ML15009A376

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2014-12-Final Outlines
ML15009A376
Person / Time
Site: Callaway Ameren icon.png
Issue date: 12/17/2014
From: Vincent Gaddy
Operations Branch IV
To:
Union Electric Co
laura hurley
References
50-483/14-012
Download: ML15009A376 (57)


Text

Rev 3 ES-401 PWR Examination Outline (RO)

Form ES-401-2 Facility: Callaway Date of Exam:

December 2014 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

4 2

2 N/A 3

4 N/A 3

18 6

2 1

2 0

2 1

3 9

4 Tier Totals 5

4 2

5 5

6 27 10

2.

Plant Systems 1

3 3

2 3

2 2

3 2

3 3

2 28 5

2 1

0 1

2 1

1 1

1 1

1 0

10 3

Tier Totals 4

3 3

5 3

3 4

3 4

4 2

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

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 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).

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 and justified; 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 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.

Rev 3 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)

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

EA1.04 Ability to operate and monitor the following as they apply to a reactor trip:

RCP operation and flow rates.

(CFR 41.7 / 45.5 / 45.6) 3.6 43 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 X

EK3.15 Knowledge of the reasons for the following responses as the apply to the small break LOCA: Closing of RCP thermal barrier outlet valves.

(CFR 41.5 / 41.10 / 45.6 / 45.13) 3.2 44 000011 Large Break LOCA / 3 X

EK2.02 Knowledge of the interrelations between the Large Break Loca and the following: Pumps.

(CFR 41.7 / 45.7) 2.6 45 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 X

AA2.04 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: How long PZR level can be maintained within limits.

(CFR 43.5/ 45.13) 2.9 46 000025 Loss of RHR System / 4 X

AK1.01 Knowledge of the operational implication of the following concepts as they apply to A loss of Residual Heat Removal System: Loss of RHRS during all modes of operation.

(CFR 41.8 / 41.10 / 45.3) 3.9 47 000026 Loss of Component Cooling Water / 8 X

AA1.01 Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water:

CCW temperature indications.

(CFR 41.7 / 45.5 / 45.6) 3.1 48 000027 Pressurizer Pressure Control System Malfunction / 3 X

AA2.03 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: Effects of RCS pressure changes on key components in plant.

(CFR: 43.5 / 45.13) 3.3 49 000029 ATWS / 1 X

EK1.01 Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermo-hydraulics behavior.

(CFR 41.8 / 41.10 / 45.3) 2.8 50

Rev 3 000038 Steam Gen. Tube Rupture / 3 X

EA1.19 Ability to operate and monitor the following as they apply to a SGTR: MFW System status indicator.

(CFR 41.7 / 45.5 / 45.6) 3.4 9

000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture - Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 X

EK1.01 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : Effect of battery discharge rates on capacity.

(CFR 41.8 / 41.10 / 45.3) 3.3 10 000056 Loss of Off-site Power / 6 X

AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer (CFR 41.5,41.10 / 45.6 / 45.13) 3.5 11 000057 Loss of Vital AC Inst. Bus / 6 X

AA2.03 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: RPS panel alarm annunciators and trip indicators.

(CFR: 43.5 / 45.13) 3.7 12 000058 Loss of DC Power / 6 X

2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12) 3.6 13 000062 Loss of Nuclear Svc Water / 4 X

AA2.02 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss.

(CFR: 43.5 / 45.13) 2.9 14 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 X

EK1.2 Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment): Normal, abnormal and emergency operating procedures associated with (LOCA Outside Containment).

(CFR: 41.8 / 41.10, 45.3) 3.5 15 W/E11 Loss of Emergency Coolant Recirc. / 4 X

EK2.1 Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) 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) 3.6 16

Rev 3 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 X

2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12) 4.0 17 000077 Generator Voltage and Electric Grid Disturbances / 6 X

2.1.27 Knowledge of system purpose and/or function.

(CFR: 41.7) 3.9 18 K/A Category Totals:

4 2

2 3

4 3

Group Point Total:

18

Rev 3 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 X

AK1.02 Knowledge of the operatinal implicaitons of the following concepts as they apply to Dropped Control Rod: Effects of turbine-reactor power mismatch on rod control.

3.1 26 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 X

AK2.01 Knowledge of the interrelations between Emergency Boration and the following: Valves.

(CFR 41.7 / 45.7) 2.7 19 000028 Pressurizer Level Malfunction / 2 X 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 20 000032 Loss of Source Range NI / 7 X

AA2.02 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Expected change in source range count rate when rods are moved.

(CFR: 43.5 / 45.13) 3.6 21 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 X

AA1.04 Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak:

Condensate air ejector exhaust radiation monitor and failure indicator.

(CFR 41.7 / 45.5 / 45.6) 3.6 22 000051 Loss of Condenser Vacuum / 4 X 2.1.20 Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12) 4.6 23 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 X 2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13) 3.6 24 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8

Rev 3 000069 (W/E14) Loss of CTMT Integrity / 5 X

EA1.3 Ability to operate and / or monitor the following as they apply to the (High Containment Pressure):

Desired operating results during abnormal and emergency situations.

(CFR: 41.7 / 45.5 / 45.6) 3.3 25 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 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 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 X

EK2.2 Knowledge of the interrelations between the (Pressurized Thermal Shock) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

(CFR: 41.7 / 45.7) 3.6 27 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

1 2

0 2

1 3

Group Point Total:

9

Rev 3 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

A1.09 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including:

Seal flow and D/P (CFR: 41.5 / 45.5) 2.8 28 003 Reactor Coolant Pump X

2.1.28 Knowledge of the purpose and function of major system components and controls.

(CFR: 41.7) 4.1 29 004 Chemical and Volume Control X

A2.25 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Uncontrolled boration or dilution.

(CFR: 41.5/ 43.5 / 45.3 / 45.5) 3.8 30 005 Residual Heat Removal X

A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: Heatup/cooldown rates (CFR: 41.5 / 45.5) 3.5 31 006 Emergency Core Cooling X

K5.06 Knowledge of the operational implications of the following concepts as they apply to ECCS: Relationship between ECCS flow and RCS pressure.

(CFR: 41.5 / 45.7) 3.5 32 007 Pressurizer Relief/Quench Tank X

A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal pressure in the PRT.

(CFR: 41.5 / 43.5 / 45.3 / 45.13) 2.6 33

Rev 3 008 Component Cooling Water X

A4.01 Ability to manually operate and/or monitor in the control room: CCW indications and controls.

(CFR: 41.7 / 45.5) 3.3 34 008 Component Cooling Water X

K1.05 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems:

Sources of makeup water.

(CFR: 41.2 to 41.9 / 45.7 to 45.9) 3.0 35 010 Pressurizer Pressure Control X

K2.03 Knowledge of bus power supplies to the following:

Indicator for PORV position (CFR: 41.7) 2.8 36 012 Reactor Protection X

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

Trip setpoint adjustment.

(CFR: 41.5 / 45.5) 2.9 37 012 Reactor Protection X

K3.03 Knowledge of the effect that a loss or malfunction of the RPS will have on the following:

SDS.

(CFR: 41.7 / 45.6) 3.1 38 013 Engineered Safety Features Actuation X

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) 2.7 39 022 Containment Cooling X

A4.01 Ability to manually operate and/or monitor in the control room: CCS fans.

(CFR: 41.7 / 45.5 to 45.8) 3.6 40 025 Ice Condenser 026 Containment Spray X

K4.04 Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following:

Reduction of temperature and pressure in containment after a LOCA by condensing steam, to reduce radiological hazard, and protect equipment from corrosion damage (spray).

(CFR: 41.7) 3.7 41

Rev 3 039 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: 441.5 / 45.7) 3.6 42 059 Main Feedwater X

A3.04 Ability to monitor automatic operation of the MFW, including: Turbine driven feed pump.

(CFR: 41.7 / 45.5) 2.5 1

061 Auxiliary/Emergency Feedwater X

2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13) 3.6 2

061 Auxiliary/Emergency Feedwater X

K1.01 Knowledge of the physical connections and/or causeeffect relationships between the AFW and the following systems: S/G system.

(CFR: 41.2 to 41.9 / 45.7 to 45.8) 4.1 3

062 AC Electrical Distribution X

K4.03 Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following:

Interlocks between automatic bus transfer and breakers.

(CFR: 41.7) 2.8 4

062 AC Electrical Distribution X

K2.01 Knowledge of bus power supplies to the following: Major system loads.

(CFR: 41.7) 3.3 5

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 6

064 Emergency Diesel Generator X

K6.07 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers.

(CFR: 41.7 / 45.7) 2.7 7

064 Emergency Diesel Generator X

K3.02 Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following: ESFAS controlled or actuated systems.

(CFR: 41.7 / 45.6) 4.2 8

Rev 3 073 Process Radiation Monitoring X

K1.01 Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: Those systems served by PRMs.

(CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.6 51 076 Service Water X

A3.02 Ability to monitor automatic operation of the SWS, including: Emergency heat loads.

(CFR: 41.7 / 45.5) 3.7 52 076 Service Water X

K2.01 Knowledge of bus power supplies to the following: Service water.

(CFR: 41.7) 2.7 53 078 Instrument Air X

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

Cross-over to other air systems.

(CFR: 41.7) 3.2 54 103 Containment X

A4.01 Ability to manually operate and/or monitor in the control room: Flow control, pressure control, and temperature control valves, including pneumatic valve controller..

