ML16158A449

Callaway-2016-06 Draft Outlines

ML16158A449
Person / Time
Site:	Callaway
Issue date:	06/01/2016
From:	Vincent Gaddy Operations Branch IV
To:
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Download: ML16158A449 (57)

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ES-401 PWR Examination Outline Form ES-401-2 Facility: Callaway Plant (RO, Rev. 0) Date of Exam:

2016 Tier Group RO K/A Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total 1. Emergency & Abnormal Plant Evolutions 1 3 3 3 N/A 3 3 N/A 3 18 6 2 1 1 2 1 2 2 9 4 Tier Totals 4 4 5 4 5 5 27 10 2. Plant Systems 1 3 2 3 3 2 3 3 2 3 2 2 28 5 2 1 1 1 1 1 1 1 1 1 1 10 3 Tier Totals 4 3 4 4 3 4 4 3 4 3 2 38 8 3. Generic Knowledge and Abilities Categories 1 3 2 2 3 2 4 3 10 1 2 3 4 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO

-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category). 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO

-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site

-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES

-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant

-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO

-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES

-401 for the applicable K/As. 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO

-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO

-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 1 (RO / SRO) E/APE # / Name

/ Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization

- Recovery / 1 X EK2.02 - Knowledge of the interrelations between a reactor trip

and the following:

Breakers, relays and disconnects (CFR 41.7/45.7) 2.6 1 000008 Pressurizer Vapor Space Accident / 3 X 2.4.11 - Knowledge of abnormal condition procedures.

(CFR: 41.10/43.5/45.13) 4.0 1 000009 Small Break LOCA / 3 X EK1.02 - Knowledge of the operational implications of the following concepts as they apply to the small break LOCA:

Use of steam tables (CFR 41.8/41.10/45.3) 3.5 1 000011 Large Break LOCA / 3 X EK3.02 - Knowledge of the reasons for the following responses as the apply to the Large Break LOCA

Feedwater isolation (CFR 41.5/41.10/45.6/45.13) 3.5 1 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 X AK3.01 - Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System: Shift to alternate flowpath (CFR 41.5,41.10/45.6/45.13) 3.1 1 000026 Loss of Component Cooling Water / 8 X AA2.06 - Ability to determine and interpret the following as they apply

to the Loss of Component Cooling Water: The length of time after the loss of CCW flow to a component

before that component may be damaged (CFR: 43.5/45.13) 2.8 1 000027 Pressurizer Pressure Control System Malfunction / 3 X AK1.01 - Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions:

Definition of saturation temperature (CFR 41.8 / 41.10 / 45.3) 3.1 1 000029 ATWS / 1 X EK2.06 - Knowledge of the interrelations between the and the following an ATWS:

Breakers, relays, and disconnects (CFR 41.7 / 45.7) 2.9 1 000038 Steam Gen. Tube Rupture / 3 X EK3.01 - Knowledge of the reasons for the following responses as the apply to the SGTR: Equalizing pressure on primary and secondary sides of ruptured S/G (CFR 41.5/41.10/45.6/45.13) 4.1 1 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture

- Excessive Heat Transfer / 4 X W/E12, EA1.3

- Ability to operate and

/ or monitor the following as they apply to the (Uncontrolled Depressurization of all Steam Generators):

Desired operating results during abnormal and emergency

situations. (CFR: 41.7 / 45.5 /

45.6) 3.4 1 000054 (CE/E06) Loss of Main Feedwater / 4 X AA2.02 - Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW):

Differentiation between loss of all

MFW and trip of one MFW pump (CFR: 43.5 / 45.13) 4.1 1 000055 Station Blackout / 6 000056 Loss of Off

-site Power / 6 X AA1.08 - Ability to operate and/

or monitor the following as they apply to the Loss of Offsite Power:

HVAC chill water pump and unit (CFR 41.7/45.5/45.6) 2.5 1 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 X AA1.02 - Ability to operate and / or monitor the following as they apply

to the Loss of DC Power:

Static inverter dc input breaker, frequency meter, ac output breaker, and ground fault detector (CFR 41.7 / 45.5 /

45.6) 3.1 1 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 X 2.1.28 - Knowledge of the purpose and function of major system components and controls. (CFR: 41.7) 4.1 1 W/E04 LOCA Outside Containment / 3 X EK2.1 - Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. (CFR:

41.7 / 45.7) 3.5 1 W/E11 Loss of Emergency Coolant Recirc. / 4 X EK1.3 - Knowledge of the operational implications of the following concepts as they apply to the (Loss of Emergency Coolant Recirculation):

Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Emergency Coolant Recirculation). (CFR: 41.8 / 41.10 / 45.3) 3.6 1 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 X W/E05, EA2.1

- Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink): Facility conditions and selection of appropriate procedures

during abnormal and emergency operations. (CFR: 43.5 / 45.13) 3.4 1 000077 Generator Voltage and Electric Grid Disturbances / 6 X 2.1.19 - Ability to use plant computers to evaluate system or component status. (CFR: 41.10 /

45.12) 3.9 1 K/A Category Totals:

3 3 3 3 3 3 Group Point Total:

18/6 ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 2 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 X AK1.03 - Knowledge of the operational implications of the following concepts as they apply to Inoperable / Stuck Control Rod:

Xenon transient (CFR 41.8 / 41.10 / 45.3) 3.2 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 X 2.4.21 - Knowledge of the parameters and logic used to

assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5

/ 45.12) 4.0 1 000060 Accidental Gaseous Radwaste Rel. / 9 X AA1.01 - Ability to operate and

/ or monitor the following as they apply to the Accidental Gaseous Radwaste:

Area radiation monitors (CFR 41.7 /

45.5 / 45.6) 2.8 1 000061 ARM System Alarms

/ 7 000067 Plant Fire On

-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07)

Inad. Core Cooling

/ 4 000076 High Reactor Coolant Activity / 9 X AK2.01 - Knowledge of the interrelations between the High Reactor Coolant Activity and the following:

Process radiation monitors (CFR 41.7 /

45.7) 2.6 1 W/EO1 & E02 Rediagnosis

& SI Termination

/ 3 X W/E02, EA2.2

- Ability to determine and interpret the following as they apply to the (SI Termination):

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments. (CFR:

43.5 / 45.13) 3.5 1 W/E13 Steam Generator Over

-pressure / 4

W/E15 Containment Flooding / 5 X EK3.1 - Knowledge of the reasons for the following responses as they apply to the (Containment Flooding):

Facility operating

characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics. (CFR: 41.5 /

41.10, 45.6, 45.13) 2.7 1 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI

-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 X EK3.2 - Knowledge of the reasons for the following responses as they apply to the (LOCA Cooldown and Depressurization):

Normal, abnormal and emergency

operating procedures associated with (LOCA Cooldown and Depressurization). (CFR:

41.5 / 41.10, 45.6 / 45.13) 3.4 1 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 X 2.4.45 - Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 /

45.12) 4.1 1 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling

- PTS / 4 X EA2.1 - Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations. (CFR: 43.5 / 45.13) 3.1 1 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

1 1 2 1 2 2 Group Point Total:

9/4 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 1 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 003 Reactor Coolant Pump X K6.04 - Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:

Containment isolation valves affecting RCP operation (CFR: 41.7 / 45/5) 2.8 1 004 Chemical and Volume Control X K3.04 - Knowledge of the effect that a loss or

malfunction of the CVCS will have on the following:

RCPS (CFR: 41.7/45/6) 3.7 1 005 Residual Heat Removal X X K3.01 -- Knowledge of the effect that a loss or malfunction of the RHRS will have on the following:

RCS (CFR: 41.7 / 45.6)

A4.03 - Ability to manually operate and/or monitor in the control room:

RHR temperature, PZR heaters and flow, and nitrogen (CFR: 41.7 / 45.5 to 45.8) 3.9

2.8 1

1 006 Emergency Core Cooling X K5.08 - Knowledge of the operational implications of the following concepts as they apply to ECCS:

Operation of pumps in parallel (CFR: 41.5 / 45.7) 2.9 1 007 Pressurizer Relief/Quench Tank X A1.03 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including:

Monitoring quench tank temperature (CFR: 41.5 /

45.5) 2.6 1 008 Component Cooling Water X A3.04 - Ability to monitor automatic operation of the CCWS, including:

Requirements on and for the CCWS for different conditions of the power plant (CFR: 41.7 / 45.5) 2.9 1 010 Pressurizer Pressure Control X X K1.01 -- Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the

following systems:

RPS (CFR:

41.2 to 41.9 / 45.7 to 45.8)

A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those

predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Spray valve failures (CFR: 41.5 / 43.5 /

