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[en] This paper describes recent advances in the transport theory by the method of invariant embedding applicable to the transport problems of gamma rays and neutrons with realistic cross-sections. It includes the following three topics: 1. Generation of gamma ray albedo data for semi-infinite medium of various materials by the method of invariant embedding. 2. Development of a new semi-analytical method, called 'angular eigenvalue method', for radiation transport problems in slabs, and applications of the method to the penetrations of obliquely incident gamma rays in slabs. The analytical formulae for the buildup factors of gamma rays from a point isotropic source in an infinite homogeneous medium as well as those from a plane isotropic source were derived based on the method. 3. Calculations of gamma ray albedos for slabs of finite thickness by the invariant embedding method. (author)

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Japan Atomic Energy Research Inst., Tokyo (Japan); Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan); 615 p; Sep 1996; p. A/90-A/99; PHYSOR96: international conference on the physics of reactors 1996; Mito (Japan); 16-20 Sep 1996; Available from Atomic Energy Society of Japan

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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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AbstractAbstract

[en] In this paper the streaming term of the one-speed transport equation is expressed as a duality operation of modern tensor analysis. This representation allows one to write the equation easily in any orthogonal coordinate system. An application of the present transformation technique is made to three different geometries of current interest in controlled thermonuclear reactor design, namely cylindrical and toroidal (with both circular and general elliptic cross sections). (author)

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 19(3); p. 175-185

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Takeda, T.; Yamasaki, M.; Ikeda, H.; Nishigori, T.

Proceedings of the 7th topical meeting on nuclear code development

Proceedings of the 7th topical meeting on nuclear code development

AbstractAbstract

[en] Results are reported of NEACRP '3D Neutron Transport Benchmarks' proposed from Osaka UNiversity, and of recent progress in the development of a 3D neutron transport code. Takeda et al. proposed four problems to NEACRP as 3D neutron transport benchmarks, and 22 results from 20 organizations were submitted. A variety of methods have been used, such as the Monte Carlo, Sn, Pn, synthetic, and nodal method. The results for k-eff, control-rod worths, and region-averaged fluxes are summarized with the conclusions that (1) in XYZ geometry the Sn method with n=8 shows a good agreement with the Monte-Carlo method, and gives even better results in some cases, (2) the Pn method has significant spatial mesh effects, and (3) the Sn method is not satisfactory in hexagonal-Z geometry, and improvements in accuracy are desirable. Improvement of a 3D neutron transport code is in progress to resolve the problem in the hexagonal-Z geometry by considering new diamond difference schemes and an improved coarse-mesh method, and also by applying the nodal method. (author)

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Japan Atomic Energy Research Inst., Tokyo (Japan); 189 p; Mar 1992; p. 96-111; 7. topical meeting on nuclear code development; Tokai, Ibaraki (Japan); 30-31 Oct 1991

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[en] A Monte Carlo code, MORSE-CV, which can calculate the covariance of the scalar neutron flux spectrum, was developed. The method and input instructions are given in this report. (author)

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Bulletin of the Research Laboratory for Nuclear Reactors (Tokyo Institute of Technology); ISSN 0387-6144; ; CODEN BRLTD; v. 13 p. 3-11

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AbstractAbstract

[en] A Monte-Carlo photon and neutron transport code was developed at OAEP. The code was written in C and C++ languages in an object-oriented programming style. Constructive solid geometry (CSG), rather than combinatorial, was used such that making its input file more readable and recognizable. As the first stage of code validation, data from some ENDF files, in the MCNP's specific format, were used and compared with experimental data. The neutron (from a 300 mCi Am/Be source) attenuation by water was chosen to compare the results. The agreement of the quantity 1/Σ among the calculation from SIPHON and MCNP, and the experiment - which are 10.39 cm, 9.71 cm and 10.25 cm respectively - was satisfactorily well within the experimental uncertainties. These results also agree with the 10.8 cm result of N.M., Mirza, et al. (author)

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Yamano, Naoki (ed.) (Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)); Fukahori, Tokio (ed.) (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment); Japan Atomic Energy Research Inst., Tokyo (Japan); 364 p; Mar 2001; p. 45-50; 2000 symposium on nuclear data; Tokai, Ibaraki (Japan); 16-17 Nov 2000; 8 refs., 4 figs., 2 tabs.

