1CAN101904, AN1-2020-04 Draft Outlines
| ML20233A923 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/05/2020 |
| From: | Martin R Entergy Operations |
| To: | Kelly Clayton Operations Branch IV |
| References | |
| 1CAN101904 | |
| Download: ML20233A923 (47) | |
Text
Entergy Operations, Inc.
untelg)1 Russellville,AR 72802 Te1479 8586124 D. Gregory Kilpatrick Training Manager Arkansas Nuclear One October 30, 2019 1CAN1O19O4 Mr. Kelly Clayton Chief Examiner U. S. Nuclear Regulatory Commission 1600 E. Lamar Blvd Arlington, TX 76011-4511
SUBJECT:
RO and SRO (Ui) Exam Outlines Submittal Arkansas Nuclear One Unit 1 Docket No. 50-313 Renewed Facility Operating License No. DPR-51
Dear Mr. Clayton:
Enclosed are the operating tests outlines. Additionally, enclosed are the written examinations outlines in their current state after being supplied by the NRC. The operating tests outlines were developed under NUREG 1021 Revision 11, attached are the necessary forms and supporting information as required by ES-201 -1.
We request that these documents be withheld from public disclosure in accordance with 1 OCFR2.790.
Should you require further information regarding the content of this submittal, please contact Randal Martin at (479) 858-6844.
Sincerely, DGK/rkm Randal Martin Enclosures Facility Reviewer cc w/o enclosures:
Mr. Greg Werner, Chief Operations Branch Division of Reactor Safety U. S. Nuclear Regulatory Commission Region IV 1600 E. Lamar Blvd Arlington, TX 76011-4511 CMS File Licensing, ANO-DCC
ES-401, Page 40 of 52 ES-401 PWR Examination Outline Form ES-401-2 Facility: ANO U1 Date of Exam: 4/8/2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total 1.
Emergency and Abnormal Plant Evolutions 1
3 3
3 3
3 3
18 6
2 2
1 1
1 2
2 9
4 Tier Totals 5
4 4
4 5
5 27 10 2.
Plant Systems 1
3 2
3 3
2 2
3 2
3 2
3 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
3 4
38 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
3 2
2 3
Note:
- 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only 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, Page 41 of 52 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 R
EK2.2 Knowledge of the interrelations between the Vital System Status Verification and the following:
facilities heat removal system, including primary coolant, emergency coolant, decay heat removal systems, and relations between the proper operations of these systems to operation of the facility 4.2 1
Bank 415 000008 (APE 8) Pressurizer Vapor Space Accident / 3 R
AK2.02 Knowledge of the interrelations between the PZR Vapor space accident and the following:
Sensors and detectors 2.7*
2 Bank 1029 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 R
EK3.04 Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: Placing containment fan cooler damper in accident position 4.0 3
TBD 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 R
AK1.03 Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow): The basis for operating at a reduced power level when one RCP is out of service 3.0 4
TBD 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 R
R AK1.02 Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Relationship of charging flow to pressure differential between charging and RCS AK1.01 Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Consequences of thermal shock to RCP seals.
2.7 2.8 Rej.
5 TBD 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Cooling Water / 8 R
AA1.05 Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: The CCWS surge tank, including level control and level alarms, and radiation alarm 3.1 6
TBD 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient Without Scram / 1 R
EK1.02 Knowledge of the operational implications of the following concepts as they apply to the ATWS:
Definition of reactivity 2.6 7
TBD 000038 (EPE 38) Steam Generator Tube Rupture / 3 R
EA2.13 Ability to determine or interpret the following as they apply to a SGTR: magnitude of rupture 3.1 8
Bank 856 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 R
AK3.02 Knowledge of the reasons for the following responses as they apply to the Steam Line Rupture:
ESFAS initiation 4.4 9
TBD 000054 (APE 54; CE E06) Loss of Main Feedwater /4 R
AA1.03 Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW): AFW auxiliaries, including oil cooling water supply 3.5 10 TBD 000055 (EPE 55) Station Blackout / 6 R
EA2.03 Ability to determine or interpret the following as they apply to a Station Blackout: Actions necessary to restore power 3.9 11 Bank 1097 000056 (APE 56) Loss of Offsite Power / 6 R
2.1.31 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired lineup.
4.6 12 TBD 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 R
AA1.01 Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Manual Inverter swapping 3.7 13 TBD 000058 (APE 58) Loss of DC Power / 6 R
AA2.01 Ability to determine and interpret the following as they apply to the Loss of DC Power:
That a loss of dc power has occurred; verification that substitute power sources have come on line 3.7 14 TBD
ES-401, Page 42 of 52 000062 (APE 62) Loss of Nuclear Service Water / 4 R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
4.3 15 TBD 000065 (APE 65) Loss of Instrument Air / 8 R 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrumentation 4.4 16 TBD 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 R
AK3.01 Knowledge of the reasons for the following responses as they apply to Generator Voltage and Electric Grid Disturbances: Reactor and turbine trip criteria 3.9 17 TBD (W E04) LOCA Outside Containment / 3 N/A for this design type (W E11) Loss of Emergency Coolant Recirculation / 4 N/A for this design type (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 R
EK2.1 Knowledge of the interrelations between the Inadequate Heat Transfer and the following:
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 3.8 18 TBD K/A Category Totals:
3 3
3 3
3 3
Group Point Total:
18
ES-401, Page 43 of 52 ES-401 3
Form ES-401-2 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 Not sampled 000003 (APE 3) Dropped Control Rod / 1 Not sampled 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 Not sampled 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 Not sampled 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 Not sampled 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 R
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 2.7 19 Mod 425 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 Not sampled 000037 (APE 37) Steam Generator Tube Leak / 3 Not sampled 000037 (APE 37) Steam Generator Tube Leak / 3 Selected instead of APE78 000051 (APE 51) Loss of Condenser Vacuum / 4 R 2.1.20 Ability to interpret and execute procedure steps.
4.6 20 Bank 1062 000059 (APE 59) Accidental Liquid Radwaste Release / 9 R
AK2.01 Knowledge of the interrelations between the Accidental Liquid Radwaste Release and the following:
Radioactive-liquid monitors 2.7 21 Bank 951 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 R
AA1.02 Ability to operate and / or monitor the following as they apply to the Accidental gaseous Radwaste: Ventilation system 2.9 22 TBD 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 R
AA2.06 Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:
Required actions if alarm channel is out of service 3.2 23 TBD 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 Not sampled 000069 (APE 69; W E14) Loss of Containment Integrity / 5 R 2.1.19 Ability to use plant computers to evaluate system or component status.
3.9 24 TBD 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 R
AK1.3 Knowledge of the operational implications of the following concepts as they apply to the Inadequate Core Cooling:
Processes for removing decay heat from the core 4.5 25 TBD 000076 (APE 76) High Reactor Coolant Activity / 9 Not sampled 000078 (APE 78*) RCS Leak / 3
(W E01 & E02) Rediagnosis & SI Termination / 3 N/A for this design type (W E13) Steam Generator Overpressure / 4 N/A for this design type (W E15) Containment Flooding / 5 N/A for this design type (W E16) High Containment Radiation /9 N/A for this design type
ES-401, Page 44 of 52 (BW A01) Plant Runback / 1 R
AA2.2 Ability to determine and interpret the following as they apply to the (Plant Runback)
Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
AA2.1 Ability to determine and interpret the following as they apply to the (Plant Runback)
Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
3.5 3.0 Rej.
26 TBD (BW A02 & A03) Loss of NNI-X/Y/7 Not sampled (BW A04) Turbine Trip / 4 Not sampled (BW A05) Emergency Diesel Actuation / 6 Not sampled (BW A07) Flooding / 8 Not sampled (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 R
EK3.3 Knowledge of the reasons for the following responses as they apply to the (LOCA Cooldown):
Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.
