ML17355A644
| ML17355A644 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/15/2017 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | Wolf Creek |
| References | |
| 50-482/17-11 50-482/OL-17 | |
| Download: ML17355A644 (50) | |
Text
ES-401 RO PWR Examination Outline Form ES-401-2 Facility: Wolf Creek (RO Exam)
Date of Exam: November 2017 Tier Group RO K/A Category Points SRO-Only Points K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G*
Total A2 G*
Total
- 1.
Emergency &
Abnormal Plant Evolutions 1
3 3
3 N/A 3
3 N/A 3
18 6
2 2
1 1
2 2
1 9
4 Tier Totals 5
4 4
5 5
4 27 10
- 2.
Plant Systems 1
3 2
3 3
2 2
3 2
3 3
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
3 4
4 3
38 8
- 3. Generic Knowledge and Abilities Categories 1
3 2
2 3
2 4
3 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G*
Generic K/As
ES-401 RO Tier 1 / Group 1 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)
E/APE # / Name / Safety Function K
1 K
2 K
3 A
1 A
2 G*
K/A Topic(s)
IR 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 X
EK3.01: Knowledge of the reasons for the following as the apply to a reactor trip: Actions contained in EOP for reactor trip (CFR 41.5 /41.10 / 45.6 / 45.13) 4.0 11 000008 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 12 000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 X
EK3.13: Knowledge of the reasons for the following responses as the apply to the Large Break LOCA:
Hot-leg injection/recirculation (CFR 41.5 / 41.10 / 45.6 / 45.13) 3.8 13 000015/17 RCP Malfunctions / 4 X
AK1.02: Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow):
Consequences of an RCPS failure (CFR 41.8 / 41.10 / 45.3) 3.7 14 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 X
AK2.03: Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: Service water or closed cooling water pumps (CFR 41.7 / 45.7) 2.7 15 000026 Loss of Component Cooling Water / 8 X
2.1.27: Knowledge of system purpose and/or function.(CFR: 41.7) 3.9 16 000027 Pressurizer Pressure Control System Malfunction / 3 X
AK2.03: Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners (CFR 41.7 / 45.7) 2.6 17 000029 ATWS / 1 000038 Steam Gen. Tube Rupture / 3 X
EA2.01: Ability to determine or interpret the following as they apply to a SGTR: When to isolate one or more S/Gs (CFR 43.5 / 45.13) 4.1 18 000040 (BW/E05; CE/E05; W/E12)
Steam Line Rupture - Excessive Heat Transfer / 4 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) 4.0 19 000054 (CE/E06) Loss of Main Feedwater / 4 X
AK1.02: Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): Effects of feedwater introduction on dry S/G (CFR 41.8 / 41.10 / 45.3) 3.6 20 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 X
AK3.02: Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Actions contained in EOP for loss of offsite power (CFR 41.5,41.10 / 45.6 / 45.13) 4.4 21 000057 Loss of Vital AC Inst. Bus / 6 X
AA1.06: Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Manual control of components for which automatic control is lost (CFR 41.7 / 45.5 /
45.6) 3.5 22 000058 Loss of DC Power / 6 X
AA2.03: Ability to determine and interpret the following as they apply to the Loss of DC Power:
DC loads lost; impact on ability to operate and monitor plant systems(CFR: 43.5 / 45.13) 3.5 23
000062 Loss of Nuclear Svc Water / 4 X
AA1.05: Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): The CCWS surge tank, including level control and level
- alarms, and radiation alarm (CFR 41.7 / 45.5 / 45.6) 3.1 24 000065 Loss of Instrument Air / 8 X
AA2.06: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to trip reactor if instrument air pressure is de-creasing (CFR: 43.5 / 45.13) 3.6 25 W/E04 LOCA Outside Containment / 3 X
W/E04 EK2.2: Knowledge of the interrelations between the (LOCA Outside Containment) and the following: 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.(CFR: 41.7 / 45.7) 3.8 26 W/E11 Loss of Emergency Coolant Recirc. / 4 X
W/E11 EA1.3: Ability to operate and / or monitor the following as they apply to the (Loss of Emergency Coolant Recirculation): Desired operating results during abnormal and emergency situations. (CFR:
41.7 / 45.5 / 45.6) 3.7 27 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 X
AK1.02: Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: Over-excitation (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8) 3.3 28 K/A Category Totals:
3 3 3 3 3 3
Group Point Total:
18/
6
ES-401 RO Tier 1 / Group 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)
E/APE # / Name / Safety Function K
1 K
2 K
3 A
1 A
2 G*
K/A Topic(s)
IR 000001 Continuous Rod Withdrawal / 1 X
AA2.05: Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal: Uncontrolled rod withdrawal, from available indications (CFR:
43.5 / 45.13) 4.4 29 000003 Dropped Control Rod / 1 X
AA1.07: Ability to operate and / or monitor the following as they apply to the Dropped Control Rod: In-core and ex-core instrumentation (CFR 41.7 / 45.5 / 45.6) 3.8 30 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 X
AK3.02: Knowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions:
Relationships between PZR pressure increase and reactor makeup/letdown imbalance (CFR 41.5,41.10 / 45.6 / 45.13) 2.9 31 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 X
AK1.01: Knowledge of the operational implications of the following concepts as they apply to Loss of Intermediate Range Nuclear Instrumentation: Effects of voltage changes on performance CFR 41.8 / 41.10 /
45.3) 2.7 32 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 X
2.4.6: Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13) 3.7 33 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 X
AK1.02: Knowledge of the operational implications of the following concepts as they apply to Plant Fire on Site: Fire fighting (CFR 41.8 / 41.10 / 45.3) 3.1 35 000068 (BW/A06) Control Room Evac. / 8 X
AA2.09: Ability to determine and interpret the following as they apply to the Control Room Evacuation: Saturation margin. (CFR:
43.5 / 45.13) 4.1 34 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 X
W/E13 EA1.1: Ability to operate and / or monitor the following as they apply to the (Steam Generator Overpressure):
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.(CFR: 41.7 /
45.5 / 45.6) 3.1 36 W/E15 Containment Flooding / 5
W/E16 High Containment Radiation / 9 X
W/E16 EK2.2: Knowledge of the interrelations between the (High Containment Radiation) and the following:
Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.(CFR:
41.7 / 45.7) 2.6 37 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 Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:
2 1 1 2 2 1 Group Point Total:
9/4
ES-401 RO Tier 2 / Group 1 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)
System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G*
K/A Topic(s)
IR 003 Reactor Coolant Pump X
X K2.02: Knowledge of bus power supplies to the following: CCW pumps(CFR: 41.7)
A1.09: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: Seal flow and D/P (CFR: 41.5 / 45.5) 2.5 2.8 1
2 004 Chemical and Volume Control X
K5.09: Knowledge of the operational implications of the following concepts as they apply to the CVCS: Thermal shock:
high component stress due to rapid temperature change (CFR: 41.5/45.7) 3.7 3
005 Residual Heat Removal X
X K4.07: Knowledge of RHRS design feature(s) and/or interlock(s)which provide or the following: System protection logics, including high-pressure interlock, reset controls, and valve interlocks (CFR: 41.7)
A4.02: Ability to manually operate and/or monitor in the control room: Heat exchanger bypass flow control (CFR: 41.7 / 45.5 to 45.8) 3.2 3.4 4
5 006 Emergency Core Cooling X
2.1.20: Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 /
45.12) 4.6 6
007 Pressurizer Relief/Quench Tank X
A3.01: Ability to monitor automatic operation of the PRTS, including:
Components which discharge to the PRT (CFR: 41.7 / 45.5) 2.7 7
008 Component Cooling Water X
A2.03: Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High/low surge tank level (CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.3 8
010 Pressurizer Pressure Control X X K5.01: Knowledge of the operational implications of the following concepts as the apply to the PZR PCS:
Determination of condition of fluid in PZR, using steam tables (CFR: 41.5 /
45.7)
K6.03: Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PZR sprays and heaters (CFR: 41.7 / 45.7) 3.5 3.2 9
10 012 Reactor Protection X
K3.01: Knowledge of the effect that a loss or malfunction of the RPS will have on the following: CRDS (CFR:
41.7 / 45.6) 3.9 38
013 Engineered Safety Features Actuation X
X A1.06: Ability to predict and/or monitor changes in parameters (to Prevent exceeding design limits) associated with operating the ESFAS controls including:
RWST level(CFR: 41.5 / 45.5)
A3.01: Ability to monitor automatic operation of the ESFAS including: Input channels and logic (CFR: 41.7 / 45.5) 3.6 3.7 39 40 022 Containment Cooling X
K4.04: Knowledge of CCS design feature(s) and/or interlock(s)which provide for the following: Cooling of control rod drive motors (CFR: 41.7) 2.8 41 025 Ice Condenser 026 Containment Spray X
K1.01: Knowledge of the physical connections and/or cause effect relationships between the CSS and the following systems: ECCS (CFR: 41.2 to 41.9 / 45.7 to 45.8 4.2 42 039 Main and Reheat Steam X
A3.02: Ability to monitor automatic operation of the MRSS, including:
Isolation of the MRSS(CFR: 41.5 / 45.5) 3.1 43 059 Main Feedwater X
X K3.03: Knowledge of the effect that a loss or malfunction of the MFW will have on the following: S/Gs (CFR: 41.7 / 45.6)
A4.03: Ability to manually operate and monitor in the control room: Feedwater control during power increase and decrease (CFR: 41.7 / 45.5 to 45.8) 3.5 2.9 44 45 061 Auxiliary/Emergency Feedwater X
K6.01: Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Controllers and positioners(CFR: 41.7 / 45.7) 2.5 46 062 AC Electrical Distribution X
