (IR 3.3)
RO T2/G2 034 A3.02 Unable to write a satisfactory question to KA - low operational validity as the fuel movement is a separate qual, and normally performed by non-licensed DCPP personnel.
A3.03 (IR 2.9) selected - this is a parameter monitored by the RO in the Control Room during fuel movement.
ES-401 PWR Examination Outline Form ES-401-2 Rev. 11 Facility: Diablo Canyon Date of Exam: January 25, 2019 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 2
1 3
Tier Totals 38 5
3 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 76 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X 2.4.6 Knowledge of EOP mitigation strategies.
(CFR: 41.10 / 43.5 / 45.13) 4.7 77 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 X
Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: (CFR: 43.5 / 45.13)
AA2.06 Existence of proper RHR overpressure protection 3.4*
78 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 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)
EA2.16 Actions to be taken if S/G goes solid and water enters steam line 4.6 79 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer Uncontrolled Depressurization of all Steam Generators/ 4 Replaced KA X
2.4.44 Knowledge of emergency plan protective action recommendations. (CFR: 41.10 / 41.12 / 43.5
/ 45.11) 2.4.6 - Knowledge of EOP mitigation strategy 4.4 4.7 80 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X
2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (CFR: 41.10 / 43.5 /
45.13) 4.1 81 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
ES-401 3
Form ES-401-2 Rev. 11 (W E04) LOCA Outside Containment / 3 (W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals:
0 0
0 0
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 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 X
2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR:
41.7 / 43.5 / 45.12) 4.6 82 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 X
Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR:
43.5 / 45.13)
AA2.12 Maximum allowable channel disagreement 3.1*
83 000036 (APE 36; BW/A08) 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 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 X
2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 /
43.5 / 45.3) 4.0 84 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 X
Ability to determine and interpret the following as they apply to the (High Containment Radiation)
(CFR: 43.5 / 45.13)
EA2.2 Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
3.3 85 (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
ES-401 5
Form ES-401-2 Rev. 11 (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:
0 0
0 0
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 X
Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; 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)
A2.01 Problems with RCP seals, especially rates of seal leak-off 3.9 86 004 (SF1; SF2 CVCS) Chemical and Volume Control 005 (SF4P RHR) Residual Heat Removal X
Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, 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)
A2.04 RHR valve malfunction 2.9 87 006 (SF2; SF3 ECCS) Emergency Core Cooling X 2.1.32 Ability to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 /
45.12) 4.0 88 007 (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 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam X
Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on 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 Flow paths of steam during a LOCA 3.2 89 059 (SF4S MFW) Main Feedwater X 2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1
/ 45.13) 4.5 90 061 (SF4S AFW)
Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution 063 (SF6 ED DC) DC Electrical Distribution
ES-401 7
Form ES-401-2 Rev. 11 064 (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:
0 0
0 0
0 0
0 3
0 0
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 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication X
Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those 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)
A2.04 Misaligned rod 3.9 91 015 (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 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 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) 4.7 92 068 (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 Replaced KA w/KA 056 X 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) 4.7 92 086 Fire Protection X
Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection 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)
A2.01 Manual shutdown of the FPS 3.1 93 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals:
0 0
0 0
0 0
0 2
0 0
1 Group Point Total:
3
ES-401 9
Form ES-401-2 Rev. 11
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev. 11 Facility: Diablo Canyon Date of Exam: January 25, 2019 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.14 Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.
3.1 94 2.1.32 Ability to explain and apply system limits and precautions.
4.0 95 Subtotal
- 2. Equipment Control 2.2.6 Knowledge of the process for making changes to procedures.
3.6 96 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
4.2 97 Subtotal
- 3. Radiation Control 2.3.6 Ability to approve release permits.
3.8 98 Subtotal
- 4. Emergency Procedures/Plan Replaced KA 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.
4.3 99 2.4.50 2.4.45 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
Ability to prioritize and interpret the significance of each annunciator or alarm.
4.0 4.3 100 Subtotal Tier 3 Point Total 10 7
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev. 11 Tier /
Group Randomly Selected K/A Reason for Rejection SRO -
T1/G1 E12 G2.4.44 Not a testable tie between PARs and Uncontrolled Depressurization of all Steam Generators Randomly selected G2.4.6 (IR 4.7) as replacement.
SRO T3G4 G2.4.50 Not an SRO KA. Randomly selected 2.4.45 (IR 4.3)
SRO T2/G2 079 G2.2.40 Not a testable SRO system. Randomly 056 selected as replacement system.
Kept G2.2.40
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Diablo Canyon Date of Examination:
01/14/2019 Examination Level: RO SRO Operating Test Number:
L171 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations (NRCL171-A1)
M, R Determine Quadrant Power Tilt Ratio 2.1.7 Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.
(4.4)
(Bank: LJC-078)
Conduct of Operations (NRCL171-A2)
M, R Determine Spent Fuel Pool Heat Load/Removal Parameters 2.1.42 Knowledge of new and spent fuel movement procedures.
(2.5)
(From NRCL141-A2)
Equipment Control (NRCL171-A3)
M, R Prepare Outage Safety Checklist 2.2.37 Ability to determine operability and/or availability of safety related equipment.
(3.6)
(Bank: LJAEC-01R)
Radiation Control (NRCL171-A4)
M, R Determine Entry Conditions for Radiation Area Clearance 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. (3.5)
(modified from CP-2016-RA4)
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 ( 1)
(P)revious 2 exams ( 1, randomly selected)
Rev 0
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Diablo Canyon Date of Examination:
01/14/2019 Examination Level: RO SRO Operating Test Number:
L171 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations (NRCL171-A5)
D, R Review Operator Logs 2.1.3 Knowledge of shift or short-term relief turnover practices (3.9)
(From NRCL081LJA_SROA2)
Conduct of Operations (NRCL171-A6)
M, R Review Determination of Spent Fuel Pool Heat Load/Removal Parameters 2.1.42 Knowledge of new and spent fuel movement procedures.