(CFR: 41.7 / 45.5 to 45.8) 3.2 55 K/A Category Point Totals:

3 3

2 3

2 2

3 2

3 3

2 Group Point Total:

28

Rev 3 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

K4.14 Knowledge of CRDS design feature(s) and/or interlock(s) which provide for the following: Operation parameters, including proper rod speed (CFR: 41.7) 2.6 56 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor X

K1.02 Knowledge of the physical connections and/or causeeffect relationships between the ITM system and the following systems:

RCS.

(CFR: 41.2 to 41.9 / 45.7 to 45.8) 3.3 57 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control X

K5.01 Knowledge of the operational implications of the following concepts as they apply to the HRPS: Explosive hydrogen concentration.

(CFR: 41.5 / 45.7) 3.4 58 029 Containment Purge X

A1.02 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment Purge System controls including: Radiation levels.

(CFR: 41.5 / 45.5) 3.4 59 033 Spent Fuel Pool Cooling X

K4.01 Knowledge of design feature(s) and/or interlock(s) which provide for the following:

Maintenance of spent fuel level.

(CFR: 41.7) 2.9 60 034 Fuel Handling Equipment

Rev 3 035 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 61 041 Steam Dump/Turbine Bypass Control X

K3.01 Knowledge of the effect that a loss or malfunction of the SDS will have on the following: S/G.

(CFR: 41.7 / 45.6) 3.2 62 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring X

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

Check source for operability demonstration.

(CFR: 41.7 / 45.5 to 45.8) 3.1 63 075 Circulating Water X

A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of Circulating water pumps.

(CFR: 41.5 / 43.5 / 45.3 / 45.13) 2.5 64 079 Station Air 086 Fire Protection X

A3.03 Ability to monitor automatic operation of the Fire Protection System including: Actuation of fire detectors.

(CFR: 41.7 / 45.5) 2.9 65 K/A Category Point Totals:

1 0

1 2

1 1

1 1

1 1

0 Group Point Total:

10

Rev 3 ES-401 Generic Knowledge and Abilities Outline (Tier 3) (RO)

Form ES-401-3 Facility: Callaway Date of Exam: December 2014 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) 2.9 66 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

(CFR: 41.10 / 45.12) 3.4 67 2.1.34 Knowledge of primary and secondary plant chemistry limits.

(CFR: 41.10 / 43.5 / 45.12) 2.7 68 Subtotal 3

2.

Equipment Control 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems.

(CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.9 69 2.2.13 Knowledge of tagging and clearance procedures.

(CFR: 41.10 / 45.13) 4.1 70 2.2.43 Knowledge of the process used to track inoperable alarms.

(CFR: 41.10 / 43.5 / 45.13) 3.0 71 Subtotal 3

3.

Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10) 3.2 72 2.3.11 Ability to control radiation releases.

(CFR: 41.11 / 43.4 / 45.10) 3.8 73 Subtotal 2

4.

Emergency Procedures /

Plan 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

(CFR: 41.10 / 43.5 / 45.13) 4.2 74 2.4.43 Knowledge of emergency communications systems and techniques.

(CFR: 41.10 / 45.13) 3.2 75 Subtotal 2

Tier 3 Point Total 10 7

Rev 1 Page 1 of 2 ES-401 Record of Rejected K/As (RO)

Form ES-401-4 Random Selection method was performed for re-selecting K/As using pennies with dates ending in 0-9. The pennies were drawn in sequence to designate K-1 thru A4, then drawn to pick the number of the K/A in sequence. Pennies 1-9 were drawn from, then 0-9, then 0-9 again.

Generic K/A replacements were drawn from the generic section using the same number selection method. If a K/A could be selected from within the same System and K/A section (ie. A1, K2, etc.) then only the last two numbers were reselected. If a K/A could not be replaced within the same section, then a new section and K/A number were selected within the same system using the method listed above. If a no usable K/A could be selected within the system, then a new system, and K/A was selected from within the Same Tier and Group ensuring no resampling of systems within the Tier and Group unless all Systems were already sampled.

Tier /

Group Randomly Selected K/A Reason for Rejection 1 / 1 009 EK3.15 009 EK3.06 - Unable to write question to meet K/A with plausible distractors. Random selection from within 009 EK3.

1 / 1 011 EK2.02 011 EK1.01 - Audit Overlap. Unable to write another question to K/A and maintain separation from NRC and AUDIT exams.

Random selection from within topic 00011.

1 / 1 025 AK1.01 025 AK3.03 - Site specific procedure for Loss of RHR does not have any Immediate Operator Actions. Cannot write a question for this K/A. Random selection from within topic 00025 1 / 2 032 AA2.02 032 AA2.03 - Unable to write discriminating question pertaining to the automatic removal of high voltage to SRNIs. Random selection from within 032 AA2.

1 / 2 003 AK1.02 W/E15 EK1.2 - Rejected due to oversampling. Multiple other K/As all apply to the Containment. Random selection from within unsampled Tier 1/ Group 2 topics, with random K/A sampled within that Topic.

2 / 1 008 A4.01 008 A4.07 - Question redundancy on this sample plan. CCW surge tank control is sampled twice on this sample plan. Random selection from within 008 A4.

2 / 1 022 A4.01 022 A4.02 - Rejected due to equipment not applicable to Callaway.

The site does not have CCS pumps. Random selection from within 022 A4.

2 / 1 061 2.2.38 061 2.4.2 - Unable to write question to this generic K/A as it would apply to the AFW system. Random selection from Generic K/A section.

2 / 1 062 K4.03 062 K4.07 - Unable to write question to meet K/A with plausible distractors. Random selection from within 062 K4.

Rev 0 Page 2 of 2 ES-401 Record of Rejected K/As (RO)

Form ES-401-4 Tier /

Group Randomly Selected K/A Reason for Rejection 2 / 1 103 A4.01 103 A4.04 - Audit Overlap. Unable to write another question to K/A and maintain separation from NRC and AUDIT exams.

Random selection from within 103 A4.

2 / 2 001 K4.14 001 K4.08 - Rejected due to equipment not applicable to Callaway.

The site does not have interlocks preventing excessive rod movement. Interlocks for Rod control that were looked at for question writing could be better tied to other K/As. Random selection from within 001 K4.

2 / 2 072 A4.03 072 A2.02 - Rejected due to equipment not applicable to Callaway.

Area Radiation Monitors do not cause any automatic actuations; therefore a failure of an ARM has no impact on plant operation.

Random selection from within topic 072.

2 / 2 075 A2.02 075 K2.03 - Rejected due to oversampling. Other K/A in sample plan tests the same knowledge, Power supply to Service Water Pumps. Random selection from within topic 075.

G G 2.2.39 G2.2.7 - Unable to write RO level question to meet K/A. Random selection from within Generic K/As.

G G 2.2.43 G 2.2.41 - Unable to write question to meet K/A. K/A is better suited to a JPM. Initially Random selection from within Generic K/As resulted in selection of G2.2.44. Could not write Generic Question not associated with specific system to G2.2.44. Randomly reselected within Generic K/As.

Rev 2 ES-401 PWR Examination Outline (SRO)

Form ES-401-2 Facility: Callaway Date of Exam:

December 2014 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

1 1

3 Tier Totals 38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

2 2

1 2

Note:

1.

Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only 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).

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 and justified; 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 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.

Rev 2 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 (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

2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12) 4.3 76 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 X

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) 4.7 77 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 000038 Steam Gen. Tube Rupture / 3 000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture - Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 X

2.2.12 Knowledge of surveillance procedures.

(CFR: 41.10 / 45.13) 4.1 78 000057 Loss of Vital AC Inst. Bus / 6 X

AA2.19 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus.

(CFR: 43.5 / 45.13) 4.3 79 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 X

AA2.04 Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Typical conditions which could cause a compressor trip (ie. high temperature)

(CFR: 43.5 / 45.13) 2.7 80

Rev 2 W/E04 LOCA Outside Containment / 3 X

EA2.1 Ability to determine and interpret the following as they apply to the (LOCA Outside Containment): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(CFR: 43.5 / 45.13) 4.3 81 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals:

3 3

Group Point Total:

6

Rev 2 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 (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 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 X

AA2.09 Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Conditions which allow bypass of an intermediate-range level trip switch.

(CFR: 43.5 / 45.13) 3.7 82 000036 (BW/A08) Fuel Handling Accident / 8 X 2.4.41 Knowledge of the emergency action level thresholds and classifications.

(CFR: 41.10 / 43.5 / 45.11) 4.6 83 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 W/E15 Containment Flooding / 5 X 2.4.18 Knowledge of the specific bases for EOPs.

(CFR: 41.10 / 43.1 / 45.13) 4.0 84 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

Rev 2 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 X

EA2.2 Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

(CFR: 43.5 / 45.13) 4.1 85 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals:

2 2

Group Point Total:

4

Rev 2 ES-401 4

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (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

2.4.6 Knowledge of EOP Mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) 4.7 90 004 Chemical and Volume Control 005 Residual Heat Removal X

A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation (CFR: 41.5 / 43.5 / 45.3 / 45.13) 2.9 86 006 Emergency Core Cooling X

A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of flow path.

