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AbstractAbstract

[en] Described in this paper is an analytic methodology for the solution of the neutron transport equation in slab geometry using P

_{N}method. The first part of the present methodology consists of obtaining a local general solution for the P_{N}equations with arbitrary order 𝑵 L ≤ Nand degree 𝑳 ≤ 𝑵 of scattering anisotropy. In the second part, the local general solution for the P_{N}equations was replaced in the scattering source of a simplified version of the linear Boltzmann transport equation, i. e., stationary, slab-geometry, monoenergetic, azimuthally symmetric, for non-multiplying media and isotropic internal source. This methodology has been implemented in a computer code developed on the MatLab® platform for Windows. As a result, in addition to generating numerical results for the scalar flux through the P_{N}method, the computer code generates numerical results for the angular flux at any position in the domain and for any direction not perpendicular to the domain. To evaluate the applicability of the P_{N}method and the analytic methodology, as described in this paper, numerical results for a model problem are presented. (author)Original Title

Implementação computacional de metodologia analítica de solução da equação de transporte de nêutrons em geometria planar utilizando o método P

_{N}Primary Subject

Source

Available from: file:///C:/Users/smvvianna/Downloads/1203-8294-1-PB.pdf

Record Type

Journal Article

Journal

Brazilian Journal of Radiation Sciences; ISSN 2319-0612; ; v. 9(1); 19 p

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INIS VolumeINIS Volume

INIS IssueINIS Issue

Oliva, A.M.; Alves Filho, H.; Silva, D.M.; García, C.R., E-mail: aoliva@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: cgh@instec.cu

AbstractAbstract

[en] In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that generates numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as a study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (S

_{N}) formulation, in slab geometry, multigroup approximation, with linearly anisotropic scattering, considering a heterogeneous domain with fixed-source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional fine-mesh Diamond Difference (DD) method and the coarse-mesh spectral Green's function (SGF) method to illustrate the method's accuracy and stability. The solution algorithms problem is implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)Primary Subject

Source

21. meeting on nuclear reactor physics and thermal hydraulics - ENFIR; Santos, SP (Brazil); 21-25 Oct 2019; Available from: https://www.bjrs.org.br/revista/index.php/REVISTA/article/view/618/688

Record Type

Journal Article

Literature Type

Conference

Journal

Brazilian Journal of Radiation Sciences; ISSN 2319-0612; ; v. 8(3B); 16 p

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AbstractAbstract

[en] In this work the latest developments a Monte Carlo simulator with continuous energy is reported. This simulator makes use of a sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum and the Maxwell-Boltzmann distribution. The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution, such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies. It is then shown how this procedure can emulate the up-scattering effect by the increase in the kinetic energy of the neutron population. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process. (author)

Primary Subject

Source

21. meeting on nuclear reactor physics and thermal hydraulics - ENFIR; Santos, SP (Brazil); 21-25 Oct 2019; Available from: https://www.bjrs.org.br/revista/index.php/REVISTA/article/view/401/681

Record Type

Journal Article

Literature Type

Conference

Journal

Brazilian Journal of Radiation Sciences; ISSN 2319-0612; ; v. 8(3B); 12 p

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AbstractAbstract

[en] This paper examines radiation-shielding abilities of oxyfluoro-tellurite-zinc glasses in the chemical form of AlF${}_{3}$-TeO${}_{2}$-ZnO under the substitution of AlF${}_{3}$ by ZnO. Gamma-ray- and neutron-shielding properties were tested in terms of mass attenuation coefficient (μ/ρ), half value layer, mean free path, effective atomic numbers (Z${}_{eff}$), effective electron density (N${}_{eff}$) and removal cross-section (Σ${}_{R}$). The μ/ρ values of the glasses were generated by Geant4 simulations over an extended energy range and then the generated data were confirmed via XCOM software. The results showed that both gamma-ray- and neutron-shielding efficiencies of the selected glasses evolved by substituting of AlF${}_{3}$ by ZnO. Nuclear radiation-shielding abilities of the current glass systems were compared with that of some conventional shielding materials and newly developed HMO glasses. It can be concluded that oxyfluoro-tellurite-zinc glasses could be useful to design novel shields for radiation protection applications.