(3.4)
(From NRCL141-A6)
Equipment Control (NRCL171-A7)
M, R Verify Outage Safety Checklist 2.2.37 Ability to determine operability and/or availability of safety related equipment.
(4.6)
(Bank: LJAEC-01S)
Radiation Control (NRCL171-A8)
M, R Determine Entry Conditions for Radiation Area Clearance and Associated Risk Assessment 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. (3.6) 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.
(3.8)
(modified from CP-2016-SA4)
Emergency Plan (NRCL171-A9)
D, R Perform an Emergency Classification 2.4.41 Knowledge of the emergency action level thresholds and classifications.
(4.6)
(Bank: LJE-004)
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 ( 1)
(P)revious 2 exams ( 1, randomly selected)
Rev 0A
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Diablo Canyon Date of Examination:
01/14/2019 Exam Level: RO SRO-I SRO-U Operating Test Number:
L171 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.A2.03) Verify Misaligned Rod Is Not Stuck (Bank LJC-066)
D,S 1
A,D,EN,L,S 2
- c. (S3) (010.A2.03) Depressurize RCS To Minimize Break Flow And Refill PZR (Modified LJC-033)
A,L,M,S 3
A,D,E,L,S 4P
- e. (S5) (026.A4.01) Align RHR to Containment Spray for Iodine Scrubbing A,E,L,N,S 5
- f. (S6) (055.EA1.07) Backfeed the Unit During a Loss of All AC Power L,N,S 6
- g. (S7) (015.A2.02) Remove Power Range Channel N42 from Service (LJC-051)
D,S 7
- h. (S8) (041.A4.08) Initiate RCS Cooldown to Cold Shutdown A,L,N,S 4S In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
D,E,R 1
- j. (P2) (055.EA1.04) Shed Non-Essential DC Loads (LJP-099)
D,E,L 6
- k. (P3) (067.AA1.08) Manually Operate the Cardox System - Normal Path (LJP-138)
D 8
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 /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 2 9/ 8/ 4 1/ 1/ 1 1/ 1/ 1 (control room system) 1/ 1/ 1 2/ 2/ 1 3/ 3/ 2 (randomly selected) 1/ 1/ 1 Rev 0
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Diablo Canyon Date of Examination:
01/14/2019 Exam Level: RO SRO-I SRO-U Operating Test Number:
L171 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.A2.03) Verify Misaligned Rod Is Not Stuck (Bank LJC-066)
D,S 1
A,D,EN,L,S 2
- c. (S3) (010.A2.03) Depressurize RCS To Minimize Break Flow And Refill PZR (Modified LJC-033)
A,L,M,S 3
A,D,E,L,S 4P
- e. (S5) (026.A4.01) Align RHR to Containment Spray for Iodine Scrubbing A,E,L,N,S 5
- f.
- g. (S7) (015.A2.02) Remove Power Range Channel N42 from Service (LJC-051)
D,S 7
- h. (S8) (041.A4.08) Initiate RCS Cooldown to Cold Shutdown A,L,N,S 4S In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
D,E,R 1
- j. (P2) (055.EA1.04) Shed Non-Essential DC Loads (LJP-099)
D,E,L 6
- k. (P3) (067.AA1.08) Manually Operate the Cardox System - Normal Path (LJP-138)
D 8
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 /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 8/ 4 1/ 1/ 1 1/ 1/ 1 (control room system) 1/ 1/ 1 2/ 2/ 1 3/ 3/ 2 (randomly selected) 1/ 1/ 1 Rev 0
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Diablo Canyon Date of Examination:
01/14/2019 Exam Level: RO SRO-I SRO-U Operating Test Number:
L171 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.
A,D,EN,L,S 2
- c.
- d.
- e. (S5) (026.A4.01) Align RHR to Containment Spray for Iodine Scrubbing A,E,L,N,S 5
- f.
- g.
- h. (S8) (041.A4.08) Initiate RCS Cooldown to Cold Shutdown A,L,N,S 4S In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
D,E,R 1
- j.