4.0 27 TBD (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not sampled (BW E13 & E14) EOP Rules and Enclosures Not sampled (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 N/A for this design type (CE A16) Excess RCS Leakage / 2 N/A for this design type (CE E09) Functional Recovery N/A for this design type (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 N/A for this design type K/A Category Point Totals:
2 1
1 1
2 2
Group Point Total:
9
ES-401, Page 45 of 52 ES-401 4
Form ES-401-2 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 R
K6.04 Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: Containment isolation valves affecting RCP operation 2.8 28 TBD 004 (SF1; SF2 CVCS) Chemical and Volume Control R
K4.15 Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following: Interlocks associated with operation of orifice isolation valves 3.0 29 TBD 005 (SF4P RHR) Residual Heat Removal R
R K1.04 Knowledge of the physical connections and/or cause-effect relationships between the RHRS and the following systems: CVCS K4.02 Knowledge of RHRS design feature(s) and/or interlock(s) which provide or the following: Modes of Operation 2.9 3.2 30 TBD 31 TBD 006 (SF2; SF3 ECCS) Emergency Core Cooling R
R K3.01 Knowledge of the effect that a loss or malfunction of the ECCS will have on the following: RCS A1.18 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: PZR level and pressure 4.1 4.0 32 TBD 33 TBD 007 (SF5 PRTS) Pressurizer Relief/Quench Tank R
R R
A1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Monitoring quench tank temperature A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Monitoring quench tank pressure A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Overpressurization of the PZR 2.6 2.7 3.6
K2.02 Knowledge of bus power supplies to the following: CCW pump, including emergency backup 3.0 36 New 1253 010 (SF3 PZR PCS) Pressurizer Pressure Control R
K4.01 Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following: Spray valve warm-up 2.7 37 TBD 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection R
R K3.02 Knowledge of the effect that a loss or malfunction of the RPS will have on the following: T/G A3.05 Ability to monitor automatic operation of the RPS, including: Single and multiple channel trip indicators 3.2 3.6 38 TBD 39 TBD 013 (SF2 ESFAS) Engineered Safety Features Actuation R
A4.03 Ability to manually operate and/or monitor in the control room: ESFAS initiation 4.5 40 TBD 022 (SF5 CCS) Containment Cooling R 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
3.9 41 TBD
ES-401, Page 46 of 52 025 (SF5 ICE) Ice Condenser N/A 026 (SF5 CSS) Containment Spray R
A3.02 Ability to monitor automatic operation of the CSS, including: Pump starts and correct MOV positioning 4.3 42 TBD 039 (SF4S MSS) Main and Reheat Steam R
K5.05 Knowledge of the operational implications of the following concepts as the apply to the MRSS: Bases for RCS cooldown limits 2.7 43 Bank 910 059 (SF4S MFW) Main Feedwater R
R K3.02 Knowledge of the effect that a loss or malfunction of the MFW will have on the following: AFW system 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.
3.6 3.9 44 TBD 45 TBD 061 (SF4S AFW)
Auxiliary/Emergency Feedwater R
R K6.02 Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Pumps 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.
2.6 4.0 46 Mod 1252 47 TBD 062 (SF6 ED AC) AC Electrical Distribution R
R K1.04 Knowledge of the physical connections and/or cause effect relationships between the ac distribution system and the following systems: off-site power sources K2.01 Knowledge of bus power supplies to the following: Major system loads 3.7 3.3 48 TBD 49 TBD 063 (SF6 ED DC) DC Electrical Distribution R
K1.03 Knowledge of the physical connections and/or cause-effect relationships between the DC electrical system and the following systems:
Battery charger and battery 2.9 50 Bank 670 064 (SF6 EDG) Emergency Diesel Generator R
A2.19 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: Consequences of high VARS on ED/G integrity 2.5 51 TBD 073 (SF7 PRM) Process Radiation Monitoring R
K5.02 Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: Radiation intensity changes with source distance 2.5 52 TBD 076 (SF4S SW) Service Water R
A4.01 Ability to manually operate and/or monitor in the control room: SWS pumps 2.9 53 Bank 794 078 (SF8 IAS) Instrument Air R
A3.01 Ability to monitor automatic operation of the IAS, including: Air pressure 3.1 54 TBD 103 (SF5 CNT) Containment R
A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including:
Containment pressure, temperature, and humidity 3.7 55 New 1254 053 (SF1; SF4P ICS*) Integrated Control Cant sample until rev 3 of KA catalogs K/A Category Point Totals:
3 2
3 3
2 2
3 2
3 2
3 Group Point Total:
28
ES-401, Page 47 of 52 ES-401 5
Form ES-401-2 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 R
K5.12, Knowledge of the following operational implications as they apply to the CRDS: Effects on power of inserting axial shaping rods 3.4 56 TBD 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication R 2.4.6 Knowledge of EOP mitigation strategies.
3.7 57 TBD 015 (SF7 NI) Nuclear Instrumentation Not sampled 016 (SF7 NNI) Nonnuclear Instrumentation Not sampled 017 (SF7 ITM) In-Core Temperature Monitor R
A4.02 Ability to manually operate and/or monitor in the control room: Temperature values used to determine RCS/RCP operation during inadequate core cooling (i.e., if applicable, average of five highest values) 3.8 58 TBD 027 (SF5 CIRS) Containment Iodine Removal R
AK1.02 Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: Expansion of liquids as temperature increases (This is for APE 027, not SYS 027)
K1.01 Knowledge of the physical connections and/or cause-effect relationships between the CIRS and the following systems: CSS 2.8 3.4 Rej.
59 TBD 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control Not sampled 029 (SF8 CPS) Containment Purge Not sampled 033 (SF8 SFPCS) Spent Fuel Pool Cooling Not sampled 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator R
K6.02 Knowledge of the effect of a loss or malfunction on the following will have on the S/GS: secondary porv 3.1 60 New 1255 041 (SF4S SDS) Steam Dump/Turbine Bypass Control Not sampled 045 (SF 4S MTG) Main Turbine Generator R
K3.01 Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following: Remainder of the Plant 2.9 61 TBD 055 (SF4S CARS) Condenser Air Removal Not sampled 056 (SF4S CDS) Condensate R
A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: loss of Condensate pumps 2.6 62 Bank 765 068 (SF9 LRS) Liquid Radwaste Not sampled 071 (SF9 WGS) Waste Gas Disposal R
K4.04 Knowledge of the design features and/or interlocks which provide for the following:
Isolation of waste gas release tanks 2.9 63 Bank 470 072 (SF7 ARM) Area Radiation Monitoring R
A3.01 Ability to monitor automatic operation of the ARM system, including: Changes in ventilation alignment 2.9 64 Mod 153
ES-401, Page 48 of 52 075 (SF8 CW) Circulating Water Not sampled 079 (SF8 SAS**) Station Air Not sampled 086 Fire Protection R
A1.03 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits associated with operating the Fire Protection System controls, including: Fire doors 2.7 65 TBD 050 (SF 9 CRV*) Control Room Ventilation N/A until rev3 of KA catalogs K/A Category Point Totals:
1 0
1 1
1 1
1 1
1 1
1 Group Point Total:
10
ES-401, Page 49 of 52 ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility:
Date of Exam:
Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.34 Knowledge of primary and secondary plant chemistry limits.
2.7 66 TBD 2.1.40 2.1.32 Knowledge of refueling administrative requirements.
Ability to explain and apply system limits and precautions.
2.8 3.8 Rej.
67 TBD 2.1.21 Ability to verify the controlled procedure copy 3.5 68 Bank 389 Subtotal 3
- 2. Equipment Control 2.2.13 Knowledge of tagging and clearance procedures.
4.1 69 TBD 2.2.20 2.2.43 Knowledge of the process for managing troubleshooting activities.
Knowledge of the process used to track inoperable alarms 2.6 3.0 Rej.
70 TBD Subtotal 2
- 3. Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
2.9 71 TBD 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.
3.5 72 Bank 996 Subtotal 2
- 4. Emergency Procedures/Plan 2.4.6 Knowledge of EOP mitigation strategies.
3.7 73 Bank 803 2.4.9 2.4.25 Knowledge of low power/shutdown implications in accident (e.g.,
loss of coolant accident or loss of residual heat removal) mitigation strategies.
Knowledge of fire protection procedures.
3.8 3.3 Rej.
74 Bank 848 2.4.46 2.4.3 Ability to verify that the alarms are consistent with the plant conditions.
Ability to identify post-accident instrumentation.
4.2 3.7 Rej.
