ML20233A925

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AN1-2020-04 Final Outlines
ML20233A925
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 06/05/2020
From: Greg Werner
Operations Branch IV
To:
Entergy Operations
References
ES-401-2
Download: ML20233A925 (39)
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Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: ANO U1 Date of Exam: 4/8/2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total 3 3 3 3 3 3 18 3 3 6

1. 1 Emergency and 2 1 1 1 2 2 9 2 2 4 Abnormal Plant 2 Evolutions 5 4 4 4 5 5 27 5 5 10 Tier Totals 3 2 3 3 2 2 3 2 3 2 3 28 2 3 5 1
2. 1 0 1 1 1 1 1 1 1 1 1 10 1 1 1 3 Plant 2 Systems 4 2 4 4 3 3 4 3 4 3 4 38 4 4 8 Tier Totals
3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 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, Page 40 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 IR #

Topic(s) 000007 (EPE 7; BW E02&E10; CE E02) R EK2.2 Knowledge of the interrelations between the 4.2 1 Reactor Trip, Stabilization, Recovery / 1 Vital System Status Verification and the following: Bank facilities heat removal system, including primary 415 coolant, emergency coolant, decay heat removal systems, and relations between the proper operations of these systems to operation of the facility 000008 (APE 8) Pressurizer Vapor Space R AK2.02 Knowledge of the interrelations between 2.7* 2 Accident / 3 the PZR Vapor space accident and the following: Bank Sensors and detectors 371 000009 (EPE 9) Small Break LOCA / 3 S EA2.06 Ability to determine or interpret the 4.3 76 following as they apply to a small break LOCA: New Whether PZR water inventory loss is imminent 1243 000011 (EPE 11) Large Break LOCA / 3 R EK3.04 Knowledge of the reasons for the following 4.0 3 responses as the apply to the Large Break LOCA: New Placing containment fan cooler damper in accident 1298 position 000015 (APE 15) Reactor Coolant Pump R AK1.03 Knowledge of the operational implications 3.0 4 Malfunctions / 4 of the following concepts as they apply to Reactor New Coolant Pump Malfunctions (Loss of RC Flow): The 1269 basis for operating at a reduced power level when one RCP is out of service 000022 (APE 22) Loss of Reactor Coolant R AK1.01 Knowledge of the operational implications Makeup / 2 of the following concepts as they apply to Loss of 2.8 5 Reactor Coolant Makeup: Consequences of Bank thermal shock to RCP seals. 183 000025 (APE 25) Loss of Residual Heat S AA2.05 Ability to determine and interpret the 3.5 77 Removal System / 4 following as they apply to the Loss of Residual Heat Mod Removal System: Limitations on LPI flow and 1244 temperature rates of change 000026 (APE 26) Loss of Component R AA1.02 Ability to operate and / or monitor the 3.2 6 Cooling Water / 8 following as they apply to the Loss of Component New Cooling Water: Loads on the CCWS in the control 1313 room.

000027 (APE 27) Pressurizer Pressure S AA2.11 Ability to determine and interpret the 4.1 78 Control System Malfunction / 3 following as they apply to the Pressurizer Pressure New Control Malfunctions: RCS Pressure 1256 000029 (EPE 29) Anticipated Transient R EK1.01 Knowledge of the operational implications 2.8 7 Without Scram / 1 of the following concepts as they apply to the Bank ATWS: Reactor nucleonics and thermo-hydraulics 887 behavior.

000038 (EPE 38) Steam Generator Tube R EA2.13 Ability to determine or interpret the 3.1 8 Rupture / 3 following as they apply to a SGTR: magnitude of Bank rupture 856 4.6 79 S 2.4.41 Knowledge of the emergency action level New thresholds and classifications. 1301 000040 (APE 40; BW E05; CE E05; W E12) R AK3.02 Knowledge of the reasons for the following 4.4 9 Steam Line RuptureExcessive Heat responses as they apply to the Steam Line New Transfer / 4 Rupture: ESFAS initiation 1271 S 2.4.6 Knowledge of EOP mitigating strategies 4.7 80 New 1246 000054 (APE 54; CE E06) Loss of Main R AA1.03 Ability to operate and / or monitor the 3.5 10 Feedwater /4 following as they apply to the Loss of Main New Feedwater (MFW): AFW auxiliaries, including oil 1272 cooling water supply ES-401, Page 41 of 52

000055 (EPE 55) Station Blackout / 6 R EA2.03 Ability to determine or interpret the 3.9 11 following as they apply to a Station Blackout: Bank Actions necessary to restore power 1097 000056 (APE 56) Loss of Offsite Power / 6 R 2.1.31 Ability to locate control room switches, 4.6 12 controls, and indications, and to determine that they New correctly reflect the desired lineup. 1273 S 2.4.20 Knowledge of the operational implications of 4.3 81 EOP warnings, cautions, and notes. New 1258 000057 (APE 57) Loss of Vital AC R AA1.01 Ability to operate and / or monitor the 3.7 13 Instrument Bus / 6 following as they apply to the Loss of Vital AC New Instrument Bus: Manual Inverter swapping 1314 000058 (APE 58) Loss of DC Power / 6 R AA2.01 Ability to determine and interpret the 3.7 14 following as they apply to the Loss of DC Power: New That a loss of dc power has occurred; verification 1274 that substitute power sources have come on line 000062 (APE 62) Loss of Nuclear Service R 2.1.23 Ability to perform specific system and 4.3 15 Water / 4 integrated plant procedures during all modes of Bank plant operation. 890 000065 (APE 65) Loss of Instrument Air / 8 R 2.1.7 Ability to evaluate plant performance and 4.4 16 make operational judgments based on operating Bank characteristics, reactor behavior, and 1102 000077 (APE 77) Generator Voltage and R AK3.01 Knowledge of the reasons for the following 3.9 17 Electric Grid Disturbances / 6 responses as they apply to Generator Voltage and Mod Electric Grid Disturbances: Reactor and turbine trip 1308 criteria (W E04) LOCA Outside Containment / 3 N/A for this design type (W E11) Loss of Emergency Coolant N/A for this design type Recirculation / 4 (BW E04; W E05) Inadequate Heat R EK2.1 Knowledge of the interrelations between the 3.8 18 TransferLoss of Secondary Heat Sink / 4 Inadequate Heat Transfer and the following: New Components, and functions of control and safety 1315 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features K/A Category Totals: 3 3 3 3 3/3 3/3 Group Point Total: 18/6 ES-401, Page 42 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 S 2.2.38 Knowledge of conditions 4.5 82 and limitations in the facility New license. 1245 000024 (APE 24) Emergency Boration / 1 Not sampled 000028 (APE 28) Pressurizer (PZR) Level Control Not sampled Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Not sampled Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear R AK1.01 Knowledge of the 2.7 19 Instrumentation / 7 operational implications of the Bank following concepts as they apply 425 to Loss of Intermediate Range Nuclear instrumentation: Effects of voltage changes on 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 Notf sampled 000037 (APE 37) Steam Generator Tube Leak / 3 S 2.2.40 Ability to apply Technical 4.7 84 (Selected instead of APE78 below) Specifications for a system. New 1260 000051 (APE 51) Loss of Condenser Vacuum / 4 R 2.1.20 Ability to interpret and 4.6 20 execute procedure steps. Bank 1062 000059 (APE 59) Accidental Liquid Radwaste Release / 9 R AK2.01 Knowledge of the 2.7 21 interrelations between the Repeat Accidental Liquid Radwaste 951 Release and the following:

Radioactive-liquid monitors 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 R AA1.02 Ability to operate and / or 2.9 22 monitor the following as they Bank apply to the Accidental gaseous 272 Radwaste: Ventilation system 000061 (APE 61) Area Radiation Monitoring System Alarms R AA2.06 Ability to determine and 3.2 23

/7 interpret the following as they New apply to the Area Radiation 1276 Monitoring (ARM) System Alarms: Required actions if alarm channel is out of service 000067 (APE 67) Plant Fire On Site / 8 S AA2.17 Ability to determine and 4.3 83 interpret the following as they Bank apply to plant fire on site: 1027 systems that may be affected by the fire.