(CFR: 41.5 / 45.5) 4.3 87 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features Actuation X

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) 4.7 88 022 Containment Cooling 025 Ice Condenser

Rev 2 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution X

A2.10 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Effects of switching power supplies on instruments and controls (CFR: 41.5 / 43.5 / 45.3 / 45.13) 3.3 89 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals:

3 2

Group Point Total:

5

Rev 2 ES-401 5

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (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 X 2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1) 4.5 91 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 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 92 034 Fuel Handling Equipment X

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

(CFR: 41.7) 4.4 93 035 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

Rev 2 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals:

1 1

1 Group Point Total:

3

Rev 2 ES-401 Generic Knowledge and Abilities Outline (Tier 3) (SRO)

Form ES-401-3 Facility: Callaway Date of Exam: December 2014 Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.35 Knowledge of the fuel-handling responsibilities of SROs.

(CFR: 41.10 / 43.7) 3.9 94 2.1.13 Knowledge of facility requirements for controlling vital/controlled access.

(CFR: 41.10 / 43.5 / 45.9 / 45.10) 3.2 95 Subtotal 2

2.

Equipment Control 2.2.20 Knowledge of the process for managing troubleshooting activities.

(CFR: 41.10 / 43.5 / 45.13) 3.8 96 2.2.5 Knowledge of the process for making design or operating changes to the facility.

(CFR: 41.10 / 43.3 / 45.13) 3.2 97 Subtotal 2

3.

Radiation Control 2.3.6 Ability to approve release permits.

(CFR: 41.13 / 43.4 / 45.10) 3.8 98 Subtotal 1

4.

Emergency Procedures /

Plan 2.4.27 Knowledge of fire in the plant procedures.

(CFR: 41.10 / 43.5 / 45.13) 3.9 99 2.4.37 Knowledge of the lines of authority during implementation of the emergency plan.

(CFR: 41.10 / 45.13) 4.1 100 Subtotal 2

Tier 3 Point Total 7

Rev 1 Page 1 of 1 ES-401 Record of Rejected K/As (SRO)

Form ES-401-4 Random Selection method was performed for re-selecting K/As using pennies with dates ending in 0-9. The pennies were drawn in sequence to designate K-1 thru A4, then drawn to pick the number of the K/A in sequence. Pennies 1-9 were drawn from, then 0-9, then 0-9 again.

Generic K/A replacements were drawn from the generic section using the same number selection method. If a K/A could be selected from within the same System and K/A section (ie. A1, K2, etc.) then only the last two numbers were reselected. If a K/A could not be replaced within the same section, then a new section and K/A number were selected within the same system using the method listed above. If a no usable K/A could be selected within the system, then a new system, and K/A was selected from within the Same Tier and Group ensuring no resampling of systems within the Tier and Group unless all Systems were already sampled.

Tier /

Group Randomly Selected K/A Reason for Rejection 1 / 1 065 AA2.04 065 AA2.05 - Rejected due to not having direction to shut down the plant within the Loss of Instrument Air procedure at Callaway.

Random selection from within 065 AA2.

1 / 2 033 AA2.09 005 AA2.03 - Overlap with Audit Exam. Unable to write another question to meet K/A and maintain separation from NRC and AUDIT exams. Random selection from within unsampled topics within Tier 1 / Group 2.

2 / 1 006 A2.02 008 A2.02 - Rejected due to oversampling of the CCW system.

Random selection from within unsampled topics within Tier 2 /

Group 1.

2 / 1 003 2.4.6 103 2.4.6 - Rejected due to oversampling of the Containment system. Using the same Generic K/A, Random selection from the unsampled topics within Tier 2 / Group 1.

2 / 2 034 K4.02 068 2.1.23 - Rejected due to overlap with generic K/A section.

Limited required knowledge at Callaway pertaining to release permits. Unable to write question to this topic and K/A. Random selection from the unsampled topics within Tier 2 / Group 2, with random selection of new K/A number. Because system 034 (Fuel Handling) was selected the K/A was randomly selected from ALL K/As.

ES-301 Administrative Topics Outline Form ES-301-1 Rev 2 NUREG-1021, Revision 9 Facility:

Callaway Date of Examination:

12/8/2014 Examination Level:

RO Operating Test Number:

2014-1 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations A1 R, M 2.1.25 (3.9)

Ability to interpret reference materials, such as graphs, curves, tables, etc.

JPM:

Determine Time to Boil for a Loss of Shutdown Cooling Conduct of Operations A2 R, D 2.1.18 (3.6)

Ability to make accurate, clear, and concise logs, records, status boards, and reports.

JPM:

Complete RCS Inventory Balance Equipment Control A3 R, D, P*

2.2.12 (3.7)

Knowledge of surveillance procedures.

JPM:

Review Pump Run Data and Determine if Acceptance Criteria are Met Radiation Control A4 R, M 2.3.14 (3.4)

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

JPM:

Determine if Respirator Needs to be Worn 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 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)

(S)imulator

  • The JPMs from the last 2 exams were randomly selected by placing 8 slips of paper labeled A1.a 2011 through A4 2013 in a hardhat. A2 2011was drawn from the hardhat.

ES-301 Administrative Topics Outline Form ES-301-1 Rev 2 NUREG-1021, Revision 9 A1 This is a MODIFIED JPM. The parent JPM was used on the 2011 NRC SRO exam. The initial conditions (days after shutdown and RCS temperature) were changed. These changes result in different Figures being used for the time to boil determination. This results in the time to boil to change from the parent JPM. The applicant will be assigned the task of determining time to boil given a loss of RHR IAW OTO-EJ-00003, Loss of RHR While Operating at Reduced Inventory or Mid-Loop Conditions. Upon completion of this JPM the operator will have determined the Time to Boil for Condition 1 and Condition

2.

A2 This is a BANK JPM. The applicant will be assigned the task of determining the RCS leakage using OSP-BB-00009 ADD 1, RCS Inventory Balance Excessive Leakage or Manual Calculation. Upon completion of the task the Candidate will have identified Total RCS leakage is approximately 0.6809 gpm, (range 0.680 to 0.681 gpm) Identified RCS Leakage is 0.0764 gpm, and Unidentified RCS Leakage is approximately 0.6045 gpm (range 0.604 to 0.605 gpm).

A3 This BANK JPM was used on the 2011 ILT NRC Exam. The applicant will be assigned the task of completing the calculations and determining if the Acceptance Criteria has been satisfied in accordance with OSP-EN-P001B, TRAIN B CONTAINMENT SPRAY PUMP INSERVICE TEST. Upon completion of this JPM the operator will have calculated pump flow and D/P, completed Attachment 4 of OSP-EN-P001B, TRAIN B CONTAINMENT SPRAY PUMP INSERVICE TEST, and determined that the acceptance criteria is met for the B Containment Spray Pump.

A4 This is a MODIFIED JPM. The parent JPM (Set 1 G1 A.3) has not been used on an NRC Exam administered at Callaway between 2004 and 2013. The initial conditions (dose rate and internal dose without a respirator) were changed. In the parent JPM airborne dose was given to the applicant. In the modified JPM the applicant must obtain the date from a survey map and convert from DAC to mrem. The applicant will be assigned the task of determining if personnel working on the Spent Fuel Pool rack should wear a respirator.

Upon completion of this JPM the operator will have calculated dose with and without a respirator and selected performing the job with a respirator.

ES-301 Administrative Topics Outline Form ES-301-1 Rev 2 Page 1 of 2 Facility:

Callaway Date of Examination:

12/8/2014 Examination Level:

SRO Operating Test Number:

2014 - 1 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations A5 R, N 2.1.18 (3.8)

Ability to make accurate, clear, and concise logs, records, status boards, and reports.

JPM:

Determine Reportability for a Required Shutdown Conduct of Operations A6 R, D, P*

2.1.37 (4.6)

Knowledge of procedures, guidelines, or limitations associated with reactivity management.

JPM:

Review ECP Calculation Equipment Control A7 R, D 2.2.40 (4.7)

Ability to apply Technical Specifications for a system.

JPM:

Determine Applicable Technical Specifications Radiation Control A8 R, M 2.3.14 (3.8)

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

JPM:

Determine if Respirator Needs to be Worn Emergency Procedures/Plan A9 R or S, D 2.4.43 (3.8)

Knowledge of emergency communications systems and techniques JPM:

Make appropriate Offsite Notifications during an ALERT 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)

  • The JPMs from the last 2 exams (the 2013 re-exam was also included) were randomly selected by placing 15 slips of paper labeled A1.a 2011 through A4 2013R in a hardhat. A1.a 2013Rwas drawn from the hardhat.

ES-301 Administrative Topics Outline Form ES-301-1 Rev 2 Page 2 of 2 A5 This is a NEW JPM. The applicant will be assigned the task of determining the initial reportability requirements of a TS required shutdown per APA-ZZ-00520, REPORTING REQUIREMENTS AND RESPONSIBILITIES. Upon completion of this JPM the operator will have determined that a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report is required to be made to the NRC Operation Center.

A6 This BANK JPM was used on the 2013 ILT NRC Re-Exam. The applicant will be assigned the task of reviewing and approving the completed Attachment 1 of OSP-SF-00005, Estimated Critical Position Calculation using the printouts and information provided by the Reactor engineer. Upon completion of this JPM, the Applicant will have identified the following errors in Attachment 1 of OSP-SF-00005 Estimated Critical Position Calculation:

1.

Step 6.8.1 incorrect Anticipated RCS avg temp at startup

2.

Step 6.12.2 incorrect Differential boron worth

3.

Step 6.14.4 incorrect Maximum Rod Height A7 This is a BANK JPM. The applicant will be assigned the task of determining the impact of hanging a Hold Off tag on plant safety systems and identifying all applicable Technical Specifications Limiting Condition for Operation (LCO) entries that will apply. Upon completion of the JPM the candidate should determine that the applicable T/S are 3.7.4.A and 3.7.5.C.

