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)
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
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.)
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
International Nuclear Information System (INIS)
Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Coste-Delclaux, Mireille; M'Backe Diop, Cheikh; Nicolas, Anne; Andrieux, Catherine; Archier, Pascal; Baudron, Anne-Marie; Bernard, David; Biaise, Patrick; Blanc-Tranchant, Patrick; Bonin, Bernard; Bouland, Olivier; Bourganel, Stephane; Calvin, Christophe; Chiron, Maurice; Damian, Frederic; Dumonteil, Eric; Fausser, Clement; Fougeras, Philippe; Gabriel, Franck; Gagnier, Emmanuel; Gallo, Daniele; Hudelot, Jean-Pascal; Hugot, Francois-Xavier; Dat Huynh, Tan; Jouanne, Cedric; Lautard, Jean-Jacques; Laye, Frederic; Lee, Yi-Kang; Lenain, Richard; Leray, Sylvie; Litaize, Olivier; Magnaud, Christine; Malvagi, Fausto; Mijuin, Dominique; Mounier, Claude; Naury, Sylvie; Nicolas, Anne; Noguere, Gilles; Palau, Jean-Marc; Le Pallec, Jean-Charles; Peneliau, Yannick; Petit, Odile; Poinot-Salanon, Christine; Raepsaet, Xavier; Reuss, Paul; Richebois, Edwige; Roque, Benedicte; Royer, Eric; Saint-Jean, Cyrille de; Santamarina, Alain; Serot, Olivier; Soldevila, Michel; Tommasi, Jean; Trama, Jean-Christophe; Tsilanizara, Aime; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre
2013-10-01
This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality
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)
International Nuclear Information System (INIS)
Buckel, G.
1983-01-01
The objectives are the development, testing and cultivation of reliable, efficient and user-optimized neutron-physical calculation methods and conformity with users' requirements concerning design of power reactors, planning and analysis of experiments necessary for their protection as well as research on physical key problems. A short outline of available computing programmes for the following objectives is given: - Provision of macroscopic group constants, - Calculation of neutron flux distribution in transport theory and diffusion approximation, - Evaluation of neutron flux-distribution, - Execution of disturbance calculations for the determination reactivity coefficients, and - graphical representation of results. (orig./RW) [de
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)
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
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
General-purpose Monte Carlo codes for neutron and photon transport calculations. MVP version 3
International Nuclear Information System (INIS)
Nagaya, Yasunobu
2017-01-01
JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
International Nuclear Information System (INIS)
Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
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
Resolution of the neutron transport equation by massively parallel computer in the Cronos code
International Nuclear Information System (INIS)
Zardini, D.M.
1996-01-01
The feasibility of neutron transport problems parallel resolution by CRONOS code's SN module is here studied. In this report we give the first data about the parallel resolution by angular variable decomposition of the transport equation. Problems about parallel resolution by spatial variable decomposition and memory stage limits are also explained here. (author)
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
The neutron transport code DTF-Traca users manual and input data
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C
1979-07-01
This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.
The neutron transport code DTF-Traca users manual and input data
International Nuclear Information System (INIS)
Ahnert, C.
1979-01-01
This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs
The neutron transport code DTF-TRACA. User's manual and input data
International Nuclear Information System (INIS)
Anhert, C.
1979-01-01
A user's manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data description is given. The new options developped at JEN are included too. (author)
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)
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
International Nuclear Information System (INIS)
Kirk, B.L.; West, J.T.
1984-06-01
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided
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)
Spallation neutron production and the current intra-nuclear cascade and transport codes
International Nuclear Information System (INIS)
Filges, D.; Goldenbaum, F.
2001-01-01
A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models. (orig.)
Spallation neutron production and the current intra-nuclear cascade and transport codes
Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.
A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.
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
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)
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
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
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
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
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
International Nuclear Information System (INIS)
Deng Li; Xie Zhongsheng
1999-01-01
The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1979-11-01
The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
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
International Nuclear Information System (INIS)
Attaya, H.
1995-01-01
The primary goal of this task is to provide the capabilities in the activation code RACC, to treat pulsed operation modes. In addition, it is required that the code utilizes the same spatial mesh and geometrical models as employed in the one or multidimensional neutron transport codes used in ITER design. This would ensure the use of the same neutron flux generated by those codes to calculate the different activation parameters. It is also required to have the capabilities for generating graphical outputs for the calculated activation parameters
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1978-07-01
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables
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.
The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)
Energy Technology Data Exchange (ETDEWEB)
Rhoades, W.A.; Simpson, D.B.
1997-10-01
TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.
International Nuclear Information System (INIS)
Ahnert, C.; Aragones, J. M.
1981-01-01
This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs
International Nuclear Information System (INIS)
Ahnert, C.; Aragones, J.M.
1981-01-01
A user's manual of the neutron transport code Twotran-Traca is presented; it is a version of the original Twotran-GG from the Los Alamos Laboratory, with some modifications made at J.E.N., Spain. A detailed input data description is given as well as the new modifications developped at J.E.N. (author) [es
International Nuclear Information System (INIS)
Chauvet, Y.
1985-01-01
This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in neutron transport problems, the authors briefly describe the work done in order to get a vector code. Vectorization principles are presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples are presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example they propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion they prove that Monte Carlo algorithms are very well suited to future vector and parallel computers
Resonance self-shielding methodology of new neutron transport code STREAM
International Nuclear Information System (INIS)
Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung; Hong, Ser Gi
2015-01-01
This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)
International Nuclear Information System (INIS)
Chauvet, Y.
1985-01-01
This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in our neutron transport problems, we briefly describe the work we have done in order to get a vector code. Vectorization principles will be presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples will be presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example we propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion we prove that Monte Carlo algorithms are very well suited to future vector and parallel computers. (orig.)
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
International Nuclear Information System (INIS)
Cullen, D.E
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
Cullen, D
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.
Performance of a neutron transport code with full phase space decomposition on the Cray Research T3D
International Nuclear Information System (INIS)
Dorr, M.R.; Salo, E.M.
1995-01-01
We present performance results obtained on a 128-node Cray Research T3D computer by a neutron transport code implementing a standard mtiltigroup, discrete ordinates algorithm on a three-dimensional Cartesian grid. After summarizing the implementation strategy used to obtain a full decomposition of phase space (i.e., simultaneous parallelization of the neutron energy, directional and spatial variables), we investigate the scalability of the fundamental source iteration step with respect to each phase space variable. We also describe enhancements that have enabled performance rates approaching 10 gigaflops on the full 128-node machine
Monte Carlo codes use in neutron therapy
International Nuclear Information System (INIS)
Paquis, P.; Mokhtari, F.; Karamanoukian, D.; Pignol, J.P.; Cuendet, P.; Iborra, N.
1998-01-01
Monte Carlo calculation codes allow to study accurately all the parameters relevant to radiation effects, like the dose deposition or the type of microscopic interactions, through one by one particle transport simulation. These features are very useful for neutron irradiations, from device development up to dosimetry. This paper illustrates some applications of these codes in Neutron Capture Therapy and Neutron Capture Enhancement of fast neutrons irradiations. (authors)
International Nuclear Information System (INIS)
Clancy, B.E.
1986-01-01
This chapter begins with a neutron transport equation which includes the one dimensional plane geometry problems, the one dimensional spherical geometry problems, and numerical solutions. The section on the ANISN code and its look-alikes covers problems which can be solved; eigenvalue problems; outer iteration loop; inner iteration loop; and finite difference solution procedures. The input and output data for ANISN is also discussed. Two dimensional problems such as the DOT code are given. Finally, an overview of the Monte-Carlo methods and codes are elaborated on
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1975-10-01
The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level
International Nuclear Information System (INIS)
Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.
1976-01-01
The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components
International Nuclear Information System (INIS)
Maeng, YoungJae; Lim, MiJoung; Kim, KyungSik; Cho, YoungKi; Yoo, ChoonSung; Kim, ByoungChul
2015-01-01
Age-related degradation mechanisms are including the irradiation-assisted stress corrosion cracking(IASCC), void swelling, stress relaxation, fatigue, and etc. A lot of Baffle Former Bolts(BFBs) was installed at the former plate ends between baffle and barrel structure. These would undergo severe experiences, which are high temperature and pressure, bypass water flow and neutron exposure and have some radioactive limitation in inspecting their integrity. The objectives of this paper is to evaluate fast neutron fluence(n/cm 2 , E>1.0MeV) for PWR internals using 3D transport calculation code, RAPTOR-M3G, and to figure out a strategy to manage the effects of aging in PWR internals. One of age-related degradation mechanisms, IASCC, which is affected by fast neutron exposure rate, has been currently issued for PWR internals and has 2 x 10 21 (n/cm 2 ) of the threshold value by MRP-175. Because a lot of BFBs was installed around the internal components, closer inspections are required. As part of an aging management for Kori unit 2, 3D transport calculation code, RAPTOR-M3G, was applied for determining fast neutron fluence at baffle, barrel and former plates regions. As a result, the fast neutron fluence exceeds the screening or threshold values of IASCC in all of baffle, barrel and former plate region. And the most significant region is the baffle because it is located closest to the core. In addition, some regions including former plate tend to be more damaged because of less moderate ability than water. In conclusion, Ice's has been progressed for PWR internals of Kori unit 2. Several regions of internal components were damaged by fast neutron exposure and increase in size as time goes by
International Nuclear Information System (INIS)
Bogacz, J.; Loskiewicz, J.; Zazula, J.M.
1991-01-01
The use of universal neutron transport codes in order to calculate the parameters of well-logging probes presents a new approach first tried in U.S.A. and UK in the eighties. This paper deals with first such an attempt in Poland. The work is based on the use of MORSE code developed in Oak Ridge National Laboratory in U.S.A.. Using CG MORSE code we calculated neutron detector response when surrounded with sandstone of porosities 19% and 38%. During the work it come out that it was necessary to investigate different methods of estimation of the neutron flux. The stochastic estimation method as used currently in the original MORSE code (next collision approximation) can not be used because of slow convergence of its variance. Using the analog type of estimation (calculation of the sum of track lengths inside detector) we obtained results of acceptable variance (∼ 20%) for source-detector spacing smaller than 40 cm. The influence of porosity on detector response is correctly described for detector positioned at 27 cm from the source. At the moment the variances are quite large. (author). 33 refs, 8 figs, 8 tabs
A new philosophy for calibrating oil well logging tools based on neutron transport codes
International Nuclear Information System (INIS)
Butler, J.; Clayton, C.G.
1984-01-01
The current practice of calibrating neutron borehole logging probes is limited by an inability to match calibration conditions to those which pertain in an operational situation. In addition, test boreholes are expensive to construct and, when natural materials are used, rely on an exact correspondence in composition and in structure between the materials of the test facility and representative samples which may not be valid. Now that neutron tansport codes have been developed to a point at which they are able to cope with realistic, complex situations an alternative approach to calibration can be considered. The basis of this philosophy is the construction of a limited number of calibration facilities which are composed of artificial rocks of controlled but variable porosity and accurately known nuclear characteristics
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)
International Nuclear Information System (INIS)
Arita, Yutaka; Kihara, Yuji; Mitsuhasi, Junichi; Niita, Koji; Takai, Mikio; Ogawa, Izumi; Kishimoto, Tadafumi; Yoshihara, Tsutomu
2007-01-01
The simulation of a thermal-neutron-induced single-event upset (SEU) was performed on a 0.4-μm-design-rule 4 Mbit static random access memory (SRAM) using particle and heavy-ion transport code system (PHITS): The SEU rates obtained by the simulation were in very good agreement with the result of experiments. PHITS is a useful tool for simulating SEUs in semiconductor devices. To further improve the accuracy of the simulation, additional methods for tallying the energy deposition are required for PHITS. (author)
International Nuclear Information System (INIS)
Sabundjian, G.; Nakata, H.