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 3.9 24 computers to evaluate system or New component status. 1277 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling / R EK1.03 Knowledge of the 4.5 25 4 operational implications of the Bank following concepts as they apply 289 to the Inadequate Core Cooling:

Processes for removing decay heat from the core 000076 (APE 76) High Reactor Coolant Activity / 9 Not sampled 000078 (APE 78*) RCS Leak / 3 N/A for this rev of KA catalogs (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 ES-401, Page 43 of 52

(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 R AA2.1 Ability to determine and 3.0 26 interpret the following as they New apply to the (Plant Runback) 1278 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(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 EA2.1 Ability to determine and 85 S interpret the following as they 4.0 New apply to the (Inadequate 1251 Subcooling Margin) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(BW E08; W E03) LOCA CooldownDepressurization / 4 R EK3.3 Knowledge of the reasons 4.0 27 for the following responses as New they apply to the (LOCA 1302 Cooldown): Manipulation of controls required to obtain desired operating results during abnormal, and emergency (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 N/A for this design type Shock / 4 (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 2/2 Group Point Total: 9/4 ES-401, Page 44 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 IR #

Topic(s) 003 (SF4P RCP) Reactor Coolant R K6.04 Knowledge of the effect of a loss or 2.8 28 Pump malfunction on the following will have on the Repeat RCPS: Containment isolation valves affecting 53 RCP operation 004 (SF1; SF2 CVCS) Chemical and R K4.15 Knowledge of CVCS design feature(s) 3.0 29 Volume Control and/or interlock(s) which provide for the New following: Interlocks associated with operation 1279 of orifice isolation valves 005 (SF4P RHR) Residual Heat R K1.04 Knowledge of the physical connections 2.9 30 Removal and/or cause-effect relationships between the New RHRS and the following systems: CVCS 1280 R K4.02 Knowledge of RHRS design feature(s) 3.2 31 and/or interlock(s) which provide or the New following: Modes of Operation 1281 006 (SF2; SF3 ECCS) Emergency R K3.01 Knowledge of the effect that a loss or 4.1 32 Core Cooling malfunction of the ECCS will have on the Bank following: RCS 197 R A1.18 Ability to predict and/or monitor 4.0 33 changes in parameters (to prevent exceeding New design limits) associated with operating the 1282 ECCS controls including: PZR level and pressure 007 (SF5 PRTS) Pressurizer R A1.02 Ability to predict and/or monitor Relief/Quench Tank changes in parameters (to prevent exceeding 2.7 34 design limits) associated with operating the New PRTS controls including: Monitoring quench 1283 tank pressure A2.03 Ability to (a) predict the impacts of the R following malfunctions or operations on the P 3.6 35 S; and (b) based on those predictions, use Mod procedures to correct, control, or mitigate the 1300 consequences of those malfunctions or operations: Overpressurization of the PZR 008 (SF8 CCW) Component Cooling R K2.02 Knowledge of bus power supplies to 3.0 36 Water the following: CCW pump, including New emergency backup 1253 010 (SF3 PZR PCS) Pressurizer R K4.01 Knowledge of PZR PCS design 2.7 37 Pressure Control feature(s) and/or interlock(s) which provide New for the following: Spray valve warm-up 1284 010 (SF3 PZR PCS) Pressurizer A2.01 Ability to (a) predict the impacts of the 3.6 86 Pressure Control S following malfunctions or operations on the New PZR PCS; and (b) based on those 1261 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater 012 (SF7 RPS) Reactor Protection R F il Knowledge of the effect that a loss or K3.02 3.2 38 malfunction of the RPS will have on the New following: T/G 1285 R A3.05 Ability to monitor automatic operation 3.6 39 of the RPS, including: Single and multiple Bank channel trip indicators 1093 ES-401, Page 45 of 52

013 (SF2 ESFAS) Engineered R A4.03 Ability to manually operate and/or 4.5 40 Safety Features Actuation monitor in the control room: ESFAS initiation New 1286 S 2.2.37 Ability to determine operability and/or 4.6 87 availability of safety related equipment Bank 1052 022 (SF5 CCS) Containment Cooling R 2.2.42 Ability to recognize system parameters 3.9 41 that are entry-level conditions for Technical New Specifications. 1287 025 (SF5 ICE) Ice Condenser N/A 026 (SF5 CSS) Containment Spray R A3.01 Ability to monitor automatic operation 4.3 42 of the CSS, including: Pump starts and Bank correct MOV positioning 223 S 2.4.21 Knowledge of the parameters and 4.6 89 logic used to assess the status of safety New functions, such as reactivity control, core 1311 cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

039 (SF4S MSS) Main and Reheat R K5.05 Knowledge of the operational 2.7 43 Steam implications of the following concepts as the Bank apply to the MRSS: Bases for RCS cooldown 910 limits S 2.2.25 Knowledge of the bases in Technical 4.2 88 Specifications for limiting conditions for New operations and safety limits 1248 059 (SF4S MFW) Main Feedwater R K3.02 Knowledge of the effect that a loss or 3.6 44 malfunction of the MFW will have on the Bank following: AFW system 912 R 2.1.25 Ability to interpret reference materials, 3.9 45 such as graphs, curves, tables, etc. New 1288 061 (SF4S AFW) R K6.02 Knowledge of the effect of a loss or 2.6 46 Auxiliary/Emergency Feedwater malfunction of the following will have on the Mod AFW components: Pumps 1252 R 2.4.21 Knowledge of the parameters and 4.0 47 logic used to assess the status of safety Bank functions, such as reactivity control, core 668 cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

062 (SF6 ED AC) AC Electrical R K1.04 Knowledge of the physical connections 3.7 48 Distribution and/or cause effect relationships between the Bank ac distribution system and the following 790 systems: off-site power sources R K2.01 Knowledge of bus power supplies to 3.3 49 the following: Major system loads New 1289 063 (SF6 ED DC) DC Electrical R K1.03 Knowledge of the physical connections 2.9 50 Distribution and/or cause-effect relationships between the Mod DC electrical system and the following 670 systems: Battery charger and battery ES-401, Page 46 of 52

064 (SF6 EDG) Emergency Diesel R A2.06 Ability to (a) predict the impacts of the 2.9 51 Generator following malfunctions or operations on the New ED/G system; and (b) based on those 1299 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Operating unloaded, lightly loaded, and highly loaded time limit.

073 (SF7 PRM) Process Radiation R K5.01 Knowledge of the operational 2.5 52 Monitoring implications as they apply to concepts as they Mod apply to the PRM system: Radiation theory, 215 including sources, types, units, and effects 076 (SF4S SW) Service Water R A4.01 Ability to manually operate and/or 2.9 53 monitor in the control room: SWS pumps New 1268 078 (SF8 IAS) Instrument Air R A3.01 Ability to monitor automatic operation 3.1 54 of the IAS, including: Air pressure Repeat 673 103 (SF5 CNT) Containment R A1.01 Ability to predict and/or monitor 3.7 55 changes in parameters (to prevent exceeding New design limits) associated with operating the 1309 containment system controls including:

Containment pressure, temperature, and humidity S A2.03 Ability to (a) predict the impacts of the 3.8 90 following malfunctions or operations on the Mod containment system and (b) based on those 980 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Phase A and B isolation.