Primary Subject

Source

Available from: http://dx.doi.org/10.1007/s00339-019-3265-6; AID: 88

Record Type

Journal Article

Literature Type

Numerical Data

Journal

Applied Physics. A, Materials Science and Processing; ISSN 0947-8396; ; CODEN APAMFC; v. 126(2); p. 1-12

Country of publication

BARYON REACTIONS, BARYONS, CHALCOGENIDES, CROSS SECTIONS, DATA, ELECTROMAGNETIC RADIATION, ELEMENTARY PARTICLES, FERMIONS, FLUORINE COMPOUNDS, HADRON REACTIONS, HADRONS, HALIDES, HALOGEN COMPOUNDS, INFORMATION, IONIZING RADIATIONS, NEUTRAL-PARTICLE TRANSPORT, NEUTRONS, NUCLEAR REACTIONS, NUCLEON REACTIONS, NUCLEONS, NUMERICAL DATA, OXIDES, OXYGEN COMPOUNDS, OXYHALIDES, RADIATION TRANSPORT, RADIATIONS, SIMULATION, TELLURIUM COMPOUNDS, ZINC COMPOUNDS

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INIS VolumeINIS Volume

INIS IssueINIS Issue

AbstractAbstract

[en] The hypothesis of the present investigation underlined with determination of possible synergistic effects of serpentine mineral additive on Li${}_{2}$B${}_{4}$O${}_{7}$ glasses. A group of Li${}_{2}$B${}_{4}$O${}_{7}$ glasses with serpentine mineral additive were synthesized by melt-quenching technique. The elemental analysis of two different Li${}_{2}$B${}_{4}$O${}_{7}$ glasses with different amount of serpentine additive is tested using energy-dispersive X-ray (EDX) technique. Next, the surface morphology of synthesized serpentine glasses was investigated with scanning electron microscopy (SEM). The optical features of synthesized serpentine glasses were determined along the wavelength ranged from 200 to 900 nm. Lastly, nuclear radiation shielding properties of Li${}_{2}$B${}_{4}$O${}_{7}$ glasses with serpentine mineral additive were determined for gamma rays, neutrons and charged particles. MCNPX (version 2.6.0) general-purpose Monte Carlo code has been utilized for mass attenuation coefficients calculations. The results showed that the spectra are decreasing with wavelength with an observed peak centered at 450 nm. Moreover, it is observed that serpentine mineral additive improves the gamma protecting capacity of Li${}_{2}$B${}_{4}$O${}_{7}$ glasses. It was also noticed that the addition of serpentine mineral also enhanced the neutron and charged particle absorption of the glasses.

Primary Subject

Source

Available from: http://dx.doi.org/10.1007/s00339-020-3397-8; AID: 208

Record Type

Journal Article

Literature Type

Numerical Data

Journal

Applied Physics. A, Materials Science and Processing (Print); ISSN 0947-8396; ; CODEN APAMFC; v. 126(3); p. 1-19

Country of publication

ABSORPTION SPECTRA, ATTENUATION, BORATES, CHARGED-PARTICLE TRANSPORT, EXPERIMENTAL DATA, GAMMA RADIATION, GLASS, KEV RANGE, LITHIUM COMPOUNDS, MEV RANGE 01-10, NEUTRON REACTIONS, NEUTRON TRANSPORT, NEUTRONS, OPACITY, PHOTON TRANSPORT, SCANNING ELECTRON MICROSCOPY, SERPENTINE, TOTAL CROSS SECTIONS, X RADIATION, X-RAY SPECTRA