1983-02-01
The neutronic calculational procedure in a 4,75% w/O enriched UO 2 fueled light water moderated critical assembly was tested, using the transport codes and diffusin code available at the Instituto de Pesquisas Energeticas e Nucleares. The results of the tested codes, LEOPARD, CITHAMMER, LASER, GELS and CITATION, were found to be satisfatory and only a slight advantage is presented by CITHAMMER code. (Author) [pt
International Nuclear Information System (INIS)
Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.
1976-01-01
The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components. Since 1964, the Center has been involved in the international exchange of information, encouraged and supported by both government and interagency agreements; and to achieve an equally viable and successful program in fusion research, the reciprocal exchange of CTR data and computing technology is encouraged and welcomed
International Nuclear Information System (INIS)
Masukawa, Fumihiro; Takano, Makoto; Naito, Yoshitaka; Yamazaki, Takao; Fujisaki, Masahide; Suzuki, Koichiro; Okuda, Motoi.
1993-11-01
In order to improve the accuracy and calculating speed of shielding analyses, MCNP 4, a Monte Carlo neutron and photon transport code system, has been parallelized and measured of its efficiency in the highly parallel distributed memory type computer, AP1000. The code has been analyzed statically and dynamically, then the suitable algorithm for parallelization has been determined for the shielding analysis functions of MCNP 4. This includes a strategy where a new history is assigned to the idling processor element dynamically during the execution. Furthermore, to avoid the congestion of communicative processing, the batch concept, processing multi-histories by a unit, has been introduced. By analyzing a sample cask problem with 2,000,000 histories by the AP1000 with 512 processor elements, the 82 % of parallelization efficiency is achieved, and the calculational speed has been estimated to be around 50 times as fast as that of FACOM M-780. (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
International Nuclear Information System (INIS)
Iwamoto, Y.; Ogawa, T.
2016-01-01
The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials
MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. 3 - Restrictions on the complexity of the problem: None noted
Shi, Xue-Ming; Peng, Xian-Jue
2016-09-01
Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.
International Nuclear Information System (INIS)
Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.
1977-07-01
The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)
High performance 3D neutron transport on peta scale and hybrid architectures within APOLLO3 code
International Nuclear Information System (INIS)
Jamelot, E.; Dubois, J.; Lautard, J-J.; Calvin, C.; Baudron, A-M.
2011-01-01
APOLLO3 code is a common project of CEA, AREVA and EDF for the development of a new generation system for core physics analysis. We present here the parallelization of two deterministic transport solvers of APOLLO3: MINOS, a simplified 3D transport solver on structured Cartesian and hexagonal grids, and MINARET, a transport solver based on triangular meshes on 2D and prismatic ones in 3D. We used two different techniques to accelerate MINOS: a domain decomposition method, combined with an accelerated algorithm using GPU. The domain decomposition is based on the Schwarz iterative algorithm, with Robin boundary conditions to exchange information. The Robin parameters influence the convergence and we detail how we optimized the choice of these parameters. MINARET parallelization is based on angular directions calculation using explicit message passing. Fine grain parallelization is also available for each angular direction using shared memory multithreaded acceleration. Many performance results are presented on massively parallel architectures using more than 103 cores and on hybrid architectures using some tens of GPUs. This work contributes to the HPC development in reactor physics at the CEA Nuclear Energy Division. (author)
International Nuclear Information System (INIS)
Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.
1977-03-01
TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided
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
International Nuclear Information System (INIS)
Calloo, A.A.
2012-01-01
In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice
Energy Technology Data Exchange (ETDEWEB)
Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.
2014-04-15
In this study, the molten salt-heavy metal mixtures 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % NpO{sub 2}, 93-85 % Li{sub 20}Sn{sub 80} + 5 % SFG-PuO{sub 2} and 2-10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion-fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. (orig.)
International Nuclear Information System (INIS)
Tyobeka, B.; Ivanov, K.; Pautz, A.
2007-01-01
In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters
LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory
International Nuclear Information System (INIS)
Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.
1976-04-01
The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented
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)
International Nuclear Information System (INIS)
White, Morgan C.
2000-01-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
Energy Technology Data Exchange (ETDEWEB)
White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
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
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
International Nuclear Information System (INIS)
Briesmeister, J.F.
1986-09-01
This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs
Modelling of the non-stationary thermal neutron transport in hydrogenous media using the MCNP code
International Nuclear Information System (INIS)
Wiacek, Urszula
2006-01-01
The work is aimed to compare results of the Monte Carlo simulations of pulsed neutron experiments with results of real experiments. The simulations have been performed for homogenous and two-zone system. In the first case the cylinder of Plexiglas has been used. In two-zone systems, Plexiglas has been used as an outer moderator and aqueous solutions of H 3 BO 3 or KCl of known concentrations have been used as the inner samples. The simulations have been performed (1) using for neutron scattering in Plexiglas a standard library for H in polyethylene (a commonly suggested way) and (2) using a modification of this library. The modification gives very good agreement between simulations and experimental results
Energy Technology Data Exchange (ETDEWEB)
Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko
2017-04-01
Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.
The CORSYS neutronics code system
International Nuclear Information System (INIS)
Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.
1994-01-01
The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs
Installation of Monte Carlo neutron and photon transport code system MCNP4
International Nuclear Information System (INIS)
Takano, Makoto; Sasaki, Mikio; Kaneko, Toshiyuki; Yamazaki, Takao.
1993-03-01
The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)
Vectorization and multitasking with a Monte-Carlo code for neutron transport problems
International Nuclear Information System (INIS)
Chauvet, Y.
1985-04-01
This paper summarizes two improvements of a Monte Carlo code by resorting to vectorization and multitasking techniques. After a short presentation of the physical problem to solve and a description of the main difficulties to produce an efficient coding, this paper introduces the vectorization principles employed and briefly describes how the vectorized algorithm works. Next, measured performances on CRAY 1S, CYBER 205 and CRAY X-MP are compared. The second part of this paper is devoted to multitasking technique. Starting from the standard multitasking tools available with FORTRAN on CRAY X-MP/4, a multitasked algorithm and its measured speed-ups are presented. In conclusion we prove that vector and parallel computers are a great opportunity for such Monte Carlo algorithms
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Leppaenen, J.
2007-01-01
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
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)
Assessment of neutron transport codes for application to CANDU fuel lattices analysis
International Nuclear Information System (INIS)
Roh, Gyu Hong; Choi, Hang Bok
1999-08-01
In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS
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
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.
Updated version of the DOT 4 one- and two-dimensional neutron/photon transport code
International Nuclear Information System (INIS)
Rhoades, W.A.; Childs, R.L.
1982-07-01
DOT 4 is designed to allow very large transport problems to be solved on a wide range of computers and memory arrangements. Unusual flexibilty in both space-mesh and directional-quadrature specification is allowed. For example, the radial mesh in an R-Z problem can vary with axial position. The directional quadrature can vary with both space and energy group. Several features improve performance on both deep penetration and criticality problems. The program has been checked and used extensively
International Nuclear Information System (INIS)
Vergnaud, T.; Nimal, J.C.; Ulpat, J.P.; Faucheux, G.
1988-01-01
The Tripoli-2 computer code has been adapted to calculate, in addition to energy deposit in matter by neutrons (Kerma) the energy deposit by gamma produced in neutronic impacts and the induced recoil electron current. The energy deposit conduces at air ionization, consequently at a conductibility. This knowledge added at that of electron current permit to resolve the Maxwell equations of electromagnetic field. The study is realized for an atmospheric explosion 100 meters high. The calculations of energy deposit and electron current have been conducted as far as 2.5km [fr
Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier
2018-01-01
Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using
Neutron transportation simulator
International Nuclear Information System (INIS)
Uenohara, Yuzo.
1995-01-01
In the present invention, problems in an existent parallelized monte carlo method is solved, and behaviors of neutrons in a large scaled system are accurately simulated at a high speed. Namely, a neutron transportation simulator according to the monte carlo method simulates movement of each of neutrons by using a parallel computer. In this case, the system to be processed is divided based on a space region and an energy region to which neutrons belong. Simulation of neutrons in the divided regions is allotted to each of performing devices of the parallel computer. Tarry data and nuclear data of the neutrons in each of the regions are memorized dispersedly to memories of each of the performing devices. A transmission means for simulating the behaviors of the neutrons in the region by each of the performing devices, as well as transmitting the information of the neutrons, when the neutrons are moved to other region, to the performing device in a transported portion are disposed to each of the performing devices. With such procedures, simulation for the neutrons in the allotted region can be conducted with small capacity of memories. (I.S.)
International Nuclear Information System (INIS)
Thomas, J.W.; Lee, H.C.; Downar, T.J.; Sofu, T.; Weber, D.P.; Joo, H.G.; Cho, J.Y.
2003-01-01
As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)
Neutron transport by TRIPOLI-2 code in the lower part of a PWR pit and in the pit-access cell
International Nuclear Information System (INIS)
Vergnaud, T.; Bourdet, L.; Gonnord, J.; Nimal, J.C.; Champion, G.
1984-09-01
The neutrons, which exit the reactor vessel, leak between the reactor vessel and the primary concrete shield and provide high dose rate in the lower part of the reactor pit. In this part of the reactor the biological shield is reduced at the level of the pit-access cell door and of the ventilation duct. Consequently there is a pin-point increase of dose rate at reactor place where access must be free during power operation. Studies concern 900 MWe and 1300 MWe plants; calculation results are compared with dose rate measurements which have been done at several interesting points of reactor. Neutron transport is studied in two steps: - first Monte Carlo calculation carried out by the TRIPOLI-2 system gives the current of neutrons entering into the pit-access cell; - a second calculation is about neutron transport in this cell; it is also performed with the TRIPOLI-2 code which uses a very realistic model for cell geometry. Then the door efficiency is evaluated by a SN one dimension code
Monte Carlo codes use in neutron therapy; Application de codes Monte Carlo en neutrontherapie
Energy Technology Data Exchange (ETDEWEB)
Paquis, P.; Mokhtari, F.; Karamanoukian, D. [Hopital Pasteur, 06 - Nice (France); Pignol, J.P. [Hopital du Hasenrain, 68 - Mulhouse (France); Cuendet, P. [CEA Centre d' Etudes de Saclay, 91 - Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Fares, G.; Hachem, A. [Faculte des Sciences, 06 - Nice (France); Iborra, N. [Centre Antoine-Lacassagne, 06 - Nice (France)
1998-04-01
Monte Carlo calculation codes allow to study accurately all the parameters relevant to radiation effects, like the dose deposition or the type of microscopic interactions, through one by one particle transport simulation. These features are very useful for neutron irradiations, from device development up to dosimetry. This paper illustrates some applications of these codes in Neutron Capture Therapy and Neutron Capture Enhancement of fast neutrons irradiations. (authors)
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C; Aragones, J M
1981-07-01
This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs.
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro
2007-02-01
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
International Nuclear Information System (INIS)
Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog
2005-03-01
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too
Energy Technology Data Exchange (ETDEWEB)
Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog
2005-03-15
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.
