Solution and study of nodal neutron transport equation applying the LTSN-DiagExp method

International Nuclear Information System (INIS)

Hauser, Eliete Biasotto; Pazos, Ruben Panta; Vilhena, Marco Tullio de; Barros, Ricardo Carvalho de

2003-01-01

In this paper we report advances about the three-dimensional nodal discrete-ordinates approximations of neutron transport equation for Cartesian geometry. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtained the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. (author)

Spatially adaptive hp refinement approach for PN neutron transport equation using spectral element method

International Nuclear Information System (INIS)

Nahavandi, N.; Minuchehr, A.; Zolfaghari, A.; Abbasi, M.

2015-01-01

Highlights: • Powerful hp-SEM refinement approach for P N neutron transport equation has been presented. • The method provides great geometrical flexibility and lower computational cost. • There is a capability of using arbitrary high order and non uniform meshes. • Both posteriori and priori local error estimation approaches have been employed. • High accurate results are compared against other common adaptive and uniform grids. - Abstract: In this work we presented the adaptive hp-SEM approach which is obtained from the incorporation of Spectral Element Method (SEM) and adaptive hp refinement. The SEM nodal discretization and hp adaptive grid-refinement for even-parity Boltzmann neutron transport equation creates powerful grid refinement approach with high accuracy solutions. In this regard a computer code has been developed to solve multi-group neutron transport equation in one-dimensional geometry using even-parity transport theory. The spatial dependence of flux has been developed via SEM method with Lobatto orthogonal polynomial. Two commonly error estimation approaches, the posteriori and the priori has been implemented. The incorporation of SEM nodal discretization method and adaptive hp grid refinement leads to high accurate solutions. Coarser meshes efficiency and significant reduction of computer program runtime in comparison with other common refining methods and uniform meshing approaches is tested along several well-known transport benchmarks

Development of a CAD-based neutron transport code with the method of characteristics

International Nuclear Information System (INIS)

Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing

2012-01-01

The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)

Multi-group transport methods for high-resolution neutron activation analysis

International Nuclear Information System (INIS)

Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

2009-01-01

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

Method for calculating anisotropic neutron transport using scattering kernel without polynomial expansion

International Nuclear Information System (INIS)

Takahashi, Akito; Yamamoto, Junji; Ebisuya, Mituo; Sumita, Kenji

1979-01-01

A new method for calculating the anisotropic neutron transport is proposed for the angular spectral analysis of D-T fusion reactor neutronics. The method is based on the transport equation with new type of anisotropic scattering kernels formulated by a single function I sub(i) (μ', μ) instead of polynomial expansion, for instance, Legendre polynomials. In the calculation of angular flux spectra by using scattering kernels with the Legendre polynomial expansion, we often observe the oscillation with negative flux. But in principle this oscillation disappears by this new method. In this work, we discussed anisotropic scattering kernels of the elastic scattering and the inelastic scatterings which excite discrete energy levels. The other scatterings were included in isotropic scattering kernels. An approximation method, with use of the first collision source written by the I sub(i) (μ', μ) function, was introduced to attenuate the ''oscillations'' when we are obliged to use the scattering kernels with the Legendre polynomial expansion. Calculated results with this approximation showed remarkable improvement for the analysis of the angular flux spectra in a slab system of lithium metal with the D-T neutron source. (author)

An introduction to neutron transport

International Nuclear Information System (INIS)

Wiesenfeld, Bernard

2015-01-01

Neutron transport science is the study of neutron transport in a nuclear reactor and of associated nuclear reactions, notably fission reactions. Heat released by these reactions can be used for several purposes: electricity production, hydrogen production, sea water desalination, urban heating, naval propulsion, space propulsion, and so on. This publication contains the course proposed at Mines ParisTech and at the Arts et Metiers ParisTech. It is an introduction to neutron transport science and aims at presenting fundamental physical principles of this original branch of nuclear physics, a so called 'low energies' branch whereas 'high energy' nuclear physics focuses on elementary particles. It addresses complex computation methods which have been developed during the last decades with computation codes of always higher performance. The first part presents elements of atom physics: origin of matter, properties of nuclei and atoms, notion of quantum mechanics, interaction between radiation and matter (ray absorption, Compton Effect and scattering, photoelectric effect). The second part introduces neutron transport by addressing the following issues: nuclear structure, the various aspects of the interaction between neutrons and matter, the evolution of the reactivity of a reactor in normal operation, the chain fission reaction kinetics, and neutron slowing down. The third part addresses various aspects of neutron transport calculation: expression of neutron assessment, scattering approximation, critical condition of a nuclear reactor, introduction to transport theory, peculiarities of fast breeder reactors. The last chapter 'from theory to practice' addresses the approach of the neutron scientist, proposes an overview of the main calculation codes, and presents fields of application (within or without nuclear fission)

Novel Parallel Numerical Methods for Radiation and Neutron Transport

International Nuclear Information System (INIS)

Brown, P N

2001-01-01

In many of the multiphysics simulations performed at LLNL, transport calculations can take up 30 to 50% of the total run time. If Monte Carlo methods are used, the percentage can be as high as 80%. Thus, a significant core competence in the formulation, software implementation, and solution of the numerical problems arising in transport modeling is essential to Laboratory and DOE research. In this project, we worked on developing scalable solution methods for the equations that model the transport of photons and neutrons through materials. Our goal was to reduce the transport solve time in these simulations by means of more advanced numerical methods and their parallel implementations. These methods must be scalable, that is, the time to solution must remain constant as the problem size grows and additional computer resources are used. For iterative methods, scalability requires that (1) the number of iterations to reach convergence is independent of problem size, and (2) that the computational cost grows linearly with problem size. We focused on deterministic approaches to transport, building on our earlier work in which we performed a new, detailed analysis of some existing transport methods and developed new approaches. The Boltzmann equation (the underlying equation to be solved) and various solution methods have been developed over many years. Consequently, many laboratory codes are based on these methods, which are in some cases decades old. For the transport of x-rays through partially ionized plasmas in local thermodynamic equilibrium, the transport equation is coupled to nonlinear diffusion equations for the electron and ion temperatures via the highly nonlinear Planck function. We investigated the suitability of traditional-solution approaches to transport on terascale architectures and also designed new scalable algorithms; in some cases, we investigated hybrid approaches that combined both

Monte Carlo simulation of neutron transport phenomena

International Nuclear Information System (INIS)

Srinivasan, P.

2009-01-01

Neutron transport is one of the central problems in nuclear reactor related studies and other applied sciences. Some of the major applications of neutron transport include nuclear reactor design and safety, criticality safety of fissile material handling, neutron detector design and development, nuclear medicine, assessment of radiation damage to materials, nuclear well logging, forensic analysis etc. Most reactor and dosimetry studies assume that neutrons diffuse from regions of high to low density just like gaseous molecules diffuse to regions of low concentration or heat flow from high to low temperature regions. However while treatment of gaseous or heat diffusion is quite accurately modeled, treatment of neutron transport as simple diffusion is quite limited. In simple diffusion, the neutron trajectories are irregular, random and zigzag - where as in neutron transport low reaction cross sections (1 barn- 10 -24 cm 2 ) lead to long mean free paths which again depend on the nature and irregularities of the medium. Hence a more accurate representation of the neutron transport evolved based on the Boltzmann equation of kinetic gas theory. In fact the neutron transport equation is a linearized version of the Boltzmann gas equation based on neutron conservation with seven independent variables. The transport equation is difficult to solve except in simple cases amenable to numerical methods. The diffusion (equation) approximation follows from removing the angular dependence of the neutron flux

A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods

International Nuclear Information System (INIS)

Zhang, H.; Zheng, Y.; Wu, H.; Cao, L.

2013-01-01

A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)

A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods

Energy Technology Data Exchange (ETDEWEB)

Zhang, H.; Zheng, Y.; Wu, H.; Cao, L. [School of Nuclear Science and Technology, Xi' an Jiaotong University, No. 28, Xianning West Road, Xi' an, Shaanxi 710049 (China)

2013-07-01

A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)

Resolution of the neutron transport equation by a three-dimensional least square method

International Nuclear Information System (INIS)

Varin, Elisabeth

2001-01-01

The knowledge of space and time distribution of neutrons with a certain energy or speed allows the exploitation and control of a nuclear reactor and the assessment of the irradiation dose about an irradiated nuclear fuel storage site. The neutron density is described by a transport equation. The objective of this research thesis is to develop a software for the resolution of this stationary equation in a three-dimensional Cartesian domain by means of a deterministic method. After a presentation of the transport equation, the author gives an overview of the different deterministic resolution approaches, identifies their benefits and drawbacks, and discusses the choice of the Ressel method. The least square method is precisely described and then applied. Numerical benchmarks are reported for validation purposes

UN Method For The Critical Slab Problem In One-Speed Neutron Transport Theory

International Nuclear Information System (INIS)

Oeztuerk, Hakan; Guengoer, Sueleyman

2008-01-01

The Chebyshev polynomial approximation (U N method) is used to solve the critical slab problem in one-speed neutron transport theory using Marshak boundary condition. The isotropic scattering kernel with the combination of forward and backward scattering is chosen for the neutrons in a uniform finite slab. Numerical results obtained by the U N method are presented in the tables together with the results obtained by the well-known P N method for comparison. It is shown that the method converges rapidly with its easily executable equations.

Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

International Nuclear Information System (INIS)

Shafii, Mohammad Ali; Meidianti, Rahma; Wildian,; Fitriyani, Dian; Tongkukut, Seni H. J.; Arkundato, Artoto

2014-01-01

Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation

Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

Energy Technology Data Exchange (ETDEWEB)

Shafii, Mohammad Ali, E-mail: mashafii@fmipa.unand.ac.id; Meidianti, Rahma, E-mail: mashafii@fmipa.unand.ac.id; Wildian,, E-mail: mashafii@fmipa.unand.ac.id; Fitriyani, Dian, E-mail: mashafii@fmipa.unand.ac.id [Department of Physics, Andalas University Padang West Sumatera Indonesia (Indonesia); Tongkukut, Seni H. J. [Department of Physics, Sam Ratulangi University Manado North Sulawesi Indonesia (Indonesia); Arkundato, Artoto [Department of Physics, Jember University Jember East Java Indonesia (Indonesia)

2014-09-30

Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation.

Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

International Nuclear Information System (INIS)

Chen, G.S.

1997-01-01

We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used. TFQMR (transpose free quasi-minimal residual algorithm), CGS (conjuage gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These sub-routines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. (author)

3-D neutron transport benchmarks

International Nuclear Information System (INIS)

Takeda, T.; Ikeda, H.

1991-03-01

A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes

Numerical simulation for fractional order stationary neutron transport equation using Haar wavelet collocation method

Energy Technology Data Exchange (ETDEWEB)

Saha Ray, S., E-mail: santanusaharay@yahoo.com; Patra, A.

2014-10-15

Highlights: • A stationary transport equation has been solved using the technique of Haar wavelet collocation method. • This paper intends to provide the great utility of Haar wavelets to nuclear science problem. • In the present paper, two-dimensional Haar wavelets are applied. • The proposed method is mathematically very simple, easy and fast. - Abstract: In this paper the numerical solution for the fractional order stationary neutron transport equation is presented using Haar wavelet Collocation Method (HWCM). Haar wavelet collocation method is efficient and powerful in solving wide class of linear and nonlinear differential equations. This paper intends to provide an application of Haar wavelets to nuclear science problems. This paper describes the application of Haar wavelets for the numerical solution of fractional order stationary neutron transport equation in homogeneous medium with isotropic scattering. The proposed method is mathematically very simple, easy and fast. To demonstrate about the efficiency and applicability of the method, two test problems are discussed.

TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222

International Nuclear Information System (INIS)

Shen, H.; Li, Z.; Wang, K.; Yu, G.

2010-01-01

A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)

Symmetrized neutron transport equation and the fast Fourier transform method

International Nuclear Information System (INIS)

Sinh, N.Q.; Kisynski, J.; Mika, J.

1978-01-01

The differential equation obtained from the neutron transport equation by the application of the source iteration method in two-dimensional rectangular geometry is transformed into a symmetrized form with respect to one of the angular variables. The discretization of the symmetrized equation leads to finite difference equations based on the five-point scheme and solved by use of the fast Fourier transform method. Possible advantages of the approach are shown on test calculations

A domian Decomposition Method for Transient Neutron Transport with Pomrning-Eddington Approximation

International Nuclear Information System (INIS)

Hendi, A.A.; Abulwafa, E.E.

2008-01-01

The time-dependent neutron transport problem is approximated using the Pomraning-Eddington approximation. This approximation is two-flux approximation that expands the angular intensity in terms of the energy density and the net flux. This approximation converts the integro-differential Boltzmann equation into two first order differential equations. The A domian decomposition method that used to solve the linear or nonlinear differential equations is used to solve the resultant two differential equations to find the neutron energy density and net flux, which can be used to calculate the neutron angular intensity through the Pomraning-Eddington approximation

Neutron transport study based on assembly modular ray tracing MOC method

International Nuclear Information System (INIS)

Tian Chao; Zheng Youqi; Li Yunzhao; Li Shuo; Chai Xiaoming

2015-01-01

It is difficulty for the MOC method based on Cell Modular Ray Tracing to deal with the irregular geometry such as the water gap between the PWR lattices. Hence, the neutron transport code NECP-Medlar based on Assembly Modular Ray Tracing is developed. CMFD method is used to accelerate the transport calculation. The numerical results of the 2D C5G7 benchmark and typical PWR lattice prove that NECP-Medlar has an excellent performance in terms of accuracy and efficiency. Besides, NECP-Medlar can describe clearly the flux distribution of the lattice with water gap. (authors)

Neutronics methods for thermal radiative transfer

International Nuclear Information System (INIS)

Larsen, E.W.

1988-01-01

The equations of thermal radiative transfer are time discretized in a semi-implicit manner, yielding a linear transport problem for each time step. The governing equation in this problem has the form of a neutron transport equation with fission but no scattering. Numerical methods are described, whose origins lie in neutron transport, and that have been successfully adapted to this new problem. Acceleration methods that have been developed specifically for the radiative transfer problem, but may have generalizations applicable in neutronics problems, are also discussed

Nodal methods for problems in fluid mechanics and neutron transport

International Nuclear Information System (INIS)

Azmy, Y.Y.

1985-01-01

A new high-accuracy, coarse-mesh, nodal integral approach is developed for the efficient numerical solution of linear partial differential equations. It is shown that various special cases of this general nodal integral approach correspond to several high efficiency nodal methods developed recently for the numerical solution of neutron diffusion and neutron transport problems. The new approach is extended to the nonlinear Navier-Stokes equations of fluid mechanics; its extension to these equations leads to a new computational method, the nodal integral method which is implemented for the numerical solution of these equations. Application to several test problems demonstrates the superior computational efficiency of this new method over previously developed methods. The solutions obtained for several driven cavity problems are compared with the available experimental data and are shown to be in very good agreement with experiment. Additional comparisons also show that the coarse-mesh, nodal integral method results agree very well with the results of definitive ultra-fine-mesh, finite-difference calculations for the driven cavity problem up to fairly high Reynolds numbers

Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

International Nuclear Information System (INIS)

Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T.H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew

2016-01-01

A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

Energy Technology Data Exchange (ETDEWEB)

Collins, Benjamin, E-mail: collinsbs@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Stimpson, Shane, E-mail: stimpsonsg@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Kelley, Blake W., E-mail: kelleybl@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Young, Mitchell T.H., E-mail: youngmit@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Kochunas, Brendan, E-mail: bkochuna@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Graham, Aaron, E-mail: aarograh@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Larsen, Edward W., E-mail: edlarsen@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Downar, Thomas, E-mail: downar@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Godfrey, Andrew, E-mail: godfreyat@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

2016-12-01

A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

Solution and study of nodal neutron transport equation applying the LTS{sub N}-DiagExp method

Energy Technology Data Exchange (ETDEWEB)

Hauser, Eliete Biasotto; Pazos, Ruben Panta [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Faculdade de Matematica]. E-mail: eliete@pucrs.br; rpp@mat.pucrs.br; Vilhena, Marco Tullio de [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Instituto de Matematica]. E-mail: vilhena@mat.ufrgs.br; Barros, Ricardo Carvalho de [Universidade do Estado, Nova Friburgo, RJ (Brazil). Instituto Politecnico]. E-mail: ricardo@iprj.uerj.br

2003-07-01

In this paper we report advances about the three-dimensional nodal discrete-ordinates approximations of neutron transport equation for Cartesian geometry. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S{sub N} equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS{sub N} method, first applying the Laplace transform to the set of the nodal S{sub N} equations and then obtained the solution by symbolic computation. We include the LTS{sub N} method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS{sub N} approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. (author)

Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry

International Nuclear Information System (INIS)

Valdes Parra, J.J.

1986-01-01

One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as 'acceleration of synthetic diffusion' which has been applied to solve the neutron transport equation with 'classical schemes of spatial integration' obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author)

Analytical synthetic methods of solution of neutron transport equation with diffusion theory approaches energy multigroup

International Nuclear Information System (INIS)

Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C.

2013-01-01

In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation

Analysis of the neutrons dispersion in a semi-infinite medium based in transport theory and the Monte Carlo method

International Nuclear Information System (INIS)

Arreola V, G.; Vazquez R, R.; Guzman A, J. R.

2012-10-01

In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., μο=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)

Application of the finite element method to the neutron transport equation

International Nuclear Information System (INIS)

Martin, W.R.

1976-01-01

This paper examines the theoretical and practical application of the finite element method to the neutron transport equation. It is shown that in principle the system of equations obtained by application of the finite element method can be solved with certain physical restrictions concerning the criticality of the medium. The convergence of this approximate solution to the exact solution with mesh refinement is examined, and a non-optical estimate of the convergence rate is obtained analytically. It is noted that the numerical results indicate a faster convergence rate and several approaches to obtain this result analytically are outlined. The practical application of the finite element method involved the development of a computer code capable of solving the neutron transport equation in 1-D plane geometry. Vacuum, reflecting, or specified incoming boundary conditions may be analyzed, and all are treated as natural boundary conditions. The time-dependent transport equation is also examined and it is shown that the application of the finite element method in conjunction with the Crank-Nicholson time discretization method results in a system of algebraic equations which is readily solved. Numerical results are given for several critical slab eigenvalue problems, including anisotropic scattering, and the results compare extremely well with benchmark results. It is seen that the finite element code is more efficient than a standard discrete ordinates code for certain problems. A problem with severe heterogeneities is considered and it is shown that the use of discontinuous spatial and angular elements results in a marked improvement in the results. Finally, time-dependent problems are examined and it is seen that the phenomenon of angular mode separation makes the numerical treatment of the transport equation in slab geometry a considerable challenge, with the result that the angular mesh has a dominant effect on obtaining acceptable solutions

Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods

International Nuclear Information System (INIS)

Lefvert, T.

1975-11-01

Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)

Solving the equation of neutron transport

International Nuclear Information System (INIS)

Nasfi, Rim

2009-01-01

This work is devoted to the study of some numerical methods of resolution of the problem of transport of the neutrons. We started by introducing the equation integro-differential transport of the neutrons. Then we applied the finite element method traditional for stationary and nonstationary linear problems in 2D. A great part is reserved for the presentation of the mixed numerical diagram and mixed hybrid with two types of uniform grids: triangular and rectangular. Thereafter we treated some numerical examples by implementations in Matlab in order to test the convergence of each method. To finish, we had results of simulation by the Monte Carlo method on a problem of two-dimensional transport with an aim of comparing them with the results resulting from the finite element method mixed hybrids. Some remarks and prospects conclude this work.

Reliability analysis of neutron transport simulation using Monte Carlo method

International Nuclear Information System (INIS)

Souza, Bismarck A. de; Borges, Jose C.

1995-01-01

This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs

Direct Discrete Method for Neutronic Calculations

International Nuclear Information System (INIS)

Vosoughi, Naser; Akbar Salehi, Ali; Shahriari, Majid

2002-01-01

The objective of this paper is to introduce a new direct method for neutronic calculations. This method which is named Direct Discrete Method, is simpler than the neutron Transport equation and also more compatible with physical meaning of problems. This method is based on physic of problem and with meshing of the desired geometry, writing the balance equation for each mesh intervals and with notice to the conjunction between these mesh intervals, produce the final discrete equations series without production of neutron transport differential equation and mandatory passing from differential equation bridge. We have produced neutron discrete equations for a cylindrical shape with two boundary conditions in one group energy. The correction of the results from this method are tested with MCNP-4B code execution. (authors)

Presentation of some methods for the solution of the monoenergetic neutrons transport equation

International Nuclear Information System (INIS)

Valle G, E. del.

1978-01-01

The neutrons transport theory problems whose solution has been reached were collected in order to show that the transport equation is so complicated that different techniques were developed so as to give approximative numerical solutions to problems concerning the practical application. Such a technique, which had not been investigated in the literature dealing with these problems, is described here. The results which were obtained through this technique in undimensional problems of criticity are satisfactory and speaking in a conceptual way this method is extremely simple because it times. There is no limitation to deal with problems related neutrons sources with an arbitrary distribution and in principle the application of this technique can be extended to unhomogeneous environments. (author)

Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

CERN Document Server

2002-01-01

This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

Synergism of the method of characteristics and CAD technology for neutron transport calculation

International Nuclear Information System (INIS)

Chen, Z.; Wang, D.; He, T.; Wang, G.; Zheng, H.

2013-01-01

The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, a new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)

The study of neutron transport by oscillation method; Etude du transport des neutrons par la methode de modulation

Energy Technology Data Exchange (ETDEWEB)

Raievski, V

1959-07-01

The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [French] La methode de modulation est un procede tres general d'etude des proprietes neutroniques des milieux contenant des neutrons thermiques. Le present rapport a pour but de decrire les principales de ces experiences et d'en donner une theorie generale. Cette theorie, exposee dans la premiere partie, est etablie dons le cadre de l'approximation a deux groupes de vitesse qui a prouve son efficacite dons le cas des piles a neutrons thermiques. Le domaine de validite de l'approximation a deux groupes est rappele au debut, ce qui permet de definir avec precision la signification des parametres qui entrent dons la theorie et qui font l'objet de ces mesures. La deuxieme partie decrit les experiences realisees, en particulier celles effectuees au Centre d'Etudes Nucleaires de Saclay ou la methode a ete considerablement developpee. Ces experiences sont interpretees dans le cadre de la theorie generale exposee precedemment. On prouve ainsi l'identite des parametres mesures par cette methode et de ceux figurant

A method for solving the spherical harmonics equations applied for space-energy transport of fast and resonance neutrons

International Nuclear Information System (INIS)

Matausek, M.

1972-01-01

A new proposed method for solving the space-energy dependent spherical harmonics equations represents a methodological contribution to neutron transport theory. The proposed method was applied for solving the problem of spec-energy transport of fast and resonance neutrons in multi-zone, cylindrical y symmetric infinite reactor cell and is related to previously developed procedure for treating the thermal energy region. The advantages of this method are as follows: a unique algorithm was obtained for detailed determination of spatial and energy distribution of neutrons (from thermal to fast) in the reactor cell; these detailed distributions enable more precise calculations of criticality conditions, obtaining adequate multigroup data and better interpretation of experimental data; computing time is rather short

A coordinate transform method for one-speed neutron transport in composite slabs

International Nuclear Information System (INIS)

Haidar, N.H.S.

1988-01-01

The optical path transformation is applied to reduce the one-speed neutron transport equation for a class of composite subcritical slabs to single-region problems. The class idealises, within the uncertainty of the one-speed model, a variety of practical situations such as U-D 2 O-C-Zr-Pb or Pu-U-Na-Fe symmetric reactor assemblies; which may possibly contain a symmetrically anisotropic neutron source. A closed form double series solution, which turns out to be quite convenient for design and optimisation purposes, has been obtained, in terms of discontinuous functions for the multi-regional angular flux by application of a double finite Legendre transform. Disadvantage factor evaluations for a U-C lattice cell resulting from a low-order P 0 P 1 approximation of this method are found to be in full agreement with hybrid diffusion-transport estimates. (author)

Neutron stochastic transport theory with delayed neutrons

International Nuclear Information System (INIS)

Munoz-Cobo, J.L.; Verdu, G.

1987-01-01

From the stochastic transport theory with delayed neutrons, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation. From this theory a general expression is obtained for the Feynman Y-function when delayed neutrons are included. The single mode approximation for the particular case of a subcritical assembly is developed, and it is shown that Y-function reduces to the familiar expression quoted in many books, when delayed neutrons are not considered, and spatial and source effects are not included. (author)

Neutron transport and Montecarlo method: analysis and revision

International Nuclear Information System (INIS)

Perlado, J.M.

1982-01-01

The resolution of the neutron transport equation by the Montecarlo method is presented. Coming from an extensive discussion on the best formulation of that equation in order to be treated through the mentioned method, the theoretical bases of the estimator and random-walk generation is extensively explained. The most general expression for the estimators in different physical situations, each with a diverse random-walk, is included in this basical theoretical part. Furthemore, a large revision on the variance reduction methods is made. Its theoretical presentation is claimed to be in connection with the need for each one of them. The use of the adjoint equation, as a part of the importance sampling, Russian Roulette, splitting, exponential transform, conditional and correlated Montecarlo, and one-collision and next-event extimators, are discussed. Finally, come comments in the presentation of the last works on the theoretical prediction of errors in the generation of estimators-random walks are made. (author)

The discrete cones method for two-dimensional neutron transport calculations

International Nuclear Information System (INIS)

Watanabe, Y.; Maynard, C.W.

1986-01-01

A novel method, the discrete cones method (DC/sub N/), is proposed as an alternative to the discrete ordinates method (S/sub N/) for solutions of the two-dimensional neutron transport equation. The new method utilizes a new concept, discrete cones, which are made by partitioning a unit spherical surface that the direction vector of particles covers. In this method particles in a cone are simultaneously traced instead of those in discrete directions so that an anomaly of the S/sub N/ method, the ray effects, can be eliminated. The DC/sub N/ method has been formulated for X-Y geometry and a program has been creaed by modifying the standard S/sub N/ program TWOTRAN-II. Our sample calculations demonstrate a strong mitigation of the ray effects without a computing cost penalty

Coarse-mesh rebalance methods compatible with the spherical harmonic fictitious source in neutron transport calculations

International Nuclear Information System (INIS)

Miller, W.F. Jr.

1975-10-01

The coarse-mesh rebalance method, based on neutron conservation, is used in discrete ordinates neutron transport codes to accelerate convergence of the within-group scattering source. Though very powerful for this application, the method is ineffective in accelerating the iteration on the discrete-ordinates-to-spherical-harmonics fictitious sources used for ray-effect elimination. This is largely because this source makes a minimum contribution to the neutron balance equation. The traditional rebalance approach is derived in a variational framework and compared with new rebalance approaches tailored to be compatible with the fictitious source. The new approaches are compared numerically to determine their relative advantages. It is concluded that there is little incentive to use the new methods. (3 tables, 5 figures)

Comparison of two Ssub(infinity) methods for solving the neutron transport equation

International Nuclear Information System (INIS)

Mennig, J.; Brandt, D.; Haelg, W.

1978-01-01

A semianalytic method (S 0 sub(infinity)) is presented for solving the monoenergetic multi-region transport equation. This method is compared with results from S 1 sub(infinity)-theory given in the literature. Application of S 1 sub(infinity)-theory to reactor shields may lead to negative neutron fluxes and to flux oscillations. These unphysical effects are completely avoided by the new method. Numerical results demonstrate the limitations of S 1 sub(infinity) and confirm the numerical stability of (S 0 sub(infinity)). (Auth.)

Improved method for solving the neutron transport problem by discretization of space and energy variables

International Nuclear Information System (INIS)

Bosevski, T.

1971-01-01

The polynomial interpolation of neutron flux between the chosen space and energy variables enabled transformation of the integral transport equation into a system of linear equations with constant coefficients. Solutions of this system are the needed values of flux for chosen values of space and energy variables. The proposed improved method for solving the neutron transport problem including the mathematical formalism is simple and efficient since the number of needed input data is decreased both in treating the spatial and energy variables. Mathematical method based on this approach gives more stable solutions with significantly decreased probability of numerical errors. Computer code based on the proposed method was used for calculations of one heavy water and one light water reactor cell, and the results were compared to results of other very precise calculations. The proposed method was better concerning convergence rate, decreased computing time and needed computer memory. Discretization of variables enabled direct comparison of theoretical and experimental results

Multi-group neutron transport theory

International Nuclear Information System (INIS)

Zelazny, R.; Kuszell, A.

1962-01-01

Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr

Numerical method for solving integral equations of neutron transport. II

International Nuclear Information System (INIS)

Loyalka, S.K.; Tsai, R.W.

1975-01-01

In a recent paper it was pointed out that the weakly singular integral equations of neutron transport can be quite conveniently solved by a method based on subtraction of singularity. This previous paper was devoted entirely to the consideration of simple one-dimensional isotropic-scattering and one-group problems. The present paper constitutes interesting extensions of the previous work in that in addition to a typical two-group anisotropic-scattering albedo problem in the slab geometry, the method is also applied to an isotropic-scattering problem in the x-y geometry. These results are compared with discrete S/sub N/ (ANISN or TWOTRAN-II) results, and for the problems considered here, the proposed method is found to be quite effective. Thus, the method appears to hold considerable potential for future applications. (auth)

Numerical solution of the time dependent neutron transport equation by the method of the characteristics

International Nuclear Information System (INIS)

Talamo, Alberto

2013-01-01

This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps

Numerical solution of the time dependent neutron transport equation by the method of the characteristics

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto, E-mail: alby@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)

2013-05-01

This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps.

Particle-transport simulation with the Monte Carlo method

International Nuclear Information System (INIS)

Carter, L.L.; Cashwell, E.D.

1975-01-01

Attention is focused on the application of the Monte Carlo method to particle transport problems, with emphasis on neutron and photon transport. Topics covered include sampling methods, mathematical prescriptions for simulating particle transport, mechanics of simulating particle transport, neutron transport, and photon transport. A literature survey of 204 references is included. (GMT)

Heterogeneity effects in neutron transport computations

International Nuclear Information System (INIS)

Gelbard, E.M.

1975-01-01

A nuclear reactor is, generally, an intricate heterogeneous structure whose adjacent components may differ radically in their neutronic properties. The heterogeneities in the structure of the reactor complicate the work of the reactor analyst and tend to degrade the efficiency of the numerical methods used in reactor computations. Two types of heterogeneity effects are considered. First, certain singularities in the solution of the neutron transport equation, induced by heterogeneities, are briefly described. Second, the effect of heterogeneities on neutron leakage rates, and consequently on effective diffusion coefficients, are discussed. (5 figures) (U.S.)

Albedo's determination by the method of neutron impulse

International Nuclear Information System (INIS)

Flores Calderon, J.E.