A8 This is a MODIFIED JPM. The parent JPM (Set 1 G1 A.3) has not been used on an NRC Exam administered at Callaway between 2004 and 2013. The initial conditions (dose rate and internal dose without a respirator) were changed. In the parent JPM airborne dose was given to the applicant. In the modified JPM the applicant must obtain the date from a survey map and convert from DAC to mrem. The applicant will be assigned the task of determining if personnel working on the Spent Fuel Pool rack should wear a respirator.

Upon completion of this JPM the operator will have calculated dose with and without a respirator and selected performing the job with a respirator.

A9 This is a BANK JPM. The applicant will be assigned the task of determining the Emergency Event Classification, completing the Sentry Notification form, and performing the notification of off-site agencies within the allotted amount of time. Upon completion of this JPM the operator will have determined the event classification to be an Alert based on EAL HA-1.2, Tornado and damage to table H-1 system, within 15 minutes of start and completed the Sentry notification form and sent it within 15 minutes of the event classification time.

ES-301 Control Room/In-Plant Systems Outline (Rev 2)

Form ES-301-2 Page 1 of 4 Facility: ____Callaway_____________________

Date of Examination: _12/8/14_____

Exam Level: RO SRO-I SRO-U Operating Test No.: __2014-1____

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code*

Safety Function S1 004 CVCS (BG) / Perform a Dilution N, S 1

S2 006 ECCS (SB) / Initial RCS Depressurization and SI Block L, M, S 2

S3 010 Reactor Coolant System (BB) / Perform System Surveillance -

BBHV8000A Stroke Test D, P*, S 3

S4 003 Reactor Coolant Pump System (BB) / Start Reactor Coolant Pump during RCS Natural Circulation Cooldown A, M, S 4P S5 026 Containment Spray (EN) / Align Containment Spray for Recirculation A, D, EN, S 5

S6 064 SR Elect Gen and Dist (NE) / Manually Start Diesel Generators A, M, S 6

S7 016 Non-Nuclear Instrumentation System (NN) / Selection of available Instrumentation N, S 7

S8 008 Component Cooling Water System (EG) / Respond to a loss of CCW to the RCPs A, D, S 8

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1 068 Control Room Evacuation / Operate the PZR Heaters at the Aux Shutdown Panel E, M, R 3

P2 015 Reactor Coolant Pump Malfunction / Local RCP Seal Isolation A, D, E, R 4P P3 068 Control Room Evacuation / Locally Trip Reactor Trip and Bypass Breakers E, M, R 1

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 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 (Rev 2)

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 9 / 8 / 4 1 / 1 / 1

- / - / 1 (control room system) 1 / 1 / 1 2 / 2 / 1 3 / 3 / 2 (randomly selected) 1 / 1 / 1

  • The JPMs from the last 2 exams were randomly selected by placing 22 slips of paper labeled a through k for the 2011 exam and S1 through P3 from the 2013 exam in a hardhat. Two of these items (S2 and P1) were drawn from the hardhat. S2 was used as S3 on this exam. P1 was originally used as P1 on this exam. P1 was replaced after NRC review of the originally submitted outline.

S1 This is a NEW JPM. The applicant will be assigned the task of diluting the RCS 100 gallons by performing a nominal 120 gpm dilution per OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System. Upon completion of this JPM, the Applicant will have diluted the RCS 100 gallons (acceptable range 97 to 103 gallons).

S2 This is a MODIFIED JPM in a SHUTDOWN condition. The parent JPM (URO-NOP-06-C153J) has not been used on an NRC Exam administered at Callaway between 2004 and 2013. The last revision to this JPM was in 2005. The actions to perform this task are now covered by a different procedure. Additionally the method for lowering RCS pressure has been changed. The applicant will be assigned the task of depressurizing the RCS and blocking the SI signal, per OTG-ZZ-00006 Addendum 04, Initial RCS Depressurization and SI Block. Upon completion of this JPM, the applicant will have blocked both trains of Pressurizer Pressure and Steam Line Pressure Safety Injection without initiating a Safety Injection.

S3 This BANK JPM was used on the 2013 ILT NRC Exam. It was randomly selected using the method described above. The applicant will be assigned the task of performing the Stroke Time Test for BBHV8000A, RCS PZR OUT PWR OPER RLF HV, per OSP-BB-V0001, RCS Valve Inservice Test, Section 6.1. Upon completion of this JPM, the applicant will have completed the Stroke Time Test for BBHV8000A per OSP-BB-V0001 and reported to the CRS that the Acceptance Criteria was not satisfied.

ES-301 Control Room/In-Plant Systems Outline (Rev 2)

Form ES-301-2 Page 3 of 4 S4 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (URO-AEO C052J) has not been used on an NRC Exam administered at Callaway between 2004 and 2013.The parent JPM was not alternate path and the applicant was successful in starting the D RCP. The applicant will be assigned the task of starting D Reactor Coolant Pump in accordance with EOP Addendum 3, Starting An RCP.

The D RCP will not start and the applicant will then start the A (or B) RCP. Upon completion of this JPM, the operator will have started RCP A (or B) and secured the respective RCP Oil Lift pump.

S5 This is an ALTERNATE PATH, BANK JPM. The applicant will be assigned the task of completing the actions of ES-1.3, Transfer to Cold Leg Recirculation, Step 6.

Upon completion of this JPM, the applicant will have aligned the train B Containment Spray Pump to a recirculating lineup and secured the train A Containment Spray Pump.

S6 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (URO-AEO C174J(A)) was used on the 2011 NRC Exam administered at Callaway. The parent JPM had both EDGs start and the B ESW pump start. In the modified JPM the B ESW pump does not start and the NE02 (B EDG) must be secured. The applicant will be assigned the task of completing the actions of Step 5 of ECA-0.0, Loss of All AC Power. Upon completion of this JPM, the applicant will have manually started NE01 and manually started A ESW Pump prior to A DG auto trip.

S7 This is a NEW JPM. The applicant will be assigned the task of selecting the correct instrumentation in accordance with OTO-NN-00001, LOSS OF SAFETY RELATED INSTRUMENT POWER. Upon completion of this JPM, the operator will have selected the proper instrumentation for SG Flow and Level and PZR pressure and level.

S8 This is an ALTERNATE PATH, TIME CRITICAL, BANK JPM. The applicant will be assigned the task of performing Attachment C of OTO-BB-00002, RCP Off-Normal to address a loss of CCW to the RCPs. Upon completion of this JPM, the operator will have restored CCW to Containment (within 10 minutes) and the RCPs.

P1 This is a MODIFIED JPM. The parent JPM (URO-AEO-02-P015J) was last used on the 2007 NRC Exam administered at Callaway. The JPM was used on the 2013 ILT NRC Exam. The parent JPM had the operator taking action to control SG water level, the modified JPM has the operator taking action to control PZR pressure. The operator will be assigned the task of performing OTO-ZZ-00001, Control Room Inaccessibility Attachment G, Control Room Supervisor (CRS) Actions Without A Fire. Upon completion of this JPM the Operator will have demonstrated the ability to control RCS temperature and PZR pressure from the Aux Shutdown Panel.

P2 This is an ALTERNATE PATH, BANK JPM. The operator will be assigned the task of locally isolating RCP Seals using EOP Addendum 22, Local RCP Seal Isolation.

Upon completion of this JPM, the operator will have successfully isolate RCP seals.

ES-301 Control Room/In-Plant Systems Outline (Rev 2)

Form ES-301-2 Page 4 of 4 P3 This is a MODIFIED JPM. The operator will be assigned the task of performing OTO-ZZ-00001, Control Room Inaccessibility Attachment A, Safe Shutdown Operator Actions with Fire. Upon completion of this JPM the operator will have isolated loads on MCC NG04C, placed class 1E electrical equipment A/C unit in ISO/RUN, aligned CCW and locally tripped both reactor trip breakers and bypass breakers.

ES-301 Control Room/In-Plant Systems Outline (Rev 2)

Form ES-301-2 Page 1 of 3 Facility: ____Callaway_____________________

Date of Examination: _12/8/14_____

Exam Level: RO SRO-I SRO-U Operating Test No.: __2014-1____

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code*

Safety Function S1 004 CVCS (BG) / Perform a Dilution N, S 1

S2 006 ECCS (SB) / Initial RCS Depressurization and SI Block L, M, S 2

S3 010 Reactor Coolant System (BB) / Perform System Surveillance -

BBHV8000A Stroke Test D, P*, S 3

S4 003 Reactor Coolant Pump System (BB) / Start Reactor Coolant Pump during RCS Natural Circulation Cooldown A, M, S 4P S5 026 Containment Spray (EN) / Align Containment Spray for Recirculation A, D, EN, S 5

S6 064 SR Elect Gen and Dist (NE) / Manually Start Diesel Generators A, M, S 6

S8 008 Component Cooling Water System (EG) / Respond to a loss of CCW to the RCPs A, D, S 8

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1 068 Control Room Evacuation / Operate the PZR Heaters at the Aux Shutdown Panel E, M, R 3

P2 015 Reactor Coolant Pump Malfunction / Local RCP Seal Isolation A, D, E, R 4P P3 068 Control Room Evacuation / Locally Trip Reactor Trip and Bypass Breakers E, M, R 1

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 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 (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 9 / 8 / 4 1 / 1 / 1

- / - / 1 (control room system) 1 / 1 / 1 2 / 2 / 1 3 / 3 / 2 (randomly selected) 1 / 1 / 1

ES-301 Control Room/In-Plant Systems Outline (Rev 2)

Form ES-301-2 Page 2 of 3

  • The JPMs from the last 2 exams were randomly selected by placing 22 slips of paper labeled a through k for the 2011 exam and S1 through P3 from the 2013 exam in a hardhat. Two of these items (S2 and P1) were drawn from the hardhat. S2 was used as S3 on this exam. P1 was originally used as P1 on this exam. P1 was replaced after NRC review of the originally submitted outline.

