Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method
The purpose of this study is to investigate the neutron energy flux from reactor after attenuated by silicon, cadmium and plumbum at the end of the beam port. Neutrons can be categorized into thermal neutron, intermediate and fast neutron according to their energies. Fast neutrons in reactor needs t...
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my-utm-ep.113002017-09-27T04:13:02Z Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method 2010-04 Hamzah, Hafida QC Physics The purpose of this study is to investigate the neutron energy flux from reactor after attenuated by silicon, cadmium and plumbum at the end of the beam port. Neutrons can be categorized into thermal neutron, intermediate and fast neutron according to their energies. Fast neutrons in reactor needs to be thermalized before it can be utilized for applications such as in medicine, industrial and research purposes. The fast neutron will lost its energy into thermal neutrons due to scattering and absorption process with nucleuses. The neutron probability of interaction with nucleus depends on the microscopic cross-section, which is different for each material. All the cross-section data for every element have been compiled in ENDF (Evaluated Nuclear Data File) format and internationally recognized. In this study, the neutron transport was simulated using Monte Carlo N-Particle Transport Code, Version 5 (MCNP5). The thickness of the materials used in this research is in the range of 1 cm to 10 cm. The result shows that the neutron was reduced significantly by silicon then follows by plumbum and cadmium. The thermal neutron flux was the lowest in cadmium because it have high thermal neutron cross-section. 2010-04 Thesis http://eprints.utm.my/id/eprint/11300/ http://eprints.utm.my/id/eprint/11300/4/HafidaHamzahMFS2010.pdf application/pdf en public masters Universiti Teknologi Malaysia, Faculty of Science Faculty of Science [1] Weston M. Stacey, Nuclear Reactor Physics 2nd Edition, Weinheim, Wiley-Vch Verlag GmbH & C0. KGaA. 2007. Page xxiii [2] Everitt P. Blizzard and Lorraine S. Abbott, Reactor Handbook 2nd Edition Volume III Part B: Shielding, New York, Interscience Publishers. 1962. Page 17-18 [3] Muhd Noor Muhd Yunus, Nuclear reactor Energy Research: The Reactor TRIGA PUSPATI, The Ingeniur Volume 37. 2008. Page 38 [4] Arthur B. Chilton, J. Kenneth Shultis and Richard E. Faw, Priciples of Radiation Shielding, Englewood Cliff, New Jersey, Prentice-Hall, Inc. 1984. Page 53 [5] Esam M.A. Hussien, Radiation Mechanics: Principles and Practice, UK, Elsevier. 2007. Page 284 [6] Denise B. Pelowitz, MCNPXTM User’s Manual Version 2.5.0, Los Alamos National Laboratory. 2005. Page 1-2 [7] J.K. Shultis and R.E. Faw, An MCNP Primer, Manhattan, Dept. of Mechanical and Nuclear Engineering, Kansas State University. 2005. Page 36 [8] S.L. Shue, R.E. Faw and J.K. Shultis, Fast Neutron Thermalization And Capture Gamma-Ray Generation In Soils, Manhattan, Department of Nuclear Enggineering, Kansas State University. |
| institution |
Universiti Teknologi Malaysia |
| collection |
UTM Institutional Repository |
| language |
English |
| topic |
QC Physics |
| spellingShingle |
QC Physics Hamzah, Hafida Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method |
| description |
The purpose of this study is to investigate the neutron energy flux from reactor after attenuated by silicon, cadmium and plumbum at the end of the beam port. Neutrons can be categorized into thermal neutron, intermediate and fast neutron according to their energies. Fast neutrons in reactor needs to be thermalized before it can be utilized for applications such as in medicine, industrial and research purposes. The fast neutron will lost its energy into thermal neutrons due to scattering and absorption process with nucleuses. The neutron probability of interaction with nucleus depends on the microscopic cross-section, which is different for each material. All the cross-section data for every element have been compiled in ENDF (Evaluated Nuclear Data File) format and internationally recognized. In this study, the neutron transport was simulated using Monte Carlo N-Particle Transport Code, Version 5 (MCNP5). The thickness of the materials used in this research is in the range of 1 cm to 10 cm. The result shows that the neutron was reduced significantly by silicon then follows by plumbum and cadmium. The thermal neutron flux was the lowest in cadmium because it have high thermal neutron cross-section. |
| format |
Thesis |
| qualification_level |
Master's degree |
| author |
Hamzah, Hafida |
| author_facet |
Hamzah, Hafida |
| author_sort |
Hamzah, Hafida |
| title |
Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method |
| title_short |
Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method |
| title_full |
Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method |
| title_fullStr |
Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method |
| title_full_unstemmed |
Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method |
| title_sort |
simulation of neutron flux in silicon, cadmium and plumbum using monte carlo method |
| granting_institution |
Universiti Teknologi Malaysia, Faculty of Science |
| granting_department |
Faculty of Science |
| publishDate |
2010 |
| url |
http://eprints.utm.my/id/eprint/11300/4/HafidaHamzahMFS2010.pdf |
| _version_ |
1747814837448081408 |