75 TBD Subtotal 3
Tier 3 Point Total 10
ES-401, Page 50 of 52 ES-401 Record of Rejected K/As Form ES-401-4 Tier/Group Randomly Selected K/A Reason for Rejection RO T1/G1 022 AK1.02 (2.7)
(Q#5) 022 AK1.01 (2.8) 022 - Loss of Reactor Coolant Makeup. The original K/A concerns a potential loss of reactor coolant makeup as RCS pressure rises. This is not applicable on Unit 1 for several reasons. RCS makeup flow is controlled with a makeup valve that controls pressurizer level. By design, the shutoff head of the makeup pumps is so high that any change in RCS pressure wont be enough to affect makeup flow in the long term. Also, the Loss of Makeup AOP only addresses leaks within the makeup system or a trip of a running makeup pump. Randomly selected AK1.01 which focuses on the consequences of thermal shock to RCP seals after a loss of makeup scenario.
RO T1/G2 BW A01 AA2.2 (3.5)
(Q#26)
BW A01 AA2.1 (3.0)
A01 - Plant Runback. The original K/A concerns the adherence to appropriate procedures and operation within the limitations in the facilities license and amendments as applicable to a Plant Runback. There is no direct relationship to a Plant Runback condition and limitations within the facility license. Randomly selected AA2.1 which concerns the facility conditions and selection of appropriate procedures during abnormal and emergency conditions.
RO T2/G1 007 A1.03 (2.6)
(Q#34) 007 A1.02 (2.7) 007 - Pressurizer Relief / Quench Tank. The original K/A was the ability to predict or monitor quench tank temperature to prevent exceeding design limits. While there is indication of quench tank temperature, there is no procedural limitation on how high or low it should be. Randomly selected A1.02 which concerns quench tank pressure which has procedural limitations.
RO T2/G2 027 AK1.02 (2.8)
(Q#59) 027 K1.01 (3.4) 027 - Containment Iodine Removal. The original K/A is a typo using APE 027 not SYS 027 for Tier 2. Selected the only K1 K/A for SYS 027 which is K1.01.
RO T3 2.1.40 (2.8)
(Q#67) 2.1.32 (3.8)
The original K/A concerns the knowledge of refueling administrative limits. This is a SRO topic and shouldnt be asked on the RO portion of the exam. Randomly selected 2.1.32 which concerns the application of system limits and precautions.
RO T3 2.2.20 (2.6)
(Q#70) 2.2.43 (3.0)
The original K/A concerns the process for managing troubleshooting activities. IAW EN-MA-125, the management and approvals of troubleshooting come from the SM and or FIN SRO. The RO position is not involved in any portion of these types of activities therefore this question shouldnt be asked on the RO portion of the exam. Randomly selected 2.2.43 which concerns the process of tracking inoperable alarms.
RO T3 2.4.9 (3.8)
(Q#74) 2.4.25 (3.3)
The original K/A concerns specific mitigation strategies for a low power / shutdown accident. This topic belongs in T1, not T3. Randomly selected 2.4.25 which concerns the knowledge of fire protection procedures.
RO T3 2.4.46 (4.2)
(Q#75) 2.4.3 (3.7)
The original K/A concerns the ability to verify alarms are consistent with plant conditions. This topic belongs in T2, not T3. Randomly selected 2.4.3 which concerns the ability to identify post-accident instrumentation.
ES-401 PWR Examination Outline Form ES-401-2 Facility: ANO U1 Date of Exam: 4/8/2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total 1.
Emergency and Abnormal Plant Evolutions 1
18 3
3 6
2 9
2 2
4 Tier Totals 27 5
5 10 2.
Plant Systems 1
28 2
3 5
2 10 1
1 1
3 Tier Totals 38 4
4 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 2
1 2
Note:
- 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As
- These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
- These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 PWR Examination Outline Form ES-401-2 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 S
EA2.06 Ability to determine or interpret the following as they apply to a small break LOCA: Whether PZR water inventory loss is imminent 4.3 76 New 1243 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 S
AA2.05 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Limitations on LPI flow and temperature rates of change 3.5 77 Mod 1244 000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 S
S AA2.14 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: RCP injection flow AA2.11 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: RCS Pressure 2.9 4.1 Rej.
78 New 1256 000029 (EPE 29) Anticipated Transient Without Scram / 1 000038 (EPE 38) Steam Generator Tube Rupture / 3 S 2.4.41 Knowledge of the emergency action level thresholds and classifications.
4.6 79 New 1257 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 S
2.4.6 Knowledge of EOP mitigating strategies 4.7 80 New 1246 000054 (APE 54; CE E06) Loss of Main Feedwater /4 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 S
S 2.4.42 Ability to recognize system parameters that are entry-level conditions for technical Specifications.
2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.
4.6 4.3 Rej.
81 New 1258 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 (W E04) LOCA Outside Containment / 3 N/A for this design type (W E11) Loss of Emergency Coolant Recirculation / 4 N/A for this design type (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals:
3 3
Group Point Total:
6
ES-401 PWR Examination Outline Form ES-401-2 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 Not sampled 000003 (APE 3) Dropped Control Rod / 1 Not sampled 000005 (APE 5) Inoperable/Stuck Control Rod / 1 S 2.2.38 Knowledge of conditions and limitations in the facility license.
4.5 82 New 1245 000024 (APE 24) Emergency Boration / 1 Not sampled 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction/2 Not sampled 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 Not sampled 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 Not sampled 000037 (APE 37) Steam Generator Tube Leak / 3 Not sampled (Sampled below) 000037 (APE 37) Steam Generator Tube Leak / 3 (Selected instead of APE78 below)
S 2.2.40 Ability to apply Technical Specifications for a system.
4.7 84 New 1260 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 S
S AA2.13 Ability to determine and interpret the following as they apply to the Plant Fire on Site:
Need for emergency plant shutdown AA2.17 Ability to determine and interpret the following as they apply to plant fire on site: systems that may be affected by the fire.
4.4 4.3 Rej.
83 New 1259 000068 (APE 68; BW A06) Control Room Evacuation / 8 Not sampled 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07)
Inadequate Core Cooling / 4 000076 (APE 76) High Reactor Coolant Activity / 9 Not sampled 000078 (APE 78*) RCS Leak / 3
- Rev 2 of KAs. APE78 doesnt exist.
(Selected APE37 for SG tube leak.)
S 2.2.40 Ability to apply Technical Specifications for a system.
4.7 Rej.
(W E01 & E02) Rediagnosis & SI Termination / 3 N/A for this design type (W E13) Steam Generator Overpressure / 4 N/A for this design type (W E15) Containment Flooding / 5 N/A for this design type (W E16) High Containment Radiation /9 N/A for this design type (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 Not sampled (BW A04) Turbine Trip / 4 Not sampled
(BW A05) Emergency Diesel Actuation / 6 Not sampled (BW A07) Flooding / 8 Not sampled (BW E03) Inadequate Subcooling Margin / 4 S
EA2.1 Ability to determine and interpret the following as they apply to the (Inadequate Subcooling Margin) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
4.0 85 New 1251 (BW E08; W E03) LOCA Cooldown Depressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not sampled (BW E13 & E14) EOP Rules and Enclosures Not sampled (CE A11**; W E08) RCS Overcooling Pressurized Thermal Shock / 4 N/A for this design type (CE A16) Excess RCS Leakage / 2 N/A for this design type (CE E09) Functional Recovery N/A for this design type (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 N/A for this design type K/A Category Point Totals:
2 2
Group Point Total:
4
ES-401 PWR Examination Outline Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump 004 (SF1; SF2 CVCS) Chemical and Volume Control 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 010 (SF3 PZR PCS) Pressurizer Pressure Control S
A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater Failures 3.6 86 New 1261 012 (SF7 RPS) Reactor Protection S 2.2.37 Ability to determine operability and/or availability of safety related equipment 4.6 87 New 1247 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser N/A 026 (SF5 CSS) Containment Spray 039 (SF4S MSS) Main and Reheat Steam S 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits 4.2 88 New 1248 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)
Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical Distribution 063 (SF6 ED DC) DC Electrical Distribution 064 (SF6 EDG) Emergency Diesel Generator 073 (SF7 PRM) Process Radiation Monitoring S 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.
4.6 89 New 1262 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air
103 (SF5 CNT) Containment S
S A2.02 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 Necessary plant conditions for work in containment 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.