S1 This is a NEW JPM. The applicant will be assigned the task of diluting the RCS 100 gallons by performing a nominal 120 gpm dilution per OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System. Upon completion of this JPM, the Applicant will have diluted the RCS 100 gallons (acceptable range 97 to 103 gallons).

S2 This is a MODIFIED JPM in a SHUTDOWN condition. The parent JPM (URO-NOP-06-C153J) has not been used on an NRC Exam administered at Callaway between 2004 and 2013. The last revision to this JPM was in 2005. The actions to perform this task are now covered by a different procedure. Additionally the method for lowering RCS pressure has been changed. The applicant will be assigned the task of depressurizing the RCS and blocking the SI signal, per OTG-ZZ-00006 Addendum 04, Initial RCS Depressurization and SI Block. Upon completion of this JPM, the applicant will have blocked both trains of Pressurizer Pressure and Steam Line Pressure Safety Injection without initiating a Safety Injection.

S3 This BANK JPM was used on the 2013 ILT NRC Exam. It was randomly selected using the method described above. The applicant will be assigned the task of performing the Stroke Time Test for BBHV8000A, RCS PZR OUT PWR OPER RLF HV, per OSP-BB-V0001, RCS Valve Inservice Test, Section 6.1. Upon completion of this JPM, the applicant will have completed the Stroke Time Test for BBHV8000A per OSP-BB-V0001 and reported to the CRS that the Acceptance Criteria was not satisfied.

S4 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (URO-AEO C052J) has not been used on an NRC Exam administered at Callaway between 2004 and 2013.The parent JPM was not alternate path and the applicant was successful in starting the D RCP. The applicant will be assigned the task of starting D Reactor Coolant Pump in accordance with EOP Addendum 3, Starting An RCP. The D RCP will not start and the applicant will then start the A (or B) RCP. Upon completion of this JPM, the operator will have started RCP A (or B) and secured the respective RCP Oil Lift pump.

S5 This is an ALTERNATE PATH, BANK JPM. The applicant will be assigned the task of completing the actions of ES-1.3, Transfer to Cold Leg Recirculation, Step

6. Upon completion of this JPM, the applicant will have aligned the train B Containment Spray Pump to a recirculating lineup and secured the train A Containment Spray Pump.

ES-301 Control Room/In-Plant Systems Outline (Rev 2)

Form ES-301-2 Page 3 of 3 S6 This is an ALTERNATE PATH, MODIFIED JPM. The parent JPM (URO-AEO C174J(A)) was used on the 2011 NRC Exam administered at Callaway. The parent JPM had both EDGs start and the B ESW pump start. In the modified JPM the B ESW pump does not start and the NE02 (B EDG) must be secured.

The applicant will be assigned the task of completing the actions of Step 5 of ECA-0.0, Loss of All AC Power. Upon completion of this JPM, the applicant will have manually started NE01 and manually started A ESW Pump prior to A DG auto trip.

S8 This is an ALTERNATE PATH, TIME CRITICAL, BANK JPM. The operator will be assigned the task of performing Attachment C of OTO-BB-00002, RCP Off-Normal to address a loss of CCW to the RCPs. Upon completion of this JPM, the operator will have restored CCW to Containment (within 10 minutes) and the RCPs.

P1 This is a MODIFIED JPM. The parent JPM (URO-AEO-02-P015J) was last used on the 2007 NRC Exam administered at Callaway. The JPM was used on the 2013 ILT NRC Exam. The parent JPM had the operator taking action to control SG water level, the modified JPM has the operator taking action to control PZR pressure. The operator will be assigned the task of performing OTO-ZZ-00001, Control Room Inaccessibility Attachment G, Control Room Supervisor (CRS)

Actions Without A Fire. Upon completion of this JPM the Operator will have demonstrated the ability to control RCS temperature and PZR pressure from the Aux Shutdown Panel.

P2 This is an ALTERNATE PATH, BANK JPM. The operator will be assigned the task of locally isolating RCP Seals using EOP Addendum 22, Local RCP Seal Isolation. Upon completion of this JPM, the operator will have successfully isolate RCP seals.

P3 This is a MODIFIED JPM. The operator will be assigned the task of performing OTO-ZZ-00001, Control Room Inaccessibility Attachment A, Safe Shutdown Operator Actions with Fire. Upon completion of this JPM the operator will have isolated loads on MCC NG04C, placed class 1E electrical equipment A/C unit in ISO/RUN, aligned CCW and locally tripped both reactor trip breakers and bypass breakers.

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 1, Rev 2 Op-Test No.: 2014-1 Examiners: ____________________________ Operators:

Initial Conditions: 100%

Turnover: B CCP is tagged out for Breaker PMTs Eve nt No.

Malf. No.

Event Type*

Event Description 1

PT0456 SRO (I)

RO (I)

PZR Pressure instrument Fails high (Tech Specs) 2 ABFT0542 SRO (I)

BOP (I)

SG Steam Flow Inst Fails low 3

SFB08_DR SRO (C)

RO (C)

BOP (C)

Dropped rod (Tech Specs) 4 SF2CD2/2 SRO (M)

RO (M)

BOP (M)

Multiple Dropped Rods / Rx Trip 5

AB001_A SRO (M)

RO (M)

BOP (M)

Steam Line Break inside containment 6

SAS9XX_1, 2, 3, 4 SRO (C)

BOP (C)

MSIVs fail to close 7

PEN01A_2 and EN08RL_420TVSP SRO (C)

BOP (C)

Containment Spray Pump Auto Start Failure and Containment Spray Pump Discharge Valve 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) 7
2. Malfunctions after EOP entry (1-2) 2
3. Abnormal events (2-4) 3
4. Major transients (1-2) 2
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 2014 NRC Exam Scenario #1, rev. 2 The plant is at 100% with the B CCP tagged out After the crew takes the watch, the lower selected PZR Pressure Channel BB PT-456 fails high.

Annunciator PZR PRESS HI and RX PARTIAL TRIP will alarm to alert the operator of the malfunction.

The CRS should enter OTO-BB-00006, Pressurizer Pressure Control Malfunction, and transfer pressure control to remove the failed channel.

After the Technical Specifications from the previous event have been identified, a Steam Flow Channel on D S/G Fails Low. The crew will respond using, OTO AE 00002, Steam Generator Water Level Control Malfunctions and select SG steam flow channel selector to an operable channel on AB FS-542C and DFWCS.

After the crew has selected away from the failed channel, a dropped rod occurs as indicated by DRPI and Control Rod Alarms. The ROs will perform the immediate actions of OTO-SF-00001 to place rods in manual. The crew will establish conditions for rod recovery identify Technical Specifications and begin restoration. During rod recovery, additional control rods drop into the core. The operators will take immediate action to trip the reactor.

After the Reactor Trip, a Steam Line Break inside containment will occur. The MSIVs will fail to auto close. The crew will respond IAW E-0, Reactor Trip or Safety Injection, and transition to E-2, Faulted Steam Generator Isolation The A Containment Spray Pump fails to auto start and the B Containment Spray Pump discharge valve fails to open. The crew can manually start the A Containment Spray Pump and/or direct an OT to open EN-HV-12, CS PUMP B DISCH, in the field while performing the actions of E-0.

The crew will continue to perform the actions of E-2, and isolate the A SG, then transition to ES-1.1, SI Termination.

The scenario can be terminated after the transition to ES-1.1 occurs

Scenario Event Description Callaway 2014 NRC Exam Scenario #1, rev. 2 Critical Tasks:

Critical Tasks Isolate the faulted A SG before transition out of E-2 Manually actuate containment cooling by:

Starting the 'A' Containment Spray Pump EVENT 5

7 Safety significance 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)

If one train of containment spray is not actuated, 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 actuate at least one train of containment spray under the scenario conditions and when it is possible to do so constitutes a violation of the license condition.