A4.04: Ability to manually operate and/or monitor in the control room: Local operation of breakers (CFR: 41.7 / 45.5 /
to 45.8) 2.6 47 063 DC Electrical Distribution X
K2.01: Knowledge of bus power supplies to the following: Major DC loads (CFR: 41.7) 2.9 48 064 Emergency Diesel Generator X
X K4.02: Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following: Trips for ED/G while operating (normal or emergency)(CFR: 41.7)
A1.03: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: Operating voltages, currents, and temperatures (CFR: 41.5 / 45.5) 3.9 3.2 49 50 073 Process Radiation Monitoring X
2.1.28: Knowledge of the purpose and function of major system components and controls. (CFR: 41.7) 4.1 51 076 Service Water X
K1.19: Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: SWS emergency heat loads (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.6 52 078 Instrument Air X
K3.01: Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Containment air system (CFR: 41.7 / 45.6) 3.1 53
103 Containment X
X K1.02: Knowledge of the physical connections and/or cause effect relationships between the containment system and the following systems:
Containment isolation/containment integrity. (CFR: 41.2 to 41.9 / 45.7 to 45.8)
A2.03: 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: Phase A and B isolation (CFR: 41.5 / 43.5 / 45.3 /
45.13) 3.1 3.5 54 55 K/A Category Point Totals:
3 2 3 3 2 2 3 2 3 3 2
Group Point Total:
28/5
ES-401 RO Tier 2 / Group 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)
System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G*
K/A Topic(s)
IR 001 Control Rod Drive X
K5.07: Knowledge of the following operational implications as they apply to the CRDS: Effects of an asymmetric rod configuration on power distribution (CFR:
41.5/45.7) 3.3 56 002 Reactor Coolant 011 Pressurizer Level Control X
K3.02: Knowledge of the effect that a loss or malfunction of the PZR LCS will have on the following: RCS (CFR: 41.7 / 45.6) 3.5 57 014 Rod Position Indication 015 Nuclear Instrumentation X
2.2.42: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) 3.9 58 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor X
K4.01: Knowledge of ITM system design feature(s) and/or interlock(s) which provide for the following: Input to subcooling monitors(CFR: 41.7) 3.4 59 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X
A1.02: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Radiation monitoring systems (CFR: 41.5 / 45.5) 2.8 60 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator X
A3.11: Ability to monitor automatic operation of the MT/G system, including:
Generator trip (CFR: 41/7 / 45.5) 2.6 61 055 Condenser Air Removal 056 Condensate x
K1.03: Knowledge of the physical connections and/or cause-effect relationships between the Condensate System and the following systems: MFW (CFR: 41.2 to 41.9 / 45.7 to 45.8) 2.6 65 068 Liquid Radwaste X
K6.10: Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste System : Radiation monitors (CFR: 41.7 / 45.7) 2.5 62
071 Waste Gas Disposal X
A4.14: Ability to manually operate and/or monitor in the control room: WDGS status alarms (CFR: 41.7 / 45.5 to 45.8) 2.8 63 072 Area Radiation Monitoring 075 Circulating Water X
A2.03: Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use Procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Safety features and relationship between condenser vacuum, turbine trip, and steam dump (CFR: 41.5 / 43.5 / 45.3 / 45.13) 2.5 64 079 Station Air 086 Fire Protection K/A Category Point Totals:
1 0 1 1 1 1 1 1 1 1 1 Group Point Total:
10/
3
ES-401 RO Tier 3 Form ES-401-3 Facility: Wolf Creek (RO)
Date of Exam: November 2017 Category K/A #
Topic RO SRO-Only IR IR
- 1.
Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements. (CFR:
41.10 / 45.13) 3.8 66 2.1.36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7) 3.0 67 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication. CFR: 41.7 /
43.5 / 45.4) 4.3 68 2.1.
Subtotal
- 2.
Equipment Control 2.2.13 Knowledge of tagging and clearance procedures.
(CFR: 41.10 / 45.13) 4.1 69 2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13) 3.6 70 2.2.
Subtotal
- 3.
Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10) 3.2 71 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR:
41.12 / 43.4 / 45.9) 2.9 72 2.3.
Subtotal
- 4.
Emergency Procedures /
Plan 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions. (CFR: 41.10 / 43.5 / 45.13) 3.7 73 2.4.32 Knowledge of operator response to loss of all annunciators. (CFR: 41.10 / 43.5 / 45.13) 3.6 74 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) 4.2 75 2.4.
Subtotal Tier 3 Point Total 10 7
ES-401 SRO PWR Examination Outline Form ES-401-2 Facility: Wolf Creek (SRO)
Date of Exam: November 2017 Tier Group RO K/A Category Points SRO-Only Points K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G*
Total A2 G*
Total
- 1.
Emergency &
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 0
2 1
3 Tier Totals 38 5
3 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 2
2 3
1 4
2 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G*
Generic K/As
ES-401 SRO Tier 1 / Group 1 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)
E/APE # / Name / Safety Function K
1 K
2 K
3 A
1 A
2 G*
K/A Topic(s)
IR 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 X
EA2.11: Ability to determine or interpret the following as they apply to a small break LOCA:
Containment temperature, pressure, and humidity (CFR 43.5 / 45.13) 4.1 76 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 X
AA2.01 : Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Whether charging line leak exists (CFR 43.5/ 45.13) 3.8 77 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 X
2.4.41: Knowledge of the emergency action level thresholds and classifications.(CFR: 41.10 / 43.5 /
45.11) 4.6 78 000038 Steam Gen. Tube Rupture / 3 X
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) 4.4 79 000040 (BW/E05; CE/E05; W/E12)
Steam Line Rupture - Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 X
EA2.02: Ability to determine or interpret the following as they apply to a Station Blackout: RCS core cooling through natural circulation cooling to S/G cooling.(CFR 43.5 / 45.13) 4.6 80 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4
BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 X
2.2.37: Ability to determine operability and/or availability of safety related equipment.(CFR: 41.7 /
43.5 / 45.12) 4.6 81 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals:
3 3
Group Point Total:
18/
6
ES-401 SRO Tier 1 / Group 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)
E/APE # / Name / Safety Function K
1 K
2 K
3 A
1 A
2 G*
K/A Topic(s)
IR 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 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 82 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 X
AA2.01: Ability to determine and interpret the following as they apply to the Loss of Containment Integrity: Loss of containment integrity (CFR: 43.5 / 45.13) 4.3 84 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 X
W/E01 EA2.1: Ability to determine and interpret the following as they apply to the Reactor Trip or Safety Injection Rediagnosis:
Facility conditions and selection of appropriate procedures during abnormal and emergency operations. (CFR: 43.5 / 45.13) 4.0 85 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 X
W/E15 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) 4.7 83 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 Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures
CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:
2 2 Group Point Total:
9/4
ES-401 SRO Tier 2 / Group 1 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)
System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G*
K/A Topic(s)
IR 003 Reactor Coolant Pump 004 Chemical and Volume Control X
2.1.23: Ability to perform specific system and integrated plant procedures during all modes of plant operation.(CFR:
41.10 / 43.5 / 45.2 / 45.6) 4.4 86 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection X
A2.01: Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty bistable operation (CFR: 41.5 / 43.5 / 45.3 / 45.5) 3.6 87 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 026 Containment Spray X
2.2.40: Ability to apply Technical Specifications for a system.(CFR: 41.10
/ 43.2 / 43.5 / 45.3) 4.7 88 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater X
A2.05: 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: Automatic control malfunction (CFR: 41.5 / 43.5 / 45.3 /
45.13) 3.4 89 062 AC Electrical Distribution X
A2.12: Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Restoration of power to a system with a fault on it (CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.6 90 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring
076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals:
3 2
Group Point Total:
28/
5
ES-401 SRO Tier 2 / Group 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)
System # / Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G*
K/A Topic(s)
IR 001 Control Rod Drive 002 Reactor Coolant X
2.4.18: Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13) 4.0 91 011 Pressurizer Level Control 014 Rod Position Indication X
A2.06: 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: Loss of LVDT (CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.0 92 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator X
A2.01: Ability to (a) predict the impacts of the following malfunctions or operations on the GS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or ruptured S/Gs (CFR: 41.5 / 43.5 / 45.3 /
45.5) 4.6 93 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection
K/A Category Point Totals:
2 1
Group Point Total:
10/
3
ES-401 SRO Tier 3 Form ES-401-3 Facility: Wolf Creek (SRO)
Date of Exam: November 2017 Category K/A #
Topic RO SRO-Only IR IR
- 1.