- k. (P3) (067.AA1.08) Manually Operate the Cardox System - Normal Path (LJP-138)
D 8
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 /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 8/ 4 1/ 1/ 1 1/ 1/ 1 (control room system) 1/ 1/ 1 2/ 2/ 1 3/ 3/ 2 (randomly selected) 1/ 1/ 1 Rev 0
ES-301 Transient and Event Checklist Form ES-301-5 Group x (In, In, Rn, Rn)
Facility: Diablo Canyon Date of Exam: Jan 14, 2019 Operating Test Number: L171 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios Scenario 1 Scenario 2 Scenario 3 Scenario 4 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 1
1 0
NOR 1
1 1
1 1
I/C 2,3,4 4
2,3,6,7 2,3,4,5 4,5 2,3,4,7 1,2,3 1,3 2,3,5,6 1,3,4,5,8 1,4,5,8 1,3,4,8,9 4
4 2
MAJ 5
5 5
6,8 6,8 6,8 4
4 4
6,7 6,7 6,7 2
2 1
TS 2,3,4 3,4 1,2 2,3 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
ES-301 Transient and Event Checklist Form ES-301-5 Group 1 (U1, R1, R2)
Facility: Diablo Canyon Date of Exam: Jan 14, 2019 Operating Test Number: L171 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios Scenario 1 (Day2) Scenario 2 (Day3) Scenario 3(Spare) Scenario 4 (Day1) 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-U1 RX 1
1 1
1 0
NOR 0*
1 1
1 I/C 2,3,4 3,4,5,6,9 8 4
4 2
MAJ 5
7,8 3
2 2
1 TS 2,3,4 2,3 5
0 2
2 RO1 SRO-I SRO-U RX 0*
1 1
0 NOR 1
1 1
1 1
I/C 2,3,6,7 4,5,9 7
4 4
2 MAJ 5
7,8 3
2 2
1 TS 0
0 2
2 RO2 SRO-I SRO-U RX 1
1 1
1 0
NOR 1
1 1
1 1
I/C 4
3,4,9,10 5 4
4 2
MAJ 5
7,8 3
2 2
1 TS 0
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 1
ES-301 Transient and Event Checklist Form ES-301-5 Group 2 (I1, I2, R3)
Facility: Diablo Canyon Date of Exam: Jan 14, 2019 Operating Test Number: L171 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios Scenario 1 (Day2) Scenario 2 (Day3) Scenario 3(Spare) Scenario 4 (Day1) 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-I1 SRO-U RX 1
1 1
1 0
NOR 0*
1 1
1 I/C 4
2,3,4,5 3,4,5,6,9 10 4
4 2
MAJ 5
6,8 7,8 5
2 2
1 TS 3,4 2,3 4
0 2
2 RO SRO-I2 SRO-U RX 1
1 1
1 0
NOR 1
1 2
1 1
1 I/C 2,3,4 2,3,4,7 4,5,9 10 4
4 2
MAJ 5
6,8 7,8 5
2 2
1 TS 2,3,4 3
0 2
2 RO3 SRO-I SRO-U RX 0*
1 1
0 NOR 1
1 2
1 1
1 I/C 2,3,6,7 4,5 3,4,9,10 10 4
4 2
MAJ 5
6,8 7,8 5
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 1
ES-301 Transient and Event Checklist Form ES-301-5 Group 3 (I3, I4, R5)
Facility: Diablo Canyon Date of Exam: Jan 14, 2019 Operating Test Number: L171 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios Scenario 1 (Day2) Scenario 2 (Day3) Scenario 3(Spare) Scenario 4 (Day1) 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-I3 SRO-U RX 1
1 1
1 0
NOR 0*
1 1
1 I/C 4
2,3,4,5 3,4,5,6,9 10 4
4 2
MAJ 5
6,8 7,8 5
2 2
1 TS 3,4 2,3 4
0 2
2 RO SRO-I4 SRO-U RX 1
1 1
1 0
NOR 1
1 2
1 1
1 I/C 2,3,4 2,3,4,7 4,5,9 10 4
4 2
MAJ 5
6,8 7,8 5
2 2
1 TS 2,3,4 3
0 2
2 RO5 SRO-I SRO-U RX 0*
1 1
0 NOR 1
1 2
1 1
1 I/C 2,3,6,7 4,5 3,4,9,10 10 4
4 2
MAJ 5
6,8 7,8 5
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 1
ES-301 Transient and Event Checklist Form ES-301-5 Group 4 (Surr, R4, R6)
Facility: Diablo Canyon Date of Exam: Jan 14, 2019 Operating Test Number: L171 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios Scenario 1 (Day2) Scenario 2 (Day3) Scenario 3(Spare) Scenario 4 (Day1) 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 4
4 2
MAJ 2
2 1
TS 0
2 2
RO4 SRO-I SRO-U RX 0*
1 1
0 NOR 1
1 1
1 1
I/C 2,3,6,7 4,5,9 7
4 4
2 MAJ 5
7,8 3
2 2
1 TS 0
0 2
2 RO6 SRO-I SRO-U RX 1
1 1
1 0
NOR 1
1 1
1 1
I/C 4
3,4,9,10 5 4
4 2
MAJ 5
7,8 3
2 2
1 TS 0
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 1
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-01 r1.docx Page 1 of 3 Rev 1 Facility:
Diablo Canyon (PWR)
Scenario No:
1 Op-Test No:
L171 NRC Examiners:
Operators:
Initial Conditions:
10E-8 with AFW in service, MOL, 1390 ppm boron Turnover:
At OP L-2, step 6.1.23, ready to raise power to 2%
Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 N/A R (ATC, SRO)
Raise reactor power from 10E-8 to 2% (OP L-2, sec 6.1) 2 NH04GSS_RV8855CTVLEAK.25 TS, C (BOP, SRO)
Accum Pressure Low (PK02-05; T.S. 3.5.1.B) 3 XMT_AFW14_3 ramp =120 TS, C (BOP, SRO)
AFW Pp 1-2 High Stator Temp (PK09-17, T.S. 3.7.5.B) 4 XMT_CVC4_3 27 ramp=60 TS, I (ATC, SRO)
Charging flow control transmitter FT-128 fails high (AP-5; TS 3.3.4.A) 5 RC08RC_8948CTVLEAK 0.5 delay=2 RC08RC_8818CTVLEAK 0.1 delay=2 RC09RC_8742BTVLEAK 0.1 delay=2 MAL_RHR2B 0.2 delay=5 M (ALL)
Seismically induced intersystem LOCA 6
MAL_VEN_POV2_TRICON_HALT C (BOP)
Aux Building Ventilation Failure (POV Halt) 7 RLY_PPL9_2 CLOSED RLY_PPL11_2 CLOSED RLY_PPL27_2 CLOSED C (BOP)
SSPS Relays K605, K613, K614 (various Phase A, Train A valves and Control Room Vent to Mode 4).