053 (SF1; SF4P ICS*) Integrated Cant sample until rev 3 of KA catalogs Control K/A Category Point Totals: 3 2 3 3 2 2 3 2/2 3 2 3/3 Group Point Total: 28/5 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 3.4 56 implications as they apply to the CRDS: Effects New on power of inserting axial shaping rods 1290 002 (SF2; SF4P RCS) Reactor 2.4.8 Knowledge of how abnormal operating 4.5 91 Coolant S procedures are used in conjunction with EOPs. New 1263 011 (SF2 PZR LCS) Pressurizer S A2.03 Ability to (a) predict the impacts of the 3.9 92 Level Control following malfunctions or operations on the PZR New LCS and (b) based on those predictions, use 1249 procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of PZR level 014 (SF1 RPI) Rod Position R 2.4.6 Knowledge of EOP mitigation strategies. 3.7 57 Indication New 1291 015 (SF7 NI) Nuclear Not sampled Instrumentation 016 (SF7 NNI) Nonnuclear Not sampled Instrumentation 017 (SF7 ITM) In-Core Temperature R A4.02 Ability to manually operate and/or 3.8 58 Monitor monitor in the control room: Temperature New values used to determine RCS/RCP operation 1292 during inadequate core cooling (i.e., if applicable, average of five highest values) 027 (SF5 CIRS) Containment Iodine R K1.01 Knowledge of the physical connections 3.4 59 Removal and/or cause-effect relationships between the Bank CIRS and the following systems: CSS 209 028 (SF5 HRPS) Hydrogen Not sampled Recombiner and Purge Control 029 (SF8 CPS) Containment Purge Not sampled 033 (SF8 SFPCS) Spent Fuel Pool Not sampled Cooling 034 (SF8 FHS) Fuel-Handling S K1.02 Knowledge of the physical connections 3.2 93 Equipment and/or cause effect relationships between the New Fuel Handling System and the following 1264 systems: RHRS 035 (SF 4P SG) Steam Generator R K6.02 Knowledge of the effect of a loss or 3.1 60 malfunction on the following will have on the New S/GS: secondary porv 1255 041 (SF4S SDS) Steam Not sampled Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine R K3.01 Knowledge of the effect that a loss or 2.9 61 Generator malfunction of the MT/G system will have on Bank the following: Remainder of the Plant 533 055 (SF4S CARS) Condenser Air Not sampled Removal 056 (SF4S CDS) Condensate R A2.04 Ability to (a) predict the impacts of the 2.6 62 following malfunctions or operations on the Mod Condensate System; and (b) based on those 765 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: loss of Condensate pumps 068 (SF9 LRS) Liquid Radwaste Not sampled ES-401, Page 48 of 52

071 (SF9 WGS) Waste Gas R K4.01 Knowledge of design feature(s) and/or 2.6 63 Disposal interlock(s) which provide for the following: New Pressure capability of the waste gas decay 1305 tank.

072 (SF7 ARM) Area Radiation R A3.01 Ability to monitor automatic operation of 2.9 64 Monitoring the ARM system, including: Changes in Bank ventilation alignment 153 075 (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 2.7 65 in parameters to prevent exceeding design New limits associated with operating the Fire 1304 Protection System controls, including: Fire doors 050 (SF 9 CRV*) Control Room N/A until rev3 of KA catalogs Ventilation K/A Category Point Totals: 1/1 0 1 1 1 1 1 1/1 1 1 1/1 Group Point Total: 10/3 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 #

2.1.34 Knowledge of primary and secondary plant chemistry limits. 2.7 66 New 1294

1. Conduct of 2.1.32 Ability to explain and apply system limits and precautions. 3.8 67 Operations Bank 651 2.1.21 Ability to verify the controlled procedure copy 3.5 68 Bank 389 2.1.4 Knowledge of individual licensed operator responsibilities related to 3.8 94 shift staffing, such as medical requirements, no-solo operation, Bank maintenance of active license status, 10CFR55, etc. 407 2.1.36 Knowledge of procedures and limitations involved in core alterations 4.1 95 New 1250 Subtotal 3 2 2.2.13 Knowledge of tagging and clearance procedures. 4.1 69
2. Equipment Bank Control 231 2.2.43 Knowledge of the process used to track inoperable alarms 3.0 70 New 1296 2.2.6 Knowledge of the process for making changes to procedures 3.6 96 Bank 409 2.2.12 Knowledge of surveillance procedures 4.1 97 New 1265 Subtotal 2 2 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation 2.9 71 monitors and alarms, portable survey instruments, personnel New monitoring equipment, etc. 1310
3. Radiation 2.3.7 Ability to comply with radiation work permit requirements during 3.5 72 Control normal or abnormal conditions. Bank 996 2.3.13 Knowledge of radiological safety procedures pertaining to licensed 3.8 98 operator duties, such as response to radiation monitor alarms, New containment entry requirements, fuel handling responsibilities, 1266 access to locked high radiation areas, aligning filters, etc.

Subtotal 2 1 ES-401, Page 50 of 52

2.4.6 Knowledge of EOP mitigation strategies. 3.7 73 Bank 803 2.4.25 Knowledge of fire protection procedures. 3.3 74 Bank 848 2.4.3 Ability to identify post-accident instrumentation. 3.7 75 Bank 242

4. Emergency 2.4.14 Knowledge of general guidelines for EOP usage. 4.5 99 Procedures/Plan New 1267 2.4.30 Knowledge of events related to system operation/status that must 4.1 100 be reported to internal organizations or external agencies, such as New the State, the NRC, or the transmission system operator 1312 Subtotal 3 2 Tier 3 Point Total 10 7 ES-401, Page 51 of 52

ES-401 Record of Rejected K/As Form ES-401-4 Tier/Group Randomly Reason for Rejection Selected K/A 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 RO (Q#5) several reasons. RCS makeup flow is controlled with a makeup valve that controls T1/G1 022 pressurizer level. By design, the shutoff head of the makeup pumps is so high that any 022 AK1.01 change in RCS pressure wont be enough to affect makeup flow in the long term. Also, AK1.02 (2.8) the Loss of Makeup AOP only addresses leaks within the makeup system or a trip of a (2.7) 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 026 - Loss of Component Cooling Water. The original K/A concerns the ability to (Q#6) operate and / or monitor CCWS surge tank parameters from the control room. ANO T1/G1 026 Unit 1 has no level indication, pressure indication, or surge tank radiation indication.

026 AA1.02 Selected AA1.02 which concerns operating and / or monitoring the loads on the CCWS AA1.05 (3.2) system in the control room.

(3.1) 029 - Anticipated Transient Without Scram. The original K/A concerns the operational RO implications of the concept of the definition of reactivity as it applies to an ATWS. A (Q#7) site specific definition of reactivity does not exist in the AOPs/EOPs or AOP/EOP users T1/G1 029 guide therefore does not belong as a T1 question. With no site specific definition a 029 EK1.02 question on the definition of reactivity would be repeating GFES specific topics.

EK1.01 (2.8) Randomly selected EK1.02 which concerns the behavior of reactor nucleonics and (2.6) thermo-hydraulics during an ATWS.

A01 - Plant Runback. The original K/A concerns the adherence to appropriate RO procedures and operation within the limitations in the facilities license and amendments (Q#26)

T1/G2 as applicable to a Plant Runback. There is no direct relationship to a Plant Runback BW A01 BW A01 condition and limitations within the facility license. Randomly selected AA2.1 which AA2.1 AA2.2 concerns the facility conditions and selection of appropriate procedures during (3.0)

(3.5) abnormal and emergency conditions.