1982-01-01

Experiments with non-stationary neutron transport in large cavity moderators (l>>Σsub(tr) -1 ) (where l is the characteristic cavity length and Σsub(tr) -1 the macroscopic transport section of the moderator) led to the method reported in this study which, based on neutron impulses for determining albedo of thermal neutrons, gave a precision greater by an order of magnitude over previous methods. A sufficient time interval after introduction of the neutron flux into the moderator chamber decreased exponentially the decay constant L, which was itself related to albedo by a function called f. Numerical calculations of albedo were assisted. (author)

Asymptotic time dependent neutron transport in multidimensional systems

International Nuclear Information System (INIS)

Nagy, M.E.; Sawan, M.E.; Wassef, W.A.; El-Gueraly, L.A.

1983-01-01

A model which predicts the asymptotic time behavior of the neutron distribution in multi-dimensional systems is presented. The model is based on the kernel factorization method used for stationary neutron transport in a rectangular parallelepiped. The accuracy of diffusion theory in predicting the asymptotic time dependence is assessed. The use of neutron pulse experiments for predicting the diffusion parameters is also investigated

Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

CERN Document Server

Žukauskaitėa, A; Plukienė, R; Ridikas, D

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

Neutron Transport Methods for Accelerator-Driven Systems

International Nuclear Information System (INIS)

Nicholas Tsoulfanidis; Elmer Lewis

2005-01-01

The objective of this project has been to develop computational methods that will enable more effective analysis of Accelerator Driven Systems (ADS). The work is centered at the University of Missouri at Rolla, with a subcontract at Northwestern University, and close cooperation with the Nuclear Engineering Division at Argonne National Laboratory. The work has fallen into three categories. First, the treatment of the source for neutrons originating from the spallation target which drives the neutronics calculations of the ADS. Second, the generalization of the nodal variational method to treat the R-Z geometry configurations frequently needed for scoping calculations in Accelerator Driven Systems. Third, the treatment of void regions within variational nodal methods as needed to treat the accelerator beam tube

Neutron transport simulation in high speed moving media using Geant4

Science.gov (United States)

Li, G.; Ciungu, B.; Harrisson, G.; Rogge, R. B.; Tun, Z.; van der Ende, B. M.; Zwiers, I.

2017-12-01

A method using Geant4 to simulate neutron transport in moving media is described. The method is implanted in the source code of the software since Geant4 does not intrinsically support a moving object. The simulation utilizes the existing physical model and data library in Geant4, combined with frame transformations to account for the effect of relative velocity between neutrons and the moving media. An example is presented involving a high speed rotating cylinder to verify this method and show the effect of moving media on neutron transport.

Analysis of the neutrons dispersion in a semi-infinite medium based in transport theory and the Monte Carlo method; Analisis de la dispersion de neutrones en un medio semi-infinito en base a teoria de transporte y el metodo de Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Arreola V, G. [IPN, Escuela Superior de Fisica y Matematicas, Posgrado en Ciencias Fisicomatematicas, area en Ingenieria Nuclear, Unidad Profesional Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07730 Mexico D. F. (Mexico); Vazquez R, R.; Guzman A, J. R., E-mail: energia.arreola.uam@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

2012-10-15

In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., {mu}{omicron}=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)

Uncertainty analysis of neutron transport calculation

International Nuclear Information System (INIS)

Oka, Y.; Furuta, K.; Kondo, S.

1987-01-01

A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

Surface harmonics method for two-dimensional time-dependent neutron transport problems of square-lattice nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Boyarinov, V. F.; Kondrushin, A. E.; Fomichenko, P. A. [National Research Centre Kurchatov Institute, Kurchatov Sq. 1, Moscow (Russian Federation)

2013-07-01

Time-dependent equations of the Surface Harmonics Method (SHM) have been derived from the time-dependent neutron transport equation with explicit representation of delayed neutrons for solving the two-dimensional time-dependent problems. These equations have been realized in the SUHAM-TD code. The TWIGL benchmark problem has been used for verification of the SUHAM-TD code. The results of the study showed that computational costs required to achieve necessary accuracy of the solution can be an order of magnitude less than with the use of the conventional finite difference method (FDM). (authors)

The study of neutron transport by oscillation method

International Nuclear Information System (INIS)

Raievski, V.

1959-01-01

The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [fr

Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

International Nuclear Information System (INIS)

Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

1989-01-01

Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

Progress in multidimensional neutron transport computation

International Nuclear Information System (INIS)

Lewis, E.E.

1977-01-01

The methods available for solution of the time-independent neutron transport problems arising in the analysis of nuclear systems are examined. The merits of deterministic and Monte Carlo methods are briefly compared. The capabilities of deterministic computational methods derived from the first-order form of the transport equation, from the second-order even-parity form of this equation, and from integral transport formulations are discussed in some detail. Emphasis is placed on the approaches for dealing with the related problems of computer memory requirements, computational cost, and achievable accuracy. Attention is directed to some areas where problems exist currently and where the need for further work appears to be particularly warranted

Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

International Nuclear Information System (INIS)

Chen, G.S.; Yang, D.Y.

1998-01-01

We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used: TFQMR (transpose free quasi-minimal residual algorithm) CGS (conjugate gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These subroutines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. The reasons to choose the generalized conjugate gradient methods are that the methods have better residual (equivalent to error) control procedures in the computation and have better convergent rate. The pointwise incomplete LU factorization ILU, modified pointwise incomplete LU factorization MILU, block incomplete factorization BILU and modified blockwise incomplete LU factorization MBILU are the preconditioning techniques used in the several testing problems. In Bi-CGSTAB, CGS, TFQMR and QMRCGSTAB method, we find that either CGS or Bi-CGSTAB method combined with preconditioner MBILU is the most efficient algorithm in these methods in the several testing problems. The numerical solution of flux by preconditioned CGS and Bi-CGSTAB methods has the same result as those from Cray computer, obtained by either the point successive relaxation method or the line successive relaxation method combined with Gaussian elimination

The isotope density inverse problem in multigroup neutron transport

International Nuclear Information System (INIS)

Zazula, J.M.

1981-01-01

The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms of the flux in a group - density of an element spatial correlation, leads to a set of integral functionals for the densities weighted by the group fluxes. Some methods of approximation to make the problem uniquently solvable are proposed. Particularly P 0 angular flux information and the spherically-symetrical geometry of an infinite medium are considered. The numerical calculation using this method related to sooner evaluated direct problem data gives promising agreement with primary densities. This approach would be the basis for further application in an elemental analysis of a medium, using an isotopic neutron source and a moving, energy-dependent neutron detector. (author)

Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

CERN Document Server

Žukauskaite, A; Plukiene, R; Plukis, A

2007-01-01

Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

A Monte-Carlo method for ex-core neutron response

International Nuclear Information System (INIS)

Gamino, R.G.; Ward, J.T.; Hughes, J.C.

1997-10-01

A Monte Carlo neutron transport kernel capability primarily for ex-core neutron response is described. The capability consists of the generation of a set of response kernels, which represent the neutron transport from the core to a specific ex-core volume. This is accomplished by tagging individual neutron histories from their initial source sites and tracking them throughout the problem geometry, tallying those that interact in the geometric regions of interest. These transport kernels can subsequently be combined with any number of core power distributions to determine detector response for a variety of reactor Thus, the transport kernels are analogous to an integrated adjoint response. Examples of pressure vessel response and ex-core neutron detector response are provided to illustrate the method

Fast rigorous numerical method for the solution of the anisotropic neutron transport problem and the NITRAN system for fusion neutronics application. Pt. 2

International Nuclear Information System (INIS)

Takahashi, A.; Rusch, D.

1979-10-01

The I*-method, which is a non-approximative treatment of the neutron balance equations by the use of double-differential cross sections and a generalized angular transfer probability, is realized within the NITRAN system. It is shown, by means of test calculations for assemblies related to fusion reactor neutronics that double-differential cross section data provide substantial progress in transport problems with kinematically complicated reaction channels like (n,2n), (n,n'γ), and (n,n'α), because the I*-method is free from kinematic assumptions. The properties of the exponential method to generate the supplementary equations to the SN equations are investigated. (orig.) [de

Transportable, Low-Dose Active Fast-Neutron Imaging

Energy Technology Data Exchange (ETDEWEB)

Mihalczo, John T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wright, Michael C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McConchie, Seth M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Palles, Blake A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.

Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P{sub 3} nodal methods

Energy Technology Data Exchange (ETDEWEB)

Duerigen, Susan

2013-05-15

The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.

A modular spherical harmonics approach to the neutron transport equation

International Nuclear Information System (INIS)

Inanc, F.; Rohach, A.F.

1989-01-01

A modular nodal method was developed for solving the neutron transport equation in 2-D xy coordinates. The spherical harmonic expansion was used for approximating the second-order even-parity form of the neutron transport equation. The boundary conditions of the spherical harmonics approximation were derived in a form to have forms analogous to the partial currents in the neutron diffusion equation. Relations were developed for generating both the second-order spherical harmonic equations and the boundary conditions in an automated computational algorithm. Nodes using different orders of the spherical harmonics approximation to the transport equation were interfaced through mixed-type boundary conditions. The determination of spherical harmonic orders implemented in the nodes were determined by the scheme in an automated manner. Results of the method compared favorably to benchmark problems. (author)

Solution of two-dimensional equations of neutron transport in 4P0-approximation of spherical harmonics method

International Nuclear Information System (INIS)

Polivanskij, V.P.

1989-01-01

The method to solve two-dimensional equations of neutron transport using 4P 0 -approximation is presented. Previously such approach was efficiently used for the solution of one-dimensional problems. New an attempt is made to apply the approach to solution of two-dimensional problems. Algorithm of the solution is given, as well as results of test neutron-physical calculations. A considerable as compared with diffusion approximation is shown. 11 refs

Generalized Coarse-Mesh Rebalance Method for Acceleration of Neutron Transport Calculations

International Nuclear Information System (INIS)

Yamamoto, Akio

2005-01-01

This paper proposes a new acceleration method for neutron transport calculations: the generalized coarse-mesh rebalance (GCMR) method. The GCMR method is a unified scheme of the traditional coarse-mesh rebalance (CMR) and the coarse-mesh finite difference (CMFD) acceleration methods. Namely, by using an appropriate acceleration factor, formulation of the GCMR method becomes identical to that of the CMR or CMFD method. This also indicates that the convergence property of the GCMR method can be controlled by the acceleration factor since the convergence properties of the CMR and CMFD methods are generally different. In order to evaluate the convergence property of the GCMR method, a linearized Fourier analysis was carried out for a one-group homogeneous medium, and the results clarified the relationship between the acceleration factor and the spectral radius. It was also shown that the spectral radius of the GCMR method is smaller than those of the CMR and CMFD methods. Furthermore, the Fourier analysis showed that when an appropriate acceleration factor was used, the spectral radius of the GCMR method did not exceed unity in this study, which was in contrast to the results of the CMR or the CMFD method. Application of the GCMR method to practical calculations will be easy when the CMFD acceleration is already adopted in a transport code. By multiplying a suitable acceleration factor to a coefficient (D FD ) of a finite difference formulation, one can improve the numerical instability of the CMFD acceleration method

Concise four-vector scheme for neutron transport calculations

International Nuclear Information System (INIS)

Kulacsy, K.; Lukacs, B.; Racz, A.

1995-01-01

An explicit Riemannian geometrical form or the vectorial Neutron Streaming Term is presented. The method applies the full Riemannian technique of general covariance. There are cases when the symmetry of the neutron flux must be smaller than that of the arrangement. However, in coordinate space there are always solutions of the Neutron Transport Equation as symmetric as the arrangement, if the latter's symmetry is at least an affine collineation of the Euclidian 3-space. (author). 7 refs

A linear multiple balance method for discrete ordinates neutron transport equations

International Nuclear Information System (INIS)

Park, Chang Je; Cho, Nam Zin

2000-01-01

A linear multiple balance method (LMB) is developed to provide more accurate and positive solutions for the discrete ordinates neutron transport equations. In this multiple balance approach, one mesh cell is divided into two subcells with quadratic approximation of angular flux distribution. Four multiple balance equations are used to relate center angular flux with average angular flux by Simpson's rule. From the analysis of spatial truncation error, the accuracy of the linear multiple balance scheme is ο(Δ 4 ) whereas that of diamond differencing is ο(Δ 2 ). To accelerate the linear multiple balance method, we also describe a simplified additive angular dependent rebalance factor scheme which combines a modified boundary projection acceleration scheme and the angular dependent rebalance factor acceleration schme. It is demonstrated, via fourier analysis of a simple model problem as well as numerical calculations, that the additive angular dependent rebalance factor acceleration scheme is unconditionally stable with spectral radius < 0.2069c (c being the scattering ration). The numerical results tested so far on slab-geometry discrete ordinates transport problems show that the solution method of linear multiple balance is effective and sufficiently efficient

Comparison of Monte Carlo method and deterministic method for neutron transport calculation

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki

1987-01-01

The report outlines major features of the Monte Carlo method by citing various applications of the method and techniques used for Monte Carlo codes. Major areas of its application include analysis of measurements on fast critical assemblies, nuclear fusion reactor neutronics analysis, criticality safety analysis, evaluation by VIM code, and calculation for shielding. Major techniques used for Monte Carlo codes include the random walk method, geometric expression method (combinatorial geometry, 1, 2, 4-th degree surface and lattice geometry), nuclear data expression, evaluation method (track length, collision, analog (absorption), surface crossing, point), and dispersion reduction (Russian roulette, splitting, exponential transform, importance sampling, corrected sampling). Major features of the Monte Carlo method are as follows: 1) neutron source distribution and systems of complex geometry can be simulated accurately, 2) physical quantities such as neutron flux in a place, on a surface or at a point can be evaluated, and 3) calculation requires less time. (Nogami, K.)

Response matrix method for neutron transport in reactor lattices using group symmetry properties

International Nuclear Information System (INIS)

Mund, E.H.

1991-01-01

This paper describes a response matrix method for the approximate solution of one-velocity, multi-dimensional transport problems in reactor lattices, with isotropic neutron scattering. The transport equation is solved on a homogeneous cell by using a Petrov-Galerkin technique based on a set of trial and test functions (including polynomials and exponential functions) closely related to transport problems in infinite media. The number of non-zero elements of the response matrices reduces to a minimum when the symmetry properties of the cell are included ab initio in the span of the basis functions. To include these properties, use is made of projection operations which are performed very efficiently on symbolic manipulation programs. Numerical results of model problems in square geometry show a good agreement with reference solutions

Development of a three-dimensional neutron transport code DFEM based on the double finite element method

International Nuclear Information System (INIS)

Fujimura, Toichiro

1996-01-01

A three-dimensional neutron transport code DFEM has been developed by the double finite element method to analyze reactor cores with complex geometry as large fast reactors. Solution algorithm is based on the double finite element method in which the space and angle finite elements are employed. A reactor core system can be divided into some triangular and/or quadrangular prism elements, and the spatial distribution of neutron flux in each element is approximated with linear basis functions. As for the angular variables, various basis functions are applied, and their characteristics were clarified by comparison. In order to enhance the accuracy, a general method is derived to remedy the truncation errors at reflective boundaries, which are inherent in the conventional FEM. An adaptive acceleration method and the source extrapolation method were applied to accelerate the convergence of the iterations. The code structure is outlined and explanations are given on how to prepare input data. A sample input list is shown for reference. The eigenvalue and flux distribution for real scale fast reactors and the NEA benchmark problems were presented and discussed in comparison with the results of other transport codes. (author)

Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

Energy Technology Data Exchange (ETDEWEB)

He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)

2015-12-15

Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

A massively parallel discrete ordinates response matrix method for neutron transport

International Nuclear Information System (INIS)

Hanebutte, U.R.; Lewis, E.E.

1992-01-01

In this paper a discrete ordinates response matrix method is formulated with anisotropic scattering for the solution of neutron transport problems on massively parallel computers. The response matrix formulation eliminates iteration on the scattering source. The nodal matrices that result from the diamond-differenced equations are utilized in a factored form that minimizes memory requirements and significantly reduces the number of arithmetic operations required per node. The red-black solution algorithm utilizes massive parallelism by assigning each spatial node to one or more processors. The algorithm is accelerated by a synthetic method in which the low-order diffusion equations are also solved by massively parallel red-black iterations. The method is implemented on a 16K Connection Machine-2, and S 8 and S 16 solutions are obtained for fixed-source benchmark problems in x-y geometry

Tracking techniques for the characteristics method applied to the resolution of the neutrons transport equation in multi scale domains; Techniques de tracage pour la methode des caracteristiques appliquee a la resolution de l'equation du transport des neutrons en domaines multi-dimensionnels

Energy Technology Data Exchange (ETDEWEB)

Fevotte, F. [CEA Saclay, Dept. Modelisation de Systemes et Structures (DEN/DANS/DM2S/SERMA), 91 - Gif sur Yvette (France)

2008-07-01

At the various stages of a nuclear reactor's life, numerous studies are needed to guaranty the safety and efficiency of the design, analyse the fuel cycle, prepare the dismantlement, and so on. Due to the extreme difficulty to take extensive and accurate measurements in the reactor core, most of these studies are numerical simulations. The complete numerical simulation of a nuclear reactor involves many types of physics: neutronics, thermal hydraulics, materials, control engineering, Among these, the neutron transport simulation is one of the fundamental steps, since it allows computation - among other things - of various fundamental values such as the power density (used in thermal hydraulics computations) or fuel burn-up. The neutron transport simulation is based on the Boltzmann equation, which models the neutron population inside the reactor core. Among the various methods allowing its numerical solution, much interest has been devoted in the past few years to the Method of Characteristics in unstructured meshes (MOC), since it offers a good accuracy and operates in complicated geometries. The aim of this work is to propose improvements of the calculation scheme bound on the two dimensions MOC, in order to decrease the needed resources number. (A.L.B.)

Multi-domain/multi-method numerical approach for neutron transport equation; Couplage de methodes et decomposition de domaine pour la resolution de l'equation du transport des neutrons

Energy Technology Data Exchange (ETDEWEB)

Girardi, E

2004-12-15

A new methodology for the solution of the neutron transport equation, based on domain decomposition has been developed. This approach allows us to employ different numerical methods together for a whole core calculation: a variational nodal method, a discrete ordinate nodal method and a method of characteristics. These new developments authorize the use of independent spatial and angular expansion, non-conformal Cartesian and unstructured meshes for each sub-domain, introducing a flexibility of modeling which is not allowed in today available codes. The effectiveness of our multi-domain/multi-method approach has been tested on several configurations. Among them, one particular application: the benchmark model of the Phebus experimental facility at Cea-Cadarache, shows why this new methodology is relevant to problems with strong local heterogeneities. This comparison has showed that the decomposition method brings more accuracy all along with an important reduction of the computer time.

Some improvements in the discrete ordinate method of B.G. Carlson for solving the neutron transport equation

Energy Technology Data Exchange (ETDEWEB)

Askew, J R; Brissenden, R J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1963-08-15

This report gives an account of the DSN method for simulating neutron transport, together with methods of solution developed to deal with problems in the physics of thermal reactors, for which previously available computer programmes were unsatisfactory. The methods described are those incorporated in the programmes WINFRITH DSN written in FORTRAN language for the IBM 7090 and STRETCH computers. (author)

Radiation transport calculation methods in BNCT

International Nuclear Information System (INIS)

Koivunoro, H.; Seppaelae, T.; Savolainen, S.

2000-01-01

Boron neutron capture therapy (BNCT) is used as a radiotherapy for malignant brain tumours. Radiation dose distribution is necessary to determine individually for each patient. Radiation transport and dose distribution calculations in BNCT are more complicated than in conventional radiotherapy. Total dose in BNCT consists of several different dose components. The most important dose component for tumour control is therapeutic boron dose D B . The other dose components are gamma dose D g , incident fast neutron dose D f ast n and nitrogen dose D N . Total dose is a weighted sum of the dose components. Calculation of neutron and photon flux is a complex problem and requires numerical methods, i.e. deterministic or stochastic simulation methods. Deterministic methods are based on the numerical solution of Boltzmann transport equation. Such are discrete ordinates (SN) and spherical harmonics (PN) methods. The stochastic simulation method for calculation of radiation transport is known as Monte Carlo method. In the deterministic methods the spatial geometry is partitioned into mesh elements. In SN method angular integrals of the transport equation are replaced with weighted sums over a set of discrete angular directions. Flux is calculated iteratively for all these mesh elements and for each discrete direction. Discrete ordinates transport codes used in the dosimetric calculations are ANISN, DORT and TORT. In PN method a Legendre expansion for angular flux is used instead of discrete direction fluxes, land the angular dependency comes a property of vector function space itself. Thus, only spatial iterations are required for resulting equations. A novel radiation transport code based on PN method and tree-multigrid technique (TMG) has been developed at VTT (Technical Research Centre of Finland). Monte Carlo method solves the radiation transport by randomly selecting neutrons and photons from a prespecified boundary source and following the histories of selected particles

Toward whole-core neutron transport without spatial homogenization

International Nuclear Information System (INIS)

Lewis, E. E.

2009-01-01

Full text of publication follows: A long-term goal of computational reactor physics is the deterministic analysis of power reactor core neutronics without incurring significant discretization errors in the energy, spatial or angular variables. In principle, given large enough parallel configurations with unlimited CPU time and memory, this goal could be achieved using existing three-dimensional neutron transport codes. In practice, however, solving the Boltzmann equation for neutrons over the six-dimensional phase space is made intractable by the nature of neutron cross-sections and the complexity and size of power reactor cores. Tens of thousands of energy groups would be required for faithful cross section representation. Likewise, the numerous material interfaces present in power reactor lattices require exceedingly fine spatial mesh structures; these ubiquitous interfaces preclude effective implementation of adaptive grid, mesh-less methods and related techniques that have been applied so successfully in other areas of engineering science. These challenges notwithstanding, substantial progress continues in the pursuit for more robust deterministic methods for whole-core neutronics analysis. This paper examines the progress over roughly the last decade, emphasizing the space-angle variables and the quest to eliminate errors attributable to spatial homogenization. As prolog we briefly assess 1990's methods used in light water reactor analysis and review the lessons learned from the C5G7 benchmark exercises which were originated in 1999 to appraise the ability of transport codes to perform core calculations without homogenization. We proceed by examining progress over the last decade much of which falls into three areas. These may be broadly characterized as reduced homogenization, dynamic homogenization and planar-axial synthesis. In the first, homogenization in three-dimensional calculations is reduced from the fuel assembly to the pin-cell level. In the second

Tracking techniques for the characteristics method applied to the resolution of the neutrons transport equation in multi scale domains

International Nuclear Information System (INIS)

Fevotte, F.

2008-01-01

At the various stages of a nuclear reactor's life, numerous studies are needed to guaranty the safety and efficiency of the design, analyse the fuel cycle, prepare the dismantlement, and so on. Due to the extreme difficulty to take extensive and accurate measurements in the reactor core, most of these studies are numerical simulations. The complete numerical simulation of a nuclear reactor involves many types of physics: neutronics, thermal hydraulics, materials, control engineering, Among these, the neutron transport simulation is one of the fundamental steps, since it allows computation - among other things - of various fundamental values such as the power density (used in thermal hydraulics computations) or fuel burn-up. The neutron transport simulation is based on the Boltzmann equation, which models the neutron population inside the reactor core. Among the various methods allowing its numerical solution, much interest has been devoted in the past few years to the Method of Characteristics in unstructured meshes (MOC), since it offers a good accuracy and operates in complicated geometries. The aim of this work is to propose improvements of the calculation scheme bound on the two dimensions MOC, in order to decrease the needed resources number. (A.L.B.)

Cellular neural network to the spherical harmonics approximation of neutron transport equation in x–y geometry

International Nuclear Information System (INIS)

Pirouzmand, Ahmad; Hadad, Kamal

2012-01-01

Highlights: ► This paper describes the solution of time-dependent neutron transport equation. ► We use a novel method based on cellular neural networks (CNNs) coupled with the spherical harmonics method. ► We apply the CNN model to simulate step and ramp perturbation transients in a core. ► The accuracy and capabilities of the CNN model are examined for x–y geometry. - Abstract: In an earlier paper we utilized a novel method using cellular neural networks (CNNs) coupled with spherical harmonics method to solve the steady state neutron transport equation in x–y geometry. Here, the previous work is extended to the study of time-dependent neutron transport equation. To achieve this goal, an equivalent electrical circuit based on a second-order form of time-dependent neutron transport equation and one equivalent group of neutron precursor density is obtained by the CNN method. The CNN model is used to simulate step and ramp perturbation transients in a typical 2D core.

Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCAD

International Nuclear Information System (INIS)

Chen Qichang; Wu Hongchun; Cao Liangzhi

2008-01-01

A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively

Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCAD

Energy Technology Data Exchange (ETDEWEB)

Chen Qichang; Wu Hongchun [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China); Cao Liangzhi [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China)], E-mail: caolz@mail.xjtu.edu.cn

2008-10-15

A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively.

Comparison of neutron transport calculations with NRC test results

International Nuclear Information System (INIS)

Koban, J.; Hofmann, W.

1981-02-01

For an exactly defined reactor arrangement (PCA = Pool Critical Assembly) neutron fluxes, neutron spectra and reaction rates for several neutron detectors were calculated by means of one and two dimensional transport codes. An international comparison proved the methods applied at KWU to be adequate. There were difficulties, however, in considering the three dimensions of the assembly which result mainly from its small dimension. This fact applies to all participants who didn't use three dimensional codes. (orig.) [de

A prestorage method to measure neutron transmission of ultracold neutron guides

International Nuclear Information System (INIS)

Blau, B.; Daum, M.; Fertl, M.; Geltenbort, P.; Göltl, L.; Henneck, R.; Kirch, K.; Knecht, A.; Lauss, B.; Schmidt-Wellenburg, P.; Zsigmond, G.

2016-01-01

There are worldwide efforts to search for physics beyond the Standard Model of particle physics. Precision experiments using ultracold neutrons (UCN) require very high intensities of UCN. Efficient transport of UCN from the production volume to the experiment is therefore of great importance. We have developed a method using prestored UCN in order to quantify UCN transmission in tubular guides. This method simulates the final installation at the Paul Scherrer Institute's UCN source where neutrons are stored in an intermediate storage vessel serving three experimental ports. This method allowed us to qualify UCN guides for their intended use and compare their properties.

An outline review of numerical transport methods

International Nuclear Information System (INIS)

Budd, C.

1981-01-01

A brief review is presented of numerical methods for solving the neutron transport equation in the context of reactor physics. First the various forms of transport equation are given. Second, the various ways of classifying numerical transport methods are discussed. Finally each method (or class of methods) is outlined in turn. (U.K.)

Magnetic field devices for neutron spin transport and manipulation in precise neutron spin rotation measurements

Energy Technology Data Exchange (ETDEWEB)

Maldonado-Velázquez, M. [Posgrado en Ciencias Físicas, Universidad Nacional Autónoma de México, 04510 (Mexico); Barrón-Palos, L., E-mail: libertad@fisica.unam.mx [Instituto de Física, Universidad Nacional Autónoma de México, Apartado Postal 20-364, 01000 (Mexico); Crawford, C. [University of Kentucky, Lexington, KY 40506 (United States); Snow, W.M. [Indiana University, Bloomington, IN 47405 (United States)

2017-05-11

The neutron spin is a critical degree of freedom for many precision measurements using low-energy neutrons. Fundamental symmetries and interactions can be studied using polarized neutrons. Parity-violation (PV) in the hadronic weak interaction and the search for exotic forces that depend on the relative spin and velocity, are two questions of fundamental physics that can be studied via the neutron spin rotations that arise from the interaction of polarized cold neutrons and unpolarized matter. The Neutron Spin Rotation (NSR) collaboration developed a neutron polarimeter, capable of determining neutron spin rotations of the order of 10{sup −7} rad per meter of traversed material. This paper describes two key components of the NSR apparatus, responsible for the transport and manipulation of the spin of the neutrons before and after the target region, which is surrounded by magnetic shielding and where residual magnetic fields need to be below 100 μG. These magnetic field devices, called input and output coils, provide the magnetic field for adiabatic transport of the neutron spin in the regions outside the magnetic shielding while producing a sharp nonadiabatic transition of the neutron spin when entering/exiting the low-magnetic-field region. In addition, the coils are self contained, forcing the return magnetic flux into a compact region of space to minimize fringe fields outside. The design of the input and output coils is based on the magnetic scalar potential method.

A three-dimensional neutron transport benchmark solution

International Nuclear Information System (INIS)

Ganapol, B.D.; Kornreich, D.E.

1993-01-01

For one-group neutron transport theory in one dimension, several powerful analytical techniques have been developed to solve the neutron transport equation, including Caseology, Wiener-Hopf factorization, and Fourier and Laplace transform methods. In addition, after a Fourier transform in the transverse plane and formulation of a pseudo problem, two-dimensional (2-D) and three-dimensional (3-D) problems can be solved using the techniques specifically developed for the one-dimensional (1-D) case. Numerical evaluation of the resulting expressions requiring an inversion in the transverse plane have been successful for 2-D problems but becomes exceedingly difficult in the 3-D case. In this paper, we show that by using the symmetry along the beam direction, a 2-D problem can be transformed into a 3-D problem in an infinite medium. The numerical solution to the 3-D problem is then demonstrated. Thus, a true 3-D transport benchmark solution can be obtained from a well-established numerical solution to a 2-D problem

Exact solution of the neutron transport equation in spherical geometry

Energy Technology Data Exchange (ETDEWEB)

Anli, Fikret; Akkurt, Abdullah; Yildirim, Hueseyin; Ates, Kemal [Kahramanmaras Suetcue Imam Univ. (Turkey). Faculty of Sciences and Letters

2017-03-15

Solution of the neutron transport equation in one dimensional slab geometry construct a basis for the solution of neutron transport equation in a curvilinear geometry. Therefore, in this work, we attempt to derive an exact analytical benchmark solution for both neutron transport equations in slab and spherical medium by using P{sub N} approximation which is widely used in neutron transport theory.

Transmission probability method for solving neutron transport equation in three-dimensional triangular-z geometry

Energy Technology Data Exchange (ETDEWEB)

Liu Guoming [Department of Nuclear Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi 710049 (China)], E-mail: gmliusy@gmail.com; Wu Hongchun; Cao Liangzhi [Department of Nuclear Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi 710049 (China)

2008-09-15

This paper presents a transmission probability method (TPM) to solve the neutron transport equation in three-dimensional triangular-z geometry. The source within the mesh is assumed to be spatially uniform and isotropic. At the mesh surface, the constant and the simplified P{sub 1} approximation are invoked for the anisotropic angular flux distribution. Based on this model, a code TPMTDT is encoded. It was verified by three 3D Takeda benchmark problems, in which the first two problems are in XYZ geometry and the last one is in hexagonal-z geometry, and an unstructured geometry problem. The results of the present method agree well with those of Monte-Carlo calculation method and Spherical Harmonics (P{sub N}) method.

