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AbstractAbstract

[en] Complex eigenvalues for the monoenergetic neutron transport equation in the buckling approximation have been calculated for various combinations of linearly and quadratically anisotropic scattering. The results are discussed in terms of the time-dependent case. Tables are given of complex bucklings for real decay constants and of complex decay constants for real bucklings. The results fit nicely into the pattern of real and purely imaginary eigenvalues obtained earlier. (author)

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 18(1); p. 1-5

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[en] A method to obtain an approximate solution to the transport equation, when both sources and collisions show a linearly anisotropic behavior, is outlined and the possible implications for numerical calculations in applied neutronics as well as shielding evaluations are investigated. The form of the differential system of equations taken by the method is quite handy and looks simpler and more manageable than any other today available technique. To go deeper into the efficiency of the method, some typical calculations concerning critical dimension of multiplying systems are then performed and the results are compared with the ones coming from the classical Ssub(N) approximations. The outcome of such calculations leads us to think of interesting developments of the method which could be quite useful in alternative to other today widespread approximate procedures, for any geometry, but especially for curved ones. (author)

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 20(10); p. 822-831

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Higuchi, Kenji; Asai, Kiyoshi; Yamazaki, Takashi.

Japan Atomic Energy Research Inst., Tokyo

Japan Atomic Energy Research Inst., Tokyo

AbstractAbstract

[en] We have vectorized the multi-group Monte Carlo code MORSE-DD. The Monte Carlo method is very powerful to simulate the motion of particles. However the Monte Carlo codes spend much calculation time, it is required to accelerate the codes by using vector processing. It was known that vectorization of the Monte Carlo codes for treating with the neutron transport problem is very difficult, because each neutron has a different motion and there are many paths in a Monte Carlo step. The Monte Carlo step of MORSE-DD code is completely rewritten by using event bank method. As the result, we have obtained speedup 1.55 times on FACOM VP-100 compared with the original code in scalar mode for problem to analyze neutron behaviour in FNS (Fusion Neutron Source). In this report, the techniques employed for vectorization and the essential problems which prevent from the efficient use of vector processor are described. (author)

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Feb 1987; 53 p

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Report

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AbstractAbstract

[en] We investigate transport theory for anisotropic transport of neutrons in finite medium or injected externally. When anisotropic transport is treated by the usual transport equation, on which reversibility of collisions is shown imposed, successive collisions always induce 'self-collision' or sham collision; the fact is unavoidable as long as statistical ensemble is constructed from the reductionistic mechanical-systems. Then, irreductionistic elements, or spatial cells containing assembly of free neutrons (and implicit medium nuclei) uniformly are introduced, from which alternative Liouville equation is constructed. Successive collisions are expressed by fusing three cells; for reviving mechanical law in the collisions the law of action and reaction is applied to between first fused-cell and third cell. Extended transport equation can thus describe the process of chaotically mixing anisotropic momentum, i.e., the well-known deep penetration. (author)

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Japan Atomic Energy Research Inst., Tokyo (Japan); 906 p; Mar 2000; p. 459-465; ICRS-9: 9. international conference on radiation shielding; Tsukuba, Ibaraki (Japan); 17-22 Oct 1999; Available from the Internet at URL https://doi.org/10.1080/00223131.2000.10874928; 7 refs.; This record replaces 32001147

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[en] Neutron skyshine dose evaluation is one of safety evaluation terms for nuclear facilities. However, evaluation method of neutron skyshine dose is not fully established, so we studied following terms. 1. We developed simple neutron skyshine dose calculation code using a personal computer which could also calculate attenuation effect of simple shielding structure. This code reduce computing time by 1000 or 10000 times than present evaluation method using DOT. Neutron skyshine dose calculated with this code is more than DOT'S result by about 7 times, however this code can be used to pre-liminaly design or pre-liminaly analysis for order estimate. 2. We elavuated meteorogical and topographical effects on skyshine dose distribution. We found out cloud has negligible effect, but (1) moisture, (2) ground surface configulation and (3) evaluation point altitude have remarkable effect. And ground surface configulation is most remarkable. (author)