RO 007 - Pressurizer Relief / Quench Tank. The original K/A was the ability to predict or (Q#34) monitor quench tank temperature to prevent exceeding design limits. While there is T2/G1 007 indication of quench tank temperature, there is no procedural limitation on how high or 007 A1.02 low it should be. Randomly selected A1.02 which concerns quench tank pressure A1.03 (2.7) which has procedural limitations.

(2.6) 064 - Emergency Diesel Generator. The original K/A was to predict the consequences of high VARs on ED/G integrity and to use procedures to correct, control, or mitigate the consequences. There is no specific procedural guidance or caution/warning of RO (Q#51) high VARs impacting the integrity of an EDG. The only guidance in the normal T2/G1 operating procedure is to maintain VARs low. Further investigation into training 064 064 material and GFES material was conducted and no discussion of the effect of high A2.06 A2.19 (2.9) VARs on an operating EDG were found. Since there is no clear discussion anywhere (2.5) of high VARs impacting EDG integrity, randomly selected A2.06 which is to predict the impacts of operating unloaded, lightly loaded, and highly loaded, and to use procedures to correct, control, or mitigate the consequences of those operations.

RO 073 - Process Radiation Monitoring System. The original K/A concerns the knowledge (Q#52) of radiation intensity changes with source distance which isnt really applicable to T2/G1 073 process radiation monitors installed at ANO1. Changed K/A to K5.01 which concerns 073 K5.01 radiation theory, including sources, types, units, and effects.

K5.02 (2.5)

(2.5)

RO 027 - Containment Iodine Removal. The original K/A is a typo using APE 027 not SYS (Q#59) 027 for Tier 2. Selected the only K1 K/A for SYS 027 which is K1.01.

T2/G2 027 027 K1.01 AK1.02 (3.4)

(2.8) 071 - Waste Gas Disposal System (WGDS). The original K/A concerned the design RO feature(s) and/or interlock(s) which provide isolation of waste gas release tanks. This (Q#63)

T2/G2 action is performed by RE-4830 which is already asked in Question 22 under KA 071 071 APE060 AA1.02. Randomly selected K4.01 which concerns the knowledge of design K4.01 K4.04 feature(s) and/or interlock(s) which provide for the following: Pressure capability of the (2.6)

(2.9) waste gas decay tank.

RO The original K/A concerns the knowledge of refueling administrative limits. This is a (Q#67) SRO topic and shouldnt be asked on the RO portion of the exam. Randomly selected T3 2.1.32 2.1.32 which concerns the application of system limits and precautions 2.1.40 (3.8)

(2.8)

ES-401, Page 52 of 52

The original K/A concerns the process for managing troubleshooting activities. IAW RO EN-MA-125, the management and approvals of troubleshooting come from the SM and (Q#70)

T3 or FIN SRO. The RO position is not involved in any portion of these types of activities 2.2.43 2.2.20 therefore this question shouldnt be asked on the RO portion of the exam. Randomly (3.0)

(2.6) selected 2.2.43 which concerns the process of tracking inoperable alarms.

The original K/A concerns specific mitigation strategies for a low power / shutdown RO accident. This topic belongs in T1, not T3. Randomly selected 2.4.25 which concerns (Q#74)

T3 the knowledge of fire protection procedures.

2.4.25 2.4.9 (3.3)

(3.8)

RO The original K/A concerns the ability to verify alarms are consistent with plant (Q#75) conditions. This topic belongs in T2, not T3. Randomly selected 2.4.3 which concerns T3 2.4.3 the ability to identify post-accident instrumentation.

2.4.46 (3.7)

(4.2) 027 - Pressurizer Pressure Control System Malfunction. The original K/A concerned SRO interpreting RCP injection flow given a PZR control system malfunction. This is not (Q#78)

T1/G1 applicable on Unit 1 as RCP injection flow has its own flow controller which will 027 027 maintain flow based on its own set point. Also the 1203.015 doesnt specifically AA2.11 AA2.14 discuss any effects on RCP seal injection flow for any type of control system failure.

(4.1)

(2.9) Randomly selected AA2.11 which concerns the effect on RCS pressure.

SRO 056 - Loss of Offsite Power. Since this is a T1 K/A, generic K/As must be selected (Q#81) from the list provided in D.1.b of ES-401. The original KA was not found in this list.

T1/G1 056 Utilized 2.4.20 which was originally in the T3 section at Q#100, but was rejected since 056 2.4.20 it is a T1 K/A not a T3 K/A.

2.4.42 (4.3)

(3.8) 067 - Plant Fire On Site. The original K/A concerned the need for an emergency plant SRO shutdown based on a plant fire. 1203.49 is written generically to first allow assessment T1/G2 (Q#83) from the operators and then later determine if a reactor trip is required. There is no 067 067 specific trip criteria based on a fire alone, so any conditions given in a question that AA2.13 AA2.17 would require a reactor trip would already have guidance in another EOP or AOP.

(4.4) (4.3) Randomly selected AA2.17 which concerns systems that may be affected by an onsite fire.

SRO 078 - RCS Leak. This APE is not in Rev 2 K/A catalog but will be released in Rev 3.

(Q#84) Randomly selected APE37 Steam Generator Tube Leak instead.

T1/G2 037 078 2.2.40 2.2.40 (4.7)

(4.7)

SRO 012 - Reactor Protection. The original K/A was to determine the operability or (Q#87) availability of the RPS system. After discussion with the NRC lead examiner, selecting T2/G1 013 any system for use in 3 questions is excessive. NRC lead examiner re-selected 012 2.2.37 ESFAS system with the same K/A, leaving only 2 RPS questions on this exam.

2.2.37 (4.6)

(4.6)

SRO 073 - Process Radiation Monitoring System. The original K/A tested the rad monitor (Q#89) system and how it is used to assess the status of safety functions. This exam has a T2/G1 026 high sampling of radiation monitor questions therefore the system was changed to the 073 2.4.21 Containment Spray System 026.

2.4.21 (4.6)

(4.6) 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 SRO procedure that would be used to control or mitigate the situation. There is no clear (Q#90)

T2/G1 procedural guidance on when entry into containment is allowed based upon 103 103 containment conditions. This would be governed by Radiation Protection or industrial A2.03 A2.02 safety procedures/personnel. Randomly selected A2.03 which concerns the (3.8)

(3.2) procedures that would be used to mitigate or control the effects of containment isolations.

SRO 002- Reactor Coolant System. The original K/A concerns the knowledge of system (Q#91) operation/status that must be reported to outside agencies. This K/A belongs in T3, T2/G2 002 not T2. Since the K/A for Q#100 in T3 was rejected, moved 2.4.30 for Q#91 to Q#100.

002 2.4.8 Randomly selected K/A 2.4.8 for Q#91 which is the knowledge of how AOPs are used 2.4.30 (4.5) with EOPs.

(4.1)

SRO The original K/A concerns the recognition of entry conditions for EOPs and AOPs. This (Q#99) is RO level knowledge. Randomly selected 2.4.14 which concerns the general T3 2.4.14 guidelines for EOP usage.

2.4.4 (4.5)

(4.7)

ES-401, Page 53 of 52

SRO The original K/A concerns the knowledge of the operational implications of notes (Q#100) cautions and warnings of EOPs. This is a T1 question and was moved to Q#81 since T3 2.4.30 Q#81 K/A was rejected. Moved 2.4.30 for Q#91 to Q#100 because it is a T3 K/A.