Effect of fractional parameter on neutron transport in finite disturbed reactors with quadratic scattering

International Nuclear Information System (INIS)

Sallah, M.; Margeanu, C. A.

2016-01-01

The space-fractional neutron transport equation is used to describe the neutrons transport in finite disturbed reactors. It is approximated using the Pomraning-Eddington technique to yield two space-fractional differential equations, in terms of neutron density and net neutron flux. These resultant equations are coupled into a fractional diffusion-like equation for the neutron density whose solution is obtained by using Laplace transformation method. The solution is represented in terms of the Mittag-Leffler function and its different orders. The scattering is considered as quadratic scattering to offer a more realistic, compact representation of the system, and to increase the accuracy of the estimated neutronic parameters. The results are presented graphically to illustrate the fractional parameter effect in addition to the effect of radiative-transfer properties on the physical parameters of interest (reflection coefficient, transmission coefficient, neutron energy, and net neutron flux). The neutron transport problem in finite disturbed reactor with quadratic scattering is considered in investigating the shielding effectiveness, by using MAVRIC shielding module from SCALE6 programs package. The fractional parameter can be used to adjust the analysed data on neutron energy and flux, both for the theoretical model and the neutron transport application. (authors)

A finite element method for neutron transport

International Nuclear Information System (INIS)

Ackroyd, R.T.

1978-01-01

A variational treatment of the finite element method for neutron transport is given based on a version of the even-parity Boltzmann equation which does not assume that the differential scattering cross-section has a spherical harmonic expansion. The theory of minimum and maximum principles is based on the Cauchy-Schwartz equality and the properties of a leakage operator G and a removal operator C. For systems with extraneous sources, two maximum and one minimum principles are given in boundary free form, to ease finite element computations. The global error of an approximate variational solution is given, the relationship of one the maximum principles to the method of least squares is shown, and the way in which approximate solutions converge locally to the exact solution is established. A method for constructing local error bounds is given, based on the connection between the variational method and the method of the hypercircle. The source iteration technique and a maximum principle for a system with extraneous sources suggests a functional for a variational principle for a self-sustaining system. The principle gives, as a consequence of the properties of G and C, an upper bound to the lowest eigenvalue. A related functional can be used to determine both upper and lower bounds for the lowest eigenvalue from an inspection of any approximate solution for the lowest eigenfunction. The basis for the finite element is presented in a general form so that two modes of exploitation can be undertaken readily. The model can be in phase space, with positional and directional co-ordinates defining points of the model, or it can be restricted to the positional co-ordinates and an expansion in orthogonal functions used for the directional co-ordinates. Suitable sets of functions are spherical harmonics and Walsh functions. The latter set is appropriate if a discrete direction representation of the angular flux is required. (author)

Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)

International Nuclear Information System (INIS)

Pellegrino, Esteban

2011-01-01

Monte Carlo simulation is well suited for solving the Boltzmann neutron transport equation in an inhomogeneous media for complicated geometries. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop PC. The interest in adopting Graphics Processing Units (GPUs) for Monte Carlo acceleration is rapidly growing. This is due to the massive parallelism provided by the latest GPU technologies which is the most promising solution to the challenge of performing full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem were developed for GPU environments in order to evaluate issues associated with computational speedup using GPUs. Results obtained in this work suggest that a speedup of several orders of magnitude is possible using the state-of-the-art GPU technologies. (author) [es

Solution of the linearly anisotropic neutron transport problem in a infinite cylinder combining the decomposition and HTSN methods

International Nuclear Information System (INIS)

Goncalves, Glenio A.; Bodmann, Bardo; Bogado, Sergio; Vilhena, Marco T.

2008-01-01

Analytical solutions for neutron transport in cylindrical geometry is available for isotropic problems, but to the best of our knowledge for anisotropic problems are not available, yet. In this work, an analytical solution for the neutron transport equation in an infinite cylinder assuming anisotropic scattering is reported. Here we specialize the solution, without loss of generality, for the linearly anisotropic problem using the combined decomposition and HTS N methods. The key feature of this method consists in the application of the decomposition method to the anisotropic problem by virtue of the fact that the inverse of the operator associated to isotropic problem is well know and determined by the HTS N approach. So far, following the idea of the decomposition method, we apply this operator to the integral term, assuming that the angular flux appearing in the integrand is considered to be equal to the HTS N solution interpolated by polynomial considering only even powers. This leads to the first approximation for an anisotropic solution. Proceeding further, we replace this solution for the angular flux in the integral and apply again the inverse operator for the isotropic problem in the integral term and obtain a new approximation for the angular flux. This iterative procedure yields a closed form solution for the angular flux. This methodology can be generalized, in a straightforward manner, for transport problems with any degree of anisotropy. For the sake of illustration, we report numerical simulations for linearly anisotropic transport problems. (author)

The Application of Neutron Transport Green's Functions to Threat Scenario Simulation

Science.gov (United States)

Thoreson, Gregory G.; Schneider, Erich A.; Armstrong, Hirotatsu; van der Hoeven, Christopher A.

2015-02-01

Radiation detectors provide deterrence and defense against nuclear smuggling attempts by scanning vehicles, ships, and pedestrians for radioactive material. Understanding detector performance is crucial to developing novel technologies, architectures, and alarm algorithms. Detection can be modeled through radiation transport simulations; however, modeling a spanning set of threat scenarios over the full transport phase-space is computationally challenging. Previous research has demonstrated Green's functions can simulate photon detector signals by decomposing the scenario space into independently simulated submodels. This paper presents decomposition methods for neutron and time-dependent transport. As a result, neutron detector signals produced from full forward transport simulations can be efficiently reconstructed by sequential application of submodel response functions.

A random walk approach to stochastic neutron transport

International Nuclear Information System (INIS)

Mulatier, Clelia de

2015-01-01

One of the key goals of nuclear reactor physics is to determine the distribution of the neutron population within a reactor core. This population indeed fluctuates due to the stochastic nature of the interactions of the neutrons with the nuclei of the surrounding medium: scattering, emission of neutrons from fission events and capture by nuclear absorption. Due to these physical mechanisms, the stochastic process performed by neutrons is a branching random walk. For most applications, the neutron population considered is very large, and all physical observables related to its behaviour, such as the heat production due to fissions, are well characterised by their average values. Generally, these mean quantities are governed by the classical neutron transport equation, called linear Boltzmann equation. During my PhD, using tools from branching random walks and anomalous diffusion, I have tackled two aspects of neutron transport that cannot be approached by the linear Boltzmann equation. First, thanks to the Feynman-Kac backward formalism, I have characterised the phenomenon of 'neutron clustering' that has been highlighted for low-density configuration of neutrons and results from strong fluctuations in space and time of the neutron population. Then, I focused on several properties of anomalous (non-exponential) transport, that can model neutron transport in strongly heterogeneous and disordered media, such as pebble-bed reactors. One of the novel aspects of this work is that problems are treated in the presence of boundaries. Indeed, even though real systems are finite (confined geometries), most of previously existing results were obtained for infinite systems. (author) [fr

The spectral element approach for the solution of neutron transport problems

International Nuclear Information System (INIS)

Barbarino, A.; Dulla, S.; Ravetto, P.; Mund, E.H.

2011-01-01

In this paper a possible application of the Spectral Element Method to neutron transport problems is presented. The basic features of the numerical scheme on the one-dimensional diffusion equation are illustrated. Then, the AN model for neutron transport is introduced, and the basic steps for the construction of a bi-dimensional solver are described. The AN equations are chosen for their structure, involving a system of coupled elliptic-type equations. Some calculations are carried out on typical benchmark problems and results are compared with the Finite Element Method, in order to evaluate their performances. (author)

Development of new multigrid schemes for the method of characteristics in neutron transport theory

International Nuclear Information System (INIS)

Grassi, G.

2006-01-01

This dissertation is based upon our doctoral research that dealt with the conception and development of new non-linear multigrid techniques for the Method of the Characteristics (MOC) within the TDT code. Here we focus upon a two-level scheme consisting of a fine level on which the neutron transport equation is iteratively solved using the MOC algorithm, and a coarse level defined by a more coarsely discretized phase space on which a low-order problem is considered. The solution of this problem is then used in order to correct the angular flux moments resulting from the previous transport iteration. A flux-volume homogenization procedure is employed to evaluate the coarse-level material properties after each transport iteration. This entails the non-linearity of the methods. According to the Generalised Equivalence Theory (GET), additional degrees of freedom are introduced for the low-order problem so that the convergence of the acceleration scheme can be ensured. We present two classes of non-linear methods: transport-like methods and discussion-like methods. Transport-like methods consider a homogenized low-order transport problem on the coarse level. This problem is iteratively solved using the same MOC algorithm as for the transport problem on the fine level. Discontinuity factors are then employed, per region or per surface, in order to reconstruct the currents evaluated by the low-order operator, which ensure the convergence of the acceleration scheme. On the other hand, discussion-like methods consider a low-order problem inspired by diffusion. We studied the non-linear Coarse Mesh Finite Difference (CMFD) method, already present in literature, in the perspective of integrating it into TDT code. Then, we developed a new non-linear method on the model of CMFD. From the latter, we borrowed the idea to establish a simple relation between currents and fluxes in order to obtain a problem involving only coarse fluxes. Finally, those non-linear methods have been

A novel method for trace tritium transport studies

International Nuclear Information System (INIS)

Bonheure, Georges; Mlynar, Jan; Murari, A.; Giroud, C.; Popovichev, S.; Belo, P.; Bertalot, L.

2009-01-01

A new method combining a free-form solution for the neutron emissivity and the ratio method (Bonheure et al 2006 Nucl. Fusion 46 725-40) is applied to the investigation of tritium particle transport in JET plasmas. The 2D neutron emissivity is calculated using the minimum Fisher regularization method (MFR) (Anton et al 1996 Plasma Phys. Control. Fusion 38 1849, Mlynar et al 2003 Plasma Phys. Control. Fusion 45 169). This method is being developed and studied alongside other methods at JET. The 2D neutron emissivity was significantly improved compared with the first MFR results by constraining the emissivity along the magnetic flux surfaces. 1D profiles suitable for transport analysis are then obtained by subsequent poloidal integration. In methods on which previous JET publications are based (Stork et al 2005 Nucl. Fusion 45 S181, JET Team (prepared by Zastrow) 1999 Nucl. Fusion 39 1891, Zastrow et al 2004 Plasma Phys. Control. Fusion 46 B255, Adams et al 1993 Nucl. Instrum. Methods A 329 277, Jarvis et al 1997 Fusion Eng. Des. 34-35 59, Jarvis et al 1994 Plasma Phys. Control. Fusion 36 219), the 14.07 MeV D-T neutron line integrals measurements were simulated and the transport coefficients varied until good fits were obtained. In this novel approach, direct knowledge of tritium concentration or the fuel ratio n T /n D is obtained using all available neutron profile information, e.g both 2.45 MeV D-D neutron profiles and 14.07 MeV D-T neutron profiles (Bonheure et al 2006 Nucl.Fusion 46 725-40). Tritium particle transport coefficients are then determined using a linear regression from the dynamic response of the tritium concentration n T /n D profile. The temporal and spatial evolution of tritium particle concentration was studied for a set of JET discharges with tritium gas puffs from the JET trace tritium experiments. Local tritium transport coefficients were derived from the particle flux equation Γ = -D∇n T + Vn T , where D is the particle diffusivity and V

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Cosmic-ray neutron transport at a forest field site

DEFF Research Database (Denmark)

Andreasen, Mie; Jensen, Karsten Høgh; Desilets, Darin

2017-01-01

-ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different...

Solution of the neutron transport problem with anisotropic scattering in cylindrical geometry by the decomposition method

International Nuclear Information System (INIS)

Goncalves, G.A.; Bogado Leite, S.Q.; Vilhena, M.T. de

2009-01-01

An analytical solution has been obtained for the one-speed stationary neutron transport problem, in an infinitely long cylinder with anisotropic scattering by the decomposition method. Series expansions of the angular flux distribution are proposed in terms of suitably constructed functions, recursively obtainable from the isotropic solution, to take into account anisotropy. As for the isotropic problem, an accurate closed-form solution was chosen for the problem with internal source and constant incident radiation, obtained from an integral transformation technique and the F N method

OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

Energy Technology Data Exchange (ETDEWEB)

Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)

2017-06-15

Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.

A midway forward-adjoint coupling method for neutron and photon Monte Carlo transport

International Nuclear Information System (INIS)

Serov, I.V.; John, T.M.; Hoogenboom, J.E.

1999-01-01

The midway Monte Carlo method for calculating detector responses combines a forward and an adjoint Monte Carlo calculation. In both calculations, particle scores are registered at a surface to be chosen by the user somewhere between the source and detector domains. The theory of the midway response determination is developed within the framework of transport theory for external sources and for criticality theory. The theory is also developed for photons, which are generated at inelastic scattering or capture of neutrons. In either the forward or the adjoint calculation a so-called black absorber technique can be applied; i.e., particles need not be followed after passing the midway surface. The midway Monte Carlo method is implemented in the general-purpose MCNP Monte Carlo code. The midway Monte Carlo method is demonstrated to be very efficient in problems with deep penetration, small source and detector domains, and complicated streaming paths. All the problems considered pose difficult variance reduction challenges. Calculations were performed using existing variance reduction methods of normal MCNP runs and using the midway method. The performed comparative analyses show that the midway method appears to be much more efficient than the standard techniques in an overwhelming majority of cases and can be recommended for use in many difficult variance reduction problems of neutral particle transport

Study of a transportable neutron radiography system

International Nuclear Information System (INIS)

Souza, S.N.A. de.

1991-05-01

This work presents a study a transportable neutron radiography system for a 185 GBq 241 Am-Be (α, η) source with a neutron yield roughly 1,25 x 10 7 n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio of 14, for neutron fluxes up to 4,09 x 10 2 n.cm -2 .s -1 . Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10 2 n.cm -2 .s -1 and 2,65 x 10 2 n.cm -2 .s -1 , respectively. (author). 30 refs, 39 figs, 9 tabs

Interfacing MCNPX and McStas for simulation of neutron transport

DEFF Research Database (Denmark)

Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

2013-01-01

Stas[4, 5, 6, 7]. The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as Mc...... geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides....

Nystro¨m method applied to integral formulation of the neutron transport equation in X-Y geometry

Energy Technology Data Exchange (ETDEWEB)

Azevedo, Fabio S.; Sauter, Esequia; Konzen, Pedro H.A.; Barichello, Liliane B., E-mail: fabio.azevedo@ufrgs.br, E-mail: esequia.sauter@ufrgs.br, E-mail: pedro.konzen@ufrgs.br, E-mail: lbaric@mat.ufrgs.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Departamento de Matem´atica Pura e Aplicada

2017-07-01

Neutron transport problems in X-Y geometry have been solved with several techniques in last decades but it is still a challenge to produce a good balance between computational efficiency and accuracy. In this work, we address this problem by efficiently applying the Nystr¨om method to the integral formulation of the transport equation. Analytical techniques, modern numerical packages and optimized implementation were applied to reduce the computational time. This method presented results free of ray effects leading to high accurate numerical results for two-dimensional scalar flux. Our implementation simulates homogeneous problems with vacuum and reflective boundary conditions. Results were validated with up to seven significant digits and compared with those available in the literature. (author)

Error reduction techniques for Monte Carlo neutron transport calculations

International Nuclear Information System (INIS)

Ju, J.H.W.

1981-01-01

Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas

Application of space-and-angle finite element method to the three-dimensional neutron transport problems

International Nuclear Information System (INIS)

Fujimura, T.; Nakahara, Y.; Matsumura, M.

1983-01-01

A double finite element method (DFEM), in which both the space-and-angle finite elements are employed, has been formulated and computer codes have been developed to solve the static multigroup neutron transport problems in the three-dimensional geometry. Two methods, Galerkin's weighted residual and variational are used to apply the DFEM to the transport equation. The variational principle requires complicated formulation than the Galerkin method, but the boundary conditions can be automatically incorporated and each plane equation becomes symmetric. The system equations are solved over the planar layers which we call plane iteration. The coarse mesh rebalancing technique is used for the inner iteration and the outer iteration is accelerated by extra-polation. Numerical studies of these two DFEM algorithms have been done in comparison between them and also with THe CITATION and TWOTRAN-II results. It has been confirmed that in the case of variational formulation an adaptive acceleration method of the SSOR iteration works effectively and the ray effects are mitigated in both DFEM algorithms. (author)

Advances in the solution of three-dimensional nodal neutron transport equation

International Nuclear Information System (INIS)

Pazos, Ruben Panta; Hauser, Eliete Biasotto; Vilhena, Marco Tullio de

2003-01-01

In this paper we study the three-dimensional nodal discrete-ordinates approximations of neutron transport equation in a convex domain with piecewise smooth boundaries. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtaining the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. We give numerical results obtained with an algebraic computer system (for N up to 8) and with a code for higher values of N. We compare our results for the geometry of a box with a source in a vertex and a leakage zone in the opposite with others techniques used in this problem. (author)

Use of the light-water neutron scattering kernel in the study of neutron transport processes in mixtures of light and heavy water

International Nuclear Information System (INIS)

Tewari, S.P.

1975-01-01

A method of studying neutron transport properties in H 2 O-D 2 O mixtures, both liquid and solid, which extrapolates the neutron thermalization parameters of H 2 O is described. The decay of pulsed neutrons in the media has been investigated as an example of the application of the method. The results of the calcutions agree with the experiment for concentrations up to 50 percent D 2 O. (1 figure) (U.S.)

Neutron Flux Interpolation with Finite Element Method in the Nuclear Fuel Cell Calculation using Collision Probability Method

International Nuclear Information System (INIS)

Shafii, M. Ali; Su'ud, Zaki; Waris, Abdul; Kurniasih, Neny; Ariani, Menik; Yulianti, Yanti

2010-01-01

Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the k eff and other parameters are investigated.

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

International Nuclear Information System (INIS)

Biondo, Elliott D.; Wilson, Paul P. H.

2017-01-01

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 _± 5 • _1_0_"_4 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.

Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1, 1975--August 31, 1976

International Nuclear Information System (INIS)

Bareiss, E.H.

1976-05-01

The objectives of the work are to develop mathematically and computationally founded for the design of highly efficient and reliable multidimensional neutron transport codes to solve a variety of neutron migration and radiation problems, and to analyze existing and new methods for performance. As new analytical insights are gained, new numerical methods are developed and tested. Significant results obtained include implementation of the integer-preserving Gaussian elimination method (two-step method) in a CDC 6400 computer code, modes analysis for one-dimensional transport solutions, and a new method for solving the 1-T transport equation. Some of the work dealt with the interface and corner problem in diffusion theory

Linear stochastic neutron transport theory

International Nuclear Information System (INIS)

Lewins, J.

1978-01-01

A new and direct derivation of the Bell-Pal fundamental equation for (low power) neutron stochastic behaviour in the Boltzmann continuum model is given. The development includes correlation of particle emission direction in induced and spontaneous fission. This leads to generalizations of the backward and forward equations for the mean and variance of neutron behaviour. The stochastic importance for neutron transport theory is introduced and related to the conventional deterministic importance. Defining equations and moment equations are derived and shown to be related to the backward fundamental equation with the detector distribution of the operational definition of stochastic importance playing the role of an adjoint source. (author)

The infinite medium Green's function for neutron transport in plane geometry 40 years later

International Nuclear Information System (INIS)

Ganapol, B.D.

1993-01-01

In 1953, the first of what was supposed to be two volumes on neutron transport theory was published. The monograph, entitled open-quotes Introduction to the Theory of Neutron Diffusionclose quotes by Case et al., appeared as a Los Alamos National Laboratory report and was to be followed by a second volume, which never appeared as intended because of the death of Placzek. Instead, Case and Zweifel collaborated on the now classic work entitled Linear Transport Theory 2 in which the underlying mathematical theory of linear transport was presented. The initial monograph, however, represented the coming of age of neutron transport theory, which had its roots in radiative transfer and kinetic theory. In addition, it provided the first benchmark results along with the mathematical development for several fundamental neutron transport problems. In particular, one-dimensional infinite medium Green's functions for the monoenergetic transport equation in plane and spherical geometries were considered complete with numerical results to be used as standards to guide code development for applications. Unfortunately, because of the limited computational resources of the day, some numerical results were incorrect. Also, only conventional mathematics and numerical methods were used because the transport theorists of the day were just becoming acquainted with more modern mathematical approaches. In this paper, Green's function solution is revisited in light of modern numerical benchmarking methods with an emphasis on evaluation rather than theoretical results. The primary motivation for considering the Green's function at this time is its emerging use in solving finite and heterogeneous media transport problems

Variational formulation and projectional methods for the second order transport equation

International Nuclear Information System (INIS)

Borysiewicz, M.; Stankiewicz, R.

1979-01-01

Herein the variational problem for a second-order boundary value problem for the neutron transport equation is formulated. The projectional methods solving the problem are examined. The approach is compared with that based on the original untransformed form of the neutron transport equation

A variable-order time-dependent neutron transport method for nuclear reactor kinetics using analytically-integrated space-time characteristics

International Nuclear Information System (INIS)

Hoffman, A. J.; Lee, J. C.

2013-01-01

A new time-dependent neutron transport method based on the method of characteristics (MOC) has been developed. Whereas most spatial kinetics methods treat time dependence through temporal discretization, this new method treats time dependence by defining the characteristics to span space and time. In this implementation regions are defined in space-time where the thickness of the region in time fulfills an analogous role to the time step in discretized methods. The time dependence of the local source is approximated using a truncated Taylor series expansion with high order derivatives approximated using backward differences, permitting the solution of the resulting space-time characteristic equation. To avoid a drastic increase in computational expense and memory requirements due to solving many discrete characteristics in the space-time planes, the temporal variation of the boundary source is similarly approximated. This allows the characteristics in the space-time plane to be represented analytically rather than discretely, resulting in an algorithm comparable in implementation and expense to one that arises from conventional time integration techniques. Furthermore, by defining the boundary flux time derivative in terms of the preceding local source time derivative and boundary flux time derivative, the need to store angularly-dependent data is avoided without approximating the angular dependence of the angular flux time derivative. The accuracy of this method is assessed through implementation in the neutron transport code DeCART. The method is employed with variable-order local source representation to model a TWIGL transient. The results demonstrate that this method is accurate and more efficient than the discretized method. (authors)

Analysis of a block Gauss-Seidel iterative method for a finite element discretization of the neutron transport equation

International Nuclear Information System (INIS)

Lorence, L.J. Jr.; Martin, W.R.; Luskin, M.

1985-01-01

We prove the convergence of a finite element discretization of the neutron transport equation. The iterative solution of the resulting linear system by a block Gauss-Seidel method is also analyzed. This procedure is shown to require less storage than the direct solution by Gaussian elimination, and an estimate for the rate of convergence is used to show that fewer arithmetic operations are required

Simulation of neutron transport equation using parallel Monte Carlo for deep penetration problems

International Nuclear Information System (INIS)

Bekar, K. K.; Tombakoglu, M.; Soekmen, C. N.

2001-01-01

Neutron transport equation is simulated using parallel Monte Carlo method for deep penetration neutron transport problem. Monte Carlo simulation is parallelized by using three different techniques; direct parallelization, domain decomposition and domain decomposition with load balancing, which are used with PVM (Parallel Virtual Machine) software on LAN (Local Area Network). The results of parallel simulation are given for various model problems. The performances of the parallelization techniques are compared with each other. Moreover, the effects of variance reduction techniques on parallelization are discussed

Application of Trotter approximation for solving time dependent neutron transport equation

International Nuclear Information System (INIS)

Stancic, V.

1987-01-01

A method is proposed to solve multigroup time dependent neutron transport equation with arbitrary scattering anisotropy. The recurrence relation thus obtained is simple, numerically stable and especially suitable for treatment of complicated geometries. (author)

Analysis and development of spatial hp-refinement methods for solving the neutron transport equation

International Nuclear Information System (INIS)

Fournier, D.

2011-01-01

The different neutronic parameters have to be calculated with a higher accuracy in order to design the 4. generation reactor cores. As memory storage and computation time are limited, adaptive methods are a solution to solve the neutron transport equation. The neutronic flux, solution of this equation, depends on the energy, angle and space. The different variables are successively discretized. The energy with a multigroup approach, considering the different quantities to be constant on each group, the angle by a collocation method called SN approximation. Once the energy and angle variable are discretized, a system of spatially-dependent hyperbolic equations has to be solved. Discontinuous finite elements are used to make possible the development of hp-refinement methods. Thus, the accuracy of the solution can be improved by spatial refinement (h-refinement), consisting into subdividing a cell into sub-cells, or by order refinement (p-refinement), by increasing the order of the polynomial basis. In this thesis, the properties of this methods are analyzed showing the importance of the regularity of the solution to choose the type of refinement. Thus, two error estimators are used to lead the refinement process. Whereas the first one requires high regularity hypothesis (analytical solution), the second one supposes only the minimal hypothesis required for the solution to exist. The comparison of both estimators is done on benchmarks where the analytic solution is known by the method of manufactured solutions. Thus, the behaviour of the solution as a regard of the regularity can be studied. It leads to a hp-refinement method using the two estimators. Then, a comparison is done with other existing methods on simplified but also realistic benchmarks coming from nuclear cores. These adaptive methods considerably reduces the computational cost and memory footprint. To further improve these two points, an approach with energy-dependent meshes is proposed. Actually, as the

Importance estimation in Monte Carlo modelling of neutron and photon transport

International Nuclear Information System (INIS)

Mickael, M.W.

1992-01-01

The estimation of neutron and photon importance in a three-dimensional geometry is achieved using a coupled Monte Carlo and diffusion theory calculation. The parameters required for the solution of the multigroup adjoint diffusion equation are estimated from an analog Monte Carlo simulation of the system under investigation. The solution of the adjoint diffusion equation is then used as an estimate of the particle importance in the actual simulation. This approach provides an automated and efficient variance reduction method for Monte Carlo simulations. The technique has been successfully applied to Monte Carlo simulation of neutron and coupled neutron-photon transport in the nuclear well-logging field. The results show that the importance maps obtained in a few minutes of computer time using this technique are in good agreement with Monte Carlo generated importance maps that require prohibitive computing times. The application of this method to Monte Carlo modelling of the response of neutron porosity and pulsed neutron instruments has resulted in major reductions in computation time. (Author)

Numerical solution of the Neutron Transport Equation using discontinuous nodal methods at X-Y geometry

International Nuclear Information System (INIS)

Delfin L, A.

1996-01-01

The purpose of this work is to solve the neutron transport equation in discrete-ordinates and X-Y geometry by developing and using the strong discontinuous and strong modified discontinuous nodal finite element schemes. The strong discontinuous and modified strong discontinuous nodal finite element schemes go from two to ten interpolation parameters per cell. They are describing giving a set D c and polynomial space S c corresponding for each scheme BDMO, RTO, BL, BDM1, HdV, BDFM1, RT1, BQ and BDM2. The solution is obtained solving the neutron transport equation moments for each nodal scheme by developing the basis functions defined by Pascal triangle and the Legendre moments giving in the polynomial space S c and, finally, looking for the non singularity of the resulting linear system. The linear system is numerically solved using a computer program for each scheme mentioned . It uses the LU method and forward and backward substitution and makes a partition of the domain in cells. The source terms and angular flux are calculated, using the directions and weights associated to the S N approximation and solving the angular flux moments to find the effective multiplication constant. The programs are written in Fortran language, using the dynamic allocation of memory to increase efficiently the available memory of the computing equipment. (Author)

A time-dependent neutron transport method of characteristics formulation with time derivative propagation

Energy Technology Data Exchange (ETDEWEB)

Hoffman, Adam J., E-mail: adamhoff@umich.edu; Lee, John C., E-mail: jcl@umich.edu

2016-02-15

A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Source Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.

A time-dependent neutron transport method of characteristics formulation with time derivative propagation

International Nuclear Information System (INIS)

Hoffman, Adam J.; Lee, John C.

2016-01-01

A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Source Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.

Measurements of anomalous neutron transport in bulk graphite

International Nuclear Information System (INIS)

Bowman, C.D.; Smith, G.A.; Vogelaar, B.; Howell, C.R.; Bilpuch, E.G.; Tornow, W.

2003-01-01

The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)

Measurements of anomalous neutron transport in bulk graphite

Energy Technology Data Exchange (ETDEWEB)

Bowman, C.D.; Smith, G.A. [ADNA Corp., Los Alamos, NM (United States); Vogelaar, B. [Virginia Tech., Blacksburg, VA (United States); Howell, C.R.; Bilpuch, E.G.; Tornow, W. [Triangle Univ. Nuclear Lab., Duke Univ., Durham, NC (United States)

2003-07-01

The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)

Structures of the fractional spaces generated by the difference neutron transport operator

International Nuclear Information System (INIS)

Ashyralyev, Allaberen; Taskin, Abdulgafur

2015-01-01

The initial boundary value problem for the neutron transport equation is considered. The first, second and third order of accuracy difference schemes for the approximate solution of this problem are presented. Highly accurate difference schemes for neutron transport equation based on Padé approximation are constructed. In applications, stability estimates for solutions of difference schemes for the approximate solution of the neutron transport equation are obtained.The positivity of the neutron transport operator in Slobodeckij spaces is proved. Numerical techniques are developed and algorithms are tested on an example in MATLAB

Methods of neutron spectrometry

International Nuclear Information System (INIS)

Doerschel, B.

1981-01-01

The different methods of neutron spectrometry are based on the direct measurement of neutron velocity or on the use of suitable energy-dependent interaction processes. In the latter case the measuring effect of a detector is connected with the searched neutron spectrum by an integral equation. The solution needs suitable unfolding procedures. The most important methods of neutron spectrometry are the time-of-flight method, the crystal spectrometry, the neutron spectrometry by use of elastic collisions with hydrogen nuclei, and neutron spectrometry with the aid of nuclear reactions, especially of the neutron-induced activation. The advantages and disadvantages of these methods are contrasted considering the resolution, the measurable energy range, the sensitivity, and the experimental and computational efforts. (author)

The transport of neutrons and gamma-rays in the air

International Nuclear Information System (INIS)

Adamski, J.

1980-01-01

The transport of neutrons and gamma rays in the infinite homogeneous air has been investigated. For the calculations has been used the Multigroup One Dimensional Discrete Ordinates Transport Code ANISN-W. The calculations have been performed for three types of neutron sources. The neutrons and gamma ray doses in the air have been analyzed, and comparison to the other authors' results has been given. (author)

Three-group albedo method applied to the diffusion phenomenon with up-scattering of neutrons

International Nuclear Information System (INIS)

Terra, Andre M. Barge Pontes Torres; Silva, Jorge A. Valle da; Cabral, Ronaldo G.