3.2 3.8 Rej.
90 TBD 053 (SF1; SF4P ICS*) Integrated Control Cant sample until rev 3 of KA catalogs K/A Category Point Totals:
2 3
Group Point Total:
5
ES-401 PWR Examination Outline Form ES-401-2 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 S
S 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 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
4.1 4.5 Rej.
91 New 1263 011 (SF2 PZR LCS) Pressurizer Level Control S
A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of PZR level 3.9 92 New 1249 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation Not sampled 016 (SF7 NNI) Nonnuclear Instrumentation Not sampled 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control Not sampled 029 (SF8 CPS) Containment Purge Not sampled 033 (SF8 SFPCS) Spent Fuel Pool Cooling Not sampled 034 (SF8 FHS) Fuel-Handling Equipment S
K1.02 Knowledge of the physical connections and/or cause effect relationships between the Fuel Handling System and the following systems: RHRS 3.2 93 TBD 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control Not sampled 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal Not sampled 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste Not sampled 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water Not sampled 079 (SF8 SAS**) Station Air Not sampled 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation N/A until rev3 of KA catalogs K/A Category Point Totals:
1 1
1 Group Point Total:
3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility:
Date of Exam:
Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.
3.8 94 Bank 407 2.1.36 Knowledge of procedures and limitations involved in core alterations 4.1 95 New 1250 Subtotal 2
- 2. Equipment Control 2.2.6 Knowledge of the process for making changes to procedures 3.6 96 Bank 1082 2.2.12 Knowledge of surveillance procedures 4.1 97 TBD Subtotal 2
- 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.
3.8 98 TBD Subtotal 1
- 4. Emergency Procedures/Plan 2.4.4 2.4.14 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
Knowledge of general guidelines for EOP usage.
4.7 4.5 Rej.
99 TBD 2.4.20 2.4.30 Knowledge of the operational implications of EOP warnings, cautions, and notes.(Not T3 K/A. Moved to T1)
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 4.3 4.1 Rej.
100 Bank 411 Subtotal 2
Tier 3 Point Total 7
ES-401 Record of Rejected K/As Form ES-401-4 Tier/Group Randomly Selected K/A Reason for Rejection SRO T1/G1 027 AA2.14 (2.9)
(Q#78) 027 AA2.11 (4.1) 027 - Pressurizer Pressure Control System Malfunction. The original K/A concerned interpreting RCP injection flow given a PZR control system malfunction. This is not applicable on Unit 1 as RCP injection flow has its own flow controller which will maintain flow based on its own set point. Also the 1203.015 doesnt specifically discuss any effects on RCP seal injection flow for any type of control system failure.
Randomly selected AA2.11 which concerns the effect on RCS pressure.
SRO T1/G1 056 2.4.42 (3.8)
(Q#81) 056 2.4.20 (4.3) 056 - Loss of Offsite Power. Since this is a T1 K/A, generic K/As must be selected from the list provided in D.1.b of ES-401. The original KA was not found in this list.
Utilized 2.4.20 which was originally in the T3 section at Q#100, but was rejected since it is a T1 K/A not a T3 K/A.
SRO T1/G2 067 AA2.13 (4.4)
(Q#83) 067 AA2.17 (4.3) 067 - Plant Fire On Site. The original K/A concerned the need for an emergency plant shutdown based on a plant fire. 1203.49 is written generically to first allow assessment from the operators and then later determine if a reactor trip is required. There is no specific trip criteria based on a fire alone, so any conditions given in a question that would require a reactor trip would already have guidance in another EOP or AOP.
Randomly selected AA2.17 which concerns systems that may be affected by an onsite fire.
SRO T1/G2 078 2.2.40 (4.7)
(Q#84) 037 2.2.40 (4.7) 078 - RCS Leak. This APE is not in Rev 2 K/A catalog but will be released in Rev 3.
Randomly selected APE37 Steam Generator Tube Leak instead.
SRO T2/G1 103 A2.02 (3.2)
(Q#90) 103 A2.03 (3.8) 103 - Containment System. The original K/A concerns conditions needed for work in containment given a malfunction or operation of the containment system and a procedure that would be used to control or mitigate the situation. There is no clear procedural guidance on when entry into containment is allowed based upon containment conditions. This would be governed by Radiation Protection or industrial safety procedures/personnel. Randomly selected A2.03 which concerns the procedures that would be used to mitigate or control the effects of containment SRO T2/G2 002 2.4.30 (4.1)
(Q#91) 002 2.4.8 (4.5) 002-Reactor Coolant System. The original K/A concerns the knowledge of system operation/status that must be reported to outside agencies. This K/A belongs in T3, not T2. Since the K/A for Q#100 in T3 was rejected, moved 2.4.30 for Q#91 to Q#100.
Randomly selected K/A 2.4.8 for Q#91 which is the knowledge of how AOPs are used with EOPs.
SRO T3 2.4.4 (4.7)
(Q#99) 2.4.14 (4.5)
The original K/A concerns the recognition of entry conditions for EOPs and AOPs. This is RO level knowledge. Randomly selected 2.4.14 which concerns the general guidelines for EOP usage.
SRO T3 2.4.20 (4.3)
(Q#100) 2.4.30 (4.1)
The original K/A concerns the knowledge of the operational implications of notes cautions and warnings of EOPs. This is a T1 question and was moved to Q#81 since Q#81 K/A was rejected. Moved 2.4.30 for Q#91 to Q#100 because it is a T3 K/A.
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
ANO Unit 1 Date of Examination:
3/30/2020 Examination Level: RO SRO Operating Test Number:
1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed A1 Conduct of Operations K/A - 2.1.4, Importance Rating 3.3 R,N Determine Active / Inactive Status of Licensed Operator A1JPM-RO-ADMINAI A2 Conduct of Operations K/A - 2.1.19, Importance Rating 3.9 S,D Operate Plant Computer to disable an alarm A1JPM-RO-PMS3 A3 Equipment Control K/A - 2.2.12, Importance Rating 3.7 R,M Conduct Shift Surveillance Test to determine operability of EDG Fuel Transfer Pump A1JPM-RO-SURV3 A4 Radiation Control K/A - 2.3.7, Importance Rating 3.5 R,D,P Ability to comply with Radiation Work Permit requirements A1JPM-RO-ADMIN-RWP3 Emergency Plan 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)
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
ANO Unit 1 Date of Examination:
3/30/2020 Examination Level: RO SRO Operating Test Number:
1 Administrative Topic (see Note)
Type Code*