Conduct of Operations 2.1.35 Knowledge of the fuel-handling responsibilities of SROs.
(CFR: 41.10 / 43.7) 3.9 94 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 95 2.1.
Subtotal
- 2.
Equipment Control 2.2.11 Knowledge of the process for controlling temporary design changes. (CFR: 41.10 / 43.3 / 45.13) 3.3 96 2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc. (CFR: 41.10 / 43.5 /
45.13) 3.9 97 2.2.
Subtotal
- 3.
Radiation Control 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 98 2.3.
2.3.
Subtotal
- 4.
Emergency Procedures /
Plan 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (CFR: 41.10 / 43.5 / 45.13) 4.4 99 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11) 4.1 100 2.4.
Subtotal Tier 3 Point Total 10 7
ES-401 Record of Rejected K/As Form ES-401-4 Tier /
Group Randomly Selected K/A Reason for Rejection
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
WCNOC Date of Examination:
11/6/17 Examination Level: RO SRO Operating Test Number:
1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed A1 - PERFORM A MANUAL SHUTDOWN MARGIN CALCULATION Conduct of Operations K/A: 2.1.37 (4.3/4.6) - Knowledge of procedures, guidelines, or limitations associated with reactivity management.
R, M Perform a manual shutdown margin calculation using STS RE-004, SHUTDOWN MARGIN DETERMINATION, Attachment A, Shutdown Margin Calculation Short Form.
A2 - ESTIMATE TIME TO BOIL, TIME TO ONSET OF CORE UNCOVERING, AND TIME TO COMPLETE CORE UNCOVERING Conduct of Operations K/A: 2.1.25 (3.9/4.2) - Ability to interpret reference materials, such as graphs, curves, tables, etc.
R, D Estimate time to boil, time to onset of core uncovering, and time to complete core uncovering using Figures 5 and 6 per OFN EJ-015, LOSS OF RHR COOLING, Step 31.
A3 - DETERMINE ISOLATION POINTS FOR A MDAFW PUMP CLEARANCE ORDER Equipment Control K/A: 2.2.13 (4.1 / 4.3) - Knowledge of tagging and clearance procedures.
R, N Given a Clearance Order Worksheet, work order instructions, and drawings, determine isolation points to provide a safe work boundary for a MDAFW pump bearing oil change using AP 21E-001, CLEARANCE ORDERS.
A4 - DETERMINE DAC HOURS Radiation Control K/A: 2.3.12 (3.2/3.7) - 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.
R, D Calculate the dose an Operator would receive working a job with and without wearing a full-face respirator and determine which provides the Operator with less dose.
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).
ES-301 Administrative Topics Outline Form ES-301-1
- 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) (2)
(N)ew or (M)odified from bank ( 1) (2)
(P)revious 2 exams ( 1, randomly selected) (0)
Facility:
WCNOC Date of Examination:
11/6/17 Examination Level: RO SRO Operating Test Number:
1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed A5 - REVIEW A SHUTDOWN MARGIN CALCULATION Conduct of Operations K/A: 2.1.37 (4.3/4.6) - Knowledge of procedures, guidelines, or limitations associated with reactivity management.
R, M Review a completed manual shutdown margin calculation for approval, identify errors, and disapprove using STS RE-004, SHUTDOWN MARGIN DETERMINATION, Attachment A, Shutdown Margin Calculation Short Form.
A6 - DETERMINE LICENSED OPERATOR LICENSE STATUS Conduct of Operations K/A: 2.1.4 (3.3/3.8) - Knowledge of individual licensed Operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.
R, N Review watch standing history and Determine ROs 1 & 4 license status are inactive, ROs 2 &3 license status are active, and ROs 2 & 3 can be assigned to stand watch as the RO per AI 26C-002, CONDITIONS FOR MAINTAININGIN INDIVIDUAL LICENSES AT WOLF CREEK.
A7 - REVIEW A CLEARANCE ORDER Equipment Control K/A: 2.2.13 (4.1/4.3) - Knowledge of tagging and clearance procedures.
R, D Review a completed clearance order for approval, identify the critical errors, and disapprove the clearance order using AP 21E-001, CLEARANCE ORDERS.
A8 - REVIEW A LIQUID RELEASE PERMIT Radiation Control K/A: 2.3.6 (2.0/3.0) - Ability to approve release permits.
R, N Given a prepared liquid release permit, identify critical errors and determine if it is NOT ready to be authorized for release to the environment.
ES-301 Administrative Topics Outline Form ES-301-1 A9 - MAKE A PROTECTIVE ACTION RECOMMENDATION Emergency Plan K/A: 2.4.44 (2.4/4.4) - Knowledge of emergency plan protective action recommendations.
R, M Given a set of conditions, determine a GENERAL EMERGENCY exists per APF 06-002-01, EMERGENCY ACTION LEVELS and Make a Protective Action Recommendation using EPP 06-006, PROTECTIVE ACTION RECOMMENDATIONS, Attachment A, Protective Action Recommendation Chart.
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) (1)
(N)ew or (M)odified from bank ( 1) (4)
(P)revious 2 exams ( 1, randomly selected) (0)
Facility:
WCNOC Date of Examination:
11/6/2017 Exam Level: RO SRO-I SRO-U Operating Test Number:
1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U (Bold indicates alternate path)
System/JPM Title Type Code*
Safety Function S1 - Control Rod Drive System / PERFORM ACTIONS TO RETRIEVE A DROPPED CONTROL ROD K/A: APE 003 AA1.02 (3.6 / 3.4): Ability to operate and/or monitor the following as they apply to the Dropped Control Rod: Controls and components necessary to recover rod RO, SRO - I S, D 1 - Reactivity Control S2 - SI Termination / ESTABLISH LETDOWN DURING SI TERMINATION IAW EMG ES-03 (ESTABLISH EXCESS LETDOWN)
K/A: EA1.1 (4.0 / 3.9): Ability to operate and / or monitor the following as they apply to SI termination: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
RO, SRO - I, SRO - U S, N, A, EN, L 2 - Inventory Control S3 - Emergency Core Cooling System / OPERATE SI PUMP TO RAISE ACCUMULATOR LEVEL (RECOGNIZE SI PUMP GAS BINDING AND SECURE THE PUMP)
K/A: 006 A4.07 (4.4 / 4.4): Ability to manually operate and/or monitor in the control room: ECCS pumps and valves.
RO, SRO - I, SRO - U S, D, A, EN 3 - Pressure Control S4 - Waste Gas Disposal System / SET RM-11R SETPOINTS FOR A Radioactive Release K/A: 071 A4.25 (3.2 / 3.2): Ability to manually operate and/or monitor in the control room: Setting of process radiation monitor alarms, automatic functions, and adjustment of setpoints.
RO S, D 9 - Radioactivity Release
S5 - Reactor Coolant Pump System / START RCP D IAW EMG ES-11 K/A: E03 EA1.1 (4.0 / 4.0) Ability to operate and / or monitor the following as they apply to LOCA cooldown and depressurization:
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
RO, SRO - I S, D, L 4P - Heat Removal (Primary System)
S6 - Containment Spray System / MANUALLY ALIGN CONTAINMENT SPRAY COMPONENTS THAT FAILED TO ACTUATE FOLLOWING A LOSS OF COOLANT ACCIDENT.
K/A: 026 A4.01 (4.5 / 4.3) Ability to manually operate and/or monitor in the control room: CSS controls.
RO, SRO - I S, N, L, A, EN 5 - Containment Integrity S7 - Emergency Diesel Generators / TRANSFER NB02 FROM THE EDG TO THE NORMAL POWER SUPPLY (UNLOAD AND STOP THE EDG FOLLOWING LOSS OF ESW)
K/A: 064 A1.03 (3.2 / 3.3): Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the EDG system controls including: Operating voltages, currents, and temperatures.
