The Studies and Modelling of Cooling Accidents Laboratory (LIMAR, ex-LEMAR) is located at Cadarache in the south of France and headed by Gaétan Guillard.
Context and research themes
The overall mission of the LIMAR is to improve knowledge on the degradation of the core in situations of severe accidents and loss of primary coolant (APRP) on the degradation of fuel in the situation of an accident of handling or storage, as well as aspects of thermo-hydraulic accidents injection Reactivity (RIA). It also carries out studies to support expertise.
His skills, acquired on pressurized water reactors, are applied to research reactors in naval propulsion, the boiling water reactors, reactors changing 3rd generation and 4th generation reactors.The main issue of the laboratory is to improve the knowledge the fuel behaviour in accidental situation of Nuclear Power Plants. Studies in support to safety analyses of N.P.P are also in the scoop of the laboratory.
In the framework of these activities, the laboratory is involved in the preparation, the analysis and the interpretation of in-pile or out-of pile, integral or separated-effect tests in international programmes using computer codes.
Specialties and researchers
List of researchers and engineers:
Gaétan Guillard, head of laboratory
Adrien Abbate, PhD student (2014-2017)
Jean Baccou, research engineer
Éric Chojnacki, research engineer and expert
Olivier De Luze, engineer
Marc Forestier, engineer
Tony Glantz, engineer
Sébastien Marmin, PhD student (2014-2017)
Jimmy Martin, engineer
Pierre Ruyer, research engineer
Christine Sartoris, engineer
Tatiana Taurines, research engineer
Zhenhai Zou, PhD student (2016-2019)
Facilities and techniques
The tools available in the laboratory are listed below:
- Scientific computer codes which are developed in the laboratory it-shelf or by others laboratories or organisations, in particular:
- a computer code developed to evaluate physical phenomena involved in a course of a severe accident on a Nuclear Pressurised Water Reactor
- an advanced thermal hydraulics computer codes
- a computer code devoted the thermomechanical behaviour of the fuel rods of Pressurised Water Reactor in the nuclear fields.
Data bases : bibliography and experimental results from integral or separated-effect tests
These databases concern in-of-pile and out-of-pile experiment in the domain of the loss of cooling accident in the primary circuit, severe accident in the nuclear fields, and data on nuclear materials in particular at high temperatures regarding :
- physical properties
- thermal chemical and thermodynamics properties