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-01 r1.docx Page 2 of 3 Rev 1 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) (ECA-1.2, E-1, FR-P.1) 3
- 6. EOP contingencies requiring substantive actions (0-2) (ECA-1.2, FR-P.1) 2
- 7. Critical tasks (> 2)(See description below) 2 Critical Task Justification Reference (S1CT-1) Manually isolate LOCA outside containment by closing 8809B such that a positive RCS pressure trend* is established prior to transition out of EOP ECA E-1.2, LOCA Outside Containment.
- RCS pressure greater than 10 psig about that at time of isolation and rising continuously.
The intersystem LOCA results in a failed fission product, permitting RCS fluid into the Aux Building through an unintended path. Failure to isolate the leak allows a containment barrier to remain in a degraded state when it is within the capability of the crew to correct the problem and restore the barrier.
- WCAP-17711-NP, CT-32
- WOG Background HECA12BG_R3 (S1CT-2) 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.
L171 NRC ES-D-1-01 r1.docx Page 3 of 3 Rev 1 SCENARIO
SUMMARY
- NRC #1
- 1.
Control rods are used to raise power from 10E-8 to 2% per OP L-2, Secondary Plant Startup, step 6.1.23. ATC operator complies with 1 step pull and wait procedural requirement while monitoring relevant controls and diverse indicators. Shift Foreman provides reactivity oversight.
- 2.
Accum 1-3 Relief Valve 8855C begins leaking by bringing in PK02-05, ACCUM PRESSURE HI-LO for Accumulator 1-3 pressure below 595.5 psig and enters T.S. 3.5.1.B, Accumulators, for Accumulator less than minimum required Tech Spec limit of 579 psig. The valve reseats at approximately 569 psig and the Accumulator pressure is restored per OP B-3B:I, Accumulators - Fill and Pressurize.
- 3.
Annunciator PK09-17, MOTOR DRIVEN AUX FW PP comes into alarm due to rising stator temperature on AFW Pump 1-2. Reactor operators note corresponding rise in motor amps to above the 75 amp maximum listed on VB3 lamacoid. When dispatched, field Nuclear Operator reports motor is extremely hot. Shift Foreman directs shutdown of AFW Pump 1-2 and enters T.S. 3.7.5.B Auxiliary Feedwater (AFW) System, for one AFW train inoperable and may elect to start the Turbine Drive AFW pump if desired, however, S/G levels are adequate and it is not required.
- 4.
FT128 (charging flow) fails high, causing actual charging flow to lower. The crew responds per OP AP5, Malfunction of Eagle 21 Protection or Control Channel. FCV128 and HC459D are taken to manual, and charging flow is monitored using alternate indications (RCP seals, Pzr level, VCT level, etc) for the remainder of the scenario. TS 3.3.4.A, Remote Shutdown Systems, is implemented.
- 5.
An earthquake occurs, causing an intersystem loca into the RHR system. RCS pressure degrades quickly and the Shift Foreman directs a reactor trip and safety injection.
- 6.
The crew enters EOP E-0, Reactor Trip or Safety Injection and performs their immediate actions. Several Containment Isolation valves fail to actuate due to SSPS relay failures and ventilation systems for the Control Room and Aux Building experience alignment issues. Manual alignments are addressed as part of Appendix E.
- 7.
E-0 Diagnostic steps direct the crew to EOP ECA-1.2, Intersystem LOCA to perform the critical task of closing 8809B (S1CT-1) Manually isolate LOCA outside containment. ***
- 8.
With RCS pressure recovering, the crew transitions to EOP E-1, Loss of Reactor or Secondary Coolant.
Low initial power combined with cold injection water from the RWST result in a magenta path on Containment Integrity, and the crew transitions to EOP FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, where they perform the critical task to terminate safety injection (S1CT-2)
Terminate SI prior to rupture of PRT.***
The scenario may be terminated once Critical Task S1CT-2, SI Termination, is complete.
- Denotes Critical Task
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-02 r1.docx Page 1 of 4 Rev 1 Facility:
Diablo Canyon (PWR)
Scenario No:
2 Op-Test No:
L171 NRC Examiners:
Operators:
Initial Conditions:
75% Power with D/G 1-2 OOS; MOL, 1018 ppm boron Turnover:
At 75% power for SCCW HX Clearance Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 N/A N (ATC, BOP)
Place 120 gpm letdown in service (OP B-1A:XII, Sec 6.3, B-1A:V, Sec 6.2).
2 insert VLV_RCP5_2 0.55 delIA VLV_RCP5_2 2 cd=V2_254S_1 C (BOP, SRO) 8153A (RCP 1-1 standpipe fill) fails mid-position (AR PK05-01).
3 XMT_RMS20_3 1E+06 TS, I (BOP, SRO)
S/G Blowdown RM-19 fails high. (ECG 39.5.A)(AR PK11-17).
4 MAL_RCS4C 50 TS, C (ALL) 50 gpm SGTL; plant shutdown required (OP O-4, AP-3, AP-25, TS 3.4.13.B).
5 MAL_ROD3_B10 STATIONARY MAL_ROD3_F02 STATIONARY C (ATC, SRO)
Two dropped rods during ramp (OP AP-12C).
6 MAL_RCS4C 500 M (ALL)
SGTR (S/G 1-3) during isolation following reactor trip.
7 VLV_MSS8_1 1 delIA VLV_MSS8_1 2 cd=V3_184s_1 C (BOP)
FCV-42 (Main Steam Isolation for S/G 1-2) fails open.