RO, SRO - I S, N, A 6 - Electrical S8 - Reactor Protection System / BYPASS A FAILED PRNI CHANNEL K/A: 012 A4.03 (3.6 / 3.6): Ability to manually operate and/or monitor in the control room: Channel blocks and bypasses.
Instrumentation In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 - DC Electrical Distribution System / TRANSFER BUS NN02 POWER Supply TO NN12 FROM BYPASS CVT FOLLOWING A LOSS OF AND RESTORATION OF BUS NK02.
K/A: 058 AA1.03 (3.1 /3.3): Ability to operate and/or monitor the following as they apply to the Loss of DC Power: Vital and battery bus components.
RO, SRO - I, SRO - U D, L 6 - Electrical
P2 - Spent Fuel Pool Cooling System / MAKEUP TO THE SFP FROM ESW K/A: 033 A2.03 (3.1 / 3.5) Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal spent fuel pool water level or loss of water level.
RO, SRO - I, SRO - U N, R 8 - Plant Service Systems P3 - Chemical and Volume Control System / PERFORM LOCAL ACTIONS TO BORATE THE REACTOR COOLANT SYSTEM (USE THE ALTERNATE METHOD FOR LOCAL OPERATION)
K/A: APE 024 / AA1.04 (3.6 / 3.7): Ability to operate and / or monitor the following as they apply to emergency boration: Manual boration valve RO, SRO - I, SRO - U D, A, E, R 1 - Reactivity Control 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 (5) /4-6 (5) /2-3 (3)
(0/0/0) 9 (7) / 8 (6) / 4 (3) 1 (1) / 1 (1) / 1 (1) 1 (3) / 1 (3) / 1 (2) (control room system) 1 (4) / 1 (4) / 1 (2) 2 (4) / 2 (4) / 1 (2) 3 (0) / 3 (0) / 2 (0) (randomly selected) 1 (2) / 1 (2) / 1 (2)
(8/7/2)
ES-301 Transient and Event Checklist Form ES-301-5 Facility: Wolf Creek 1 Date of Exam: 11/6-10/2017 Operating Test No.: 2017301 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios 1
4 3
2 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
U1 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 5,7 3,4,5
,7,8 10 4
4 2 MAJ 6
6 2
2 2 1 TS 2,4 2,5 4
0 2 2 R1 RX 0
1 1 0 NOR 0
1 1 1 I/C 3,4, 5,7 4,5, 7
7 4
4 2 MAJ 6
6 2
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 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 Transient and Event Checklist Form ES-301-5 Facility: Wolf Creek 1 Date of Exam: 11/6-10/2017 Operating Test No.:
A P
P L
I C
A N
T E
V E
N T
T Y
P E
Scenarios 1
4 3
2 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
I1 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 5,7 4,5, 7
2,3,4 5,7 13 4
4 2 MAJ 6
6 6
3 2
2 1 TS 2,4 2,3 4
0 2 2 I2 RX 0
1 1 0 NOR 1
1 1
3 1
1 1 I/C 2,5 3,4,5
,7,8 2,3,4 5,7 12 4
4 2 MAJ 6
6 6
3 2
2 1 TS 2,5 2,3 4
0 2 2 R3 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 3,4, 5,7 3,8 2,4,5 9
4 4 2 MAJ 6
6 6
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 Transient and Event Checklist Form ES-301-5 Facility: Wolf Creek 1 Date of Exam: 11/6-10/2017 Operating Test No.:
A P
P L
I C
A N
T E
V E
N T
T Y
P E
Scenarios 1
4 3
2 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
I3 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 5,7 4,5, 7
2,3,4 5,7 13 4
4 2 MAJ 6
6 6
3 2
2 1 TS 2,4 2,3 4
0 2 2 I4 RX 0
1 1 0 NOR 1
1 1
3 1
1 1 I/C 2,5 3,4,5
,7,8 3,4,5
,7 11 4
4 2 MAJ 6
6 6
3 2
2 1 TS 2,5 2
0 2 2 R4 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 3,4, 5,7 3,8 2,4,5 9
4 4 2 MAJ 6
6 6
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 Transient and Event Checklist Form ES-301-5 Facility: Wolf Creek 1 Date of Exam: 11/6-10/2017 Operating Test No.:
A P
P L
I C
A N
T E
V E
N T
T Y
P E
Scenarios 1
4 3
2 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
I5 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 5,7 3,4,5
,7,8 2,4,5 13 4
4 2 MAJ 6
6 6
3 2
2 1 TS 2,4 2,5 4
0 2 2 R5 RX 0
1 1 0 NOR 1
1 1
3 1
1 1 I/C 2,5 3,8 3,4,5
,7 8
4 4 2 MAJ 6
6 6
3 2
2 1 TS 0
0 2 2 R6 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 3,4, 5,7 4,5, 7
3,4,5
,7 11 4
4 2 MAJ 6
6 6
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.
Facility: _Wolf Creek_________ Scenario No.: ____1________
Op-Test No.: _________
Examiners: ___________________________ Operators:
Initial Conditions: 95% Power, B CCP in service, 75 gpm letdown _____________________
Turnover: No equipment out of service. The crew has been directed to raise letdown flow to 120 gpm after taking the watch. Power level reduced in preparation for Heater Drain Pump A maintenance later in the shift. _____________________ _____________ _________________
Critical Tasks: CT-1 Isolate faulted S/G C before CTMT CSF Orange Path met on CTMT normal sump level >= 2003 11. CT-2 Isolate High Head ECCS flow through the BIT before overfill of the RCS results in a rupturing of the pressurizer relief tank (PRT) ____________________
Event No.
Malf.
No.
Event Type*
Event Description 1
Raise Letdown to 120 GPM 2
TS (SRO)
PZR Level Instrument BB LI-459A Fails Low 3
Steam Header Pressure Instrument PI-507 fails low 4
TS (SRO)
Spurious opening of S/G B ARV 5
RCP A Hi Vibration, requires RX trip 6
M (All)
Unisolable Loop C main feed line break in Containment 7
FWIS fails to actuate automatically (both trains)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes per Scenario (See Section D.5.d)
Actual Attributes
- 1.
Malfunctions after EOP entry (1-2) 2
- 2.
Abnormal events (2-4) 4
- 3.
Major transients (1-2) 1
- 4.
EOPs entered/requiring substantive actions (1-2) 2
- 5.
Entry into a contingency EOP with substantive actions (> 1 per scenario set) 0
- 6.
Preidentified critical tasks (> 2) 2
Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback CT 1: Isolate faulted S/G C before Containment CSF Orange Path met on CTMT normal sump level >= 2003 ft 11 in.
Failure to isolate feed flow into containment leads to an unnecessary and avoidable challenge to the containment integrity critical safety function as a result of flooding.
S/G pressure and level indications will make it possible to identify S/G C as the faulted S/G.
Lowering Level S/G C. Containment pressure and temp rise resulting in SIS. Containment radiation levels do not rise as expected if a LOCA was present.
FWIS will show as not actuated on the ESFAS Status Panel. Valve position indication on MCB and plant computer show C FRV and FWIV open.
With the failure of FWIS to actuate, the Applicant will manually close the following (6) valves to isolate C S/G.
AB HV-20 AE HV-41 AE FK-530 AL HV-11 AL HV-12 AB V-V087 Indication of FWIV closure on MCB (green lamp lit).
Pressure drop in intact S/Gs stop, and S/G levels stabilize.
CT 2: Isolate high head ECCS flow through the BIT before overfill of the RCS results in a rupturing of the pressurizer relief tank (PRT)
Continued maximum injection causes RCS to go solid and PORV to open, passing excess inventory through PORVs to the PRT.
Failure to terminate ECCS flow when it is possible to do so results in a rupture of the PRT, spread of radioactive coolant into Containment, and constitutes an avoidable degradation of a fission product barrier, as well as increased risk of stuck open PORV (SBLOCA).
RCS pressure and pressurizer level rise. PORVs open, flow indicated.
PRT level, pressure, and temperature rise.
When PRT ruptures at ~91 psig, PRT pressure drops and equalizes with Containment Pressure.
The Operator will isolate the BIT per EMG ES-03, Step 13.
Green lights LIT for the following valves:
- EM HIS-8803A
- EM HIS-8803B
- EM HIS-8801A
- EM HIS-8801B CCP To BIT Flow indicators drop to 0 GPM.
