The Transient & Accident Analysis and Mitigating Core Damage (TAA/MCD) Training Course will prepare the successful participant to perform technical support activities involving application of accident analyses. It does not provide sufficient experience to perform ab-initio accident analyses. It does not describe all categories of accidents and does not address every specific event included in the accident analyses.

This course is intended for engineers and licensed operators responsible for planning, analysis, and operational activities affected by accident analysis. No previous knowledge is required, but knowledge of major plant systems and understanding of reactor physics and core thermal-hydraulic concepts are helpful.

Consistent with the Nuclearning® Educational Model, the Transient & Accident Analysis and Mitigating Core Damage course incorporates numerous student-centered learning activities to help the student learn and remember the concepts being studied. This approach puts the student at the center of the learning process making him an active participant in all learning activities. Specifically, the course incorporates instructor facilitated discussions, classroom exercises, and group work exercises. Applicable to: PWR.

Day 1

• Course Introduction – The purpose and general methods of accident analysis 
• Radiological Aspects of Nuclear Accidents – Identification and description of radioactive nuclides and their behavior during accident conditions 
• Reactor Theory – Intrinsic reactivity effects, reactivity control, and reactor kinetics 

Day 2

• Core Thermal Hydraulic Principles – Heat transfer mechanisms, departure from nucleate boiling, and natural circulation 
• Introduction to Accident Analysis Methods 

Day 3

• Increased Heat Removal by the Secondary System – Excessive turbine load, condensate and feedwater system malfunctions, and steam line break accidents
• Reduced Heat Removal by the Secondary System – Loss of normal load, loss of normal feedwater, and feed line break accidents

Day 4

• Reduced Reactor Coolant Flow – Partial loss of forced coolant flow, total loss of forced coolant flow, and natural circulation
• Increased Reactor Coolant Inventory – Inadvertent safety injection actuation and malfunction of the chemical and volume control system

Day 5

• Reduced Reactor Coolant Inventory – Chemical and volume control pipe break outside of containment, theory of loss of reactor coolant accidents (LOCAs), and large-break LOCA

Day 6

• Reduced Reactor Coolant Inventory (continued) – Intermediate and small-break LOCA
• Steam Generator Tube Rupture

Day 7

• Introduction to Mitigating Core Damage – Concept and development of Emergency Response Guideline Critical Safety Functions 
• Subcriticality Critical Safety Function – Evaluation of anticipated transient without scram (ATWS) events 

Day 8

• Core Cooling Critical Safety Function – Evaluation of small-break LOCA without high-head Emergency Core Cooling System injection
• Heat Sink Critical Safety Function – Evaluation of loss of all secondary heat sink event

Day 9

• Primary Integrity Critical Safety Function – Evaluation of pressurized thermal shock events
• Containment Critical Safety Function – Evaluation of containment over-pressure events
• Severe Accident Phenomenology – Adverse consequences of molten core debris

Day 10

• Severe Accident Phenomenology (continued)
• Accident Response of Instrumentation
• Written examination*

*written examination will be used to test each student’s understanding of the course material. The examination questions are based on the learning objectives used in the course.