Evaluation of the thermal neutron flux in samples of Al–Au alloy irradiated in the carrousel channels of the TRIGA MARK I IPR-R1 reactor using MCNP code-Reference-Cited by-同舟云学术

Evaluation of the thermal neutron flux in samples of Al–Au alloy irradiated in the carrousel channels of the TRIGA MARK I IPR-R1 reactor using MCNP code

Published:2014-07 Issue: Volume:273 Page:576-583
ISSN:0029-5493
Container-title:Nuclear Engineering and Design
language:en
Short-container-title:Nuclear Engineering and Design

Author:

Salomé J.A.D.,Guerra B.T.,Pereira C.,de Menezes M.Â.B.C.,da Silva C.A.M.,Dalle H.M.

Publisher

Elsevier BV

Subject

Mechanical Engineering,Waste Management and Disposal,Safety, Risk, Reliability and Quality,General Materials Science,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Reference15 articles.

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2. Fundamentals of Mathematics and Statistics for Students of Chemistry and Allied Subjects;Brookes,1979

3. MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination;Da Silva;Nucl. Eng. Des.,2011

4. Simulação do reator TRIGA IPR-R1 utilizando métodos de transporte por Monte Carlo;Dalle,2005

5. Neutron Flux Evaluation in the Irradiation Channels of the IPR-R1 TRIGA Research Reactor Using Monte Carlo Method;Guerra,2011

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2. High-resolution transmission electron microscopy study on reversion of Al2CuMg precipitates in Al–Cu–Mg alloys under irradiation;Micron;2015-09

3. Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code;International Journal of Nuclear Energy;2014-09-09