- EM FI-917A
- EM FI-917B Note: Causing an unnecessary plant trip or ESF actuation may constitute a Critical Task failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1 SCENARIO # 1 NARRATIVE Turnover: The Unit is operating at 95% power MOL. B CCP is in service. No equipment is out of service. The crew has been directed to raise letdown flow to 120 gpm after taking the watch.
Event 1: Raise Letdown to 120 GPM. The crew will raise letdown flow from 75 GPM to 120 GPM in accordance with SYS BG-201, SHIFTING CHARGING PUMPS, Step 6.6 per beginning of shift guidance.
Once the PZR Level Master Controller BB LK-459 is in auto and letdown flow has been raised to 120 GPM, and/or at the discretion of the Lead Examiner, the next event will start.
Event 2: PZR Level Instrument BB LI-459 Fails LOW. The upper selected controlling PZR Level channel will fail to 20%. The crew will select an alternate PZR level channel and enter OFN SB-008, INSTRUMENT MALFUNCTIONS, to perform ATTACHMENT J. The SRO will evaluate Tech Spec LCO 3.3.1 Reactor Trip System (RTS) Instrumentation (Table 3.3.1-1 Function 8.a) and determine Conditions A (Immediate) and M - (72 Hours) are applicable. When the crew has returned PZR Level control to Auto and evaluated Technical Specifications, the next event will start.
Event 3: Steam header pressure instrument PI-507 fails LOW. The crew takes manual control of the MFP Master Speed Control and then enters OFN SB-008, INSTRUMENT MALFUNCTIONS, and performs Attachment B. After the crew has placed the Master MFP Speed Controller in manual to restore S/G levels to program, the next event will start.
Event 4: S/G B ARV AB PV-2 opens spuriously. The crew will attempt to place the ARV in manual and close, but this attempt will NOT be successful. The crew will be required to dispatch an operator for local closure of S/G B ARV isolation valve AB-V040 to stop steam flow. The SRO should evaluate and determine Tech Spec LCO 3.7.4 Atmospheric Relief Valves (ARVs) Condition A (7 days) is applicable.
When the crew has isolated B ARV and determined applicable tech specs, the next event will start.
Event 5: RCP A Hi Vibration, Requires Rx Trip, MCB Annunciators 070B-RCP VIB SYS ALERT and 070A-RCP VIB DANGER illuminate. The crew will validate Immediate Trip Criteria per OFN BB-005, RCP MALFUNCTIONS conditions exist and manually trip the reactor and the A RCP.
Events 6 and 7: Unisolable Loop C Main Feed Line Break in Containment and Failure of FWIS to automatically actuate. When the crew has tripped the reactor, stopped RCP A and completed the immediate action steps of EMG E-0, REACTOR TRIP OR SAFETY INJECTION, an unisolable feedwater line break occurs on C S/G inside containment. Both trains of FWIS fail to actuate, so the crew is required to isolated Feed Water Isolation Valves (FWIV) and Feed Water Regulation valves (FRV).
The crew will re-perform the immediate actions of EMG E-0, and then perform Foldout Page Item 3 to isolate C S/G by closing the MSIV, FWIV AE HV-41, and stopping auxiliary feedwater and main feedwater flow to C S/G. The crew will transition to EMG E-2, FAULTED STEAM GENERATOR ISOLATION, and complete isolation of C S/G. CT 1: Isolate faulted S/G C before Containment CSF Orange Path met on CTMT normal sump level >= 2003 ft 11 in.
The crew then transitions to EMG ES-03, SI TERMINATION, and reduces ECCS flow by isolating flow through the BIT. CT 2: Isolate high head ECCS flow through the BIT before overfill of the RCS results in a rupturing of the pressurizer relief tank (PRT).
The scenario is complete when the crew has completed isolation of C S/G and isolated ECCS flow through the BIT, and/or at the discretion of the Lead Examiner.
Note: Causing an unnecessary plant trip or ESF actuation may constitute a Critical Task failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1
- 2017 ILO NRC Exam, Scenario 1
- Event 1 - No Keys - Raise Letdown to 120 GPM.
- Event 2 - [Key 2] PZR Level Instrument BB LI-459A Fails LOW IMF mBB22A f
- 20 r:30 k:2
- Event 4 - [Key 4] Spurious Opening of 'B' ARV IMF mAB07F f
- 100 r:30 k:4
- REMOTE Action to Locally Close B ARV isolation AB-V040
{KEY[7} IRF rAB02B f:0 r:300
- Event 5 - [Key 5]'A' RCP High Vibration IMF mBB31A f
- 6 r:30 k:5 IMF mBB32A f:22 r:30 k:5
- Event 6 - Unisolable Feedline Break in CTMT (both Auto and Manual initiation options available)
{bkPA00107.state=0} IMF mAE14C f:1e+07
{KEY[6]} IMF mAE14C f:1e+07
- Event 7 - FWIS fails to auto actuate on both trains IMF mSA21A f
- 1 IMF mSA21B f:1
- REMOTE Action to Locally Close AB V-087
{KEY[8]} IRF rAB04B f:0 r:30
Appendix D Scenario Outline Form ES-D-1 Facility: _Wolf Creek_________ Scenario No.: ____2________
Op-Test No.: ____1_____
Examiners: ___________________________ Operators:
Initial Conditions: 100% Power, EDG B and B ESW out of service. B CCP in service, 120 gpm letdown. A and B CCW Pump in service with Service Loop aligned to A train. _____________
Turnover: Crew will place C CCW Pump in service per SYS EG-120, COMPONENT COOLING WATER SYSTEM, Section 6.3, Shifting Train A CCW Pumps____________________________
Critical Tasks: CT-1, Establish AFW Flow during SBO before 3 S/G WR level < =12%. CT-2, Energize at least one AC Emergency Bus prior to performing depressurization and cooldown Step 32 of EMG C-0.__________________________________________________________________________
Event No.
Malf. No.
Event Type*
Event Description 1
Swap CCW Pumps 2
C (SRO/RO)
TS (SRO)
B CCP trips (TS) 3 I (SRO/ BOP)
TS (SRO)
C S/G Level Channel, AE LI-553 Fails HIGH (TS) 4 C (ALL)
Condensate Pump B Trip, Downpower to 90%
5 C (ALL)
MFP A Trip, Rx Trip Required 6
M (ALL)
LOOP and A EDG trips, (Loss of All AC) 7 C (SRO, BOP)
TDAFW pump fails to Auto Start 8
SWYD East Bus Restored, Energize NB01 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes per Scenario (See Section D.5.d)
Actual Attributes
- 1. Malfunctions after EOP entry (1-2) 2
- 2. Abnormal events (2-4) 4
- 3. Major transients (1-2) 1
- 4. EOPs entered/requiring substantive actions (1-2) 1
- 5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 1
- 6. Pre-identified critical tasks (> 2) 2
Appendix D Scenario Outline Form ES-D-1 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback CT-1 Establish at least 270,000 lbm/hr AFW flow rate to the S/Gs before 3 S/G wide range levels reach 12% [28%].
The failure of the TDAFW pump to start requires the operator to take manual action to ensure the necessary decay heat removal system is maintained for the required station blackout coping duration.
- Indication and/or annunciation of station blackout AND Indication and/or annunciation that insufficient AFW flow to S/Gs is present Manipulation of controls in the control room as required to establish the minimum required AFW flow rate to the S/Gs Indication that at least the minimum required AFW flow is being delivered to the S/Gs S/G Levels rising CT-2 Energize at least one AC Emergency bus prior to conducting the depressurization and cooldown Step 32 of EMG C-0.
Reenergizing an emergency bus allows transition to a recovery procedure for normal recovery.
Conducting the cooldown would cause a Safety Injection signal and complicate the recovery.
- Indication and/or annunciation of station blackout AND
- Indication that an Offsite Power source is available Manipulation of controls in the control room as required to establish power to one AC Emergency Bus.
Normal voltage indicated on the Emergency Bus Note: Causing an unnecessary plant trip or ESF actuation may constitute a Critical Task failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1 SCENARIO 2 NARRATIVE Turnover: 100% Power, EDG B and B ESW are out of service. B CCP is in service with 120 gpm letdown. Both trains of CCW are in service with the service loop aligned to A train. The crew will swap running CCW Pumps in A train per SYS EG-120, COMPONENT COOLING WATER SYSTEM, Section 6.3, Shifting Train A CCW Pumps after taking the watch.
Event 1: BOP Swaps CCW Loop per SYS EG-120, Section 6.3. A copy of SYS EG-120 will be presented to the BOP at shift turnover. The BOP will perform the selected procedure section to start C CCW Pump and secure A CCW Pump. Upon completion of step 6.3.11, Event 2 will start.
