ML17270A169

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2017-09-DRAFT Outlines
ML17270A169
Person / Time
Site: Callaway Ameren icon.png
Issue date: 09/20/2017
From: Vincent Gaddy
Operations Branch IV
To:
Union Electric Co
References
Download: ML17270A169 (56)
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Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Callaway Date of Exam: 9/20/17 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

1. 1 3 3 3 3 3 3 18 6 Emergency and Abnormal Plant 2 1 1 2 N/A 1 2 N/A 2 9 4 Evolutions Tier Totals 4 4 5 4 5 5 27 10 1 3 2 2 3 2 2 3 3 3 2 3 28 5 2.

Plant 2 1 0 1 1 1 1 1 1 1 1 1 10 3 Systems Tier Totals 4 2 3 4 3 3 4 4 4 3 4 38 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 3 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

  • These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
    • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

Rev. 11

ES-401 2 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) Knowledge of the interrelations between a reactor Reactor Trip, Stabilization, Recovery / 1 X trip and the following: (CFR 41.7 / 45.7) 2.6 4 EK2.02 Breakers, relays and disconnects 000008 (APE 8) Pressurizer Vapor Space Knowledge of the interrelations between the Accident / 3 Pressurizer Vapor Space Accident and the X 2.7* 17 following: (CFR 41.7 / 45.7)

AK2.02 Sensors and detectors 000009 (EPE 9) Small Break LOCA / 3 Knowledge of the operational implications of the following concepts as they apply to the small break X LOCA: (CFR 41.8 / 41.10 / 45.3) 4.2 8 EK1.01 Natural circulation and cooling, including reflux boiling 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Ability to determine and interpret the following as Makeup / 2 they apply to the Loss of Reactor Coolant Makeup:

X (CFR 43.5/ 45.13) 2.9 10 AA2.04 How long PZR level can be maintained within limits 000025 (APE 25) Loss of Residual Heat Ability to operate and / or monitor the following as Removal System / 4 they apply to the Loss of Residual Heat Removal X 3.4 7 System: (CFR 41.7 / 45.5 / 45.6)

AA1.03 LPI Pumps 000026 (APE 26) Loss of Component Ability to operate and / or monitor the following as Cooling Water / 8 they apply to the Loss of Component Cooling X Water: (CFR 41.7 / 45.5 / 45.6) 3.1 15 AA1.05 The CCWS surge tank, including level control and level alarms, and radiation alarm 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient 2.4.31 Knowledge of annunciator alarms, Without Scram / 1 X indications, or response procedures. (CFR: 41.10 / 4.2 18 43.5 )

000038 (EPE 38) Steam Generator Tube Knowledge of the reasons for the following Rupture / 3 responses as the apply to the SGTR: (CFR 41.5 /

41.10 / 45.6 / 45.13)

X 4.2 12 EK3.06 Actions contained in EOP for RCS water inventory balance, S/G tube rupture, and plant shutdown procedures 000040 (APE 40; BW E05; CE E05; W E12) Ability to determine and interpret the following as Steam Line RuptureExcessive Heat they apply to the Steam Line Rupture: (CFR: 43.5 /

X 4.1 14 Transfer / 4 45.13)

AA2.05 When ESFAS systems may be secured 000054 (APE 54; CE E06) Loss of Main Knowledge of the operational implications of the Feedwater /4 following concepts as they apply to Loss of Main X 3.6 16 Feedwater (MFW): (CFR 41.8 / 41.10 / 45.3)

AK1.02 Effects of feedwater introduction on dry S/G 000055 (EPE 55) Station Blackout / 6 Rev. 11

ES-401 3 Form ES-401-2 000056 (APE 56) Loss of Offsite Power / 6 Knowledge of the reasons for the following responses as they apply to the Loss of Offsite X Power: (CFR 41.5,41.10 / 45.6 / 45.13) 4.4 6 AK3.02 Actions contained in EOP for loss of offsite power 000057 (APE 57) Loss of Vital AC Ability to determine and interpret the following as Instrument Bus / 6 they apply to the Loss of Vital AC Instrument Bus:

X (CFR: 43.5 / 45.13) 3.5 11 AA2.12 PZR level controller, instrumentation, and heater indications 000058 (APE 58) Loss of DC Power / 6 Knowledge of the reasons for the following responses as they apply to the Loss of DC Power:

X (CFR 41.5,41.10 / 45.6 / 45.1) 4.0 1 AK3.02 Actions contained in EOP for loss of dc power 000062 (APE 62) Loss of Nuclear Service Ability to operate and / or monitor the following as Water / 4 they apply to the Loss of Nuclear Service Water (SWS): (CFR 41.7 / 45.5 / 45.6)

X 2.9 5 AA1.07 Flow rates to the components and systems that are serviced by the SWS; interactions among the components 000065 (APE 65) Loss of Instrument Air / 8 2.4.46 Ability to verify that the alarms are consistent X with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 4.2 13 45.12) 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment / 3 Knowledge of the interrelations between the (LOCA Outside Containment) and the following: (CFR: 41.7

/ 45.7)

X 3.5 3 EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(W E11) Loss of Emergency Coolant Knowledge of the operational implications of the Recirculation / 4 following concepts as they apply to the (Loss of Emergency Coolant Recirculation) (CFR: 41.8 /

X 3.7 9 41.10 / 45.3)

EK1.1 Components, capacity, and function of emergency systems.

(BW E04; W E05) Inadequate Heat 2.4.4 Ability to recognize abnormal indications for TransferLoss of Secondary Heat Sink / 4 X system operating parameters that are entry-level 4.5 2 conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6)

K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18 Rev. 11

ES-401 4 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 000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Knowledge of the reasons for the Instrumentation / 7 following responses as they apply to the Loss of Source Range X Nuclear Instrumentation: (CFR 3.2 19 41.5,41.10 / 45.6 / 45.13)

AK3.01 Startup termination on source-range loss 000033 (APE 33) Loss of Intermediate Range Nuclear Ability to determine and interpret Instrumentation / 7 the following as they apply to the Loss of Intermediate Range X Nuclear Instrumentation: (CFR: 3.1 21 43.5 / 45.13)

AA2.11 Loss of compensating voltage 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 2.2.22 Knowledge of the limiting X conditions for operations and 4.0 22 safety limits (CFR: 41.5 / 43.2 /

45.2) 000051 (APE 51) Loss of Condenser Vacuum / 4 Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum:

X 3.9 23 (CFR: 43.5 / 45.13)

AA2.02 Conditions requiring reactor and/or turbine trip 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms

/7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 Ability to operate and / or monitor the following as they apply to the X Control Room Evacuation: (CFR 4.1 26 41.7 / 45.5 / 45.6)

AA1.03 S/G level 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 Knowledge of the interrelations between the High Reactor Coolant Activity and the following:

X 2.6 20 (CFR 41.7 / 45.7)

AK2.01 Process radiation monitors 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 Rev. 11

ES-401 5 Form ES-401-2 (W E16) High Containment Radiation /9 Knowledge of the reasons for the following responses as they apply to the (High Containment Radiation) (CFR: 41.5 / 41.10, X 45.6, 45.13) 2.9 27 EK3.2 Normal, abnormal and emergency operating procedures associated with (High Containment Radiation)

(BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA CooldownDepressurization / 4 2.4.18 Knowledge of the specific X bases for EOPs (CFR: 41.10 / 3.3 24 43.1 /45.13)

(BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Knowledge of the operational Shock / 4 implications of the following concepts as they apply to the (Pressurized Thermal Shock)

X (CFR: 41.8 / 41.10, 45.3) 3.5 25 EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the (Pressurized Thermal Shock).

(CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals: 1 1 2 1 2 2 Group Point Total: 9 Rev. 11

ES-401 6 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 Knowledge of the operational implications of Pump the following concepts as they apply to the X RCPS: (CFR: 41.5 / 45.7) 2.8 47 K5.02 Effects of RCP coastdown on RCS parameters 004 (SF1; SF2 CVCS) Chemical and 2.1.25 Ability to interpret reference materials, Volume Control X such as graphs, curves, tables etc. (CFR: 3.9 46 41.10 / 43.5 / 45.12) 005 (SF4P RHR) Residual Heat Ability to (a) predict the impacts of the Removal following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate X 3.5 43 the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.02 Pressure transient protection during cold shutdown 006 (SF2; SF3 ECCS) Emergency Ability to predict and/or monitor changes in Core Cooling parameters (to prevent exceeding design X limits) associated with operating the ECCS 2.9 36 controls including: (CFR: 41.5 / 45.5)

A1.12 RHR heatup limits 007 (SF5 PRTS) Pressurizer Ability to monitor automatic operation of the Relief/Quench Tank PRTS, including: (CFR: 41.7 / 45.5)

X 2.7* 52 A3.01 Components which discharge to the PRT 008 (SF8 CCW) Component Cooling Ability to (a) predict the impacts of the Water following malfunctions or operations on the CCWS, and (b) based on those predictions, X use procedures to correct, control, or mitigate 3.3 55 the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.01 Loss of CCW pump 010 (SF3 PZR PCS) Pressurizer Knowledge of bus power supplies to the Pressure Control X following: (CFR: 41.7) 3.0 40 K2.01 PZR heaters 012 (SF7 RPS) Reactor Protection Ability to manually operate and/or monitor in X the control room: (CFR: 41.7 / 45.5 to 45.8) 3.3 53 A4.04 Bistable, trips, reset and test switches 013 (SF2 ESFAS) Engineered Knowledge of the operational implications of Safety Features Actuation the following concepts as they apply to the X 2.9 38 ESFAS: (CFR: 41.5 / 45.7)