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)
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)
International Nuclear Information System (INIS)
Cai, Li
2014-01-01
In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core
International Nuclear Information System (INIS)
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki
2005-06-01
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)
Current status of high energy nucleon-meson transport code
Energy Technology Data Exchange (ETDEWEB)
Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)
Monte Carlo code for neutron radiography
International Nuclear Information System (INIS)
Milczarek, Jacek J.; Trzcinski, Andrzej; El-Ghany El Abd, Abd; Czachor, Andrzej
2005-01-01
The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms
Monte Carlo code for neutron radiography
Energy Technology Data Exchange (ETDEWEB)
Milczarek, Jacek J. [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)]. E-mail: jjmilcz@cyf.gov.pl; Trzcinski, Andrzej [Institute for Nuclear Studies, Swierk, 05-400 Otwock (Poland); El-Ghany El Abd, Abd [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland); Nuclear Research Center, PC 13759, Cairo (Egypt); Czachor, Andrzej [Institute of Atomic Energy, Swierk, 05-400 Otwock (Poland)
2005-04-21
The concise Monte Carlo code, MSX, for simulation of neutron radiography images of non-uniform objects is presented. The possibility of modeling the images of objects with continuous spatial distribution of specific isotopes is included. The code can be used for assessment of the scattered neutron component in neutron radiograms.
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)
Energy Technology Data Exchange (ETDEWEB)
Liu, T.; Ding, A.; Ji, W.; Xu, X. G. [Nuclear Engineering and Engineering Physics, Rensselaer Polytechnic Inst., Troy, NY 12180 (United States); Carothers, C. D. [Dept. of Computer Science, Rensselaer Polytechnic Inst. RPI (United States); Brown, F. B. [Los Alamos National Laboratory (LANL) (United States)
2012-07-01
Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of {approx}2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)
International Nuclear Information System (INIS)
Liu, T.; Ding, A.; Ji, W.; Xu, X. G.; Carothers, C. D.; Brown, F. B.
2012-01-01
Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of ∼2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)
Computer codes for neutron data evaluation
International Nuclear Information System (INIS)
Nakagawa, Tsuneo
1979-01-01
Data compilation codes such as NESTOR and REPSTOR, and NDES (Neutron Data Evaluation System) are mainly discussed. NDES is a code for neutron data evaluation using a TSS terminal, TEKTRONIX 4014. Users of NDES can perform plotting of data and calculation with nuclear models under conversational mode. (author)
International Nuclear Information System (INIS)
Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.
2011-01-01
Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (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)
High Energy Transport Code HETC
International Nuclear Information System (INIS)
Gabriel, T.A.
1985-09-01
The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs
DEFF Research Database (Denmark)
Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani
2014-01-01
Packet losses in wireless networks dramatically curbs the performance of TCP. This paper introduces a simple coding shim that aids IP-layer traffic in lossy environments while being transparent to transport layer protocols. The proposed coding approach enables erasure correction while being...... oblivious to the congestion control algorithms of the utilised transport layer protocol. Although our coding shim is indifferent towards the transport layer protocol, we focus on the performance of TCP when ran on top of our proposed coding mechanism due to its widespread use. The coding shim provides gains...
Spatial neutron kinetic module of ROSA code
International Nuclear Information System (INIS)
Cherezov, A.L.; Shchukin, N.V.
2009-01-01
A spatial neutron kinetic module was developed for computer code ROSA. The paper describes a numerical scheme used in the module for resolving neutron kinetic equations. Analytical integration for delayed neutrons emitters method and direct numerical integration method (Gear's method) were analyzed. The two methods were compared on their efficiency and accuracy. Both methods were verified with test problems. The results obtained in the verification studies were presented [ru
Fast neutron analysis code SAD1
International Nuclear Information System (INIS)
Jung, M.; Ott, C.
1985-01-01
A listing and an example of outputs of the M.C. code SAD1 are given here. This code has been used many times to predict responses of fast neutrons in hydrogenic materials (in our case emulsions or plastics) towards the elastic n, p scattering. It can be easily extended to other kinds of such materials and to any kind of incident fast neutron spectrum
SRAC95; general purpose neutronics code system
International Nuclear Information System (INIS)
Okumura, Keisuke; Tsuchihashi, Keichiro; Kaneko, Kunio.
1996-03-01
SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author)
SRAC95; general purpose neutronics code system
Energy Technology Data Exchange (ETDEWEB)
Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio
1996-03-01
SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).
The MC21 Monte Carlo Transport Code
International Nuclear Information System (INIS)
Sutton TM; Donovan TJ; Trumbull TH; Dobreff PS; Caro E; Griesheimer DP; Tyburski LJ; Carpenter DC; Joo H
2007-01-01
MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities
International Nuclear Information System (INIS)
Tzika, F.; Stamatelatos, I.E.
2004-01-01
Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample
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
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
CALTRANS: A parallel, deterministic, 3D neutronics code
Energy Technology Data Exchange (ETDEWEB)
Carson, L.; Ferguson, J.; Rogers, J.
1994-04-01
Our efforts to parallelize the deterministic solution of the neutron transport equation has culminated in a new neutronics code CALTRANS, which has full 3D capability. In this article, we describe the layout and algorithms of CALTRANS and present performance measurements of the code on a variety of platforms. Explicit implementation of the parallel algorithms of CALTRANS using both the function calls of the Parallel Virtual Machine software package (PVM 3.2) and the Meiko CS-2 tagged message passing library (based on the Intel NX/2 interface) are provided in appendices.
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)
Comparison of neutron spectrum unfolding codes
International Nuclear Information System (INIS)
Zijp, W.
1979-02-01
This final report contains a set of four ECN-reports. The first is dealing with the comparison of the neutron spectrum unfolding codes CRYSTAL BALL, RFSP-JUL, SAND II and STAY'SL. The other three present the results of calculations about the influence of statistical weights in CRYSTAL BALL, SAND II and RFSP-JUL
TRIPOLI-4: Monte Carlo transport code functionalities and applications
International Nuclear Information System (INIS)
Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.
2003-01-01
Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)
Validation of comprehensive space radiation transport code
International Nuclear Information System (INIS)
Shinn, J.L.; Simonsen, L.C.; Cucinotta, F.A.
1998-01-01
The HZETRN code has been developed over the past decade to evaluate the local radiation fields within sensitive materials on spacecraft in the space environment. Most of the more important nuclear and atomic processes are now modeled and evaluation within a complex spacecraft geometry with differing material components, including transition effects across boundaries of dissimilar materials, are included. The atomic/nuclear database and transport procedures have received limited validation in laboratory testing with high energy ion beams. The codes have been applied in design of the SAGE-III instrument resulting in material changes to control injurious neutron production, in the study of the Space Shuttle single event upsets, and in validation with space measurements (particle telescopes, tissue equivalent proportional counters, CR-39) on Shuttle and Mir. The present paper reviews the code development and presents recent results in laboratory and space flight validation
A guide to the AUS modular neutronics code system
International Nuclear Information System (INIS)
Robinson, G.S.
1987-04-01
A general description is given of the AUS modular neutronics code system, which may be used for calculations of a very wide range of fission reactors, fusion blankets and other neutron applications. The present system has cross-section libraries derived from ENDF/B-IV and includes modules which provide for lattice calculations, one-dimensional transport calculations, and one, two, and three-dimensional diffusion calculations, burnup calculations and the flexible editing of results. Details of all system aspects of AUS are provided but the major individual modules are only outlined. Sufficient information is given to enable other modules to be added to the system
Energy Technology Data Exchange (ETDEWEB)
Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), Service d' Etudes de Reacteurs et de Modelisation Avancee, 91 - Gif sur Yvette (France)
2003-07-01
Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)
Development of particle and heavy ion transport code system
International Nuclear Information System (INIS)
Niita, Koji
2004-01-01
Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)
Code system for fast reactor neutronics analysis
International Nuclear Information System (INIS)
Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.
1983-04-01
A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)
Electron transport code theoretical basis
International Nuclear Information System (INIS)
Dubi, A.; Horowitz, Y.S.
1978-04-01
This report mainly describes the physical and mathematical considerations involved in the treatment of the multiple collision processes. A brief description is given of the traditional methods used in electron transport via Monte Carlo, and a somewhat more detailed description, of the approach to be used in the presently developed code
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Tanaka, Shun-ichi
1991-09-01
A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)
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)
Monte Carlo method for neutron transport problems
International Nuclear Information System (INIS)
Asaoka, Takumi
1977-01-01
Some methods for decreasing variances in Monte Carlo neutron transport calculations are presented together with the results of sample calculations. A general purpose neutron transport Monte Carlo code ''MORSE'' was used for the purpose. The first method discussed in this report is the method of statistical estimation. As an example of this method, the application of the coarse-mesh rebalance acceleration method to the criticality calculation of a cylindrical fast reactor is presented. Effective multiplication factor and its standard deviation are presented as a function of the number of histories and comparisons are made between the coarse-mesh rebalance method and the standard method. Five-group neutron fluxes at core center are also compared with the result of S4 calculation. The second method is the method of correlated sampling. This method was applied to the perturbation calculation of control rod worths in a fast critical assembly (FCA-V-3) Two methods of sampling (similar flight paths and identical flight paths) are tested and compared with experimental results. For every cases the experimental value lies within the standard deviation of the Monte Carlo calculations. The third method is the importance sampling. In this report a biased selection of particle flight directions discussed. This method was applied to the flux calculation in a spherical fast neutron system surrounded by a 10.16 cm iron reflector. Result-direction biasing, path-length stretching, and no biasing are compared with S8 calculation. (Aoki, K.)
Neutron spectrum unfolding using computer code SAIPS
International Nuclear Information System (INIS)
Karim, S.
1999-01-01
The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results. (author)
CTCN: Colloid transport code -- nuclear
International Nuclear Information System (INIS)
Jain, R.
1993-01-01
This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential-algebraic equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential-algebraic systems
Fast-neutron, coded-aperture imager
Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.
2015-06-01
This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led
Fast-neutron, coded-aperture imager
International Nuclear Information System (INIS)
Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.
2015-01-01
This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led
Fast-neutron, coded-aperture imager
Energy Technology Data Exchange (ETDEWEB)
Woolf, Richard S., E-mail: richard.woolf@nrl.navy.mil; Phlips, Bernard F., E-mail: bernard.phlips@nrl.navy.mil; Hutcheson, Anthony L., E-mail: anthony.hutcheson@nrl.navy.mil; Wulf, Eric A., E-mail: eric.wulf@nrl.navy.mil
2015-06-01
This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led
A code for leakage neutron spectra through thick shields
International Nuclear Information System (INIS)
Nagarajan, P.S.; Sethulakshmi, P.; Raghavendran, C.P.
1975-01-01
An exponential transform Monte Carlo code has been developed for deep penetration of neutrons and the results of leakage neutron spectra of this code have been compared with those of a basic Monte Carlo code for small thickness. The development of the code and optimisation of certain transform parameters are discussed and results are presented for a few thick shields of concrete and water in the context of neutron monitoring in the environs of accelerator and reactor shields. (author)
SNAP - a three dimensional neutron diffusion code
International Nuclear Information System (INIS)
McCallien, C.W.J.
1993-02-01
This report describes a one- two- three-dimensional multi-group diffusion code, SNAP, which is primarily intended for neutron diffusion calculations but can also carry out gamma calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP can solve the multi-group neutron diffusion equations using finite difference methods. The one-dimensional slab, cylindrical and spherical geometries and the two-dimensional case are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. (Author)
Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Austregesilo, H.; Velkov, K. [GRS, Garching (Germany)] [and others
1997-07-01
The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.
Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes
International Nuclear Information System (INIS)
Langenbuch, S.; Austregesilo, H.; Velkov, K.
1997-01-01
The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes
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
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
Implementation of the kinetics in the transport code AZTRAN
International Nuclear Information System (INIS)
Duran G, J. A.; Del Valle G, E.; Gomez T, A. M.
2017-09-01
This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S n for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)
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
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
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
Applications of the Los Alamos High Energy Transport code
International Nuclear Information System (INIS)
Waters, L.; Gavron, A.; Prael, R.E.
1992-01-01
Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter
High energy particle transport code NMTC/JAM
International Nuclear Information System (INIS)
Niita, K.; Takada, H.; Meigo, S.; Ikeda, Y.
2001-01-01
We have developed a high energy particle transport code NMTC/JAM, which is an upgrade version of NMTC/JAERI97. The available energy range of NMTC/JAM is, in principle, extended to 200 GeV for nucleons and mesons including the high energy nuclear reaction code JAM for the intra-nuclear cascade part. We compare the calculations by NMTC/JAM code with the experimental data of thin and thick targets for proton induced reactions up to several 10 GeV. The results of NMTC/JAM code show excellent agreement with the experimental data. From these code validation, it is concluded that NMTC/JAM is reliable in neutronics optimization study of the high intense spallation neutron utilization facility. (author)
Some improved methods in neutron transport theory
Energy Technology Data Exchange (ETDEWEB)
Pop-Jordanov, J; Stefanovic, D; Kocic, A; Matausek, M; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)
1973-07-01
The methods described in this paper are: analytical approach to neutron spectra in case of energy dependent anisotropy of elastic scattering; Monte Carlo estimations of neutron absorption reaction rate during slowing down process; spherical harmonics treatment of space-angle-lethargy dependent slowing down transport equation; integral transport theory based on point-wise representation of variables.
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
Validation of SCALE code package on high performance neutron shields
International Nuclear Information System (INIS)
Bace, M.; Jecmenica, R.; Smuc, T.
1999-01-01
The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)
Neutron transport equation - indications on homogenization and neutron diffusion
International Nuclear Information System (INIS)
Argaud, J.P.
1992-06-01
In PWR nuclear reactor, the practical study of the neutrons in the core uses diffusion equation to describe the problem. On the other hand, the most correct method to describe these neutrons is to use the Boltzmann equation, or neutron transport equation. In this paper, we give some theoretical indications to obtain a diffusion equation from the general transport equation, with some simplifying hypothesis. The work is organised as follows: (a) the most general formulations of the transport equation are presented: integro-differential equation and integral equation; (b) the theoretical approximation of this Boltzmann equation by a diffusion equation is introduced, by the way of asymptotic developments; (c) practical homogenization methods of transport equation is then presented. In particular, the relationships with some general and useful methods in neutronic are shown, and some homogenization methods in energy and space are indicated. A lot of other points of view or complements are detailed in the text or the remarks
Progress in nuclear well logging modeling using deterministic transport codes
International Nuclear Information System (INIS)
Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.
2002-01-01
Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)
Results from the Coded Aperture Neutron Imaging System (CANIS)
International Nuclear Information System (INIS)
Brubaker, Erik; Steele, John T.; Brennan, James S.; Hilton, Nathan R.; Marleau, Peter
2010-01-01
Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging- a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.
Results from the coded aperture neutron imaging system
International Nuclear Information System (INIS)
Brubaker, Erik; Steele, John T.; Brennan, James S.; Marleau, Peter
2010-01-01
Because of their penetrating power, energetic neutrons and gamma rays (∼1 MeV) offer the best possibility of detecting highly shielded or distant special nuclear material (SNM). Of these, fast neutrons offer the greatest advantage due to their very low and well understood natural background. We are investigating a new approach to fast-neutron imaging - a coded aperture neutron imaging system (CANIS). Coded aperture neutron imaging should offer a highly efficient solution for improved detection speed, range, and sensitivity. We have demonstrated fast neutron and gamma ray imaging with several different configurations of coded masks patterns and detectors including an 'active' mask that is composed of neutron detectors. Here we describe our prototype detector and present some initial results from laboratory tests and demonstrations.
Particle and heavy ion transport code system; PHITS
International Nuclear Information System (INIS)
Niita, Koji
2004-01-01
Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear
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.)
Scintillator Based Coded-Aperture Imaging for Neutron Detection
International Nuclear Information System (INIS)
Hayes, Sean-C.; Gamage, Kelum-A-A.
2013-06-01
In this paper we are going to assess the variations of neutron images using a series of Monte Carlo simulations. We are going to study neutron images of the same neutron source with different source locations, using a scintillator based coded-aperture system. The Monte Carlo simulations have been conducted making use of the EJ-426 neutron scintillator detector. This type of detector has a low sensitivity to gamma rays and is therefore of particular use in a system with a source that emits a mixed radiation field. From the use of different source locations, several neutron images have been produced, compared both qualitatively and quantitatively for each case. This allows conclusions to be drawn on how suited the scintillator based coded-aperture neutron imaging system is to detecting various neutron source locations. This type of neutron imaging system can be easily used to identify and locate nuclear materials precisely. (authors)
Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.
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.
Recent advances in neutral particle transport methods and codes
International Nuclear Information System (INIS)
Azmy, Y.Y.
1996-01-01
An overview of ORNL's three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned
BERMUDA-1DG: a one-dimensional photon transport code
International Nuclear Information System (INIS)
Suzuki, Tomoo; Hasegawa, Akira; Nakashima, Hiroshi; Kaneko, Kunio.
1984-10-01
A one-dimensional photon transport code BERMUDA-1DG has been developed for spherical and infinite slab geometries. The purpose of development is to equip the function of gamma rays calculation for the BERMUDA code system, which was developed by 1983 only for neutron transport calculation as a preliminary version. A group constants library has been prepared for 30 nuclides, and it now consists of the 36-group total cross sections and secondary gamma ray yields by the 120-group neutron flux. For the Compton scattering, group-angle transfer matrices are accurately obtained by integrating the Klein-Nishina formula taking into account the energy and scattering angle correlation. The pair production cross sections are now calculated in the code from atomic number and midenergy of each group. To obtain angular flux distribution, the transport equation is solved in the same way as in case of neutron, using the direct integration method in a multigroup model. Both of an independent gamma ray source problem and a neutron-gamma source problem are possible to be solved. This report is written as a user's manual with a brief description of the calculational method. (author)
PHITS-a particle and heavy ion transport code system
International Nuclear Information System (INIS)
Niita, Koji; Sato, Tatsuhiko; Iwase, Hiroshi; Nose, Hiroyuki; Nakashima, Hiroshi; Sihver, Lembit
2006-01-01
The paper presents a summary of the recent development of the multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS. In particular, we discuss in detail the development of two new models, JAM and JQMD, for high energy particle interactions, incorporated in PHITS, and show comparisons between model calculations and experiments for the validations of these models. The paper presents three applications of the code including spallation neutron source, heavy ion therapy and space radiation. The results and examples shown indicate PHITS has great ability of carrying out the radiation transport analysis of almost all particles including heavy ions within a wide energy range
Comparison of DT neutron production codes MCUNED, ENEA-JSI source subroutine and DDT
Energy Technology Data Exchange (ETDEWEB)
Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Kodeli, Ivan [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Milocco, Alberto [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sauvan, Patrick [Departamento de Ingeniería Energética, E.T.S. Ingenieros Industriales, UNED, C/Juan del Rosal 12, 28040 Madrid (Spain); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)
2016-11-01
Highlights: • Results of three codes capable of simulating the accelerator based DT neutron generators were compared on a simple model where only a thin target made of mixture of titanium and tritium is present. Two typical deuteron beam energies, 100 keV and 250 keV, were used in the comparison. • Comparisons of the angular dependence of the total neutron flux and spectrum as well as the neutron spectrum of all the neutrons emitted from the target show general agreement of the results but also some noticeable differences. • A comparison of figures of merit of the calculations using different codes showed that the computational time necessary to achieve the same statistical uncertainty can vary for more than 30× when different codes for the simulation of the DT neutron generator are used. - Abstract: As the DT fusion reaction produces neutrons with energies significantly higher than in fission reactors, special fusion-relevant benchmark experiments are often performed using DT neutron generators. However, commonly used Monte Carlo particle transport codes such as MCNP or TRIPOLI cannot be directly used to analyze these experiments since they do not have the capabilities to model the production of DT neutrons. Three of the available approaches to model the DT neutron generator source are the MCUNED code, the ENEA-JSI DT source subroutine and the DDT code. The MCUNED code is an extension of the well-established and validated MCNPX Monte Carlo code. The ENEA-JSI source subroutine was originally prepared for the modelling of the FNG experiments using different versions of the MCNP code (−4, −5, −X) and was later extended to allow the modelling of both DT and DD neutron sources. The DDT code prepares the DT source definition file (SDEF card in MCNP) which can then be used in different versions of the MCNP code. In the paper the methods for the simulation of the DT neutron production used in the codes are briefly described and compared for the case of a
International Nuclear Information System (INIS)
McGill, B.L.; Roussin, R.W.; Trubey, D.K.; Maskewitz, B.F.
1980-01-01
The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results
Available computer codes and data for radiation transport analysis
International Nuclear Information System (INIS)
Trubey, D.K.; Maskewitz, B.F.; Roussin, R.W.
1975-01-01
The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results
BCG: a code for calculating pointwise neutron spectra and criticality in fast reactor cells
International Nuclear Information System (INIS)
Leite, S.B.; Caldeira, A.D.; Garcia, R.D.M.
1988-02-01
The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is presented. The code solves the unidimensional neutron transport equation together with interface current relations at each energy in an unionized grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstructed total microscopic cross sections of the atomic species in the cell. Results for a defined sample problem illustrate the high resolution and accuracy that can be obtained with the code. (author) [pt
BCG: a computer code for calculating neutron spectra and criticality in cells of fast reactors
International Nuclear Information System (INIS)
Leite, S.B.; Caldeira, A.D.; Garcia, R.D.M.
1988-01-01
The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is discussed. The code solves the unidimensional neutron transport equation together with interface current relations at each energy point in an unionized energy grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstruced total microscopic cross sections of the atomic species in the cell. Results for a simplified fuel cell illustrate the high resolution and accuracy that can be obtained with the code. (author) [pt
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.
Energy Technology Data Exchange (ETDEWEB)
Beckert, C.
2007-12-19
Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer
Modifications to the Monte Carlo neutronics code MONK
International Nuclear Information System (INIS)
Hutton, J.L.
1979-09-01
The Monte Carlo neutronics code MONK has been widely used for criticality calculations, and is one of the standard methods for assessing the safety of transport flasks and fuel storage facilities in the UK. Recently, attempts have been made to extend the range of applications of this calculational technique. In particular studies have been carried out using Monte Carlo to analyse reactor physics experiments. In these applications various shortcomings of the standard version MONK5 became apparent. The basic data library was found to be inadequate and additional estimates of parameters (eg power distribution) not normally included in criticality studies were required. These features which required improvement, primarily in the context of using the code for reactor physics calculations, are enumerated. To facilitate the use of the code as a reactor physics calculational tool a series of modifications have been carried out. The code has been modified so that the user can use group data tabulations of the cross sections instead of the present 'point' data values. The code can now interface with a number of reactor physics group data preparation schemes but in particular it can use WIMS-E interfaces as a source of group data. Details of the changes are outlined and a new version of MONK incorporating these modifications has been created. This version is called MONK5W. This paper provides a guide to the use of this version. The data input is described along with other details required to use this code on the Harwell IBM 3033. To aid the user, examples of calculations using the new facilities incorporated in MONK5W are given. (UK)
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
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)
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
Villarino, Eduardo; Hergenreder, Daniel
2000-01-01
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
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)
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
Energy Technology Data Exchange (ETDEWEB)
Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio
1996-05-01
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).
