Monte Carlo Calculations Using MCNP4B for an Optimal Shielding Design If A14-MeV Neutron Source PDF Download
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Languages : en
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The shielding for 14-MeV neutrons from a D-T generator at SLAC needs to be optimized in terms of shield thickness and cost. Shielding calculations were made using the MCNP4B Monte Carlo code and the ENDF/B-VI cross section set. A few materials were studied: iron, borated polyethylene, three types of normal and heavy concrete, and mixtures of iron and borated polyethylene. The effects of shielding geometry (sphere, cube, and the proposed geometry) were also examined. The transmission results of ambient dose equivalents, as well as the attenuation lengths and the average energies, for neutrons and gammas outside the shielding are presented.
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Languages : en
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Book Description
The shielding for 14-MeV neutrons from a D-T generator at SLAC needs to be optimized in terms of shield thickness and cost. Shielding calculations were made using the MCNP4B Monte Carlo code and the ENDF/B-VI cross section set. A few materials were studied: iron, borated polyethylene, three types of normal and heavy concrete, and mixtures of iron and borated polyethylene. The effects of shielding geometry (sphere, cube, and the proposed geometry) were also examined. The transmission results of ambient dose equivalents, as well as the attenuation lengths and the average energies, for neutrons and gammas outside the shielding are presented.
Author: Wendell De Marcus
Publisher:
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Category : Gamma rays
Languages : en
Pages : 12
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Category :
Languages : en
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Integral experiments performed at the Institute for Reactor Development at Juelich, West Germany, have raised doubts about the adequacy of current nuclear data and calculational methods in predicting neutron multiplication and leakage from beryllium assemblies bombarded with 14-MeV neutrons. Experimental values of inferred neutron multiplication were reported to be less than calculated values by approx. 25%. We have performed calculations of the experiments using the TART Monte Carlo code. The ratio of measured leakage multiplication to our calculated leakage multiplication is 0.85 for 8-cm-thick beryllium and 0.99 for 12-cm-thick beryllium. However, much uncertainty exists in the procedure leading to the stated experiment values of apparent multiplication. We also performed calculations for a series of experiments done at LLNL from 1955 to 1956. A 14-MeV neutron source was placed in the center of a cylindrical beryllium assembly 8 in. in radius and 24 in. in height. The beryllium assembly was encased in an aluminum can surrounded by a manganese bath. The ratio of experimental to calculated neutron leakage multiplication is 0.88. Much uncertainty exists in both series of experiments.
Author: John S. Hendricks
Publisher: Springer Nature
ISBN: 3031041291
Category : Science
Languages : en
Pages : 316
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Book Description
This open access book is a pedagogical, examples-based guide to using the Monte Carlo N-Particle (MCNP®) code for nuclear safeguards and non-proliferation applications. The MCNP code, general-purpose software for particle transport simulations, is widely used in the field of nuclear safeguards and non-proliferation for numerous applications including detector design and calibration, and the study of scenarios such as measurement of fresh and spent fuel. This book fills a gap in the existing MCNP software literature by teaching MCNP software usage through detailed examples that were selected based on both student feedback and the real-world experience of the nuclear safeguards group at Los Alamos National Laboratory. MCNP input and output files are explained, and the technical details used in MCNP input file preparation are linked to the MCNP code manual. Benefiting from the authors’ decades of experience in MCNP simulation, this book is essential reading for students, academic researchers, and practitioners whose work in nuclear physics or nuclear engineering is related to non-proliferation or nuclear safeguards. Each chapter comes with downloadable input files for the user to easily reproduce the examples in the text.
Author: Luka Snoj
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Category :
Languages : en
Pages : 25
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Author: Simon Kellman
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Category : FORTRAN II (Computer program language)
Languages : en
Pages : 110
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Author: F. H. Clark
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Category :
Languages : en
Pages : 24
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The 05R Monte Carlo code (See ORNL-3622, NASA N65-17892) was adapted to calculate fast-neutron doses within and outside of concrete slab shields. Such calculations were performed for a variety of shield thicknesses (0 to 180 cm) and for various monoenergetic sources from 0.7 to 14 MeV. The concrete chosen is ordinary concrete with approximately 6% water content. Results are plotted as tissue kerma versus penetration distance for each source energy. The source is in each case a broad normally incident beam. (Author).
Author: L. Clemons (Jr)
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Category :
Languages : en
Pages : 20
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Author:
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Category :
Languages : en
Pages :
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For heavy-ion beam driven inertial fusion ''liquid-protected'' reactor designs such as HYLIFE-II, a mixture of molten salts made of F[sup 10], Li[sup -6], Li[sup 7] and Be[sup 9] (called flibe) allows small chambers and final-focus magnets closer to the target with superconducting coils suffering higher radiation damage, though they can stand only a certain amount of energy deposited before quenching. This work has been primarily focusing on verifying that total energy deposited by fusion neutrons and induced gamma rays remain under such limit values and the final purpose is the optimization of the shielding of the magnetic lens system from the points of view of the geometrical configuration and of the physical nature of the materials adopted. The system is analyzed in terms of six geometrical models going from simplified up to much more realistic representations of a system of 192 beam lines, each focused by six magnets. A 3-D transport calculation of the radiation penetrating through ducts, that takes into account the complexity of the system, requires Monte Carlo methods. The quantities analyzed, using the two codes MCNP and TART include: neutron mean free path and total path length dependence on energy, energy deposited by neutrons and gamma photons, values of the total fluence integrated in the whole energy range, and the neutron spectrum in different zones of the system. The technical nature of the design problem and the methodology followed were presented in a previous paper by summarizing briefly the results for the deposited energy distribution on the six focal magnets. Now a much more extensive comparison of the performances of the two codes for different configurations of the system is discussed, separating the n and [gamma] contributions, in the light of the physical interpretation of the results in terms of first flight and of scattered neutron fluxes, of primary [gamma] and of secondary [gamma] generated by inelastically scattered or radiatively captured neutrons. The final conclusions indicate some guidelines and suggest possible improvements for the future neutronic design of a heavy ion beam fusion facility named IRE (Integrated Research Experiment) to be realized in the US.
Author: F. H. Clark
Publisher:
ISBN:
Category :
Languages : en
Pages : 19
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An extensive program of calculations was performed to check certain fast-neutron dose-rate measurements made at Project BREN in 1962. The response characteristic of the modified long counter was computed. Nitrogen and oxygen cross sections were reviewed and used in calculations of the transport in air of neutrons from a Godiva source to a distance of 66 g/sq cm. Transport results obtained by Monte Carlo were compared with those obtained by the moments method. A spectrum at an 80-g/sq cm penetration was determined. The perturbating effect of a structure on the neutron field in its neighborhood was estimated. The effect of variations of the spectrum with penetration distances on doses computed with different weight functions was also examined; the Snyder-Neufeld multicollision physical dose was found to vary with distance in a way markedly different from others examined. Work was also done on importance sampling techniques, especially the exponential transform, and the sampling problem was formulated in a simplified, approximate expression to facilitate statistical interpretations. Finally, a number of shield penetration calculations were performed to compare with the measurements and to determine the effects of shield shape, shield cavity size, and source angular distributions incident on the shields. The agreement between the calculations and the measurements was within the uncertainty in the detector response characteristic. (Author).