2.4.20 (4.1)

(4.3)

ES-401, Page 54 of 52

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 Describe activity to be performed Code*

A1 Calculate Minimum Required Boron to Maintain Conduct of Operations R,D Shutdown Margin K/A - 2.1.43, Importance Rating 4.1 A1JPM-RO-RBAL3 A2 Perform Time to Boil / Core Uncovery Conduct of Operations R,M Estimation K/A - 2.1.25, Importance Rating 3.9 A1JPM-RO-TTB A3 Perform Service Water Surveillance Equipment Control R,N A1JPM-RO-SURV3 K/A - 2.2.12, Importance Rating 3.7 A4 Ability to comply with Radiation Work Permit Radiation Control R,D,P requirements K/A - 2.3.7, Importance Rating 3.5 A1JPM-RO-ADMIN-RWP3 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 Describe activity to be performed Code*

A5 Review Calculated Minimum Required Boron to Conduct of Operations R,D Maintain Shutdown Margin K/A - 2.1.43, Importance Rating 4.3 A1JPM-SRO-RBAL3 A6 Determine Operability of SW Pump based on Conduct of Operations R,N Pump Curve K/A - 2.1.25, Importance Rating 4.2 A1JPM-SRO-ADMINSW A7 Apply T.S. for LPI and EDG Inoperability Equipment Control R,N A1JPM-SRO-TS20 K/A - 2.2.40, Importance Rating 4.7 A8 Ability to comply with Radiation Work Permit Radiation Control R,D requirements K/A - 2.3.7, Importance Rating 3.6 A1JPM- SRO-ADMIN-RWP3 A9 Classify an Emergency Event Emergency Plan R,N A1JPM-SRO-EAL18 K/A - 2.4.41, Importance Rating 4.6 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 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 D/S 1 004 A4.07 (RO 3.9 / SRO 3.7)

RO S2. Perform actions required for Reactor Building Sump Blockage E/EN/L/N/S 2 006 A2.02 (RO 3.9 / SRO 4.3)

RO / SRO-I / SRO-U S3. Respond to Pressurizer ERV Failure A/L/N/S 3 010 A2.03 (RO 4.1 / SRO 4.2)

RO / SRO-I / SRO-U S4. Initiate Common Feedwater N/L/S 4S E04 EA1.1 (RO 4.4 / SRO 4.2)

RO / SRO-I S5. Place Hydrogen Recombiner M55B in Operation D/L/P/EN/S 5 028 A4.01 (RO 4.0/SRO 4.0)

RO / SRO-I S6. Synchronize and Load #1 EDG with a failure of the load switch A/D/P/S 6 064 A2.05 (RO 3.1 / SRO 3.2)

RO / SRO-I S7. Bypass MSLI L/N/S 7 E02 EA1.1 (RO 4.0 / SRO 3.6)

RO / SRO-I S8. Reactor Building Purge Gaseous Release A/L/N/S 9 071 A1.06 (RO 2.5 / SRO 2.8)

RO / SRO-I / SRO-U

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 A/D/L P1. Perform ICS Startup 4S 041 A4.01 (RO 2.9 / SRO 3.1)

RO / SRO-I / SRO-U A/D/E P2. Manual Actuation of Halon System #3 8 086 A4.06 (RO 3.2 / SRO 3.2)

RO / SRO-I E/L/N/R P3. Align Boric acid via HPI Pump during Control Room Evacuation 1 068 AA1.08 (RO 4.2 / SRO 4.2)

RO / SRO-I / SRO-U

  • 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 5/5 /3 (C)ontrol room (D)irect from bank 5/4/2 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 2 / 2 / 1 (control room system)

(L)ow-Power/Shutdown 8/8/5 (N)ew or (M)odified from bank including 1(A) 6/6/3 (P)revious 2 exams 2 / 2 / 0 (randomly selected)

(R)CA 1/1/1 (S)imulator

Appendix D Scenario Outline Facility: ANO-1 Scenario No.: 1 R4 Op-Test No.: 2020-1 Examiners: ____________________________Operators: _____________________________

Initial Conditions:

- IC201 - Middle of life - EFIC failed

- 80% Power. - Group 5 Rod 12 and Group 3 Rod 1

- C-28A IA Compressor is out of service for are stuck and will not insert during overhaul (B-3256) Hang Caution card Rx trip

- Make sure vacuum alarm adjusted correctly - CV-2827 Failed Closed

- C-8B In Service Turnover:

- Boron 660 ppmB -

Event Event Event Malf. No.

No. Type* Description C-(BOP)

Loss of B3 with a failure of C-5B to start (Condenser 1 ED191 C-(SRO)

Vacuum Pump)

AOP I-(ATC)

LT-1001 Pressurizer Level Fails Low 2 TR049 I-(SRO)

(T.S. 3.3.15 Condition A - Function 11 Pressurizer Level)

TS, AOP C-(ATC)

C-(BOP) D RCP Seal Cooler Leak (35 gpm) 3 RC457 C-(SRO) (T.S. 3.4.13 Condition B)

TS, AOP C-(ATC)

FW075 B MFW Pump Trip with a failure of CV-2827 (Results in 4 C-(SRO)

CV2827 Reactor Trip)

EOP C-(ATC)

RD351 2 Stuck Control Rods - Emergency Boration within 15 5 C-(SRO)

RD350 minutes of Rx Trip CT-23 RC045 6 M-(ALL) Pressurizer Steam Space Leak increases to ~800 gpm (0.4 - 2 min)

C-(ATC) 7 FW621 EFIC Fails to actuate on low SG level C-(SRO)

C-(ATC)

C-(BOP) 8 RC045 RCPs must be secured within 2 minutes of LOSM C-(SRO)

CT-1

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 6

Appendix D Scenario Outline 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 1 scenario set)

Critical Tasks (2) 2 SCENARIO 1 OBJECTIVES

1) Evaluate individual ability to recognize and respond to a Loss of MCC (B3)
2) Evaluate individual ability to recognize and respond to a failed pressurizer level transmitter
3) Evaluate individual ability to recognize when conditions require the entry into technical specifications conditions.
4) Evaluate individual ability to estimate RCS leakage rate.
5) Evaluate individual ability to recognize and respond to excess RCS leakage.
6) Evaluate individual ability to reduce plant power.
7) Evaluate individual ability to recognize and respond to a Loss of Feedwater.
8) Evaluate individual ability to recognize and respond to Reactor Trip.
9) Evaluate individual ability to recognize and respond to a failure of EFIC
10) Evaluate individual ability to recognize and respond to a Loss of Subcooling Margin.

Page 2 of 6

Appendix D Scenario Outline SCENARIO 1 NARRATIVE Event One: Loss of B3 with a failure of C-5B to automatically start The crew will assume plant responsibility at ~80% power. Several alarms will occur due to the loss of load center but K02-C8 NON-ES Bus Loss of Voltage indicates a loss of B-3. Priority should be placed on verifying redundant equipment running for the following alarms, K05-A3 Vacuum Pump Trip, K05-C7 EH Pump P-14A/P-14B Trip, and K12-C3 Instrument Air Compressor Trouble. Actions will be directed from 1203.046 Section 3 Loss of Loadcenter B3. (BOP-C) (SRO-C)

Event Two: 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 Condition A - Function 11 Pressurizer Level.