45.3 / 45.13) 3.9 3.9 1 1 012 Reactor Protection X A4.03 - Ability to manually operate and/or monitor in the control room:

Channel blocks and bypasses (CFR:

41.7 / 45.5 to 45.8) 3.6 1 013 Engineered Safety Features Actuation X X K2.01 -- Knowledge of bus power supplies to the

following:

ESFAS/safeguards equipment control (CFR:

41.7) K6.01 - Knowledge of the effect of a loss or

malfunction on the following will have on the ESFAS:

Sensors and detectors (CFR:

41.7 / 45.5 to 45.8) 3.6 2.7 1 1 022 Containment Cooling X K4.04 - Knowledge of CCS design feature(s) and/or interlock(s) which provide for the following:

Cooling of control rod drive motors (CFR: 41.7) 2.8 1 025 Ice Condenser Not part of the plant design 026 Containment Spray X X K1.02 -- Knowledge of the physical connections and/or

cause/effect relationships between the CSS and the

following systems: Cooling water (CFR: 41.2 to 41.9 /

45.7 to 45.8)

K3.01 - Knowledge of the effect that a loss or malfunction of the CSS will have on the following:

CCS (CFR: 41.7 / 45.6) 4.1 3.9 1 1 039 Main and Reheat Steam X K5.08 - Knowledge of the operational implications of the following concepts as the apply to the MRSS:

Effect of steam removal on reactivity (CFR: 41.5 / 45.7) 3.6 1 059 Main Feedwater X X A1.07 -- Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating

the MFW controls including:

Feed Pump speed, including normal control speed for ICS (CFR: 41.5 / 45.5)

A3.06 - Ability to monitor automatic operation of the MFW, including:

Feedwater isolation (CFR: 41.7 / 45.5) 2.5

3.2 1

1 061 Auxiliary/Emergency Feedwater X X K2.0 1 -- Knowledge of bus power supplies to the following:

AFW system MOVs (CFR: 41.7)

A1.01 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including:

S/G level (CFR: 41.5 / 45.5) 3.2 3.9 1 1 062 AC Electrical Distribution X K4.10 - Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following:

Uninterruptable ac power sources (CFR: 41.7) 3.1 1 063 DC Electrical Distribution X A3.01 - Ability to monitor automatic operation of the DC electrical system, including:

Meters, annunciators, dials, recorders, and indicating lights (CFR: 41.7 / 45.5) 2.7 1 064 Emergency Diesel Generator X K6.08 - Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Fuel oil storage tanks (CFR: 41.7 / 45.7) 3.2 1 073 Process Radiation Monitoring X 2.4.31 - Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 /

45.3) 4.2 1 076 Service Water X X K1.21 -- Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems:

Auxiliary backup SWS (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K4.02 - Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following:

Automatic start features associated with SWS pump controls (CFR: 41/7) 2.7 2.9 1 1 078 Instrument Air X 2.4.18 - Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13) 3.3 1 103 Containment X A2.05 - Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b)

based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Emergency con tainment entry (CFR: 41.5 / 43.5 / 45.3 /

45.13) 2.9 1 K/A Category Point Totals:

3 2 3 3 2 3 3 2 3 2 2 Group Point Total:

28/5 ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 2 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 001 Control Rod Drive X A3.04 - Ability to monitor automatic operation of the CRDS, including:

Radial imbalance (CFR:

41.7/45.13) 3.5 1 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication X A4.02 - Ability to manually operate and/or monitor in the control room:

Control rod mode-select switch (CFR: 41.7

/ 45.5 to 45.8) 3.4 1 015 Nuclear Instrumentation X K1.01 - Knowledge of the physical connections and/or cause/effect relationships between the NIS and the following systems:

RPS (CFR:

41.2 to 41.9 / 45.7 to 45.8) 4.1 1 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal X K2.01 - Knowledge of bus power supplies to the following:

Fans (CFR: 41.7) 3.1 1 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X A1.01 - Ability to predict and/or monitor changes in

parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including:

Spent fuel pool water level (CFR:

41.5 / 45.5) 2.7 1 034 Fuel Handling Equipment X K4.03 - Knowledge of design feature(s) and/or interlock(s) which provide for the following:

Overload protection (CFR: 41.7) 2.6 1 035 Steam Generator X K6.03 - Knowledge of the effect of a loss or malfunction on the following will have on the S/GS:

S/G level detector (CFR: 41.7 /

45.7) 2.6 1 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate

068 Liquid Radwaste X A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System ;

and (b) based on those

predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Insuffi cient sampling frequency of the boric acid in the evaporator bottoms (CFR:

41.5 / 43.5 / 45.3 / 45.13) 2.5 1 071 Waste Gas Disposal 072 Area Radiation Monitoring X K3.01 - Knowledge of the effect that a loss or malfunction of the ARM system will have on the following:

Containment ventilation isolation (CFR: 41.7 / 45.6) 3.2 1 075 Circulating Water 079 Station Air 086 Fire Protection X K5.03 - Knowledge of the operational implication of the

following concepts as they apply to the Fire Protection

System: Effect of water spray on electrical components (CFR:

41.5 / 45.7) 3.1 1 K/A Category Point Totals:

1 1 1 1 1 1 1 1 1 1 Group Point Total:

10/3 ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: Callaway Plant (RO, Rev. 0) Date of Exam:

2016 Category K/A # Topic RO SRO-Only IR # IR # 1. Conduct of Operations

2.1.7 Ability

to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13) 4.4 1 2.1.36 Knowledge of procedures and limitations involved in core alterations.

(CFR: 41.10 / 43.6 / 45.7) 3.0 1 2.1.44 Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation. (CFR: 41.10 / 43.7 / 45.12) 3.9 1 2.1. 2.1. Subtotal 3 2. Equipment Control 2.2.14 Knowledge of the process for controlling equipment configuration or status. (CFR: 41.10 / 43.3 / 45.13) 3.9 1 2.2.23 Ability to track Technical Specification limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13) 3.1 1 2.2. 2.2. 2.2. Subtotal 2 3. Radiation Control 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10) 3.8 1 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(CFR: 41.12 / 43.4 / 45.10) 3.4 1 2.3. 2.3. 2.3. Subtotal 2 4. Emergency Procedures / Plan 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13) 3.7 1 2.4.29 Knowledge of the emergency plan. (CFR: 41.10 / 43.5 / 45.11) 3.1 1 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

(CFR: 41.10 / 43.5 / 45.3 / 45.12) 4.2 1 2.4. 2.4. Subtotal 3 Tier 3 Point Total 10 7 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection

ES-401 PWR Examination Outline Form ES-401-2 Facility: Callaway Plant (SRO, Rev. 0) Date of Exam:

2016 Tier Group RO K/A Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total 1. Emergency & Abnormal Plant Evolutions 1 N/A N/A 18 3 3 6 2 9 2 2 4 Tier Totals 27 5 5 10 2. Plant Systems 1 28 3 2 5 2 10 1 2 3 Tier Totals 38 4 4 8 3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 1 2 2 3 2 4 2 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO

-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO

-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site

-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES

-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant

-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO

-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES

-401 for the applicable K/As. 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO

-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO

-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES

-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 1 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization

- Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 X AA2.19 - Ability to determine and interpret the following as they

apply to the Pressurizer Vapor Space Accident:

PZR spray valve failure, using plant parameters (CFR: 43.5 /

45.13) 3.6 1 000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 X AA2.01 - Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water:

Location of a leak in the CCWS (CFR: 43.5 / 45.13) 3.5 1 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 000038 Steam Gen. Tube Rupture / 3 X 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc. (CFR: 41.10 /

43.5 / 45.12) 4.2 1 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture

- Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 X AA2.07 - Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW): Reactor trip first

-out panel indicator (CFR: 43.5 / 45.13) 3.9 1 000055 Station Blackout / 6 000056 Loss of Off

-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 X 2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator

actions and directives affect plant and system conditions.