ALKALI METAL COMPOUNDS, BARYON REACTIONS, BARYONS, BORON COMPOUNDS, CROSS SECTIONS, DATA, ELECTROMAGNETIC RADIATION, ELECTRON MICROSCOPY, ELEMENTARY PARTICLES, ENERGY RANGE, FERMIONS, HADRON REACTIONS, HADRONS, INFORMATION, IONIZING RADIATIONS, MEV RANGE, MICROSCOPY, MINERALS, NEUTRAL-PARTICLE TRANSPORT, NUCLEAR REACTIONS, NUCLEON REACTIONS, NUCLEONS, NUMERICAL DATA, OPTICAL PROPERTIES, OXYGEN COMPOUNDS, PHYSICAL PROPERTIES, RADIATION TRANSPORT, RADIATIONS, SILICATE MINERALS, SPECTRA

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INIS VolumeINIS Volume

INIS IssueINIS Issue

AbstractAbstract

[en] For 25 Li${}_{2}$O–(75 − x) B${}_{2}$O${}_{3}$–x Bi${}_{2}$O${}_{3}$ (where x = 0, 5, 10, 15, 20, 25, 30, 35, and 40 mol%) glasses, gamma-ray and neutrons attenuation features were explored by theoretical approach using ParShield/WinXCOM program, Geant4, and Penelope codes. At ${}^{133}$Ba (276, 303, 356, and 384 keV), ${}^{22}$Na (511 and 1280 keV), ${}^{137}$Cs (662 keV), ${}^{54}$Mn (835 keV), and ${}^{60}$Co (1170 and 1330 keV) photon peaks, for all samples, mass attenuation coefficient (μ/ρ), effective atomic number (Z${}_{eff}$), effective electron density (N${}_{eff}$), half-value layer (HVL), and mean free path (MFP) parameters have been evaluated using ParShield/WinXCOM program. The μ/ρ values computed by WinXCOM, Geant4, and Penelope codes were compared to check the accuracy, and satisfactory agreement among the values was identified. Moreover, using G–P fitting method as a function of penetration depth (1, 5, 10, 15, 20, 25, 30, 35, and 40 mfp) within the photon energy range of 0.015–15 MeV, exposure buildup factor (EBF) and energy absorption buildup factor (EABF) were derived. For all selected glasses, the effectiveness of the neutrons attenuation has been discussed in terms of macroscopic effective removal cross-section (Σ${}_{R}$), coherent scattering cross-section (σ${}_{cs}$), incoherent scattering cross-section (σ${}_{ics}$), absorption cross-section (σ${}_{A}$), and total neutron cross-section (σ${}_{T}$). The 'σ${}_{T}$' values have been calculated within 10${}^{\text{\xe2}\text{\u02c6}\text{\u2019}4}$–10${}^{\text{\xe2}\text{\u02c6}\text{\u2019}8}$ MeV neutron energy range using the Geant4 code. The μ/ρ possessed larger values at the lowest energy and lower values at higher energy regions for all studied glasses. The μ/ρ, Z${}_{eff}$, HVL, and MFP values showed enhanced γ-ray shielding capability with Bi${}_{2}$O${}_{3}$ content increment in the samples. The 25 Li${}_{2}$O–35 B${}_{2}$O${}_{3}$–40 Bi${}_{2}$O${}_{3}$ (mol%) sample by having larger Z${}_{eq}$ and/or Z${}_{eff}$ value, faired lower EBF and EABF values. Largest μ/ρ and Z${}_{eff}$, and minimal HVL, MFP, EBF, and EABF values of 25 Li${}_{2}$O–35 B${}_{2}$O${}_{3}$–40 Bi${}_{2}$O${}_{3}$ (mol%) glass demonstrated its superior γ-ray attenuation ability among all examined glasses. Further, among all glasses, 25 Li${}_{2}$O–75 B${}_{2}$O${}_{3}$ (mol%) sample exhibits relatively higher Σ${}_{R}$ (0.11326 cm${}^{\text{\xe2}\text{\u02c6}\text{\u2019}1}$) and ‘σ${}_{T}$’ (46.109 cm${}^{\text{\xe2}\text{\u02c6}\text{\u2019}1}$ → 0.84607 cm${}^{\text{\xe2}\text{\u02c6}\text{\u2019}1}$ from 1 × 10${}^{\text{\xe2}\text{\u02c6}\text{\u2019}8}$ MeV → 1×10${}^{\text{\xe2}\text{\u02c6}\text{\u2019}4}$ MeV neutron energy) values for fast and thermal neutrons attenuation, respectively, indicating its better neutrons absorption competence.