(ATC-C) (SRO-C) AOP, TS Event Three: D RCP Seal Cooler Leak (35 gpm) develops The crew will recognize indications of an RCP Seal Cooler leak (Skewed seal injection flow rates, RCP Seal alarms) and receive K10-B2, Process Monitor Radiation Hi alarm. The crew will calculate a leak rate of ~35 gpm based on the NUC-ICW Surge Tank level rate of change (As reported from the field). 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 B. The crew will commence a plant shutdown in accordance with OP-1203.045, Rapid Plant Shutdown to allow for securing the RCP with the seal cooler leak. (ATC-C) (BOP-C) (SRO-C) AOP, TS Event Four: B MFW Pump Trip with a failure of CV-2827 (MFW Pump Crosstie valve)

Once power is stabilized and the RCP secured, the B MFWP will trip with a subsequent failure of the crosstie valve. With no feed to the B steam generator, the ATC will trip the reactor. The SRO will enter the Reactor Trip EOP. (ATC-C) (SRO-C) AOP Event Five: 2 Stuck Control Rods During immediate actions the ATC will identify two stuck rods and commence Emergency Boration per RT-12. It is a CRITICAL TASK to commence emergency boration within 15 minutes of the trip. Since Letdown was isolated due to the RCP seal cooler leak contingency actions from RT-12 will be required.

This may be handed off to the BOP or performed by the ATC. (ATC-C) (BOP-C) (SRO-C) EOP, CT CT Justification: CT-23 Safety significance - Establish and Maintain Reactor Shutdown Requirements.

Initiating Cue - Lack of Green In Limit lights for two control rods post trip.

Measurable Performance Standard - Manual initiation of Emergency Boration per RT-12 through Step 2.B with full flow through an injection nozzle within 15 minutes of the reactor trip.

Performance Feedback - HPI flow to the RCS via an injection nozzle. (CV-1220)

Event Six and Eight: Pressurizer Steam Space Leak (~800 gpm)

After emergency boration has been initiated and at the discretion of the lead examiner, a large pressurizer steam space leak will begin. This leak will result 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.

Page 3 of 6

Appendix D Scenario Outline Performance Feedback - Change of status lamps on control console for RCPs and indicated RCS flow near zero.

Event Seven: EFIC Fails to actuate on low SG level As a result of the B MFW Pump trip and crosstie valve failure, the B steam generator will not have a source of feedwater. Steam generator level will drop to the EFIC low level setpoint but will fail to actuate.

The ATC will identify the low level and manually actuate EFW from the remote matrix on C09.

(ATC-C) (SRO-C) EOP The scenario can be terminated at the discretion of the lead examiner.

Critical Task Table:

Safety Initiating Cue Measurable Performance significance Performance Feedback Standard Failure of a fission SCM monitor on ICCMDS Trip ALL RCPs Change of status product barrier indicates <30 oF and timer within 2 minutes of lamps on control (RCS) counting. LOSM following console for RCPs CT-1 the reactor trip. and indicated RCS Train A/B ICC Event alarms flow near zero.

on K11.

Establish and Lack of Green In Limit lights Manual initiation HPI flow to the Maintain Reactor for two control rods post trip. of Emergency RCS via an Shutdown Boration per injection nozzle.

Requirements. RT-12 through (CV-1220)

CT-23 Step 2.B with full flow through an injection nozzle within 15 minutes of the reactor trip.

Page 4 of 6

Appendix D Scenario Outline Procedures:

Event 1 OP-1203.012B - K02-C8 - NON-ES Bus Loss of Voltage OP-1203.012D - K05-A3 - Vacuum Pump Trip K05-C7 - EH Pump P-14A/P-14B Trip OP-1203.012K - K12-C3 - Instrument Air Compressor Trouble.

OP-1203.046 - Loss of Loadcenter (Section 3 - Loss of Loadcenter B3)

Event 2 OP-1203.012H - K09-C3, Pressurizer Level Lo K09-A3, Pressurizer Level Lo-Lo OP-1203.015 - Pressurizer Systems Failure (Section 4 - Pressurizer Level Indication Malfunction)

T.S 3.3.15 Condition A Event 3 OP-1203.012I - K10-B2 - Process Monitor Radiation Hi alarm OP-1203.031, Reactor Coolant Pump and Motor Emergency (Exhibit A, RCP Seal Bleedoff Isolation)

OP-1203.039, Excess RCS Leakage OP-1203.045, Rapid Plant Shutdown T.S. 3.4.13 Condition B.

Event 4 OP-1203.012F - K07-A8 - B MFP Turbine Trip OP-1203.027 - Loss of Steam Generator Feed Event 5 OP-1202.001 - Reactor Trip OP-1202.012 - Repetitive Tasks (RT Emergency Boration)

(RT Initiate HPI)

Events 6, 7, 8 OP-1202.002 - Loss of Subcooling Margin OP-1202.012 - Repetitive Tasks (RT Initiate Full HPI)

(RT Proper EFW Actuation and Control Verification)

(RT Check Proper Electrical Response)

(RT Check SG Tube Integrity)

(RT Verify Proper ESAS Actuation)

(RT Control RCS Pressure)

Page 5 of 6

Appendix D Scenario Outline Simulator Instructions for Scenario 1 Ensure malfunctions agree with scenario guide (either schedule file or IC)

Event Time Event Description No. Type C-(BOP)

Loss of B3 with a failure of C-5B to start (Condenser 1 T=0 C-(SRO)

Vacuum Pump)

AOP I-(ATC)

LT-1001 Pressurizer Level Fails Low 2 T=12 I-(SRO)

(T.S. 3.3.15 Condition A - Function 11 Pressurizer Level)

TS, AOP C-(ATC)

C-(BOP) 3 T=20 RCP Seal Cooler Leak (35 gpm) (T.S. 3.4.13 Condition B)

C-(SRO)

TS, AOP C-(ATC)

B MFW Pump Trip with a failure of CV-2827 (Results in 4 T=55 C-(SRO)

Reactor Trip)

EOP C-(ATC) 5 T=55 C-(SRO) 2 Stuck Control Rods - Emergency Boration CT-23 6 T=60 M-(ALL) Pressurizer Steam Space Leak increases to ~800 gpm C-(ATC) 7 T=60 EFIC Fails to actuate on low SG level C-(SRO)

C-(ATC)

C-(BOP) 8 T=60 RCPs must be secured within 2 minutes of LOSM C-(SRO)

CT-1 Page 6 of 6

Appendix D Scenario Outline Facility: ANO-1 Scenario No.: 2 R3 Op-Test No.: 2020-1 Examiners: ____________________________Operators: _____________________________

Initial Conditions:

- IC202 - Beginning of Life - B5106 and B5110 Open

- 60% Power. - CV-2670 failed open (P7B to SG-A)

- Mablevale Line isolated for repairs - Boron 1290 ppm

- Maintain <490 MWe (Net)

Turnover:

- Perform a 2 minute delithiation -

- Reactivity Brief completed

- EFIC 1/2 Trip, I&C performing monthly surveillance Event Event Event Malf. No.

No. Type* Description N-(BOP) 1 N/A Align for 2 minute delithiation N-(SRO)

C-(SRO) Inverter Y11 failure 2 K01-A5 TS (T.S. 3.8.7 Condition A)

C-(ATC)

C-(BOP) Inadvertent EFW actuation 3 ES254 C-(SRO) (T.S. 3.7.5 Condition E)

AOP, TS C-(ATC)

B2264 C-(BOP) P-33C ICW Pump trips with an auto start failure on P-33B 4

SW071 C-(SRO) ICW Pump.