Cueing Both of the following:

Steam pressure and flow rate indications that make it possible to identify A SG as faulted AND Valve position and flow rate indication that AFW continues to be delivered to the faulted A SG Indication that containment cooling is required by Containment pressure either currently or has been greater than 27 psig Indication and/or annunciation that no train of containment spray has successfully actuated by both:

A Containment Spray Pump not running B Containment Spray Pump discharge valve NOT open Performance indicator ISOLATE AFW flow to faulted SG(s):

o CLOSE associated MD AFP Flow Control Valve(s):

AL HK-7A (SG A) o CLOSE associated TD AFP Flow Control Valve(s):

AL HK-8A (SG A)

CLOSE Steamline Low Point Drain valve from faulted SG(s):

o AB HIS-9 (SG A)

FAST CLOSE all MSIVs and Bypass valves:

o AB HS79 o

AB HS80 Starting the 'A' Containment Spray Pump Performance feedback Crew will observe the following:

Any depressurization of intact SGs stops AFW flow rate indication to faulted SG of zero Indication that 'A' Containment Spray Pump is running AND Containment pressure lowering Justification for the chosen performance limit before transition out of E-2 is in accordance with the PWR Owners 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.

before completion of Attachment A of E-0 is in accordance with the PWR Owners 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-17 Isolate faulted SG CT-3 Manually actuate containment cooling

Scenario Event Description Callaway 2014 NRC Exam Scenario #1, rev. 2 References OTO-BG-00001, Pressurizer Level Control Malfunction OTO-AE-00002, Steam Generator Water Level Control Malfunctions OTO-SF-00001, Rod Control/Malfunctions E-0, Reactor Trip or Safety Injection E-2, Faulted Steam Generator Isolation ES-1.1, SI Termination Tech Spec 3.3.1.E & M for Reactor Trip System Instrumentation Tech Spec 3.3.2.D and L for Reactor Trip System Instrumentation Tech spec 3.1.4.B for Rod Group Alignment Limits ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions

1. Main Steam Line Break Inside Containment (T(MSI))
a. Close MSIVs
b. Isolate Feedwater to the affected SG

Scenario Setup Guide:

Establish the initial conditions of IC-10, MOL 100%:

RCS boron concentration 752 ppm CCP A 765 ppm minus 5 days CCP B 775 ppm minus 15 days Rod Control Bank D 215 steps, Other banks 228 steps ENSURE BB-PT-455 / BB-PT-456 is selected on BB-PS-455F B CCP is tagged out with breaker in P-T-L and tag on handswitch

=======SCENARIO PRELOADS / SETUP ITEMS==========

B CCP Tagged out due to Breaker PMTs.

ME Schematics (BG) e23bg01b BG02NB0201_BKRTA_BKPOS, Value = 3 Steam Line Break inside containment / MSIV auto close failure (Event 5)

Insert Malfunctions (AB)AB001_A/3600 klbm per hr, cond rec0009 le 1.0 (reactor trip)

Insert Malfunction SAS9XX_1, 2, 3, 4 - Value = Enable Containment Spray Pump Auto Start Failure and / Containment Spray Pump Discharge Valve Failure (Event 6)

Inhibit start of A Containment Spray Pump Auto Start o

Insert Malfunction (EN) PEN01A_2, Value = 1 Containment Spray Pump B Disch fails to open on CISB - Event 6 o

Go To ME Schematics (EN) e23en03_b, click on 42 o relay in middle of page o

Override open relay EN08RL_42oTVSP, Value = 0

======= EVENT 1============================

PZR Pressure Channel BB PT-456 fails high Insert Malfunction (BB) BBPT0456, Value = 2500

======= EVENT 2============================

D S/G Steam Flow Channel Failure ABFT0542 Insert Malfunction (AB) ABFT0542, value = 0

======= EVENT 3============================

Dropped Rod Insert Malfunctions (SF) SF B08 = DR, Value = stationary gripper

=======EVENT 4==============================

Multiple Dropped Rods / Rx Trip Insert Malfunctions (SF) SF2CD2, Value = 2 rods

=======EVENT 5 PRELOADED============================

Steam Line Break inside containment See Preloads Above

=======EVENT 6 PRELOADED===============

Containment Spray Pump Auto Start Failure and / Containment Spray Pump Discharge Valve Failure See Preloads Above

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 2, rev. 2 Op-Test No.: 2014-1 Examiners: ____________________________ Operators:

Initial Conditions: 10-8 amps, Steady State Conditions.

Turnover: Startup is on hold due to an oil leak identified with NE02. NE02 is tagged out to work on the leak. It is unknown when it will be returned to service. On coming crew will be swapping Service Water Pumps for maintenance. The Outside OT has been briefed and is standing by to swap SW pumps.

Event No.

Malf. No.

Event Type*

Event Description 1

N/A SRO (N)

BOP (N)

Start Service Water Pump C Secure Service Water Pump A 2

N/A SRO (C)

RO (C)

RCS High Activity / Place 120 gpm Letdown in-service (Tech Spec) 3 BBTE0421B SRO (I)

RO (I)

Tavg Channel Failure - Loop 2 Tc fails high (Tech Spec) 4 NB01_F SRO (C)

BOP (C)

Loss of NB01 (Tech Spec) 5 (SB) SB001 SRO (C)

RO (C)

Reactor Trip 6

Lossofswitch yard.lsn SRO (M)

RO (M)

BOP (M)

Loss of all AC Power 7

PAL02_3 SRO (C)

BOP (C)

TDAFW fails to Auto Start 8

PA50101_1 SRO (C)

BOP (C)

AEPs EDGs fail to 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) 1
7. Critical tasks (2-3) 2

Scenario Event Description Callaway 2014 Scenario #2, rev. 0 The plant is operating at 10-8 amps, steady state power. NE02 is tagged out for an oil leak. It is unknown when it will be returned to service. On coming crew will be swapping Service Water Pumps for maintenance.

After the crew takes the watch, the crew will start C Service Water Pump and secure A Service Water Pump IAW OTN-EA-00001, Service Water System.

After the Service Water Pumps are swapped, a report from Chemistry indicates high activity in the RCS. The Crew will enter OTO-BB-00005, RCS High Activity and establish 120 gpm letdown flow IAW OTN-BG-00001 Addendum 04, Operation of CVCS Letdown. The CRS will evaluate Tech Spec 3.4.16.

After Tech Specs have been addressed for the RCS High Activity, Tcold Channel BB TE 421 B fails high. The crew will respond using OTO-BB-00004, RTD Channel Failure, and select an operable channel. The CRS will evaluate Tech Spec 3.3.1.

After the crew has addressed the Tcold Channel trip, NB01 will trip and lockout. The crew should enter OTO-NB-00001, Loss of Power to NB01. OTO-NB-00001 will direct the crew to secure NE01 and verify the electric plant line up.

After NE01 has been secured a Reactor Trip will occur. The crew should respond by entering E-0, Reactor Trip or Safety Injection. The crew will then transition to ES-0.1, Reactor Trip Response.

After the crew has transition to ES-0.1 and prior to Step 6 a Loss of Off Site Power will occur.

The crew will transition to ECA-0.0, Loss of All AC Power. The crew will manually start the TDAFW. When restoring power using EOP Addendum 39, Alternate Emergency Power Supply, the AEPs EDGs will not auto start and the operator must manually start the AEPs EDGs. After the AEPs EDGs are started power is restored to NB02.

The scenario can be terminated after power has been restored to NB02.

Scenario Event Description Callaway 2014 Scenario #2, rev. 0 Critical Tasks:

Critical Tasks Establish greater than 285,000 lbm/hr AFW flow rate to the SGs prior to SG dryout occurring.. (<10% WR on all SGs)

Energize at least one AC Emergency Bus prior to RCP seal degradation (greater than 25 gpm per pump)

EVENT 6

7 Safety significance Failure to establish minimum AFW flow in this scenario is a violation of the basic objective of ECA-0.0 and of the assumptions of the analyses upon which ECA-0.0 is based. Without AFW flow, the SGs could not support any significant plant cooldown.

Thus, the crew would lose the ability to delay the adverse consequences of core uncovery.

In the scenario, failure to energize at least one ac emergency bus results in the needless continuation of a situation in which the pumped ECCS capacity and the emergency power capacity are both in a completely degraded status, as are all other active safeguards requiring electrical power. Although the completely degraded status is 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 Both of the following:

Indication of Station Blackout AND Less than 285,000 lbm/hr AFW flow to the SGs Indication and/or annunciation that all ac emergency buses are de-energized Bus energized lamps extinguished Circuit breaker position Bus voltage EDG status Performance indicator Manipulation of the:

TDAFW Steam Supply valve(s):

AB HIS5A (SG B)

AB HIS6A (SG C)

TDAFP Mechanical Trip/Throttle valve:

FC HIS312A Manipulation of controls as required to energize at least one ac emergency bus from the AEPS Using PBXY0001 CLOSE AEPS FDR BKR PB0502 TO NB0214 o

PB0502 CLOSE NB02 AEPS SUPPLY BKR NB0214 o

NB HIS-68 Performance feedback Crew will observe greater than 285,000 lbm/hr AFW flow to the SGs.

Indication that NB02 is energized NB02 Bus energized light NB02 bus voltage Justification for the chosen performance limit Without AFW flow, decay heat would open the SG safety valves and would rapidly deplete the SG inventory, leading to a loss of secondary heat sink, or SG dryout. Decay heat would then increase RCS temperature and pressure until the pressurizer PORVs open, imposing a larger LOCA than RCP seal leakage.

Failure to perform the critical task also results in needless degradation of any barrier to fission product release, specifically of the RCS barrier at the point of the RCP seals.

Failure to perform the critical task means that RCS inventory lost through the RCP seals cannot be replaced. It also means that the RCP seals remain without cooling and gradually deteriorate. As the seals deteriorate the rate of RCS inventory loss increases.