Describe activity to be performed A5 Conduct of Operations K/A - 2.1.5, Importance Rating 3.9 M,R Perform working hour history review and select eligible operators to fill a vacancy A1JPM-NRC-WHHR A6 Conduct of Operations K/A - 2.1.25, Importance Rating 4.2 N,R Ability to interpret Rod Insertion Limits per COLR for operability A1JPM-SRO-ADMINCURV A7 Equipment Control K/A - 2.2.37, Importance Rating 4.6 D,P,R Determine OPERABILITY of MSSV and apply Technical Specifications A1JPM-SRO-ADMINMSSV A8 Radiation Control K/A - 2.3.4, Importance Rating 3.7 M,R Provided with the dose history of 5 individuals, determine which are eligible for containment power entry A1JPM-SRO-RAD3 A9 Emergency Plan K/A - 2.4.41, Importance Rating 4.6 N,R Determine appropriate Emergency Action Level A1JPM-SRO-EAL17 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)
ES-301 Administrative Topics Outline Form ES-301-1
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
ANO Unit 1 Date of Examination:
3/30/2020 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 System/JPM Title Type Code*
Safety Function S1. Perform RCS Boration BATCH Feed Method 004 A4.07 (RO 3.9 / SRO 3.7)
RO D/S 1
S2. Perform actions required for ESAS (Step 14) 006 A1.17 (RO 4.2 / SRO 4.3)
RO / SRO-I / SRO-U A/EN/L/N/S 2
S3. Manually control RCS Pressure with a Pressurizer Spray Valve Failure 010 A2.02 (RO 3.9 / SRO 3.9)
RO / SRO-I A/D/P/S 3
S4. Stop RCP at Power with Reverse Rotation 003 A2.02 (RO 3.7 / SRO 3.9)
RO / SRO-I A/D/P/S 4A S5. Initiate Common Feedwater E04 EA1.1 (RO 4.4 / SRO 4.2)
RO / SRO-I N/L/S 4B S6. Respond to Inadvertent ESAS Actuation 0026 A3.01 (RO 4.3/SRO 4.5)
RO / SRO-I / SRO-U D/EN/S 5
S7. Synchronize and Load #1 EDG with a failure of the load switch 064 A2.05 (RO 3.1 / SRO 3.2)
RO / SRO-I A/D/P/S 6
S8. Bypass MSLI E02 EA1.1 (RO 4.0 / SRO 3.6)
RO / SRO-I L/N/S 7
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. Perform ICS Startup 041 A4.01 (RO 2.9 / SRO 3.1)
RO / SRO-I / SRO-U A/D/L 7
P2. Halon System Manual Actuation 086 A4.06 (RO 3.2 / SRO 3.2)
RO / SRO-I SRO-U A/D/E 8
P3. Commence Waste Gas Release 071 A4.26 (RO 3.1 / SRO 3.9)
RO / SRO-I / SRO-U D/R 9
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 6 / 6 / 3 8 / 7 / 4 1 / 1 / 1 2 / 2 / 2 (control room system) 4 / 4 / 2 3 / 3 / 1 3 / 3 / 0 (randomly selected) 1 / 1 / 1
ES-301 Transient and Event Checklist Form ES-301-5 Facility: ANO Unit 1 Date of Exam: 3/30/2020 Operating Test No.: 1 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios 1
2 3
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
U1 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 7
2,3,4 5,6,7 9,10 12 4
4 2 MAJ 5
8 2
2 2 1 TS 3,4 2,3,4 5
0 2 2 U2 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 5,6,7 9
2,3,4 5,6,7 9,10 15 4
4 2 MAJ 8
8 2
2 2 1 TS 2,5 2,3,4 5
0 2 2 U3 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 5
2,3,4 5,6,7 9,10 12 4
4 2 MAJ 6
8 2
2 2 1 TS 1,2,3 2,3,4 6
0 2 2 I1 RX 0
1 1 0 NOR 1
1 1
3 1
1 1 I/C 2,3,4 7
3,4 6,9 10 2,3,4 5
13 4
4 2 MAJ 5
8 6
3 2
2 1 TS 3,4 1,2,3 5
0 2 2
ES-301 Transient and Event Checklist Form ES-301-5 Facility: ANO Unit 1 Date of Exam: 3/30/2020 Operating Test No.: 1 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios 1
2 3
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
I2 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 5,6,7 9
3,4,5 2,3,4 5,6,7 9,10 18 4
4 2 MAJ 8
6 8
3 2
2 1 TS 2,5 2,3,4 5
0 2 2 I3 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 2,3,4 7
3,4,6, 7,9 3,4,5 12 4
4 2 MAJ 5
8 6
3 2
2 1 TS 3,4 2
0 2 2 I4 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,3,4 6,7 2,3,4 5,6,7 10 2,3,4
,5 16 4
4 2 MAJ 5
8 6
3 2
2 1 TS 2,5 1,2,3 5
0 2 2 R9 RX 0
1 1 0 NOR 1
1 1
3 1
1 1 I/C 2,4, 7
3,4 6,9 10 2,3,4 7
12 4
4 2 MAJ 5
8 6
3 2
2 1 TS 0
0 2 2
ES-301 Transient and Event Checklist Form ES-301-5 Facility: ANO Unit 1 Date of Exam: 3/30/2020 Operating Test No.: 1 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios 1
2 3
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
R1 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 2,3,4 6,7 3,4,6 10 9
4 4 2 MAJ 5
8 2
2 2 1 TS 0
0 2 2 R2 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 2,4, 7
3,4,5 7,9, 10 9
4 4 2 MAJ 5
8 2
2 2 1 TS 0
0 2 2 R3 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 3,4, 6,9, 10 3,4,6 10 9
4 4 2 MAJ 8
8 2
2 2 1 TS 0
0 2 2 R4 RX 0
1 1 0 NOR 0
1 1 1 I/C 3,4,6 7,9 3,4,5 7,9, 10 11 4
4 2 MAJ 8
8 2
2 2 1 TS 0
0 2 2
ES-301 Transient and Event Checklist Form ES-301-5 Facility: ANO Unit 1 Date of Exam: 3/30/2020 Operating Test No.: 1 A
P P
L I
C A
N T
E V
E N
T T
Y P
E Scenarios 1
2 3
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
R5 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 3,4,5 3,4,6 10 7
4 4 2 MAJ 6
8 2
2 2 1 TS 0
0 2 2 R6 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 2,3,4 7
3,4,5 7,9, 10 10 4
4 2 MAJ 6
8 2
2 2 1 TS 0
0 2 2 R7 RX 0
1 1 0 NOR 1
1 1
1 1 I/C 2,3,4 6,7 3,4,6 7,9 3,4,6 10 14 4
4 2 MAJ 5
8 8
3 2
2 1 TS 0
0 2 2 R8 RX 0
1 1 0 NOR 1
1 2
1 1 1 I/C 2,4, 7
2,3,4 7
3,4,5 7,9, 10 13 4
4 2 MAJ 5
6 8
3 2
2 1 TS 0
0 2 2
ES-301 Transient and Event Checklist Form ES-301-5 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.
Appendix D Scenario Outline Page 1 of 5 Facility: ANO-1 Scenario No.: 1 R0 Op-Test No.: 2020-1 Examiners: ____________________________Operators: _____________________________
Initial Conditions:
- IC201
- 99.7% Power.
- C-28B IA Compressor is out of service for overhaul
- Run Schedule File Scenario 1 R0
- P-3A,B,C CW Pumps in service
- RPS Failed, will not automatically trip reactor Turnover:
- CBOT place P-33B in service & secure P-33A Event No.
Malf. No.
Event Type*
Event Description 1
N/A N-(BOP)
N-(SRO)
Shift ICW Pumps, Place P-33B in service, Secure P-33A 2
CW065 C-(ATC)
C-(BOP)
C-(SRO)
ACA P-3A Circ Water Pump Sheared Shaft 3
TR049 I-(ATC)
I-(SRO)
TS, AOP LT-1001 Pressurizer Level Fails Low (T.S. 3.3.15) 4 RC045 C-(ATC)
C-(BOP)
C-(SRO)
TS, AOP Pressurizer Steam Space Leak (15 gpm) (T.S. 3.4.13) 5 RC045 M-(ALL)
Pressurizer Steam Space Leak increases to ~800 gpm 6
RP246 -
RP249 C-(ATC)
CT RPS is failed. ATC must manually trip the reactor 7
RC045 C-(ATC)
C-(BOP)
C-(SRO)
CT RCPs must be secured within 2 minutes of LOSM (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Page 2 of 5 Target Quantitative Attributes (Section D.5.d)
Actual Attributes Malfunctions after EOP entry (1-2) 1 Abnormal Events (2-4) 3 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per scenario set) 1 Critical Tasks (2) 2 SCENARIO 1 OBJECTIVES
- 1)
Evaluate individual ability to perform switching of ICW Pumps
- 2)
Evaluate individual ability to recognize and respond to a Circulating Water Pump sheared shaft.
- 3)
Evaluate individual ability to recognize and respond to a failed pressurizer level transmitter
- 4)
Evaluate individual ability to recognize when conditions require the entry into technical specifications conditions.
- 5)
Evaluate individual ability to estimate RCS leakage rate.
- 6)
Evaluate individual ability to recognize and respond to excess RCS leakage.
- 7)
Evaluate individual ability to reduce plant power.
- 8)
Evaluate individual ability to recognize and respond to a failure of RPS.
- 9)
Evaluate individual ability to recognize and respond to Reactor Trip.
- 10) Evaluate individual ability to recognize and respond to a Loss of Subcooling Margin.