K5.02 Safety system logic and reliability 022 (SF5 CCS) Containment Cooling Knowledge of CCS design features(s) and/or interlock(s) which provide for the following:

X (CFR 41.7) 3.1 49 K 4.02 Correlation of fan speed and flowpath changes with containment pressure 025 (SF5 ICE) Ice Condenser Rev. 11

ES-401 7 Form ES-401-2 026 (SF5 CSS) Containment Spray Knowledge of the effect that a loss or malfunction of the CSS will have on the X 3.9 34 following: (CFR: 41.7 / 45.6)

K3.01 CCS 039 (SF4S MSS) Main and Reheat Knowledge of the physical connections and/or Steam cause-effect relationships between the MRSS X and the following systems: (CFR: 41.2 to 41.9 / 2.7* 37 45.7 to 45.8)

K1.08 MFW 059 (SF4S MFW) Main Feedwater Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the X 3.1* 30 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.05 Rupture in MFW suction or discharge line 061 (SF4S AFW) Ability to predict and/or monitor changes in Auxiliary/Emergency Feedwater parameters (to prevent exceeding design X limits) associated with operating the AFW 3.6 35 controls including: (CFR: 41.5 / 45.5)

A1.05 AFW flow/motor amps 062 (SF6 ED AC) AC Electrical Ability to monitor automatic operation of the ac Distribution distribution system, including: (CFR: 41.7 /

X 3.5 29 45.5)

A3.05 Safety-related indicators and controls 063 (SF6 ED DC) DC Electrical Knowledge of DC electrical system design Distribution feature(s) and/or interlock(s) which provide for X the following: (CFR: 41.7) 2.9* 44 K4.02 Breaker interlocks, permissives, bypasses and cross-ties 064 (SF6 EDG) Emergency Diesel Knowledge of the effect of a loss or Generator malfunction of the following will have on the X 2.7 45 ED/G system: (CFR: 41.7 / 45.7)

K6.07 Air receivers 073 (SF7 PRM) Process Radiation Knowledge of the physical connections and/or Monitoring cause/effect relationships between the PRM X system and the following systems: (CFR: 41.2 3.6 48 to 41.9 / 45.7 to 45.8)

K1.01 Those systems served by PRMs 076 (SF4S SW) Service Water Ability to monitor automatic operation of the X SWS, including: (CFR: 41.7 / 45.5) 3.7 39 A3.02 Emergency heat loads 078 (SF8 IAS) Instrument Air Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following:

X 2.7 33 (CFR: 41.7)

K4.02 Cross-over to other air systems 103 (SF5 CNT) Containment Knowledge of the effect that a loss or malfunction of the containment system will X have on the following: (CFR: 41.7 / 45.6) 3.8 54 K3.02 Loss of containment integrity under normal conditions Rev. 11

ES-401 8 Form ES-401-2 053 (SF1; SF4P ICS*) Integrated Control 005 (SF4P RHR) Residual Heat Knowledge of bus power supplies to the Removal X following: (CFR: 41.7) 3.0 41 K2.01 RHR pumps 008 (SF8 CCW) Component Cooling 2.2.42 Ability to recognize systems parameters Water X that are entry-level conditions for Technical 50 3.9 specifications. (CFR: 41.7 / 41.10 / 43.2 / 43.3

/ 45.3) 010 (SF3 PZR PCS) Pressurizer Ability to predict and/or monitor changes in Pressure Control parameters (to prevent exceeding design X limits) associated with operating the PZR PCS 3.7 31 controls including: (CFR: 41.5 / 45.5)

A1.07 RCS pressure 039 (SF4S MSS) Main and Reheat Ability to manually operate and/or monitor in Steam X the control room: (CFR: 41.7 / 45.5 to 45.8) 2.8* 28 A4.07 Steam dump valves 061 (SF4S AFW) Knowledge of the effect of a loss or Auxiliary/Emergency Feedwater malfunction of the following will have on the X 2.6 51 AFW components: (CFR: 41.7 / 45.7)

K6.02 Pumps 063 (SF6 ED DC) DC Electrical Knowledge of the physical connections and/or Distribution cause/effect relationships between the DC X electrical system and the following systems: 2.9 32 (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.03 Battery charger and battery 103 (SF5 CNT) Containment 2.4.4 Ability to recognize abnormal indications for system operating parameters that are X entry-level conditions for emergency and 4.5 42 abnormal operating procedures. (CFR: 41.10 /

43.2 / 45.6)

K/A Category Point Totals: 3 2 2 3 2 2 3 3 3 2 3 Group Point Total: 28 Rev. 11

ES-401 9 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 Knowledge of RCS design feature(s) and/or Coolant interlock(s) which provide for the following:

X 2.7 57 (CFR: 41.7)

K4.01 Filling and draining the RCS 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Ability to (a) predict the impacts of the following Instrumentation malfunctions or operations on the NIS; and (b based on those predictions, use procedures to X correct, control, or mitigate the consequences 3.3 60 of those malfunctions or operations: (CFR:

41.5 / 43.5 / 45.3 / 45.5)

A2.05 Core void formation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Knowledge of the operational implications of Monitor the following concepts as they apply to the ITM X 3.7 62 system: (CFR: 41.5 / 45.7)

K5.03 Indication of superheating 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment X Purge System controls including: (CFR: 41.5 / 3.0* 59 45.5)

A1.03 Containment pressure, temperature, and humidity 033 (SF8 SFPCS) Spent Fuel Pool Knowledge of the effect that a loss or Cooling malfunction of the Spent Fuel Pool Cooling X System will have on the following: (CFR: 41.7 / 3.0 56 45.6)

K3.03 Spent fuel temperature 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator Knowledge of the effect of a loss or malfunction on the following will have on the X 3.1 58 S/GS: (CFR: 41.7 / 45.7)

K6.02 Secondary PORV 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Ability to manually operate and/or monitor in Generator X the control room: (CFR: 41.7 / 45.5 to 45.8) 2.8 65 A4 06 Turbine stop valves Rev. 11

ES-401 10 Form ES-401-2 055 (SF4S CARS) Condenser Air 2.4.46 Ability to verify that the alarms are Removal X consistent with the plant conditions. (CFR: 4.2 61 41.10 / 43.5 / 45.3 / 45.12) 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste Ability to monitor automatic operation of the Liquid Radwaste System including: (CFR: 41.7 X 3.6 64

/ 45.5)

A3.02 Automatic isolation 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water Knowledge of the physical connections and/or cause/effect relationships between the X circulating water system and the following 2.5 63 systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.01 SWS 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10 Rev. 11

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Callaway Date of Exam: 9/20/17 Category K/A # Topic RO SRO-only IR # IR #

2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical 3.4 67 requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10 / 43.2) 2.1.18 Ability to make accurate, clear, and concise logs,

1. Conduct of records, status boards, and reports. (CFR: 41.10 / 45.12 3.6 68 Operations / 45.13) 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high 3.3 71 pressure, caustic, chlorine, oxygen and hydrogen).

(CFR: 41.10 / 45.12)

Subtotal 3 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with 2.6 70 the transmission system operator. (CFR: 41.10 / 43.5 /

45.13)

2. Equipment Control 2.2.13 Knowledge of tagging and clearance procedures (CFR:

4.1 69 41.10 / 45.13) 2.2.43 Knowledge of the process used to track inoperable 3.0 73 alarms. (CFR: 41.10 / 43.5 / 45.13)

Subtotal 3 2.3.4 Knowledge of radiation exposure limits under normal or 3.2 66 emergency conditions. (CFR: 41.12 / 43.4 / 45.10)

3. Radiation 2.3.7 Ability to comply with radiation work permit requirements Control during normal or abnormal conditions. (CFR: 41.12 / 3.5 72 45.10)

Subtotal 2 2.4.29 Knowledge of the emergency plan. (CFR: 41.10 / 43.5 /

3.1 74 45.11)

4. Emergency Procedures/Plan 2.4.14 Knowledge of general guidelines for EOP Usage. (CFR:

3.8 75 41.10 / 45.13)

Subtotal 2 Tier 3 Point Total 10 Rev. 11

ES-401 PWR Examination Outline Form ES-401-2 Facility: Callaway Date of Exam: 9/20/17 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

1. 1 18 3 3 6 Emergency and Abnormal Plant 2 N/A N/A 9 2 2 4 Evolutions Tier Totals 27 5 5 10 1 28 2 3 5 2.

Plant 2 10 0 2 1 3 Systems Tier Totals 38 4 4 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 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.