8 MAL_MSS2B 4.68e+06 cd=H_C2_013CTL_11 ramp=30 M (ALL)
Steamline break inside containment on S/G 1-2 following E-3 cooldown of RCS for SGTR.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-02 r1.docx Page 2 of 4 Rev 1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4)
Actual Attributes
- 1. Total malfunctions (5-8) (Events 2,3,4,5,6,7,8) 7
- 2. Malfunctions after EOP entry (1-2) (Event 7) 1
- 3. Abnormal events (1-4) (Events 2,3,4,5) 4
- 4. Major transients (1-2) (Event 6,8) 2
- 5. EOPs entered/requiring substantive actions (1-2) (E-3, E-2) 2
- 6. EOP contingencies requiring substantive actions (0-2) 0
- 7. Critical tasks (> 2)(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-108 (MDAFW Level Control Valve)
LCV-115 (TDAFW Level Control Valve)
Isolate steamflow by closing FCV-38 (AFW S/G 1-3 supply to TD AFW Pp)
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:
MAGENTA PATH on RCS INTEGRITY is avoided and RCS subcooled margin still exists following the cooldown.
- For 40% steam dumps, maximum rate limit is 100 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 40% dumps are not available or if steam line isolation occurs, maximum rate cooldown requires 10% steam dumps on intact S/Gs to be atleast 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 (continued)
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-02 r1.docx Page 3 of 4 Rev 1 Critical Task Justification Reference (S2CT-3) Isolate the faulted S/G before severe challenge to Integrity Safety Function develops (magenta path on F-0.4 RCS Integrity) by completing the following:
Isolate steamflow by closing:
FCV-37 (AFW S/G 1-2 supply to TD AFW Pp)
Isolate feedwater by ensuring closed:
LCV-107 (MDAFW Level Control Valve)
LCV-111 (TDAFW Level Control Valve)
Events which lead 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
- WCAP-17711-NP, CT-12
- WOG Background HE2BG_R3 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.
L171 NRC ES-D-1-02 r1.docx Page 4 of 4 Rev 1 SCENARIO
SUMMARY
- NRC #2
- 1.
Crew follows guidance of OP B-1A:V, section 6.2 for starting a second charging pump and places 120 gpm letdown in service per OP B-1A:XII, section 6.3 for RCS cleanup.
- 2.
RCP 1-1 standpipe fill valve 8153A drifts to mid-position. The crew responds per AR PK05-01, RCP NO. 11 for RCP 1-1 Standpipe Level High. The valve is manually closed and the crew verifies there are no corroborating conditions to indicate a leak on either the number 2 or 3 RCP seal.
- 3.
The S/G Blowdown radiation monitor (RM-19) fails high, but all associated automatic valve actuations actions fail. The Shift Foreman enters AR PK11-17, SG BLOW DOWN HI RAD and directs the board operator to manually close the associated blowdown isolation valves and realign blowdown discharge to the Equipment Drain Receiver. The crew diagnoses that the high reading on the radiation monitor is an instrument failure based on comparisons with other monitors, the rate of failure, etc. Shift Foreman enters ECG 39.5.A, Steam Generator Blowdown Liquid Sample Monitor RE-19, for Steam Generator Blowdown Tank (RM-19) inoperable
- 4.
Steam Generator 1-3 develops a 50 gpm tube leak as indicated by rising counts on various radiation monitors. The crew enters OP AP-3, Steam Generator Tube Failure, to address the leak and OP O-4, Primary to Secondary Steam Generator Tube Leak Detection, which directs the crew to reduce power by 50% in the next hour and be in Mode 3 within two hours. 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 to reduce power, two control rods drop, requiring the crew to enter OP AP-12C, Dropped Control Rod. The Shift Foreman directs a reactor trip and the crew transitions to EOP E-0, REACTOR TRIP OR SAFETY INJECTION after completing their immediate actions.
- 6.
Following the trip, the tube leak on S/G 1-3 rapidly develops into a 500 gpm rupture. The crew identifies the failure based on notable changes in the associated radiation monitor readings, RCS pressure and level lowering, and S/G 1-3 level beginning to rise. The crew transitions to EOP E-3, Steam Generator Tube Rupture, where they address critical tasks (S2CT-1) Isolate the ruptured steam generator from the intact steam generators and (S2CT-2) Perform RCS cooldown at maximum rate to CETC target temperature.***
- 7.
When the crew begins the depressurization phase of EOP E-3, a steamline break occurs inside containment on S/G 1-2. The crew implements EOP E-3 fold out page item #5 for Secondary Integrity and transitions to EOP E-2, Faulted Steam Generator Isolation to address the casualty.
- 8.
The Main Steam Line Isolation Valve for S/G 1-2, FCV-42, fails open and must be closed. Appendix HH, Isolate Faulted Steam Generator is implemented to meet the critical task of isolating S/G 1-2 (S2CT-3)
Isolate the faulted S/G before severe challenge to Integrity Safety Function develops.***
The scenario may be terminated once Critical Task S2CT-3, Isolate the faulted S/G is complete, which corresponds with completion of EOP E-2, Appendix HH.
- Denotes Critical Task
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-03 r0.docx Page 1 of 4 Rev 0 Facility:
Diablo Canyon (PWR)
Scenario No:
3 Op-Test No:
L171 NRC Examiners:
Operators:
Initial Conditions:
100% Power with CSP Pp 1-1 & CFCU 1-1 OOS; MOL, 954 ppm boron Turnover:
At 100% power. CFCU 1-1 failure late last shift.
Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 XMT_RCS54_3 679 ramp=180 TS, I (ATC, SRO)
Loop 2 Tcold slow failure high, requiring manual control of rods. (AP-5, TS 3.3.1.E,X; 3.3.2.M).
2 MAL_CCW1B 0.1 ramp=1 PMP_VEN6_MTRF 6.32 delay=600 ramp=10 TS, C (BOP, SRO)
CFCU 1-2 leak (AP-11, Sec C, TS 3.6.6.C & D).
