MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination-Reference-Cited by-同舟云学术

MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination

Published:2011-12 Issue:12 Volume:241 Page:4989-4993
ISSN:0029-5493
Container-title:Nuclear Engineering and Design
language:en
Short-container-title:Nuclear Engineering and Design

Author:

da Silva Clarysson A.M.,Pereira Claubia,Guerra Bruno T.,Veloso Maria Auxiliadora F.,Costa Antonella L.,Dalle Hugo M.,Menezes Maria Ângela de B.C.

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

Reference11 articles.

1. MCNP-A General Monte Carlo N-Particle Transport Code (Version 5);Booth,2003

2. Dalle, H.M., 2005. Simulação do Reator IPR-R1Utilizando Métodos de Transporte por Monte Carlo, Doctoral Thesis, Faculdade de Engenharia Química, Universidade Estadual de Campinas, Campinas, Brazil.

3. Neutronic calculation to the TRIGA IPR-R1 Reactor using the WIMSD4 and CITATION codes;Dalle;Annals of Nuclear Energy,2002

4. Guimarães, R.R.R., 1985. Levantamento das Distribuições dos Fluxos de Nêutrons Térmicos e Rápidos no Núcleo do Reator IPR-R1, Master Thesis, Escola de Engenharia, Universidade Federal de Minas Gerais, Belo Horizonte, Brazil.

5. Experimental heat transfer analysis of the IPR-R1TRIGA Reactor;Mesquita,2006

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1. Neutronic assessment of IPR-R1 TRIGA using Serpent 2.1.32 code;Nuclear Engineering and Design;2024-08

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

3. 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;Nuclear Engineering and Design;2014-07