(CFR: 41.5 /

43.5 / 45.12) 4.4 1 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4

BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 X 2.4.9 - Knowledge of low power/shutdown implications in

accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 / 43.5 / 45.13) 4.2 1 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals:

3 3 Group Point Total:

18/6 ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 2 (RO / SRO) E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 X 2.1.20 - Ability to interpret and execute procedure steps. (CFR: 41.10 /

43.5 / 45.12) 4.6 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07)

Inad. Core Cooling

/ 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis

& SI Termination

/ 3 W/E13 Steam Generator Over

-pressure / 4 X EA2.2 - Ability to determine and interpret the following as they apply to the (Steam Generator Overpressure):

Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. (CFR: 43.5 / 45.13) 3.4 1 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 X 2.2.40 - Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 /

43.5 / 45.3) 4.7 1 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI

-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 X E10, EA2.2

- Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without

RVLIS): Adherence to appropriate procedures and

operation within the limitations in the facility's license and amendments. (CFR:

43.5 / 45.13) 3.9 1 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling

- PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

2 2 Group Point Total:

9/4 ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 1 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal X 2.4.35 - Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41.10 / 43.5

/ 45.13) 4.0 1 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank X A2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Overpressurization of the waste gas vent header (CFR:

41.5 / 43.5 / 45.3 / 45.13) 2.9 1 008 Component Cooling Water X 2.4.41 - Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10

/ 43.5 / 45.11) 4.6 1 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam X A2.02 - Ability to (a) predict the impacts of the

following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Decrease in turbine load as it relates to steam escaping

from relief valves (CFR:

41.5 / 43.5 / 45.3 / 45.13) 2.7 1 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution

064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air X A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the IAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Air dryer and filter malfun ctions (CFR:

41.5 / 43.5 / 45.3 / 45.13) 2.9 1 103 Containment K/A Category Point Totals:

3 2 Group Point Total:

28/5 ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

- Tier 2/Group 2 (RO / SRO) System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation X 2.1.23 - Ability to perform specific system and integrated plant procedures during all

modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 /

45.6) 4.4 1 017 In-Core Temperature Monitor X 2.4.30 - Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 /

43.5 / 45.11) 4.1 1 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling X A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadequate SDM (CFR: 41.5 /

43.5 / 45.3 / 45.13) 3.5 1 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air

086 Fire Protection K/A Category Point Totals:

1 2 Group Point Total:

10/3 ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: Callaway Plant (SRO, Rev. 0) Date of Exam:

2016 Category K/A # Topic RO SRO-Only IR # IR # 1. Conduct of Operations

2.1.5 Ability

to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12) 3.9 1 2.1. 2.1. 2.1. 2.1. Subtotal 1 2. Equipment Control 2.2.7 Knowledge of the process for conducting special or infrequent tests.

(CFR: 41.10 / 43.3 / 45.13) 3.6 1 2.2.35 Ability to determine Technical Specification Mode of Operation. (CFR: 41.7 / 41.10 / 43.2 / 45.13) 4.5 1 2.2. 2.2. 2.2. Subtotal 2 3. Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 / 41.12 / 43.4 / 45.9) 2.9 1 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator l duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high

-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 / 45.10) 3.8 1 2.3. 2.3. 2.3. Subtotal 2 4. Emergency Procedures / Plan 2.4.28 Knowledge of procedures relating to a security event (non-safeguards information). (CFR: 41.10 / 43.5 / 45.13) 4.1 1 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 / 43.5 / 45.12) 4.2 1 2.4. 2.4. 2.4. Subtotal 2 Tier 3 Point Total 10 7 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection

ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 NUREG-1021, Revision 10 Page 1 of 2 Facility: Callaway Date of Examination: 5/23/2016 Examination Level: RO Operating Test Number: 2016-1 Administrative Topic (see Note) Type Code* Describe activity to be performed Conduct of Operations A1 S, D 2.1.26 (3.4) Knowledge of industrial safety procedures JPM: Respond to an Industrial Injury Conduct of Operations A2 R, M 2.1.25 (3.9) Ability to interpret reference materials such as graphs, curves, tables, etc. JPM: Determine RV Venting Time (EOP ADD 33) Equipment Control A3 R, D, P 2.2.37 (3.6) Ability to determine operability and/or availability of safety related equipment. JPM: Determine Amperage Limits for 480 VAC Safety Related busses. Radiation Control A4 R, M 2.3.7 (3.5) Ability to comply with radiation work permit requirements during normal or abnormal conditions. JPM: Determine entry requirements for HRA in the RCA. 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 (N)ew or (M)odified from *The JPMs from the 2013 exam were randomly selected by placing 4 s2 2013 ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 NUREG-1021, Revision 10 Page 2 of 2 A1 This is a BANK JPM. The parent JPM (Set 4 RSA-1) has not been used on an ILT NRC Exam administered at Callaway between 2004 and 2014. This JPM is based on the Injuries A2 This is a MODIFIED JPM. The parent JPM was used on the 2009 ILT NRC exam. The candidate is to determine the maximum RV Venting time using EOP Addendum 33. A marked up FR-I.3 will be provided. A3 This BANK JPM was used on the 2013 ILT NRC Exam. The applicant will review planned maintenance which requires load centers NG01 and NG03 to be cross-connected. The applicant will be required to determine what equipment can be started on the cross-connected load centers without overloading the buses. A4 This is a MODIFIED JPM from the 2013 Palo Verde ILT NRC Exam. This JPM requires the RO to review given conditions and determine dose received for a task, required authorization for that dose, and posting requirements for the area where the task will be performed; in accordance with APA-ZZ-01004, Radiological Work standards, and HDP-ZZ-01500, Radiological Postings.

ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 NUREG-1021, Revision 10 Page 1 of 2 Facility: Callaway Date of Examination: 5/23/2016 Examination Level: SRO Operating Test Number: 2016 - 1 Administrative Topic (see Note) Type Code* Describe activity to be performed Conduct of Operations A5 R, D 2.1.37 (4.6) Knowledge of procedures, guidelines, or limitations associated with reactivity management JPM: Review a QPTR Calculation Conduct of Operations A6 R, M 2.1.25 (4.2) Ability to interpret reference materials such as graphs, curves, tables, etc JPM: Determine RV Venting Time (EOP ADD 33) Equipment Control A7 R, D, P 2.2.37 (4.6) Ability to determine operability and/or availability of safety related equipment JPM: Determine Amperage Limits for 480 VAC Safety Related busses Radiation Control A8 R, M 2.3.4 (3.7) Knowledge of radiation exposure limits under normal or emergency conditions JPM: Select Volunteer for Emergency Exposure Emergency Procedures/Plan A9 R, M 2.4.44 (4.4) Make a Protective Action Recommendation JPM: Determine the Protective Action Recommendation (PAR) 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 (N)ew or (M)odified from *No JPMs from the last 2 SRO exams (including the 2013 re-exam) were selected for this exam. JPM A7 was on the 2013 RO exam. This JPMs was randomly selected by placing 4 s20132 2013 ES-301 Administrative Topics Outline Form ES-301-1 Rev 0 NUREG-1021, Revision 10 Page 2 of 2 A5 This is a BANK JPM. The parent JPM (SRO-MAS-04-A006J) has not been used on an NRC Exam administered at Callaway between 2004 and 2014. The SRO candidate will be required to review a QPTR calculation and determine that an error occurred in the calculation and determine the QTPR is not within the limits of TS 3.2.4 and that required actions A.1, A.2, A.3, A.4, A.5 AND A.6 must be performed. A6 This is a MODIFIED JPM. The parent JPM was used on the 2009 ILT NRC exam. The candidate is to determine the maximum RV Venting time using EOP Addendum 33. A marked up FR-I.3 will be provided. A7 This BANK JPM was used on the 2013 ILT NRC Exam. The applicant will review planned maintenance which requires load centers NG01 and NG03 to be cross-connected. The applicant will be required to determine what equipment can be started on the cross-connected load centers without overloading the buses. A8 This is a MODIFIED JPM. The parent JPM (SRO-RER-03-A203J) was used on the 2009 ILT NRC exam. The SRO candidate will be given a set of conditions and the appropriate procedures in an emergency radiological situation. The SRO candidate, acting as the Emergency Coordinator, will determine which volunteer is the most eligible to receive an emergency dose. A9 This is a MODIFIED JPM. The parent JPM (SRO-RER-02-A031J(TC)) was used on the 2011 ILT NRC exam. The applicant will be assigned the task of determining the Protective Action Recommendation (PAR) within the allotted amount of time. Upon completion of this JPM the operator will have determined the PAR to be Evacuate 5 miles all sectors and Evacuate 10 miles sectors J, H, and G.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Page 1 of 4 Facility: __Callaway___________________________ Date of Examination: _5/23/2016___ Exam Level: RO SRO-I SRO-U Operating Test No.: __2016-1_______ Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S1 001 Control Rod Drive System (SF) / Perform Control Rod Partial Movement Test D, S 1 S2 004 CVCS (BG) / Swap From the NCP to 'B' CCP A, D, S 2 S3 010 Pressurizer Pressure Control System (BB) / Initiate Cold Overpressure Mitigation With PORV Malfunction A, D, L, S 3 S4 059 Main Feedwater System (AE) / Transfer Steam Generator Water Level Control A, N, S 4S S5 005 Residual Heat Removal System (EJ) / Transfer to Cold Leg Recirculation A, D, P1, EN, S 4P S6 062 A.C. Electrical Distribution (PA) / Perform Operational Testing of the Alternate Emergency Power Source D, P1, S 6 S7 015 Nuclear Instrumentation System (SE) / Respond to a Failed Power Range Instrument D, S 7 S8 Containment Purge System (GT) / Remove Shutdown Purge System From Service N, L, S 8 In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) P1 006 Emergency Core Cooling System (EP) / Secure Safety Injection Accumulators D, L 2 P2 035 Main and Reheat Steam System (AB) / Isolate a Failed Open Atmospheric Steam Dump A, M, E, R 4S P3 062 AC Electrical Distribution System (NN) / Transfer NN01 from Manual Bypass to Normal M 6