2007-01-01

The main objective of this research is to develop a three-group neutron Albedo algorithm considering the up-scattering of neutrons in order to analyse the diffusion phenomenon in nonmultiplying media. The neutron Albedo method is an analytical method that does not try to solve describing explicit equations for the neutron fluxes. Thus the neutron Albedo methodology is very different from the conventional methodology, as the neutron diffusion theory model. Graphite is analyzed as a model case. One major application is in the determination of the nonleakage probabilities with more understandable results in physical terms than conventional radiation transport method calculations. (author)

Computational Modeling of a Time-Independent, Heterogeneous Reactor Core Using Simplified Discrete Ordinates Neutron Transport Techniques

National Research Council Canada - National Science Library

Labowski, Kristofer

2001-01-01

The Linear Characteristic (LC) method on rectangular boxoid meshes is a discrete ordinate neutron transport technique that uses both zeroth and first moments of the angular neutron flux to construct a relatively accurate...

Completely boundary-free minimum and maximum principles for neutron transport and their least-squares and Galerkin equivalents

International Nuclear Information System (INIS)

Ackroyd, R.T.

1982-01-01

Some minimum and maximum variational principles for even-parity neutron transport are reviewed and the corresponding principles for odd-parity transport are derived by a simple method to show why the essential boundary conditions associated with these maximum principles have to be imposed. The method also shows why both the essential and some of the natural boundary conditions associated with these minimum principles have to be imposed. These imposed boundary conditions for trial functions in the variational principles limit the choice of the finite element used to represent trial functions. The reasons for the boundary conditions imposed on the principles for even- and odd-parity transport point the way to a treatment of composite neutron transport, for which completely boundary-free maximum and minimum principles are derived from a functional identity. In general a trial function is used for each parity in the composite neutron transport, but this can be reduced to one without any boundary conditions having to be imposed. (author)

Transport of accelerator produced high energy neutrons though concrete

International Nuclear Information System (INIS)

Prabhakar Rao, G.; Sarkar, P.K.

1996-01-01

Development of a computational system for estimating the production and transport of high energy neutrons in particle accelerators is reported. The energy-angle distribution of neutrons from accelerated ions bombarding thick targets is calculated by a hybrid nuclear reaction model code, ALICE-91, modified to suit the purpose. Subsequent transmission of these neutrons through concrete slabs is treated using the anisotropic source-flux iteration technique (ASFIT) in the framework of a coupled neutron-gamma transport. Several parameters of both the codes have been optimized to obtain the transmitted dose through concrete. The calculations are found to be accurate and at the same time faster compared to the detailed Monte Carlo calculations. (author). 8 refs., 2 figs

Analysis of a HP-refinement method for solving the neutron transport equation using two error estimators

International Nuclear Information System (INIS)

Fournier, D.; Le Tellier, R.; Suteau, C.; Herbin, R.

2011-01-01

The solution of the time-independent neutron transport equation in a deterministic way invariably consists in the successive discretization of the three variables: energy, angle and space. In the SNATCH solver used in this study, the energy and the angle are respectively discretized with a multigroup approach and the discrete ordinate method. A set of spatial coupled transport equations is obtained and solved using the Discontinuous Galerkin Finite Element Method (DGFEM). Within this method, the spatial domain is decomposed into elements and the solution is approximated by a hierarchical polynomial basis in each one. This approach is time and memory consuming when the mesh becomes fine or the basis order high. To improve the computational time and the memory footprint, adaptive algorithms are proposed. These algorithms are based on an error estimation in each cell. If the error is important in a given region, the mesh has to be refined (h−refinement) or the polynomial basis order increased (p−refinement). This paper is related to the choice between the two types of refinement. Two ways to estimate the error are compared on different benchmarks. Analyzing the differences, a hp−refinement method is proposed and tested. (author)

Two-group neutron transport theory in adjacent space with lineary anisotropic scattering

International Nuclear Information System (INIS)

Maiorino, J.R.

1978-01-01

A solution method for two-group neutron transport theory with anisotropic scattering is introduced by the combination of case method (expansion method of self singular function) and the invariant imbedding (invariance principle). The numerical results for the Milne problem in light water and borated water is presented to demonstrate the avalibility of the method [pt

Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1976--November 30, 1977

International Nuclear Information System (INIS)

Bareiss, E.H.

1977-08-01

The objectives of this research are to develop mathematically and computationally founded criteria for the design of highly efficient and reliable multidimensional neutron transport codes to solve a variety of neutron migration and radiation problems, and to analyze existing and new methods for performance

Experimental validation of GADRAS's coupled neutron-photon inverse radiation transport solver

International Nuclear Information System (INIS)

Mattingly, John K.; Mitchell, Dean James; Harding, Lee T.

2010-01-01

Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of α-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.

A numerical method for two-dimensional anisotropic transport problem in cylindrical geometry

International Nuclear Information System (INIS)

Du Mingsheng; Feng Tiekai; Fu Lianxiang; Cao Changshu; Liu Yulan

1988-01-01

The authors deal with the triangular mesh-discontinuous finite element method for solving the time-dependent anisotropic neutron transport problem in two-dimensional cylindrical geometry. A prior estimate of the numerical solution is given. Stability is proved. The authors have computed a two dimensional anisotropic neutron transport problem and a Tungsten-Carbide critical assembly problem by using the numerical method. In comparision with DSN method and the experimental results obtained by others both at home and abroad, the method is satisfactory

Two-dimensional Haar wavelet Collocation Method for the solution of Stationary Neutron Transport Equation in a homogeneous isotropic medium

International Nuclear Information System (INIS)

Patra, A.; Saha Ray, S.

2014-01-01

Highlights: • A stationary transport equation has been solved using the technique of Haar wavelet Collocation Method. • This paper intends to provide the great utility of Haar wavelets to nuclear science problem. • In the present paper, two-dimensional Haar wavelets are applied. • The proposed method is mathematically very simple, easy and fast. - Abstract: This paper emphasizes on finding the solution for a stationary transport equation using the technique of Haar wavelet Collocation Method (HWCM). Haar wavelet Collocation Method is efficient and powerful in solving wide class of linear and nonlinear differential equations. Recently Haar wavelet transform has gained the reputation of being a very effective tool for many practical applications. This paper intends to provide the great utility of Haar wavelets to nuclear science problem. In the present paper, two-dimensional Haar wavelets are applied for solution of the stationary Neutron Transport Equation in homogeneous isotropic medium. The proposed method is mathematically very simple, easy and fast. To demonstrate about the efficiency of the method, one test problem is discussed. It can be observed from the computational simulation that the numerical approximate solution is much closer to the exact solution

Angular neutron transport investigation in the HZETRN free-space ion and nucleon transport and shielding computer program

International Nuclear Information System (INIS)

Singleterry, R.C. Jr.; Wilson, J.W.

1997-01-01

Extension of the high charge and energy (HZE) transport computer program HZETRN for angular transport of neutrons is considered. For this paper, only light ion transport, He 4 and lighter, will be analyzed using a pure solar proton source. The angular transport calculator is the ANISN/PC program which is being controlled by the HZETRN program. The neutron flux values are compared for straight-ahead transport and angular transport in one dimension. The shield material is aluminum and the target material is water. The thickness of these materials is varied; however, only the largest model calculated is reported which is 50 gm/cm 2 of aluminum and 100 gm/cm 2 of water. The flux from the ANISN/PC calculation is about two orders of magnitude lower than the flux from HZETRN for very low energy neutrons. It is only a magnitude lower for the neutrons in the 10 to 20 MeV range in the aluminum and two orders lower in the water. The major reason for this difference is in the transport modes: straight-ahead versus angular. The angular treatment allows a longer path length than the straight-ahead approximation. Another reason is the different cross section sets used by the ANISN/PC-BUGLE-80 mode and the HZETRN mode. The next step is to investigate further the differences between the two codes and isolate the differences to just the angular versus straight-ahead transport mode. Then, create a better coupling between the angular neutron transport and the charged particle transport

Monte Carlo method in radiation transport problems

International Nuclear Information System (INIS)

Dejonghe, G.; Nimal, J.C.; Vergnaud, T.

1986-11-01

In neutral radiation transport problems (neutrons, photons), two values are important: the flux in the phase space and the density of particles. To solve the problem with Monte Carlo method leads to, among other things, build a statistical process (called the play) and to provide a numerical value to a variable x (this attribution is called score). Sampling techniques are presented. Play biasing necessity is proved. A biased simulation is made. At last, the current developments (rewriting of programs for instance) are presented due to several reasons: two of them are the vectorial calculation apparition and the photon and neutron transport in vacancy media [fr

Effect of granulation of geological samples in neutron transport measurements

International Nuclear Information System (INIS)

Woznicka, Urszula; Drozdowicz, Krzysztof; Gabanska, Barbara; Krynicka, Ewa; Igielski, Andrzej

2001-01-01

The thermal neutron absorption cross section is one of the parameters describing the transport of thermal neutrons in a medium. Theoretical descriptions and experiments which determine the absorption cross section have a wide literature for homogeneous media. The situation comes true e.g. for fluids or amorphous solids. There are many other media which should be treated as heterogeneous. Among others - geological materials. The material heterogeneity for the thermal neutron transport in a considered volume is understood here as an existence of many small regions which differ significantly in their macroscopic neutron diffusion parameters (defined by the absorption and transport cross sections). The final difference, which influences the neutron transport, comes from a combination of the absolute differences between the parameters and of sizes of regions (related to the neutron mean free paths). A rock can be naturally heterogeneous in the above meaning. Besides, it can happen that a preparation of the rock sample for a neutron measurement can increase its natural heterogeneity. (For example, when the rock material is crushed and the measured sample consists of the obtained grains). The question is which granulation is allowed to treat the sample material as still homogeneous, and from which size of the rock grains we have to consider a two-component medium. It has been experimentally proved that the effective absorption of thermal neutrons in a heterogeneous two-component material can significantly differ from the absorption in a homogeneous one which consists of the same elements. The final effect is dependent on a few factors: the macroscopic absorption cross sections of the components, their total mass contributions, and the size of the grains. The ratio of the effective absorption cross section of the heterogeneous material to the cross section of the equivalent homogeneous, is a measure of the heterogeneity effect on the thermal neutron absorption

Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations. Interim report; Weiterentwicklung moderner Verfahren zu Neutronentransport und Unsicherheitsanalysen fuer Kernberechnungen. Zwischenbericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, Winfried; Aures, Alexander; Bostelmann, Friederike; Pasichnyk, Ihor; Perin, Yann; Velkov, Kiril; Zilly, Matias

2016-12-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1536 ''Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations'' as of the 3{sup rd} quarter of 2016. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts, in particular fast reactors cooled by liquid metal. The contributing individual goals are the further optimization and validation of deterministic calculation methods with high spatial and energy resolution, the development of a coupled calculation system using the Monte Carlo method for the neutron transport to describe time-dependent reactor core states, the processing and validation of nuclear data, particularly with regard to covariance data, the development, validation, and application of sampling-based methods for uncertainty and sensitivity analyses, the creation of a platform for performing systematic uncertainty analyses for fast reactor systems, as well as the description of states of severe core damage with the Monte Carlo method. Moreover, work regarding the European NURESAFE project, started in the preceding project RS1503, are being continued and completed.

Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

International Nuclear Information System (INIS)

Thiagu Supramaniam

2007-01-01

The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

Safety improvement of start-up neutron source handling work by preparing new transport containers

International Nuclear Information System (INIS)

Shimazaki, Yosuke; Sawahata, Hiroaki; Yanagida, Yoshinori; Shinohara, Masanori; Kawamoto, Taiki; Takada, Shoji

2016-01-01

The conventional transport containers that have been used in HTTR start-up neutron source replacement work are not specialized type for HTTR start-up neutron source. As the risks associated with the safe handling of neutron source holders due to the above fact, the following three risks have been confirmed: (1) exposure risk due to leakage of neutron source or gamma rays, (2) risk of erroneous fall of neutron source holders, and (3) accident due to incorrect handling of transport containers. This paper reports the risks confirmed in the handling of neutron source holders associated with transport containers and the risk reduction measures, as well as the fabrication of new transport containers. By employing the size-reduction and simple structure, new transport containers have been completed at the same level of costs compared with the continuous use of the conventional transport containers, while satisfying the criteria of Type A transport materials and serving as risk preventive measures. Thus, new transport containers aimed at the risk prevention measures for the handling work of neutron source holders have been completed, and the safety of operation has been improved. (A.O.)

Neutron and gamma-ray transport experiments in liquid air

International Nuclear Information System (INIS)

Farley, W.E.

1976-01-01

Accurate estimates of neutron and gamma radiations from a nuclear explosion and their subsequent transport through the atmosphere are vital to nuclear-weapon employment studies: i.e., for determining safety radii for aircraft crews, casualty and collateral-damage risk radii for tactical weapons, and the kill range from a high-yield defensive burst for a maneuvering reentry vehicle. Radiation transport codes, such as the Laboratory's TARTNP, are used to calculate neutron and gamma fluences. Experiments have been performed to check and update these codes. Recently, a 1.3-m-radius liquid-air (21 percent oxygen) sphere, with a pulsed source of 14-MeV neutrons at its center, was used to measure the fluence and spectra of emerging neutrons and secondary gamma rays. Comparison of measured radiation dose with TARTNP showed agreement within 10 percent

Positive solution of a time and energy dependent neutron transport problem

International Nuclear Information System (INIS)

Pao, C.V.

1975-01-01

A constructive method is given for the determination of a solution and an existence--uniqueness theorem for some nonlinear time and energy dependent neutron transport problems, including the linear transport system. The geometry of the medium under consideration is allowed to be either bounded or unbounded which includes the geometry of a finite or infinite cylinder, a half-space and the whole space R/subm/ (m=1,2,center-dotcenter-dotcenter-dot). Our approach to the problem is by successive approximation which leads to various recursion formulas for the approximations in terms of explicit integrations. It is shown under some Lipschitz conditions on the nonlinear functions, which describe the process of neutrons absorption, fission, and scattering, that the sequence of approximations converges to a unique positive solution. Since these conditions are satisfied by the linear transport equation, all the results for the nonlinear system are valid for the linear transport problem. In the general nonlinear problem, the existence of both local and global solutions are discussed, and an iterative process for the construction of the solution is given

AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

International Nuclear Information System (INIS)

1990-01-01

1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

On the importance of effective convergence velocity of synthetic acceleration methods in neutron transport

International Nuclear Information System (INIS)

Coppa, G.G.M.; Ravetto, P.; Colombo, V.

1996-01-01

The present work concerns some aspects of the optimization of the synthesis acceleration techniques in neutron transport. The importance of non-asymptotic convergence velocity as a theoretical means to characterize and optimize acceleration methods is discussed in detail for isotropic as well as highly anisotropic scattering cases; this shows the innacuracy of results based only on the usual asyptotic analysis. A detailed study of convergence velocity behaviour for space discretized schemes and multidimensional problems is also presented. Finally, various kinds of theoretical-evaluated convergence velocities are reported to study the effective behaviour of some modifications of the classic DSA technique recently proposed to face its loss of effectiveness and optimize performances when dealing with highly anisotropic scattering; comparisons with results of already assessed DSA modification techniques are reported for various scattering cross-section configurations. (Author)

A Green function of neutron transport equation

International Nuclear Information System (INIS)

Simovic, R.

1993-01-01

In this paper the angularly dependent Green function of the neutron transport equation is derived analytically and approximately. By applying the analytical FDPN approximation up to eighth order, numerical values of the Green functions are obtained with the accuracy of six significant figures in the whole range of parameter c, angle cosine μ and distances x up to the ten optical lengths from the neutron source. (author)

Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

Science.gov (United States)

Colonna, N; Altieri, S

2002-06-01

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

Discrete nodal integral transport-theory method for multidimensional reactor physics and shielding calculations

International Nuclear Information System (INIS)

Lawrence, R.D.; Dorning, J.J.

1980-01-01

A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method

On the spectrum of the one-speed slab-geometry discrete ordinates operator in neutron transport theory

International Nuclear Information System (INIS)

Abreu, Marcos Pimenta de

1998-01-01

We describe a numerical method applied to the first-order form of one-speed slab-geometry discrete ordinates equations modelling time-independent neutron transport problems with anisotropic scattering, with no interior source and defined in a nonmultiplying homogeneous host medium. Our numerical method is concerned with the generation of the spectrum and of a vector basis for the null space of the one-speed slab-geometry discrete ordinates operator. Moreover, it allows us to overcome the difficulties introduced in previous methods by anisotropic scattering and by angular quadrature sets of high order. To illustrate the positive features of our numerical method, we present numerical results for one-speed slab-geometry neutron transport model problems with anisotropic scattering

Neutron measurement by transportable spectrometer

International Nuclear Information System (INIS)

Anon.

1990-01-01

Two levels of neutron spectrometry are in regular use at nuclear power plants: some techniques used in the laboratory produce detailed spectra but require specialist operators, while simple instruments used by non-specialists to measure the neutron dose-rate to operators provide little spectral information. The standard portable instruments are therefore of no use when anomalous readings are obtained which require further investigation. AEA Technology at Winfrith has developed a Transportable Neutron Spectrometer (TNS) which is designed to produce reasonable spectra in routine use by staff with no specialist skill in spectroscopy, and high-quality spectra in the hands of skilled staff. The TNS provides a level of information intermediate between those currently available, and is also designed to solve the problem of imperfect dose response which is common in portable dosimeters. The TNS system consists of a power supply, a probe and a signal processing and data acquisition unit. (author)

In situ neutron depth profiling: A powerful method to probe lithium transport in micro-batteries

NARCIS (Netherlands)

Oudenhoven, J.F.M.; Labohm, F.; Mulder, M.; Niessen, R.A.H.; Mulder, F.M.; Notten, P.H.L.

2011-01-01

In situ neutron depth profiling (NDP) offers the possibility to observe lithium transport inside micro-batteries during battery operation. It is demonstrated that NDP results are consistent with the results of electrochemical measurements, and that the use of an enriched6LiCoO2 cathode offers more

MC++: A parallel, portable, Monte Carlo neutron transport code in C++

International Nuclear Information System (INIS)

Lee, S.R.; Cummings, J.C.; Nolen, S.D.

1997-01-01

MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms

Mathematical models for volume rendering and neutron transport

International Nuclear Information System (INIS)

Max, N.

1994-09-01

This paper reviews several different models for light interaction with volume densities of absorbing, glowing, reflecting, or scattering material. They include absorption only, glow only, glow and absorption combined, single scattering of external illumination, and multiple scattering. The models are derived from differential equations, and illustrated on a data set representing a cloud. They are related to corresponding models in neutron transport. The multiple scattering model uses an efficient method to propagate the radiation which does not suffer from the ray effect

Continuous energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the neutron energy spectrum by target nuclei tagging

Energy Technology Data Exchange (ETDEWEB)

Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Grupo de Estudos Nucleares; Leite, Sergio Q. Bogado, E-mail: sbogado@ibest.com.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

In this work a Monte Carlo simulator with continuous energy is used. This simulator distinguishes itself by using the 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 (for high energy neutrons) and the Maxwell-Boltzmann distribution (for thermal neutrons). The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. It is common practice in neutron transport calculations, e.g. multi-group transport, to consider that the neutrons only lose energy with each scattering reaction and then to use a thermal group with 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, i.e. in the thermal energy region, in which it can be regarded as a Maxwell-Boltzmann distribution for thermal equilibrium. 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. It is then shown how this procedure can emulate the up-scattering effect by the increase in the neutron population kinetic energy. 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. This work contains some preliminary results obtained from a Monte Carlo simulator for neutron transport that is being developed at Federal University of Rio Grande do Sul. (author)

The potential of detecting intermediate-scale biomass and canopy interception in a coniferous forest using cosmic-ray neutron intensity measurements and neutron transport modeling

Science.gov (United States)

Andreasen, M.; Looms, M. C.; Bogena, H. R.; Desilets, D.; Zreda, M. G.; Sonnenborg, T. O.; Jensen, K. H.

2014-12-01

The water stored in the various compartments of the terrestrial ecosystem (in snow, canopy interception, soil and litter) controls the exchange of the water and energy between the land surface and the atmosphere. Therefore, measurements of the water stored within these pools are critical for the prediction of e.g. evapotranspiration and groundwater recharge. The detection of cosmic-ray neutron intensity is a novel non-invasive method for the quantification of continuous intermediate-scale soil moisture. The footprint of the cosmic-ray neutron probe is a hemisphere of a few hectometers and subsurface depths of 10-70 cm depending on wetness. The cosmic-ray neutron method offers measurements at a scale between the point-scale measurements and large-scale satellite retrievals. The cosmic-ray neutron intensity is inversely correlated to the hydrogen stored within the footprint. Overall soil moisture represents the largest pool of hydrogen and changes in the soil moisture clearly affect the cosmic-ray neutron signal. However, the neutron intensity is also sensitive to variations of hydrogen in snow, canopy interception and biomass offering the potential to determine water content in such pools from the signal. In this study we tested the potential of determining canopy interception and biomass using cosmic-ray neutron intensity measurements within the framework of the Danish Hydrologic Observatory (HOBE) and the Terrestrial Environmental Observatories (TERENO). Continuous measurements at the ground and the canopy level, along with profile measurements were conducted at towers at forest field sites. Field experiments, including shielding the cosmic-ray neutron probes with cadmium foil (to remove lower-energy neutrons) and measuring reference intensity rates at complete water saturated conditions (on the sea close to the HOBE site), were further conducted to obtain an increased understanding of the physics controlling the cosmic-ray neutron transport and the equipment used

Research on new methods in transport theory

International Nuclear Information System (INIS)

Stefanovicj, D.

1975-01-01

Neutron transport theory is the basis for development of reactor theory and reactor calculational methods. It has to be acknowledged that recent applications of these disciplines have influenced considerably the development of power reactor concepts and technology. However, these achievements were implemented in a rather heuristic way, since the satisfaction of design demands were of utmost importance. Often this kind of approach turns out to be very restrictive and not even adequate for rather typical reactor applications. Many aspects and techniques of reactor theory and calculations ought to be reevaluated and/or reformulated on the more sound physical and mathematical foundations. At the same time, new reactor concepts and operational demands give rise to more sophisticated and complex design requirements. These new requirements can be met only by the development of new design techniques, which in the case of reactor neutronic calculation lead directly to the advanced transport theory methods. In addition, the rapid development of computer technology opens new opportunities for applications of advanced transport theory in practical calculations

Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

International Nuclear Information System (INIS)

Feghhi, S. A. H.; Afarideh, H.; Shahriari, M.

2007-01-01

The purpose of the present work is to develop an efficient solution method to calculate neutron importance function in fissionable assemblies for all criticality conditions, using Monte Carlo Method. The neutron importance function has a well important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating adjoint flux through out solving the Adjoint weighted transport equation with deterministic methods. However, in complex geometries these calculations are very difficult. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on physical concept of neutron importance has been introduced for calculating neutron importance function in sub-critical, critical and supercritical conditions. For this means a computer program has been developed. The results of the method has been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries and their correctness has been approved for all three criticality conditions. Ultimately, the efficiency of the method for complex geometries has been shown by calculation of neutron importance in MNSR research reactor

CMADR acceleration and its convergence analysis of the method of characteristics for neutron transport calculation

International Nuclear Information System (INIS)

Young, Ryong Park; Nam, Zin Cho

2005-01-01

As the nuclear reactor core becomes more complex, heterogeneous, and geometrically irregular, the method of characteristics (MOC) is gaining its wide use in the neutron transport calculations. However, the long computing times require good acceleration methods. In this paper, the concept of coarse-mesh angular dependent re-balance (CMADR) acceleration is described and applied to the MOC calculation in x-y-z (z-infinite, uniform) geometry. The method is based on the angular dependent re-balance factors defined only on the coarse-mesh boundaries; a coarse-mesh consists of several fine meshes that may be heterogeneous and of mixed geometries with irregular or unstructured mesh shapes. In addition, the coarse-mesh boundaries may not coincide with the structural interfaces of the problem and can be chosen artificially for convenience. CMADR acceleration is tested on several test problems and the results show that CMADR is very effective in reducing the number of iterations and computing times of MOC calculations. Fourier analysis is also provided to investigate convergence of the CMADR method analytically and the results show that CMADR acceleration is unconditionally stable. (authors)

Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

International Nuclear Information System (INIS)

Lee, Gil Soo

2006-02-01

To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

SUSD, Sensitivity and Uncertainty in Neutron Transport and Detector Response

International Nuclear Information System (INIS)

Furuta, Lazuo; Kondo, Shunsuke; Oka, Yoshika

1991-01-01

1 - Description of program or function: SUSD calculates sensitivity coefficients for one and two-dimensional transport problems. Variance and standard deviation of detector responses or design parameters can be obtained using cross-section covariance matrices. In neutron transport problems, this code is able to perform sensitivity-uncertainty analysis for secondary angular distribution (SAD) or secondary energy distribution (SED). 2 - Method of solution: The first-order perturbation theory is used to obtain sensitivity coefficients. The method described in the distributed report is employed to consider SAD/SED effect. 3 - Restrictions on the complexity of the problem: Variable dimension is used so that there is no limitation in each array size but the total core size

Neutron radiography using a transportable superconducting cyclotron

Energy Technology Data Exchange (ETDEWEB)

Allen, D.A. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Hawkesworth, M.R. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Beynon, T.D. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Green, S. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Rogers, J.D. (Rolls-Royce, Derby (United Kingdom)); Allen, M.J. (Rolls-Royce, Derby (United Kingdom)); Plummer, H.C. (Rolls-Royce, MatEval, Derby (United Kingdom)); Boulding, N.J. (Oxford Instruments (United Kingdom)); Cox, M. (Oxford Instruments (United Kingdom)); McDougall, I. (Oxford Instruments (United Kingdom))

1994-12-30

A thermal neutron radiography system based on a compact 12 MeV superconducting proton cyclotron is described. Neutrons are generated using a thick beryllium target and moderated in high density polyethylene. Monte Carlo computer simulations have been used to model the neutron and photon transport in order to optimise the performance of the system. With proton beam currents in excess of 100 [mu]A, it can provide high thermal neutron fluxes with L/D ratios of between 50 and 300 for various applications. Both film and electronic imaging are used to produce radiographs. The electronic imaging system consists of a [sup 6]Li-loaded ZnS intensifier screen, and a low light CCD or SIT camera. High resolution images can be recorded and computer-controlled data processing, analysis and display are possible. ((orig.))

Performance of the discrete ordinates method-like neutron transport computation with equivalent group condensation and angle-collapsing

International Nuclear Information System (INIS)

Yoo, Han Jong; Won, Jong Hyuck; Cho, Nam Zin

2011-01-01

In computational studies of neutron transport equations, the fine-group to few-group condensation procedure leads to equivalent total cross section that becomes angle dependent. The difficulty of this angle dependency has been traditionally treated by consistent P or extended transport approximation in the literature. In a previous study, we retained the angle dependency of the total cross section and applied directly to the discrete ordinates equation, with additional concept of angle-collapsing, and tested in a one-dimensional slab problem. In this study, we provide further results of this discrete ordinates-like method in comparison with the typical traditional methods. In addition, IRAM acceleration (based on Krylov subspace method) is tested for the purpose of further reducing the computational burden of few-group calculation. From the test results, it is ascertained that the angle-dependent total cross section with angle-collapsing gives excellent estimation of k_e_f_f and flux distribution and that IRAM acceleration effectively reduces the number of outer iterations. However, since IRAM requires sufficient convergence in inner iterations, speedup in total computer time is not significant for problems with upscattering. (author)

Numerical solution of the Neutron Transport Equation using discontinuous nodal methods at X-Y geometry; Solucion numerica de la ecuacion de transporte de neutrones usando metodos nodales discontinuos en geometria X-Y

Energy Technology Data Exchange (ETDEWEB)

Delfin L, A

1997-12-31

The purpose of this work is to solve the neutron transport equation in discrete-ordinates and X-Y geometry by developing and using the strong discontinuous and strong modified discontinuous nodal finite element schemes. The strong discontinuous and modified strong discontinuous nodal finite element schemes go from two to ten interpolation parameters per cell. They are describing giving a set D{sub c} and polynomial space S{sub c} corresponding for each scheme BDMO, RTO, BL, BDM1, HdV, BDFM1, RT1, BQ and BDM2. The solution is obtained solving the neutron transport equation moments for each nodal scheme by developing the basis functions defined by Pascal triangle and the Legendre moments giving in the polynomial space S{sub c} and, finally, looking for the non singularity of the resulting linear system. The linear system is numerically solved using a computer program for each scheme mentioned . It uses the LU method and forward and backward substitution and makes a partition of the domain in cells. The source terms and angular flux are calculated, using the directions and weights associated to the S{sub N} approximation and solving the angular flux moments to find the effective multiplication constant. The programs are written in Fortran language, using the dynamic allocation of memory to increase efficiently the available memory of the computing equipment. (Author).

A DETERMINISTIC METHOD FOR TRANSIENT, THREE-DIMENSIONAL NUETRON TRANSPORT

International Nuclear Information System (INIS)

S. GOLUOGLU, C. BENTLEY, R. DEMEGLIO, M. DUNN, K. NORTON, R. PEVEY I.SUSLOV AND H.L. DODDS

1998-01-01

A deterministic method for solving the time-dependent, three-dimensional Boltzmam transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement can also be modeled. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multidimensional neutronic systems

Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

International Nuclear Information System (INIS)

Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.

2007-01-01

The purpose of the present work is to develop an efficient solution method for the calculation of neutron importance function in fissionable assemblies for all criticality conditions, based on Monte Carlo calculations. The neutron importance function has an important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating the adjoint flux while solving the adjoint weighted transport equation based on deterministic methods. However, in complex geometries these calculations are very complicated. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on the physical concept of neutron importance has been introduced for calculating the neutron importance function in sub-critical, critical and super-critical conditions. For this propose a computer program has been developed. The results of the method have been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries. The correctness of these results has been confirmed for all three criticality conditions. Finally, the efficiency of the method for complex geometries has been shown by the calculation of neutron importance in Miniature Neutron Source Reactor (MNSR) research reactor

Biomedical applications of two- and three-dimensional deterministic radiation transport methods

International Nuclear Information System (INIS)

Nigg, D.W.