Rev. 11

ES-401 2 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 000011 (EPE 11) Large Break LOCA / 3 Ability to determine and interpret the following as they apply to a Large Break LOCA: (CFR 43.5/

45.13)

X 4.7 78 EA2.01 Actions to be taken, based on RCS temperature and pressure - saturated and superheated 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 000025 (APE 25) Loss of Residual Heat Removal System / 4 000026 (APE 26) Loss of Component Ability to determine and interpret the following as Cooling Water / 8 they apply to the Loss of Component Cooling Water: (CFR: 43.5 / 45.13)

X 2.5* 76 AA2.05 The normal values for CCW-header flow rate and the flow rates to the components cooled by the CCWS 000027 (APE 27) Pressurizer Pressure 2.2.25 Knowledge of the bases in Technical Control System Malfunction / 3 X Specifications for limiting conditions for operations 4.2 81 and safety limits. (CFR: 41.5 / 41.7 / 43.2) 000029 (EPE 29) Anticipated Transient 2.4.41 Knowledge of the emergency action level Without Scram / 1 X thresholds and classifications. (CFR: 41.10 / 43.5 / 4.6 79 45.11) 000038 (EPE 38) Steam Generator Tube Rupture / 3 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 000058 (APE 58) Loss of DC Power / 6 000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 Rev. 11

ES-401 3 Form ES-401-2 (W E04) LOCA Outside Containment / 3 2.4.30 Knowledge of events related to system operation/status that must be reported to internal X organizations or external agencies, such as the 4.1 77 State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)

(W E11) Loss of Emergency Coolant Ability to determine and interpret the following as Recirculation / 4 they apply to the (Loss of Emergency Coolant Recirculation)

X 4.2 80 EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations. (CFR: 43.5 / 45.13)

(BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 K/A Category Totals: 3 3 Group Point Total: 6 Rev. 11

ES-401 4 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 000003 (APE 3) Dropped Control Rod / 1 Ability to determine and interpret the following as they apply to the Dropped Control Rod: (CFR: 43.5 X 3.7 84

/ 45.13)

AA2.01 Rod position indication to actual rod position 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 2.1.25 Ability to interpret X reference materials, such as 4.2 82 graphs, curves, tables,etc. (CFR:

41.10 / 43.5 / 45.12) 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 2.4.30 Knowledge of events related to system operation/status that must be reported to internal X organizations or external 4.1 85 agencies, such as the State, the NRC, or the transmission system operator (CFR: 41.10 / 43.5 /

45.11) 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms

/7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /

4 000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 (W E13) Steam Generator Overpressure / 4 (W E15) Containment Flooding / 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 (BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip / 4 (BW A05) Emergency Diesel Actuation / 6 (BW A07) Flooding / 8 (BW E03) Inadequate Subcooling Margin / 4 Rev. 11

ES-401 5 Form ES-401-2 (BW E08; W E03) LOCA CooldownDepressurization / 4 Ability to determine and interpret the following as they apply to the (Containment Flooding) (CFR:

43.5 / 45.13)

X 4.2 83 EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals: 2 2 Group Point Total: 4 Rev. 11

ES-401 6 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 Ability to (a) predict the impacts of the Pump following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate X the consequences of those malfunctions or 3.1 88 operations: (CFR: 41.5 / 43.5/ 45.3 / 45/13)

A2.03 Problems associated with RCP motors, including faulty motors and current, and winding and bearing temperature problems 004 (SF1; SF2 CVCS) Chemical and 2.2.38 Knowledge of conditions and limitations Volume Control X in the facility license. (CFR: 41.7 / 41.10 / 43.1 4.5 87

/ 45.13) 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency 2.2.25 Knowledge of the bases in Technical Core Cooling X Specifications for limiting conditions for 4.2 89 operations and safety limits. (CFR: 41.5 / 41.7

/ 43.2) 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling Water 010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 2.2.37 Ability to determine operability and/or X availability of safety related equipment. (CFR: 4.6 90 41.7 / 43.5 / 45.12) 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use X procedures to correct, control, or mitigate the 4.4 86 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.03 Failure of ESF 039 (SF4S MSS) Main and Reheat Steam 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 Rev. 11

ES-401 7 Form ES-401-2 073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals: 2 3 Group Point Total: 5 Rev. 11

ES-401 8 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 2.4.21 Knowledge of the parameters and logic Coolant used to assess the status of safety functions, X such as reactivity control, core cooling and 4.6 91 heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12) 011 (SF2 PZR LCS) Pressurizer Level Control 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Ability to (a) predict the impacts of the following Monitor malfunctions or operations on the ITM system; and (b) based on those predictions, use X procedures to correct, control or mitigate the 4.1 93 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

A2.02 Core damage 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Ability to (a) predict the impacts of the following Generator malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the X consequences of those malfunctions or 3.1* 92 operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

A2.08 Steam dumps are not cycling properly at low load, or stick open at higher load (isolate and use atmospheric reliefs when necessary) 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water Rev. 11

ES-401 9 Form ES-401-2 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals: 2 1 Group Point Total: 3 Rev. 11

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 #

2.1.1 Knowledge of conduct of operations section. (CFR:

4.2 96 41.10 / 43.13)

1. Conduct of 2.1.13 Knowledge of facility requirements for controlling Operations vital/controlled access. (CFR: 41.10 / 43.5 / 45.9 / 3.2 97 45.10)

Subtotal 2 2.2.20 Knowledge of the process for managing troubleshooting 3.8 100 activities. (CFR: 41.10 / 43.5 / 45.13)

2. Equipment Control 2.2.21 Knowledge of pre- and post-maintenance operability 4.1 99 requirements. (CFR: 41.10 / 43.2)

Subtotal 2 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry

3. Radiation 3.8 95 requirements, fuel handling responsibilities, access to Control locked high-radiation areas, aligning filters, etc. (CFR:

41.12 / 43.4 / 45.9 / 45.10)

Subtotal 1 2.4.44 Knowledge of emergency plan protective action 4.4 94 recommendations. (CFR: 41.10 / 41.12 / 43.5 / 45.11)

4. Emergency Procedures/Plan 2.4.25 Knowledge of fire protection procedures. (CFR: 41.10 /

3.7 98 43.5 / 45.13)

Subtotal 2 Tier 3 Point Total 7 Rev. 11

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/1 00025 AA1.03 Q#7 - 00025 AA1.18 was randomly replaced within the same ability due to the inability to write a question with plausible distractors.

1/1 E00009 EK1.01 Q#8 - 00027 AK1.01 was replaced due to overlap with question #31. While maintaining the same k/a, a new topic that was not previously sampled was randomly selected.

1/1 00029 G2.4.31 Q#18 - 00029 G2.4.41 was randomly replaced with a generic K/A in the same group because the selected K/A was at the SRO level.

1/2 00037 G2.2.22 Q#20 00037 AA1.06 was randomly replaced due to the inability to write a question to the original K/A without implying a ruptured SG at the same time and to prevent overlap with question #22 1/2 00033 AA2.11 Q#21 00033 AA2.12 was replaced by selecting an adjacent K/A (one with K/A >2.5) within the same ability due to the inability to write a question with plausible distractors and avoid overlap with the audit exam.

1/2 W/E03 G2.4.18 Q#24 W/E03 G2.2.22 was randomly replaced with a generic K/A as the original K/A was not applicable to the topic.

1/2 W/E 16 EK3.2 Q#27 W/E 16 EK3.4 was replaced within the same topic and knowledge due to the inability to write a question with plausible distractors.

2/1 062 A3.05 Q#29 062 A3.01 was replaced within the same topic and ability due to the inability to write a question with plausible distractor based on available control room indications.

2/1 00004 G2.1.25 Q#46 G2.4.3 was randomly replaced with a generic K/A as the original K/A was not applicable to the topic. Per TS table 3.3.3-1, there are no CVCS PAM Instruments.

2/1 00022 K4.02 Q#49 00022 K3.01 was replaced within the same topic due to the inability to write a question with plausible distractors.

2/1 00008 G2.2.42 Q#50 - 00008 G2.2.40 was randomly replaced with a generic K/A in the same group because the selected k/a was at the SRO level as there are no one hour or less Technical Specifications for the CCW system. .

2/1 00012 A4.04 Q#53 00012 A4.03 was replaced by selecting an adjacent K/A within the same ability to avoid overlap with Question #19 and #21.

2/1 103 K3.02 Q#54 103 A4.01 was randomly replaced due to the inability to write a question with plausible distractor to the original K/A.

While maintaining the same system, a new K/A was selected to maintain an even spread of k/a across the tier.

Rev. 11

ES-401 Record of Rejected K/As Form ES-401-4 3 G2.2.13 Q#69 G2.2.39 was randomly replaced with a generic K/A in the same group (Equipment control) due to the inability to write a Tier 3 question to the original K/A.

3 G2.2.17 Q#70 G2.2.36 was randomly replaced with a generic K/A in the same group (Equipment control) due to the inability to write a Tier 3 question to the original K/A.

3 G2.4.14 Q#75 G2.4.46 was randomly replaced with a generic K/A in the same group (Equipment control) due to the inability to write a Tier 3 question to the original K/A and to remove overlap with Question #73.

SRO 1/1 00011 EA2.01 Q#78 00022 AA2.04 was randomly replaced while maintaining a A2 to maintain a random distribution of topics as question #10 on the RO exam was the same K/A SRO 1/1 00029 G2.4.41 Q#79 00029 G2.4.11 was randomly replaced with a generic K/A in the same group due to the inability to write an SRO question to this K/A as there are no abnormal procedures associated with ATWS, only EOPs. Additionally, this K/A was removed in Rev 11 of NUREG 1021.

SRO 1/1 W/E11 EA2.1 Q#80 W/E11 2.2.42 was randomly replaced within the A2 category (to maintain an even spread) due to the inability to write an SRO question to this K/A as entry level Technical Specification is RO knowledge and would not be immediately addressed during the implementation of Emergency Operating Procedures.

SRO 1/1 00027 G2.2.25 Q#81 00027 AA2.16 was randomly replaced due to the inability to write an SRO question to this K/A. Most, if not all, of the actions in the abnormal procedure are RO level knowledge. To maintain tier and group k/a spread, new k/a was randomly selected from generic K/As.

SRO 1/2 00024 G2.1.25 Q#82 00024 G2.2.39 was replaced as knowledge of less than or equal to one hour T.S. action statement s is RO knowledge. New k/A was a randomly selected generic K/A.

SRO 2/1 W/E03 EA2.1 Q#83 W/E 15 EA2.1 was replaced due to the inability to write a question at the SRO level for this K/A. The procedure is an orange path procedure which is RO knowledge that contains 4 steps. A new safety function was selected while maintaining the same K/A.