25 gpm per pump is the limit used in Section 8.3A.3 CALLAWAY STATION BLACKOUT DURATION (Table 8.3A-1 Section D)

PWR Owners Group Appendix CT - 23, Establish AFW flow during SBO CT - 24, Energize at least one ac emergency bus

Scenario Event Description Callaway 2014 Scenario #2, rev. 0 References OTN-EA-00001, Service Water System OTO-BB-00005, RCS High Activity OTN-BG-00001 Addendum 04, Operation of CVCS Letdown OTO-BB-00004, RTD Channel Failure OTO-NB-00001, Loss of Power to NB01 E-0, Reactor Trip or Safety Injection ES-0.1, Reactor Trip Response ECA-0.0, Loss of All AC Power EOP Addendum 39, Alternate Emergency Power Supply PRA Systems, Events or Operator Actions

1. Station Blackout (T(1S))
a. Manually Operate TDAFW pump
b. AEPS used to power AC Bus

Scenario Setup Guide:

Establish the initial conditions of IC-164, MOL 10-8 amps:

RCS boron concentration 1367 ppm CCP A 1372 ppm minus 2 days CCP B 1370 ppm minus 1 days Rod Control Bank D 105 steps, Other banks 228 steps NE02 is tagged out with breaker in P-T-L and tag on handswitch Ensure A and B SW pumps running with C SW pump in STBY Ensure A CCW pump is running, suppling the service loop NCP recirc should be open < 100 gpm SB-S701 A and B should be NORMAL

=======SCENARIO PRELOADS / SETUP ITEMS==========

Decay Heat Plant parameters, (BB) RRSPDDH - current 1 NE02 is tagged out for an LCO Remote Function (KJ) DGBLOCK_2, Block_Both Remote Function (KJ) KJHS0109, Local_Manual ME Schematic (NE) e23nell_b, NE15NB0211_BKRTA_BKPOS, Value=3 TDAFW failure to auto start Insert Remote Function (AL) PAL02_3, Value = 1 AEPS EDG fail to start Insert Remote Function (PA) PA50101_1, Value = 1

======= EVENT 1============================

Start Service Water Pump C and Secure Service Water Pump A No malfunction

======= EVENT 2============================

RCS High Activity / Place 120 gpm Letdown in-service No malfunction

=======EVENT 3============================

Tavg Channel Failure - Loop 2 Tc Insert Malfunction (BB) BBTE0421B, Value = 630

=======EVENT 4============================

Loss of NB01 Insert Malfunction (NB) NB01_F, Value=Fault

=======EVENT 5======================

Reactor Trip Insert Malfunction (SB) SB001, Value=Enable

=======EVENT 6 PRELOADED=================

Loss of all AC Power Run lesson: All/Generic/lossofswitchyard.lsn

=======EVENT 7 PRELOADED=================

TDAFW fails to Auto Start SEE PRELOADS ABOVE

=======EVENT 8 PRELOADED=================

AEPS EDG fail to start SEE PRELOADS ABOVE

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 3, Rev 2 Op-Test No.: 2014-1 Examiners: ____________________________ Operators:

Initial Conditions: 100%

Turnover: B CCP is tagged out for Breaker PMTs Even t No.

Malf. No.

Event Type*

Event Description 1

SEN0044 SRO (I)

RO (I)

Nuclear Instrument Fails high. (Tech Specs) 2 AELT0519 SRO (I)

BOP (I)

SG Level Instrument Fails to 75% (Tech Specs) 3 PAE01A_2 PAE01A_3 PAE01A_4 SRO (R)

RO (R)

BOP (C)

MFP Vibrations Forcing Downpower 4

PAE0`A.1 SRO (C)

RO (C)

BOP (C)

MFP Trip / Rx Trip 5

SF006 SRO (M)

RO (M)

BOP (M)

Reactor Trip Failure / ATWS 6

TBB03 SRO (M)

RO (M)

Pressurizer Steam Space Leak 7

SIS_A SIS_B SRO (C)

RO (C)

Automatic SI Actuation Failure Both Trains (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes (Per Scenario; See Section D.5.d)

Actual Attributes

1. Total malfunctions (5-8) 7
2. Malfunctions after EOP entry (1-2) 2
3. Abnormal events (2-4) 3
4. Major transients (1-2) 2
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 2014-1 NRC Scenario #3, rev. 2 The plant is steady at 100% power.

After the crew takes the watch, N44 will fail high. The crew will respond with OTO-SE-00001, Nuclear Instrument Malfunction, to select an operable channel and bypass the failed instrument.

The CRS will evaluate Tech Spec 3.3.1.

After the failed NI has been bypassed and the CRS has identified the applicable Tech Specs, Steam Generator A controlling level channel slowly fails to 75%. The crew will respond using OTO-AE-00002, Steam Generator Water Level Control Instrument Malfunctions, to control SG level. The CRS will evaluate Tech Spec 3.3.1. and 3.3.2 After Tech Specs have been addressed, A MFP will develop vibrations. The crew will respond with OTO-MA-00008, Rapid Load Reduction to lower power to below 65%. The crew will lower turbine load and borate to lower power.

After the crew has lowered power 5 to 10 MWe and prior to reducing power less than 70%, A MFP will trip. The crew will respond IAW OTO-AE-00001, Feedwater System Malfunction and attempt to trip the reactor.

When the crew attempts to trip the reactor rods will not insert. The crew will respond by entering E-0, Reactor Trip or Safety Injection and transitioning to FR-S.1, Response to Nuclear Power Generation/ATWS. The crew will be able to successfully de-energize PG19 and 20, insert control rods, and emergency borate the RCS. After rods have been inserted and the reactor has been shutdown the crew will transition back to E-0, Reactor Trip or Safety Injection.

After the transition back to E-0, a leak in the PZR steam space develops. When the RCS depressurizes SI fails to automatically initiate. The crew will manually initiate SI and transition to E-1, Loss of Reactor or Secondary Coolant.

The scenario can be terminated after Step 5 of E-1.

Scenario Event Description Callaway 2014-1 NRC Scenario #3, rev. 2 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 Manually actuate at least one train of SIS-actuated safeguards equipment before transition to E-1.

EVENT 5

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.

In the scenario, failure to manually actuate Sl results in a significant reduction of safety margin beyond that irreparably introduced by the scenario. Failure to manually actuate Sl in the scenario would be a violation of the facility license condition. Although the degraded status is not due to the crew's action, continuation in the degraded status would be a result of the crew's failure to manually actuate Sl.

Cueing Both of the following:

Indication of ATWS (the reactor is not tripped and that a manual reactor trip is not effective)

Indication and/or annunciation that that Sl is required PRZR pressure or SG pressure less than Sl actuation setpoint No indication or annunciation that Sl is actuated 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 Manipulation of controls as required to actuate at least one train of Sl SB HS-27 SB HS-28 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 Indication that both Trains of SI - Actuated LOCA Sequencer annunciator 30A - Lit LOCA Sequencer annunciator 30B - Lit SB069 SI Actuate Red Light - Lit SOLID (NOT blinking)

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.

The crew has had ample opportunity to recognize the need for Sl and the fact that Sl has not automatically actuated.

Given the postulated plant conditions, transition from E-0 to ES-0.1 constitutes an error in using the E-0 procedure. The crew is in the wrong procedure; however, the crew is allowed to recover from this error up through Step 3.a of ES-0.1.

The ERG network is designed to "catch" errors in procedure usage. Step 3.a is designed to get the crew back to E-0, if that is in fact where the crew should be. If the crew members pass through Step 3.a and remain in ES-0.1, they have missed the last step that would return them to the correct procedure.

PWR Owners Group Appendix CT-52, Insert negative reactivity into the core CT-2, Manually actuate Sl

Scenario Event Description Callaway 2014-1 NRC Scenario #3, rev. 2 References OTO-SE-00001, Nuclear Instrument Malfunction OTO-AE-00002, Steam Generator Water Level Control Instrument Malfunctions OTA-RK-00026 Add 120E, MFP A VIB ALERT OTO-MA-00008, Rapid Load Reduction OTO-AE-00001, Feedwater System Malfunction E-0, Reactor Trip or Safety Injection E-1, Loss of Primary or Secondary Reactor coolant FR-S.1, Response to Nuclear Power Generation/ATWS 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. ATWS from Loss of Main Feedwater (TAT2)
a. Manually open control rod drive motor generator breakers
b. Manually drive in control rods
2. LOCA (S(2 or 3))
a. Manually Initiate SI

Scenario Setup Guide:

Establish the initial conditions of IC-10, MOL 100% power:

RCS boron concentration 756 ppm CCP A 760 ppm minus 3 days CCP B 765 ppm minus 7 days Rod Control Bank D 215 steps, Other banks 228 steps B CCP is tagged out for Breaker PMTs

=======SCENARIO PRELOADS / SETUP ITEMS==========

Reactor Fails to Trip in Auto and Manual (Event 5)

Insert Malfunction (SF) SF006, Value = Both_modes PZR steam leak (Event 6)

Insert Malfunction (BB) TBB03, Value = 200, Ramp = 1 min, conditional rec0009 le 1.0 with a delay of 5 minutes SI Fails to Automatically Actuate (Event 7)

Malfunctions (SB)SIS_A_Block_Auto, Value = Block Malfunctions (SB)SIS_B_Block_Auto, Value = Block

======= EVENT 1============================

N44 fails high Insert Malfunction (SE) SEN0044, Value = 120, Ramp = 10 sec

======= EVENT 2============================

Steam Generator A controlling level channel slowly fails to 75%

Insert Malfunction (AE) AELT0519, Value = 75, ramp = 1 min

=======EVENT 3============================

Main Feed Pump Vibration-A MFP Insert Malfunctions (AE) PAE01A_2, Value = 0.042 (AE) PAE01A_3, Value = 0.038 (AE) PAE01A_4, Value = 0.03

=======EVENT 4============================

Main Feed Pump Trip-A MFP Insert Malfunction (AE) PAE01A_1, Value = Trip

=======EVENT 5============================

Reactor Fails to Trip in Auto and Manual SEE PRELOADS ABOVE

=======EVENT 6 PRELOADED=================

PZR steam leak SEE PRELOADS ABOVE

=======EVENT 7 PRELOADED===============

SI Fails to Automatically Actuate SEE PRELOADS ABOVE

Appendix D Scenario Outline Form ES-D-1 Facility: Callaway Scenario No.: 4, Rev 2 Op-Test No.: 2014-1 Examiners: ____________________________ Operators:

Initial Conditions: 80%

Turnover:

Holding at 80% per Transmission Operations A MDAFWP is tagged out for Breaker PMTs Covers for containment purge in place on GTRT22 and GTRT33 Even t No.