ACA I-(ATC)

I-(BOP) 5 TR580 Controlling Header Pressure fails to 800 psig I-(SRO)

AOP 6 MS131 M-(ALL) A Main Steam Line break inside containment I-(ATC)

ES263 I-(BOP) 7 ESAS Channels 5 & 6 fail to actuate automatically ES264 I-(SRO)

CT-19 C-(ATC)

CV2670 EFIC Vector Isolation failure for one EFW flow path to 8 CT-11 HIC2646 failed generator CT-16

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 7

Appendix D Scenario Outline 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 1 scenario set)

Critical Tasks (2) 2 SCENARIO 2 OBJECTIVES

1) Evaluate individual ability to perform RCS Delithiation.
2) Evaluate individual ability to recognize and respond to 120 VAC Inverter Annunciators on K-1.
3) Evaluate individual ability to recognize and respond to inadvertent EFIC actuation
4) Evaluate individual ability to recognize when conditions require the entry into Technical Specifications conditions.
5) Evaluate individual ability to recognize and respond to ICW System Annunciators on K-12.
6) Evaluate individual ability to recognize and recover letdown flow following high letdown temperature.
7) Evaluate individual ability to recognize and respond to ICS input failures.
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.

Page 2 of 7

Appendix D Scenario Outline SCENARIO 2 NARRATIVE Event One: Normal operation align for two minute delithiation The crew will assume plant responsibility at ~60% 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: Inverter Y11 failure Following the delithiation, K01-A5, RS1 Inverter Trouble annunciator will alert the crew of an inverter failure. The BOP will dispatch the NLO to investigate. Report from the field will be that the Y11 Inverter has failed and 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) ACA, TS Event Three: Inadvertent EFW actuation Once the T.S. entry has been announced to the crew, I&C will cause an inadvertent EFIC actuation during their monthly surveillance. The crew should respond to K12-A5, EFW Actuation Signal. The ACA will direct stopping the EFW pumps and then send the RO to 1105.005, Emergency Feedwater Initiation and Control for resetting EFIC. During the restoration, EFW will be inoperable and the SRO will enter T.S. 3.7.5 Condition E.

(ATC-C) (BOP-C) (SRO-C) ACA, TS Event Four: P-33C ICW Pump trips with a failure of P-33B to automatically start Once EFIC is reset and TS entry announced, 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: Controlling Header Pressure fails to 800 psig Once Letdown has been restored, the controlling header pressure will fail to 800 psig. This will cause ICS to pull rods and close governor valves to recover 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. The ATC and BOP will work together to maintain <490 MWe while recovering header pressure. Once header pressure is recovered and a good signal selected, the Turbine, TBVs, and SG/RX can be returned to automatic control.

(ATC-I) (BOP-I) (SRO-I) AOP Event Six: A Main Steam Line break inside containment The major event will be an overcooling due to a main steam line break inside containment that 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 Page 3 of 7

Appendix D Scenario Outline Event Seven: 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-I) (BOP-I) (SRO-I) 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.

Event Eight: EFIC Vector Isolation failure for one EFW flow path to failed 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.

The critical task of manually controlling EFW must be performed prior to the completion of RT-6.

(ATC-C) (SRO-C) EOP(RT), CT CT Justification: CT-11 and CT-16 Safety significance - Excessive primary to secondary heat transfer due to overfeeding faulted steam generator.

Initiating Cue - RCS overcooling in progress, skewed EFW flow to faulted SG preventing depressurization and causing continued overcooling conditions.

Measurable Performance Standard - Manually control EFW flow to A Steam Generator prior to reporting the completion of RT-6.

Performance Feedback - EFW flowrate near zero, A Steam Generator pressure trending towards zero, RCS temperature stable or rising.

The scenario can be terminated at the discretion of the lead examiner.

Page 4 of 7

Appendix D Scenario Outline Critical Task Table:

Safety Initiating Cue Measurable Performance significance Performance Feedback Standard Isolate possible RB Pressure HI (2.175 Manually actuate Annunciators for RCS leak paths psig) alarm on K11 and Channels 5 & 6 RB Isolation and assure Reactor Building prior to reporting Channel 5 & 6 on containment pressure >4 psig (ESAS the completion of K11 and CT-19 integrity. setpoint). RT-10. components reposition to ESAS position for Channels 5 & 6.

Excessive primary RCS overcooling in Manually control EFW flowrate to secondary heat progress, skewed EFW EFW flow to A near zero, A transfer due to flow to faulted SG Steam Generator Steam Generator CT-11 overfeeding preventing prior to reporting pressure trending CT-16 faulted steam depressurization and the completion of towards zero, generator. causing continued RT-6. RCS temperature overcooling conditions. stable or rising.

Page 5 of 7

Appendix D Scenario Outline Procedures:

Event 1 OP-1104.002 Step 10.0 Event 2 OP-1203.012A - K01-A5, RS1 INVERTER TROUBLE Attachment E, RS1 Inverter Trouble Alarm Response T.S 3.8.7 Condition A Event 3 OP-1203.012K - K12-A5 - EFW Actuation Signal OP-1105.005Emergency Feedwater Initiation and Control T.S. 3.7.5 Condition E Event 4 OP-1203.012K - K12-B4 - ICW Flow Lo OP-1203.012I - K10-A Letdown Temp Hi OP-1104.002, Makeup & Purification System Operation Event 5 OP-1203.012F - K07-B4 - SASS Mismatch OP-1203.001, ICS Abnormal Operations OP-1105.004, Integrated Control System Events 6, 7, 8 OP-1202.001 - Reactor Trip OP-1202.003 - Overcooling OP-1202.012 - Repetitive Tasks (RT Proper EFW Actuation and Control Verification)

(RT Proper MSLI Actuation and Control Verification)

(RT Verify Proper ESAS Actuation)

(RT Control RCS Pressure)

Page 6 of 7

Appendix D Scenario Outline 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 Time Event Description No. Type N-(BOP) 1 T=0 Align for 2 minute delithiation N-(SRO)

C-(SRO) Inverter Y11 failure 2 T=5 TS (T.S. 3.8.7 Condition A)

C-(ATC)

Inadvertent EFW actuation 3 T=15 C-(SRO)

(T.S. 3.7.5 Condition E)

AOP, TS C-(ATC)

C-(BOP) P-33C ICW Pump trips with an auto start failure on P-33B ICW 4 T=35 C-(SRO) Pump.

ACA I-(ATC)

I-(BOP) 5 T=50 Controlling Header Pressure fails to 800 psig I-(SRO)

AOP 6 T=60 M-(ALL) A Main Steam Line break inside containment I-(ATC)

I-(BOP) 7 T=60 ESAS Channels 5 & 6 fail to actuate automatically I-(SRO)

CT-19 C-(ATC)

EFIC Vector Isolation failure for one EFW flow path to failed 8 T=60 CT-11 generator CT-16 Page 7 of 7

Appendix D Scenario Outline Facility: ANO-1 Scenario No.: 4 R2 Op-Test No.: 2020-1 Examiners: ____________________________Operators: _____________________________

Initial Conditions:

- IC205 - Beginning of Life - Have Group 6 on Remote Display 5% Power.

- Boron 1710 ppm Turnover:

- Place A MFWP in service - Provide figures / attachments for

- Raise power to 10% startup

- Lead Examiner will need the EFIC Maintenance Bypass key Event Event Event Malf. No.

No. Type* Description N-(ATC) 1 N/A N-(BOP) Place A MFWP in service and secure P-75 Aux FW Pump N-(SRO)

N-(BOP) 2 N/A Reset ARTS N-(SRO)

R-(ATC) 3 N/A Raise power to 10%

R-(SRO)

C-(ATC)

C-(BOP) Group 7 Rods do not sequence on resulting in no overlap 4 RD271 C-(SRO) between Groups 6 & 7. (T.S. 3.2.1 Condition C)

TS I-(BOP)

I-(CRS) A OTSG EFIC Low Range Level fails low (LI-2618) 5 TR596 ACA (T.S. 3.3.11 Condition A)

TS C-(ATC)

C-(BOP) 6 ED452 Loss of ICS Power C-(SRO)

AOP 7 ED183 M-(ALL) Loss of Offsite Power C-(BOP) 8 DG174 #2 EDG Failure C-(SRO)

C-(ATC)

P-7A EFW Pump trips on trip throttle linkage failure.