SRO 2/1 006 G2.25 Q#89 G2.4.11 was randomly replaced with a generic K/A due to the inability to write an SRO question to this K/A.

Additionally, this K/A was removed in Rev 11 of NUREG 1021.

SRO 2/1 00012 G2.2.37 Q#90 G2.2.44 was randomly replaced with a generic K/A in the same category due to the inability to write an SRO question to this K/A.

Rev. 11

ES-401 Record of Rejected K/As Form ES-401-4 SRO G2.1.1 Q#96, G2.1.4 was replaced as the same K/A was on the RO section of the exam thereby being oversampled. New K/A 3/Generic was randomly selected from same conduct of operations generic section.

SRO G2.4.44 Q#94 G2.4.1 was randomly replaced with a generic K/A in the same category due to the inability to write an SRO question 3/Generic to this K/A.

Rev. 11

ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 Facility: Callaway Date of Examination: 9/11/2017 Examination Level: RO Operating Test Number: 2017-1 Administrative Topic Type Code* Describe activity to be performed (see Note) 2.1.25 (3.9) Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc.

R,M A1.a JPM: Calculate volume of water to transfer between RWST and SFP 2.1.18 (3.6) Ability to make accurate, clear, and concise Conduct of Operations logs, records, status boards, and reports R,D,P*

A1.b JPM: Complete RCS Inventory Balance 2.2.37 (3.6) Ability to determine operability and/or availability Equipment Control of safety related equipment.

S,N A2 JPM: Verify AC Sources 2.3.12 (3.2) Knowledge of radiological safety principles Radiation Control pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation A3 R,M areas, aligning filters, etc.

JPM: Determine Estimated Dose and Make Shielding Recommendation 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 & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1; randomly selected)

  • A JPM from the 2014 exam was randomly selected by placing 4 slips of paper labeled A1.a through A3 in a hardhat. A1.bwas drawn from the hardhat.

NUREG-1021, Revision 11 Page 1 of 2

ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 A1.a This is a MODIFIED, BANK JPM. The parent JPM (A2 rev 3) was last used on an ILT NRC Exam administered at Callaway in 2013. Upon completion of this JPM, the Applicant will have determined 5053 gallons (+/- 10 gallons) of water needed to actuate the SFP high level alarm and the RWST Low Level Alarm will actuate.

A1.b This is a BANK JPM. The parent JPM (A1.b) was last used on the 2014 ILT NRC Exam.

Upon completion of the task the applicant will have identified Total RCS leakage is 0.681 gpm, Identified RCS Leakage is 0.076 gpm, and Unidentified RCS Leakage is 0.605 gpm.

A2 This is a NEW JPM. Upon completion of this JPM, the applicant will have verified Path B is aligned for Offsite Source #1, verified either Path A, Path B, or both Path A & B are aligned for Offsite Source #2, and determined Category 8 Alarm Status is SAT.

A3 This is a MODIFIED, BANK JPM. The parent JPM (A8, Rev 3) was last used on an ILT NRC Exam administered at Callaway in 2013. Upon completion of this JPM, the applicant will have calculated total estimated dose for the work without installing shielding to be 33.3 mrem and with shielding to be 36.6 mrem. The applicant does not recommend shielding be requested.

NUREG-1021, Revision 11 Page 2 of 2

ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 Facility: Callaway Date of Examination: 9/11/2017 Examination Level: SRO Operating Test Number: 2017 - 1 Administrative Topic Type Code* Describe activity to be performed (see Note) 2.1.5 (3.9) Ability to use procedures related to shift staffing, Conduct of Operations such as minimum crew complement, overtime R,M limitations, etc.

A1.a (SRO)

JPM: Determine Shift Staffing (Fatigue Rule) 2.1.37 (4.6) Knowledge of procedures, guidelines, or Conduct of Operations limitations associated with reactivity R,M,P* management.

A1.b (SRO)

JPM: Review ECP Calculation 2.2.17 (3.8) Knowledge of the process for managing Equipment Control maintenance activities during power operations, such as risk assessments, work prioritization, R,N and coordination with the transmission system A2 operator.

JPM: Prioritize Job and Assess Risk 2.3.4 (3.7) Knowledge of radiation exposure limits under Radiation Control normal or emergency conditions R,M A3 JPM: Determine Maximum Allowable Stay Time 2.4.44 (4.4) Knowledge of emergency plan protective action Emergency Procedures/Plan recommendations.

R,N A4 JPM: Determine the Protective Action Recomendation NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1; randomly selected)

  • A JPM from the 2014 exam was randomly selected by placing 5 slips of paper labeled "A1.a" through "A4" in a hardhat. "A1.b" was drawn from the hardhat.

NUREG-1021, Revision 10 Page 1 of 2

ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 A1.a This is a MODIFIED, BANK JPM. The parent JPM (Set 2 G2 A.1.b) was last used on an ILT NRC Exam administered at Callaway in 2011. The applicant is provided a schedule for 5 Senior Reactor Operators and is directed to determine which operators are eligible for an overtime shift.

A1.b This is a MODIFIED JPM. The parent JPM (RSA-1 rev 1) was last used on the 2014 ILT NRC Exam. The applicant is directed to review the ECP performed by Reactor Engineering per OSP-SF-00005, Estimated Critical Position Calculation. The applicant will find errors in the Critical Rod Position, Minimum Rod Height, and Maximum Rod Height.

A2 This is a NEW JPM. The applicant is given a list of current plant status and provided a new Job against 'B' Train ESW UHS Bypass Valve EFHV0066. The applicant is directed to screen and assess the risk of the condition on plant status.

A3 This is a MODIFIED JPM. The parent JPM (RSA-4) was last used on the 2013 ILT NRC Retake Exam. The applicant is directed to determine the maximum allowable stay time for an operator hanging local control tags on the RHR Heat Exchanger Bypass Flow Control Valve EJFCV0618 in support of an upcoming AOV diagnostic test.

A4 This is a NEW JPM. Upon completion of this JPM, the candidate will have recommended Sheltering within a 2 mile radius and 5 miles downwind (sectors B,C,D,E) and marked the PAR Flowchart accordingly.

NUREG-1021, Revision 10 Page 2 of 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0 Facility: __Callaway___________________________ Date of Examination: _9/11/2017___

Exam Level: RO SRO-I SRO-U Operating Test No.: __2017-1_______

Control Room Systems: 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1 004 Chemical and Volume Control System / Loss of D,L,S 1 Shutdown Margin S2 013 Engineered Safety Features Actuation System / A,D,EN,S 2 ESFAS - CSAS failure to auto actuate S3 010 Pressurizer Pressure Control System / Initiate Cold A,D,L,S 3 Overpressure Mitigation with PORV Malfunction 1

S4 003 Reactor Coolant Pump System / Start Reactor Coolant A,L,M,P ,S 4P Pump during RCS Natural Circulation Cooldown S5 026 Containment Spray System / Stroke Time Test of N,S 5 BNHV0004 S6 062 AC Electrical Distribution System / Energize NB Bus D,L,S 6 from AEPS diesel generators S7 073 Process Radiation Monitoring System / Radiation N,S 7 Monitor Source Check S8 029 Containment Purge System / Reinitiate Containment D,L,S 8 Purge following CPIS In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) 1 P1 054 Loss of Main Feedwater / Local TDAFP Start Assuming D,P 4S Loss of AC and DC Power P2 055 Loss of Offsite and Onsite Power / Manually load A,D,E 6 equipment on to an AC Bus P3 008 Component Cooling Water System / Local Bypass and D,R 8 Isolation of CCW to the Seal Water HX

  • 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; in-plant systems and functions may overlap those tested in the control room.

Page 1 of 3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA 1/1/1 (S)imulator Note 1. The JPMs from the 2014 exam were randomly selected by placing 11 slips of paper labeled A through K in a container. "D" and "J" were drawn from the hardhat.

S1 This is a BANK JPM. The JPM (BG-RO-S-001) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. At the completion of this JPM, emergency boration of the RCS will have been established from the boric acid storage tanks through BG HV-8104.

S2 This is an ALTERNATE PATH, BANK JPM. The JPM (EN-RO-S-002) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM , the Operator will have manually stopped all Reactor Coolant Pumps and initiated Containment Spray Actuation Signal (CSAS).

S3 This is an ALTERNATE PATH, BANK JPM. The JPM (BB-RO-S-002A) was used on the 2007 ILT NRC Exam. Upon completion of this JPM, the operator will have armed both Pressurizer PORVs for Cold Overpressure Mitigation and isolated or closed BB PV-455A after it fails open.

S4 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (SIM D) was used on the 2014 ILT NRC Exam. The parent was modified into an alternate path JPM. Upon completion of this JPM, the operator will have started RCP A or B.

S5 This is a NEW JPM. The applicant is directed to perform a stroke time test of BNHV0004, RWST TO CTMT SPRAY PUMP A, using OSP-EN-V001A, Train A Containment Spray Valve Operability. Upon completion of this JPM, the operator will have stroked time tested BNHV0004 with times inside the normal stroke range (45.5 - 61.5 seconds).

Page 2 of 3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0 S6 This is a BANK JPM. The JPM (NB-RO-S-001) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM, the operator will have established AEPS power to NB01.

S7 This is a NEW JPM. Upon completion of this JPM, he operator will have successfully source checked all three channels for GH-RE-10B (105, 108, 109).

S8 This is a BANK JPM. The JPM (GT-RO-S-001) was last used on the 2011 ILT NRC Exam. At the completion of this JPM, the operator will have established containment mini-purge intake and exhaust for both trains.

P1 This is a BANK JPM last used on the 2014 ILT NRC Exam. Upon completion of this JPM, the operator will have started the TDAFP and notified the EC that flow to the Steam Generators is available.