1992-01-01

Multidimensional deterministic radiation transport methods are routinely used in support of the Boron Neutron Capture Therapy (BNCT) Program at the Idaho National Engineering Laboratory (INEL). Typical applications of two-dimensional discrete-ordinates methods include neutron filter design, as well as phantom dosimetry. The epithermal-neutron filter for BNCT that is currently available at the Brookhaven Medical Research Reactor (BMRR) was designed using such methods. Good agreement between calculated and measured neutron fluxes was observed for this filter. Three-dimensional discrete-ordinates calculations are used routinely for dose-distribution calculations in three-dimensional phantoms placed in the BMRR beam, as well as for treatment planning verification for live canine subjects. Again, good agreement between calculated and measured neutron fluxes and dose levels is obtained

Range calculations using multigroup transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

1979-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

A spin-transport system for a longitudinally polarized epithermal neutron beam

International Nuclear Information System (INIS)

Crawford, B.E.; Bowman, J.D.; Penttilae, S.I.; Roberson, N.R.

2001-01-01

The TRIPLE (Time Reversal and Parity at Low Energies) collaboration uses a polarized epithermal neutron beam and a capture γ-ray detector to study parity violation in neutron-nucleus reactions. In order to preserve the spin polarization of the neutrons as they travel the 60-m path to the target, the beam pipes are wrapped with wire to produce a solenoidal magnetic field of about 10 G along the beam direction. The flanges and bellows between sections of the beam pipe cause gaps in the windings which in turn produce radial fields that can depolarize the neutron spins. A computer code has been developed that numerically evaluates the effect of these gaps on the polarization. A measurement of the neutron depolarization for neutrons in the actual spin-transport system agrees with a calculation of the neutron depolarization for the TRIPLE system. Features that will aid in designing similar spin-transport systems are discussed

Determination of the mean free path of the thermal neutrons transport by the measure of a complex diffusion length; Determination du libre parcours moyen de transport des neutrons thermiques par la mesure d'une longueur de diffusion complexe

Energy Technology Data Exchange (ETDEWEB)

Raievski, V; Horowitz, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The further method is the outcome of a technique used in the study of neutrons in scattering and slowing-down environment. In this technique, we replace the constant sources used in the classic experiences by modulated sources with a variable frequency. The object of this article is to describe the extension of the method for the mean free path for transport of thermal neutrons and also to indicate the possible applications for other sizes, as the slowing length, or the absolute value of the cross-section of the boron. (M.B.) [French] La methode qui va etre decrite est l'aboutissement d'une technique utilisee dans l'etude des milieux ou diffusent et se ralentissent des neutrons. Dans cette technique, on remplace les sources constantes utilisees dans les experiences classiques par des sources modulees, a frequence variable. L'objet de cet article est de decrire l'extension de la methode a la mesure du libre parcours moyen de transport des neutrons thermiques et egalement d'indiquer les applications possibles a la mesure d'autres grandeurs, telles que la longueur de ralentissement, ou la valeur absolue de la section de capture du bore. (M.B.)

Spherical harmonics solutions of multi-dimensional neutron transport equation by finite Fourier transformation

International Nuclear Information System (INIS)

Kobayashi, Keisuke

1977-01-01

A method of solution of a monoenergetic neutron transport equation in P sub(L) approximation is presented for x-y and x-y-z geometries using the finite Fourier transformation. A reactor system is assumed to consist of multiregions in each of which the nuclear cross sections are spatially constant. Since the unknown functions of this method are the spherical harmonics components of the neutron angular flux at the material boundaries alone, the three- and two-dimensional equations are reduced to two- and one-dimensional equations, respectively. The present approach therefore gives fewer unknowns than in the usual series expansion method or in the finite difference method. Some numerical examples are shown for the criticality problem. (auth.)

ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

International Nuclear Information System (INIS)

1985-01-01

1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

Domain decomposition methods for the neutron diffusion problem

International Nuclear Information System (INIS)

Guerin, P.; Baudron, A. M.; Lautard, J. J.

2010-01-01

The neutronic simulation of a nuclear reactor core is performed using the neutron transport equation, and leads to an eigenvalue problem in the steady-state case. Among the deterministic resolution methods, simplified transport (SPN) or diffusion approximations are often used. The MINOS solver developed at CEA Saclay uses a mixed dual finite element method for the resolution of these problems. and has shown his efficiency. In order to take into account the heterogeneities of the geometry, a very fine mesh is generally required, and leads to expensive calculations for industrial applications. In order to take advantage of parallel computers, and to reduce the computing time and the local memory requirement, we propose here two domain decomposition methods based on the MINOS solver. The first approach is a component mode synthesis method on overlapping sub-domains: several Eigenmodes solutions of a local problem on each sub-domain are taken as basis functions used for the resolution of the global problem on the whole domain. The second approach is an iterative method based on a non-overlapping domain decomposition with Robin interface conditions. At each iteration, we solve the problem on each sub-domain with the interface conditions given by the solutions on the adjacent sub-domains estimated at the previous iteration. Numerical results on parallel computers are presented for the diffusion model on realistic 2D and 3D cores. (authors)

Monte Carlo methods for neutron transport on graphics processing units using Cuda - 015

International Nuclear Information System (INIS)

Nelson, A.G.; Ivanov, K.N.

2010-01-01

This work examined the feasibility of utilizing Graphics Processing Units (GPUs) to accelerate Monte Carlo neutron transport simulations. First, a clean-sheet MC code was written in C++ for an x86 CPU and later ported to run on GPUs using NVIDIA's CUDA programming language. After further optimization, the GPU ran 21 times faster than the CPU code when using single-precision floating point math. This can be further increased with no additional effort if accuracy is sacrificed for speed: using a compiler flag, the speedup was increased to 22x. Further, if double-precision floating point math is desired for neutron tracking through the geometry, a speedup of 11x was obtained. The GPUs have proven to be useful in this study, but the current generation does have limitations: the maximum memory currently available on a single GPU is only 4 GB; the GPU RAM does not provide error-checking and correction; and the optimization required for large speedups can lead to confusing code. (authors)

Finite element based composite solution for neutron transport problems

International Nuclear Information System (INIS)

Mirza, A.N.; Mirza, N.M.

1995-01-01

A finite element treatment for solving neutron transport problems is presented. The employs region-wise discontinuous finite elements for the spatial representation of the neutron angular flux, while spherical harmonics are used for directional dependence. Composite solutions has been obtained by using different orders of angular approximations in different parts of a system. The method has been successfully implemented for one dimensional slab and two dimensional rectangular geometry problems. An overall reduction in the number of nodal coefficients (more than 60% in some cases as compared to conventional schemes) has been achieved without loss of accuracy with better utilization of computational resources. The method also provides an efficient way of handling physically difficult situations such as treatment of voids in duct problems and sharply changing angular flux. It is observed that a great wealth of information about the spatial and directional dependence of the angular flux is obtained much more quickly as compared to Monte Carlo method, where most of the information in restricted to the locality of immediate interest. (author)

Arbitrary quadrature: its application in the solution of one-dimensional, planar neutron transport problems

International Nuclear Information System (INIS)

Sanchez, J.

2010-10-01

A standard numerical procedure for the solution of singular integral equations is applied to the one-dimensional transport equation for monoenergetic neutrons. As is usual in quadrature methods, the procedure yields an Eigen system whose solution provide, for the critical slab, both the eigenvalue which is proportional to the number of secondary neutrons per collision, and the density as a function of position. The results obtained with two versions of the procedure, differing only in the extent of the basic region to which they are applied, are compared with analytically derived results available for benchmarking. The procedures considered yield consistent results for the calculated neutron densities and eigenvalues. Since the one-dimensional transport kernel and its spatial moments are integrable and their integrals can be put in terms of exponential integral functions, the resulting approximations to the neutron density yield somewhat lengthy but closed, forms. These approximate expressions of the neutron density can be used to render, after they are operated on, closed-form formulas for build-up factors, extrapolation distances or angular densities or employed for other purposes that require an analytical expression of the neutron density. As an example of this latter capability, the results of the calculation of the angular density at the surface of the slab are provided. (Author)

An application of reactor noise techniques to neutron transport problems in a random medium

International Nuclear Information System (INIS)

Sahni, D.C.

1989-01-01

Neutron transport problems in a random medium are considered by defining a joint Markov process describing the fluctuations of one neutron population and the random changes in the medium. Backward Chapman-Kolmogorov equations are derived which yield an adjoint transport equation for the average neutron density. It is shown that this average density also satisfied the direct transport equation as given by the phenomenological model. (author)

A new Monte Carlo method for neutron noise calculations in the frequency domain

International Nuclear Information System (INIS)

Rouchon, Amélie; Zoia, Andrea; Sanchez, Richard

2017-01-01

Neutron noise equations, which are obtained by assuming small perturbations of macroscopic cross sections around a steady-state neutron field and by subsequently taking the Fourier transform in the frequency domain, have been usually solved by analytical techniques or by resorting to diffusion theory. A stochastic approach has been recently proposed in the literature by using particles with complex-valued weights and by applying a weight cancellation technique. We develop a new Monte Carlo algorithm that solves the transport neutron noise equations in the frequency domain. The stochastic method presented here relies on a modified collision operator and does not need any weight cancellation technique. In this paper, both Monte Carlo methods are compared with deterministic methods (diffusion in a slab geometry and transport in a simplified rod model) for several noise frequencies and for isotropic and anisotropic noise sources. Our stochastic method shows better performances in the frequency region of interest and is easier to implement because it relies upon the conventional algorithm for fixed-source problems.

A New Monte Carlo Neutron Transport Code at UNIST

International Nuclear Information System (INIS)

Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung

2014-01-01

Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results

Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1, 1974--August 31, 1975

International Nuclear Information System (INIS)

Bareiss, E.H.

1975-01-01

The objectives of the research remain the same as outlined in the original proposal. They are in short as follows: Develop mathematically and computationally founded criteria for the design of highly efficient and reliable multi-dimensional neutron transport codes to solve a variety of neutron migration and radiation problems and analyze existing and new methods for performance. (U.S.)

Generalized diffusion theory for calculating the neutron transport scalar flux

International Nuclear Information System (INIS)

Alcouffe, R.E.

1975-01-01

A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)

Study of a method to solve the one speed, three dimensional transport equation using the finite element method and the associated Legendre function

International Nuclear Information System (INIS)

Fernandes, A.

1991-01-01

A method to solve three dimensional neutron transport equation and it is based on the original work suggested by J.K. Fletcher (42, 43). The angular dependence of the flux is approximated by associated Legendre functions and the finite element method is applied to the space components is presented. When the angular flux, the scattering cross section and the neutrons source are expanded in associated Legendre functions, the first order neutron transport equation is reduced to a coupled set of second order diffusion like equations. These equations are solved in an iterative way by the finite element method to the moments. (author)

Design studies for a high-resolution, transportable neutron radiography/radioscopy system

International Nuclear Information System (INIS)

Gillespie, G.H.; Micklich, B.J.; McMichael, G.E.

1996-01-01

A preliminary design has been developed for a high-resolution, transportable neutron radiology system (TNRS) concept. The primary system requirement is taken to be a thermal neutron flux of 10[sup 6] n/(cm[sup 2]-sec) with a L/D ratio of 100. The approach is to use an accelerator-driven neutron source, with a radiofrequency quadrupole (RFQ) as the primary accelerator component. Initial concepts for all of the major components of the system have been developed,and selected key parts have been examined further. An overview of the system design is presented, together with brief summaries of the concepts for the ion source, low energy beam transport (LEBT), RFQ, high energy beam transport (HEBT), target, moderator, collimator, image collection, power, cooling, vacuum, structure, robotics, control system, data analysis, transport vehicle, and site support. The use of trade studies for optimizing the TNRS concept are also described

Parameter estimation for hydrogen analysis by using transport method

International Nuclear Information System (INIS)

Selvi, S.; Can, N.

1992-01-01

A transport method is described which reduces greatly the number of calibration standards needed for hydrogen analysis by neutron induced prompt γ-rays. The counts in the photopeaks from neutron capture in hydrogen for various standard concentrations, the distribution of the source neutron rate entering the thermal group and the reaction rates in the samples are investigated theoretically using 100 energy group cross sections and experimental 252 Cf spectra for a test configuration. Comparison of theoretical results with those measured from the test configuration shows good agreement. (author)

Homogenization of the critically spectral equation in neutron transport

International Nuclear Information System (INIS)

Allaire, G.; Paris-6 Univ., 75; Bal, G.

1998-01-01

We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)

Homogenization of the critically spectral equation in neutron transport

Energy Technology Data Exchange (ETDEWEB)

Allaire, G. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie]|[Paris-6 Univ., 75 (France). Lab. d' Analyse Numerique; Bal, G. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches

1998-07-01

We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)

Solution of neutron transport equation using Daubechies' wavelet expansion in the angular discretization

International Nuclear Information System (INIS)

Cao Liangzhi; Wu Hongchun; Zheng Youqi

2008-01-01

Daubechies' wavelet expansion is introduced to discretize the angular variables of the neutron transport equation when the neutron angular flux varies very acutely with the angular directions. An improvement is made by coupling one-dimensional wavelet expansion and discrete ordinate method to make two-dimensional angular discretization efficient and stable. The angular domain is divided into several subdomains for treating the vacuum boundary condition exactly in the unstructured geometry. A set of wavelet equations coupled with each other is obtained in each subdomain. An iterative method is utilized to decouple the wavelet moments. The numerical results of several benchmark problems demonstrate that the wavelet expansion method can provide more accurate results by lower-order expansion than other angular discretization methods

Analysis and development of spatial hp-refinement methods for solving the neutron transport equation; Analyse et developpement de methodes de raffinement hp en espace pour l'equation de transport des neutrons

Energy Technology Data Exchange (ETDEWEB)

Fournier, D.

2011-10-10

The different neutronic parameters have to be calculated with a higher accuracy in order to design the 4. generation reactor cores. As memory storage and computation time are limited, adaptive methods are a solution to solve the neutron transport equation. The neutronic flux, solution of this equation, depends on the energy, angle and space. The different variables are successively discretized. The energy with a multigroup approach, considering the different quantities to be constant on each group, the angle by a collocation method called SN approximation. Once the energy and angle variable are discretized, a system of spatially-dependent hyperbolic equations has to be solved. Discontinuous finite elements are used to make possible the development of hp-refinement methods. Thus, the accuracy of the solution can be improved by spatial refinement (h-refinement), consisting into subdividing a cell into sub-cells, or by order refinement (p-refinement), by increasing the order of the polynomial basis. In this thesis, the properties of this methods are analyzed showing the importance of the regularity of the solution to choose the type of refinement. Thus, two error estimators are used to lead the refinement process. Whereas the first one requires high regularity hypothesis (analytical solution), the second one supposes only the minimal hypothesis required for the solution to exist. The comparison of both estimators is done on benchmarks where the analytic solution is known by the method of manufactured solutions. Thus, the behaviour of the solution as a regard of the regularity can be studied. It leads to a hp-refinement method using the two estimators. Then, a comparison is done with other existing methods on simplified but also realistic benchmarks coming from nuclear cores. These adaptive methods considerably reduces the computational cost and memory footprint. To further improve these two points, an approach with energy-dependent meshes is proposed. Actually, as the

Methodology of Continuous-Energy Adjoint Monte Carlo for Neutron, Photon, and Coupled Neutron-Photon Transport

International Nuclear Information System (INIS)

Hoogenboom, J. Eduard

2003-01-01

Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems

Some results on the neutron transport and the coupling of equations

International Nuclear Information System (INIS)

Bal, G.

1997-01-01

Neutron transport in nuclear reactors is well modeled by the linear Boltzmann transport equation. Its resolution is relatively easy but very expensive. To achieve whole core calculations, one has to consider simpler models, such as diffusion or homogeneous transport equations. However, the solutions may become inaccurate in particular situations (as accidents for instance). That is the reason why we wish to solve the equations on small area accurately and more coarsely on the remaining part of the core. It is than necessary to introduce some links between different discretizations or modelizations. In this note, we give some results on the coupling of different discretizations of all degrees of freedom of the integral-differential neutron transport equation (two degrees for the angular variable, on for the energy component, and two or three degrees for spatial position respectively in 2D (cylindrical symmetry) and 3D). Two chapters are devoted to the coupling of discrete ordinates methods (for angular discretization). The first one is theoretical and shows the well posing of the coupled problem, whereas the second one deals with numerical applications of practical interest (the results have been obtained from the neutron transport code developed at the R and D, which has been modified for introducing the coupling). Next, we present the nodal scheme RTN0, used for the spatial discretization. We show well posing results for the non-coupled and the coupled problems. At the end, we deal with the coupling of energy discretizations for the multigroup equations obtained by homogenization. Some theoretical results of the discretization of the velocity variable (well-posing of problems), which do not deal directly with the purposes of coupling, are presented in the annexes. (author)

Neutron imaging integrated circuit and method for detecting neutrons

Science.gov (United States)

Nagarkar, Vivek V.; More, Mitali J.

2017-12-05

The present disclosure provides a neutron imaging detector and a method for detecting neutrons. In one example, a method includes providing a neutron imaging detector including plurality of memory cells and a conversion layer on the memory cells, setting one or more of the memory cells to a first charge state, positioning the neutron imaging detector in a neutron environment for a predetermined time period, and reading a state change at one of the memory cells, and measuring a charge state change at one of the plurality of memory cells from the first charge state to a second charge state less than the first charge state, where the charge state change indicates detection of neutrons at said one of the memory cells.

Improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell

International Nuclear Information System (INIS)

Bosevski, T.

1986-01-01

An improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell has been developed. Expanding the neutron flux and source into a series of even powers of the radius, one' gets a convenient method for integration of the one-energy group integral transport equation. It is shown that it is possible to perform an analytical integration in the x-y plane in one variable and to use the effective Gaussian integration over another one. Choosing a convenient distribution of space points in fuel and moderator the transport matrix calculation and cell reaction rate integration were condensed. On the basis of the proposed method, the computer program DISKRET for the ZUSE-Z 23 K computer has been written. The suitability of the proposed method for the calculation of the thermal-neutron-flux distribution in a reactor cell can be seen from the test results obtained. Compared with the other collision probability methods, the proposed treatment excels with a mathematical simplicity and a faster convergence. (author)

Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

Energy Technology Data Exchange (ETDEWEB)

Modak, R S; Kumar, Vinod; Menon, S V.G. [Theoretical Physics Div., Bhabha Atomic Research Centre, Mumbai (India); Gupta, Anurag [Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai (India)

2005-09-15

The numerical solution of linear multi-energy-group neutron transport equation is required in several analyses in nuclear reactor physics and allied areas. Computer codes based on the discrete ordinates (Sn) method are commonly used for this purpose. These codes solve external source problem and K-eigenvalue problem. The overall solution technique involves solution of source problem in each energy group as intermediate procedures. Such a single-group source problem is solved by the so-called Source Iteration (SI) method. As is well-known, the SI-method converges very slowly for optically thick and highly scattering regions, leading to large CPU times. Over last three decades, many schemes have been tried to accelerate the SI; the most prominent being the Diffusion Synthetic Acceleration (DSA) scheme. The DSA scheme, however, often fails and is also rather difficult to implement. In view of this, in 1997, Ramone and others have developed a new acceleration scheme called Transport Synthetic Acceleration (TSA) which is much more robust and easy to implement. This scheme has been recently incorporated in 2-D and 3-D in-house codes at BARC. This report presents studies on the utility of TSA scheme for fairly general test problems involving many energy groups and anisotropic scattering. The scheme is found to be useful for problems in Cartesian as well as Cylindrical geometry. (author)

Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

International Nuclear Information System (INIS)

Modak, R.S.; Vinod Kumar; Menon, S.V.G.; Gupta, Anurag

2005-09-01

The numerical solution of linear multi-energy-group neutron transport equation is required in several analyses in nuclear reactor physics and allied areas. Computer codes based on the discrete ordinates (Sn) method are commonly used for this purpose. These codes solve external source problem and K-eigenvalue problem. The overall solution technique involves solution of source problem in each energy group as intermediate procedures. Such a single-group source problem is solved by the so-called Source Iteration (SI) method. As is well-known, the SI-method converges very slowly for optically thick and highly scattering regions, leading to large CPU times. Over last three decades, many schemes have been tried to accelerate the SI; the most prominent being the Diffusion Synthetic Acceleration (DSA) scheme. The DSA scheme, however, often fails and is also rather difficult to implement. In view of this, in 1997, Ramone and others have developed a new acceleration scheme called Transport Synthetic Acceleration (TSA) which is much more robust and easy to implement. This scheme has been recently incorporated in 2-D and 3-D in-house codes at BARC. This report presents studies on the utility of TSA scheme for fairly general test problems involving many energy groups and anisotropic scattering. The scheme is found to be useful for problems in Cartesian as well as Cylindrical geometry. (author)

Physical qualification and improvements of the numerical model of a method of characteristics for the resolution of the neutron transport equation in non-structured grids

International Nuclear Information System (INIS)

Santandrea, Simone

2001-01-01

This research thesis addresses the resolution of the neutron transport equation inside reactor cells in non-structured grids and in general geometry by using the method of characteristics (MoC) and two acceleration methods developed during this research. The author introduces the MoC with a flat approximation of the neutron collision source within each computation area. This formulation leads to a linear approximation. The next part presents the mathematical framework for the use of the Lanczos iterative scheme. A new acceleration method is then introduced. The last part reports realistic cases with a high spatial and angular heterogeneity. Results obtained by using the Apollo2-TDT code are compared with those obtained with the Tripoli4 Monte-Carlo code [fr

Numerical study of the particle transport in fast neutron detectors with conversion layer

International Nuclear Information System (INIS)

Sedlackova, K.; Zatko, B.; Necas, V.

2012-01-01

This paper deals with fast neutron and recoil proton transport simulation using statistical analysis of Monte Carlo radiation transport code (MCNPX). Its possibilities in the detector design and optimization are presented. MCNPX proved as a very advantageous self-contained simulation program for fast neutron and secondary proton tracking. Simulations of respective particle transport through conversion layer of HDPE and further in the active volume of detector let us to follow important characteristics as neutron/proton flux density, reaction rate of elastic scattering on hydrogen nuclei and deposited energy as well as their dependencies on incident neutron energy and conversion layer/active region thickness. The efficiency of neutrons to protons conversion has been calculated and its maximum was reached for 500 μm thick conversion layer. The minimum active region thickness has been estimated to be about 300 μm.(authors)

On the reciprocity-like relations in linear neutron transport theory

International Nuclear Information System (INIS)

Modak, R.S.; Sahni, D.C.

1997-01-01

The existence of certain reciprocity-like relations in neutron transport theory was shown earlier under some quite restrictive conditions. Here, these relations are shown to be valid in more general situations by using a different approach based on individual neutron trajectories. (author)

Reevaluation of the case, de Hoffman, and Placzek one-group neutron transport benchmark solution in plane geometry

International Nuclear Information System (INIS)

Ganapol, B.D.

1986-01-01

In a course on neutron transport theory and also in the analytical neutron transport theory literature, the pioneering work of Case et al. (CdHP) is often referenced. This work was truly a monumental effort in that it treated the fundamental mathematical properties of the one-group neutron Boltzmann equation in detail as well as the numerical evaluation of most of the resulting solutions. Many mathematically and numerically oriented dissertations were based on this classic monograph. In light of the considerable advances made both in numerical methods and computer technology since 1953, when the historic CdHP monograph first appeared, it seems appropriate to reevaluate the numerical benchmark solutions found therein with present-day computational technology. In most transport theory courses, the subject of proper benchmarking of numerical algorithms and transport codes is seldom addressed at any great length. This may be the reason that the benchmarking procedure is so rarely practiced in the nuclear community and when practiced is improperly applied. In this presentation, the development of a new benchmark for the one-group neutron flux in an infinite medium will be detailed with emphasis placed on the educational aspects of the benchmarking activity

Method for the determination of the dominant eigenvalue of the neutron transport equation in a slab using fractional derivative

International Nuclear Information System (INIS)

Sperotto, Fabiola Aiub; Segatto, Cynthia Feijo; Zabadal, Jorge

2002-01-01

In this work, we determine the dominant eigenvalue of the one-dimensional neutron transport equation in a slab constructing an integral form for the neutron transport equation which is the expressed in terms of fractional derivative of the angular flux. Equating the fractional derivative of the angular flux to the integrate equation, we determine the unknown order of the fractional derivative comparing the kernel of the integral equation with the one of Riemann-Liouville definition of fractional derivative. Once known the angular flux the dominant eigenvalue is calculated solving a transcendental equation resulting from the application of the boundary conditions. We report the methodology applied, for comparison with available results in literature. (author)

MVP/GMVP version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

2017-03-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

An analytical approach for a nodal scheme of two-dimensional neutron transport problems

International Nuclear Information System (INIS)

Barichello, L.B.; Cabrera, L.C.; Prolo Filho, J.F.

2011-01-01

Research highlights: → Nodal equations for a two-dimensional neutron transport problem. → Analytical Discrete Ordinates Method. → Numerical results compared with the literature. - Abstract: In this work, a solution for a two-dimensional neutron transport problem, in cartesian geometry, is proposed, on the basis of nodal schemes. In this context, one-dimensional equations are generated by an integration process of the multidimensional problem. Here, the integration is performed for the whole domain such that no iterative procedure between nodes is needed. The ADO method is used to develop analytical discrete ordinates solution for the one-dimensional integrated equations, such that final solutions are analytical in terms of the spatial variables. The ADO approach along with a level symmetric quadrature scheme, lead to a significant order reduction of the associated eigenvalues problems. Relations between the averaged fluxes and the unknown fluxes at the boundary are introduced as the usually needed, in nodal schemes, auxiliary equations. Numerical results are presented and compared with test problems.

BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

International Nuclear Information System (INIS)

Fanaro, L.C.C.B.

1984-01-01

It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt

SITHA program for simulating hadron transport in the 100-1010eV energy range. Simulation of neutron transport with E6 eV

International Nuclear Information System (INIS)

Daniehl', A.V.; Dushin, V.N.

1987-01-01

The methods for simulation of neutron transport with Z<20 MeV used in the SITHA (simulation transport hadron) program, the original library of group microconstants (175 groups) with subgroup description of resonance range and a set of programs for its creation are described. The results of a number of integral experiments are discussed

The Random Ray Method for neutral particle transport

Energy Technology Data Exchange (ETDEWEB)

Tramm, John R., E-mail: jtramm@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science Engineering, 77 Massachusetts Avenue, 24-107, Cambridge, MA 02139 (United States); Argonne National Laboratory, Mathematics and Computer Science Department 9700 S Cass Ave, Argonne, IL 60439 (United States); Smith, Kord S., E-mail: kord@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science Engineering, 77 Massachusetts Avenue, 24-107, Cambridge, MA 02139 (United States); Forget, Benoit, E-mail: bforget@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science Engineering, 77 Massachusetts Avenue, 24-107, Cambridge, MA 02139 (United States); Siegel, Andrew R., E-mail: siegela@mcs.anl.gov [Argonne National Laboratory, Mathematics and Computer Science Department 9700 S Cass Ave, Argonne, IL 60439 (United States)

2017-08-01

A new approach to solving partial differential equations (PDEs) based on the method of characteristics (MOC) is presented. The Random Ray Method (TRRM) uses a stochastic rather than deterministic discretization of characteristic tracks to integrate the phase space of a problem. TRRM is potentially applicable in a number of transport simulation fields where long characteristic methods are used, such as neutron transport and gamma ray transport in reactor physics as well as radiative transfer in astrophysics. In this study, TRRM is developed and then tested on a series of exemplar reactor physics benchmark problems. The results show extreme improvements in memory efficiency compared to deterministic MOC methods, while also reducing algorithmic complexity, allowing for a sparser computational grid to be used while maintaining accuracy.

The Random Ray Method for neutral particle transport

International Nuclear Information System (INIS)

Tramm, John R.; Smith, Kord S.; Forget, Benoit; Siegel, Andrew R.

2017-01-01

A new approach to solving partial differential equations (PDEs) based on the method of characteristics (MOC) is presented. The Random Ray Method (TRRM) uses a stochastic rather than deterministic discretization of characteristic tracks to integrate the phase space of a problem. TRRM is potentially applicable in a number of transport simulation fields where long characteristic methods are used, such as neutron transport and gamma ray transport in reactor physics as well as radiative transfer in astrophysics. In this study, TRRM is developed and then tested on a series of exemplar reactor physics benchmark problems. The results show extreme improvements in memory efficiency compared to deterministic MOC methods, while also reducing algorithmic complexity, allowing for a sparser computational grid to be used while maintaining accuracy.

Neutron transport in Eulerian coordinates with bulk material motion

Energy Technology Data Exchange (ETDEWEB)

Baker, Randal S., E-mail: rsb@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Dahl, Jon A., E-mail: dahl@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Fichtl, Erin J., E-mail: efichtl@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Morel, Jim E., E-mail: morel@tamu.edu [Department of Nuclear Engineering, Texas A& M University, College Station, TX (United States)

2015-12-15

A consistent, numerically stable algorithm for the solution of the neutron transport equation in the presence of a moving material background is presented for one-dimensional spherical geometry. Manufactured solutions are used to demonstrate the correctness and stability of our numerical algorithm. The importance of including moving material corrections is shown for the r-process in proto-neutron stars.

Finite moments approach to the time-dependent neutron transport equation

International Nuclear Information System (INIS)

Kim, Sang Hyun

1994-02-01

Currently, nodal techniques are widely used in solving the multidimensional diffusion equation because of savings in computing time and storage. Thanks to the development of computer technology, one can now solve the transport equation instead of the diffusion equation to obtain more accurate solution. The finite moments method, one of the nodal methods, attempts to represent the fluxes in the cell and on cell surfaces more rigorously by retaining additional spatial moments. Generally, there are two finite moments schemes to solve the time-dependent transport equation. In one, the time variable is treated implicitly with finite moments method in space variable (implicit finite moments method), the other method uses finite moments method in both space and time (space-time finite moments method). In this study, these two schemes are applied to two types of time-dependent neutron transport problems. One is a fixed source problem, the other a heterogeneous fast reactor problem with delayed neutrons. From the results, it is observed that the two finite moments methods give almost the same solutions in both benchmark problems. However, the space-time finite moments method requires a little longer computing time than that of the implicit finite moments method. In order to reduce the longer computing time in the space-time finite moments method, a new iteration strategy is exploited, where a few time-stepwise calculation, in which original time steps are grouped into several coarse time divisions, is performed sequentially instead of performing iterations over the entire time steps. This strategy results in significant reduction of the computing time and we observe that 2-or 3-stepwise calculation is preferable. In addition, we propose a new finite moments method which is called mixed finite moments method in this thesis. Asymptotic analysis for the finite moments method shows that accuracy of the solution in a heterogeneous problem mainly depends on the accuracy of the

A Wavelet-Based Finite Element Method for the Self-Shielding Issue in Neutron Transport

International Nuclear Information System (INIS)

Le Tellier, R.; Fournier, D.; Ruggieri, J. M.