Appendix D Scenario Outline Page 3 of 5 SCENARIO 1 NARRATIVE Event One: Normal operation to swap ICW Pumps The crew will assume plant responsibility at ~100% power. The SRO will direct swapping ICW Pumps in accordance with OP-1104.028 Section 10.0. The BOP will open cross-connect valves, direct venting the pump, then start P-33B and after 3 minutes stop P-33A. (BOP-N)
(SRO-N)
Event Two: P-3A Circ Water Pump sheared shaft The first indication will be lowering vacuum. The crew will then receive K05-B2, Condenser Vacuum Low alarm and take actions as directed by the ACA. The SRO will enter Loss of Condenser Vacuum AOP and direct the ATC to maintain power less than 100% and gradually lower power to stabilize vacuum. The BOP should diagnose the sheared shaft with the help of the Outside AOs report of back flow coming from the A CW Bay. The SRO will direct starting P-3D and stopping P-3A. Once vacuum is stabilized the down power will be stopped. (ATC-C)
(BOP-C) (SRO-C) AOP Event Three: Pressurizer level transmitter (LT-1001) fails low Once the plant is stabilized, the controlling pressurizer level transmitter will fail low. This will cause K09-C3, Pressurizer Level Lo and K09-A3, Pressurizer Level Lo-Lo. The crew should diagnose the failure by comparing diverse indication of pressurizer level and the lack of RCS pressure drop. The SRO will utilize OP-1203.015, Pressurizer Systems Failure, Section 4 -
pressurizer Level Indication Malfunction. The ATC will determine that LT-1002 is the valid signal and select it for level control. The SRO will enter Post Accident Monitoring (PAM)
Instrumentation T.S 3.3.15. (ATC-C) (SRO-C) AOP, TS Event Four: Pressurizer Steam Space Leak (15 gpm) develops The crew will recognize indications of an RCS leak (Makeup flow rising, Reactor Building temperature rising, RCS pressure lowering) and receive K10-B2, Process Monitor Radiation Hi alarm. The crew will calculate a leak rate of ~15 gpm. The SRO will enter OP-1203.039, Excess RCS Leakage AOP and direct the crew accordingly. The SRO will also enter RCS Operational Leakage T.S. 3.4.13 Condition A. The crew will commence a plant shutdown at a rate ~ 5%/min in accordance with OP-1203.045, Rapid Plant Shutdown (ATC-C) (BOP-C)
(SRO-C) AOP, TS Event Five / Six: Pressurizer Steam Space Leak (~800 gpm) RPS failed, will not automatically trip the reactor Once the TS entry is announced RCS leak will get larger which will ensure that a plant maneuver has started. Additionally RPS is failed such that a manual reactor trip will be required when an RPS trip setpoint is exceeded. The critical task of tripping the reactor must be performed within one minute of exceeding an RPS trip setpoint. This is a time critical operator action as described in EN-OP-123, Time Critical Action/Time Sensitive Action Program Standard. This critical task will be met even if the crew decides to trip prior to exceeding a setpoint. (ATC-C) (BOP-C) (SRO-C) EOP, CT CT Justification: CT-23 Safety significance - Failure of Reactor Protection System (RPS)
Initiating Cue - RCS Pressure <1800 psig Measurable Performance Standard - Manually trip the reactor within 1 minute of exceeding the RPS trip setpoint. This criterion is based on EN-OP-123, Time Critical Action / Time Sensitive Action Program Standard, which describes the following: On failure of RPS trip and failure of Main Reactor Trip pushbuttons. Within 1 minute, trip CRD backup trip breakers (A-501 and B-631) from C03. While this is not exactly the same situation, one minute provides adequate time for the crew to recognize the condition and trip the plant as required.
Performance Feedback - Reactor tripped, all rods inserted, and reactor power dropping.
Appendix D Scenario Outline Page 4 of 5 NOTE to Kelly - This CT will likely be completed before the RCS pressure reaches 1800 psig (it takes ~10 seconds to reach setpoint). In other words, conservative action by the crew will take place when they realize the rate of pressure drop due to the steam space leak. I want to make sure with you, during this initial submittal that you agree we still have a CT even if we dont exceed an RPS setpoint prior to the trip.
Event Seven: Steam Space Leak results in a Loss of Subcooling Margin Following the reactor trip RCS pressure will continue to drop resulting in a Loss of Subcooling Margin, this will require a transition to the contingency EOP. The second critical task of tripping all RCPs within two minutes of a LOSM will be required during this event. The SRO will direct the crew in accordance with OP-1202.002 Loss of Subcooling Margin. (ATC-C) (BOP-C) (SRO-C) EOP, CT CT Justification: CT-1 Safety significance - Failure of a fission product barrier (RCS)
Initiating Cue - SCM monitor on ICCMDS indicates <30 oF and timer counting. Train A/B ICC Event alarms on K11.
Measureable Performance Standard - Trip ALL RCPs within 2 minutes of LOSM following the reactor trip.
Performance Feedback - Change of status lamps on control console for RCPs and indicated RCS flow near zero.
The scenario can be terminated at the discretion of the lead examiner.
Appendix D Scenario Outline Page 5 of 5 Simulator Instructions for Scenario 1 Reset simulator to MOL ~100% power IC steady state Ensure malfunctions agree with scenario guide (either schedule file or IC)
Event No.
Time Event Type Description 1
T=0 N-(BOP)
N-(SRO)
Shift ICW Pumps, Place P-33B in service, Secure P-33A 2
T=5 C-(ATC)
C-(BOP)
C-(SRO)
ACA P-3A Circ Water Pump Sheared Shaft 3
T=15 I-(ATC)
I-(SRO)
TS, AOP LT-1001 Pressurizer Level Fails Low (T.S. 3.3.15) 4 T=20 C-(ATC)
C-(BOP)
C-(SRO)
TS, AOP Pressurizer Steam Space Leak (15 gpm) (T.S. 3.4.13) 5 T=30 M-(ALL)
Pressurizer Steam Space Leak increases to ~800 gpm 6
T=40 C-(ATC)
CT RPS is failed. ATC must manually trip the reactor 7
T=45 C-(ATC)
C-(BOP)
C-(SRO)
CT RCPs must be secured within 2 minutes of LOSM
Appendix D Scenario Outline Page 1 of 6 Facility: ANO-1 Scenario No.: 2 R0 Op-Test No.: 2020-1 Examiners: ____________________________Operators: _____________________________
Initial Conditions:
- IC202
- 99.7% Power.
- P-75 Auxiliary Feedwater Pump Tagged Out for oil sample / change
- Run Schedule File Scenario 2 R0
- P-33B ICW Pump auto start failure Turnover:
Event No.
Malf. No.
Event Type*
Event Description 1
N/A N-(ATC)
N-(SRO)
Align for 2 minute delithiation 2
K09-E3 C-(SRO)
TS Pressurizer Heater Ground Fault 3
C-(BOP)
C-(SRO)
AOP Degrading Vacuum 4
B2564 SW071 C-(ATC)
C-(BOP)
C-(SRO)
ACA P-33C ICW Pump trips with an auto start failure on P-33B ICW Pump.
5 K01-A5 C-(SRO)
TS Inverter Y11 failure 6
IC09 I-(ATC)
I-(BOP)
I-(SRO)
AOP Main Steam Header Pressure bias failure 7
AI_TIC4018S C-(BOP)
C-(SRO)
ACA, CT Generator H2 Temp Controller Setpoint (TIC-4018) failure 8
MS131 M-(ALL)
A Main Steam Line break inside containment 9
ES263 ES264 I-(ATC)
I-(BOP)
I-(SRO)
CT ESAS Channels 5 & 6 fail to actuate automatically 10 CV2670 HIC2646 C-(ATC)
CT EFIC Vector Isolation failure for one EFW flow path to failed generator (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Page 2 of 6 Target Quantitative Attributes (Section D.5.d)
Actual Attributes Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 4 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per scenario set) 1 Critical Tasks (2) 2 SCENARIO 2 OBJECTIVES
- 1)
Evaluate individual ability to perform RCS Delithiation.
- 2)
Evaluate individual ability to recognize and respond to ICW System Annunciators on K-12.
- 3)
Evaluate individual ability to recognize and recover letdown flow following high letdown temperature.
- 4)
Evaluate individual ability to recognize and respond to Inoperative Pressurizer Heaters.
- 5)
Evaluate individual ability to recognize when conditions require the entry into Technical Specifications conditions.
- 6)
Evaluate individual ability to recognize and respond to high generator gas temperatures.