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Whitlow, J.D.; Neuhauser, K.S.

The 10th international symposium on the packaging and transportation of radioactive materials

The 10th international symposium on the packaging and transportation of radioactive materials

AbstractAbstract

[en] This paper will describe a methodology which has been developed to allow accident probabilities associated with one severity category scheme to be transferred to another severity category scheme, permitting some comparisons of different studies at the category level. In this methodology, the severity category schemes to be compared are mapped onto a common set of axes. The axes represent critical accident environments (e.g., impact, thermal, crush, puncture) and indicate the range of accident parameters from zero (no accident) to the most sever credible forces. The choice of critical accident environments for the axes depends on the package being transported and the mode of transportation. The accident probabilities associated with one scheme are then transferred to the other scheme. This transfer of category probabilities is based on the relationships of the critical accident parameters to probability of occurrence. The methodology can be employed to transfer any quantity between category schemes if the appropriate supporting information is available. (J.P.N.)

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1709 p; 1993; v. 2 p. 799-806; PATRAM'92: 10. international symposium on the packaging and transportation of radioactive materials; Yokohama (Japan); 13-18 Sep 1992

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AbstractAbstract

[en] As a practical variance reduction technique applicable to Monte Carlo shielding calculations, the present article shows a new simple biased sampling technique on particle flight directions. Scattered particles not directed towards the detector positions are killed if they are not so important, that is, if the particle weights are sufficiently small compared to the source weight. In this way, we can reduce the sample size required for obtaining an accurate estimate for the detector response. The present technique was incorporated into the multigroup neutron and γ-ray transport code MORSE, and sample calculations were performed on spherical fast neutron systems. The results have shown that this biased technique is effective for dealing with neutron multiplication as well as neutron transmission problems. The neutron flux or the effective multiplication factor of a nuclear reactor is estimated more accurately than from the method of path-length stretching with about the same computation time. In addition, it is shown that the flight-direction biasing can further effectively be used by combining it with other variance reduction techniques. (auth.)

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Journal of Nuclear Science and Technology (Tokyo); v. 14(8); p. 603-609

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Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

Japan Atomic Energy Research Inst., Tokyo (Japan)

Japan Atomic Energy Research Inst., Tokyo (Japan)

AbstractAbstract

[en] In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)

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Jun 2005; 412 p; Also available from JAEA; 65 refs., 41 figs.

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Nishida, Takahiko; Horikami, Kunihiko; Suzuki, Tadakazu; Nakahara, Yasuaki; Taji, Yukichi

Japan Atomic Energy Research Inst., Tokyo

Japan Atomic Energy Research Inst., Tokyo

AbstractAbstract

[en] The coarse-mesh rebalancing technique is introduced into the general-purpose neutron and gamma-ray Monte Carlo transport code MORSE, to accelerate the convergence rate of the iteration process for eigenvalue calculation in a nuclear reactor system. Two subroutines are thus attached to the code. One is bookkeeping routine 'COARSE' for obtaining the quantities related with the neutron balance in each coarse mesh cell, such as the number of neutrons absorbed in the cell, from random walks of neutrons in a batch. The other is rebalance factor calculation routine 'REBAL' for obtaining the scaling factor whereby the neutron flux in the cell is multiplied to attain the neutron balance. The two subroutines and algorithm of the coarse mesh rebalancing acceleration in a Monte Carlo game are described. (auth.)

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Sep 1975; 36 p

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ALGORITHMS, COMPUTER CALCULATIONS, EIGENVALUES, GAMMA TRANSPORT THEORY, ITERATIVE METHODS, M CODES, MONTE CARLO METHOD, NEUTRON TRANSPORT, NEUTRON TRANSPORT THEORY, NUMERICAL SOLUTION, ONE-DIMENSIONAL CALCULATIONS, PHOTON TRANSPORT, RANDOMNESS, THREE-DIMENSIONAL CALCULATIONS, TWO-DIMENSIONAL CALCULATIONS

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