• 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 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Page 2 of 4 A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 Note 1. The JPMs from the 2013 exam were randomly selected by placing 11 slips of paper labe in a hardhat. Two of these items (S6 and S7) were drawn from the hardhat. S1 This is a BANK JPM. The JPM (URO-SSF-01-C005J) was used on the 2009 ILT NRC Exam. The applicant will be assigned the task of performing control rod partial movement for all shutdown banks, per OSP-SF-00002, Control Rod Partial Movement, beginning at step 6.1 Upon completion of this JPM, the applicant will have inserted at least 12 steps into the core and restored them to their pretest position. S2 This is an ALTERNATE PATH, BANK JPM. The JPM (URO-SBG-02-C160J (A)) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014.The applicant will perform the actions of OTN-BG-00001, Addendum 1 to shift from the NCP to the B CCP. After the B CCP is started and during the transition from the NCP flow controller to the B CCP flow controller, the B CCP will Trip, requiring the applicant to restore charging flow. Upon completion of this JPM the applicant will have restored charging flow to normal. S3 This is an ALTERNATE PATH, BANK JPM. The JPM (URO-SBB-02-C065J (A)) was used on the 2007 ILT NRC Exam. The applicant will be directed to ARM the Pressurizer Power Operated Relief Valves for Cold Overpressure Mitigation in accordance with Section 5.6 of OTN-BB-PORV BB-HIS-456A will open. Upon completion of this JPM, the applicant will have armed both Pressurizer PORVs for Cold Overpressure Mitigation and isolated or closed BB PV-456A after it fails open.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Page 3 of 4 S4 This is an ALTERNATE PATH, NEW JPM. The applicant will be assigned the task transferring Steam Generator Water Level Control from the MFRV Bypass Valves to the Main Feedwater Regulating Valves using OTN-AE-00001, will abort the automatic valve transfer and manually maintain SGWL. Upon completion of this JPM, the applicant will have transferred Steam Generator Water Level Control from the MFRV Bypass Valves to the MFRVs without causing a Feedwater Isolation Signal due to high or low Steam Generator water level. S5 This is an ALTERNATE PATH, BANK JPM that was used on the 2013 ILT NRC Exam (S7 on 2013 exam). It was randomly selected using the method described above. The simulator will be set up following a large Loss of Coolant Accident. The applicant will be directed to transfer the Emergency Core Cooling System to the recirculation mode in accordance with ES-1.3, Transfer to Cold Leg Recirculation. During performance, the applicant finds valves out of position and must use the Response Not Obtained column to complete the task. Upon completion of this JPM, the applicant will have aligned the RHR pumps for cold leg recirculation and aligned the SI pumps and CCPs suction to the RHR pumps IAW ES-1.3. S6 This is a BANK JPM that was used on the 2013 ILT NRC Exam (S6 on 2013 exam). It was randomly selected using the method described above. The applicant will be assigned the task of performing an online test of Alternate Emergency Power Source Diesel Generator #4 from the Control Room. The diesel will be started, readings taken and then secured from the Control Room. S7 This is a BANK JPM. The JPM (URO-SSE-03-C126J) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014. The applicant will perform the actions of OTO-SE-00001, Nuclear Instrument Malfunction, Attachment A to bypass the Power Range NIS Channel N41 current comparator and rod stop inputs. Upon Completion of this JPM, Power Range NIS channel N41 current comparator and rod stop inputs will be bypassed. The control power fuses for N41 will be removed. S8 This is a NEW JPM. The applicant will perform the actions of OTN-GT-00001, Containment Purge System, to remove containment shutdown purge from service. Upon completion of this JPM, the applicant will have removed containment shutdown purge from service IAW OTN-GT-00001. P1 This is a BANK JPM. The JPM (RO-SRO Au j) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014. The applicant will be assigned the task of locally securing Safety Injection accumulators per OTG-ZZ-00006, Addendum. Upon completion of this JPM, the applicant will have closed the SI Accumulator Outlet Isolation Valves and opened the feeder breakers to the SI accumulator outlet isolation valves.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Page 4 of 4 P2 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (EOP-SAB08077J(A)) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014. The applicant will be assigned the task of locally closing Atmospheric Steam Dumps, AB PV-3 AND AB PV-4. Upon completion of this JPM, the applicant will have closed AB PV-3 and isolated AB PV-4. AB PV-3 was closed by isolating Air/N2 from the valve. AB PV-4 was isolated by closing the manual isolation valve, ABV0007. P3 This is a MODIFIED JPM. The parent JPM (EOS-SNN-03-P010J) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014. The applicant will be assigned the task of transferring NN01 to the normal power source per OTN-NN-00001. Upon completion of this JPM the applicant will have transferred NN01 to the normal power supply (inverter and NK01) without a loss of voltage.

ES-301 Control Room/In

-Plant Systems Outline Form ES-301-2 Page 1 of 4 Facility: __Callaway___________________________

Date of Examination: _5/23/2016___ Exam Level:

RO SRO-I SRO-U Operating Test No.: __2016-1_______ Control Room Systems

• 8 for RO; 7 for SRO

-I; 2 or 3 for SRO-U System / JPM Title Type Code*

Safety Function S1 001 Control Rod Drive System (SF) / Perform Control Rod Partial Movement Test D, S 1 S2 004 CVCS (BG) / Swap From the NCP to 'B' CCP A, D, S 2 S3 010 Pressurizer Pressure Control System (BB) / Initiate Cold Overpressure Mitigation With PORV Malfunction A, D, L, S 3 S4 059 Main Feedwater System (AE) / Transfer Stea m Generator Water Level Control A, N, S 4S S5 005 Residual Heat Removal System (EJ) / Transfer to Cold Leg Recirculation A, D, P 1, EN, S 4P S6 062 A.C. Electrical Distribution (PA) / Perform Operational Testing of the Alternate Emergency Power Source D, P 1 , S 6 S7 015 Nuclear Instrumentation System (SE) / Respond to a Failed Power Range Instrument D, S 7 In-Plant Systems

• (3 for RO); (3 for SRO

-I); (3 or 2 for SRO

-U) P1 006 Emergency Core Cooling System (EP)

/ Secure Safety Injection Accumulators D, L 2 P2 035 Main and Reheat Steam System (AB)

/ Isolate a Failed Open Atmospheric Steam Dump A, M, E, R 4S P3 062 AC Electrical Distribution System (NN)

/ Transfer NN01 from Manual Bypass to Normal M 6

• 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 ES-301 Control Room/In

-Plant Systems Outline Form ES-301-2 Page 2 of 4 A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in

-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 Note 1. The JPMs from the 2013 exam were randomly selected by placing 11 slips of paper labeled "S1" through "P3" in a hardhat. Two of these items (S6 and S7) were drawn from the hardhat

.

S1 This is a BANK JPM. The JPM (URO-SSF-01-C005J) was used on the 2009 ILT NRC Exam. The applicant will be assigned the task of performing control rod partial movement for all shutdown banks, per OSP

-SF-00002, Control Rod Partial Movement, beginning at step 6.1 Upon completion of this JPM, the applicant will have inserted all shutdown bank 'A' control rods at least 12 steps into the core and restored them to their pretest position.

S2 This is an ALTERNATE PATH, BANK JPM. The JPM (URO-SBG-02-C160J (A)) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014

.The applicant will perform the actions of OTN

-BG-00001, Addendum 1 to shift from the NCP to the B CCP.

After the B CCP is started and during the transition from the NCP flow controller to the B CCP flow controller, the B CCP will Trip, requiring the applicant to restore charging flow

. Upon completion of this JPM the applicant will have restored charging flow to normal.

S3 This is an ALTERNATE PATH, BANK JPM. The JPM (URO-SBB-02-C065J (A)) was used on the 2007 ILT NRC Exam.