3 MAL_CVC1 0.12 delay=30 ramp=180 C
(ALL) 35 gpm letdown leak inside containment (AP-18).
4 MAL_SEI1 0.3 delay=0 ramp=12 MAL_RCS1B 100%_DBA cd='jmlsei1' M
(ALL)
Seismically induced 100% DBA LBLOCA.
5 PMP_ASW2_2 OVERLOAD_DEV_FAIL cd='jmlsei1' delay=12 PMP_CCW1_2 OVERLOAD_DEV_FAIL cd='jmlsei1' delay=30 PMP_ASW1_1 AS_IS delIA PMP_ASW2_1 2 delay=0 cd='v1
_242s_3 OR V1_116S_2' C
(BOP)
ASW Pp 1-2 and CCW 1-1 OC on Seismic. ASW Pp 1-1 fails to start, manual start available.
6 RLY_PPL75_2 CLOSED C
(BOP)
Fail SSPS Relay K645 (CSP2 failed start)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-03 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) 6
- 2. Malfunctions after EOP entry (1-2) (Events 5,6) 2
- 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-1, E-1.3) 2
- 6. EOP contingencies requiring substantive actions (0-2) 0
- 7. Critical tasks (> 2)(See description below) 3 Critical Task Justification Reference (S3CT-1) Manually start ASW Pump 1-1 by completion of EOP E-0, Appendix E, step
- 11.
ASW train is required to remove accident generated and core decay heat following a design basis LOCA. Without ASW, CFCUs cannot remove heat from the containment atmosphere.
Additionally, ASW serves as the ultimate heat sink during the recirculation mode of ECCS cooling.
Failure to start the minimum number of required ASW pumps places the plant in an unanalyzed condition.
- FSAR, Section 6.2. & 9.2.7.
- Westinghouse Owners Group WCAP-17711-NP
- Technical Specification Basis B.3.7.8 (S3CT-2) Manually align at least 1 train of Containment Spray (1 pump and associated valves) by completion of EOP E-0, Appendix E, step 11.
Failure to initiate the minimum required Containment Spray equipment as a means of pressure suppression represents a severe challenge to Containment Safety Function.
- EOP FR-Z.1 Background Document
- Westinghouse Owners Group WCAP-17711-NP (S3CT-3) Establish one train of cold leg recirculation before insufficient RWST level results in ECCS pump cavitation as indicated by rapid swings in pump amperage. Recirc alignment verified as:
Train A (RHR HX 1-1 in Service) 8700A CLOSED 8982A OPEN 8804A OPEN Flow Indicated on FI-970A/B OR Train B (RHR HX 1-2 in Service) 8700B CLOSED 8982B OPEN 8804B OPEN 8807A OR 8807B OPEN Flow Indicated on FI-971A/B Failing to perform RHR suction realignment can lead to inadequate RHR NPSH and degraded the emergency core cooling system performance.
- STA-061 (07938-3-21, W Letter PGE-99-546 (07711-1153)
- OP1.ID2, Time Critical Operator Actions Rev 10,
- 8.
- 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.
L171 NRC ES-D-1-03 r0.docx Page 3 of 4 Rev 0 SCENARIO
SUMMARY
- NRC #3
- 1.
Loop 2 Tcold slowly fails high causing rods to drive in unexpectedly and charging flow to increase momentarily (until the Process Control System kicks out the bad signal). After determining auto rod motion is unwarranted, board operator takes rods to manual. Shift Foreman enters OP AP-5, Malfunction of Eagle 21 Protection or Control Channel, and directs manual operation of failed automatic controls to return parameters to normal. Once rods are returned to their starting all rods out position and temperature returned to normal, rod control can be placed back in Auto. Shift Foreman enters Tech Specs 3.3.1.E,X, Reactor Trip System Instrumentation, and 3.3.2.M, ESFAS Instrumentation for the failed channel.
- 2.
CFCU 1-2 develops a CCW leak bringing in drain collection system alarms and causing the CCW Surge tank level to drop. The crew responds, per OP AP-11, section C (outleakage), and OVIDs (system prints) to stop and isolate the CFCU. Shift Foreman enters Tech Spec 3.6.6.C and 3.6.6.D Containment Cooling Systems.
- 3.
A 35 gpm letdown line leak inside containment, downstream RO 27/28/29 ramps in over 3 minutes.
Diagnostic brief by the crew identifies lowering letdown flow, VCT level, and rising charging and determines letdown line inside containment as likely leak source (pressurizer pressure stable, structure sumps rising, RM-12 in alarm). Crew enters OP AP-18, Letdown Line Failure to address the leak. Normal letdown is isolated and excess letdown is placed in service.
(Note: If leak is not identified as being on the letdown line during initial diagnosis, crew will enter OP AP-1, and be directed to OP AP-18).
- 4.
A large seismic event results in a 100% DBA LBLOCA. Reactor Trip and Safety Injection occur immediately and the crew enters E-0, Reactor Trip or Safety Injection. The crew performs their immediate actions and checks for actuation of emergency safeguards equipment, diagnosing conditions consistent with a large break LOCA (high containment pressure, loss of pressurizer pressure and level, loss of subcooling, high containment sump levels). The crew identifies RCP trip criteria have been met, with Shift Foreman concurrence, trip all four RCPs (TCOA).** Shift Foreman directs the BOP Operator to complete Appendix E, ESF AUTO ACTIONS, SECONDARY AND AUXILIARIES STATUS, and continues on in E-0.
- 5.