- 7)
Evaluate individual ability to recognize and respond to 120 VAC Inverter Annunciators on K-1.
- 8)
Evaluate individual ability to recognize and respond to Overcooling.
- 9)
Evaluate individual ability to recognize and respond to ESAS Actuation Annunciators on K-11.
- 10) Evaluate individual ability to recognize the need to perform manual intervention to control EFW.
Appendix D Scenario Outline Page 3 of 6 SCENARIO 2 NARRATIVE Event One: Normal operation align for two minute delithiation The crew will assume plant responsibility at ~100% power. The SRO will direct commencing Chemistry requested two minute delithiation in accordance with OP-1104.002 Section10.0.
(ATC-N) (SRO-N)
Event Two: Pressurizer Heater ground fault Following the delithiation, an annunciator will alert the crew of a pressurizer heater ground fault K09-E3, the SRO will direct the crew from OP-1203.015, Pressurizer System Failure -
Inoperative Pressurizer Heater section. The BOP will direct the NLO to investigate breakers in accordance with the ACA. Report from the field will be that the breaker for Bank 2 vital powered heaters cannot be closed. The SRO will declare Bank 2 Inoperable and enter T.S.
(T.S 3.4.9 Condition C) (SRO-C) TS Event Three: Degrading Vacuum due to low seal water pressure Once the T.S. entry has been announced to the crew, vacuum will begin to degrade due to a loss of seal water to the in service vacuum pump. K05-B2, Condenser Vacuum Lo, will alert the crew to the degrading vacuum. The SRO will direct the crew from OP-1203.016, Loss of Condenser Vacuum and OP-1203.045, Rapid Plant Shutdown. The ATC will take SG/RX to HAND and initially maintain power <100%, then will continue to lower power in an attempt to stabilize vacuum. The BOP will direct the filed operators to look for issues that could be causing the lowering vacuum. A report from the field will indicate that a loss of seal water pressure to in service vacuum pump, C-5A is the cause of the transient. The BOP will start the standby vacuum pump and secure the running vacuum pump. Vacuum will then begin to recover. The ATC should then stabilize reactor power.
Event Four: P-33C ICW Pump trips with a failure of P-33B to automatically start Once vacuum and power are stabilized, the running Nuclear ICW Pump (P-33C) will trip with a failure of the standby pump P-33B to automatically start. This will require the ATC to diagnose and manually start P-33B ICW pump. It will also result in a high temperature automatic isolation of the letdown flowpath. The BOP will utilize OP-1104.002 Section 14.0, Recovery of Letdown Following High Letdown Temperature. (ATC-C) (BOP-C) (SRO-C) ACA Event Five: Inverter Y11 Failure Once letdown is recovered, Annunciator K01-A5 will alarm to inform the control room of an RS1 Inverter Trouble. The BOP will dispatch the NLO to investigate. Report from the field will be that the Y11 Inverter has failed and has automatically selected the Alternate Power Supply. The SRO will declare Y11 Inoperable and enter T.S. 3.8.7 Condition A and direct the NLO to make preparations for placing Y13 into service. (SRO-C) TS Event Six: Main Steam Header Pressure Bias Failure Once the T.S. entry has been announced to the crew, the 50 psig bias to the Turbine Bypass Valves (TBVs) will fail to zero. This will cause the TBVs to open slightly, which will lower header pressure and MWe output. The Turbine Governor Valves will close and rods will pull in an attempt to raise header pressure. The ATC will take SG/RX to hand and keep power less than 100% while the BOP places the Turbine in Manual and the TBVs in Hand and closed. The ATC and BOP will work together to maintain power <100% while recovering header pressure. Once header pressure is recovered, the Turbine and SG/RX can be returned to automatic control.
The TBVs will remain in Hand.
Appendix D Scenario Outline Page 4 of 6 Event Seven: Generator H2 Temperature Controller Setpoint Failure When directed by the lead examiner, the setpoint for the Main Generator H2 temperature controller will fail high causing the controller to raise H2 temperature. The crew will respond to the rising temperatures or the annunciator K04-B6. The BOP will take manual control of the control valve with TIC-4018 and return temperatures to pre-transient values. (BOP-C) (SRO-C)
ACA, CT CT Justification:
Safety significance - Prevent an unnecessary reactor trip.
Initiating Cue - Generator Hydrogen temperature rising and Annunciator K04-B6 Measurable Performance Standard - Manually control Hydrogen temperature with TIC-4018 and prevent the need for a Turbine / Reactor Trip Performance Feedback - Hydrogen temperatures and ACW flow returns to near pre-transient values Event Eight: Main Steam Line break inside containment Once H2 temperature is recovering, the major event will begin. A main steam line break inside containment will result in RB pressure rising to the ESAS actuation setpoint of Channels 1-6.
After the immediate actions of Reactor Trip are completed, the SRO should identify the need to transition to the Overcooling EOP. (ATC-C) (BOP-C) (SRO-C) EOP Event Nine: ESAS Channels 5 & 6 fail to actuate automatically A main steam line break inside containment will result in RB pressure rising to the ESAS actuation setpoint of Channels 1-6. However, Channels 5 & 6 are failed and will not automatically actuate. The ATC will identify the failure and manually actuate Channels 5 & 6. If the ATC does not recognize the failure, then the BOP will identify the failure during the performance of RT-10 which verifies actuation of ESAS. The critical task of actuating Channels 5 & 6 must be performed prior to the completion of RT-10. (ATC-C) (BOP-C) (SRO-C) EOP, CT CT Justification: CT-19 Safety significance - Isolate possible RCS leak paths and assure containment integrity.
Initiating Cue - RB Pressure HI (2.175 psig) alarm on K11 and Reactor Building pressure >4 psig (ESAS setpoint).
Measurable Performance Standard - Manually actuate Channels 5 & 6 prior to reporting the completion of RT-10.
Performance Feedback - Annunciators for RB Isolation Channel 5 & 6 on K11 and components reposition to ESAS position for Channels 5 & 6.
Appendix D Scenario Outline Page 5 of 6 Event Ten: EFIC Vector Isolation failure for one EFW flow path to faulted generator A malfunction of the Vector Isolation signal to the faulted generator will result in continuing to feed the bad steam generator. The ATC should identify this failure and manually isolate the flowpath by closing CV-2670 or CV-2646 or both in accordance with RT-6. This will stop the feed source of overcooling. Once the bad steam generator depressurizes the overcooling will be terminated and the crew will stabilize RCS temperature.
CT Justification: CT-16 Safety significance - Excessive primary to secondary heat transfer due to overfeeding faulted steam generator.
Initiating Cue - RCS overcooling in progress, lower than normal steam generator pressure, skewed EFW flow to faulted steam generator.
Measurable Performance Standard - Manually isolate all FW flow (MFW and EFW) to the faulted steam generator prior to steam generator reaching 400 inches (Carry-over into main steam lines)
Performance Feedback - EFW flowrate near zero, faulted steam generator pressure trending towards zero, RCS temperature stable or rising.
The scenario can be terminated at the discretion of the lead examiner.
Appendix D Scenario Outline Page 6 of 6 Simulator Instructions for Scenario 2 Reset simulator to MOL ~100% power IC steady state Ensure malfunctions agree with scenario guide (either schedule file or IC)
Event No.
Time Event Type Description 1
T=0 N-(ATC)
N-(SRO)
Align for 2 minute delithiation 2
T=5 C-(SRO)
TS Pressurizer Heater Ground Fault 3
T=20 C-(ATC)
C-(BOP)
C-(SRO)
AOP Degrading Vacuum 4
T=25 C-(ATC)
C-(BOP)
C-(SRO)
ACA P-33C ICW Pump trips with an auto start failure on P-33B ICW Pump.
5 T=35 C-(SRO)
TS Inverter Y11 failure 6
T=40 I-(ATC)
I-(BOP)
I-(SRO)
AOP Main Steam Header Pressure bias failure 7
T=45 C-(BOP)
C-(SRO)
ACA, CT Generator H2 Temp Controller Setpoint (TIC-4018) failure 8
T=50 M-(ALL)
A Main Steam Line break inside containment 9
T=52 I-(ATC)
I-(BOP)
I-(SRO)
CT ESAS Channels 5 & 6 fail to actuate automatically 10 55 C-(ATC)
CT EFIC Vector Isolation failure for one EFW flow path to failed generator
Appendix D Scenario Outline Page 1 of 5 Facility: ANO-1 Scenario No.: 5 R0 Op-Test No.: 2020-1 Examiners: ____________________________Operators: _____________________________
Initial Conditions:
- IC205 5% Power.