Event 2: B CCP Trips. MCB Annunciators 42A, CHG LINE FLOW HILO and 42E, CHARGING PMP TROUBLE will actuate. The crew is expected to respond by performing ALR 00-042E. There is a memory Action step for the RO to isolate letdown if NO Charging pump is running. The CRS will determine that LCO 3.5.2, ECCS-OPERATING, Condition A - 72 Hours and TRM 3.1.9, BORATION INJECTION SYSTEM - OPERATING, CONDITION A - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are applicable. Once charging and letdown are re-established and returned to automatic control, and the CRS has determined technical specifications, Event 3 will start.
Event 3: C S/G Level Channel, AE LI-553 Fails HIGH. MCB Annunciators 110B, SG C LEV DEV will actuate. The BOP has a memory action to take manual control of C S/G Feed Reg Valve (AE FK-530) and match steam and feed flows to restore S/G Level to program 50%. The crew will determine which instrument failed and remove it from control before restoring to automatic. The CRS will determine LCO 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Conditions A-Immediately, and E-72 hours are applicable per Table 3.3.1-1, Function 14. LCO 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION, CONDITIONS A-Immediately, D-72 hours and I-72 hours are applicable per Table 3.3.2-1, Functions 5.c and 6.d. Once the CRS has determined applicable technical specifications Event 4 will start.
Event 4: Condensate Pump B Trip, Downpower to 90%. MCB Annunciators 120B, MFP A SUCT PRESS LO, 123B, MFP B SUCT PRESS LO, and 126D, CONDS PUMP TRIP will actuate. ALR 00-126D, directs the crew to reduce Turbine load using OFN MA-038, RAPID PLANT SHUTDOWN to restore MFP suction pressure to >340 psig. The crew is expected to commence down power using reactivity numbers that were pre-briefed before the scenario. Once the reactor turbine load reduction has stopped at 90%, or at the direction of the lead examiner, Event 5 will start.
Event 5: Main Feed Pump A Trip, Rx Trip Required. MCB Annunciator 120A, MFP A TRIP will actuate and S/G levels will lower rapidly, causing S/G Level alarms to also actuate. Step 2 of ALR 00-120A directs tripping the reactor and going to EMG E-0, REACTOR TRIP OR SAFETY INJECTION if power level is >62%. Depending on crew response, they may manually trip the reactor or the reactor will automatically trip on S/G LoLo level. The next two events automatically start with the Reactor Trip.
Event 6: Loss of Offsite Power and A EDG Trips, (Loss of All AC). The crew will perform immediate actions of EMG E-0, where at Step 3 the RO will realize neither NB01 nor NB02 are able to be powered and the crew is expected to transition to EMG C-0, LOSS OF ALL AC POWER. The crew will perform actions of EMG C-0 until, with the concurrence of the lead examiner, power from the Lacygne/Waverly line will be made available to reenergize the switchyard East Bus and start Event 8.
Event 7: TDAFW pump fails to auto start. With the Loss of all AC and failure of the TDAFW Pump to auto start, the crew must manually start the TDAFW Pump to provide heat sink. EMG C-0, Step 5 provides direction to start TDAFW Pump.
CT-1 Establish at least 270,000 lbm/hr AFW flow rate to the S/Gs before 3 S/G wide range levels reach 12% [28%].
Appendix D Scenario Outline Form ES-D-1 Event 8, SWYD East Bus Restored, Energize NB01, The crew will restore power to bus NB01 using OFN NB-030, LOSS OF AC EMERGENCY BUS NB01 (NB02). Upon re-energization of bus NB01 and with concurrence of the lead examiner, the exam scenario is complete [Freeze Simulator].
CT-2 Energize at least one AC Emergency bus prior to conducting the depressurization and cooldown Step 32 of EMG C-0.
Note: Causing an unnecessary plant trip or ESF actuation may constitute a Critical Task failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1
- ILO 2017 Scenario #2
- Tag out 'B' EDG scn SimGroup\\TAGDGB
- Event 1, No Keys
- Event 3, [Key 3] S/G 'C' level channel AE LT-553 fails high IMF mAE15C4 f
- 101 r:10 k:3
- Event 4, [Key 4] Condensate Pump 'B' trips ICM bkrDPAD01B t
- 1 d:0 k:4
- Event 5, [Key 5] 'A' MFP Trips IMF mFC01A f
- 1 k:5
- Event 6, Loss of Offsite power on Rx trip, 'A' EDG Trips
{jpplp4} IMF mSY01 i:-1 f:-1 d:10 IMF mNE04A i:-1 f:-1
- Event 7, TDAFWP fails to auto start IMF mAL01 i
- -1 f:-1
- [Key 6] Restore Waverly/LaCygne Line, Swyd energizes
{Key[6]} DMF mSY01 IMF mSY03E i:-1 f:-1 k:6 IRF rSY05 f:1 k:6 IRF rSY06 f:1 k:6
- [Key 7] Isolate RCP seals
- LOCALLY CLOSE SEAL WTR INJ FILTER INLET ISO'S BGV101/105 IRF rBG09A f
- 0 d:30 r:30 k:7 IRF rBG09B f:0 d:60 r:30 k:7
- LOCALLY CLOSE EG HV61 ICM movEGHV0061 t
- 2 d:0 k:7
Appendix D Scenario Outline Form ES-D-1 Facility: _Wolf Creek_________ Scenario No.: ____3________
Op-Test No.: ___1__ __
Examiners: ___________________________ Operators:
Initial Conditions: 100% Power, No equipment out of service, No maintenance in Progress.______
Turnover:__ Continue 100% Power operations.____________________________ ____________
Critical Tasks: CT Manually Trip The Turbine CT Isolate an intact S/G from the ruptured S/G.
CT Depressurize RCS to meet SI termination criteria prior to overfilling ruptured S/G.
Event No.
Malf. No.
Event Type*
Event Description 1
BG PK-131, LTDN HX OUTLET PRESS CTRL fails closed in Auto 2
PORV 455A fails Open (TS) 3 I (SRO / BOP)
D S/G Press Channel PI-545 fails HIGH (TS) 4 C (ALL)
Heater Drain Pump trip (Downpower) 5 M (ALL)
SGTR on A S/G, A MSIV fails Open 6
Turbine will not Auto Trip, Manual available 7
B ESW pump fails to Auto Start (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes per Scenario (See Section D.5.d)
Actual Attributes
- 1. Malfunctions after EOP entry (1-2) 2
- 2. Abnormal events (2-4) 4
- 3. Major transients (1-2) 1
- 4. EOPs entered/requiring substantive actions (1-2) 2
- 5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 1
- 6. Pre-identified critical tasks (> 2) 3
Appendix D Scenario Outline Form ES-D-1 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback CT-1 Manually trip the main turbine before a severe orange-path challenge develops to either the subcriticality or the integrity Critical Safety Function or before transition to EMG E-2, whichever happens first.
Failure to stop an uncontrolled cooldown constitutes a potential challenge to both Integrity and Subcriticality Critical Safety Functions, resulting in a reduction of margin to safety.
Turbine Stop valves do not close on SI signal RCS Temperature drops due to excessive steaming after reactor trip.
Manipulation of controls as required to trip the turbine.
Indication of Turbine Stop Valves closed.
RCS uncontrolled cooldown stopped.
CT-2, Isolate an intact S/G from the ruptured S/G prior to exiting EMG E-3, or attempting to perform a controlled RCS cooldown.
Failure to decouple the ruptured S/G from the RCS by failing to close any non-ruptured S/G MSIV extends the loss of RCS inventory and raises the amount of contamination that bypasses containment into the secondary system.
Radiation monitor alarms.
All MSIVs indicate open on MCB.
Ruptured S/G pressure dropping while attempting to use a non-ruptured S/G for cooldown.
Manipulation of controls to close at least 1 MSIV:
B: AB HIS-17 C: AB HIS-20 D: AB HIS-11 B/C/D:
AB HIS-79/80 B, C, and/or D MSIVs close providing at least one intact S/G that is thermo-dynamically isolated from the ruptured S/G.
CT-3, Commence controlled RCS depressurization to allow for SI termination prior to overfilling the ruptured S/G (90%
WR).
Depressurizing the RCS to equalize with Ruptured S/G pressure prior to overfilling the ruptured S/G minimizes radioactive release to the environment from the ruptured S/G, minimizes stress to the Main Steam Lines, and allows for a subcooled recovery vice a potential saturated recovery.
S/G Level rising in an uncontrolled manner with feed flow isolated.
Radiation monitor alarms Manipulation of controls as required to depressurize the RCS.
RCS Pressure reducing in a controlled
- manner, subcooling maintained, leak rate to ruptured S/G
- drops, PZR Level >6%,
Ruptured S/G Level <93% NR.