A neutron spectrum unfolding code based on iterative procedures
International Nuclear Information System (INIS)
Ortiz R, J. M.; Vega C, H. R.
2012-10-01
In this work, the version 3.0 of the neutron spectrum unfolding code called Neutron Spectrometry and Dosimetry from Universidad Autonoma de Zacatecas (NSDUAZ), is presented. This code was designed in a graphical interface under the LabVIEW programming environment and it is based on the iterative SPUNIT iterative algorithm, using as entrance data, only the rate counts obtained with 7 Bonner spheres based on a 6 Lil(Eu) neutron detector. The main features of the code are: it is intuitive and friendly to the user; it has a programming routine which automatically selects the initial guess spectrum by using a set of neutron spectra compiled by the International Atomic Energy Agency. Besides the neutron spectrum, this code calculates the total flux, the mean energy, H(10), h(10), 15 dosimetric quantities for radiation protection porpoises and 7 survey meter responses, in four energy grids, based on the International Atomic Energy Agency compilation. This code generates a full report in html format with all relevant information. In this work, the neutron spectrum of a 241 AmBe neutron source on air, located at 150 cm from detector, is unfolded. (Author)
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
Villarino, E.A.
1990-01-01
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author) [es
Radiation transport phenomena and modeling - part A: Codes
International Nuclear Information System (INIS)
Lorence, L.J.
1997-01-01
The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped
In-facility transport code review
International Nuclear Information System (INIS)
Spore, J.W.; Boyack, B.E.; Bohl, W.R.
1996-07-01
The following computer codes were reviewed by the In-Facility Transport Working Group for application to the in-facility transport of radioactive aerosols, flammable gases, and/or toxic gases: (1) CONTAIN, (2) FIRAC, (3) GASFLOW, (4) KBERT, and (5) MELCOR. Based on the review criteria as described in this report and the versions of each code available at the time of the review, MELCOR is the best code for the analysis of in-facility transport when multidimensional effects are not significant. When multi-dimensional effects are significant, GASFLOW should be used
Unfolding code for neutron spectrometry based on neural nets technology
International Nuclear Information System (INIS)
Ortiz R, J. M.; Vega C, H. R.
2012-10-01
The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Neural Networks have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This unfolding code called Neutron Spectrometry and Dosimetry by means of Artificial Neural Networks was designed in a graphical interface under LabVIEW programming environment. The core of the code is an embedded neural network architecture, previously optimized by the R obust Design of Artificial Neural Networks Methodology . The main features of the code are: is easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 Lil(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, only seven rate counts measurement with a Bonner spheres spectrometer are required for simultaneously unfold the 60 energy bins of the neutron spectrum and to calculate 15 dosimetric quantities, for radiation protection porpoises. This code generates a full report in html format with all relevant information. (Author)
Unfolding code for neutron spectrometry based on neural nets technology
Energy Technology Data Exchange (ETDEWEB)
Ortiz R, J. M.; Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Apdo. Postal 336, 98000 Zacatecas (Mexico)
2012-10-15
The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Neural Networks have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This unfolding code called Neutron Spectrometry and Dosimetry by means of Artificial Neural Networks was designed in a graphical interface under LabVIEW programming environment. The core of the code is an embedded neural network architecture, previously optimized by the {sup R}obust Design of Artificial Neural Networks Methodology{sup .} The main features of the code are: is easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a {sup 6}Lil(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, only seven rate counts measurement with a Bonner spheres spectrometer are required for simultaneously unfold the 60 energy bins of the neutron spectrum and to calculate 15 dosimetric quantities, for radiation protection porpoises. This code generates a full report in html format with all relevant information. (Author)
NASA space radiation transport code development consortium
International Nuclear Information System (INIS)
Townsend, L. W.
2005-01-01
Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)
Implementation of the quasi-static method for neutron transport
International Nuclear Information System (INIS)
Alcaro, Fabio; Dulla, Sandra; Ravetto, Piero; Le Tellier, Romain; Suteau, Christophe
2011-01-01
The study of the dynamic behavior of next generation nuclear reactors is a fundamental aspect for safety and reliability assessments. Despite the growing performances of modern computers, the full solution of the neutron Boltzmann equation in the time domain is still an impracticable task, thus several approximate dynamic models have been proposed for the simulation of nuclear reactor transients; the quasi-static method represents the standard tool currently adopted for the space-time solution of neutron transport problems. All the practical applications of this method that have been proposed contain a major limit, consisting in the use of isotropic quantities, such as scalar fluxes and isotropic external neutron sources, being the only data structures available in most deterministic transport codes. The loss of the angular information produces both inaccuracies in the solution of the kinetic model and the inconsistency of the quasi-static method itself. The present paper is devoted to the implementation of a consistent quasi-static method. The computational platform developed by CEA in Cadarache has been used for the creation of a kinetic package to be coupled with the existing SNATCH solver, a discrete-ordinate multi-dimensional neutron transport solver, employed for the solution of the steady-state Boltzmann equation. The work aims at highlighting the effects of the angular treatment of the neutron flux on the transient analysis, comparing the results with those produced by the previous implementations of the quasi-static method. (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)
International Nuclear Information System (INIS)
Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.
2002-01-01
A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA
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.)
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
Transportable type neutron level indicators
International Nuclear Information System (INIS)
Khatskevich, M.V.; Kalinin, O.V.; Moskovkin, V.N.; Molchanov, A.V.; Bobkov, A.D.; Rabotnov, Yu.A.
1979-01-01
Some peculiarities of designing level neutron converters (LNC) for portable indicators or level neutron relays are considered. The effect of the LNC geometry and other factors on measurement errors has been studied. Calibration results of the LNC with a neutron reflector and without it are presented. It is shown that the problem of level monitoring with the help of portable indicators can be solved practically for any volume, provided two LNC modifications with reflectors are available: the NPU-G modification with horizontal location of a counter for large volumes and the NPU-V with vertical location of a counter for lesser volumes. A possibility of perfecting LNC performances by shielding the counter with thermal neutron absorbers has been studied. The design of the NPU-V modification for the NIUP-2 level indicator is described. It is intended for tubes and cylinders 30-100 mm in diameter. Measurements carried out on different steel and aluminium vessels with a diameter ranging from 300 to 100 mm and a wall thickness of up to 16 mm with the help of the NPU-V and NPU-G modifications proved the efficiency of the LNC to control a variety of products (kerosine, gasoline, oils, acids, alkalis) [ru
Lectures on neutron transport theory
International Nuclear Information System (INIS)
Benoist, P.
1986-02-01
This note is divided in two parts. In the first one the basis of transport theory, that is, the principal forms of the transport equation and the resulting theorems, are presented. The second part is particularly devoted to the applications of integral transport theory to reactor lattice problems [fr
Parallel processing Monte Carlo radiation transport codes
International Nuclear Information System (INIS)
McKinney, G.W.
1994-01-01
Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine
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.)
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
International Nuclear Information System (INIS)
Odano, N.; Miura, T.; Yamaji, A.
1996-01-01
Measurement of activation reaction rates was carried out for fast neutrons penetrating through graphite and water from the core of JRR-4 research reactor of JAERI, with paying attention to the energy above 10 MeV. Analysis of the experiment was made using a vectorized continuous energy Monte Carlo code MVP to verify the code. The analysis shows good agreements between the measurement and calculation and the MVP code has been confirmed its validity for the fast neutron transport calculations above 10 MeV in fission neutron field. (author)
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.))
Optimization of a neutron detector design using adjoint transport simulation
International Nuclear Information System (INIS)
Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.
2012-01-01
A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)
Culbertson, C N; Wangerin, K; Ghandourah, E; Jevremovic, T
2005-08-01
The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.
An Auto sequence Code to Integrate a Neutron Unfolding Code with thePC-MCA Accuspec
International Nuclear Information System (INIS)
Darsono
2000-01-01
In a neutron spectrometry using proton recoil method, the neutronunfolding code is needed to unfold the measured proton spectrum to become theneutron spectrum. The process of the unfolding neutron in the existingneutron spectrometry which was successfully installed last year was doneseparately. This manuscript reports that the auto sequence code to integratethe neutron unfolding code UNFSPEC.EXE with the software facility of thePC-MCA Accuspec has been made and run successfully so that the new neutronspectrometry become compact. The auto sequence code was written based on therules in application program facility of PC-MCA Accuspec and then it wascompiled using AC-EXE. Result of the test of the auto sequence code showedthat for binning width 20, 30, and 40 giving a little different spectrumshape. The binning width around 30 gives a better spectrum in mean of givingsmall error compared to the others. (author)
Calculation of neutron spectra produced in neutron generator target: Code testing.
Gaganov, V V
2018-03-01
DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.
Using MCNP code for neutron and photon skyshine analysis
Energy Technology Data Exchange (ETDEWEB)
Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Netecha, M.E. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)
2000-03-01
The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector's dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field. (author)
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)
3D neutron transport modelization
International Nuclear Information System (INIS)
Warin, X.
1996-12-01
Some nodal methods to solve the transport equation in 3D are presented. Two nodal methods presented at an OCDE congress are described: a first one is a low degree one called RTN0; a second one is a high degree one called BDM1. The two methods can be made faster with a totally consistent DSA. Some results of parallelization show that: 98% of the time is spent in sweeps; transport sweeps are easily parallelized. (K.A.)
3D neutron transport modelization
Energy Technology Data Exchange (ETDEWEB)
Warin, X.
1996-12-01
Some nodal methods to solve the transport equation in 3D are presented. Two nodal methods presented at an OCDE congress are described: a first one is a low degree one called RTN0; a second one is a high degree one called BDM1. The two methods can be made faster with a totally consistent DSA. Some results of parallelization show that: 98% of the time is spent in sweeps; transport sweeps are easily parallelized. (K.A.). 10 refs.
Coupled geochemical and solute transport code development
International Nuclear Information System (INIS)
Morrey, J.R.; Hostetler, C.J.
1985-01-01
A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code
Intracoin - International Nuclide Transport Code Intercomparison Study
International Nuclear Information System (INIS)
1984-09-01
The purpose of the project is to obtain improved knowledge of the influence of various strategies for radionuclide transport modelling for the safety assessment of final repositories for nuclear waste. This is a report of the first phase of the project which was devoted to a comparison of the numerical accuracy of the computer codes used in the study. The codes can be divided into five groups, namely advection-dispersion models, models including matrix diffusion and chemical effects and finally combined models. The results are presented as comparisons of calculations since the objective of level 1 was code verification. (G.B.)
Computer codes in particle transport physics
International Nuclear Information System (INIS)
Pesic, M.
2004-01-01
Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option
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.
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
Transport calculation of medium-energy protons and neutrons by Monte Carlo method
International Nuclear Information System (INIS)
Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.
1978-09-01
A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)
Application of an integrated PC-based neutronics code system to criticality safety
International Nuclear Information System (INIS)
Briggs, J.B.; Nigg, D.W.