P2 This is an ALTERNATE PATH, BANK JPM. The JPM (EOP-NLO-P-002A) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM the following equipment will be loaded on NB01 bus:

o 480V buses NG01 and NG03 o Battery chargers NK21 and NK 23 o Instrument buses NN01 and NN03 o Control Room emergency lighting P3 This is a BANK JPM. The JPM (EG-NLO-P-001) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM, The operator will successfully isolate Seal Water Heat Exchanger.

Page 3 of 3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0 Facility: __Callaway___________________________ Date of Examination: _9/11/2017___

Exam Level: RO SRO-I SRO-U Operating Test No.: __2017-1_______

Control Room Systems: 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1 004 Chemical and Volume Control System / Loss of D,L,S 1 Shutdown Margin S2 013 Engineered Safety Features Actuation System / A,D,EN,S 2 ESFAS - CSAS failure to auto actuate S3 010 Pressurizer Pressure Control System / Initiate Cold A,D,L,S 3 Overpressure Mitigation with PORV Malfunction 1

S4 003 Reactor Coolant Pump System / Start Reactor Coolant A,L,M,P ,S 4P Pump during RCS Natural Circulation Cooldown S5 026 Containment Spray System / Stroke Time Test of N,S 5 BNHV0004 S6 062 AC Electrical Distribution System / Energize NB Bus D,L,S 6 from AEPS diesel generators S8 029 Containment Purge System / Reinitiate Containment D,L,S 8 Purge following CPIS In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) 1 P1 054 Loss of Main Feedwater / Local TDAFP Start Assuming D,P 4S Loss of AC and DC Power P2 055 Loss of Offsite and Onsite Power / Manually load A,D,E 6 equipment on to an AC Bus P3 008 Component Cooling Water System / Local Bypass and D,R 8 Isolation of CCW to the Seal Water HX

  • 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; in-plant systems and functions may overlap those tested in the control room.

Page 1 of 3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA 1/1/1 (S)imulator Note 1. The JPMs from the 2014 exam were randomly selected by placing 11 slips of paper labeled A through K in a container. "D" and "J" were drawn from the hardhat.

S1 This is a BANK JPM. The JPM (BG-RO-S-001) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. At the completion of this JPM, emergency boration of the RCS will have been established from the boric acid storage tanks through BG HV-8104.

S2 This is an ALTERNATE PATH, BANK JPM. The JPM (EN-RO-S-002) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM , the Operator will have manually stopped all Reactor Coolant Pumps and initiated Containment Spray Actuation Signal (CSAS).

S3 This is an ALTERNATE PATH, BANK JPM. The JPM (BB-RO-S-002A) was used on the 2007 ILT NRC Exam. Upon completion of this JPM, the operator will have armed both Pressurizer PORVs for Cold Overpressure Mitigation and isolated or closed BB PV-455A after it fails open.

S4 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (SIM D) was used on the 2014 ILT NRC Exam. The parent was modified into an alternate path JPM. Upon completion of this JPM, the operator will have started RCP A or B.

S5 This is a NEW JPM. The applicant is directed to perform a stroke time test of BNHV0004, RWST TO CTMT SPRAY PUMP A, using OSP-EN-V001A, Train A Containment Spray Valve Operability. Upon completion of this JPM, the operator will have stroked time tested BNHV0004 with times inside the normal stroke range (45.5 - 61.5 seconds).

Page 2 of 3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0 S6 This is a BANK JPM. The JPM (NB-RO-S-001) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM, the operator will have established AEPS power to NB01.

S8 This is a BANK JPM. The JPM (GT-RO-S-001) was last used on the 2011 ILT NRC Exam. At the completion of this JPM, the operator will have established containment mini-purge intake and exhaust for both trains.

P1 This is a BANK JPM last used on the 2014 ILT NRC Exam. Upon completion of this JPM, the operator will have started the TDAFP and notified the EC that flow to the Steam Generators is available.

P2 This is an ALTERNATE PATH, BANK JPM. The JPM (EOP-NLO-P-002A) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM the following equipment will be loaded on NB01 bus:

o 480V buses NG01 and NG03 o Battery chargers NK21 and NK 23 o Instrument buses NN01 and NN03 o Control Room emergency lighting P3 This is a BANK JPM. The JPM (EG-NLO-P-001) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM, The operator will successfully isolate Seal Water Heat Exchanger.

Page 3 of 3

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0 Facility: __Callaway___________________________ Date of Examination: _9/11/2017___

Exam Level: RO SRO-I SRO-U Operating Test No.: __2017-1_______

Control Room Systems: 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S2 013 Engineered Safety Features Actuation System / A,D,EN,S 2 ESFAS - CSAS failure to auto actuate 1

S4 003 Reactor Coolant Pump System / Start Reactor Coolant A,L,M,P ,S 4P Pump during RCS Natural Circulation Cooldown In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) 1 P1 054 Loss of Main Feedwater / Local TDAFP Start Assuming D,P 4S Loss of AC and DC Power P2 055 Loss of Offsite and Onsite Power / Manually load A,D,E 6 equipment on to an AC Bus P3 008 Component Cooling Water System / Local Bypass and D,R 8 Isolation of CCW to the Seal Water HX

  • 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; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA 1/1/1 (S)imulator Note 1. The JPMs from the 2014 exam were randomly selected by placing 11 slips of paper labeled A through K in a container. "D" and "J" were drawn from the hardhat.

Page 1 of 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Rev 0 S2 This is an ALTERNATE PATH, BANK JPM. The JPM (EN-RO-S-002) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM , the Operator will have manually stopped all Reactor Coolant Pumps and initiated Containment Spray Actuation Signal (CSAS).

S4 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (SIM D) was used on the 2014 ILT NRC Exam. The parent was modified into an alternate path JPM. Upon completion of this JPM, the operator will have started RCP A or B.

P1 This is a BANK JPM last used on the 2014 ILT NRC Exam. Upon completion of this JPM, the operator will have started the TDAFP and notified the EC that flow to the Steam Generators is available.

P2 This is an ALTERNATE PATH, BANK JPM. The JPM (EOP-NLO-P-002A) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM the following equipment will be loaded on NB01 bus:

o 480V buses NG01 and NG03 o Battery chargers NK21 and NK 23 o Instrument buses NN01 and NN03 o Control Room emergency lighting P3 This is a BANK JPM. The JPM (EG-NLO-P-001) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2016. Upon completion of this JPM, The operator will successfully isolate Seal Water Heat Exchanger.

Page 2 of 2

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Callaway Date of Exam: 9/11/2017 Operating Test No. : 2017-1 A E Scenarios: Team 1: S1, S2, S3 P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION A I

M C S A B S A B S A B S A B L U A T R T O R T O R T O R T O M(*)

N Y O C P O C P O C P O C P R I U T P E S1 S2 S3 S2 S3 S1 S3 S1 S2 SRO-I RX 4 1 1 1 0 S1 NOR 0* 1 1 1 I/C 1,2,3 3,4 2,4 7 4 4 2 MAJ 5 6 5 3 2 2 1 TS 1,2 2 0 2 2 SRO-I RX 4 1 1 1 0 S2 NOR 1 1 1 1 1 I/C 1,2 2,3,4,5 1,2,3 9 4 4 2 MAJ 5 6 5 3 2 2 1 TS 2,4,5 3 0 2 2 SRO-I RX 4 1 1 1 0 S3 NOR 1 1 1 1 1 I/C 1,3 2,5 1,2,3,4 8 4 4 2 MAJ 5 6 5 3 2 2 1 TS 1,2 2 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs 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 Instant SRO 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 instrument or component malfunctions on a 1-for-1 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.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Callaway Date of Exam: 9/11/2017 Operating Test No. : 2017-1 A E Scenarios: Team 2: S4, S5, R1 P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION A I

M C S A B S A B S A B S A B L U A T R T O R T O R T O R T O M(*)

N Y O C P O C P O C P O C P R I U T P E S4 S5 R1 S5 R1 S4 S5 S4 R1 SRO-I RX 4 1 1 1 0 S4 NOR 0* 1 1 1 I/C 1,2,3 3,4 2,4 7 4 4 2 MAJ 5 6 5 3 2 2 1 TS 1,2 2 0 2 2 SRO-I RX 4 1 1 1 0 S5 NOR 1 1 1 1 1 I/C 1,2 2,3,4,5 1,2,3,4 10 4 4 2 MAJ 5 6 5 3 2 2 1 TS 2,4,5 1,2 5 0 2 2 RO RX 4 1 1 1 0 R1 NOR 1 1 1 1 1 I/C 1,3 2,5 1,2,3 7 4 4 2 MAJ 5 6 5 3 2 2 1 TS 0 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs 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 Instant SRO 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 instrument or component malfunctions on a 1-for-1 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.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Callaway Date of Exam: 9/11/2017 Operating Test No. : 2017-1 A E Scenarios: Team 3: S6, U1, Surrogate BOP P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION A I

M C S A B S A B S A B S A B L U A T R T O R T O R T O R T O M(*)

N Y O C P O C P O C P O C P R I U T P E S6 U1 U1 S6 SRO-I RX 0* 1 1 0 S6 NOR 1 1 1 1 1 I/C 2,3,4,5 2,4 6 4 4 2 MAJ 6 5 2 2 2 1 TS 2,4,5 3 0 2 2 SRO-U RX 0* 1 1 0 U1 NOR 1 1 1 1 1 I/C 2,5 1,2,3,4 6 4 4 2 MAJ 6 5 2 2 2 1 TS 1,2 2 0 2 2 RX 1 1 0 NOR 1 1 1 I/C 4 4 2 MAJ 2 2 1 TS 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs 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 Instant SRO 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 instrument or component malfunctions on a 1-for-1 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.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Callaway Date of Exam: 9/11/2017 Operating Test No. : 2017-1 A E Scenarios: Spare P V 1 2 3 4 T M P E O I L N CREW POSITION CREW CREW CREW T N I T POSITION POSITION POSITION A I

M C S A B S A B S A B S A B L U A T R T O R T O R T O R T O M(*)

N Y O C P O C P O C P O C P R I U T P E

RX 4 1 1 0 NOR 1 1 1 I/C 2,3 4 4 2 MAJ 5 2 2 1 TS 3,4 0 2 2 RX 4 1 1 0 NOR 1 1 1 I/C 1,2,3 4 4 2 MAJ 5 2 2 1 TS 0 2 2 RX 4 1 1 0 NOR 1 1 1 I/C 2,3 4 4 2 MAJ 5 2 2 1 TS 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs 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 Instant SRO 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 instrument or component malfunctions on a 1-for-1 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.

