Topic

# Neutron

About: Neutron is a research topic. Over the lifetime, 70066 publications have been published within this topic receiving 903999 citations. The topic is also known as: N & n.

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TL;DR: In this paper, a structure refinement method was described which does not use integrated neutron powder intensities, single or overlapping, but employs directly the profile intensities obtained from step-scanning measurements of the powder diagram.

Abstract: A structure refinement method is described which does not use integrated neutron powder intensities, single or overlapping, but employs directly the profile intensities obtained from step-scanning measurements of the powder diagram. Nuclear as well as magnetic structures can be refined, the latter only when their magnetic unit cell is equal to, or a multiple of, the nuclear cell. The least-squares refinement procedure allows, with a simple code, the introduction of linear or quadratic constraints between the parameters.

13,024 citations

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Los Alamos National Laboratory

^{1}, Brookhaven National Laboratory^{2}, Oak Ridge National Laboratory^{3}, Rensselaer Polytechnic Institute^{4}, Argonne National Laboratory^{5}, Lawrence Livermore National Laboratory^{6}, International Atomic Energy Agency^{7}, National Institute of Standards and Technology^{8}, Japan Atomic Energy Agency^{9}, Idaho National Laboratory^{10}, Jožef Stefan Institute^{11}, Nuclear Research and Consultancy Group^{12}, University of Vienna^{13}TL;DR: The ENDF/B-VII.1 library as mentioned in this paper is the most widely used data set for nuclear data analysis and has been updated several times over the last five years. But the most recent version of the ENDF-B-VI.0 library is based on the JENDL-4.0 standard.

Abstract: The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment.
The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary.
Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications.
We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [M. B. Chadwick, P. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Br, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, and S. C. van der Marck, “ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,” Nuclear Data Sheets 107, 2931 (2006)].

1,865 citations

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01 Jan 1976

TL;DR: In this article, the fundamental scientific principles governing nuclear fission reactors and the methods used in modern nuclear reactor analysis and design are discussed. But they do not cover the fundamental design of modern power reactors.

Abstract: This comprehensive introduction covers the fundamental scientific principles governing nuclear fission reactors and the methods used in modern nuclear reactor analysis and design. The book is divided into four parts: Part 1 presents a relatively elementary and qualitative discussion of the basic concepts of nuclear fission chain reactions, including a brief review of relevant nuclear physics and a survey of modern power reactors. Part 2 develops a simple model of nuclear reactor behavior, assuming that the neutrons sustaining the fission chain reactions diffuse from point to point in the reactor in such a way that their energy and direction of motion can be ignored (one-speed diffusion theory). Part 3 generalizes this model to develop the primary tool of nuclear reactor analysis, multigroup diffusion theory. Part 4 illustrates the methods of nuclear reactor analysis by considering several important applications in reactor engineering, including a brief introduction to reactor core design. The text is written exclusively in S1 units and features numerous problems and exercises, many requiring digital computation.

1,753 citations

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Los Alamos National Laboratory

^{1}, National Nuclear Data Center^{2}, Oak Ridge National Laboratory^{3}, Argonne National Laboratory^{4}, Lawrence Livermore National Laboratory^{5}, National Institute of Standards and Technology^{6}, Rensselaer Polytechnic Institute^{7}, Idaho National Laboratory^{8}, Westinghouse Electric^{9}, Atomic Energy of Canada Limited^{10}, Nuclear Research and Consultancy Group^{11}TL;DR: The ENDF/B-VII.0 as discussed by the authors file contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes, based on experimental data and theory predictions.

Abstract: We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes, based on experimental data and theory predictions. The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, 6 Li, 10 B, Au and for 235,238 U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron- and proton-induced evaluations up to 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; (10) New methods for uncertainties and covariances, together with covariance evaluations for some sample cases; and (11) New actinide fission energy deposition. The paper provides an overview of this library, consisting of 14 sublibraries in the same ENDF-6 format as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched uranium thermal assemblies is removed; (b) The 238 U and 208 Pb reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good agreement for simulations of thermal high-enriched uranium assemblies is preserved; (d) The underprediction of fast criticality of 233,235 U and 239 Pu assemblies is removed; and (e) The intermediate spectrum critical assemblies are predicted more accurately. We anticipate that the new library will play an important role in nuclear technology applications, including transport simulations supporting national security, nonproliferation, advanced reactor and fuel cycle concepts, criticality safety, fusion, medicine, space applications, nuclear astrophysics, and nuclear physics facility design. The ENDF/B-VII.0 library is archived at the National Nuclear Data Center, BNL, and can be retrieved from www.nndc.bnl.gov .

1,752 citations