Primary Subject

Source

Available from: http://dx.doi.org/10.1007/s00339-020-3418-7; AID: 249

Record Type

Journal Article

Literature Type

Numerical Data

Journal

Applied Physics. A, Materials Science and Processing (Print); ISSN 0947-8396; ; CODEN APAMFC; v. 126(4); p. 1-16

Country of publication

ATTENUATION, BISMUTH OXIDES, BORON OXIDES, ELECTRON DENSITY, FAST NEUTRONS, GAMMA RADIATION, GLASS, INTEGRAL CROSS SECTIONS, KEV RANGE 100-1000, KEV RANGE 10-100, LITHIUM OXIDES, MEAN FREE PATH, MEV RANGE 01-10, MEV RANGE 10-100, NEUTRON REACTIONS, NEUTRON TRANSPORT, PHOTON TRANSPORT, THEORETICAL DATA, THERMAL NEUTRONS, TOTAL CROSS SECTIONS

ALKALI METAL COMPOUNDS, BARYON REACTIONS, BARYONS, BISMUTH COMPOUNDS, BORON COMPOUNDS, CHALCOGENIDES, CROSS SECTIONS, DATA, ELECTROMAGNETIC RADIATION, ELEMENTARY PARTICLES, ENERGY RANGE, FERMIONS, HADRON REACTIONS, HADRONS, INFORMATION, IONIZING RADIATIONS, KEV RANGE, LITHIUM COMPOUNDS, MEV RANGE, NEUTRAL-PARTICLE TRANSPORT, NEUTRONS, NUCLEAR REACTIONS, NUCLEON REACTIONS, NUCLEONS, NUMERICAL DATA, OXIDES, OXYGEN COMPOUNDS, RADIATION TRANSPORT, RADIATIONS

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Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

AbstractAbstract

[en] Throughout history, energy has played a fundamental role in human's progress living. To promote nuclear power to meet the future energy needs, ten countries including Argentina, South Africa, the United States, the United Kingdom, Brazil, Japan, Switzerland, France, Canada and Korea in a global effort (Generation IV International Forum - GIF) have agreed to investigate the next generation of nuclear energy systems known as 4${}^{th}$ generation. These reactors are expected to enter the market after 2030. Fundamental changes in the configuration of the systems and the forms of the old reactors have led to the production of new reactors, which require basic research and development, careful examination, and the construction of semi-industrial units. The capabilities of fourth-generation reactors are seawater desalination, and thermal applications in addition to the production of electricity. In 2000, the founding countries of GIF formed their first meeting to discuss the need for conduct research on the design of next-generation reactors. Subsequently, a strategy was put forward to direct the activities, and the implementation responsibility was assigned to the US Department of Energy. In this research, we investigate the neutron behavior of the advanced reactor core with lead coolant ALFRED. The purpose of the neutron calculations of the core of a reactor is to calculate the distribution of neutron flux in the center and to calculate the effective reproduction coefficient. Given the necessity of performing lattice pitch neutron calculations, it is initially required to determine the real geometry of the core, as well as the order and fuel richness, the lattice pitch the grid, the radius and height of the fuel rods, the composition and location of the fuel absorbents, the types and locations of the control rods, the fuel complex arrangement, and radial and axial peaking factor. The MCNPX code is used to perform neutron calculations.

Primary Subject

Record Type

Journal Article

Journal

Atw. Internationale Zeitschrift fuer Kernenergie; ISSN 1431-5254; ; v. 65(3); p. 142-144

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INIS IssueINIS Issue

Tasaki, S.; Idobata, Y.; Adachi, Y.; Funama, F.; Abe, Y.