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.:1 , Rev 0 Op-Test No.: 2017-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100%

Turnover: No equipment out of service.

Event Malf. No. Event Event No. Type* Description Trip of the Normal Charging Pump.

1 PBG04 RO (C)

OTO-BG-00001, Pressurizer Level Control Malfunction SRO (C)

Condensate Pump 'A' Trips.

2 PAD01A RO (C)

OTO-AE-00001, Feedwater System Malfunction BOP (C)

Power Range Channel N43 fails high.

SRO (I)

OTO-SF-00001, Rod Control Malfunctions 3 SEN0043 RO (I)

OTO-SE-00001, Nuclear Instrument Malfunction BOP (I)

(Tech Spec 3.3.1)

SRO (R) Tube Leak in Steam Generator 'D' 4 EBB01D RO (R) OTO-BB-00001, Steam Generator Tube Leak BOP (R) (Tech Spec 3.4.13)

SRO (M)

Tube Rupture in Steam Generator 'D' 5 EBB01D RO (M)

E-3, Steam Generator Tube Rupture BOP (M)

Safety Injection Pump 'A' fails to start.

6 PEM01A_2 RO (C) E-0, Reactor Trip or Safety Injection, Attachment A, Automatic Action Verification

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
1. Total malfunctions (5-8) 6
2. Malfunctions after EOP entry (1-2) 1
3. Abnormal events (2-4) 4
4. Major transients (1-2) 1
5. EOPs entered/requiring substantive actions (1-2) 1
6. EOP contingencies requiring substantive actions (0-2) 0
7. Critical tasks (2-3) 2 Page 1 of 4

Scenario #1 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 The plant is stable at 100% with no equipment out of service.

After the reactivity brief is complete the Normal Charging Pump trips. The RO should perform the Immediate Actions of OTO-BG-00001, Pressurizer Level Control Malfunction, to start a CCP.

After immediate action verification is complete, Condensate Pump 'A' Trips. The crew should respond per OTO-AE-00001, Feedwater System Malfunction, by reducing power to less than 97% per Attachment A, Load Reduction.

Once power is stabilized, Power Range Nuclear Instrument Channel N43 fails high causing an automatic rod insertion. The crew should respond to the rod insertion by performing the immediate actions of OTO-SF-00001, Rod Control Malfunctions. Following verification of immediate actions the crew will enter OTO-SE-00001, Nuclear Instrument Malfunction, to bypass channel N43 and restore control rods to desired position. Tech Spec 3.3.1 applies.

Once the crew has initiated restoration of control rod positions, a 20 gpm tube leak develops in Steam Generator 'D'. The crew should enter OTO-BB-00001, Steam Generator Tube Leak, and quantify the leak to be greater than 150 gpd. The crew should initiate a load reduction to below 50% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using OTO-MA-00008, Rapid Load Reduction. Tech Spec 3.4.13 applies.

Once the crew has reduced power below 90%, the tube leak becomes a rupture. The crew will trip the reactor and actuate safety injection and enter E-0, Reactor Trip or Safety Injection.

Safety Injection pump 'A' fails to start on the LOCA sequencer. The crew should start the pump per E-0 Attachment A, Automatic Action Verification. The crew should also monitor Steam Generator 'D' Level and isolate AFW flow to the ruptured Steam Generator when level rises above 7% NR per the Foldout Page Actions of E-0.

The crew should transition to E-3, Steam Generator Tube Rupture, and isolate the ruptured SG.

Once SG 'D' is isolated, the crew should initiate a cooldown at maximum rate to below the target temperature for maintaining RCS subcooling following RCS depressurization and SI termination.

The scenario is complete when the crew has completed the RCS cooldown.

Page 2 of 4

Scenario #1 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 Critical Tasks:

Critical Tasks Isolate feedwater flow into and steam flow from the ruptured SG before a transition to Establish/maintain an RCS temperature so that transition from E-3 does not occur ECA-3.1 occurs. because the RCS temperature is in either of the following conditions:

  • Too high to maintain minimum required subcooling OR
  • Below the RCS temperature that causes an extreme (RED path) or a severe (ORANGE path) challenge to the subcriticality and/or the integrity CSF EVENT 5 5 Safety Isolating the ruptured SG maintains a differential pressure between the ruptured SG and Failure to establish and maintain the correct RCS temperature during a SGTR leads to a significance the intact SGs. The differential pressure (250 psi) ensures that minimum RCS transition from E-3 to a contingency ERG. This failure constitutes an incorrect subcooling remains after RCS depressurization. performance that necessitates the crew taking compensating action that would complicate the event mitigation strategy....

Cueing All of the following: All of the following:

  • Indication and/or annunciation of SGTR in one SG
  • Indication and/or annunciation of SGTR in one SG o Increasing SG water level o Increasing SG water level o Radiation o Radiation
  • Indication and/or annunciation of SI
  • Indication and/or annunciation of SI
  • Indication of ruptured SG pressure greater than minimum required pressure Performance Manipulation of controls as required to isolate the ruptured SG Manipulation of controls as required to establish and maintain RCS temperature indicator
  • Close Steam line low point drain valve from ruptured SG
  • Steam dump valve position lamps and/or indicators indicate closed o AB HIS-10 (SG D)
  • SG PORV valve position lamps and/or indicators indicate closed
  • Stop feed flow to ruptured SG o CLOSE AL HK-5A and AL HK-6A Performance Crew will observe the following: Indication of steam flow rate greater than zero feedback
  • Indication of stable or increasing pressure in the ruptured SG
  • Indication of RCS temperature decreasing
  • Indication of decreasing or zero feedwater flow rate in the ruptured SG OR
  • Indication of RCS temperature less than target value Justification for When the crew cannot maintain the 250 psi differential, the ERGs require a transition to Terminating the RCS cooldown before reaching the target temperature prevents the chosen contingency ERG ECA-3.1. This transition unnecessarily delays the sequence of actions achieving the minimum RCS subcooling. Failure to achieve the required RCS performance limit leading to RCS depressurization and Sl termination. subcooling results in a condition that forces the crew to transition to contingency ERG ECA-3.1, thereby delaying the RCS depressurization and SI termination. Such a delay allows the excessive inventory increase of the ruptured SG to continue until the SG overpressure components release water or until SG overfill occurs.

Terminating the cooldown too late challenges either the subcriticality CSF or the integrity CSF. Because the crew is directed to cool down at the maximum rate, late termination of cooldown could force the RCS temperature low enough to challenge the integrity CSF. The crew must then transition to one of the integrity FRGs. The transition also delays RCS depressurization and SI termination.

PWR Owners CT-18, Isolate the Ruptured SG CT-19, Control initial RCS cooldown Group Appendix Page 3 of 4

Scenario #1 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 References OTO-BG-00001, Pressurizer Level Control Malfunction OTO-AE-00001, Feedwater System Malfunction OTO-SF-00001, Rod Control Malfunction OTO-SE-00001, Nuclear Instrument Malfunction OTO-BB-00001, Steam Generator Tube Leak OTO-MA-00008, Rapid Load Reduction E-0, Reactor Trip or Safety Injection E-3, Steam Generator Tube Rupture Tech Spec 3.3.1 Tech Spec 3.4.13 ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions

1. SG Tube Rupture Page 4 of 4

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 2 , Rev 0 Op-Test No.: 2017-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100%

Turnover: Service Water Pump 'A' is out of service for maintenance.

Event Malf. No. Event Event No. Type* Description SRO (C) Battery Charger NK21 loss 1 NG0103 RO (C) OTO-NK-00001, Failure of NK Battery Charger BOP (C) (Tech Spec 3.8.4)

Pressurizer Level Channel BB LI-461 fails high SRO (I) 2 BBLT0461 OTO-BG-00001, Pressurizer Level Control Malfunction RO (I)

(Tech Spec 3.3.1)

SRO (C) Service Water Pump 'B' trips 3 PEA2101B BOP (C) OTA-RK-00014 Addendum 12A, Service Water Pump Lockout SRO (R)

Turbine Vibration requiring load reduction 4 ACYE0017 RO (R)

OTO-AC-00002, Turbine Vibration BOP (R)

SRO (M)

AE002 Feedwater break in containment with auto reactor trip failure 5 RO (M)

SF006 E-2, Faulted Steam Generator Isolation BOP (M)

FWIV 'B' Failure to close 6 AEFV40 BOP (C)

E-0, Reactor Trip or Safety Injection

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
1. Total malfunctions (5-8) 6
2. Malfunctions after EOP entry (1-2) 1
3. Abnormal events (2-4) 4
4. Major transients (1-2) 1
5. EOPs entered/requiring substantive actions (1-2) 1
6. EOP contingencies requiring substantive actions (0-2) 0
7. Critical tasks (2-3) 2 Page 1 of 4

Scenario #2 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 The plant is stable at 100% with Service Water Pump 'A' out of service.