2009-01-01

This paper describes a new approach for treating the energy variable of the neutron transport equation in the resolved resonance energy range. The aim is to avoid recourse to a case-specific spatially dependent self-shielding calculation when considering a broad group structure. This method consists of a discontinuous Galerkin discretization of the energy using wavelet-based elements. A Σ t -orthogonalization of the element basis is presented in order to make the approach tractable for spatially dependent problems. First numerical tests of this method are carried out in a limited framework under the Livolant-Jeanpierre hypotheses in an infinite homogeneous medium. They are mainly focused on the way to construct the wavelet-based element basis. Indeed, the prior selection of these wavelet functions by a thresholding strategy applied to the discrete wavelet transform of a given quantity is a key issue for the convergence rate of the method. The Canuto thresholding approach applied to an approximate flux is found to yield a nearly optimal convergence in many cases. In these tests, the capability of such a finite element discretization to represent the flux depression in a resonant region is demonstrated; a relative accuracy of 10 -3 on the flux (in L 2 -norm) is reached with less than 100 wavelet coefficients per group. (authors)

Progress toward the development and testing of source reconstruction methods for NIF neutron imaging.

Science.gov (United States)

Loomis, E N; Grim, G P; Wilde, C; Wilson, D C; Morgan, G; Wilke, M; Tregillis, I; Merrill, F; Clark, D; Finch, J; Fittinghoff, D; Bower, D

2010-10-01

Development of analysis techniques for neutron imaging at the National Ignition Facility is an important and difficult task for the detailed understanding of high-neutron yield inertial confinement fusion implosions. Once developed, these methods must provide accurate images of the hot and cold fuels so that information about the implosion, such as symmetry and areal density, can be extracted. One method under development involves the numerical inversion of the pinhole image using knowledge of neutron transport through the pinhole aperture from Monte Carlo simulations. In this article we present results of source reconstructions based on simulated images that test the methods effectiveness with regard to pinhole misalignment.

Adjacent-cell Preconditioners for solving optically thick neutron transport problems

International Nuclear Information System (INIS)

Azmy, Y.Y.

1994-01-01

We develop, analyze, and test a new acceleration scheme for neutron transport methods, the Adjacent-cell Preconditioner (AP) that is particularly suited for solving optically thick problems. Our method goes beyond Diffusion Synthetic Acceleration (DSA) methods in that it's spectral radius vanishes with increasing cell thickness. In particular, for the ID case the AP method converges immediately, i.e. in one iteration, to 10 -4 pointwise relative criterion in problems with dominant cell size of 10 mfp or thicker. Also the AP has a simple formalism and is cell-centered hence, multidimensional and high order extensions are easier to develop, and more efficient to implement

Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory

International Nuclear Information System (INIS)

2008-01-01

The developers of computer codes involving neutron transport theory for nuclear engineering applications seldom apply analytical benchmarking strategies to ensure the quality of their programs. A major reason for this is the lack of analytical benchmarks and their documentation in the literature. The few such benchmarks that do exist are difficult to locate, as they are scattered throughout the neutron transport and radiative transfer literature. The motivation for this benchmark compendium, therefore, is to gather several analytical benchmarks appropriate for nuclear engineering applications under one cover. We consider the following three subject areas: neutron slowing down and thermalization without spatial dependence, one-dimensional neutron transport in infinite and finite media, and multidimensional neutron transport in a half-space and an infinite medium. Each benchmark is briefly described, followed by a detailed derivation of the analytical solution representation. Finally, a demonstration of the evaluation of the solution representation includes qualified numerical benchmark results. All accompanying computer codes are suitable for the PC computational environment and can serve as educational tools for courses in nuclear engineering. While this benchmark compilation does not contain all possible benchmarks, by any means, it does include some of the most prominent ones and should serve as a valuable reference. (author)

MINARET: Towards a time-dependent neutron transport parallel solver

International Nuclear Information System (INIS)

Baudron, A.M.; Lautard, J.J.; Maday, Y.; Mula, O.

2013-01-01

We present the newly developed time-dependent 3D multigroup discrete ordinates neutron transport solver that has recently been implemented in the MINARET code. The solver is the support for a study about computing acceleration techniques that involve parallel architectures. In this work, we will focus on the parallelization of two of the variables involved in our equation: the angular directions and the time. This last variable has been parallelized by a (time) domain decomposition method called the para-real in time algorithm. (authors)

Progress on RMC: a Monte Carlo neutron transport code for reactor analysis

International Nuclear Information System (INIS)

Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin

2011-01-01

This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)

Effects of fuel particle size distributions on neutron transport in stochastic media

International Nuclear Information System (INIS)

Liang, Chao; Pavlou, Andrew T.; Ji, Wei

2014-01-01

Highlights: • Effects of fuel particle size distributions on neutron transport are evaluated. • Neutron channeling is identified as the fundamental reason for the effects. • The effects are noticeable at low packing and low optical thickness systems. • Unit cells of realistic reactor designs are studied for different size particles. • Fuel particle size distribution effects are not negligible in realistic designs. - Abstract: This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (k eff ) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron

KAMCCO, a reactor physics Monte Carlo neutron transport code

International Nuclear Information System (INIS)

Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.

1976-06-01

KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de

Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, W.; Aures, A.; Bernnat, W.; and others

2013-06-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

Discrete-ordinates electron transport calculations using standard neutron transport codes

International Nuclear Information System (INIS)

Morel, J.E.

1979-01-01

The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

A pulsed neutron Ramsey's method

Energy Technology Data Exchange (ETDEWEB)

Masuda, Y. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan)]. E-mail: yasuhiro.masuda@kek.jp; Ino, T. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Jeong, S.C. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Muto, S. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Skoy, V. [Joint Institute for Nuclear Reasearch, 141980 Dubna (Russian Federation); Watanabe, Y. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan)

2005-02-15

A Ramsey's method with pulsed neutrons is proposed. A Ramsey signal, which is a neutron spin rotation about a static magnetic field for a time interval between two separated oscillatory fields, is observed as a function of a neutron time of flight (TOF) in this method. The neutron spin rotation or the RF oscillation is used as a clock of the neutron velocity measurement which ranges from cold to epithermal neutron energies. This method together with the TOF measurement can be used for neutron inelastic scattering experiments. In addition, this method can be applied to the measurement of magnetic and pseudomagnetic fields in matter, and also to neutron spin manipulation for spin dependent scattering.

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

The importance of anisotropic scattering in high energy neutron transport problems

International Nuclear Information System (INIS)

Prillinger, G.; Mattes, M.

1984-01-01

To describe the highly anisotropic scattering of very fast neutrons adequately the transport code ANISN has been improved. Fokker-Planck terms have been introduced into the transport equation which accurately describe the small changes in energy and angle. The new code has been tested for a d(50)-Be neutron source in a deep penetration iron problem. The influence of the forward peaked elastic scattering on the fast neutron spectrum is shown to be significant and can be handled efficiently in the new ANISN version. Since common cross-section libraries are limited by Legendre expansion, or by their upper energy boundary, or exclude elastic scattering above 20 MeV a special library has been created. (Auth.)

Transport theory calculation for a heterogeneous multi-assembly problem by characteristics method with direct neutron path linking technique

International Nuclear Information System (INIS)

Kosaka, Shinya; Saji, Etsuro

2000-01-01

A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted, resulting in a significant reduction in the wall-clock time of the calculation. By all these efforts, coupled with the results of many verification studies, a whole LWR core heterogeneous transport theory calculation finally became practical. CHAPLET is thought to be a useful tool which can produce the reference solutions for analyses of an LWR (author)

Statistics of Monte Carlo methods used in radiation transport calculation

International Nuclear Information System (INIS)

Datta, D.

2009-01-01

Radiation transport calculation can be carried out by using either deterministic or statistical methods. Radiation transport calculation based on statistical methods is basic theme of the Monte Carlo methods. The aim of this lecture is to describe the fundamental statistics required to build the foundations of Monte Carlo technique for radiation transport calculation. Lecture note is organized in the following way. Section (1) will describe the introduction of Basic Monte Carlo and its classification towards the respective field. Section (2) will describe the random sampling methods, a key component of Monte Carlo radiation transport calculation, Section (3) will provide the statistical uncertainty of Monte Carlo estimates, Section (4) will describe in brief the importance of variance reduction techniques while sampling particles such as photon, or neutron in the process of radiation transport

EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

International Nuclear Information System (INIS)

Jones, D.B.

1986-01-01

EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

Two-dimensional time dependent Riemann solvers for neutron transport

International Nuclear Information System (INIS)

Brunner, Thomas A.; Holloway, James Paul

2005-01-01

A two-dimensional Riemann solver is developed for the spherical harmonics approximation to the time dependent neutron transport equation. The eigenstructure of the resulting equations is explored, giving insight into both the spherical harmonics approximation and the Riemann solver. The classic Roe-type Riemann solver used here was developed for one-dimensional problems, but can be used in multidimensional problems by treating each face of a two-dimensional computation cell in a locally one-dimensional way. Several test problems are used to explore the capabilities of both the Riemann solver and the spherical harmonics approximation. The numerical solution for a simple line source problem is compared to the analytic solution to both the P 1 equation and the full transport solution. A lattice problem is used to test the method on a more challenging problem

Solution of the neutron transport equation by the Method of Characteristics using a linear representation of the source within a mesh

International Nuclear Information System (INIS)

Mazumdar, Tanay; Degweker, S.B.

2017-01-01

Highlights: • In Method of Characteristics, the neutron source within a mesh is expanded up to linear term. • This expansion reduces the number of meshes as compared to flat source assumption. • Poor representation of circular geometry with coarser meshes is corrected. • Few benchmark problems are solved to show the advantages of linear expansion of source. • The advantage of the present formalism is quite visible in problems with large flux gradient. - Abstract: A common assumption in the solution of the neutron transport equation by the Method of Characteristics (MOC) is that the source (or flux) is constant within a mesh. This assumption is adequate provided the meshes are small enough so that the spatial variation of flux within a mesh may be ignored. Whether a mesh is small enough or not depends upon the flux gradient across a mesh, which in turn depends on factors like the presence of strong absorbers, localized sources or vacuum boundaries. The flat flux assumption often requires a very large number of meshes for solving the neutron transport equation with acceptable accuracy as was observed in our earlier work on the subject. A significant reduction in the required number of meshes is attainable by using a higher order representation of the flux within a mesh. In this paper, we expand the source within a mesh up to first order (linear) terms, which permits the use of larger sized (and therefore fewer) meshes and thereby reduces the computation time without compromising the accuracy of calculation. Since the division of the geometry into meshes is through an automatic triangulation procedure using the Bowyer-Watson algorithm, representation of circular objects (cylindrical fuel rods) with coarse meshes is poorer and causes geometry related errors. A numerical recipe is presented to make a correction to the automatic triangulation process and thereby eliminate this source of error. A number of benchmark problems are analyzed to emphasize the

Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures. 1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is th...

Simulation of neutron transport process, photons and charged particles within the Monte Carlo method

International Nuclear Information System (INIS)

Androsenko, A.A.; Androsenko, P.A.; Artamonov, S.N.; Bolonkina, G.V.; Lomtev, V.L.; Pupko, S.V.

1991-01-01

Description is given to the program system BRAND designed for the accurate solution of non-stationary transport equation of neutrons, photons and charged particles in the conditions of real three-dimensional geometry. An extensive set of local and non-local estimates provides an opportunity of calculating a great set of linear functionals normally being of interest in the calculation of reactors, radiation protection and experiment simulation. The process of particle interaction with substance is simulated on the basis of individual non-group data on each isotope of the composition. 24 refs

Parallel processing of neutron transport in fuel assembly calculation

International Nuclear Information System (INIS)

Song, Jae Seung

1992-02-01

Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

Practical adjoint Monte Carlo technique for fixed-source and eigenfunction neutron transport problems

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1981-01-01

An adjoint Monte Carlo technique is described for the solution of neutron transport problems. The optimum biasing function for a zero-variance collision estimator is derived. The optimum treatment of an analog of a non-velocity thermal group has also been derived. The method is extended to multiplying systems, especially for eigenfunction problems to enable the estimate of averages over the unknown fundamental neutron flux distribution. A versatile computer code, FOCUS, has been written, based on the described theory. Numerical examples are given for a shielding problem and a critical assembly, illustrating the performance of the FOCUS code. 19 refs

Neutron transport on the connection machine

International Nuclear Information System (INIS)

Robin, F.

1991-12-01

Monte Carlo methods are heavily used at CEA and account for a a large part of the total CPU time of industrial codes. In the present work (done in the frame of the Parallel Computing Project of the CEL-V Applied Mathematics Department) we study and implement on the Connection Machine an optimised Monte Carlo algorithm for solving the neutron transport equation. This allows us to investigate the suitability of such an architecture for this kind of problem. This report describes the chosen methodology, the algorithm and its performances. We found that programming the CM-2 in CM Fortran is relatively easy and we got interesting performances as, on a 16 k, CM-2 they are the same level as those obtained on one processor of a CRAY X-MP with a well optimized vector code

Transport methods: general. 3. An Additive Angular-Dependent Re-balance Acceleration Method for Neutron Transport Equations

International Nuclear Information System (INIS)

Cho, Nam Zin; Park, Chang Je

2001-01-01

An additive angular-dependent re-balance (AADR) factor acceleration method is described to accelerate the source iteration of discrete ordinates transport calculation. The formulation of the AADR method follows that of the angular-dependent re-balance (ADR) method in that the re-balance factor is defined only on the cell interface and in that the low-order equation is derived by integrating the transport equation (high-order equation) over angular subspaces. But, the re-balance factor is applied additively. While the AADR method is similar to the boundary projection acceleration and the alpha-weighted linear acceleration, it is more general and does have distinct features. The method is easily extendible to DP N and low-order S N re-balancing, and it does not require consistent discretizations between the high- and low-order equations as in diffusion synthetic acceleration. We find by Fourier analysis and numerical results that the AADR method with a chosen form of weighting functions is unconditionally stable and very effective. There also exists an optimal weighting parameter that leads to the smallest spectral radius. The AADR acceleration method described in this paper is simple to implement, unconditionally stable, and very effective. It uses a physically based weighting function with an optimal parameter, leading to the best spectral radius of ρ<0.1865, compared to ρ<0.2247 of DSA. The application of the AADR acceleration method with the LMB scheme on a test problem shows encouraging results

A hybrid source-driven method to compute fast neutron fluence in reactor pressure vessel - 017

International Nuclear Information System (INIS)

Ren-Tai, Chiang

2010-01-01

A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribution of the peripheral assemblies in a nuclear reactor obtained from reactor core depletion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U-238 (such as U-234, U-235, U-236, Pu-239 etc) are replaced with U-238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly developed hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors. (authors)

Comparison of neutronic transport equation resolution nodal methods

International Nuclear Information System (INIS)

Zamonsky, O.M.; Gho, C.J.

1990-01-01

In this work, some transport equation resolution nodal methods are comparatively studied: the constant-constant (CC), linear-nodal (LN) and the constant-quadratic (CQ). A nodal scheme equivalent to finite differences has been used for its programming, permitting its inclusion in existing codes. Some bidimensional problems have been solved, showing that linear-nodal (LN) are, in general, obtained with accuracy in CPU shorter times. (Author) [es

Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code

International Nuclear Information System (INIS)

Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen

2013-01-01

Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)

Continuous energy adjoint Monte Carlo for coupled neutron-photon transport

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.

2001-07-01

Although the theory for adjoint Monte Carlo calculations with continuous energy treatment for neutrons as well as for photons is known, coupled neutron-photon transport problems present fundamental difficulties because of the discrete energies of the photons produced by neutron reactions. This problem was solved by forcing the energy of the adjoint photon to the required discrete value by an adjoint Compton scattering reaction or an adjoint pair production reaction. A mathematical derivation shows the exact procedures to follow for the generation of an adjoint neutron and its statistical weight. A numerical example demonstrates that correct detector responses are obtained compared to a standard forward Monte Carlo calculation. (orig.)

Application of Walsh functions to neutron transport problems. I. Theory

International Nuclear Information System (INIS)

Seed, T.J.; Albrecht, R.W.

1976-01-01

An approximation to the neutron transport equation is made by representing the angular flux with an expansion of the angular dependence in the orthogonal, complete, and binary valued sets of Walsh function. The Walsh approximation is applied to the one-speed, isotropic-scattering, rectangular-geometry form of the neutron transport equation. Sets of partial differential equations for the expansion coefficients are derived along with appropriate boundary conditions for their solution. The sets of the Walsh expansion to one- and two-dimensional forms of the transport equation are also obtained. The two-dimensional expansion coefficient equations are shown to be not only hyperbolic but also transformable to a set of S/sub N/-like equations that are coupled only through the scattering term. Such transformal sets of equations are termed Walsh-derived quadrature sets

Killing symmetries in neutron transport

International Nuclear Information System (INIS)

Lukacs, B.; Racz, A.

1992-10-01

Although inside the reactor zone there is no exact continuous spatial symmetry, in certain configurations neutron flux distribution is close to a symmetrical one. In such cases the symmetrical solution could provide a good starting point to determine the non-symmetrical power distribution. All possible symmetries are determined in the 3-dimensional Euclidean space, and the form of the transport equation is discussed in such a coordinate system which is adapted to the particular symmetry. Possible spontaneous symmetry breakings are pointed out. (author) 6 refs

On generating neutron transport tables with the NJOY system

International Nuclear Information System (INIS)

Caldeira, Alexandre D.; Claro, Luiz H.

2013-01-01

Incorrect values for the product of the average number of neutrons released per fission and the fission microscopic cross-section were detected in several energy groups of a neutron transport table generated with the most updated version of the NJOY system. It was verified that the problem persists when older versions of this system are utilized. Although this problem exists for, at least, ten years, it is still an open question. (author)

On discontinuous Galerkin and discrete ordinates approximations for neutron transport equation and the critical eigenvalue

International Nuclear Information System (INIS)

Asadzadeh, M.; Thevenot, L.

2010-01-01

The objective of this paper is to give a mathematical framework for a fully discrete numerical approach for the study of the neutron transport equation in a cylindrical domain (container model,). More specifically, we consider the discontinuous Galerkin (D G) finite element method for spatial approximation of the mono-energetic, critical neutron transport equation in an infinite cylindrical domain ??in R3 with a polygonal convex cross-section ? The velocity discretization relies on a special quadrature rule developed to give optimal estimates in discrete ordinate parameters compatible with the quasi-uniform spatial mesh. We use interpolation spaces and derive optimal error estimates, up to maximal available regularity, for the fully discrete scalar flux. Finally we employ a duality argument and prove superconvergence estimates for the critical eigenvalue.

Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

International Nuclear Information System (INIS)

Dunley, Leonardo Souza

2002-01-01

The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron

MVP/GMVP Version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

International Nuclear Information System (INIS)

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

2017-03-01

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

International Nuclear Information System (INIS)

Lee, S.R.; Cummings, J.C.; Nolen, S.D.

1997-01-01

We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed

Study of a transportable neutron radiography system; Estudo de um sistema neutrongrafico transportavel

Energy Technology Data Exchange (ETDEWEB)

Souza, S N.A. de

1991-05-01

This work presents a study a transportable neutron radiography system for a 185 GBq {sup 241} Am-Be ({alpha}, {eta}) source with a neutron yield roughly 1,25 x 10{sup 7} n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio of 14, for neutron fluxes up to 4,09 x 10{sup 2} n.cm{sup -2}.s{sup -1}. Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10{sup 2} n.cm{sup -2}.s{sup -1} and 2,65 x 10{sup 2} n.cm{sup -2}.s{sup -1}, respectively. (author). 30 refs, 39 figs, 9 tabs.

Parallel computing for homogeneous diffusion and transport equations in neutronics

International Nuclear Information System (INIS)

Pinchedez, K.

1999-06-01

Parallel computing meets the ever-increasing requirements for neutronic computer code speed and accuracy. In this work, two different approaches have been considered. We first parallelized the sequential algorithm used by the neutronics code CRONOS developed at the French Atomic Energy Commission. The algorithm computes the dominant eigenvalue associated with PN simplified transport equations by a mixed finite element method. Several parallel algorithms have been developed on distributed memory machines. The performances of the parallel algorithms have been studied experimentally by implementation on a T3D Cray and theoretically by complexity models. A comparison of various parallel algorithms has confirmed the chosen implementations. We next applied a domain sub-division technique to the two-group diffusion Eigen problem. In the modal synthesis-based method, the global spectrum is determined from the partial spectra associated with sub-domains. Then the Eigen problem is expanded on a family composed, on the one hand, from eigenfunctions associated with the sub-domains and, on the other hand, from functions corresponding to the contribution from the interface between the sub-domains. For a 2-D homogeneous core, this modal method has been validated and its accuracy has been measured. (author)

Application of a discrete-energy, discrete-ordinates technique to the study of neutron transport in iron

International Nuclear Information System (INIS)

Ching, J.T.

1975-01-01

An algebraic equivalence between the point-energy and multigroup forms of the Boltzmann transport equation is demonstrated which allows the development of a discrete-energy, discrete-ordinates method for the solution of radiation transport problems. The method utilizes a modified version of a cross section processing scheme devised for the moments method code BMT and the transport equation solution algorithm from the one-dimensional discrete-ordinates transport code ANISN. The combined system, identified as MOMANS, computes fluxes directly from point cross sections in a single operation. In the cross-section processing, the group averaging required for multigroup calculations is replaced by a fast numerical scheme capable of generating a set of transfer cross sections containing all the physical features of interest, thereby increasing the detail in the calculated results. Test calculations in which the discrete-energy method was compared with the multigroup method have shown that for the same energy grid (number of points = number of groups), the discrete-energy method is faster but somewhat less accurate than the multigroup method. However, the accuracy of the discrete-energy method increases rapidly as the spacing between energy points is decreased, approaching that of multigroup calculations. For problems requiring great detail in the energy spectrum the discrete-energy method has therefore proven to be as accurate as, and more economical than, the multigroup technique. This was demonstrated by the application of the method to the study of the transport of neutrons in an iron sphere. Using the capability of the discrete-energy method for rapidly treating changes in cross-section sets, the propagation of neutrons from a 14 MeV source in a 22 cm radius sphere of iron was analyzed for sensitivity to changes in the microscopic scattering mechanisms

Cellular neural networks (CNN) simulation for the TN approximation of the time dependent neutron transport equation in slab geometry

International Nuclear Information System (INIS)

Hadad, Kamal; Pirouzmand, Ahmad; Ayoobian, Navid

2008-01-01

This paper describes the application of a multilayer cellular neural network (CNN) to model and solve the time dependent one-speed neutron transport equation in slab geometry. We use a neutron angular flux in terms of the Chebyshev polynomials (T N ) of the first kind and then we attempt to implement the equations in an equivalent electrical circuit. We apply this equivalent circuit to analyze the T N moments equation in a uniform finite slab using Marshak type vacuum boundary condition. The validity of the CNN results is evaluated with numerical solution of the steady state T N moments equations by MATLAB. Steady state, as well as transient simulations, shows a very good comparison between the two methods. We used our CNN model to simulate space-time response of total flux and its moments for various c (where c is the mean number of secondary neutrons per collision). The complete algorithm could be implemented using very large-scale integrated circuit (VLSI) circuitry. The efficiency of the calculation method makes it useful for neutron transport calculations

Numerical solution of neutron transport equations in discrete ordinates and slab geometry

International Nuclear Information System (INIS)

Serrano Pedraza, F.

1985-01-01

An unified formalism to solve numerically, between other equation, the neutron transport in discrete ordinates, slab geometry, several energy groups and independents of time, has been developed recently. Such a formalism cover some of the conventional schemes as diamond difference, (WDD) characteristic step (SC) lineal characteristic (LC), quadratic characteristic (QC) and lineal discontinuous. Unified formation gives before hand the convergence order of the previously selected scheme. In fact it allows besides to generate a big amount of numerical schemes, with which is also possible to solve numerical equations as soon as neutron transport. The essential purpose of this work was to solve the neutron transport equations in slab geometry and discrete ordinates considering several energy groups without to take under advisement time dependence based in the above mentioned unified formalism. To reach this purpose it was necesary to design a computer code with the name TNOD1 (Neutron transport in discrete ordinates and 1 dimension) which includes each one of the schemes already pointed out. there exist two numerical schemes, also recently developed, quadratic continuous (QC) and cubic continuous (CN), although covered by unified formalism, it has been possible to include them inside this computer code without make substantial changes in its structure. In chapter I, derivative of neutron transport equation independent of time is taken, for angular flux, including boundary conditions and discontinuity. In chapter II the neutron transport equations are obtained in multigroups, independents of time, for approximation of discrete ordinates. Description of theory related with unified formalism and its relationship with mentioned discretization schemes is presented in chapter III. Chapter IV describes the computer code developed and finally, in chapter V different numerical results obtained with TNOD1 program are shown. In Appendix A theorems and mathematical arguments used

Considerations in the design of an improved transportable neutron spectrometer

CERN Document Server

Williams, A M; Brushwood, J M; Beeley, P A

2002-01-01

The Transportable Neutron Spectrometer (TNS) has been used by the Ministry of Defence for over 15 years to characterise neutron fields in workplace environments and provide local correction factors for both area and personal dosimeters. In light of advances in neutron spectrometry, a programme to evaluate and improve TNS has been initiated. This paper describes TNS, presents its operation in known radioisotope fields and in a reactor environment. Deficiencies in the operation of the instrument are highlighted, together with proposals for updating the response functions and spectrum unfolding methodologies.

FURNACE; a toroidal geometry neutronic program system method description and users manual

International Nuclear Information System (INIS)

Verschuur, K.A.

1984-12-01

The FURNACE program system performs neutronic and photonic calculations in 3D toroidal geometry for application to fusion reactors. The geometry description is quite general, allowing any torus cross section and any neutron source density distribution for the plasma, as well as simple parametric representations of circular, elliptic and D-shaped tori and plasmas. The numerical method is based on an approximate transport model that produces results with sufficient accuracy for reactor-design purposes, at acceptable calculational costs. A short description is given of the numerical method, and a user manual for the programs of the system: FURNACE, ANISN-PT, LIBRA, TAPEMA and DRAWER is presented

The analysis of RPV fast neutron flux calculation for PWR with three-dimensional SN method

International Nuclear Information System (INIS)

Yang Shouhai; Chen Yixue; Wang Weijin; Shi Shengchun; Lu Daogang

2011-01-01

Discrete ordinates (S N ) method is one of the most widely used method for reactor pressure vessel (RPV) design. As the fast development of computer CPU speed and memory capacity and consummation of three-dimensional discrete-ordinates method, it is mature for 3-D S N method to be used to engineering design for nuclear facilities. This work was done specifically for PWR model, with the results of 3-D core neutron transport calculation by 3-D core calculation, 3-D RPV fast neutron flux distribution obtain by 3-D S N method were compared with gained by 1-D and 2-D S N method and the 3-D Monte Carlo (MC) method. In this paper, the application of three-dimensional S N method in calculating RPV fast neutron flux distribution for pressurized water reactor (PWR) is presented and discussed. (authors)

Calculations of Neutron Flux Distributions by Means of Integral Transport Methods

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I

1967-05-15

Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.

Transport-level description of the 252Cf-source method using the Langevin technique

International Nuclear Information System (INIS)

Stolle, A.M.; Akcasu, A.Z.

1991-01-01

The fluctuations in the neutron number density and detector outputs in a nuclear reactor can be analyzed conveniently by using the Langevin equation approach. This approach can be implemented at any level of approximation to describe the time evolution of the neutron population, from the most complete transport-level description to the very basic point reactor analysis of neutron number density fluctuations. In this summary, the complete space- and velocity-dependent transport-level formulation of the Langevin equation approach is applied to the analysis of the 252 Cf-source-driven noise analysis (CSDNA) method, an experimental technique developed by J.T. Mihalczo at Oak Ridge National Laboratory, which makes use of noise analysis to determine the reactivity of subcritical media. From this analysis, a theoretical expression for the subcritical multiplication factor is obtained that can then be used to interpret the experimental data. Results at the transport level are in complete agreement with an independent derivation performed by Sutton and Doub, who used the probability density method to interpret the CSDNA experiment, but differed from other expressions that have appeared in the literature

Neutron transport in two dissimilar media anisotropic scattering

International Nuclear Information System (INIS)

Burkart, A.R.; Ishiguro, Y.; Siewert, C.E.

1976-01-01

The elementary solution of the one-speed neutron-transport equation with linearly anisotropic scattering are used in conjunction with Chandrasekhar's invariance principles to solve in a concise manner the Milne problem for two adjoining half-spaces and the critical reactor problem for a reflected slab

Least-squares finite element discretizations of neutron transport equations in 3 dimensions

Energy Technology Data Exchange (ETDEWEB)

Manteuffel, T.A [Univ. of Colorado, Boulder, CO (United States); Ressel, K.J. [Interdisciplinary Project Center for Supercomputing, Zurich (Switzerland); Starkes, G. [Universtaet Karlsruhe (Germany)

1996-12-31

The least-squares finite element framework to the neutron transport equation introduced in is based on the minimization of a least-squares functional applied to the properly scaled neutron transport equation. Here we report on some practical aspects of this approach for neutron transport calculations in three space dimensions. The systems of partial differential equations resulting from a P{sub 1} and P{sub 2} approximation of the angular dependence are derived. In the diffusive limit, the system is essentially a Poisson equation for zeroth moment and has a divergence structure for the set of moments of order 1. One of the key features of the least-squares approach is that it produces a posteriori error bounds. We report on the numerical results obtained for the minimum of the least-squares functional augmented by an additional boundary term using trilinear finite elements on a uniform tesselation into cubes.

Coupling of neutron transport equations. First results; Couplage d`equations en transport neutronique. premiere approche 1D monocinetique

Energy Technology Data Exchange (ETDEWEB)

Bal, G.

1995-07-01

To achieve whole core calculations of the neutron transport equation, we have to follow this 2 step method: space and energy homogenization of the assemblies; resolution of the homogenized equation on the whole core. However, this is no more valid when accidents occur (for instance depressurization causing locally strong heterogeneous media). One solution consists then in coupling two kinds of resolutions: a fine computation on the damaged cell (fine mesh, high number of energy groups) coupled with a coarse one everywhere else. We only deal here with steady state solutions (which already live in 6D spaces). We present here two such methods: The coupling by transmission of homogenized sections and the coupling by transmission of boundary conditions. To understand what this coupling is, we first restrict ourselves to 1D with respect to space in one energy group. The first two chapters deal with a recall of basic properties of the neutron transport equation. We give at chapter 3 some indications of the behaviour of the flux with respect to the cross sections. We present at chapter 4 some couplings and give some properties. Chapter 5 is devoted to a presentation of some numerical applications. (author). 9 refs., 7 figs.

