J
J.N. Wilson
Researcher at Centre national de la recherche scientifique
Publications - 9
Citations - 126
J.N. Wilson is an academic researcher from Centre national de la recherche scientifique. The author has contributed to research in topics: Fission & Neutron. The author has an hindex of 4, co-authored 9 publications receiving 114 citations. Previous affiliations of J.N. Wilson include University of Paris-Sud.
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Journal ArticleDOI
Neutron-induced fission cross sections of short-lived actinides with the surrogate reaction method
Grégoire Kessedjian,Beatriz Jurado,M. Aïche,G. Barreau,Adrien Bidaud,S. Czajkowski,D. Dassié,B. Haas,L. Mathieu,L. Audouin,N. Capellan,L. Tassan-Got,J.N. Wilson,E. Berthoumieux,F. Gunsing,Ch. Theisen,Olivier Serot,E. Bauge,I. Ahmad,John P. Greene,R. V. F. Janssens +20 more
TL;DR: In this paper, the authors used the surrogate reaction method to obtain the neutron-induced fission cross-sections for the first time up to the onset of second-chance fission.
Journal ArticleDOI
Comparative analysis of high conversion achievable in thorium-fueled slightly modified CANDU and PWR reactors
A. Nuttin,P. Guillemin,Adrien Bidaud,N. Capellan,Richard Chambon,Sylvain David,O. Meplan,J.N. Wilson +7 more
TL;DR: In this paper, the conversion performance of thorium-fueled standard or only slightly modified CANDU and PWR reactors with unchanged core envelope and equipments, to be eventually used as the third and last tier of symbiotic scenarios, is investigated.
Study of CANDU Thorium-based Fuel Cycles by Deterministic and Monte Carlo Methods
Alexis Nuttin,P. Guillemin,T. Courau,G. Marleau,O. Meplan,Sylvain David,F. Michel-Sendis,J.N. Wilson +7 more
TL;DR: In this article, the authors evaluate the economic competitiveness of once-through thorium-based fuel cycles in CANDU with the deterministic Canadian cell code and MURE, a C++ tool for reactor evolution calculations based on the Monte Carlo code MCNP.
Sodium-cooled fast reactors: void coefficient and waste minimization. Neutronic studies using MURE
TL;DR: In this paper, a C++ object-oriented evolution code that couples the Monte-Carlo transport code MCNP with a fuel depletion code under given conditions is used to perform neutronic studies on innovative (or evolutive) sodium-cooled reactors.
Journal ArticleDOI
Criticality experiments for validation of cross sections: The neptunium case
L.S. Leong,L.S. Leong,L. Tassan-Got,L. Tassan-Got,L. Audouin,L. Audouin,C. Paradela,J.N. Wilson,J.N. Wilson,D. Tarrío,B. Berthier,B. Berthier,I. Duran,C. Le Naour,C. Le Naour,Christoph A. Stephan,Christoph A. Stephan +16 more
TL;DR: In this article, the authors simulate a criticality experiment performed at Los Alamos with a 6-kilogram sphere of 237 Np. This sphere was surrounded by enriched uranium 235 U so as to approach criticality with fast neutrons.