Malf. No.

Event Type*

Event Description 1

N/A SRO (N)

BOP (N)

Place Mini-Purge in service 2

WEIUBGTC013 0TVSPNO AUTO SRO (C)

RO (C)

LTDN HX Temp Cont Valve Controller Fails 3

(AB)PV004A SRO (C)

BOP (C)

D SG ASD Fails Open (Tech Specs) 4 (BB)

EBB01D SRO (R)

RO (R)

BOP (C)

SG Tube leak / Downpower (Tech Specs) 5 (BB)

EBB01D SRO (M)

RO (M)

BOP (M)

SG Tube Rupture 6

CPIS_A_BL OCK_AUTO CPIS_B_BL OCK_AUTO SRO (C)

BOP (C)

Mini-Purge Failure to isolate 7

(BB)BBPCV0 455A SRO (C)

RO (C)

IA valve failure / PZR PORV 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) 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 2014 NRC Scenario #4, rev. 2 The plant is steady at 80% power steady state. A MDAFWP is tagged out for Breaker PMTs.

After the crew takes the watch, they will place Mini-Purge in service in IAW OTN-GT-00001, Containment Purge System, in preparation for Containment entry.

After the Mini-Purge system has been placed in service, the Letdown HX temperature control valve will closed. The crew will take actions IAW OTA-RK-00018 ADD 39B.

After the crew has addressed the letdown HX temperature control valve controller failure the SG D ASD fails open. The crew will respond IAW OTO-AB-00001, Steam Dump Malfunction. The crew will close the D ASD. The CRS will evaluate Tech Spec 3.7.4.A After Tech Specs have been evaluated, D SG will develop a tube leak. The crew will respond IAW OTO-BB-00003. The Crew will lower power and the CRS will evaluate Tech Spec 3.4.13, RCS Operational Leakage After the power reduction has been commenced, the tube leak will increase. The crew will trip the reactor due to the increased leak and respond IAW E-0, Reactor Trip or Safety Injection. The crew will identify that Mini-Purge failed to isolate. The crew will manually isolate Mini-Purge. The crew will transition to E-3, Steam Generator Tube Rupture and isolate the ruptured D SG Instrument Air cannot be restored to containment due to a valve failure, therefore normal pressurizer spray is NOT available. When the crew opens the PZR PORV to lower pressure the PORV will stick open. The crew will then have to close the PORV block valve.

The scenario can be terminated after the cooldown is in progress and the PORV has been isolated.

Scenario Event Description Callaway 2014 NRC Scenario #4, rev. 2 Critical Tasks:

Critical Tasks Isolate feedwater flow into and steam flow from the ruptured SG before a transition to ECA-3.1 occurs.

Terminate the RCS depressurization by closing PORV block Valve prior to Loss of Subcooling.

EVENT 5

7 Safety significance Isolating the ruptured SG maintains a differential pressure between the ruptured SG and the intact SGs. The differential pressure (250 psi) ensures that minimum RCS subcooling remains after RCS depressurization.

In the scenario, closing the block MOV constitutes a task that is essential to safety, because its improper performance or omission by an operator will result in direct adverse consequences or significant degradation in the mitigative capability of the plant.

In particular, the crew has failed to prevent degradation of any barrier to fission product release. In this case, the RCS fission product barrier can be restored to full integrity simply by closing the block MOV.

Cueing All of the following:

Indication and/or annunciation of SGTR in one SG o

Increasing SG water level o

Radiation Indication and/or annunciation of reactor trip Indication and/or annunciation of Sl All of the following:

Valve position indication and/or annunciation that the PRZR PORV is open, even after attempts to close it manually from the control room Indication and/or annunciation of decreasing RCS pressure Indication and/or annunciation consistent with the discharge of PRZR fluid to the PRT o

PRT temperature, level, pressure o

Tailpipe RTDs Valve position indication that the block MOV upstream of the stuck open PRZR PORV is open Performance indicator Manipulation of controls as required to isolate the ruptured SG Close Steam line Low point Drain valve from ruptured SG o

AB HIS-10 (SG D)

Close MSIV and MSIV bypass valve from ruptured SG o

AB HIS-11 (SG D)

Stop feed flow to ruptured SG o

CLOSE AL HK-5A and AL HK-6A Manipulation of controls as required to close the block MOV for the stuck open PRZR PORV Performance feedback Crew will observe the following:

Indication of stable or increasing pressure in the ruptured SG Indication of decreasing or zero feedwater flow rate in the ruptured SG Valve position indication that the block MOV upstream of the stuck open PRZR PORV is closed Justification for the chosen performance limit When the crew cannot maintain the 250 psi differential, the ERGs require a transition to contingency ERG ECA-3.1. This transition unnecessarily delays the sequence of actions leading to RCS depressurization and Sl termination.

prior to Loss of Subcooling. is in accordance with the PWR Owners Group Emergency Response Guidelines. The PWR Owners Group Emergency Response Guidelines states: by completion of the first step in the ERG network that directs the crew to close the block MOV. If the crew does not close the block MOV in Step 17 of E-3, Step 18 directs the crew to close the PZR PORV Block Valve if RCS pressure is NOT RISING. If the crew does not close the valves and RCS pressure continues to lower, the crew is directed to transition to ECA-3.1 PWR Owners Group Appendix CT 18, Isolate the Ruptured SG.

CT 10, Close PRZ PORV Block Valve.

Scenario Event Description Callaway 2014 NRC Scenario #4, rev. 2 References OTN-GT-00001, Containment Purge System OTN-BG-00001 ADD 4, Operation of CVCS Letdown OTA-RK-00018, ADD 39B, Letdown Heat Exchanger Discharge Temperature High.

OTO-AB-00001, Steam Dump Malfunction OTO-BB-00001, Steam Generator Tube Leak OTO-MA-00008, Rapid Load Reduction OTN-BG-00002, Attachment 8, Borate Mode of RMCS Operation E-0, Reactor Trip or Safety Injection E-3, Steam Generator Tube Rupture Tech Spec 3.7.4.A, Atmospheric Steam Dump Valves (ASDs)

Tech Spec 3.4.13, RCS Operational Leakage ODP-ZZ-00025, EOP/OTO User's Guide PRA Systems, Events or Operator Actions

1. Steam Generator Tube Rupture (T(SG))
a. Isolate Ruptured Steam Generator
b. Cool Down and Depressurizes the RCS

Scenario Setup Guide:

Establish the initial conditions of IC-2, MOL 80% power:

RCS boron concentration 958 ppm CCP A 968 ppm minus 2 days CCP B 973 ppm minus 5 days Rod Control Bank D 190 steps, Other banks 228 steps A MDAFWP is tagged out for Breaker PMT

=======SCENARIO PRELOADS / SETUP ITEMS====

A MDAFWP is tagged out Insert ME schematics (AL) e23al01a/AL01NB0105IA_BKPOS/Value = 3 Mini-Purge Failure to isolate (Inhibit CPIS Actuation with damper GT-HZ-11 stuck OPEN after CPIS manually actuated) (Event 6)

Insert Malfunction (SB) CPIS_A_BLOCK_AUTO, Value = Block Insert Malfunction (SB) CPIS_B_BLOCK_AUTO, Value = Block Insert ME Schematic (GT) m22gt01_b, GTHZ0011_ATVFAILSP, Value = 1.0, cond HWX20O49R eq 1

Insert ME Schematic (GT) m22gt01_b, GTHZ0012_ATVFAILSP, Value = 1.0, cond HWX20O52R eq 1

Delete ME Schematic (GT) m22gt01_b, GTHZ0012_ATVFAILSP, Value = 1.0, cond X20I52C eq 1 PZR PORV Failure (Event 7)

Malfunctions/BB/BBPCV0455A/100/cond=HWX21o149req1 Malfunctions/BB/BBPCV0465A/100/cond=HWX21o150req1 Override sw (KA) X24I32o Value=0

======= EVENT 1============================

Place Mini-Purge in service

======= EVENT 2============================

LTDN HX Temp Cont Valve Fails Closed Insert ME schematics (EG) m22eg03, X8809d58s009 analog 053247, WEIUBGTC0130TVSPNO AUTO, Value = 1

======= EVENT 3============================

D SG ASD Fails Open Insert Malfunction (AB)PV0004A_1, ramp 30 sec, value 1

=======EVENT 4============================

D SG Tube leak / Downpower Insert Malfunction (BB) EBB01D, value=30, ramp 5 min

=======EVENT 5============================

D SG Tube Rupture / Reactor Trip Insert Malfunction (BB) EBB01D, value=400, ramp 10 min

=======EVENT 6 PRELOADED=================

Minipurge fails to isolate See preloads above

=======EVENT 7 PRELOADED=================

PZR PORV Failure See preloads above