1991-01-01
An integrated system of neutronics and radiation transport software suitable for operation in an IBM PC-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past four years. Four modules within the system are particularly useful for criticality safety applications. Using the neutronics portion of the integrated code system, effective neutron multiplication values (k eff values) have been calculated for a variety of benchmark critical experiments for metal systems (Plutonium and Uranium), Aqueous Systems (Plutonium and Uranium) and LWR fuel rod arrays. A description of the codes and methods used in the analysis and the results of the benchmark critical experiments are presented in this paper. In general, excellent agreement was found between calculated and experimental results. (Author)
Application of direct discrete method (DDM) to multigroup neutron transport problems
International Nuclear Information System (INIS)
Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid
2003-01-01
The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)
FLUKA: A Multi-Particle Transport Code
Energy Technology Data Exchange (ETDEWEB)
Ferrari, A.; Sala, P.R.; /CERN /INFN, Milan; Fasso, A.; /SLAC; Ranft, J.; /Siegen U.
2005-12-14
This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner's guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.
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
MC2-2: a code to calculate fast neutron spectra and multigroup cross sections
International Nuclear Information System (INIS)
Henryson, H. II; Toppel, B.J.; Stenberg, C.G.
1976-06-01
MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers
Simulations of the neutronic REP behaviour using the codes DRAGON/DONJON
International Nuclear Information System (INIS)
Le Mer, J.
2007-01-01
Neutron flux calculation is necessary to understand how a nuclear reactor works. This flux is derived from the transport equation on the whole core. Because of its really complex structure and the angular dependence of the transport equation, it is impossible to compute the flux directly and several neutronic calculation codes must be used to solve the equation for different discretizations which require different modelisations. This chain of successive models, known as a calculation scheme, compute the neutron flux of a reactor from its geometry, its isotopic compositions and a cross-section library. Pressurised light Water Reactor (PWR) are the most common nuclear reactor used today. It is necessary for each neutronic code to be validated for this type of reactor. The goal of this work is to create a complete calculation scheme which can be applied to the evolution of the core of a pressurised light water nuclear reactor using the lattice code DRAGON and the reactor code DONJON. Each step of this scheme will be validated by comparisons with other codes or with experimental results. The unit cell calculation will be computed for a benchmark submitted by R. Mosteller. The assembly calculations will be used to compare the results given by DRAGON, APOLLO2 and MCNP for an assembly used by EDF for code testing. The core calculations will show that the codes DRAGON and DONJON can produce accurate macroscopic results for a real core. Those studies will be used to show the effects of many factors on the flux distribution including the cross section library, the number of energy groups, spatial discretization of the unit cell, the tracking model, the self-shielding of the resonant isotopes or the burnup steps. (author)
Neutron spallation source and the Dubna cascade code
Kumar, V; Goel, U; Barashenkov, V S
2003-01-01
Neutron multiplicity per incident proton, n/p, in collision of high energy proton beam with voluminous Pb and W targets has been estimated from the Dubna cascade code and compared with the available experimental data for the purpose of benchmarking of the code. Contributions of various atomic and nuclear processes for heat production and isotopic yield of secondary nuclei are also estimated to assess the heat and radioactivity conditions of the targets. Results obtained from the code show excellent agreement with the experimental data at beam energy, E < 1.2 GeV and differ maximum up to 25% at higher energy. (author)
Colloid transport code-nuclear user's manual
International Nuclear Information System (INIS)
Jain, R.
1992-01-01
This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems
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
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
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
Testing the new stochastic neutronic code ANET in simulating safety important parameters
International Nuclear Information System (INIS)
Xenofontos, T.; Delipei, G.-K.; Savva, P.; Varvayanni, M.; Maillard, J.; Silva, J.; Catsaros, N.
2017-01-01
Highlights: • ANET is a new neutronics stochastic code. • Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted. • Simulations of thermal, lower epithermal and fast neutron fluence rates were performed. • Axial fission rate distributions in standard and MOX fuel pins were computed. - Abstract: ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20 MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement
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)
Jetto a free boundary plasma transport code
International Nuclear Information System (INIS)
Cenacchi, G.; Taroni, A.
1988-01-01
JETTO is a one-and-a-half-dimensional transport code calculating the evolution of plasma parameters in a time dependent axisymmetric MHD equilibrium configuration. A splitting technique gives a consistent solution of coupled equilibrium and transport equations. The plasma boundary is free and defined either by its contact with a limiter (wall) or by a separatrix or by the toroidal magnetic flux. The Grad's approach to the equilibrium problem with adiabatic (or similar) constraints is adopted. This method consists of iterating by alternately solving the Grad-Schluter-Shafranov equation (PDE) and the ODE obtained by averaging the PDE over the magnetic surfaces. The bidimensional equation of the poloidal flux is solved by a finite difference scheme, whereas a Runge-Kutta method is chosen for the averaged equilibrium equation. The 1D transport equations (averaged over the magnetic surfaces) for the electron and ion densities and energies and for the rotational transform are written in terms of a coordinate (ρ) related to the toroidal flux. Impurity transport is also considered, under the hypothesis of coronal equilibrium. The transport equations are solved by an implicit scheme in time and by a finite difference scheme in space. The centering of the source terms and transport coefficients is performed using a Predictor-Corrector scheme. The basic version of the code is described here in detail; input and output parameters are also listed
TUTANK a two-dimensional neutron kinetics code
International Nuclear Information System (INIS)
Watts, M.G.; Halsall, M.J.; Fayers, F.J.
1975-04-01
TUTANK is a two-dimensional neutron kinetics code which treats two neutron energy groups and up to six groups of delayed neutron precursors. A 'theta differencing' method is used to integrate the time dependence of the equations. A position dependent exponential transformation on the time variable is available as an option, which in many circumstances can remove much of the time dependence, and thereby allow longer time steps to be taken. A further manipulation is made to separate the solutions of the neutron fluxes and the precursor concentrations. The spatial equations are based on standard diffusion theory, and their solution is obtained from alternating direction sweeps with a transverse buckling - the so-called ADI-B 2 method. Other features of the code include an elementary temperature feedback and heat removal treatment, automatic time step adjustment, a flexible method of specifying cross-section and heat transfer coefficient variations during a transient, and a restart facility which requires a minimal data specification. Full details of the code input are given. An example of the solution of a NEACRP benchmark for an LWR control rod withdrawal is given. (author)
High energy particle transport code NMTC/JAM
International Nuclear Information System (INIS)
Niita, Koji; Meigo, Shin-ichiro; Takada, Hiroshi; Ikeda, Yujiro
2001-03-01
We have developed a high energy particle transport code NMTC/JAM, which is an upgraded version of NMTC/JAERI97. The applicable energy range of NMTC/JAM is extended in principle up to 200 GeV for nucleons and mesons by introducing the high energy nuclear reaction code JAM for the intra-nuclear cascade part. For the evaporation and fission process, we have also implemented a new model, GEM, by which the light nucleus production from the excited residual nucleus can be described. According to the extension of the applicable energy, we have upgraded the nucleon-nucleus non-elastic, elastic and differential elastic cross section data by employing new systematics. In addition, the particle transport in a magnetic field has been implemented for the beam transport calculations. In this upgrade, some new tally functions are added and the format of input of data has been improved very much in a user friendly manner. Due to the implementation of these new calculation functions and utilities, consequently, NMTC/JAM enables us to carry out reliable neutronics study of a large scale target system with complex geometry more accurately and easily than before. This report serves as a user manual of the code. (author)
A study of the responses of neutron dose equivalent survey meters with computer codes
International Nuclear Information System (INIS)
Sartori, D.E.; Beer, G.P. de
1983-01-01
The ANISN and DOT discrete-ordinates radiation transport codes for one and two dimensions have been proved as effective and simple techniques to study the response of dose equivalent neutron detectors. Comparisons between results of an experimental calibration of the Harwell 95/0075 survey meter and calculated results rendered satisfactory agreement, considering the different techniques and sources of error involved. Possible improvements in the methods and designs and causes of error are discussed. (author)
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...
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.)
Integrated transport code system for a multicomponent plasma in a gas dynamic trap
International Nuclear Information System (INIS)
Anikeev, A.V.; Karpushov, A.N.; Noak, K.; Strogalova, S.L.
2000-01-01
This report is focused on the development of the theoretical and numerical models of multicomponent high-β plasma confinement and transport in the gas-dynamic trap (GDT). In order to simulate the plasma behavior in the GDT as well as that in the GDT-based neutron source the Integrated Transport Code System is developed from existing stand-alone codes calculating the target plasma, the fast ions and the neutral gas in the GDT. The code system considers the full dependence of the transport phenomena on space, time, energy and angle variables as well as the interactions between the particle fields [ru
New evaluated neutron cross section libraries for the GEANT4 code
International Nuclear Information System (INIS)
Mendoza, E.; Cano-Ott, D.; Guerrero, C.; Capote, R.
2012-04-01
The so-called High Precision neutron physics model implemented in the GEANT4 simulation package allows simulating the transport of neutrons with energies up to 20 MeV. It relies on the G4NDL cross section libraries, prepared by the GEANT4 collaboration from evaluated cross section files and distributed freely together with the code. Even though the performance of the G4NDL library has been improved over the time, users running complex simulations which involve the transport of neutrons do need more flexibility, in particular when assessing the uncertainties in the simulation results due to the neutron (and hence the nuclear) data library used. For this reason, a software tool has been developed for transforming any evaluated neutron cross section library in the ENDF-6 format into the G4NDL format. Furthermore, eight different releases of ENDF-B, JEFF, JENDL, CENDL and BROND national libraries have been translated into the G4NDL format and are distributed by the IAEA nuclear data service at www-nds.iaea.org/geant4. In this way, GEANT4 users have access to the complete list of standard evaluated neutron data libraries when performing Monte Carlo simulations with GEANT4. Consistency checks and a first validation of the libraries have been made following the methods described in this report. (author)
International Nuclear Information System (INIS)
2001-05-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
International Nuclear Information System (INIS)
2001-01-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
Energy Technology Data Exchange (ETDEWEB)
Duran G, J. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: redfield1290@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2017-09-15
This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S{sub n} for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)
Hydrogen recycle modeling in transport codes
International Nuclear Information System (INIS)
Howe, H.C.
1979-01-01
The hydrogen recycling models now used in Tokamak transport codes are reviewed and the method by which realistic recycling models are being added is discussed. Present models use arbitrary recycle coefficients and therefore do not model the actual recycling processes at the wall. A model for the hydrogen concentration in the wall serves two purposes: (1) it allows a better understanding of the density behavior in present gas puff, pellet, and neutral beam heating experiments; and (2) it allows one to extrapolate to long pulse devices such as EBT, ISX-C and reactors where the walls are observed or expected to saturate. Several wall models are presently being studied for inclusion in transport codes
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)
3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others
1997-07-01
This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.
Benchmark of neutron production cross sections with Monte Carlo codes
Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun
2018-02-01
Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first
Neutron transport simulation in high speed moving media using Geant4
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.
General features of the neutronics design code EQUICYCLE
International Nuclear Information System (INIS)
Jirlow, K.