The applicant will be directed to ARM the Pressurizer Power Operated Relief Valves for Cold Overpressure Mitigation in accordance with Section 5.6 of OTN

-BB-00005, "Pressurizer and Pressurizer Pressure Control." When the Train B COM Switch is placed in ARM, Pressurizer PORV BB-HIS-456A will open. Upon completion of this JPM, the applicant will have armed both Pressurizer PORVs for Cold Overpressure Mitigation and isolated or closed BB PV

-456A after it fails open.

ES-301 Control Room/In

-Plant Systems Outline Form ES-301-2 Page 3 of 4 S4 This is an ALTERNATE PATH, NEW JPM. The applicant will be assigned the task transferring Steam Generator Water Level Control from the MFRV Bypass Valves to the Main Feedwater Regulating Valves using OTN

-AE-00001, Feedwater System. During the transfer the 'D' MFRV will not open. The applicant will abort the automatic valve transfer and manually maintain SGWL.

Upon completion of this JPM, the applicant will have transferred Steam Generator Water Level Control from the MFRV Bypass Valves to the MFRVs for SG 'A', 'B', and 'C" and taken manual control of SG 'D' water level without causing a Feedwater Isolation Signal due to high or low Steam Generator water level. S5 This is an ALTERNATE PATH, BANK JPM that was used on the 2013 ILT NRC Exam (S7 on 2013 exam)

. It was randomly selected using the method described above. The simulator will be set up following a large Loss of Coolant Accident. The applicant will be directed to transfer the Emergency Core Cooling System to the recirculation mode in accordance with ES

-1.3, Transfer to Cold Leg Recirculation. During performance, the applicant finds valves out of position and must use the Response Not Obtained column to complete the task. Upon completion of this JPM, the applicant will have aligned the RHR pumps for cold leg recirculation and aligned the SI pumps and CCPs suction to the RHR pumps IAW ES-1.3.

S6 This is a BANK JPM that was used on the 2013 ILT NRC Exam (S6 on 2013 exam). It was randomly selected using the method described above.

The applicant will be assigned the task of performing an online test of Alternate Emergency Power Source Diesel Generator #4 from the Control Room. The diesel will be started, readings taken and then secured from the Control Room.

S7 This is a BANK JPM.

The JPM (URO-SSE-03-C126J) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014. The applicant will perform the actions of OTO

-SE-00001, Nuclear Instrument Malfunction, Attachment A to bypass the Power Range NIS Channel N41 current comparator and rod stop inputs.

Upon Completion of this JPM, Power Range NIS channel N41 current comparator and rod stop inputs will be bypassed. The control power fuses for N41 will be removed.

P1 This is a BANK JPM. The JPM (RO-SRO Au j) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014. The applicant will be assigned the task of locally securing Safety Injection accumulators per OTG

-ZZ-00006, Addendum. Upon completion of this JPM, the applicant will have closed the SI Accumulator Outlet Isolation Valves and opened the feeder breakers to the SI accumulator outlet isolation valves.

ES-301 Control Room/In

-Plant Systems Outline Form ES-301-2 Page 4 of 4 P2 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (EOP

-SAB08077J(A)) has not been used on an NRC ILT Exam administered a t Callaway between 2004 and 2014

. The applicant will be assigned the task of locally closing Atmospheric Steam Dumps, AB PV

-3 AND AB PV

-4. Upon completion of this JPM, the applicant will have closed AB PV

-3 and isolated AB PV-4. AB PV-3 was closed by isolating Air/N2 from the valve. AB PV

-4 was isolated by closing the manual isolation valve, ABV0007.

P3 This is a MODIFIED JPM. The parent JPM (EOS

-SNN-03-P010J) has not been used on an NRC ILT Exam administered at Callaway between 2004 and 2014

. The applicant will be assigned the task of transferring NN01 to the normal power source per OTN-NN-00001. Upon completion of this JPM the applicant will have transferred NN01 to the normal power supply (inverter and NK01) without a loss of voltage.

ES-301 Transient and Event Checklist Form ES-301-5 Page 1 of 3 Facility: Callaway Date of Exam: 05/23/2016 Operating Test No.: 2016-1 Team 1 & 2 SRO-I: S1, S2, S3, S4, S5, and S6 A P P L I C A N T E V E N T T Y P E Scenarios 1 2 3 T O T A L M I N I M U M(*) CREW POSITION CREW POSITION CREW POSITION S1/S4 S2/ S5 S3/ S6 S2/ S5 S3/ S6 S1/S4 S3/ S6 S1/S4 S2/ S5 S R O A T C B O P S R O A T C B O P S R O A T C B O P R I U SRO-I (S1 / S4) RX 6 4 2 1 NOR 1 1 2 1 I/C 2,4,5 2,4 1 6 4 MAJ 7 6 5 3 2 TS 2,4 2 2 SRO-I (S2 / S5) RX 4 1 1 NOR 3 1 2 1 I/C 2,6 2,3,4,5 2 7 4 MAJ 7 6 5 3 2 TS 2,3 2 2 SRO-I (S3 / S6) RX 6 2 4 3 1 NOR 1 1 1 I/C 4,5 3,5 1,2 6 4 MAJ 7 6 5 3 2 TS 1,3 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 -the- -of-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 minimum requirements specified for the appl-hand columns.

ES-301 Transient and Event Checklist Form ES-301-5 Page 2 of 3 Facility: Callaway Date of Exam: 05/23/2016 Operating Test No.: 2016-1 Team 3 SRO-I: S7 / RO: R1 A P P L I C A N T E V E N T T Y P E Scenarios 1 2 T O T A L M I N I M U M(*) CREW POSITION CREW POSITION CREW POSITION Surrogate S7 R1 S7 R1 Surrogate S R O A T C B O P S R O A T C B O P S R O A T C B O P R I U SRO-I (S7) RX 0* 1 NOR 3 1 2 1 I/C 2,6 2,3,4,5 6 4 MAJ 7 6 2 2 TS 2,3 2 2 RO (R1) RX 6 2 2 1 NOR 1 1 1 I/C 4,5 3,5 4 4 MAJ 7 6 2 2 TS RX NOR I/C MAJ TS Instructions: 3. 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 -the- -of-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. 4. 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 evel in the right-hand columns.