After the reactivity brief is complete, battery charger NK21 is lost. The crew should use OTO-NK-00002, Failure of NK Battery Charger, to select away from NN01 powered instrumentation and to direct placing the swing charger NK25 into service in the field. Tech Spec 3.8.4 applies.

Once Tech Specs have been determined pressurizer level channel BB LI-461 fails high. The crew should enter OTO-BG-00001, Pressurizer Level Control Malfunction, to restore pressurizer level to program in manual, select away, and restore automatic control. Tech Spec 3.3.1 applies.

Service Water Pump 'B' Trips leaving the crew with 1 pump supplying all plant loads. The crew should perform the OTA-RK-00014 Addendum 12, Service Water Pump Lockout, actions and start both trains of Essential Service Water using OTN-EF-00001, Essential Service Water System.

Once both trains of ESW are in manual, a turbine vibration develops. The crew should enter OTO-AC-00001, Turbine Vibration, and perform a load reduction while monitoring vibration levels.

Once the crew has reduced load 5-10% the 'B' Loop feedwater line breaks inside containment with no automatic reactor trip. The crew should manually trip the reactor and enter E-0, Reactor Trip or Safety Injection.

The crew should close the 'B' FWIV and perform foldout page criteria for Faulted SG Isolation Criteria to Fast Close MSIVs and stop auxiliary feedwater flow to the 'B' SG. The crew should transition to E-2, Faulted Steam Generator Isolation, and complete isolation of the 'B' SG.

The scenario is complete when the crew has transitioned out of E-2 to ES-1.1, SI Termination.

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Scenario #2 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 Critical Tasks:

Critical Tasks Manually trip the reactor before 'A', 'C' and 'D' SG Levels lower to 10% WR (dryout). Isolate faulted SG 'B' before transition out of E-2 EVENT 5 6&7 Safety Failure to manually trip the reactor causes a challenge to the subcriticality CSF beyond Failure to isolate a faulted SG that can be isolated causes challenges to CSFs (Integrity, significance that irreparably introduced by the postulated conditions. Additionally, it constitutes an Subcriticality, & Containment) beyond those irreparably introduced by the postulated incorrect performance that necessitates the crew taking compensating action that conditions.

would complicate the event mitigation strategy and demonstrates the inability of the crew to recognize a failure or an incorrect automatic actuation of an ESF system or component.

Cueing Indication and/or annunciation that plant parameter(s) exist that should result in Both of the following:

automatic reactor trip but reactor does not automatically trip Steam pressure and flow rate indications that make it possible to identify 'B' SG as

  • Annunciator 85A, SG LEV LOLO RX TRIP faulted AND Valve position and flow rate indication that AFW continues to be delivered to the 'B' SG Performance Manipulation of control room reactor trip switches (SB HS-1 or SB HS-42) as required to CLOSE AL HK-9A & AL HK-10A to isolate AFW flow to faulted SG indicator trip the reactor. Close SG B low point drain valve AB HIS-9
  • Reactor trip and bypass breakers indicate open Fast Close all MSIVs and Bypass Valves using AB HS-79 or AB HS-80 Close FWIV AE-HS-40 Performance Indications of reactor trip Indication of the following feedback
  • Any depressurization of intact SGs stops
  • Neutron flux decreasing
  • AFW flow rate indication to faulted SG of zero Justification for Not tripping the reactor prior to SG reaching dryout conditions when it is possible to do before transition out of E-2 is in accordance with the PWR Owners Group Emergency the chosen so forces an immediate extreme challenge to the subcriticality CSF, availability of the Response Guidelines. It allows enough time for the crew to take the correct action while performance limit heat sink, and containment. Additionally, the incorrect performance of failing to trip the at the same time preventing avoidable adverse consequences.

reactor necessitates the crew taking compensating action that seriously complicates the event mitigation strategy. This misoperation constitutes a significant reduction of safety margin beyond that irreparably introduced by the scenario.

PWR Owners CT-1, Manually Trip the Reactor CT-17, Isolate Faulted SG Group Appendix Page 3 of 4

Scenario #2 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 References OTO-NK-00002, Failure of NK Battery Charger OTO-BG-00001, Pressurizer Level Control Malfunction OTA-RK-00014 Addendum 12, Service Water Pump Lockout OTN-EF-00001, Essential Service Water System OTO-AC-00001, Turbine Vibration E-0, Reactor Trip or Safety Injection E-2, Faulted Steam Generator Isolation ES-1.1, SI Termination PRA Systems, Events or Operator Actions

1. ATWS (12% of contribution to Core Damage Frequency)
2. Steam Line Breaks (5% of contribution to Core Damage Frequency)

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Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 3 , Rev 0 Op-Test No.: 2017-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: Mode 3 with shutdown banks withdrawn Turnover: No equipment out of service. The crew is directed to dilute the RCS to desired ECP boron concentration.

Event Malf. No. Event Event No. Type* Description Dilute RCS to ECP boron concentration SRO (N) 1 NA OTN-BG-00001, Reactor Makeup Control and Boron Thermal RO (N)

Regeneration system Source Range Channel N31 fails high to 700 cps SRO (I) 2 SEN0031 OTO-SE-00001, Nuclear Instrument Malfunction RO (I)

Tech Specs 3.3.1 and 3.3.9 SFP Cooling Pump 'B' Trips SRO (C) 3 PEC01B OTA-RK-00022 Addendum 76E, Spent Fuel Pool Cooling BOP (C)

Pump B Trip Atmospheric Steam Dump 'D' fails open with manual control SRO (C) 4 ABPV0004 OTO-AB-00001, Steam Dump Malfunction BOP (C)

Tech Spec 3.7.4 A 25 gpm RCS leak develops.

SRO (C) 5 BB001_C OTO-BB-00003, RCS Excessive Leakage RO (C)

Tech Spec 3.4.13 SRO (M)

Small break LOCA 6 BB001_C RO (M)

E-1, Loss of Reactor or Secondary Coolant BOP (M)

EFHV0037 fails to open on Safety Injection 7 EFHV0037 RO (C)

E-0 Attachment A, Automatic Action Verification

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
1. Total malfunctions (5-8) 6
2. Malfunctions after EOP entry (1-2) 1
3. Abnormal events (2-4) 4
4. Major transients (1-2) 1
5. EOPs entered/requiring substantive actions (1-2) 1
6. EOP contingencies requiring substantive actions (0-2) 0
7. Critical tasks (2-3) 2 Page 1 of 4

Scenario #3 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 The plant is stable in Mode 3 with shutdown banks withdrawn. The crew has been directed to dilute the RCS to the desired Estimated Critical Position boron concentration per step 5.3.10.a of OTG-ZZ-00001, Plant Heatup Cold Shutdown to Hot Standby, using OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System. The crew will begin a RCS dilution that will last 25 minutes.

Once the dilution has been initiated, a failure of Source Range Channel N31 high occurs. The failure will cause a charging pump suction swapover from the VCT to the RWST on flux doubling and a high flux at shutdown alarm. The crew should respond per OTO-SE-00001, Nuclear Instrument Malfunction, to stop the RCS dilution evolution and restore the charging pump suction to the VCT. Tech Specs 3.3.1 and 3.3.9 apply.

Once Tech Specs have been determined, SFP Cooling Pump 'B' trips. The crew should respond per OTA-RK-00022 Addendum 76E, Spent Fuel Pool Cooling Pump B Trip. The crew should start a Train 'A' CCW pump and start the Train 'A' SFP Cooling Pump using OTN-EC-00001, Fuel Pool Cooling and Cleanup System.

Once Train 'A' SFP Cooling is in-service, the 'D' Atmospheric Steam Dump fails open. The BOP operator should close the dump valve using manual control. The crew should enter OTO-AB-00001, Steam Dump Malfunction. Tech Spec 3.7.4 applies.

A 25 gpm RCS leak develops. The crew should enter OTO-BB-00003, RCS Excessive Leakage, to quantify the leakage. Once letdown is isolated, the leak becomes a Small Break LOCA. The crew should take action per foldout page to trip the reactor and safety inject when it is determined that pressurizer level can not be maintained. Tech Spec 3.4.13 applies.

When SI actuates, a NB02 bus fault occurs causing a loss of Train 'B' equipment. In addition, ESW return valve EF HV-37 fails to open on the LOCA sequencer causing a loss of containment cooling. The crew will enter E-0, Reactor Trip or Safety Injection, and should open EF HV-37 when performing Attachment A. Also, the crew will need to secure RCPs when RCP Trip Criteria is met.

The scenario is complete when the crew has transitioned out of E-1, Loss of Reactor or Secondary Coolant, and then to ES-1.2, Post LOCA Cooldown and Depressurization.

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Scenario #3 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 Critical Tasks:

Critical Tasks Manually actuate containment cooling by establishing 'A' Train ESW flow by opening Trip all RCPs within 5 minutes of meeting RCP trip criteria.

EFHV0037 prior to completion of Attachment A of E-0.

EVENT 7 6 Safety If one train of containment cooling is not actuated, the FSAR assumptions and results Failure to trip the RCPs under the postulated plant conditions leads to core uncovery significance are invalid. Because compliance with the assumptions of the FSAR is part of the facility and to fuel cladding temperatures in excess of 2200°F, which is the limit specified in the license condition, failure to manually actuate at least one train of containment cooling ECCS acceptance criteria. Thus, failure to perform the task represents misoperation or under the scenario conditions and when it is possible to do so constitutes a violation of incorrect crew performance in which the crew has failed to prevent degradation of...{the the license condition. fuel cladding} ...barrier to fission product release and which leads to violation of the facility license condition.

