Design basis accidents for light water reactors and numerical simulation tools

RELAP5

 

Dr. V. Sanchez-Espinoza, INR, KIT

Dr. Sanchez-Espinoza

 

Course

009

Title

Design Basis Accidents for Light Water Reactors and Numerical Simulation Tools

Prerequisites

Thermal hydraulic of nuclear reactors, nuclear power plants

Goal of the course

Familiarization with the theoretical basis and practical use of thermal hydraulic system codes such as RELAP5, TRACE for the analysis of design basis accidents of nuclear power plants

Content

  • Reactor safety fundamentals
  • Safety analysis and deterministic analysis methodologies
  • Design basis accidents and their classification
  • Numerical simulation tools for the analysis of design basis accidents
  • Physical and mathematical models of thermal hydraulic system codes
  • Validation and uncertainty quantification of system codes
  • Stepwise approach for the development of integral LWR plant models
  • Developing hands-on training developing and running models for BWR and PWR
  • Code systems to be used: e.g. RELAP5, TRACE, SNAP (GUI Pre- and Postprocessor)

Lecturer

Dr. V. Sanchez-Espinoza, INR, KIT

Schedule
shifted!!!

(23.01. - 27.01.2017); 9 am to 5 pm daily

Location

KIT Campus North, FTU, Building 101, Room 155

Dead line for application:  09.01.2017

Application form download

 

Examples of TRACE Applications for PWR including Uncertianty Quantification with SUSA

Trace PWR Trace SUSA
TRACE PWR Integral Plant Model for Safety Evaluations TRACE/SUSA: Uncertainty Quantification of best-estimate codes: Cladding temperature for LOCA

 

Lecturer:

Dr. Victor Hugo Sanchez Espinoza
Head of Group “Reactor Physics and Dynamics"
Project Leader “LWR Methods and Analysis”
Dr.-Ing. in Nuclear Engineering of TU Dresden
Dipl.-Ing. Nuclear Engineering, TU Dresden

Areas of Expertise - Modeling of LWR in-vessel severe accident phenomena
- Improvement of thermal hydraulic system codes
- Transient analysis for PWR and BWR with system and coupled codes
- Transient analysis of Gen-4 reactors with thermal hydraulic system codes
- Development of multi-physical and multi-scale coupling for local prediction of pin power
- Validation, qualification and uncertainty quantification of thermal hydraulic and neutron kinetics coupled codes
Lectureship KIT Mechanical Engineering Faculty since 2008:
Nuclear Safety Assessment of Nuclear Power Plants
Kontact:
Email:
Homepage:
Dr. V. Sánchez Espinoza
victor sanchez∂kit edu
www.inr.kit.edu