9 FW076 C-(SRO)

Control EFW flow with P-7B when #1 EDG is recovered.

CT-16 C-(BOP) 10 DG175 C-(SRO) #1 EDG fails to auto start CT-8

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 7

Appendix D Scenario Outline 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) 2 EOP contingencies requiring substantive actions ( 1per 1 scenario set)

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 a loss of ICS power.
6) Evaluate individual ability to recognize and respond to Electrical Distribution Annunciators on K-02.
7) Evaluate individual ability to perform an RCS Leakage Investigation.
8) Evaluate individual ability to recognize and respond to Excess RCS Leakage.
9) Evaluate individual ability to recognize and respond to Degraded Power.
10) Evaluate individual ability to recognize and respond to EDG failure to start.
11) Evaluate individual ability to perform DG1 Start from Control Room.

Page 2 of 7

Appendix D Scenario Outline SCENARIO 4 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. ARTS must be reset prior exceeding 7% NI Power.

(BOP-N) (SRO-N)

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.

(T.S. 3.2.1 Condition C)

(ATC-C) (BOP-C) (SRO-C)TS Event Five: A OTSG EFIC Low Range Level fails low (LT-2618)

This failure will result in a half trip of EFIC as indicated by K12-C7, EFIC System Trouble. As a result of the failure, the SRO will enter T.S. 3.3.11 Condition A and direct the BOP to take EFIC Channel A to Maintenance Bypass in order to reset the half trip.

(ATC-C) (BOP-C) (SRO-C) ACA, TS Event Six: Loss of ICS Power Once the Tech Spec entry is announced to the crew, a Loss of ICS power will occur which will require entry into 1203.001, ICS Abnormal Operations AOP. The CRS will direct a manual reactor trip and will continue in the AOP while also performing 1202.001, Reactor Trip EOP (ATC-C) (BOP-C) (SRO-C) AOP Event Seven: Loss of Offsite Power Once the immediate actions are completed in Reactor Trip and the Atmospheric Dump Isolation valves are opened by the AOP, 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: #2 EDG Failure Shortly following the loss of offsite power, the #2 EDG will trip resulting in a momentary station blackout if the #1 EDG wasnt already started.

(BOP-C) (SRO-C) EOP Page 3 of 7

Appendix D Scenario Outline Event Nine: P-7A EFW Pump trips on trip throttle linkage failure. Control of EFW flow with P-7B when #1 EDG is recovered Shortly after the loss of offsite power, P-7A steam driven EFW Pump will trip due to a trip throttle linkage failure. Mechanical maintenance assistance will be required for repairs of the linkage and the pump will be unavailable for the duration of the scenario. In addition to the linkage failure, P-7B electric EFW Pump will not be available due to a failure of the #1 EDG to automatically start. The BOP will start the diesel which will energize P-7B. Due to the failed EFIC level transmitter, P-7B will be feeding full flow to the A Steam Generator. The ATC will identify this and manually control EFW flow. This is a CRITICAL TASK to be completed prior to the steam generator level reaching 400 inches.

(ATC-C) (SRO-C) CT 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 control EFW flow to the A steam generator prior to steam generator reaching 400 inches (Carry-over into main steam lines)

Performance Feedback - EFW flow being controlled such that Steam Generator level is approaching 300 - 340 at a rate of 2-8/minute.

Event Ten: #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. Starting #1 EDG is 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.

The scenario can be terminated at the discretion of the lead examiner.

Page 4 of 7

Appendix D Scenario Outline Critical Task Table:

Safety Initiating Cue Measurable Performance significance Performance Feedback Standard Prevent Loss of Offsite Power, BOP will start the Change of status degradation of the A3 de-energized, #1 EDG within 15 lamps on control mitigative normal lighting in the minutes of console for capability of the control room off. Blackout. EAL #1 EDG, CT-8 plant escalation is components on required if the C18 and normal EDG is not lighting returns for started within 15 the control room.

minutes.

Excessive primary RCS overcooling in Manually control EFW flow being to secondary heat progress, lower than EFW flow to the controlled such transfer due to normal steam generator A steam that Steam overfeeding steam pressure, skewed EFW generator prior Generator level is generator. flow to steam generator approaching to steam CT-16 Initiating Cue - with failed level 300 - 340 at a Measurable transmitter.

generator rate of 2-8 per Performance reaching 400 minute.

Standard - inches (Carry-over into main steam lines)

Page 5 of 7

Appendix D Scenario Outline Procedures:

Events 1 and 2 OP-1102.002, Plant Startup OP-1106.016, Condensate, Feedwater and Steam System Operation OP-1105.004, Integrated Control System Event 3 OP-1203.012K - K12-C7 - EFIC System Trouble K12-F7 - EFIC Maintenance Bypass OP-1105.005, Emergency Feedwater Initiation and Control T.S. 3.3.11 Condition A - 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time Event 4 OP-1102.002, Plant Startup Event 5 OP-1105.009, CRD System Operating Procedure T.S. 3.2.1 Condition C - 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time Events 6 OP-1203.012F - K07-A4 - ICS/AUX SYS Power Supply OP-1203.001, ICS Abnormal Operations AOP OP-1202.001, Reactor Trip OP-1202.012, Repetitive Tasks (RT Proper EFW Actuation and Control Verification)

(RT Check Proper Electrical Response)

(RT NNI and ICS power available)

(RT Control RCS Pressure)

Events 7, 8, 9, 10 OP-1203.012A - K01-A4 - EDG2 TRIP OP-1203.012B - K02-C1 - SU 1 Trouble K02-D3 - SU 2 Trouble OP-1203.012K - K12-B5 - P-7A Turbine Trip OP-1202.007, Degraded Power OP-1107.002, ES Electrical System Operation OP-1202.012, Repetitive Tasks (RT Check EDG Operation)

(RT Proper MSLI Actuation and Control Verification)

(RT Maximize RB Cooling)

RT Check SG Tube Integrity)

Page 6 of 7

Appendix D Scenario Outline Simulator Instructions for Scenario 4 Reset simulator to BOL ~2-5% power IC steady state Ensure malfunctions agree with scenario guide (either schedule file or IC)

Event Time Event Description No. Type N-(ATC) 1 T=0 N-(BOP) Place A MFWP in service and secure P-75 Aux FW Pump N-(SRO)

N-(BOP) 2 T=15 Reset ARTS N-(SRO)

I-(BOP)

A OTSG EFIC Low Range Level fails low 3 T=20 I-(CRS)

(T.S. 3.3.11 Condition A)

TS R-(ATC) 4 T=40 Raise power to 10%

R-(SRO)

C-(ATC)

C-(BOP) Group 7 Rods do not sequence on resulting in no overlap between 5 T=60 C-(SRO) Groups 6 & 7. (T.S. 3.2.1 Condition C)

TS C-(ATC)

C-(BOP) 6 T=65 Loss of ICS Power C-(SRO)

AOP 7 T=70 M-(ALL) Loss of Offsite Power T=70 C-(BOP) 8 #2 EDG Failure C-(SRO)

T=70 C-(BOP) 9 C-(SRO) #1 EDG fails to auto start CT-16 C-(ATC)

P-7A EFW Pump trips on trip throttle linkage failure.

10 T=75 C-(SRO)

Control EFW flow with P-7B when #1 EDG is recovered.

CT-8 Page 7 of 7

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