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Journal of Nuclear Materials

Resumen/Descripción – provisto por la editorial en inglés

The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome.

The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example.

The following list describes topics within the scope of the Journal.Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components; fission product behavior.Materials aspects of the entire fuel cycle.Materials aspects of the actinides and their compounds.Performance of nuclear waste materials; immobilization of wastes.Fusion reactor materials, including first walls, blankets, insulators and magnets.Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties.Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials.Please note: a manuscript must focus on materials science and engineering related to nuclear applications. Thus, all topics within nuclear science and engineering or materials science and engineering do not necessarily fall within the scope. Topics outside the scope of the Journal will not be considered, and potential alternative journals are indicated in brackets in the list below. Typical examples of topics which are out of scope for the Journal include:

(1) Topics in nuclear engineering and other areas not addressing materials, such as: Particle transport, cross-sections or isotope ratios (Radiation Physics and Chemistry; Annals of Nuclear Energy) Process engineering (Materials Science and Engineering A; Materials and Design) Thermal hydraulics (Nuclear engineering and design) Isotope separation processes (Nuclear Instrument and Methods B) Fusion reactor design and technology (Fusion Engineering & Design) Plasma physics (Physics Letters A)(2) Materials topics not addressing nuclear applications, such as general studies in:Physical properties including modeling and simulation (Materials Science and Engineering A, Materials Letters) Metallurgy (Journal of Alloys and Compounds; Materials Science and Engineering A) Corrosion (Corrosion Science) Welding and joining (Journal of Alloys and Compounds; Materials and Design) Ceramics (Journal of the European Ceramics Society; Ceramics international)

Palabras clave – provistas por la editorial

No disponibles.

Disponibilidad

Institución detectada	Período	Navegá	Descargá	Solicitá
No detectada	desde abr. 1959 / hasta dic. 2023	ScienceDirect

Información

Tipo de recurso:

revistas

ISSN impreso

0022-3115

ISSN electrónico

1873-4820

Fecha de publicación

1959-

Cobertura temática

Ingeniería mecánica

Tabla de contenidos

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doi: 10.1016/0022-3115(59)90008-x

Sintering of uranium oxides of composition UO2 to U3O8 in various atmospheres

J. Williams; E. Barnes; R. Scott; A. Hall

Palabras clave: Nuclear Energy and Engineering; General Materials Science; Nuclear and High Energy Physics.

Pp. 28-38

doi: 10.1016/0022-3115(77)90097-6

The diffusion of oxygen in alpha-zirconium

I.G Ritchie; A Atrens

Palabras clave: Nuclear and High Energy Physics; General Materials Science; Nuclear Energy and Engineering.

Pp. 254-264

doi: 10.1016/0022-3115(77)90266-5

The matrix swelling rate of UO2

H. Assmann; R. Manzel

Palabras clave: Nuclear Energy and Engineering; General Materials Science; Nuclear and High Energy Physics.

Pp. 360-364

doi: 10.1016/0022-3115(79)90013-8

In-pile densification of intergranular porosity in oxide fuels

C.C. Dollins

Palabras clave: Nuclear Energy and Engineering; General Materials Science; Nuclear and High Energy Physics.

Pp. 302-310

doi: 10.1016/0022-3115(85)90433-7

Release of metal fission products from UO2 kernel of coated fuel particle

T. Ogawa; K. Fukuda; H. Sekino; M. Numata; K. Ikawa

Palabras clave: Nuclear and High Energy Physics; General Materials Science; Nuclear Energy and Engineering.

Pp. 18-31

doi: 10.1016/0022-3115(90)90294-w

On mechanisms by which a soft neutron spectrum may induce accelerated embrittlement

L.K. Mansur; K. Farrell

Palabras clave: Nuclear and High Energy Physics; General Materials Science; Nuclear Energy and Engineering.

Pp. 236-245

doi: 10.1016/0022-3115(91)90386-l

Effect of sintering atmosphere on the densification of UO2Gd2O3 compacts

Ryoichi Yuda; Katsumi Une

Palabras clave: Nuclear Energy and Engineering; General Materials Science; Nuclear and High Energy Physics.

Pp. 195-203

doi: 10.1016/0022-3115(92)90335-i

Recommendations on damage exposure units for ferritic steel embrittlement correlations

R.E. Stoller; G.R. Odette

Pp. 203-205

doi: 10.1016/0022-3115(93)90068-a

Experimental studies of free defect generation during irradiation — Implications for reactor environments

L.E. Rehn; R.C. Birtcher

Palabras clave: Nuclear and High Energy Physics; General Materials Science; Nuclear Energy and Engineering.

Pp. 31-39

doi: 10.1016/s0022-3115(96)00739-8

Mechanisms of radiation-induced degradation of reactor vessel materials

L.K. Mansur; K. Farrell

Pp. 212-218