Note: Causing an unnecessary plant trip or ESF actuation may constitute a Critical Task failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1 Scenario 3 Narrative Turnover, 100% Power, MOL, Boron Concentration 976 ppm, No equipment out of service, No maintenance in Progress.
Event 1, BG PK-131 fails closed in Auto. MCB Annunciators 039C, LP LTDN RELIEF TEMP HI and 039D, LTDN HX DISCH PRESS HI will both actuate. Crew may perform either ALR procedure as the procedural guidance is the same. The RO will take manual control of BG PK-131 and adjust controller output to establish and maintain a letdown pressure of 350 psig. Upon completion of the ALR, and with the concurrence of the lead examiner, Event 2 will start.
Event 2, PORV 455A Fails Open, MCB Annunciator 035B, PORV OPEN will actuate. Step 1 of ALR 00-035B is a memory action step for the Operator to close the affected PZR PORV if PZR Pressure is <2315 psig and then close the associated block valve when the PORV fails to indicate fully shut. The CRS will determine that Technical Specification 3.4.11, Condition B is applicable with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> completion time to close and remove power to the associated block valve. Upon completion of the ALR, and after the CRS has determined technical specification, and with the concurrence of the Lead examiner, Event 3 will start.
Event 3, D S/G Pressure Channel AB PI-545 fails HIGH, MCB Annunciator 111C SG D FLOW MISMATCH will actuate. The crew is expected to respond by performing OFN SB-008, INSTRUMENT MALFUNCTIONS, ATTACHMENT C. The BOP has a memory action to take manual control of D S/G Feed Reg Valve (AE FK-540) and match steam and feed flows to restore S/G Level to program 50%.
The crew will determine which instrument failed (AE PI-545) and remove it from control before returning AD FK-540 Feed Reg Valve to automatic. The CRS will determine LCO 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION, CONDITIONS A-Immediately, and D-72 hours are applicable per Table 3.3.2-1, Functions 1.e and 4.e. Once the SRO has determined applicable technical specifications and with the concurrence of the lead examiner, Event 4 will start.
Event 4, A Heater Drain Pump Trip, (Downpower) MCB Annunciators 103E, HEATER DRN TK DUMP will actuate 60-70 seconds after the pump trips. The crew may notice a change in Main Generator load trending down, or a NPIS alarm prior to Annunciator 103E alarming. Expect the crew to perform ALR 103E and/or reference OFN AF-025, UNIT LIMITATIONS. OFN AF-025 directs a Maximum unit load of 95% and performance of SYS AF-121 to address continued operations with one Heater Drain Pump Out Of Service. The crew may choose to rapidly reduce power using OFN MA-038, RAPID PLANT SHUTDOWN or conduct a reactivity brief and lower power at a slower rate to 95% using GEN 00-004 hard card. Once the reactor turbine load reduction has stopped at 95%, or at the direction of the lead examiner, Event 5 will start.
Event 5, SGTR on A S/G, with A MSIV failed open. The crew will diagnose a SGTR with RCS pressure and PZR Level dropping and MCB Annunciator 061A, PROCESS RAD HIHI alarm with corresponding RED radiation levels on RM-11 Panel for GE RE-92, CONDENSER AIR REMOVAL DISCHARGE Process Radiation Monitor. The crew will Trip the reactor and actuate SI and perform EMG E-0, REACTOR TRIP OR SAFETY INJECTION until directed to transition to EMG E-3, STEAM GENERATOR TUBE RUPTURE at step 17. Prior to exiting EMG E-0, The BOP Operator will have to address a failure of the turbine to trip (Event 6) and perform foldout page direction to isolate A S/G while the RO addresses failure of B ESW Pump to auto start. Events 5-7 all occur simultaneously. While performing EMG E-3, the Crew will identify the inability to close A S/G MSIV and may transition to EMG C-31, SGTR WITH LOSS OF REACTOR COOLANT - SUBCOOLED RECOVERY DESIRED. Once the crew starts depressurizing the RCS and /or with the direction of the lead evaluator the scenario will be complete.
CT-2: Isolate an intact S/G from the ruptured S/G prior to exiting EMG E-3, or attempting to perform a controlled RCS cooldown.
CT-3: Commence controlled RCS depressurization to allow for SI termination prior to overfilling the ruptured S/G (90% WR).
Appendix D Scenario Outline Form ES-D-1 Event 6, Turbine will not trip in Auto. The BOP will note the turbine stop valves are not closed when looking at the Ovation Graphic Screen. The turbine should have tripped on Safety Injection. The RNO action is to depress BOTH turbine trip buttons which successfully trips the turbine.
CT-1: Manually trip the main turbine before a severe orange-path challenge develops to either the subcriticality or the integrity CSF or before transition to EMG E-2, whichever happens first.
Event 7, B ESW Pump fails to Auto Start, The B EDG will be running unloaded without cooling water.
One of the Operators may notice the pump failed to start and will manually start the pump per AP15C-003, PROCEDURE USERs GUIDE FOR ABNORMAL PLANT CONDITIONS Manual Backup. The RO is also directed to check this pump running during performance of EMG E-0, ATTACHMENT F. Per OFN EF-033, LOSS OF ESW, the diesel generator may run for up to 30 minutes unloaded before over temperature condition develop. NOT a critical task since both NB busses are powered from off-site power, but failure to take appropriate action within the required timelines would eliminate an emergency source of power and unnecessarily complicate the scenario and reduce the safety margin.
Note: Causing an unnecessary plant trip or ESF actuation may constitute a Critical Task failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1
- Event 1 - [Key 1] BG PK-131 fails closed IMF mBG17A f
- 0 r:5 k:1
- Event 2 - [Key 2]PORV 455A fails Open, Manual Available
{Key[2]}IMF mSA27BB07
{Key[2]}set cnS829A_NO.input=1
{Key[2]}set cnK811A_NO.input=1 IOR P21053A f:1 k:2
{Key[11]} IRF rBB31A f:1
- Event 3 - [Key 3]SG Press channel AB PT-545 fails HIGH IMF mAB01D2 f
- 1313 k:3
- Event 4 - [Key 4]Heater Drain Pump 'A' trips ICM bkrDPAF01A t
- 1 d:0 k:4
- Event 5 - [Key 5]SGTR on 'A' SG IMF mBB02A f
- 350 r:30 k:5
- Event 6 - Turbine fails to Auto trip IMF mAC02B i
- -1 f:-1
- Event 7 - ESW 'B' does not Auto Start IMF mEF05B i
- -1 f:-1
- Remote Local Action [Key 6]Locally close AB V-062 IRF rAB03A f
- 0 r:30 k:6
- Remote Local Action [Key 7] Place 'A' EDG in Standby
{Key[7]}scn SimGroup\\EDGA_STBY
- Remote Local Action [Key 8] Place 'B' EDG in Standby
{Key[8]}scn SimGroup\\EDGB_STBY
- Remote Local Action [Key 9] Close NG01AFF4 and NG02AAF4 (Bat Pumps)
IRF rBG40A f:1 k:9 IRF rBG40B f:1 k:9
- Remote Local Action [Key 10] Close NG04CPF2 (BG HV-8104)
IRF rBG41 f:1 k:10
Appendix D Scenario Outline Form ES-D-1 Facility: _Wolf Creek_________ Scenario No.: ___4________ Op-Test No.: ___1______
Examiners: ___________________________ Operators:
Initial Conditions: 2% power. A MFP in service. Power ascension on hold due to emergent work on A CCP. A CCP tagged out. LCO. 3.5.2 Condition A entered. Estimate return in 30 minutes. Both CCW trains are in service with CCW Pumps A and B running, with Service Loop aligned to B train.
Turnover: Maintain current power level while the crew briefs for the upcoming MODE Change. Swap Operating Spent Fuel Pool (SFP) Cooling Pumps to support Mechanical Maintenance on B SFP Cooling Pump.________________________________________________________________
Critical Tasks: CT-1, Manually Actuate Safety Injection. CT-2, Manually Start CCW Pump D.
CT-3, Isolate LOCA outside Containment by closing EJ HIS-8701A.
Event No.
Malf.
No.