ES-301 Transient and Event Checklist Form ES-301-5 Page 3 of 3 Facility: Callaway Date of Exam: 05/23/2014 Operating Test No.: 2016-1 A P P L I C A N T E V E N T T Y P E Scenarios 4 T O T A L M I N I M U M(*) CREW POSITION CREW POSITION CREW POSITION CREW POSITION S R O A T C B O P S R O A T C B O P S R O A T C B O P S R O A T C B O P R I U SPARE RX NOR I/C 2,3,4 2,4 3,4 MAJ 5 5 5 TS 1,2 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 -the- -of-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 -hand columns. SRO ATC BOP Scenario 1 05/24/16 0730-0900 Team 1 S1 S2 S3 05/24/16 1000-1130 Team 2 S4 S5 S6 05/24/16 1230-1400 Team 3 Surrogate S7 R1 Scenario 2 05/25/16 0730-0900 Team 1 S2 S3 S1 05/25/16 1000-1130 Team 2 S5 S6 S4 05/25/16 1230-1400 Team 3 S7 R1 Surrogate Scenario 3 05/26/16 0730-0900 Team 1 S3 S1 S2 05/26/16 1000-1130 Team 2 S6 S4 S5 Page 1 of 4 Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 1, Rev 0 Op-Test No.: 2016-1 Examiners: ____________________________ Operators: _____________________________ ____________________________ _____________________________ ____________________________ _____________________________ Initial Conditions: 100% Turnover: seal. The applicable Tech Spec is 3.5.2 A (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). The Balance of Plant (BOP) is directed to shift Event No. Malf. No. Event Type* Event Description 1 NA SRO (N) BOP (N) 2 BBLT459 SRO (I) RO (I) Pressurizer Level Transmitter BB LT-459 Fails Low(Tech Spec) 3 NA RO (N) Restore Letdown 4 ABPT0514 SRO (I) BOP (I) A S/G Steam Pressure Channel PT-514 Fails Low (Tech Spec) 5 PEG01B_1 SRO (C) BOP (C) B CCW Pump Trip / D CCW Pump Failure to Auto Start 6 KAL03 SRO (R) RO (C) BOP (R) Loss of Instrument Air to Containment 7 BB002_C SRO (M) RO (M) BOP (M) RCS Leak LOCA 8 NF039A_1 SRO (C) BOP (C) LOCA Sequencer Train A Failure * (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 Scenario Event Description Callaway 2016-1 NRC Scenario #1, rev. 0 Page 2 of 4 The plant is stable at 100ago to replace a shaft seal. The applicable Tech Spec is 3.5.2 A (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). The Balance of Plant (BOP) is directed to sh After the CCW service loop has been swapped, Pressurizer Level Channel BB LT-459 fails low, resulting in a loss of letdown. The crew will respond IAW OTO-BG-00001, Pressurizer Level Control Malfunction, select an operable pressurizer level channel and restore letdown to service. Tech Spec 3.3.1 applies. After Tech Specs have been addressed, Steam Generator A Pressure Channel 514 fails low. This causes a feedwater flow reduction and a lowering SG level. The crew should respond per OTO-AE-00002, Steam Generator Water Level Control Malfunctions, select an operable channel for control, and stabilize SG level. Tech Spec 3.3.2 applies. After Tech Specs have been addressed, -EG-00001, CCW System Malfunction, ally. The CRS should review Tech Spec 3.7.7 for "B" CCW Train. When plant conditions are stable, the crew will experience a failure of instrument air in CTMT. The initial indication will be a loss of letdown. The crew may respond with OTO-BG-00001, Pressurizer Level Control Malfunction. When it is recognized that a loss of air to containment has occurred the crew should then enter OTO-KA-0001, Partial or Total Loss of Instrument Air, to respond to the loss of air inside CTMT. The crew will begin a rapid down power per OTO-KA-00001, Attachment A. When a sufficient downpower (MWe < 1100) is achieved, the scenario continues with the next event. Once Turbine Load is reduced to1100 MWe, a leak in the RCS develops which will be seen by the crew as PZR level lowering and containment pressure rising. The crew will manually trip the reactor based on these plant conditions. The crew should enter E-0, Reactor Trip or Safety Injection. train of the LOCA sequencer fails to actuate. This will be indicated to the crew by the , and RHR pump not stating. The crew should manually start these pumps in accordance with E-0, Reactor Trip or Safety Injection, Attachment A. The crew will transition to E-1, Loss of Reactor or Secondary Coolant. The crew will then stop all RCPs within 5 minutes of meeting the RCP trip criteria. This action may be completed in E-0 per the foldout page or per step 12. The scenario will end after the crew has performed E-1 and transitions to ES-1.2, Post LOCA Cooldown and Depressurization Scenario Event Description Callaway 2016-1 NRC Scenario #1, rev. 0 Page 3 of 4 Critical Tasks: Critical Tasks Trip all RCPs within 5 minutes of meeting RCP trip criteria. Establish flow from 'A' CCP before completion of E-0 Attachment A EVENT 7 8 Safety significance Failure to trip the RCPs under the postulated plant conditions leads to core uncovery and to fuel cladding temperatures in excess of 2200°F, which is the limit specified in the ECCS acceptance criteria. Thus, failure to perform the task represents misoperation or on of...{the The acceptable results obtained in the FSAR analysis of a small-break LOCA are predicated on the assumption of minimum ECCS pumped injection. The analysis assumes that a minimum pumped ECCS flow rate, which varies with RCS pressure, is injected into the core. The flow rate values assumed for minimum pumped injection are based on operation of one each of the following ECCS pumps: Charging/SI pump (HP plants only), high-head SI pump, and low-head SI pump. Operation of this minimum required complement of ECCS injection pumps is consistent with the FSAR assumption that only minimum safeguards are actuated. Because compliance with the assumptions of the FSAR is part of the facility license condition, failure to perform the critical task (under the postulated plant conditions) constitutes a violation of the license condition. Cueing Indications of a SBLOCA AND Indication and/or annunciation of safety injection AND Indication and/or annunciation that only one train of actuates AND Indication that RCS pressure Indication and/or annunciation that Charging/SI pump injection is required SI actuation RCS pressure below the shutoff head of the Charging/SI pump Indication and/or annunciation that no Charging/SI pump is injecting into the core Control switch indication that the circuit breakers or contactors for both Charging/SI pumps are open All Charging/SI pump discharge pressure indicators read zero All flow rate indicators for Charging/SI pump injection read zero Performance indicator Manipulation of controls as required to trip all RCPs RCP breaker position lights indicate breaker open Manipulation of controls in the control room as required to start the 'A' CCP Performance feedback Indication that all RCPs are stopped: RCP breaker position lights RCP flow decreasing RCP motor amps decreasing CCP is injecting Justification for the chosen performance limit -117, March 1984) (also known as the Sheppard letter), the WOG provided 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. completion of Attachment A of E-Group Emergency Response Guidelines. It allows enough time for the crew to take the correct action while at the same time preventing avoidable adverse consequences. PWR Owners Group Appendix CT- 16, Manually Trip RCPS CT-6, Establish flow from at least one Charging/SI pump Page 4 of 4 References OTO-BG-00001, Pressurizer Level Control Malfunction OTO-AE-00002, Steam Generator Water Level Control Malfunctions OTO-EG-00001, CCW System Malfunction OTO-KA-00001 Partial or Total Loss of Instrument Air E-0, Reactor Trip or Safety Injection E-1, Loss of Reactor or Secondary Coolant Tech Spec 3.3.1 Tech spec 3.3.2 ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions 1. Small LOCA (S(2)) a. Manually start one CCP Page 1 of 4 Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 2, Rev 0 Op-Test No.: 2016-1 Examiners: ____________________________ Operators: _____________________________ ____________________________ _____________________________ ____________________________ _____________________________ Initial Conditions: 100% Turnover: The Balance of Plant (BOP) is directed to perform Control Valve Partial Stroke Test on CV-1 in accordance with Section 6.2.1, OSP-AC-00003,Turbine Control Valve Stroke Test. Event No. Malf. No. Event Type* Event Description 1 NA SRO (N) BOP (N) Perform Control Valve Partial Stroke Test on CV-1 2 ACPT0505 SRO (I) RO (R) BOP (I) First Stage Turbine Pressure Indicator Failure (Tech Spec) 3 M04_DA SRO (I) RO (I) Loss of DRPI (Rod M-4) (Tech Spec) 4 AEFCV0520 SRO (C) BOP (C) SG MFRV Failure 5 CRCPV2 SRO (C) RO (C) C RCP High Vibration 6 SF006 SRO (M) RO (M) BOP (M) Nuclear Power Generation / ATWS 7 SA075A SRO (C) BOP (C) S/G C ASD Sticks Open * (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 Scenario Event Description Callaway 2016-1 NRC Scenario #2, rev. 0 Page 2 of 4 The plant is stable at 100%. The Balance of Plant (BOP) is directed to perform Control Valve Partial Stroke Test on CV-1 in accordance with Section 6.2.1, OSP-AC-00003,Turbine Control Valve Stroke Test. After Turbine Control Valve testing is complete, Turbine First Stage Pressure Indicator AC PI-505 fails low. This causes the control rods to step in. The crew should respond per OTO-AC-00003, Turbine Impulse Pressure Channel Failure, take manual control of control rods, select and operable turbine first stage pressure channel, and restore RCS Tavg to within 1°F of Tref. Tech Spec 3.3.1 applies. After Tech Specs have been addressed, DRPI for rod M-4 will fail. The crew will be alerted to the failure by annunciator 80A and 80B. The crew should take actions per OTA-RK-00022 Addendum 80A to place rod control in Manual and record RCS Tavg once per hour. Technical Specification 3.1.7 applies. After Tech Specs have been addressed, 'B' MFRV fails closed over 120 sec. The crew should respond per OTO-AE-00001, Feedwater System Malfunctions, and manual and restore SG NR level to between 45 and 55%. After SG level has been returned to between 45% and 55%, a mechanical failure causes RCP C vibrations to rise rapidly above the immediate trip setpoint. This will drive the crew to enter OTO-BB-00002, RCP Off Normal. The crew will recognize the need to immediately trip the Reactor and the C RCP. When the crew attempts to trip the reactor it will NOT trip. The crew should enter E-0 and transition to FR-S.1, Response to Nuclear Power Generation / ATWS, at step 1 of E-0. The C RCP should NOT be tripped until Reactor power is Less than 5%. During the performance of FR-S.1, rods will drop into the core after PG19 and PG20 feeder breakers are opened to deenergize the rod drive MG sets. The crew will return to E-0 and continue with the recovery. During FR-S.1, the C S/G ASD will Fail to Close after opening during the ATWS. An SI will occur and the crew will continue through E-0. The crew will isolate steam flow from and feed flow to the C S/G per fold out page of E-0. The ASD will NOT be able to be manually closed from the Control Room and Local Operator action will be required to close the isolation valve for the ASD. The crew will transition to E-2, Faulted Steam Generator, and then transition to ES-1.1, SI Termination. The scenario may be terminated after transition to ES-1.1, SI Termination Scenario Event Description Callaway 2016-1 NRC Scenario #2, rev. 0 Page 3 of 4 Critical Tasks: Critical Tasks Insert negative reactivity into the core by at least one of the following methods before dispatching operators to locally Trip the Reactor Deenergize PG19 and PG20 Insert Control Rods Establish emergency boration flow to the RCS -2. EVENT 6 7 Safety significance In the scenario, failure to insert negative reactivity by one of the methods listed previously can result in the needless continuation of an extreme or a severe challenge to the subcriticality CSF. Although the challenge was not initiated by the crew (was not initiated by operator error), continuation of the challenge is a result of the crew's failure to insert negative reactivity. Failure to isolate a faulted SG that can be isolated causes challenges to CSFs beyond those irreparably introduced by the postulated conditions. Failure to isolate a faulted SG can result in challenges to the following CSFs: Integrity Subcriticality Containment (if the break is inside containment) Cueing In the scenario, failure to insert negative reactivity by one of the methods listed previously can result in the needless continuation of an extreme or a severe challenge to the subcriticality CSF. Although the challenge was not initiated by the crew (was not initiated by operator error), continuation of the challenge is a result of the crew's failure to insert negative reactivity. Both of the following: Steam pressure and flow rate indications that faulted AND Valve position and flow rate indication that AFW continues to be delivered to the Performance indicator Manipulation of controls in the control room as required