Shift Foreman directs the BOP to complete EOP E-0, Appendix E, ESF AUTO ACTIONS, SECONDARY AND AUXILIARIES STATUS, and continues on in E-0. Following the guidance provided in Appendix E, critical tasks (S3CT-1) Manually start ASW Pump 1-1*** and (S3CT-2) Manually align at least 1 train of Containment Spray*** are completed and status reported back to the Shift Foreman.
(continued)
- TCOA note: This DBA LBLOCA was evaluated against TCOA #8, and is similar to the TCOA bases event, so TCOA time limits will be applied to the scenario (operators have 10 min to align to cold leg recirculation, as timed from the RWST reaching 33% [alarm comes in] and finishing the alignment). Phase B, RCP Trip Criteria in this scenario was evaluated against TCOA #67 and determined to apply. Operators have 5 minutes to trip all four RCPs from the initial Phase B actuation signal.
L171 NRC ES-D-1-03 r0.docx Page 4 of 4 Rev 0 SCENARIO
SUMMARY
- NRC #3
- 6.
Diagnostic steps of EOP E-0 direct the crew to E-1, Loss of Reactor or Secondary Coolant. When RWST level will reach 33%, Both RHR pumps trip when RWST level will reach 33%, and the crew transitions immediately to E-1.3, Transfer to Cold Leg Recirculation, performing the TCOA** of Cold Leg Recirculation Alignment (S3CT-3) Transfer to Cold Leg Recirculation and establish at least 1 train of ECCS flow within 10 minutes of the RWST level reaching 33%). The Shift Foreman assigns one operator (usually the BOP) Appendix EE to perform RHR Hx alignment, while continuing with the remaining operator to complete the TCOA alignment steps of E-1.3. The loss of one train of ASW early in the scenario will limit the crew to a single RHR Hx, requiring the crew to clearly communicate and choreograph the final valve alignment process.
The scenario may be terminated once a single train of Cold Leg Recirculation Alignment is complete per EOP E-1.3, Transfer to Cold Leg Recirculation.
- TCOA note: This DBA LBLOCA was evaluated against TCOA #8, and is similar to the TCOA bases event, so TCOA time limits will be applied to the scenario (operators have 10 min to align to cold leg recirculation, as timed from the RWST reaching 33% [alarm comes in] and finishing the alignment). Phase B, RCP Trip Criteria in this scenario was evaluated against TCOA #67 and determined to apply. Operators have 5 minutes to trip all four RCPs from the initial Phase B actuation signal.
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-04 r1.docx Page 1 of 3 Rev 1 Facility:
Diablo Canyon (PWR)
Scenario No:
4 Op-Test No:
L171 NRC Examiners:
Operators:
Initial Conditions:
75% Power with SI Pump 1-2 OOS; MOL, 1018 ppm boron Turnover:
At 75% power due to grid instability Event No Malf No.
Event Type*
Event Description (See Summary for Narrative Detail) 1 N/A N (ATC, BOP)
Places Pressurizer Steam Space Purge in Service to VCT (OP B-9:I, Sect 6.2, OP A-4A:I, Sec 6.6).
2 XMT_RCS44_3 -16 TS ONLY (SRO)
RCS flow transmitter FT-414 Fails Low (PK04-06, AP-5)(TS 3.3.1.M).
3 EE18PANEL_1G_5TC52_10 BREAKER_OPEN PK0102_0169 FAIL_TO_TRUE TS, C (BOP, SRO)
ASW Pp 1-2 Room Exhaust Fan Trip/High Temp (PK01-02)(TS 3.7.8.A).
4 GHDP_BYP 1 PMP_CND10_MTRF 6 ramp=10 C (ALL)
Heater 2 Drn Pump Trip - No Auto Ramp (AP-15, AP-25) 5 MAL_ROD6A MOTION_IN C (ATC, SRO)
Continuous rod insertion at completion of ramp (AP-12A) 6 MAL_RCP1A 15 delay=0 ramp=120 C (ATC, SRO) #
RCP 1-1 #1 Seal Leak (AP-28).
- Not counted as a verifiable action for board operators due to Event 6 (ATWS) 7 MAL_PPL5A 3, MAL_PPL5B 3 V5_245S_1 0, V5_239S_1 0 M (ALL)
ATWS with rod control malfunction.
8 MAL_RCP4A 1 cd='H_V2_268G_1' MAL_RCS3A 0.75 cd='H_V2_268 G_1' delay=30 M (ALL)
RCP 1-1 Shaft Seizure and LOCA once reactor trip breakers are opened.
9 RLY_PPL15_2, RLY_PPL7_2 CLOSED MAL_EPS4D_2 DIFFERENTIAL C (ALL)
SSPS relay K604 & K608 actuation failures and 4kV bus G differential - SI flowpath and ECCS pumps impacted.
10 MAL_PPL1A FAILURE_TO_INIT VLV_CVC5_1 1 C (BOP)
Containment Isolation Flowpath fails to isolate
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D (rev 11)
Scenario Outline Form ES-D-1 L171 NRC ES-D-1-04 r1.docx Page 2 of 3 Rev 1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) (from form ES301-4)
Actual Attributes
- 1. Total malfunctions (5-8) (Events 3,4,5,7,8,9,10) 7
- 2. Malfunctions after EOP entry (1-2) (Events 9,10) 2
- 3. Abnormal events (1-4) (Events 3,4,5) 3
- 4. Major transients (1-2) (Event 7,8) 2
- 5. EOPs entered/requiring substantive actions (1-2) (FR-S.1) 1
- 6. EOP contingencies requiring substantive actions (0-2) (FR-S.1) 1
- 7. Critical tasks (> 2)(See description below) 3 Critical Task Justification Reference (S4CT-1) Insert negative reactivity into the core following per EOP FRS.1 guidance so that power is reduced to less than 5% by the completion of step 19.
