J. Dumas

发表

In-pile studies of inert matrices with emphasis on magnesia and magnesium aluminate spinel

D. Lespiaux, J. Noirot, N. Chauvin, 1999 .

TAF-ID: An international thermodynamic database for nuclear fuels applications

M. Kurata, T. Besmann, S. Gossé, 2021, Calphad.

An improved method to evaluate the “Joint Oxyde-Gaine” formation in (U,Pu)O2 irradiated fuels using the GERMINAL V2 code coupled to Calphad thermodynamic computations

B. Sundman, J. Dumas, K. Samuelsson, 2020, EPJ Nuclear Sciences & Technologies.

Simulation of the chemical state of high burnup (U,Pu)O2 fuel in fast reactors based on thermodynamic calculations

C. Guéneau, B. Sundman, J. Dumas, 2020 .

Fission Product Chemistry in Oxide Fuels

B. Lewis, W. Thompson, F. Iglesias, 2020, Comprehensive Nuclear Materials.

Current status and progression of germinal fuel performance code for sfr oxide fuel pins

B. Michel, J. Dumas, K. Samuelsson, 2017 .

Fuel Melting Margin Assessment of Fast Reactor Oxide Fuel Pin Using a Statistical Approach

V. Dupont, B. Michel, F. Charollais, 2017 .

Thermodynamic modelling of advanced oxide and carbide nuclear fuels: Description of the U–Pu–O–C systems

Christine Guéneau, Dario Manara, Rudy J. M. Konings, 2011 .

Thermodynamic assessment of the Cs–Te binary system

S. Gossé, C. Guéneau, J. Rogez, 2015 .

Thermodynamic modelling assessment of the ternary system Cs-Mo-O

C. Guéneau, N. Dupin, A.L. Smith, 2021, Calphad.

Thermodynamic Modelling of the Uranium-Carbon-Oxygen System in the Frame of the Uranium Oxide and Carbon Interaction in the TRISO Fuel Particle of High Temperature Reactor

S. Chatain, C. Guéneau, J. Piron, 2006 .

A thermodynamic approach for advanced fuels of gas-cooled reactors

S. Chatain, S. Gossé, C. Guéneau, 2005 .

Calphad thermodynamic description of some binary systems involving U

A. Berche, C. Guéneau, B. Sundman, 2011 .

Application of the CALPHAD method to the thermodynamic modeling of a miscibility gap in the U-Nd-O phase diagram

C. Guéneau, Y. Pontillon, L. Desgranges, 2014 .

New recommendation for the thermal conductivity of irradiated (U, Pu)O2 fuels under fast reactor conditions. Comparison with recent experimental data.

G. Montagnier, J. Lamontagne, I. Boukhris, 2023, Journal of Nuclear Materials.

GERMINAL, a fuel performance code of the PLEIADES platform to simulate the in-pile behaviour of mixed oxide fuel pins for sodium-cooled fast reactors

I. Ramière, B. Michel, J. Dumas, 2019, Journal of Nuclear Materials.

Two fuel performance codes of the PLEIADES platform: ALCYONE and GERMINAL

J. Sercombe, I. Ramière, B. Michel, 2021 .