to initiate the insertion of negative reactivity into the core (at least one of the following) Open supply breakers to PG19 and PG20. o PG HIS-16 and PG HIS-18 Insert Control Rods at the Maximum Rate. ALIGN emergency boration flow path: o Start boric acid transfer pumps BG HIS-5A and BG HIS-6A o OPEN Emergency Borate To Charging Pump Suction valve: BG HIS-8104 ISOLATE AFW flow to faulted SG(s): CLOSE associated MD AFP Flow Control Valve(s): o AL HK-11A (SG C) CLOSE associated TD AFP Flow Control Valve(s): o AL HK-12A (SG C) CLOSE Steamline Low Point Drain valve from faulted SG(s): o AB HIS-7 (SG C) FAST CLOSE all MSIVs and Bypass valves: o AB HS79 o AB HS80 Performance feedback Crew will observe the following: Indication of a negative SUR on the intermediate range of the excore NIS Indication of less than 5% power on the power range of the excore NIS Crew will observe the following: Any depressurization of intact SGs stops AFW flow rate indication to faulted SG of zero Justification for the chosen performance limit Local operator actions would result in reactor trip, which would shut down the reactor faster than boration (and faster than rod insertion). However, it is anticipated that effecting the local actions will be time-consuming and that actions that can be implemented from the control room should be given precedence. Thus, before dispatching operators to perform local actions to trip the reactor, the crew should perform or initiate performance of at least one of the three methods listed previously for shutting down the reactor and providing shutdown margin. -Response Guidelines. It allows enough time for the crew to take the correct action while at the same time preventing avoidable adverse consequences. PWR Owners Group Appendix CT- 52, Insert negative reactivity into the core CT-17 Isolate faulted SG Page 4 of 4 References OSP-AC-00003, Turbine Control Valve Stroke Test OTO-AC-00003, Turbine Impulse Pressure Channel Failure OTA-RK-00022 Addendum 80A, Rod Position Indication Urgent Alarm OTO-AE-00001, Feedwater System Malfunction OTO-BB-00002, RCP Off Normal E-0, Reactor Trip or Safety Injection E-2, Faulted Steam Generator Isolation FR-S.1, Response to Nuclear Power Generation / ATWS Tech Spec 3.3.2 Tech spec 3.3.1 ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions 1. ATWS TAT3 a. Manual Control Rod Insertion Page 1 of 4 Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 3, Rev 0 Op-Test No.: 2016-1 Examiners: ____________________________ Operators: _____________________________ ____________________________ _____________________________ ____________________________ _____________________________ Initial Conditions: 100% Turnover: The Auxiliary Feedpump has been out of service for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Work is scheduled to complete next shift. Event No. Malf. No. Event Type* Event Description 1 BBTE0411A1 SRO (I) RO (I) RTD Fails High (Tech Spec) 2 PCE01A SRO (C) BOP (C) Stator Cooling Pump Trip with AUTO Start Failure 3 BNLT0932 SRO Refueling Water Storage Tank (RWST) Level Channel Fails Low (Tech Spec) 4 EAD05A SRO (R) BOP (R) RO (R) Partial Loss of Condenser Vacuum 5 AB003 9XX_2 & 6 SRO (M) RO (M) BOP (M) Large Steam Line Rupture in Turbine Building with failing open 6 PAL02_3 PAL01B_1 SRO (C) BOP (C) MD AFP B trips 2 minutes after starting and TDAFP fails to automatically start * (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) 2 3. Abnormal events (2-4) 3 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 Scenario Event Description Callaway 2016-1 NRC Scenario #3, rev. 0 Page 2 of 4 The plant is steady at 100% power. The Auxiliary Feedpump is tagged out for maintenance and will not be returned until next shift. Once the crew takes the watch, the Loop 1 Hot Leg RTD will fail high causing the control rods to drive in. The Reactor Operator will take manual control of the control rods and respond in accordance with OTO-BB-00004, RCS RTD Channel Failures. Tech Specification 3.3.1 applies. After Tech Specs have been addressed, the running SCW Pump trips and the standby pump does not auto-start. A turbine runback begins as indicated by load reduction & annunciator 132C. The crew will take action to start the standby SCW Pump (prudent action, OTA directed, or OTO-MA-00001 directed). OTO-MA-00001, Turbine Load Rejection, will be entered with actions taken to stabilize the plant and initiate recovery. When plant conditions are stable, a Refueling Water Storage Tank (RWST) level channel fails low. The crew will respond IAW OTO-BN-00001, RWST Level Channel Malfunction, Tech Spec 3.3.2 applies. When plant conditions are stable, a partial loss of Condenser vacuum will occur. The crew will perform actions per OTO-AD-00001, Loss of Condenser Vacuum. The crew will commence a down power in an attempt to restore vacuum. When a sufficient downpower (MWe < 1100) is achieved, the scenario continues with the next event. Once Turbine Load is reduced to1100 MWe, a steam leak develops in the Turbine Building which will be seen by the crew as RCS pressure and temperature rapidly lower. The crew may Manually trip the reactor based on these plant conditions. The crew should enter E-0, Reactor Trip or Safety Injection. The automatic steamline isolation fails to occur. The crew should manually initiate MSLIS. The remains open. The crew should make efforts to complete the -2, Faulted S/G Isolation, but the isolated. and then trips after running for 2 minutes. The TDAFP must be started manually due to malfunction inserted during the setup. The crew will then restore adequate feed to the intact Steam Generators. The scenario will end after the crew has completed E-2 and starts to transition to ES-1.1, SI Termination Scenario Event Description Callaway 2016-1 NRC Scenario #3, rev. 0 Page 3 of 4 Critical Tasks: Critical Tasks Manually actuate main steamline isolation before a severe (ORANGE path) challenge develops to either the subcriticality or the integrity CSF or before transition to ECA-2.1 (whichever happens first) Establish 285,000 lbm/hr feedflow to the SGs before transition out of E-0 EVENT 5 6 Safety significance Failure to isolate the SGs from the steamline break such that all SGs are allowed to blow down uncontrollably significantly worsens the power excursion. This worsening of the power excursion is unnecessary; it could be prevented simply by closing the MSIVs Under the postulated plant conditions, failure to manually establish the minimum safety margin beyond that irreparably introduced by the Cueing Indication that main steamline isolation is required AND Indication that main steamline isolation has not actuated automatically MSIVs indicate open Indication of uncontrolled depressurization of all SGs Indication and/or annunciation that SI is actuated AND Indication and/or annunciation that the AFW flow rate is less than the minimum required Total AFW flow rate indicates less than the minimum required Control switch indication that the circuit breakers or contactors for the motor-driven AFW pumps are open Control switch indication that the steam supply valves to the turbine-driven AFW pump are closed Performance indicator Manipulation of controls as required to manually actuate steamline isolation MSIVs undergo fast-closure MSIVs (except B) indicate closed Manipulation of controls in the control room as required to establish the minimum required AFW flow rate to the SGs Performance feedback Crew will observe the following: Steam flow indication from all SGs except B decreases to zero All SGs except B stop depressurizing RCS cooldown rate slows Indication that at least the minimum required AFW flow rate is being delivered to the SGs SG levels increasing Justification for the chosen performance limit Uncontrolled depressurization of all SGs causes an excessive rate of RCS cooldown, well beyond the conditions typically analyzed in the FSAR. The excessive cooldown rate creates large thermal stresses in the reactor pressure vessel and causes rapid insertion of a large amount of positive reactivity. Thus, failure to close the MSIVs under the postulated conditions can result in challenges to the following CSFs: Integrity Subcriticality The acceptable results obtained in the FSAR analyses are predicated on the assumption that, at the very least, one train of safeguards actuates. If AFW flow commensurate with minimum safeguards actuation is not established, the FSAR assumptions and results are invalid. Because compliance with the assumptions of the FSAR is part of the facility license condition, failure to manually establish at least the minimum required AFW flow rate (under the postulated conditions and when it is possible to do so) constitutes a violation of the license condition. PWR Owners Group Appendix CT- 12, Manually actuate main steamline isolation CT-4, Establish AFW flow to SGs Page 4 of 4 References OTO-BB-00004, RCS RTD Channel Failures OTA-RK-00026 Add 132C, Generator Protection Runback Circuit OTO-MA-00001, Turbine Load Rejection OTO-BN-00001, RWST Level Channel Malfunction OTO-AD-00001, Loss of Condenser Vacuum E-0, Reactor Trip or Safety Injection E-2, Faulted S/G Isolation Tech Spec 3.3.1 for Reactor Trip System Instrumentation Tech spec 3.3.2 for ESFAS Instrumentation ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions 1. Main Steam Line Break Outside Containment (T(MSI)) a. MSIV Closure b. AFW Pump Start Page 1 of 4 Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 4, Rev 0 Op-Test No.: 2016-1 Examiners: ____________________________ Operators: _____________________________ ____________________________ _____________________________ ____________________________ _____________________________ Initial Conditions: 2% Turnover: The Plant is being maintained at 2% power prior a shutdown. The crew performing the reactor shutdown is receiving Just-In-Time Training on the Simulator and expected to be back within the hour. AEPS is OOS for breaker repair on PB0501. The crew is to maintain plant conditions until the oncoming crew completes Just In Time Training. Event No. Malf. No. Event Type* Event Description 1 HWXST1E21A SRO NE01 Starting Air Receiver air pressure low (Tech Spec) 2 NIS02B SRO (I) RO (I) Intermediate Range Channel Failure (Tech Spec) 3 MSS09A SRO (C) BOP (C) Steam Dump Valves fail open 4 Lossofswitchyard.lsn SRO (C) RO (C) BOP (C) Loss of Offsite Power 5 PEF01B SRO (M) RO (M) BOP (M) Loss of All AC Power 6 NE01 SRO (C) BOP (C) Loss of All AC) A ESW pump fails to AUTO start 7 PCV455A SRO (C) RO (C) PZR PORV PCV-455 Fails Open with Manual Control Available * (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) 7 2. Malfunctions after EOP entry (1-2) 2 3. Abnormal events (2-4) 3 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 Scenario Event Description Callaway 2016-1 NRC Scenario #4, rev. 0 Page 2 of 4 The Plant is being maintained at 2% power prior a shutdown. The crew performing the reactor shutdown is receiving Just-In-Time Training on the Simulator and expected to be back within the hour. AEPS is OOS for breaker repair on PB0501. The crew is to maintain plant conditions until the oncoming crew completes Just In Time Training. Once the crew takes the watch, the Secondary OT reports a worker accidently lowered the air r receivers are at 300 psig. The SRO reviews the applicable TS for the EDG air receivers, Tech Spec 3.8.3 applies. After Tech Specs have been addressed, the Intermediate Range channel N36 will fail low. The operator will respond in accordance with OTO SE-and, Tech Spec 3.3.1 applies. After Tech Specs have been addressed, Steam Dump Valves fail open. The operator will respond in accordance with OTO-AB-required to close the valves manually to control the cooldown. After the Steam Dumps have been closed, a fault at the Montgomery substation results in a loss of all offsite power. The reactor does not automatically trip (RCP loss) since power is below the P-7 setpoint. However, it should be manually tripped when it is realized that no RCPs are running. The crew should implement E-0, Reactor Trip or Safety Injection. Emergency Diesel Generator NE02 starts and energizes Essential Bus NB02, but ESW Pump B trips upon manual start attempt. NE02 trips 10 minutes after starting due to lack of cooling water if it is not manually secured by the crew. The crew should enter ECA-0.0, Loss of All AC Power. When NB02 is deenergized, PZR PORV BB PCV 455A fails partially open. The crew should close the failed PORV in step 3 of ECA-0.0. The crew should begin making attempts to reenergize one of the busses by dispatching operators to locally check the EDG. 5 minutes after the loss of NB02, the crew can start the A EDG locally. After the A EDG is started and energizes NB01, the A ESW pump will fail to AUTO start and must be manually started. The scenario is complete when the crew has transitioned out of ECA-0.0.