Failure to insert negative reactivity as procedurally directed constitutes a failure to provide appropriate reactivity control and represents an unnecessary and avoidable challenge to the criticality safety function.
- WCAP-17711-NP, CT-52
- FRS.1 Background Document, Rev. 3.
(S4CT-2) Manually align at least one train of SIS actuated safeguards by the completion of EOP E-0, Reactor Trip or Safety Injection, Appendix E, ESF Auto Actions, Secondary and Auxiliaries Status.
Component ID*
Position Intermediate Head Inj SIP 1-1 Running High Head Injection CCP 1-1 Running Charging Injection 8803A OPEN FSAR analysis predicates acceptable results on the assumption that, at the very least, one train of safeguards has actuated and is providing flow to the core. Failure to start and manually align the minimum required safeguards equipment results in the persistence of degraded emergency core cooling system capacity.
- WCAP-17711-NP, CT-2
- Backgrnd HE0BG_R3 (S4CT-3) Close containment isolation valves such that at least one valve is closed on each containment:
Close 8112 (RCP Seal Water Rtn) by the completion of EOP E-0, Reactor Trip or Safety Injection, Appendix E, ESF Auto Actions, Secondary and Auxiliaries Status.
Given the major event in this scenario (SBLOCA),
the open path between containment and the Auxiliary building represents the loss of a second fission product barrier. Closing the isolation valves is essential to safety since failure to do results in radiation release continuing beyond containment.
- Westinghouse Owners Group WCAP-17711-NP
- DCPP Emergency Action Level Technical Basis Manual 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.
L171 NRC ES-D-1-04 r1.docx Page 3 of 3 Rev 1 SCENARIO
SUMMARY
- NRC #4
- 1.
(Normal Evolution) Places Pressurizer Steam Space Purge in Service to VCT per OP B-9:I, Primary Sampling System - Make Available and Place in Service, Section 6.2 and OP A-4A:I, Pressurizer - Make Available, Sec 6.6).
- 2.
RCS flow transmitter FT-414 Fails Low bringing in PK04-06. PROTECT CHANNEL ACTIVATED, for low loop flow on RCS Loop 1. No automatic actions occur as the result of a single flow channel failure and the Shift Foreman enters OP AP-5, Malfunction of Eagle 21 Protection or Control Channel, to address Tech Spec 3.3.1.M, Reactor Trip System (RTS) Instrumentation).
- 3.
PK01-02, AUX SALT WTR PPS ROOM comes into alarm when the ASW Pp 1-2 room exhaust fan breaker trips open. The alarm reflashes shortly thereafter due to high pump room temperature. Field Operators report back that the breaker shows thermal damage and cannot be reclosed. The Shift Foreman directs swapping to the other ASW train per OP E-5:IV, Auxiliary Saltwater System-Swapping Pumps or HXs During Single CCW HX Operation and enters Tech Spec 3.7.8.A, Auxiliary Saltwater (ASW) System for one ASW train inoperable.
- 4.
Heater 2 Drn Pump Trips on overcurrent. The programmed ramp fails to initiate and crew must manually ramp to 770 MWe @ 40 mw/min per OP AP-15, Loss of Feedwater Flow, Section C: Heater 2 Drain Pump Trip. The crew will respond to the ramp, borate as needed, and stabilize the plant following the guidance of OP AP-25, Rapid Load Reduction or Shutdown.
- 5.
Rods continue to insert after ramp completes. Crew enters OP AP-12A, Continuous Withdrawal or Insertion of a Control Rod Bank and places rod control in manual. Tave is matched to Tref within 1.5 OF.
- 6.
RCP 1-1 seal leak ramps in over 2 minutes resulting in high seal return flow. Crew responds to AR PK05-01, RCP NO 11 for seal leakoff flow greater than 5.0 gpm and transitions to OP AP-28, Section B, RCP Number 1 Seal Failure. When seal leakoff and radial out bearing temperatures begin to rise, SFM directs Rx Trip and subsequent tripping of RCP 1-1 and closure of associated pressurizer spray valve once Rx Trip has been verified.
- 7.
The crew identifies the ATWS condition and the Shift Foreman enters EOP FRS.1, Response to Nuclear Power Generation / ATWS, (directly or from the step 1, response not obtained column of EOP E0, Reactor Trip or Safety Injection). Attempts to trip the reactor from the Control Room are unsuccessful and Autorod motion has failed. The crew performs the critical task of adding negative reactivity by manually driving rods (S4CT1) Insert negative reactivity into the core so that power is less than 5%.***
The crew continues working through FRS.1 until field operators are able to locally open the reactor trip breakers.
- 8.
RCP 1-1 fails due to a seized shaft which subsequently results in a 750 gpm SBLOCA. The reactor is verified subcritical and the crew transitions to EOP E0, Reactor Trip or Safety Injection. The Shift Foreman directs actuation of the Safety Injection signal. A bus differential occurs on vital 4kV Bus G when the Safety Injection system is actuated and three SSPS relays fail to actuate. The crew performs manual alignments and pump starts per EOP E-0, Appendix E, ESF AUTO ACTIONS, SECONDARY AND AUXILIARIES STATUS to satisfy critical tasks (S4CT-2) Manually align at least one train of SIS actuated safeguards by the completion of EOP E-0, Reactor Trip or Safety Injection, Appendix E*** and (S4CT-3)
Close 8112 (RCP Seal Water Rtn)*** before, transitioning to EOP E-1, Loss of Reactor or Secondary Coolant.
The scenario may be terminated once the transition has been made to EOP E-1, Loss of Reactor or Secondary Coolant.
- Denotes Critical Task