EPJ Web of Conferences, Proceedings of the 8. international meeting of union for compact accelerator-driven neutron sources (UCANS-8)

EPJ Web of Conferences, Proceedings of the 8. international meeting of union for compact accelerator-driven neutron sources (UCANS-8)

AbstractAbstract

[en] Neutron moderation properties from the cold mesitylene moderator have been studied. The Kyoto University Accelerator driven Neutron Source (KUANS) has been used for these experiments. In KUANS neutrons are produced by

^{9}Be(p,n)^{9}B reaction using a pulsed 3.5 MeV proton beam. The neutrons are moderated by the polyethylene. The container of the mesitylene moderator is situated in front of the polyethylene moderator and the change of the time of flight spectrum has been recorded as a function of the temperature of the mesitylene moderator. By fitting the Maxwell distribution to the obtained TOF spectra, the neutron temperature corresponding to the mesitylene temperature (from room temperature to 28 K) has been estimated. Measured neutron spectra have changed corresponding to the mesitylene temperature. As the mesitylene temperature decreases, the peak of the neutron spectrum shifts to longer wavelength sidePrimary Subject

Source

Ott, F. (ed.); Menelle, A. (ed.); Alba-Simionesco, C. (ed.); EDP Sciences, 17, Avenue du Hoggar, Parc d'Activite de Courtaboeuf, BP 112, F-91944 Les Ulis Cedex A (France); v. 231 [168 p.]; 2020; p. 04005.p.1-04005.p.3; UCANS-8: 8. international meeting of union for compact accelerator-driven neutron sources; Paris (France); 8-11 Jul 2019; Available from doi: https://doi.org/10.1051/epjconf/202023104005; 3 refs.

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Book

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Related RecordRelated Record

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Libotte, R.B.; Alves Filho, H.; Barros, R.C., E-mail: rafaellibotte@hotmail.com, E-mail: halves@iprj.uerj.br, E-mail: rcbarros@pq.cnpq

AbstractAbstract

[en] In this paper, we propose a new deterministic numerical methodology to solve the one-dimensional linearized Boltzmann equation applied to neutron shielding problems (fixed-source), using the transport equation in the discrete ordinates formulation (SN) considering the multigroup theory. This is a hybrid methodology, entitled Modified Spectral Deterministic Method (SDM-M), that derives from the Spectral Deterministic Method (SDM) and Diamond Difference (DD) methods. This modification in the SDM method aims to calculate neutron scalar fluxes with lower computational cost. Two model-problems are solved using the SDM-M, and the results are compared to the coarse-mesh methods SDM, Spectral Green's Function (SGF) and Response Matrix (RM), and the fine-mesh method DD. The numerical results were obtained in the programming language JAVA version 1.8.0

_{9}1. (author)Primary Subject

Source

21. meeting on nuclear reactor physics and thermal hydraulics - ENFIR; Santos, SP (Brazil); 21-25 Oct 2019; 6. meeting on nuclear industry - ENIN; Santos, SP (Brazil); 21-25 Oct 2019; Available from: https://www.bjrs.org.br/revista/index.php/REVISTA/article/view/1421/660

Record Type

Journal Article

Literature Type

Conference

Journal

Brazilian Journal of Radiation Sciences; ISSN 2319-0612; ; v. 8(3A); 16 p

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AbstractAbstract

[en] Investigation of the unexplained changes of neutron flux fluctuation magnitudes observed in KWU-built PWRs has drawn attention to long known but still incompletely understood spatial correlation patterns of the neutron flux fluctuations in the frequency range 0-2 Hz. These patterns, namely an out-of-phase behavior of signals from oppositely located core quadrants and an in-phase behavior of signals from axially aligned locations, are the dominant fluctuation phenomena because the range 0-2 Hz carries more than 95 % of the power of the signal fluctuations and the coherence functions of respective signal pairings have values between 0.5 and 1.0 in this frequency range. Therefore, finding the mechanism effecting the measured fluctuation patterns is believed to be key to explain the changes of the fluctuation amplitudes.

Primary Subject

Record Type

Journal Article

Journal

Atw. Internationale Zeitschrift fuer Kernenergie; ISSN 1431-5254; ; v. 65(6-7); p. 346-349

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