Stress measurement by neutron diffraction method. Standard method using angular dispersion method and trial of using neutron IP

International Nuclear Information System (INIS)

Sasaki, Toshihiko; Takago, Shigeki

2016-01-01

This paper outlined a stress measurement method using neutrons, and introduced the application examples to stress measurement for metal-based composite materials. In the angular dispersion type measurement using a steady-state reactor type neutron source, the white beams taken out from a nuclear reactor are monochromatized (wavelength λ is a constant value) with a single crystal monochromator and utilized. As an example of measurement, there was the case as follows: the stress of a sintered material which has been put to practical use as valve seat part for automobiles was measured by the neutron method, and the deformation behavior during load was studied. This study performed neutron diffraction measurement using a residual stress analyzer (RESA: Diffractometer for Residual Stress Analysis) installed at JAEA's experimental reactor JRR-3. As a result, it was found that the stress state of the sintered composite material of Fe-Cr and TiN can be predicted with a micromechanics model. A neutron diffraction ring can be obtained using a neutron image plate (IP), where fine powder of gadolinium (Gd) was incorporated into IP for X-rays, and it can be used as an IP reader in the same way as the case of X-rays. A report has been introduced on the examination results of the highly accurate stress measurement by applying the cos α method devised for X-ray stress measurement to neutron diffraction ring. (A.O.)

Cellular neural network to the spherical harmonics approximation of neutron transport equation in x-y geometry. Part I: Modeling and verification for time-independent solution

International Nuclear Information System (INIS)

Pirouzmand, Ahmad; Hadad, Kamal

2011-01-01

Highlights: → This paper describes the solution of time-independent neutron transport equation. → Using a novel method based on cellular neural networks (CNNs) coupled with P N method. → Utilize the CNN model to simulate spatial scalar flux distribution in steady state. → The accuracy, stability, and capabilities of CNN model are examined in x-y geometry. - Abstract: This paper describes a novel method based on using cellular neural networks (CNN) coupled with spherical harmonics method (P N ) to solve the time-independent neutron transport equation in x-y geometry. To achieve this, an equivalent electrical circuit based on second-order form of neutron transport equation and relevant boundary conditions is obtained using CNN method. We use the CNN model to simulate spatial response of scalar flux distribution in the steady state condition for different order of spherical harmonics approximations. The accuracy, stability, and capabilities of CNN model are examined in 2D Cartesian geometry for fixed source and criticality problems.

Application of neutron/gamma transport codes for the design of explosive detection systems

International Nuclear Information System (INIS)

Elias, E.; Shayer, Z.

1994-01-01

Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs

Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons

International Nuclear Information System (INIS)

Sedlačková, Katarína; Zaťko, Bohumír; Šagátová, Andrea; Nečas, Vladimír

2013-01-01

Several Monte Carlo all-particle transport codes are under active development around the world. In this paper we focused on the capabilities of the MCNPX code (Monte Carlo N-Particle eXtended) to follow the particle transport in semiconductor detector of fast neutrons. Semiconductor detector based on semi-insulating GaAs was the object of our investigation. As converter material capable to produce charged particles from the (n, p) interaction, a high-density polyethylene (HDPE) was employed. As the source of fast neutrons, the 239 Pu–Be neutron source was used in the model. The simulations were performed using the MCNPX code which makes possible to track not only neutrons but also recoiled protons at all interesting energies. Hence, the MCNPX code enables seamless particle transport and no other computer program is needed to process the particle transport. The determination of the optimal thickness of the conversion layer and the minimum thickness of the active region of semiconductor detector as well as the energy spectra simulation were the principal goals of the computer modeling. Theoretical detector responses showed that the best detection efficiency can be achieved for 500 μm thick HDPE converter layer. The minimum detector active region thickness has been estimated to be about 400 μm. -- Highlights: ► Application of the MCNPX code for fast neutron detector design is demonstrated. ► Simulations of the particle transport through conversion film of HDPE are presented. ► Simulations of the particle transport through detector active region are presented. ► The optimal thickness of the HDPE conversion film has been calculated. ► Detection efficiency of 0.135% was reached for 500 μm thick HDPE conversion film

Development of a discrete-ordinate approximation of the neutron transport equation for coupled xy-R-geometry

International Nuclear Information System (INIS)

Maertens, H.D.

1982-01-01

The inhomogenious structure of modern heavy water reactor fuel elements result in a strong spacial dependence of the neutron flux. The flux distribution can be calculated in detail by numerical methods, which describe exactly the geometrical heterogeniety and take into account the neutron flux anisotropy by higher transport theoretical approximations. Starting from the discrete ordinate method an approximation of the neutron transport equation has been developed, allowing for a cylindrical representation of the fuel-elements in a rectangular array of rods. The discretisation of the space variables, is based on the finite-difference approximation, defining a rectangular lattice in a two-dimensional cartesian coordinate system, which can be cut and replaced by circular mesh elements of a partially one-dimensional cylindrical coordinate system at arbitrary space points. To couple the two spacial regions the outer circle line of a cylindrical geometry is approximated in the cartesian system by a polygon with n >= 8. A cylindrical geometry is approximated in the cartesian system by a polygon with n>=8. A cylindrical geometry is thus enclosed by a system of two-dimensional rectangular, triangular and trapezoid mesh elements. The directional distribution of the neutron flux is conserved when switching from the xy-system to the cylindrical coordinate system. The angle discretisation by balanced sets of squares (EQsub(n)) allows a simple definition of transfer-coefficients for the redistribution of the neutron flux due to coordinate transformations. The procedure is verified and tested by selected problems. Possible applications and limits are discussed. (orig.) [de

METHODS OF ASSESSMENT OF THE RELATIVE BIOLOGICAL EFFECTIVENESS OF NEUTRONS IN NEUTRON THERAPY

Directory of Open Access Journals (Sweden)

V. A. Lisin

2017-01-01

Full Text Available The relative biological effectiveness (RBE of fast neutrons is an important factor influencing the quality of neutron therapy therefore, the assessment of RBE is of great importance. Experimental and clinical studies as well as different mathematical and radiobiological models are used for assessing RBE. Research is conducted for neutron sources differing in the method of producing particles, energy and energy spectrum. Purpose: to find and analyze the dose-dependence of fast neutron RBE in neutron therapy using the U-120 cyclotron and NG-12I generator. Material and methods: The optimal method for assessing the relative biological effectiveness of neutrons for neutron therapy was described. To analyze the dependence of the RBE on neutron dose, the multi-target model of cell survival was applied. Results: The dependence of the RBE of neutrons produced from the U-120 cyclotron and NG-120 generator on the dose level was found for a single irradiation of biological objects. It was shown that the function of neutron dose was consistent with similar dependencies found by other authors in the experimental and clinical studies.

The “neutron channel design”—A method for gaining the desired neutrons

Directory of Open Access Journals (Sweden)

G. Hu

2016-12-01

Full Text Available The neutrons with desired parameters can be obtained after initial neutrons penetrating various structure and component of the material. A novel method, the “neutron channel design”, is proposed in this investigation for gaining the desired neutrons. It is established by employing genetic algorithm (GA combining with Monte Carlo software. This method is verified by obtaining 0.01eV to 1.0eV neutrons from the Compact Accelerator-driven Neutron Source (CANS. One layer polyethylene (PE moderator was designed and installed behind the beryllium target in CANS. The simulations and the experiment for detection the neutrons were carried out. The neutron spectrum at 500cm from the PE moderator was simulated by MCNP and PHITS software. The counts of 0.01eV to 1.0eV neutrons were simulated by MCNP and detected by the thermal neutron detector in the experiment. These data were compared and analyzed. Then this method is researched on designing the complex structure of PE and the composite material consisting of PE, lead and zirconium dioxide.

Sensitivity of neutron air transport to nitrogen cross section uncertainties

International Nuclear Information System (INIS)

Niiler, A.; Beverly, W.B.; Banks, N.E.

1975-01-01

The sensitivity of the transport of 14-MeV neutrons in sea level air to uncertainties in the ENDF/B-III values of the various Nitrogen cross sections has been calculated using the correlated sampling Monte Carlo neutron transport code SAMCEP. The source consisted of a 14.0- to 14.9-MeV band of isotropic neutrons and the fluences (0.5 to 15.0 MeV) were calculated at radii from 50 to 1500 metres. The maximum perturbations, assigned to the ENDF/B-III or base cross section set in the 6.0- to 14.5-MeV energy range were; (1) 2 percent to the total, (2) 10 percent to the total elastic, (3) 40 percent to the inelastic and absorption and (4) 20 percent to the first Legendre coefficient and 10 percent to the second Legendre coefficient of the elastic angular distribtuions. Transport calculations were carried out using various physically realistic sets of perturbed cross sections, bounded by evaluator-assigned uncertainties, as well as the base set. Results show that in some energy intervals at 1500 metres, the differential fluence level with a perturbed set differed by almost a factor of two from the differential fluence level with the base set. 5 figures

The experimental method for neutron dose-equivalent detection

International Nuclear Information System (INIS)

Ji Changsong

1992-01-01

A new method, for getting neutron dose-equivalent Cd rode absorption method is described. The method adopts Cd-rode-swarm buck absorption, which greatly improved the neutron sensitivity and simplified the adjustment method. By this method, the author has developed BH3105 model neutron dose equivalent meter, the sensitivity of this instrument reach 10 cps/μSvh -1 . γ-ray depression rate reaches 4000:1, the measurement range is 0.1 μSv/h-10 6 μSv/h. The energy response is good (from thermal neutron-14 MeV neutron), this instrument can be used to measure the dose equivalent of the neutron areas

FPGA hardware acceleration for high performance neutron transport computation based on agent methodology - 318

International Nuclear Information System (INIS)

Shanjie, Xiao; Tatjana, Jevremovic

2010-01-01

The accurate, detailed and 3D neutron transport analysis for Gen-IV reactors is still time-consuming regardless of advanced computational hardware available in developed countries. This paper introduces a new concept in addressing the computational time while persevering the detailed and accurate modeling; a specifically designed FPGA co-processor accelerates robust AGENT methodology for complex reactor geometries. For the first time this approach is applied to accelerate the neutronics analysis. The AGENT methodology solves neutron transport equation using the method of characteristics. The AGENT methodology performance was carefully analyzed before the hardware design based on the FPGA co-processor was adopted. The most time-consuming kernel part is then transplanted into the FPGA co-processor. The FPGA co-processor is designed with data flow-driven non von-Neumann architecture and has much higher efficiency than the conventional computer architecture. Details of the FPGA co-processor design are introduced and the design is benchmarked using two different examples. The advanced chip architecture helps the FPGA co-processor obtaining more than 20 times speed up with its working frequency much lower than the CPU frequency. (authors)

CACTUS, a characteristics solution to the neutron transport equations in complicated geometries

International Nuclear Information System (INIS)

Halsall, M.J.

1980-04-01

CACTUS has been written to solve the multigroup neutron transport equation in a general two-dimensional geometry. The method is based upon a characteristics formulation for the problem in which the transport equation is integrated explicitly along straight line tracks that are suitably distributed throughout the problem. Source distributions and scattering are assumed to be isotropic, but the only restriction on geometry is that the outer boundary should be rectangular. Within this rectangular boundary the user is free to build his problem geometry using any combination of intersecting straight lines and circular arcs. The theory of the method is described, followed by some details of a coding, a sensitivity study on the number of tracks required to integrate fluxes in a particular problem, a user's guide, and a few test cases. (author)

Dosimetry methods in boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D. [Universita degli Studi di Milano, Department of Physics, Via Festa del Patrono 7, 20122 Milano (Italy); Agosteo, S.; Barcaglioni, L. [Istituto Nazionale di Fisica Nucleare, Milano (Italy); Campi, F.; Garlati, L. [Politecnico di Milano, Energy Department, Piazza Leonardo Da Vinci 32, 20133 Milano (Italy); De Errico, F. [Universita degli Studi di Pisa, Department of Civil and Industrial Engineering, Lungamo Pacinotti 43, 56126 Pisa (Italy); Borroni, M.; Carrara, M. [Fondazione IRCCS Istituto Nazionale Tumori, Medical Physics Unit, Via Venezian 1, 20133 Milano (Italy); Burian, J.; Klupak, V.; Viererbl, L.; Marek, M. [Research Centre Rez, Department of Neutron Physics, 250-68 Husinec-Rez (Czech Republic)

2014-08-15

Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)

Dosimetry methods in boron neutron capture therapy

International Nuclear Information System (INIS)

Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D.; Agosteo, S.; Barcaglioni, L.; Campi, F.; Garlati, L.; De Errico, F.; Borroni, M.; Carrara, M.; Burian, J.; Klupak, V.; Viererbl, L.; Marek, M.

2014-08-01

Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)

The behaviour of neutron bursts in matter

International Nuclear Information System (INIS)

Syros, C.

1978-01-01

An exact method is developed for solving the time-dependent linear transport equation for neutrons. The problem of finding the behaviour of neutron bursts in matter have been considered. The method leads to a new kind of perturbation theory applicable to the transport theoretical reactor dynamics. Applications of the theory are given for discontinuously or continuously distributed initial values of the neutron population. The boundary and initial conditions are exactly fulfilled. (author)

Cosmic-ray neutron transport at a forest field site: the sensitivity to various environmental conditions with focus on biomass and canopy interception

Science.gov (United States)

Andreasen, Mie; Jensen, Karsten H.; Desilets, Darin; Zreda, Marek; Bogena, Heye R.; Looms, Majken C.

2017-04-01

Cosmic-ray neutron intensity is inversely correlated to all hydrogen present in the upper decimeters of the subsurface and the first few hectometers of the atmosphere above the ground surface. This correlation forms the base of the cosmic-ray neutron soil moisture estimation method. The method is, however, complicated by the fact that several hydrogen pools other than soil moisture affect the neutron intensity. In order to improve the cosmic-ray neutron soil moisture estimation method and explore the potential for additional applications, knowledge about the environmental effect on cosmic-ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different parameters describing the subsurface to match measured height profiles and time series of thermal and epithermal neutron intensities at a field site in Denmark. Overall, modeled thermal and epithermal neutron intensities are in satisfactory agreement with measurements; however, the choice of forest canopy conceptualization is found to be significant. Modeling results show that the effect of canopy interception, soil chemistry and dry bulk density of litter and mineral soil on neutron intensity is small. On the other hand, the neutron intensity decreases significantly with added litter-layer thickness, especially for epithermal neutron energies. Forest biomass also has a significant influence on the neutron intensity height profiles at the examined field site, altering both the shape of the profiles and the ground-level thermal-to-epithermal neutron ratio. This ratio increases with increasing amounts of biomass, and was confirmed by measurements from three sites representing agricultural, heathland and forest land cover. A much smaller effect of canopy interception on the ground

Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

International Nuclear Information System (INIS)

Brenner, D.J.; Prael, R.E.; Little, R.C.

1987-01-01

Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs

Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

Energy Technology Data Exchange (ETDEWEB)

Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France); Méchin, Laurence [CNRS, UCBN, Groupe de Recherche en Informatique, Image, Automatique et Instrumentation de Caen, 14050 Caen (France); Hamel, Matthieu [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France)

2016-08-21

Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

International Nuclear Information System (INIS)

Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu

2016-01-01

Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

Transport calculations for a 14.8 MeV neutron beam in a water phantom

International Nuclear Information System (INIS)

Goetsch, S.J.

1981-01-01

A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented

A proposal on evaluation method of neutron absorption performance to substitute conventional neutron attenuation test

International Nuclear Information System (INIS)

Kim, Je Hyun; Shim, Chang Ho; Kim, Sung Hyun; Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo; Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho

2016-01-01

For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers

A proposal on evaluation method of neutron absorption performance to substitute conventional neutron attenuation test

Energy Technology Data Exchange (ETDEWEB)

Kim, Je Hyun; Shim, Chang Ho [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of); Kim, Sung Hyun [Nuclear Fuel Cycle Waste Treatment Research Division, Research Reactor Institute, Kyoto University, Osaka (Japan); Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo [Ionizing Radiation Center, Nuclear Fuel Cycle Waste Treatment Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho [Ionizing Radiation Center, Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)

2016-12-15

For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

Neutron transport by collision probability method in complicated geometries

International Nuclear Information System (INIS)

Constantin, Marin

2000-01-01

For the first flight collision probability (FFCP) method a rapidly increasing of the memory requirements and execution time with the number of discrete regions occurs. Generally, the use of the method is restricted at cell/supercell level. However, the amazing developments both in computer hardware and computer architecture allow a real extending of the problems' domain and a more detailed treatment of the geometry. Two ways are discussed into the paper: the direct design of new codes and the improving of the mainframe old versions. The author's experience is focused on the performances' improving of the 3D integral transport code PIJXYZ (from an old version to a modern one) and on the design and developing of the 2D transport code CP 2 D in the last years. In the first case an optimization process have been performed before the parallelization. In the second a modular design and the newest techniques (factorization of the geometry, the macrobands method, the mobile set of chords, the automatic calculation of the integration error, optimal algorithms for the innermost programming level, the mixed method for tracking process and CPs calculation, etc.) were adopted. In both cases the parallelization uses a PCs network system. Some short examples for CP 2 D and PIJXYZ calculation are presented: reactivity void effect in typical CANDU cells using a multistratified coolant model, a problem of some adjacent fuel assemblies, CANDU reactivity devices 3D simulation. (author)

Calibration of the JET neutron yield monitors using the delayed neutron counting technique

International Nuclear Information System (INIS)

van Belle, P.; Jarvis, O.N.; Sadler, G.; de Leeuw, S.; D'Hondt, P.; Pillon, M.

1990-01-01

The time-resolved neutron yield is routinely measured on the JET tokamak using a set of fission chambers. At present, the preferred technique is to employ activation reactions to determine the neutron fluence at a well-chosen position and to relate the measured fluence to the total neutron emission by means of neutron transport calculations. The delayed neutron counting method is a particularly convenient method of performing the activation measurement and the fission cross sections are accurately known. This paper outlines the measurement technique as used on JET

Development of a transportable neutron radiography system for non-destructive tests application

International Nuclear Information System (INIS)

Silva, Ademir X. da; Crispim, Verginia R.

1999-01-01

This paper presents a study of a transportable neutron radiography system utilizing californium-252. Studies about moderation, collimation and shielding are showed. A Monte Carlo Code, MCNP3b, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio 7,5, for neutron flux up to 6 X 10 -6 cm -2 .s -1 per neutron source. (author)

Domain decomposition and CMFD acceleration applied to discrete-ordinate methods for the solution of the neutron transport equation in XYZ geometries

International Nuclear Information System (INIS)

Masiello, Emiliano; Martin, Brunella; Do, Jean-Michel

2011-01-01

A new development for the IDT solver is presented for large reactor core applications in XYZ geometries. The multigroup discrete-ordinate neutron transport equation is solved using a Domain-Decomposition (DD) method coupled with the Coarse-Mesh Finite Differences (CMFD). The later is used for accelerating the DD convergence rate. In particular, the external power iterations are preconditioned for stabilizing the oscillatory behavior of the DD iterative process. A set of critical 2-D and 3-D numerical tests on a single processor will be presented for the analysis of the performances of the method. The results show that the application of the CMFD to the DD can be a good candidate for large 3D full-core parallel applications. (author)

Multigroup neutron transport equation in the diffusion and P{sub 1} approximation

Energy Technology Data Exchange (ETDEWEB)

Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1970-07-01

Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)

Numerical methods: Analytical benchmarking in transport theory

International Nuclear Information System (INIS)

Ganapol, B.D.

1988-01-01

Numerical methods applied to reactor technology have reached a high degree of maturity. Certainly one- and two-dimensional neutron transport calculations have become routine, with several programs available on personal computer and the most widely used programs adapted to workstation and minicomputer computational environments. With the introduction of massive parallelism and as experience with multitasking increases, even more improvement in the development of transport algorithms can be expected. Benchmarking an algorithm is usually not a very pleasant experience for the code developer. Proper algorithmic verification by benchmarking involves the following considerations: (1) conservation of particles, (2) confirmation of intuitive physical behavior, and (3) reproduction of analytical benchmark results. By using today's computational advantages, new basic numerical methods have been developed that allow a wider class of benchmark problems to be considered

New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code

Energy Technology Data Exchange (ETDEWEB)

Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

2016-10-15

The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.

Variation of the critical slab thickness with the degree of strongly anisotropic scattering in one-speed neutron transport theory

International Nuclear Information System (INIS)

Yildiz, C.

1998-01-01

The critical slab problem is studied in one-speed neutron transport theory using a linearly anisotropic kernel which combines forward and backward scattering. It is shown that, the recently observed non-monotonic variation of the thickness also exists in this strongly anisotropic case. In addition, the influence of the linear anisotropy on the critical thickness is analysed in detail. Numerical analysis for the critical thickness are performed using the spherical harmonics method and results are tabulated for selected illustrative cases as a function of different degrees of anisotropic scattering. Finally, some results are discussed and compared with those already obtained by other methods, the agreement is satisfactory. The spherical harmonic method gives generally accurate results in one dimensional geometry, and it is very suitable for the numerical solution of the neutron transport equation with linearly anisotropic scattering

Theoretical methods for neutronics calculations of core-blanket and core-reflector systems in fast reactors

International Nuclear Information System (INIS)

Corcuera, Roberto.

1975-12-01

The present work is a contribution to the neutronics calculational methods of fast neutron reactors. The first step is devoted to the analysis of the validity of the few-groups (of the order of 25) multigroup scheme, and of the transport-correction approximation for the treatment of the scattering anisotropy. This analysis includes both the reactor core, where the usual approximations are found to be satisfactory, and the reflector, where it turns out that the rapid variations of the neutron flux and of it's spectrum necessitate the improvement of the multigroup cross-sections' generation. Therefore, a zero-dimensional simple and accurate model for the average spectrum in the reflector is developed by the space-energy synthesis method. Finally using the Rayleigh-Ritz method, a model is developed in which the flux is spatially represented by an analytical function. This model is applied to the analysis of the sensitivity of reflector neutronics parameters to the variations of the cross sections [fr

Solution of the neutron transport equation by means of Hermite-Ssub(infinity)-theory

International Nuclear Information System (INIS)

Brandt, D.; Haelg, W.; Mennig, J.

1979-01-01

A stable numerical approximation Hsub(α)-Ssub(infinity) is obtained through the use of Hermite's method of order α(Hsub(α)) in the spatial integration of the ID neutron transport equation. The theory for α = 1 is applied to a one-group shielding problem. Numerical calculations show the new method to converge much faster than earlier versions of Ssub(infinity)-theory. Comparison of H 1 - Ssub(infinity) with the well-known Ssub(N)-code ANISN indicates a large gain in computing time for the former. (Auth.)

Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

International Nuclear Information System (INIS)

Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

2006-01-01

The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such

Solution and Study of the Two-Dimensional Nodal Neutron Transport Equation

International Nuclear Information System (INIS)

Panta Pazos, Ruben; Biasotto Hauser, Eliete; Tullio de Vilhena, Marco

2002-01-01

In the last decade Vilhena and coworkers reported an analytical solution to the two-dimensional nodal discrete-ordinates approximations of the neutron transport equation in a convex domain. The key feature of these works was the application of the combined collocation method of the angular variable and nodal approach in the spatial variables. By nodal approach we mean the transverse integration of the SN equations. This procedure leads to a set of one-dimensional S N equations for the average angular fluxes in the variables x and y. These equations were solved by the old version of the LTS N method, which consists in the application of the Laplace transform to the set of nodal S N equations and solution of the resulting linear system by symbolic computation. It is important to recall that this procedure allow us to increase N the order of S N up to 16. To overcome this drawback we step forward performing a spectral painstaking analysis of the nodal S N equations for N up to 16 and we begin the convergence of the S N nodal equations defining an error for the angular flux and estimating the error in terms of the truncation error of the quadrature approximations of the integral term. Furthermore, we compare numerical results of this approach with those of other techniques used to solve the two-dimensional discrete approximations of the neutron transport equation. (authors)

Approximate solution to neutron transport equation with linear anisotropic scattering

International Nuclear Information System (INIS)

Coppa, G.; Ravetto, P.; Sumini, M.

1983-01-01

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)

Beam-transport optimization for cold-neutron spectrometer

Directory of Open Access Journals (Sweden)

Nakajima Kenji

2015-01-01

Full Text Available We report the design of the beam-transport system (especially the vertical geometry for a cold-neutron disk-chopper spectrometer AMATERAS at J-PARC. Based on the elliptical shape, which is one of the most effective geometries for a ballistic mirror, the design was optimized to obtain, at the sample position, a neutron beam with high flux without serious degrading in divergence and spacial homogeneity within the boundary conditions required from actual spectrometer construction. The optimum focal point was examined. An ideal elliptical shape was modified to reduce its height without serious loss of transmission. The final result was adapted to the construction requirements of AMATERAS. Although the ideas studied in this paper are considered for the AMATERAS case, they can be useful also to other spectrometers in similar situations.

properties of the SN - equivalent integral transport operator in slab geometry and the iterative acceleration of neutron transport methods

International Nuclear Information System (INIS)

Massimiliano, Rosa; Azmy, Y.Y.; Morel, J.E.

2005-01-01

solution of the neutron transport equation. (authors)

Discontinuous Galerkin discretization and hp-refinement for the resolution of the neutron transport equation

International Nuclear Information System (INIS)

Fournier, Damien; Le-Tellier, Romain; Herbin, Raphaele

2013-01-01

This paper presents an hp-refinement method for a first order scalar transport reaction equation discretized by a discontinuous Galerkin method. First, the theoretical rates of convergence of h- and p-refinement are recalled and numerically tested. Then, in order to design some meshes, we propose two different estimators of the local error on the spatial domain. These quantities are analyzed and compared depending on the regularity of the solution so as to find the best way to lead the refinement process and the best strategy to choose between h- and p-refinement. Finally, the different possible refinement strategies are compared first on analytical examples and then on realistic applications for neutron transport in a nuclear reactor core. (authors)

MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

International Nuclear Information System (INIS)

2005-01-01

A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be

Methods for Probing Magnetic Films with Neutrons

Science.gov (United States)

Kozhevnikov, S. V.; Ott, F.; Radu, F.

2018-03-01

We review various methods in the investigation of magnetic films with neutrons, including those based on the effects of Larmor precession, Zeeman spatial splitting of the beam, neutron spin resonance, and polarized neutron channeling. The underlying principles, examples of the investigated systems, specific features, applications, and perspectives of these methods are discussed.

Neutron source multiplication method

International Nuclear Information System (INIS)

Clayton, E.D.

1985-01-01

Extensive use has been made of neutron source multiplication in thousands of measurements of critical masses and configurations and in subcritical neutron-multiplication measurements in situ that provide data for criticality prevention and control in nuclear materials operations. There is continuing interest in developing reliable methods for monitoring the reactivity, or k/sub eff/, of plant operations, but the required measurements are difficult to carry out and interpret on the far subcritical configurations usually encountered. The relationship between neutron multiplication and reactivity is briefly discussed and data presented to illustrate problems associated with the absolute measurement of neutron multiplication and reactivity in subcritical systems. A number of curves of inverse multiplication have been selected from a variety of experiments showing variations observed in multiplication during the course of critical and subcritical experiments where different methods of reactivity addition were used, with different neutron source detector position locations. Concern is raised regarding the meaning and interpretation of k/sub eff/ as might be measured in a far subcritical system because of the modal effects and spectrum differences that exist between the subcritical and critical systems. Because of this, the calculation of k/sub eff/ identical with unity for the critical assembly, although necessary, may not be sufficient to assure safety margins in calculations pertaining to far subcritical systems. Further study is needed on the interpretation and meaning of k/sub eff/ in the far subcritical system

Discrete energy formulation of neutron transport theory applied to solving the discrete ordinates equations

International Nuclear Information System (INIS)

Ching, J.; Oblow, E.M.; Goldstein, H.

1976-01-01

An algebraic equivalence between the point-energy and multigroup forms of the Boltzmann transport equation is demonstrated that allows the development of a discrete energy, discrete ordinates method for the solution of radiation transport problems. In the discrete energy method, the group averaging required in the cross-section processing for multigroup calculations is replaced by a faster numerical quadrature scheme capable of generating transfer cross sections describing all the physical processes of interest on a fine point-energy grid. Test calculations in which the discrete energy method is compared with the multigroup method show that, for the same energy grid, the discrete energy method is much faster, although somewhat less accurate, than the multigroup method. However, the accuracy of the discrete energy method increases rapidly as the spacing between energy grid points is decreased, approaching that of multigroup calculations. For problems requiring great detail in the energy spectrum, the discrete energy method is therefore expected to be far more economical than the multigroup technique for equivalent accuracy solutions. This advantage of the point method is demonstrated by application to the study of neutron transport in a thick iron slab

GANTRAS - a system of codes for the solution of the multigroup transport equation with a rigorous treatment of anisotropic neutron scattering

International Nuclear Information System (INIS)

Schwenk-Ferrero, A.

1986-11-01

GANTRAS is a system of codes for neutron transport calculations in which the anisotropy of elastic and inelastic (including (n,n'x)-reactions) scattering is fully taken into account. This is achieved by employing a rigorous method, so-called I * -method, to represent the scattering term of the transport equation and with the use of double-differential cross-sections for the description of the emission of secondary neutrons. The I * -method was incorporated into the conventional transport code ONETRAN. The ONETRAN subroutines were modified for the new purpose. An implementation of the updated version ANTRA1 was accomplished for plane and spherical geometry. ANTRA1 was included in GANTRAS and linked to another modules which prepare angle-dependent transfer matrices. The GANTRAS code consists of three modules: 1. The CROMIX code which calculates the macroscopic transfer matrices for mixtures on the base of microscopic nuclide-dependent data. 2. The ATP code which generates discretized angular transfer probabilities (i.e. discretizes the I * -function). 3. The ANTRA1 code to perform S N transport calculations in one-dimensional plane and spherical geometries. This structure of GANTRAS allows to accommodate the system to various transport problems. (orig.) [de

Some results on the neutron transport and the coupling of equations; Quelques resultats sur le transport neutronique et le couplage d`equations

Energy Technology Data Exchange (ETDEWEB)

Bal, G. [Electricite de France (EDF), Direction des Etudes et Recherches, 92 - Clamart (France)

1997-12-31

Neutron transport in nuclear reactors is well modeled by the linear Boltzmann transport equation. Its resolution is relatively easy but very expensive. To achieve whole core calculations, one has to consider simpler models, such as diffusion or homogeneous transport equations. However, the solutions may become inaccurate in particular situations (as accidents for instance). That is the reason why we wish to solve the equations on small area accurately and more coarsely on the remaining part of the core. It is than necessary to introduce some links between different discretizations or modelizations. In this note, we give some results on the coupling of different discretizations of all degrees of freedom of the integral-differential neutron transport equation (two degrees for the angular variable, on for the energy component, and two or three degrees for spatial position respectively in 2D (cylindrical symmetry) and 3D). Two chapters are devoted to the coupling of discrete ordinates methods (for angular discretization). The first one is theoretical and shows the well posing of the coupled problem, whereas the second one deals with numerical applications of practical interest (the results have been obtained from the neutron transport code developed at the R and D, which has been modified for introducing the coupling). Next, we present the nodal scheme RTN0, used for the spatial discretization. We show well posing results for the non-coupled and the coupled problems. At the end, we deal with the coupling of energy discretizations for the multigroup equations obtained by homogenization. Some theoretical results of the discretization of the velocity variable (well-posing of problems), which do not deal directly with the purposes of coupling, are presented in the annexes. (author). 34 refs.