1978-10-01
The neutronics code EQUICYCLE has been developed and improved over a long period of time. It is expecially adapted to survey type design calculations of large fast power reactors with particular emphasis on the nuclear parameters for a realistic equilibrium fuel cycle. Thus the code is used to evaluate the breeding performance, the power distributions and the uranium and plutonium mass balance for realistic refuelling schemes. In addition reactivity coefficients can be calculated and the influence of burnup could be assessed. The code is two-dimensional and treats the reactor core in R-Z geometry. The basic ideas of the calculating scheme are successive iterative improvement of cross-section sets and flux spectra and use of the mid-cycle flux for burning the fuel according to a specified refuelling scheme. Normally given peak burn-ups and maximum power densities are used as boundary conditions. The code is capable of handling the unconventional, so called heterogeneous cores. (author)
Alignment effects on a neutron imaging system using coded apertures
International Nuclear Information System (INIS)
Thfoin, Isabelle; Landoas, Olivier; Caillaud, Tony; Vincent, Maxime; Bourgade, Jean-Luc; Rosse, Bertrand; Disdier, Laurent; Sangster, Thomas C.; Glebov, Vladimir Yu.; Pien, Greg; Armstrong, William
2010-01-01
A high resolution neutron imaging system is being developed and tested on the OMEGA laser facility for inertial confinement fusion experiments. This diagnostic uses a coded imaging technique with a penumbral or an annular aperture. The sensitiveness of these techniques to misalignment was pointed out with both experiments and simulations. Results obtained during OMEGA shots are in good agreement with calculations performed with the Monte Carlo code GEANT4. Both techniques are sensitive to the relative position of the source in the field of view. The penumbral imaging technique then demonstrates to be less sensitive to misalignment compared to the ring. These results show the necessity to develop a neutron imaging diagnostic for megajoule class lasers taking into account our alignment capabilities on such facilities.
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
Present status of nucleon-meson transport code NMTC/JAERI
International Nuclear Information System (INIS)
Takada, H.; Meigo, S.
2001-01-01
The nucleon-meson transport code NMTC/JAM has been developed for the neutronics design study of the joint project for high-intensity proton accelerators with a power of mega-watts. The applicable energy range is extended by the inclusion of the jet AA microscopic transport model (JAM). The nucleon-nucleus cross sections are also updated for accurate transport calculation. The applicability of NMTC/JAM is studied through the analyses of thick target experiments such as neutron transmission through shield and activation reaction rate measurements. (orig.)
AUS98 - The 1998 version of the AUS modular neutronic code system
International Nuclear Information System (INIS)
Robinson, G.S.; Harrington, B.V.
1998-07-01
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 AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module
AUS98 - The 1998 version of the AUS modular neutronic code system
Energy Technology Data Exchange (ETDEWEB)
Robinson, G.S.; Harrington, B.V
1998-07-01
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 AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module refs., tabs.
Neutronics code VALE for two-dimensional triagonal (hexagonal) and three-dimensional geometries
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.
1981-08-01
This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180 0 rotational symmetry, and the rotational symmetry conditions for the 30 0 , 60 0 , and 120 0 triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source
Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes
Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.
2018-06-01
To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.
Depletion methodology in the 3-D whole core transport code DeCART
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun
2005-02-01
Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.
International Nuclear Information System (INIS)
Pashchenko, A.B.; Wienke, H.; Ganesan, S.
1996-01-01
Selected neutron reaction nuclear data evaluations for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into ACE format using the NJOY system by R.E. MacFarlane. This document summarizes the resulting continuous energy cross-section data library FENDL/MC version 1.1. The data are available cost free, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 1 tab
International Nuclear Information System (INIS)
Kawasaki, Hiromitu; Maki, Koichi; Seki, Yasushi.
1991-03-01
A code APPLE was produced in 1976 for calculating and plotting tritium breeding ratio and tritium production rate distributions. That code was improved as 'APPLE-2' in 1982, to calculate and plot not only tritium breeding ratio but also distributions of neutron and gamma-ray fluxes, their spectra, nuclear heating rates and other reaction rates, and dose rate distributions during operation and after shutdown in 1982. The code APPLE-2 can calculate and plot these nuclear properties derived from neutron and gamma-ray fluxes by ANISN (one dimensional transport code), DOT3.5 (two dimensional transport code) and MORSE (three dimensional Monte Carlo code). We revised the code APPLE-2 as 'APPLE-3' by adding many functions to the APPLE-2 code in accordance with users' requirements proposed in recent progress of fusion reaction nuclear design. With minor modification of APPLE-2, a number of inconsistencies have been found between the code manual and the input data in the code. In the present report, the new functions added to APPLE-2 and improved users' manual are explained. (author)
Discrete elements method of neutron transport
International Nuclear Information System (INIS)
Mathews, K.A.
1988-01-01
In this paper a new neutron transport method, called discrete elements (L N ) is derived and compared to discrete ordinates methods, theoretically and by numerical experimentation. The discrete elements method is based on discretizing the Boltzmann equation over a set of elements of angle. The discrete elements method is shown to be more cost-effective than discrete ordinates, in terms of accuracy versus execution time and storage, for the cases tested. In a two-dimensional test case, a vacuum duct in a shield, the L N method is more consistently convergent toward a Monte Carlo benchmark solution
Vectorization of three-dimensional neutron diffusion code CITATION
International Nuclear Information System (INIS)
Harada, Hiroo; Ishiguro, Misako
1985-01-01
Three-dimensional multi-group neutron diffusion code CITATION has been widely used for reactor criticality calculations. The code is expected to be run at a high speed by using recent vector supercomputers, when it is appropriately vectorized. In this paper, vectorization methods and their effects are described for the CITATION code. Especially, calculation algorithms suited for vectorization of the inner-outer iterative calculations which spend most of the computing time are discussed. The SLOR method, which is used in the original CITATION code, and the SOR method, which is adopted in the revised code, are vectorized by odd-even mesh ordering. The vectorized CITATION code is executed on the FACOM VP-100 and VP-200 computers, and is found to run over six times faster than the original code for a practical-scale problem. The initial value of the relaxation factor and the number of inner-iterations given as input data are also investigated since the computing time depends on these values. (author)
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....
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
DANDE-a linked code system for core neutronics/depletion analysis
International Nuclear Information System (INIS)
LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.
1986-01-01
This report describes DANDE-a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of the reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two actual problems
DANDE: a linked code system for core neutronics/depletion analysis
International Nuclear Information System (INIS)
LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.
1986-01-01
This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs
DANDE: a linked code system for core neutronics/depletion analysis
International Nuclear Information System (INIS)
LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.
1985-06-01
This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem
International Nuclear Information System (INIS)
Kawasaki, Hiromitsu; Seki, Yasushi.
1982-07-01
A computer code APPLE-2 which plots the spatial distribution of energy spectra of multi-group neutron and/or gamma ray fluxes, and reaction rates has been developed. This code is an improved version of the previously developed APPLE code and has the following features: (1) It plots energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT and MORSE. (2) It calculates and plots the spatial distribution of neutron and gamma ray fluxes and various types of reaction rates such as nuclear heating rates, operational dose rates, displacement damage rates. (3) Input data specification is greatly simplified by the use of standard, response libraries and by close coupling with radiation transport calculation codes. (4) Plotting outputs are given in camera ready form. (author)
Gaganov, V. V.
2017-12-01
The correctness of calculations performed with the SRIANG code for modeling the spectra of DT neutrons is estimated by comparing the obtained spectra to the results of calculations carried out with five different codes based on the Monte Carlo method.
The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL
International Nuclear Information System (INIS)
Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.
1983-03-01
This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)
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
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.
SNAP-3D: a three-dimensional neutron diffusion code
International Nuclear Information System (INIS)
McCallien, C.W.J.
1975-10-01
A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)
International Nuclear Information System (INIS)
Zazula, J.M.
1983-01-01
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
Energy Technology Data Exchange (ETDEWEB)
Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-10-01
A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)
International Nuclear Information System (INIS)
Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.
2014-01-01
Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement
CONIFERS: a neutronics code for reactors with channels
International Nuclear Information System (INIS)
Davis, R.S.
1977-04-01
CONIFERS is a neutronics code for nuclear reactors whose fuel is in channels that are separated from each other by several neutron mean-free-path lengths of moderator. It can treat accurately situations in which the usual homogenized-cell diffusion equation becomes inaccurate, but is more economical than other advanced methods such as response-matrix and source-sink formalisms. CONIFERS uses exact solutions of the neutron diffusion equation within each cell. It allows for the breakdown of this equation near a channel by means of data that almost any cell code can supply. It uses the results of these cell analyses in a reactor equations set that is as readily solvable as the familiar finite-difference equations set. CONIFERS can model almost any configuration of channels and other structures in two or three dimensions. It can use any number of energy groups and any reactivity scales, including scales based on control operations. It is also flexible from a programming point of view, and has convenient input and output provisions. (author)
The MCUCN simulation code for ultracold neutron physics
Zsigmond, G.
2018-02-01
Ultracold neutrons (UCN) have very low kinetic energies 0-300 neV, thereby can be stored in specific material or magnetic confinements for many hundreds of seconds. This makes them a very useful tool in probing fundamental symmetries of nature (for instance charge-parity violation by neutron electric dipole moment experiments) and contributing important parameters for the Big Bang nucleosynthesis (neutron lifetime measurements). Improved precision experiments are in construction at new and planned UCN sources around the world. MC simulations play an important role in the optimization of such systems with a large number of parameters, but also in the estimation of systematic effects, in benchmarking of analysis codes, or as part of the analysis. The MCUCN code written at PSI has been extensively used for the optimization of the UCN source optics and in the optimization and analysis of (test) experiments within the nEDM project based at PSI. In this paper we present the main features of MCUCN and interesting benchmark and application examples.
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
Application of Monte Carlo codes to neutron dosimetry
International Nuclear Information System (INIS)
Prevo, C.T.
1982-01-01
In neutron dosimetry, calculations enable one to predict the response of a proposed dosimeter before effort is expended to design and fabricate the neutron instrument or dosimeter. The nature of these calculations requires the use of computer programs that implement mathematical models representing the transport of radiation through attenuating media. Numerical, and in some cases analytical, solutions of these models can be obtained by one of several calculational techniques. All of these techniques are either approximate solutions to the well-known Boltzmann equation or are based on kernels obtained from solutions to the equation. The Boltzmann equation is a precise mathematical description of neutron behavior in terms of position, energy, direction, and time. The solution of the transport equation represents the average value of the particle flux density. Integral forms of the transport equation are generally regarded as the formal basis for the Monte Carlo method, the results of which can in principle be made to approach the exact solution. This paper focuses on the Monte Carlo technique
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)
Random mask optimization for fast neutron coded aperture imaging
Energy Technology Data Exchange (ETDEWEB)
McMillan, Kyle [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Univ. of California, Los Angeles, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Brubaker, Erik [Sandia National Lab. (SNL-CA), Livermore, CA (United States)
2015-05-01
In coded aperture imaging, one of the most important factors determining the quality of reconstructed images is the choice of mask/aperture pattern. In many applications, uniformly redundant arrays (URAs) are widely accepted as the optimal mask pattern. Under ideal conditions, thin and highly opaque masks, URA patterns are mathematically constructed to provide artifact-free reconstruction however, the number of URAs for a chosen number of mask elements is limited and when highly penetrating particles such as fast neutrons and high-energy gamma-rays are being imaged, the optimum is seldom achieved. In this case more robust mask patterns that provide better reconstructed image quality may exist. Through the use of heuristic optimization methods and maximum likelihood expectation maximization (MLEM) image reconstruction, we show that for both point and extended neutron sources a random mask pattern can be optimized to provide better image quality than that of a URA.