Event Type*
Event Description 1
N (SRO/BOP)
Swap Running SFP pumps per SYS EC-120 2
TS (SRO)
Dual Data Failure on one Shutdown Bank Rod (TS) 3 C (SRO/BOP)
A Main Feed Reg Bypass valve fails open 4
C (SRO/RO)
Letdown HX BG TCV-130 controller fails closed in Auto 5
I (SRO/ RO)
TS (SRO)
PZR Press instr BB PI-456 fails high, PORV Opens (TS) 6 M (ALL)
SSE Earthquake, RCS leak to RHR system outside CTMT 7
C (SRO/RO)
SI fails to Auto actuate 8
C (SRO/BOP)
CCW Pump B trips on SI, CCW Pump D fails to Auto start (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes per Scenario (See Section D.5.d)
Actual Attributes
- 1. Malfunctions after EOP entry (1-2) 2
- 2. Abnormal events (2-4) 4
- 3. Major transients (1-2) 1
- 4. EOPs entered/requiring substantive actions (1-2) 1
- 5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 1
- 6. Pre-identified critical tasks (> 2) 3
Appendix D Scenario Outline Form ES-D-1 Critical Task Safety Significance Cueing Measurable Performance Indicators Performance Feedback CT-1 Manually actuate at least one train of SI before any of the following:
Transition to any EMG E-1 series, EMG E-2 series, or EMG E-3 series procedure or transition to any FRG.
Completion of Step 16 of EMG ES-02.
Failure to manually actuate SI under the postulated conditions constitutes misoperation or incorrect crew performance in which the crew does not prevent degraded emergency core cooling system (ECCS)capacity.
PRZR pressure less than SI actuation setpoint.
PRZR water level less than the foldout page criterion for SI actuation.
No indication or annunciation that SI is actuated Manipulation of Either:
SB HS-27 or SB HS-28 to manually actuate at least one train of SI.
Indication and/or annunciation that at least one train of SI is actuated:
Annunciator 030A or 031A, Red SI Actuate Light on Panel SB 069 CT-2 Manually start D CCW pump prior to transition out of EMG E-0, or completion of Attachment F, whichever occurs later.
Failure to maintain CCW flow to ECCS components results in a reduction of safety margin. Red Train ECCS equipment is degraded due to LOCA outside Containment through Loop Hot Leg to RHR Pump A Suction Isolation Valve, requiring Yellow Train CCW for Safe Shutdown.
Indication and/or annunciation that SI is actuated AND Indication and/or annunciation that less than the minimum number of CCW pumps required to provide adequate component cooling for the operating safeguards train(s) are running Manipulation of EG HIS-24 to start D CCW Pump.
CCW low pressure condition clear; indication of pressure CCW low flow condition clear; indication of flow; Red running light on EG HIS-24.
CT-3 Isolate the LOCA outside containment before transition out of EMG C-12.
Failure to isolate a LOCA outside containment degrades containment integrity beyond the level of degradation irreparably introduced by the postulated conditions.
Indication and/or annunciation that SI is actuated and is required AND Indication and/or annunciation of abnormally high radiation in the auxiliary building Manipulation of EJ HIS-8701A to isolate the LOCA Indication of rising RCS pressure Note: Causing an unnecessary plant trip or ESF actuation may constitute a Critical Task failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1 SCENARIO NARRATIVE Initial Conditions: 2% power. A MFP in service. Power ascension on hold due to emergent work on A CCP. A CCP tagged out. LCO. 3.5.2 Condition A entered. Estimate return in 30 minutes. Maintain current power level while the crew briefs for the upcoming MODE Change.
Event 1: Swap Running SFP pumps from B to A per SYS EC-120. Crew turnover includes direction to swap operating Spent Fuel Pool (SFP) Cooling Pumps to support mechanical maintenance on B SFP Cooling Pump. A copy of SYS EC-120, FUEL POOL COOLING AND CLEANUP SYSTEM STARTUP will be provided with instructions to perform section 6.1, starting with step 6.1.3. Maintenance has NOT just been performed on the A SFP Cooling Pump and SFP Cleanup System is NOT in service.
Event 2: Dual Data Failure on one Shutdown Bank Rod. Numerous alarms associated with rod control.
080A, RPI URG ALARM is the higher tier alarm and should be entered by the SRO. DRPI indication will show General Warning (GW) and Rod Bottom (RB) lights for shutdown rod N11. There are no verifiable actions for the board operators. The SRO will review and enter T.S. 3.1.4, Condition B, T.S. 3.1.7, Condition A, and review T.R. 3.1.7, which is N/A for this Mode.
Event 3: A MFRV Bypass valve fails open. Level in A S/G rises. Annunciator 108B, SG A LEV DEV will alarm if level gets to 55%. SRO should perform ALR 108B. There is no OFN guidance for this failure.
At this low power and low flows this will be a slow moving event.
Event 4: Letdown HX BG TCV-130 controller fails closed in Auto. Temperature will rise on BG TI-130 on MCB RL002. Annunciator 039B, LTDN HX TEMP HI, will alarm. 039A, LTDN HX TEMP HI DIVERT, may also alarm later. THE SRO should enter one of the ALRs. Direction form the ALRS is to place BG TCV-130 in manual and control temperature between 110 and 120°F.
Event 5: PZR Press instrument BB PI-456 fails high, PORV Opens. This is the lower controlling channel and will only affect PORV BB HIS-456A. Block valves will close at 2185 psig but will reopen when pressure rises above 2192 psig. Annunciator 039B, PORV OPEN will alarm. ALR has memory action to close PORV however SRO may enter OFN SB-008, INSTRUMENT MALFUNCTIONS, directly if the RO recognizes the channel failure. The RO should select P455/P458 on BB PS-455F to select out the failed channel. SRO will determine LCO 3.3.1, Conditions A, E, and M are applicable (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action) and LCO 3.3.2, Conditions A, D (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action) and L (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action to verify P-11 interlock is in correct state).
Event 6: SSE Earthquake, RCS leak to RHR system outside CTMT. The LOCA is delayed 10 seconds after the earthquake. The crew should observe PZR pressure and level lowering quickly and the SRO should order a reactor trip and safety injection. The crew should progress through EMG C-12 and close Loop Hot Leg to RHR Pump A Suction Isolation Valve EJ-HV-8701A to isolate the LOCA. When RCS pressure is noted to be rising, the scenario can be terminated at the Lead Examiners direction.
CT-3: Isolate the LOCA outside containment before transition out of EMG C-12 Event 7: SI fails to Auto actuate. Manual is available. Crew will manually actuate SI, either with pre-emptive action or as a manual backup to failed auto actuation during performance of immediate actions.
CT-1: Manually actuate at least one train of SI before any of the following:
Transition to any EMG E-1, EMG E-2, or EMG E-3 series procedure or transition to any FRG.
Completion of Step 16 of EMG ES-02.
Event 8: CCW Pump B trips on SI, CCW Pump D does not auto start. The BOP may identify and correct the deficiency following performance of immediate actions, or the RO will have procedure guidance to verify correct CCW System alignment when performing EMG E-0, ATTACHMENT F.
CT-2: Manually start at least one CCW pump in the train with required ECCS equipment prior to transition out of EMG E-0, or completion of Attachment F, whichever occurs later.
Note: Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-1021.
Appendix D Scenario Outline Form ES-D-1
- 2017 ILO Scenario #4
- Initial Conditions - Position EJ HV-8701A, Ack alarm when done
+1 ICM movEJHV8701A t:4 f:0.2 d:0 r:0
+03 ICR movEJHV8701A t:1 d:0
+4 ICR movEJHV8701A t:0 d:0
+2 IMF ANN-B049 f:1
- Event 1 - No Keys
- Event 2 - [Key 2] DRPI Failure Shutdown Rod N11 IMF mSF10N11 f
- 2 k:2
- Event 3 - [Key 3] 'A' MFRV Bypass fails open IMF mAE20A f
- 99.1 r:10 k:3
- Event 4 - [KEY 4] BG TCV-130 fails closed IMF mBG18A f
- 0 r:10 k:4
- Event 5 - [Key 5] BB PI-456 fails high IMF mBB21B f
- 2508 r:10 k:5
- Event 6 - [Key 6] SSE Earthquake, RCS Leak to RHR and pipe rupture IMF mSG01 f
- 60 k:6 ICM vlBBPV8702A t:1 f:0.02 d:10 r:0 k:6 IMF mEJ04A f:3500 d:10 k:6
- Local Remote Action - [Key 7] Restore Power to Loop Isolation Valves
{Key[7]}DCM movEJHV8701A ICR movBBPV8702A t:1 d:0 k:7 ICR movBBPV8702B t:1 d:3 k:7 ICR movEJHV8701A t:1 d:30 k:7 ICR movEJHV8701B t:1 d:33 k:7 IOR P17091A f:1 k:7
- Event 7 - automatic SI fails - No Key IMF mSA14B f
- 1 IMF mSA14A f:1
{jpplsi}ICM bkrDPEG01B t:1 d:0 IMF mEG14D f:-1