An improved fast neutron radiography quantitative measurement method

International Nuclear Information System (INIS)

Matsubayashi, Masahito; Hibiki, Takashi; Mishima, Kaichiro; Yoshii, Koji; Okamoto, Koji

2004-01-01

The validity of a fast neutron radiography quantification method, the Σ-scaling method, which was originally proposed for thermal neutron radiography was examined with Monte Carlo calculations and experiments conducted at the YAYOI fast neutron source reactor. Water and copper were selected as comparative samples for a thermal neutron radiography case and a dense object, respectively. Although different characteristics on effective macroscopic cross-sections were implied by the simulation, the Σ-scaled experimental results with the fission neutron spectrum cross-sections were well fitted to the measurements for both the water and copper samples. This indicates that the Σ-scaling method could be successfully adopted for quantitative measurements in fast neutron radiography

Multi-level nonlinear diffusion acceleration method for multigroup transport k-Eigenvalue problems

International Nuclear Information System (INIS)

Anistratov, Dmitriy Y.

2011-01-01

The nonlinear diffusion acceleration (NDA) method is an efficient and flexible transport iterative scheme for solving reactor-physics problems. This paper presents a fast iterative algorithm for solving multigroup neutron transport eigenvalue problems in 1D slab geometry. The proposed method is defined by a multi-level system of equations that includes multigroup and effective one-group low-order NDA equations. The Eigenvalue is evaluated in the exact projected solution space of smallest dimensionality, namely, by solving the effective one- group eigenvalue transport problem. Numerical results that illustrate performance of the new algorithm are demonstrated. (author)

Development of a neutron transport code many-group two-dimensional heterogeneous calculations by the method of characteristics

International Nuclear Information System (INIS)

Petkov, P.T.

2000-01-01

The method of characteristics (MOC) is gaining increased popularity in the reactor physics community all over the world because it gives a new degree of freedom in nuclear reactor analysis. The MARIKO code solves the neutron transport equation by the MOC in two-dimensional real geometry. The domain of solution can be a rectangle or right hexagon with periodic boundary conditions on the outer boundary. Any reasonable symmetry inside the domain can be fully accounted for. The geometry is described in three levels-macro-cells, cells, and regions. The macro-cells and cells can be any polygon. The outer boundary of a region can be any combination of straight line and circular arc segments. Any level of embedded regions is allowed. Procedures for automatic geometry description of hexagonal fuel assemblies and reflector macro-cells have been developed. The initial ray tracing procedure is performed for the full rectangular or hexagonal domain, but only azimuthal angles in the smallest symmetry interval are tracked. (Authors)

Transport of D-D fusion neutrons in thick concrete

International Nuclear Information System (INIS)

Ku, L.P.; Kolibal, J.G.

1982-07-01

By altering the collision mechanism in the numerical transport calculations, and by constructing an analytical model based on age-diffusion theory, the outstanding feature in the life history of D-D fusion neutrons penetrating deeply into ordinary concrete is shown to be the transport in the 2.3 MeV oxygen anti-resonance. This result is used to assess the impact of the cross-section uncertainties and the uncertainties due to variations in the D-D fusion spectrum and temperature

Neutron lifetime well logging methods and apparatus

International Nuclear Information System (INIS)

Paap, H.J.; Pitts, R.W.

1974-01-01

A method for investigating the earth formations surrounding a well borehole, comprising the steps of: continuously generating high energy neutrons in the borehole and bombarding the surrounding media with such neutrons to develop a cloud of thermal neutrons therein; modulating the intensity of said high energy neutrons harmonically as a function of time in order to intensity modulate said cloud of thermal neutrons as a function of time; and measuring a time-dependant thermal neutron characteristic of said intensity modulated cloud of thermal neutrons

The solution of the multigroup neutron transport equation using spherical harmonics

International Nuclear Information System (INIS)

Fletcher, K.

1981-01-01

A solution of the multi-group neutron transport equation in up to three space dimensions is presented. The flux is expanded in a series of unnormalised spherical harmonics. Using the various recurrence formulae a linked set of first order differential equations is obtained for the moments psisup(g)sub(lm)(r), γsup(g)sub(lm)(r). Terms with odd l are eliminated resulting in a second order system which is solved by two methods. The first is a finite difference formulation using an iterative procedure, secondly, in XYZ and XY geometry a finite element solution is given. Results for a test problem using both methods are exhibited and compared. (orig./RW) [de

Exact analytical solution of time-independent neutron transport equation, and its applications to systems with a point source

International Nuclear Information System (INIS)

Mikata, Y.

2014-01-01

Highlights: • An exact solution for the one-speed neutron transport equation is obtained. • This solution as well as its derivation are believed to be new. • Neutron flux for a purely absorbing material with a point neutron source off the origin is obtained. • Spherically as well as cylindrically piecewise constant cross sections are studied. • Neutron flux expressions for a point neutron source off the origin are believed to be new. - Abstract: An exact analytical solution of the time-independent monoenergetic neutron transport equation is obtained in this paper. The solution is applied to systems with a point source. Systematic analysis of the solution of the time-independent neutron transport equation, and its applications represent the primary goal of this paper. To the best of the author’s knowledge, certain key results on the scalar neutron flux as well as their derivations are new. As an application of these results, a scalar neutron flux for a purely absorbing medium with a spherically piecewise constant cross section and an isotropic point neutron source off the origin as well as that for a cylindrically piecewise constant cross section with a point neutron source off the origin are obtained. Both of these results are believed to be new

Neutron transport calculations of some fast critical assemblies

Energy Technology Data Exchange (ETDEWEB)

Martinez-Val Penalosa, J A

1976-07-01

To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs.

Neutron transport calculations of some fast critical assemblies

International Nuclear Information System (INIS)

Martinez-Val Penalosa, J. A.

1976-01-01

To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs

Study of influence of transport performance of the neutron guide

International Nuclear Information System (INIS)

Li Xinxi; Wang Yan; Huang Chaoqiang; Chen Bo; Chen Liang

2009-01-01

For the sake of improving the performance of the neutron scattering instrument, usually we need use the neutron guide, it's very important to select the right type and optimizing of neutron guide. The papers calculate the focus neutron guide and the single channel neutron guide by numeric method. The results shows that the choice of neutron guide should consult the resolution requirement of neutron scattering instrument, and the length of the neutron guide should be optimized. The calculation results can be the theoretical reference for the design of neutron scattering instrument. (authors)

Applying the response matrix method for solving coupled neutron diffusion and transport problems

International Nuclear Information System (INIS)

Sibiya, G.S.

1980-01-01

The numerical determination of the flux and power distribution in the design of large power reactors is quite a time-consuming procedure if the space under consideration is to be subdivided into very fine weshes. Many computing methods applied in reactor physics (such as the finite-difference method) require considerable computing time. In this thesis it is shown that the response matrix method can be successfully used as an alternative approach to solving the two-dimension diffusion equation. Furthermore it is shown that sufficient accuracy of the method is achieved by assuming a linear space dependence of the neutron currents on the boundaries of the geometries defined for the given space. (orig.) [de

Calculation of neutron shielding using an unidimensional model of transportation in formulation of discrete ordinates with scattering linearly anisotropic and a speed

International Nuclear Information System (INIS)

Libotte, Rafael Barbosa; Alves Filho, Hermes; Oliva, Amaury Muñoz

2017-01-01

The physical phenomenon of transport of neutral particles in a host environment is of interest in various scientific applications, e.g., nuclear reactors, shielding calculations, radiological protection, nuclear medicine, agronomy, materials science, oil prospecting, etc. In all these areas there is a need for an accurate description of the transport of the particles in the host medium. In this class of applications are the neutron shielding problems, also referred to as 'fixed-source' problems, where the interaction of the particles with the medium does not produce new neutrons, i.e., non-multiplicative medium. In this context, the development of tools that model these problems is relevant and of a beneficial return to society. In this work, we propose the development of deterministic mathematical and computational modeling of neutron transport using the linearized equation of Boltzmann applied to neutron shielding problems. Here we present also the development of a spectro-nodal method (coarse mesh) considering the scattering phenomenon as being linearly anisotropic. We show the results using a computational application, developed in Java language, version 1.8.0 9 1

Method of solution of the neutron transport equation in multidimensional cartesian geometries using spherical harmonics and spatially orthogonal polynomials

International Nuclear Information System (INIS)

Fenstermacher, T.E.

1981-01-01

The solution of the neutron transport equation has long been a subject of intense interest to nuclear engineers. Present computer codes for the solution of this equation, however, are expensive to run for large, multidimensional problems, and also suffer from computational problems such as the ray effect. A method has been developed which eliminates many of these problems. It consists of transforming the transport equation into a set of linear partial differential equations by the use of spherical harmonics. The problem volume is divided into mesh boxes, and the flux components are approximated within each mesh box by spatially orthogonal quadratic polynomials, which need not be continuous at mesh box interfaces. A variational principle is developed, and used to solve for the unknown coefficients of these polynomials. Both one dimensional and two dimensional computer codes using this method have been written. The codes have each been tested on several test cases, and the solutions checked against solutions obtained by other methods. While the codes have some difficulty in modeling sharp transients, they produce excellent results on problems where the characteristic lengths are many mean free paths. On one test case, the two dimensional code, SHOP/2D, required only one-fourth the computer time required by the finite difference, discrete ordinates code TWOTRAN to produce a solution. In addition, SHOP/2D converged much better than TWOTRAN and produced more physical-appearing results

FMCEIR: a Monte Carlo program for solving the stationary neutron and gamma transport equation

International Nuclear Information System (INIS)

Taormina, A.

1978-05-01

FMCEIR is a three-dimensional Monte Carlo program for solving the stationary neutron and gamma transport equation. It is used to study the problem of neutron and gamma streaming in the GCFR and HHT reactor channels. (G.T.H.)

Monte Carlo calculation of correction factors for radionuclide neutron source emission rate measurement by manganese bath method

International Nuclear Information System (INIS)

Li Chunjuan; Liu Yi'na; Zhang Weihua; Wang Zhiqiang

2014-01-01

The manganese bath method for measuring the neutron emission rate of radionuclide sources requires corrections to be made for emitted neutrons which are not captured by manganese nuclei. The Monte Carlo particle transport code MCNP was used to simulate the manganese bath system of the standards for the measurement of neutron source intensity. The correction factors were calculated and the reliability of the model was demonstrated through the key comparison for the radionuclide neutron source emission rate measurements organized by BIPM. The uncertainties in the calculated values were evaluated by considering the sensitivities to the solution density, the density of the radioactive material, the positioning of the source, the radius of the bath, and the interaction cross-sections. A new method for the evaluation of the uncertainties in Monte Carlo calculation was given. (authors)

Neutron exposure

International Nuclear Information System (INIS)

Prillinger, G.; Konynenburg, R.A. van

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 6, LWR-PV neutron transport calculations and dosimetry methods and how they are combined to evaluate the neutron exposure of the steel of pressure vessels are discussed. An effort to correlate neutron exposure parameters with damage is made

Influence of Neutron Spectra Unfolding Method on Fast Neutron Dose Determination

International Nuclear Information System (INIS)

Marinkovic, P.

1991-01-01

Full text: Accuracy of knowing the fast neutron spectra has great influence on equivalent dose determination. In usual fast neutron spectrum measurements with scintillation detectors based on proton recoil, the main difficulty is confidence of unfolding method. In former ones variance of obtained result is usually great and negative values are possible too, which does means that we don't now exactly is obtained neutron spectrum real one. The new unfolding method based on Shanon's information theory, which gives non-negative spectrum and relative low variance, is obtained and appropriate numerical code for application in fast neutron spectrometry based on proton recoil is realized. In this method principle of maximum entropy and maximum likelihood are used together. Unknown group density distribution functions, which are considered as desired normalized mean neutron group flux, are constl u cted using only constrain of knowing mean value. Obtained distributions are consistent to available information (counts in NCA from proton recoil), while being maximally noncommittal with respect to all other unknown circumstances. For maximum likelihood principle, distribution functions around mean value of counts in the channels of MCA are taken to be Gauss function shape. Optimal non-negative solution is searched by means of Lagrange parameter method. Nonlinear system of equations, is solved using gradient and Newton iterative algorithm. Error covariance matrix is obtained too. (author)

Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

International Nuclear Information System (INIS)

Ganapol, B.D.; Kornreich, D.E.

1997-01-01

Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade

Development of three-dimensional transport code by the double finite element method

International Nuclear Information System (INIS)

Fujimura, Toichiro

1985-01-01

Development of a three-dimensional neutron transport code by the double finite element method is described. Both of the Galerkin and variational methods are adopted to solve the problem, and then the characteristics of them are compared. Computational results of the collocation method, developed as a technique for the vaviational one, are illustrated in comparison with those of an Ssub(n) code. (author)

A time-dependent neutron transport model and its coupling to thermal-hydraulics

International Nuclear Information System (INIS)

Pautz, A.

2001-01-01

A new neutron transport code for time-dependent analyses of nuclear systems has been developed. The code system is based on the well-known Discrete Ordinates code DORT, which solves the steady-state neutron/photon transport equation in two dimensions for an arbitrary number of energy groups and the most common regular geometries. For the implementation of time-dependence a fully implicit first-order scheme was employed to minimize errors due to temporal discretization. This requires various modifications to the transport equation as well as the extensive use of elaborated acceleration mechanisms. The convergence criteria for fluxes, fission rates etc. had to be strongly tightened to ensure the reliability of results. To perform coupled analyses, an interface to the GRS system code ATHLET has been developed. The nodal power densities from the neutron transport code are passed to ATHLET to calculate thermal-hydraulic system parameters, e.g. fuel and coolant temperatures. These are in turn used to generate appropriate nuclear cross sections by interpolation of pre-calculated data sets for each time step. Finally, to demonstrate the transient capabilities of the coupled code system, the research reactor FRM-II has been analysed. Several design basis accidents were modelled, like the loss of off site power, loss of secondary heat sink and unintended control rod withdrawal. (author)

Removal, transportation and disposal of the Millstone 2 neutron thermal shield

International Nuclear Information System (INIS)

Snedeker, D.F.; Thomas, L.S.; Schmoker, D.S.; Cade, M.S.

1985-01-01

Some PWR reactors equipped with neutron thermal shields (NTS) have experienced severe neutron shield degradation to the extent that removal and disposal of these shields has become necessary. Due to the relative size and activation levels of the thermal shield, disposal techniques, remote material handling and transportation equipment must be carefully evaluated to minimize plant down time and maintain disposal costs at a minimum. This paper describes the techniques, equipment and methodology employed in the removal, transportation and disposal of the NTS at the Millstone 2 Nuclear Generating Station, a PWR facility owned and operated by Northeast Utilities of Hartford, CT. Specific areas addressed include: (1) remote underwater equipment and tooling for use in segmenting and loading the thermal shield in a disposal liner; (2) adaptation of the General Electric IF-300 Irradiated Fuel Cask for transportation of the NTS for disposal; (3) equipment and techniques used for cask handling and liner burial at the Low Level Radioactive Waste (LLRW) disposal facility

Proposal of a New Method for Neutron Dosimetry Based on Spectral Information Obtained by Application of Artificial Neural Networks

International Nuclear Information System (INIS)

Fehrenbacher, G.; Schuetz, R.; Hahn, K.; Sprunck, M.; Cordes, E.; Biersack, J.P.; Wahl, W.

1999-01-01

A new method for the monitoring of neutron radiation is proposed. It is based on the determination of spectral information on the neutron field in order to derive dose quantities like the ambient dose equivalent, the dose equivalent, or other dose quantities which depend on the neutron energy. The method uses a multi-element system consisting of converter type silicon detectors. The unfolding procedure is based on an artificial neural network (ANN). The response function of each element is determined by a computational model considering the neutron interaction with the dosemeter layers and the subsequent transport of produced ions. An example is given for a multi-element system. The ANN is trained by a given set of neutron spectra and then applied to count responses obtained in neutron fields. Four examples of spectra unfolded using the ANN are presented. (author)

Rapid Measurement of Neutron Dose Rate for Transport Index

International Nuclear Information System (INIS)

Morris, R.L.

2000-01-01

A newly available neutron dose equivalent remmeter with improved sensitivity and energy response has been put into service at Rocky Flats Environmental Technology Site (RFETS). This instrument is being used to expedite measurement of the Transport Index and as an ALARA tool to identify locations where slightly elevated neutron dose equivalent rates exist. The meter is capable of measuring dose rates as low as 0.2 μSv per hour (20 μrem per hour). Tests of the angular response and energy response of the instrument are reported. Calculations of the theoretical instrument response made using MCNPtrademark are reported for materials typical of those being shipped

Comparison of analytical transport and stochastic solutions for neutron slowing down in an infinite medium

International Nuclear Information System (INIS)

Jahshan, S.N.; Wemple, C.A.; Ganapol, B.D.

1993-01-01

A comparison of the numerical solutions of the transport equation describing the steady neutron slowing down in an infinite medium with constant cross sections is made with stochastic solutions obtained from tracking successive neutron histories in the same medium. The transport equation solution is obtained using a numerical Laplace transform inversion algorithm. The basis for the algorithm is an evaluation of the Bromwich integral without analytical continuation. Neither the transport nor the stochastic solution is limited in the number of scattering species allowed. The medium may contain an absorption component as well. (orig.)

Neutron transport solver parallelization using a Domain Decomposition method

International Nuclear Information System (INIS)

Van Criekingen, S.; Nataf, F.; Have, P.

2008-01-01

A domain decomposition (DD) method is investigated for the parallel solution of the second-order even-parity form of the time-independent Boltzmann transport equation. The spatial discretization is performed using finite elements, and the angular discretization using spherical harmonic expansions (P N method). The main idea developed here is due to P.L. Lions. It consists in having sub-domains exchanging not only interface point flux values, but also interface flux 'derivative' values. (The word 'derivative' is here used with quotes, because in the case considered here, it in fact consists in the Ω.∇ operator, with Ω the angular variable vector and ∇ the spatial gradient operator.) A parameter α is introduced, as proportionality coefficient between point flux and 'derivative' values. This parameter can be tuned - so far heuristically - to optimize the method. (authors)

Measurement and analysis of fast neutron spectra in reactor materials by time-of-flight method

International Nuclear Information System (INIS)

Hayashi, Shuhei; Kimura, Itsuro; Kobayashi, Shohei; Yamamoto, Shuji; Nishihara, Hiroshi.

1982-01-01

The LINAC-TOF experiments have been done to measure the neutron energy spectra in the assemblies of reactor materials. The sample materials to be measured were iron, stainless steel, aluminum, nickel, zirconium, thorium, lithium, and so on. The shapes of assemblies were piles (rectangular parallelopiped, sphere, and polyhedron) and slab. A photoneutron target was set at the center of the pile assemblies. Each assembly has an electron injection hole and a re-entrant hole. In case of a slab, a photo neutron target was placed at the outside of the slab. Neutrons were generated by using an electron linear accelerator (LINAC). The length of the flight path was 20 m. The neutron detectors were a Li-6 glass scintillator and a B-10 vaseline-NaI(Tl) scintillator. The spatial distributions of neutrons in the piles were measured by the foil activation method. The neutron transport calculation was performed, and the evaluation of group constants was made. (Kato, T.)

Calculated characteristics of subcritical assembly with anisotropic transport of neutrons

International Nuclear Information System (INIS)

Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.

2003-01-01

There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)

Large subcriticality measurement by pulsed neutron method

International Nuclear Information System (INIS)

Yamane, Y.; Yoshida, A.; Nishina, K.; Kobayashi, K.; Kanda, K.

1985-01-01

To establish the method determining large subcriticalities in the field of nuclear criticality safety, the authors performed pulsed neutron experiments using the Kyoto University Critical Assembly (KUCA) at Research Reactor Institute, Kyoto University and the Cockcroft-Walton type accelerator attached to the assembly. The area-ratio method proposed by Sjoestrand was employed to evaluate subcriticalities from neutron decay curves measured. This method has the shortcomings that the neutron component due to a decay of delayed neutrons remarkably decreases as the subcriticality of an objective increases. To overcome the shortcoming, the authors increased the frequency of pulsed neutron generation. The integral-version of the area-ratio method proposed by Kosaly and Fisher was employed in addition in order to remove a contamination of spatial higher modes from the decay curve. The latter becomes significant as subcriticality increases. The largest subcriticality determined in the present experiments was 125.4 dollars, which was equal to 0.5111 in a multiplication factor. The calculational values evaluated by the computer code KENO-IV with 137 energy groups based on the Monte Carlo method agreed well with those experimental values

Radiation transport methods for nuclear log assessment - an overview

International Nuclear Information System (INIS)

Badruzzaman, A.

1996-01-01

Methods of radiation transport have been applied to well-logging problems with nuclear sources since the early 1960s. Nuclear sondes are used in identifying rock compositions and fluid properties in reservoirs to predict the porosity and oil saturation. Early computational effort in nuclear logging used diffusion techniques. As computers became more powerful, deterministic transport methods and, finally, Monte Carlo methods were applied to solve these problems in three dimensions. Recently, the application has been extended to problems with a new generation of devices, including spectroscopic sondes that measure such quantities as the carbon/oxygen ratio to predict oil saturation and logging-while-drilling (LWD) sondes that take neutron and gamma measurements as they rotate in the borehole. These measurements present conditions that will be difficult to calibrate in the laboratory

Criticality analysis of thermal reactors for two energy groups applying Monte Carlo and neutron Albedo method

International Nuclear Information System (INIS)

Terra, Andre Miguel Barge Pontes Torres

2005-01-01

The Albedo method applied to criticality calculations to nuclear reactors is characterized by following the neutron currents, allowing to make detailed analyses of the physics phenomena about interactions of the neutrons with the core-reflector set, by the determination of the probabilities of reflection, absorption, and transmission. Then, allowing to make detailed appreciations of the variation of the effective neutron multiplication factor, keff. In the present work, motivated for excellent results presented in dissertations applied to thermal reactors and shieldings, was described the methodology to Albedo method for the analysis criticality of thermal reactors by using two energy groups admitting variable core coefficients to each re-entrant current. By using the Monte Carlo KENO IV code was analyzed relation between the total fraction of neutrons absorbed in the core reactor and the fraction of neutrons that never have stayed into the reflector but were absorbed into the core. As parameters of comparison and analysis of the results obtained by the Albedo method were used one dimensional deterministic code ANISN (ANIsotropic SN transport code) and Diffusion method. The keff results determined by the Albedo method, to the type of analyzed reactor, showed excellent agreement. Thus were obtained relative errors of keff values smaller than 0,78% between the Albedo method and code ANISN. In relation to the Diffusion method were obtained errors smaller than 0,35%, showing the effectiveness of the Albedo method applied to criticality analysis. The easiness of application, simplicity and clarity of the Albedo method constitute a valuable instrument to neutronic calculations applied to nonmultiplying and multiplying media. (author)

Application of the three-dimensional transport code to analysis of the neutron streaming experiment

International Nuclear Information System (INIS)

Chatani, K.; Slater, C.O.

1990-01-01

The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan

TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: TDTORT solves the time-dependent, three-dimensional neutron transport equation with explicit representation of delayed neutrons to estimate the fission yield from fissionable material transients. This release includes a modified version of TORT from the C00650MFMWS01 DOORS3.1 code package plus the time-dependent TDTORT code. GIP is also included for cross-section preparation. TORT calculates the flux or fluence of particles due to particles incident upon the system from extraneous sources or generated internally as a result of interaction with the system in two- or three-dimensional geometric systems. The principle application is to the deep-penetration transport of neutrons and photons. Reactor eigenvalue problems can also be solved. Numerous printed edits of the results are available, and results can be transferred to output files for subsequent analysis. TDTORT reads ANISN-format cross-section libraries, which are not included in the package. Users may choose from several available in RSICC's data library collection which can be identified by the keyword 'ANISN FORMAT'. 2 - Methods:The time-dependent spatial flux is expressed as a product of a space-, energy-, and angle-dependent shape function, which is usually slowly varying in time and a purely time-dependent amplitude function. The shape equation is solved for the shape using TORT; and the result is used to calculate the point kinetics parameters (e.g., reactivity) by using their inner product definitions, which are then used to solve the time-dependent amplitude and precursor equations. The amplitude function is calculated by solving the kinetics equations using the LSODE solver. When a new shape calculation is needed, the flux is calculated using the newly computed amplitude function. The Boltzmann transport equation is solved using the method of discrete ordinates to treat the directional variable and weighted finite-difference methods, in addition to Linear Nodal

Neutron diffraction utilizing the T-O-F method

Energy Technology Data Exchange (ETDEWEB)

Niimura, N [Tohoku Univ., Sendai (Japan). Lab. of Nuclear Science

1974-12-01

Characteristic features of the TOF (time of flight) neutron diffraction are summarized. In this method, i) all the reciprocal points on the rod passing through the origin in the reciprocal space can be scanned by each burst of white neutrons, ii) it is easy to measure high index reflections at the large scattering angle, iii) each reflection is not affected by the higher-order harmonics, and iv) it is easy to measure the physical properties depending on the neutron wavelength. The pulse neutron generator as well as the data acquisition system in the Laboratory of Nuclear Science of Tohoku University is described. The TOF method seems to be very powerful if it is applied to accurate structure analysis. The data correction methods are discussed. The TOF method is prospective to the study of transient phenomena. In this method one can apply to the crystalline sample an external field pulsed with the same frequency as the neutrons. By using this method, the transient state of the polarization reversal of the ferroelectric NaNO/sub 2/ has been observed. The magnetically pulsed neutron TOF spectrometer is briefly introduced after a review of the chopper history.

Constitutive laws for the neutron density current

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Morales-Sandoval, Jaime B.; Vazquez-Rodriguez, Rodolfo; Espinosa-Martinez, Erick-G.

2008-01-01

In this technical note, a fractional wave equation for the average neutron motion in nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fickian and non-Fickian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional exponent that can be manipulated to obtain the best representation of the neutron transport phenomena. The detrended fluctuation analysis (DFA) method is presented in this paper to estimate the fractional exponent

Charged-particle calculations using Boltzmann transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Dodds, H.L. Jr.; Robinson, M.T.; Holmes, D.K.

1981-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of charged particle range distributions, reflection coefficients, and sputtering yields. The Boltzmann transport approach can be implemented, with minor changes, in standard neutral particle computer codes. With the multigroup discrete ordinates code, ANISN, determination of ion and target atom distributions as functions of position, energy, and direction can be obtained without the stochastic error associated with atomistic computer codes such as MARLOWE and TRIM. With the multigroup Monte Carlo code, MORSE, charged particle effects can be obtained for problems associated with very complex geometries. Results are presented for several charged particle problems. Good agreement is obtained between quantities calculated with the multigroup approach and those obtained experimentally or by atomistic computer codes

Neutron therapy coupling brachytherapy and boron neutron capture therapy (BNCT) techniques

International Nuclear Information System (INIS)

Chaves, Iara Ferreira.

1994-12-01

In the present dissertation, neutron radiation techniques applied into organs of the human body are investigated as oncologic radiation therapy. The proposal treatment consists on connecting two distinct techniques: Boron Neutron Capture Therapy (BNCT) and irradiation by discrete sources of neutrons, through the brachytherapy conception. Biological and radio-dosimetrical aspects of the two techniques are considered. Nuclear aspects are discussed, presenting the nuclear reactions occurred in tumoral region, and describing the forms of evaluating the dose curves. Methods for estimating radiation transmission are reviewed through the solution of the neutron transport equation, Monte Carlo methodology, and simplified analytical calculation based on diffusion equation and numerical integration. The last is computational developed and presented as a quickly way to neutron transport evaluation in homogeneous medium. The computational evaluation of the doses for distinct hypothetical situations is presented, applying the coupled techniques BNTC and brachytherapy as an possible oncologic treatment. (author). 78 refs., 61 figs., 21 tabs

Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

1987-02-01

Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.

1988-01-01

The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs

Generic programming for deterministic neutron transport codes

International Nuclear Information System (INIS)

Plagne, L.; Poncot, A.

2005-01-01

This paper discusses the implementation of neutron transport codes via generic programming techniques. Two different Boltzmann equation approximations have been implemented, namely the Sn and SPn methods. This implementation experiment shows that generic programming allows us to improve maintainability and readability of source codes with no performance penalties compared to classical approaches. In the present implementation, matrices and vectors as well as linear algebra algorithms are treated separately from the rest of source code and gathered in a tool library called 'Generic Linear Algebra Solver System' (GLASS). Such a code architecture, based on a linear algebra library, allows us to separate the three different scientific fields involved in transport codes design: numerical analysis, reactor physics and computer science. Our library handles matrices with optional storage policies and thus applies both to Sn code, where the matrix elements are computed on the fly, and to SPn code where stored matrices are used. Thus, using GLASS allows us to share a large fraction of source code between Sn and SPn implementations. Moreover, the GLASS high level of abstraction allows the writing of numerical algorithms in a form which is very close to their textbook descriptions. Hence the GLASS algorithms collection, disconnected from computer science considerations (e.g. storage policy), is very easy to read, to maintain and to extend. (authors)

Extension of Applicability of integral neutron transport theory in reactor cell and core investigation

International Nuclear Information System (INIS)

Pop-Jordanov, J.; Bosevski, T.; Kocic, A.; Altiparmakov, D.

1980-01-01

A Space-Point Energy-Group integral transport theory method (SPEG) is developed and applied to the local and global calculations of the Yugoslav RA reactor. Compared to other integral transport theory methods, the SPEG distinguishes by (1) the arbitrary order of the polynomial, (2) the effective determination of integral parameters through point flux values, (3) the use of neutron balance condition. as a posterior measure of the accuracy of the calculation and (4) the elimination of the subdivisions- into zones, in realistic cases. In addition, different direct (collision probability) and indirect (Monte Carlo) approaches to integral transport theory have been investigated and Some effective acceleration procedures introduced. The study was performed on three test problems in plane and cylindrical geometry, as well as on the nine-region cell of the RA reactor. In particular, the limitations of the integral transport theory including its non-applicability to optically large material regions and to global reactor calculations were examined. The proposed strictly multipoint approach, avoiding the subdivision into zones and groups, seems to provide a good starting point to overcome these limitations of the integral transport theory. (author)

GPU-based high performance Monte Carlo simulation in neutron transport

Energy Technology Data Exchange (ETDEWEB)

Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br

2009-07-01

Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)