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Journal Article

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Denryoku Chuo Kenkyusho Hokoku; CODEN DCKHD; (no.285064); p. 1-52

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AbstractAbstract

[en] The time-dependent P

_{1}equation for two-dimensional neutron transport is numerically solved by a finite difference approximation of the explicit form along the bicharacteristics of the P_{1}equation. Applying von Neumann's stability condition to this numerical procedure in an infinite space, we can derive the condition necessary for the solution to be stable. This condition is that the mesh widths satisfy the inequality 0< lambda<=√3/2 with lambda = time mesh Δt/space mesh Δx or Δz, where the time t is measured in units of inverse neutron speed 1/v. The sufficient stability condition on the ratio lambda is to be determined by numerical experiments. It has been found that the upper bound of lambda becomes larger for smaller values of space mesh width. In respect of the stability of numerical solution, the P_{1}approximation is more advantageous than the diffusion approximation. Transient behavior of neutron flux distribution due to a stationary neutron source is numerically determined assuming zero initial values. After the transient state terminates, the steady state distribution is obtained. (author)Primary Subject

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Journal Article

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 15(9); p. 645-657

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AbstractAbstract

No abstract available

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Published in summary form only.

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Journal Article

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 17(4); p. 315-317

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Pounders, Justin; Rahnema, Farzad

Proceedings of SNA + MC2010: Joint international conference on supercomputing in nuclear applications + Monte Carlo 2010 Tokyo

Proceedings of SNA + MC2010: Joint international conference on supercomputing in nuclear applications + Monte Carlo 2010 Tokyo

AbstractAbstract

[en] A new deterministic-Monte Carlo hybrid solution technique is derived for the time-dependent transport equation. This new approach is based on dividing the time domain into a number of coarse intervals and expanding the transport solution in a series of polynomials within each interval. The solutions within each interval can be represented in terms of arbitrary source terms by using precomputed response functions. In the current work, the time-dependent response function computations are performed using the Monte Carlo method, while the global time-step march is performed deterministically. This work extends previous work by coupling the time-dependent expansions to space- and angle-dependent expansions to fully characterize the 1D transport response/solution. More generally, this approach represents and incremental extension of the steady-state coarse-mesh transport method that is based on global-local decompositions of large neutron transport problems. An example of a homogeneous slab is discussed as an example of the new developments. (author)

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Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); [1630 p.]; 2010; [6 p.]; SNA + MC2010: Joint international conference on supercomputing in nuclear applications and Monte Carlo 2010 Tokyo; Tokyo (Japan); 17-21 Oct 2010; Available from Japan Atomic Energy Agency, 4-49 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1184, Japan; Available as CD-ROM Data in PDF format, Folder Name: pdf, Paper ID: 10391.pdf

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Miscellaneous

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[en] An efficient albedo Monte Carlo method newly developed has been studied by analyzing two types of experiments on neutron streaming. The method is characterized by employing the energy-angle dependent doubly differential albedos for slab, which can be calculated in a short computer time with a one-dimensional transport theory, such as the Sn method and more efficient invariant imbedding method. This paper describes the features of the present albedo Monte Carlo method, including fundamental formulas. In the analyses of the neutron streaming experiments, the calculated results agreed with the measured data within a factor of 2 for a benchmark experiment at the YAYOI reactor and within a factor of 3 for an SNR sodium duct mock-up experiment. It is concluded that the present albedo Monte Carlo method is practical and applicable to the reactor shielding analysis concerning radiation streaming. (author)

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; CODEN JNSTA; v. 23(11); p. 937-948

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AbstractAbstract

No abstract available

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Published in summary form only.

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Journal Article

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 20(7); p. 620-623

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