- P-4B powered from A3
- Run Scenario 5 Schedule File R0 Turnover:
Event No.
Malf. No.
Event Type*
Event Description 1
N/A N-(ATC)
N-(BOP)
N-(SRO)
Place A MFWP in service and secure P-75 Aux FW Pump 2
N/A N-(BOP)
N-(SRO)
N/A R-(ATC)
R-(SRO)
Raise power to 10%
4 RD271 C-(ATC)
C-(BOP)
C-(SRO)
TS Group 7 Rods do not sequence on resulting in no overlap between Groups 6 & 7 5
ED181 C-(SRO)
TS, ACA S/U #2 failure 6
CV059 C-(ATC)
C-(BOP)
C-(SRO)
TS, AOP L/D Cooler Leak (75 gpm) 7 ED183 M-(ALL)
Loss of Offsite Power 8
DG175 SW135 C-(BOP)
C-(SRO)
- 1 EDG fails to auto start P-4C fails to auto start (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Page 2 of 5 Target Quantitative Attributes (Section D.5.d)
Actual Attributes Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 2 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions ( 1per scenario set) 1 Critical Tasks (2) 2 SCENARIO 5 OBJECTIVES
- 1)
Evaluate individual ability to perform placing Main Feedwater Pump in service and securing Aux Feedwater Pump.
- 2)
Evaluate individual ability to perform Anticipatory Reactor Trip System (ARTS) Reset.
- 3)
Evaluate individual ability to perform Power Escalation.
- 4)
Evaluate individual ability to recognize and respond to failure to have proper overlap between Group 6 and Group 7 control rods.
- 5)
Evaluate individual ability to recognize and respond to Electrical Distribution Annunciators on K-02.
- 6)
Evaluate individual ability to perform an RCS Leakage Investigation.
- 7)
Evaluate individual ability to recognize and respond to Excess RCS Leakage.
- 8)
Evaluate individual ability to recognize and respond to Degraded Power.
- 9)
Evaluate individual ability to recognize and respond to EDG failure to start.
- 10) Evaluate individual ability to perform DG1 Start from Control Room.
Appendix D Scenario Outline Page 3 of 5 SCENARIO 5 NARRATIVE Event One: Place A MFWP in service and secure P-75 Aux FW Pump The SRO will direct the ATC to place the A MFW pump in service as directed in 1102.002 Plant Startup Procedure. When the A MFW pump is supplying feedwater to the steam generators, the SRO will direct the BOP to secure P-75 the Aux. FW pump.
(ATC-N) (BOP-N) (SRO-N)
Event Two: Reset ARTS The SRO will then direct that the Anticipatory Reactor Trips (ARTS) reset in RPS for the A MFW pump. Failure to perform the ARTS reset will result in a reactor trip when reactor power exceeds 7% power. (CT) ARTS must be reset prior exceeding 7% NI Power.
(BOP-N) (SRO-N) CT CT Justification:
Safety significance - Prevent unnecessary reactor trip Initiating Cue - A MFW Pump in service feeding steam generators.
Measurable Performance Standard - Reactor does not trip on Loss of Feedwater when power is raised above 7%.
Performance Feedback - Light indication on the A MFWP contact buffer will have the top red light ON, bottom red light OFF for the A RPS Channel. Due to space restraints Channels B-D are not mimicked like the plant, for the other three channels the white TRIPPED light for the A MFWP goes dim on the contact monitors when ARTS is reset. Inform the applicant that another operator will reset the channels. A trigger is set up to reset the additional channels.
Event Three: Raise power to 10%
When ARTS is reset, the SRO will then direct that plant power will be raised to ~10% power.
(ATC-R) (SRO-R)
Event Four: Group 7 Rods do not sequence on resulting in no overlap between Groups 6 & 7 As reactor power is raised, a failure of the group 7 Control Rod Sequencing circuit will result in group 7 Control Rods not withdrawing when group 6 Control Rods are between 75% and 85%.
This should be recognized by the ATC and/or BOP and the SRO should be notified. The SRO should direct that rod withdrawal be stopped and then reference Tech Specs.
(ATC-C) (BOP-C) (SRO-C)TS Event Five: S/U #2 failure After the TS entry has been declared to the crew, K02-B3, SU2 L.O. Relay Trip, will annunciate.
The BOP will perform actions from the ACA while the SRO assesses TS.
C-(SRO) TS, ACA Event Six: L/D Cooler Leak After the TS entry has been declared to the crew, a 75 gpm Letdown Cooler leak will develop.
This will be identified by a rising ICW Temperature and alarm on K12-E4. It is possible that K10-A1 Letdown Temperature Hi will also alarm. With the given alarms the crew will diagnose a Letdown cooler leak into the Nuclear ICW System. The SRO will direct the crew from OP-1203.039, Excess RCS Leakage.
(ATC-C) (BOP-C) (SRO-C) TS, AOP
Appendix D Scenario Outline Page 4 of 5 Event Seven: Loss of Offsite Power Once the leaking Letdown Cooler is isolated, a Loss of Offsite Power will occur and the SRO will direct actions from OP-1202.007, Degraded Power. If asked, Unit 2 will report that the Alternate AC Diesel Generator is not available due to governor control issues and maintenance has been contacted for investigation.
(ATC-C) (BOP-C) (SRO-C) EOP Event Eight: #1 EDG fails to start During verification of RT-21, Check EDG Operation, the BOP will identify that the #1 EDG failed to start automatically and he will start the associated EDG from C10. In addition to the failure of the #1 EDG, the Service Water Pump, P-4C which provides cooling to the #2 EDG will fail to start following restoration of power to A4. If P-4C is not started the #2 EDG will trip in ~10 minutes. The combination of the two failures makes either starting #1 EDG or starting P-4C a critical task to ensure a 4160 Volt power supply remains available to a vital bus.
(BOP-C) (SRO-C) CT CT Justification: CT-8 Safety significance -Prevent degradation of the mitigative capability of the plant Initiating Cue - Loss of Offsite Power, A3 de-energized, normal lighting in the control room off.
Measureable Performance Standard - BOP will start the #1 EDG within 15 minutes of Blackout.
EAL escalation is required if the EDG is not started within 15 minutes.
Performance Feedback - Change of status lamps on control console for #1 EDG, components on C18 and normal lighting returns for the control room.
OR CT Justification: CT-8 Safety significance -Prevent degradation of the mitigative capability of the plant Initiating Cue - Loss of Offsite Power, low Service Water Pressure, and eventually a Critical Trouble Alarm K01-C2.
Measureable Performance Standard - BOP will start P-4C before the #1 EDG trips on high temperature.
Performance Feedback - Change of status lamps on control console for P-4C Service Water Pump and Service Water Pressure rise.
The scenario can be terminated at the discretion of the lead examiner.
Appendix D Scenario Outline Page 5 of 5 Simulator Instructions for Scenario 5 Reset simulator to BOL ~2-5% power IC steady state Ensure malfunctions agree with scenario guide (either schedule file or IC)
Event No.
Time Event Type Description 1
T=0 N-(ATC)
N-(BOP)
N-(SRO)
Place A MFWP in service and secure P-75 Aux FW Pump 2
T=15 N-(BOP)
N-(SRO)
T=20 R-(ATC)
R-(SRO)
Raise power to 10%
4 T=25 C-(ATC)
C-(BOP)
C-(SRO)
TS Group 7 Rods do not sequence on resulting in no overlap between Groups 6 & 7 5
T=35 C-(BOP)
C-(SRO)
TS, ACA S/U #2 failure 6
T=45 C-(ATC)
C-(BOP)
C-(SRO)
TS, AOP L/D Cooler Leak (75 gpm) 7 T=55 M-(ALL)
Loss of Offsite Power 8
T=55 C-(BOP)
C-(SRO)
- 1 EDG fails to auto start P-4C fails to auto start