Cueing Indications of all the following:

Indication and/or annunciation that containment cooling is required

  • SBLOCA Indication and/or annunciation that the minimum required complement of containment
  • Safety Injection cooling equipment is not entirely available.
  • Only one train of safety injection pumps actuated RCP Trip Criteria are met Performance Opening EFHV0037 Manipulation of controls as required to trip all RCPs indicator
  • RCP breaker position lights indicate breaker open Performance EFHV0037 open light and Flow established to UHS as indicated by EF FI-53 Indication that all RCPs are stopped feedback
  • RCP breaker position lights
  • RCP flow decreasing
  • RCP motor amps decreasing Justification for before completion of Attachment A of E-0 is in accordance with the PWR Owners In a letter to the NRC titled Justification of the Manual RCP Trip for Small Break LOCA the chosen Group Emergency Response Guidelines. It allows enough time for the crew to take the Events (OG-117, March 1984) (also known as the Sheppard letter), the WOG provided performance limit correct action while at the same time preventing avoidable adverse consequences. the required assurance based on the results of the analyses performed in conjunction with WCAP-9584. The WOG showed that for all Westinghouse plants, more than two minutes were available between onset of the trip criteria and depletion of RCS inventory to the critical inventory. In fact, additional analyses sponsored by the WOG in connection with OG-117 conservatively showed that manual RCP trip could be delayed for five minutes beyond the onset of the RCP trip criteria without incurring any adverse consequence.

PWR Owners CT-3, Manually actuate containment cooling CT-16, Manually Trip RCPs Group Appendix Page 3 of 4

Scenario #3 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 References OTG-ZZ-00001, Plant Heatup Cold Shutdown to Hot Standby OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System OTO-SE-00001, Nuclear Instrument Malfunction OTA-RK-00022 Addendum 76E, Spent Fuel Pool Cooling Pump B Trip OTN-EC-00001, Fuel Pool Cooling and Cleanup System OTO-AB-00001, Steam Dump Malfunction OTO-BB-00003, RCS Excessive Leakage E-0, Reactor Trip or Safety Injection E-1, Loss of Reactor or Secondary Coolant ES-1.2, Post LOCA Cooldown and Depressurization EOP Addendum 26, SIS Status Panel Alignment PRA Systems, Events or Operator Actions

1. Small Break LOCA (36% of contribution to CDF)

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Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 4 , Rev 0 Op-Test No.: 2017-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: The plant is stable at 80%.

Turnover: No equipment out of service Event Malf. No. Event Event No. Type* Description Steam pressure channel fails low SRO (I) OTO-AE-00002, Steam Generator Water Level Control 1 ABPT0535 BOP (I) Malfunctions Tech Spec 3.3.2 Turbine impulse pressure channel AC PT-505 fails low SRO (I)

OTO-SF-00001, Rod Control Malfunctions 2 ACPT0505 RO (I)

OTO-AC-00003, Turbine Impulse Pressure Channel Failure BOP (I)

Tech Spec 3.3.1 SRO (C) CCW flow to containment isolated by EG HV-71 closure 3 EGHV0071 BOP (C) OTO-EG-00001, CCW System Malfunction BBPCV045 SRO (C) Loop 1 spray valve fails 50% open 4

5B RO (C) OTO-BB-00006, Pressurizer Pressure Control Malfunction SRO (M)

Loss of Off-Site Power and plant trip 5 LOOP RO (M)

E-0, Reactor Trip or Safety Injection BOP (M) 6 PAL02_1 BOP (C) TDAFP fails to auto-start. Total loss of AFW flow.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
1. Total malfunctions (5-8) 6
2. Malfunctions after EOP entry (1-2) 1
3. Abnormal events (2-4) 4
4. Major transients (1-2) 1
5. EOPs entered/requiring substantive actions (1-2) 1
6. EOP contingencies requiring substantive actions (0-2) 1
7. Critical tasks (2-3) 2 Page 1 of 4

Scenario #4 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 The plant is stable at 80% power with no equipment out of service.

Following the reactivity brief steam pressure channel AB PT-535 fails low. The crew should take manual control of the 'C' MFRV and restore 'C' Steam Generator Water Level per OTO-AE-00002, Steam Generator Water Level Control Malfunctions. They should select away and restore control to automatic. Tech Spec 3.3.2 applies.

Once Tech Specs have been determined and level control is in automatic, a failure of turbine impulse pressure AC PT-505 occurs causing control rods to insert into the core. The crew should utilize the immediate actions of OTO-SF-00001, Rod Control Malfunctions, to place rods in manual , and then enter OTO-AC-00003, Turbine Impulse Pressure Channel Failure, to select away and restore rods at or near the original position and restore rod control to automatic. Tech Spec 3.3.1 applies.

CCW flow is lost to containment when EG HV-71 fails closed. The crew should enter OTO-EG-00001, CCW System Malfunction, and re-establish CCW flow to containment by energizing and opening bypass valve EG HV-126. This action must be accomplished within 10 minutes to prevent damage to RCPs.

Once CCW flow is re-established to containment, the loop 1 pressurizer spray valve BB PCV-455B fails 50% open. The crew should take manual control of the loop 1 spray valve and close it using OTO-BB-00006, Pressurizer Pressure Control Malfunction.

When pressurizer pressure has stabilized, a reactor trip occurs from a combination of partial loss of off-site power and turbine trip. The crew will respond per E-0, Reactor Trip or Safety Injection, and transition to ES-0.1, Reactor Trip Response.

A complete loss of auxiliary feedwater flow occurs due to motor driven AFW pump discharge valve malfunctions, a loss of NB02, and the turbine driven AFW pump fails to auto-start. The crew should transition to FR-H.1, Response to Loss of Secondary Heat Sink, to start the turbine driven AFW pump.

The loss of NB02 is due to feeder breaker NB0209 failure to open preventing the B EDG from energizing the bus. NB01 remains energized for 5 minutes post trip, and then is lost when the final off-site line de-energizes. The crew should enter ECA-0.0, Loss of all AC Power, and start the 'A' EDG to re-energize NB01. The crew should then transition back to procedure and step in effect.

The scenario is complete when the crew has transitioned out of ECA-0.0 and established a minimum required feedwater flow rate (>285,000lbm/hr).

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Scenario #4 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 Critical Tasks:

Critical Tasks Restore one AC power source to NB01 within 15 minutes of NB01 power loss. Establish 285,000 lbm/hr to the SGs before SG level indicates less than 10%.

EVENT 5 6 Safety In the scenario, failure to energize at least one ac emergency bus results in the Failure to establish the minimum required feedwater flow rate, under the postulated significance needless continuation of a situation in which the pumped ECCS capacity and the plant conditions, results in adverse consequences or significant degradation in the emergency power capacity are both in a completely degraded status, as are all other mitigative capability of the plant. In this case, the minimum required feedwater flow rate active safeguards requiring electrical power. Although the completely degraded status is can be established by performing the appropriate manual action.

not due to the crew's action (was not initiated by operator error), continuation in the completely degraded status is a result of the crew's failure to energize at least one ac emergency bus.

Cueing Indication and/or annunciation that all ac emergency buses are de-energized Indication and/or annunciation of the following:

  • Circuit breaker position Secondary Heat Sink is required
  • Bus voltage
  • Total feedwater flow rate indicates less than the minimum required
  • Total AFW flow rate indicates less than the minimum required
  • AFW valve position indication that a flow path is not established to at least one SG Performance Manipulation of controls as required to energize at least one ac emergency bus from Manipulation of controls in the control room as required toestablish the minimum indicator NE01: required feedwater flow rate to the SGs:
  • Press START/RESET on KJ HS-8A
  • Press OPEN on FC HIS-312A Performance Indication that NB02 is energized
  • Indication that at least the minimum required feedwater flow rate is being feedback
  • NB02 bus energized light delivered to the SGs
  • NB02 bus voltage
  • Indication of increasing SG levels Justification for Failure to perform the critical task would result in an unnecessary Emergency Action Because the secondary heat sink is required but not satisfactorily provided, the RCS the chosen Level declaration of a Site Area Emergency. Failure to the perform the critical task also heats up. If feedwater flow rate commensurate with core decay heat is not established, performance limit results in needless degradation of any barrier to fission product release, specifically of the heat sink CSF is eventually challenged. With continued insufficient feedwater flow, the RCS barrier at the point of the RCP seals. the SGs dry out, causing an RCS pressure increase that opens the pressurizer PORVs.

The open PORVs create a small-break LOCA that eventually challenges the core cooling CSF. Ultimately, the fuel matrix/clad (a fission-product barrier) is challenged.

PWR Owners CT-24, Energize at least one ac emergency bus CT-45, Establish minimum required feedwater flow rate to SGs before SG dryout Group Appendix Page 3 of 4

Scenario #4 Event Description Callaway 2017-1 NRC ES-D-1, rev. 0 References OTO-AE-00002, Steam Generator Water Level Control Malfunctions OTO-SF-00001, Rod Control Malfunctions OTO-AC-00003, Turbine Impulse Pressure Channel Failure OTO-EG-00001, CCW System Malfunction OTO-BB-00006, Pressurizer Pressure Control Malfunction E-0, Reactor Trip or Safety Injection ES-0.1, Reactor Trip Response ECA-0.0, Loss of all AC Power FR-H.1, Response to Loss of Secondary Heat Sink PRA Systems, Events or Operator Actions Station Blackout (contribution to CDF is 9%).

Scenario addresses failures in 3 of the top 10 risk important s ystems: AFW (3), EDGs (5), CCW (8)

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