Scenario Event Description Callaway 2016-1 NRC Scenario #4, rev. 0 Page 3 of 4 Critical Tasks: Critical Tasks Manually close the Open PORV before completing Step 3 of ECA-0.0 Manually start A ESW pump prior to A EDG tripping on high temperature. EVENT 7 6 Safety significance The open PORV greatly increases the rate at which RCS inventory is depleted, at a time when the lost inventory cannot be replaced by active injection. Thus, failure to close the PORV defeats the basic purpose of ECA-0.0. Additionally, it is critical that the PORV be closed as soon as possible. Hence, manual closure of the PORV (when the PORV is open and RCS pressure is less than [the setpoint for automatic closure]4) is imperative and urgent in order to ensure the effectiveness of subsequent actions in extending the time to core uncovery. Failure to manually start the SW pump under the postulated plant conditions means that the EDG is running without SW cooling. Running the EDG without SW cooling leads to a high-temperature condition that can result in EDG failure due to damage caused by engine overheating. Under the postulated plant conditions, the running EDG is the only operable EDG. Thus, failure to perform the critical task constitutes misoperation or incorrect crew performance in which the cemergency power Cueing Indication and/or annunciation of station blackout Valve position indication and/or annunciation that the PRZR PORV is open Indication that RCS pressure is below the setpoint at which the PRZR PORV should reclose automatically Indication and/or annunciation of decreasing RCS pressure Indication and/or annunciation consistent with the discharge of PRZR fluid to the PRT PRT temperature, level, pressure Tailpipe RTDs and/or acoustic monitors Indication and/or annunciation that one ac emergency bus is energized by an EDG Bus-energized lamp illuminated Circuit breaker position lamps indicate breaker closed Bus voltage indication shows nominal voltage present EDG status AND Indication and/or annunciation that no SW pump is running Control switch indication that the circuit breakers or contactors for all SW pumps are open SW pump discharge pressure indicator reads zero SW flow indicator reads zero Performance indicator Manipulation of controls as required to close the PRZR PORV PRZR PORV indicates closed Manipulation of controls as required to start the SW pump powered from the ac emergency bus energized by the EDG Control switch indication that the circuit breaker or contactor for a SW pump aligned to supply cooling water to the running EDG is closed Performance feedback PRZR pressure stabilizes Indication and/or annunciation that a SW pump is running, aligned to supply cooling water to the running EDG SW low flow condition clear; indication of flow SW low pressure condition clear; indication of pressure Justification for the chosen performance limit This performance standard is imposed because it is imperative and urgent that the PRZR PORV be closed in order for the strategy of ECA-0.0 to succeed. The PORV constitutes a very large leakage path. Leaving it open causes rapid depletion of RCS inventory at a time when that inventory cannot be replaced. In step 3 of ECA-0.0, the crew is directed to check the major RCS outflow paths that could contribute to rapid depletion of RCS inventory. The PRZR PORVs offer the largest potential for RCS inventory loss. Therefore, they are an outflow path that must be checked and, if necessary, closed. If the EDG trips automatically because of an engine over-temperature condition, it means that the station is again blacked out. It also means that the crew failed to start the SW pump manually as directed by ECA-0.0, Step 27 PWR Owners Group Appendix CT-22, Manually close an open PORV during SBO. CT - 25, Manually start SW pump for EDG cooling Page 4 of 4 References OTO-AB-00001, Steam Dump Malfunction OTO SE-00001, Nuclear Instrument Malfunction E-0, Reactor Trip or Safety Injection ES-0.1, Reactor Trip Response ECA-0.0, Loss of All AC Power Tech spec 3.3.1 for RTS Instrumentation Tech spec 3.8.3 for Diesel Starting Air ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions 1. Loss of Offsite Power (T(1)) a. Any Open Pressurizer PORVs Reclose