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)
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
PHISICS multi-group transport neutronic capabilities for RELAP5
Energy Technology Data Exchange (ETDEWEB)
Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)
2012-07-01
PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)
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)
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
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
Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko
2017-09-01
In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.
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
MADNIX a code to calculate prompt fission neutron spectra and average prompt neutron multiplicities
International Nuclear Information System (INIS)
Merchant, A.C.
1986-03-01
A code has been written and tested on the CDC Cyber-170 to calculate the prompt fission neutron spectrum, N(E), as a function of both the fissioning nucleus and its excitation energy. In this note a brief description of the underlying physical principles involved and a detailed explanation of the required input data (together with a sample output for the fission of 235 U induced by 14 MeV neutrons) are presented. Weisskopf's standard nuclear evaporation theory provides the basis for the calculation. Two important refinements are that the distribution of fission-fragment residual nuclear temperature and the cooling of the fragments as neutrons are emitted approximately taken into account, and also the energy dependence of the cross section for the inverse process of compound nucleus formation is included. This approach is then used to calculate the average number of prompt neutrons emitted per fission, v-bar p . At high excitation energies, where fission is still possible after neutron emission, the consequences of the competition between first, second and third chance fission on N(E) and v-bar p are calculated. Excellent agreement with all the examples given in the original work of Madland and Nix is obtained. (author) [pt
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
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.
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.)
Light-water-reactor coupled neutronic and thermal-hydraulic codes
International Nuclear Information System (INIS)
Diamond, D.J.
1982-01-01
An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented
Los Alamos radiation transport code system on desktop computing platforms
International Nuclear Information System (INIS)
Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.
1990-01-01
The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines
Multiple component codes based generalized LDPC codes for high-speed optical transport.
Djordjevic, Ivan B; Wang, Ting
2014-07-14
A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.
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)
EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes
Energy Technology Data Exchange (ETDEWEB)
Paolo Balestra; Carlo Parisi; Andrea Alfonsi
2016-02-01
The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.
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
International Nuclear Information System (INIS)
Lee, Y.K.; Hugot, F.X.
2011-01-01
The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)
Development of a neutronic analysis code using data from Monju
International Nuclear Information System (INIS)
Rooijen, W.F.G. van; Yamano, N.; Shimazu, Y.
2015-01-01
In recent years three major sets of modern evaluated nuclear data have become available: JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. The authors were involved with a research project to establish analysis method for a future commercial-scale LMFBR. This project focused on JENDL-4.0 and conventional Japanese codes. As a cross check, we decided to also apply the fast reactor code ERANOS. This necessitated to produce nuclear data (cross sections, etc) for the ERANOS code system, discussed in this paper. We developed a nuclear data processing system to produce cross sections, probability tables, delayed neutron data, and covariance data from the evaluated nuclear data files for ERANOS. A benchmark calculation on the MZA/MZB benchmark showed very satisfying results. Subsequently, we analyzed the prototype LMFBR Monju with ERANOS and our own sets of nuclear data. The results are very satisfactory. The results from ERANOS indicate that the target accuracies for nuclear data have not been met, although the three sets of evaluated nuclear data all performed very well in our analysis. In the future, the covariance on nuclear data should be reduced to meet the target accuracies on criticality and feedback coefficients. (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
Wangerin, K; Culbertson, C N; Jevremovic, T
2005-08-01
The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam.
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
RADTRAN: a computer code to analyze transportation of radioactive material
International Nuclear Information System (INIS)
Taylor, J.M.; Daniel, S.L.
1977-04-01
A computer code is presented which predicts the environmental impact of any specific scheme of radioactive material transportation. Results are presented in terms of annual latent cancer fatalities and annual early fatility probability resulting from exposure, during normal transportation or transport accidents. The code is developed in a generalized format to permit wide application including normal transportation analysis; consideration of alternatives; and detailed consideration of specific sectors of industry
International Nuclear Information System (INIS)
Khattab, K.; Omar, H.; Ghazi, N.
2009-01-01
A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)
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.
Modelling of a general purpose irradiation chamber using a Monte Carlo particle transport code
International Nuclear Information System (INIS)
Dhiyauddin Ahmad Fauzi; Sheik, F.O.A.; Nurul Fadzlin Hasbullah
2013-01-01
Full-text: The aim of this research is to stimulate the effectiveness use of a general purpose irradiation chamber to contain pure neutron particles obtained from a research reactor. The secondary neutron and gamma particles dose discharge from the chamber layers will be used as a platform to estimate the safe dimension of the chamber. The chamber, made up of layers of lead (Pb), shielding, polyethylene (PE), moderator and commercial grade aluminium (Al) cladding is proposed for the use of interacting samples with pure neutron particles in a nuclear reactor environment. The estimation was accomplished through simulation based on general Monte Carlo N-Particle transport code using Los Alamos MCNPX software. Simulations were performed on the model of the chamber subjected to high neutron flux radiation and its gamma radiation product. The model of neutron particle used is based on the neutron source found in PUSPATI TRIGA MARK II research reactor which holds a maximum flux value of 1 x 10 12 neutron/ cm 2 s. The expected outcomes of this research are zero gamma dose in the core of the chamber and neutron dose rate of less than 10 μSv/ day discharge from the chamber system. (author)
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.
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
A novel neutron energy spectrum unfolding code using particle swarm optimization
International Nuclear Information System (INIS)
Shahabinejad, H.; Sohrabpour, M.
2017-01-01
A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to match well with those of the TGASU code for both under determined and over-determined problems. In addition the SDPSO has been shown to be nearly two times faster than the TGASU code. - Highlights: • Introducing a novel method for neutron spectrum unfolding. • Implementation of a particle swarm optimization code for neutron unfolding. • Comparing results of the PSO code with those of recently published TGASU code. • Match results of the PSO code with those of TGASU code. • Greater convergence rate of implemented PSO code than TGASU code.
International Nuclear Information System (INIS)
Alonso V, G.; Viais J, J.
1990-10-01
There is developed a method to generate the library of neutron cross sections for the Thermos code by means of the database ENDF-B/IV and the NJOY code. The obtained results are compared with the version previous of the library of neutron cross sections which was processed using the version ENDF-B/III. (Author)
Investigation of coupling scheme for neutronic and thermal-hydraulic codes
International Nuclear Information System (INIS)
Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.
1988-01-01
Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described
International Nuclear Information System (INIS)
Mika, J.
1975-09-01
Originally the work was oriented towards two main topics: a) difference and integral methods in neutron transport theory. Two computers were used for numerical calculations GIER and CYBER-72. During the first year the main effort was shifted towards basic theoretical investigations. At the first step the ANIS code was adopted and later modified to check various finite difference approaches against each other. Then the general finite element method and the singular perturbation method were developed. The analysis of singularities of the one-dimensional neutron transport equation in spherical geometry has been done and presented. Later the same analysis for the case of cylindrical symmetry has been carried out. The second and the third year programme included the following topics: 1) finite difference methods in stationary neutron transport theory; 2)mathematical fundamentals of approximate methods for solving the transport equation; 3) singular perturbation method for the time-dependent transport equation; 4) investigation of various iterative procedures in reactor calculations. This investigation will help to better understanding of the mathematical basis for existing and developed numerical methods resulting in more effective algorithms for reactor computer codes
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
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)
Development Of The Computer Code For Comparative Neutron Activation Analysis
International Nuclear Information System (INIS)
Purwadi, Mohammad Dhandhang
2001-01-01
The qualitative and quantitative chemical analysis with Neutron Activation Analysis (NAA) is an importance utilization of a nuclear research reactor, and this should be accelerated and promoted in application and its development to raise the utilization of the reactor. The application of Comparative NAA technique in GA Siwabessy Multi Purpose Reactor (RSG-GAS) needs special (not commercially available yet) soft wares for analyzing the spectrum of multiple elements in the analysis at once. The application carried out using a single spectrum software analyzer, and comparing each result manually. This method really degrades the quality of the analysis significantly. To solve the problem, a computer code was designed and developed for comparative NAA. Spectrum analysis in the code is carried out using a non-linear fitting method. Before the spectrum analyzed, it was passed to the numerical filter which improves the signal to noise ratio to do the deconvolution operation. The software was developed using the G language and named as PASAN-K The testing result of the developed software was benchmark with the IAEA spectrum and well operated with less than 10 % deviation
International Nuclear Information System (INIS)
Cramer, S.N.
1985-09-01
An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given
Recent developments in the Los Alamos radiation transport code system
International Nuclear Information System (INIS)
Forster, R.A.; Parsons, K.
1997-01-01
A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results
MAX: an expert system for running the modular transport code APOLLO II
International Nuclear Information System (INIS)
Loussouarn, O.; Ferraris, C.; Boivineau, A.
1990-01-01
MAX is an expert system built to help users of the APOLLO II code to prepare the input data deck to run a job. APOLLO II is a modular transport-theory code for calculating the neutron flux in various geometries. The associated GIBIANE command language allows the user to specify the physical structure and the computational method to be used in the calculation. The purpose of MAX is to bring into play expertise in both neutronic and computing aspects of the code, as well as various computational schemes, in order to generate automatically a batch data set corresponding to the APOLLO II calculation desired by the user. MAX is implemented on the SUN 3/60 workstation with the S1 tool and graphic interface external functions
A computer code for calculating neutron cross-sections from resonance parameter data
International Nuclear Information System (INIS)
Mill, A.J.
1979-08-01
A computer code, XSEC, has been written which calculates neutron cross-sections from resonance data. Although the program was originally written in order to identify neutron 'windows' in enriched nuclides, it may be used to evaluate the total neutron cross-section of any medium mass nuclide at intermediate energies. XSEC has proved very useful in identifying suitable nuclides for use as neutron filters at intermediate energies. (author)
Properties of SiC semiconductor detector of fast neutrons investigated using MCNPX code
International Nuclear Information System (INIS)
Sedlakova, K.; Sagatova, A.; Necas, V.; Zatko, B.
2013-01-01
The potential of silicon carbide (SiC) for use in semiconductor nuclear radiation detectors has been long recognized. The wide bandgap of SiC (3.25 eV for 4H-SiC polytype) compared to that for more conventionally used semiconductors, such as silicon (1.12 eV) and germanium (0.67 eV), makes SiC an attractive semiconductor for use in high dose rate and high ionization nuclear environments. The present work focused on the simulation of particle transport in SiC detectors of fast neutrons using statistical analysis of Monte Carlo radiation transport code MCNPX. Its possibilities in detector design and optimization are presented.(authors)
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.
Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem
International Nuclear Information System (INIS)
William Charlton
2007-01-01
Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions
PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation
International Nuclear Information System (INIS)
Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.
1995-01-01
1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the
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)
Validation of the 3D finite element transport theory code EVENT for shielding applications
International Nuclear Information System (INIS)
Warner, Paul; Oliveira, R.E. de
2000-01-01
This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P N ) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the code are compared with predictions of the Monte Carlo code MCBEND and also with the results from benchmark shielding experiments. In particular, results are presented for the ASPIS experimental configuration for both neutron and gamma ray calculations using the BUGLE 96 nuclear data library. (author)
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
International Nuclear Information System (INIS)
Ganapol, B.D.
1987-01-01
For almost 20 yr, the main thrust of the author's research has been the generation of as many benchmark solutions to the time-dependent monoenergetic neutron transport equation as possible. The major motivation behind this effort has been to provide code developers with highly accurate numerical solutions to serve as standards in the assessment of numerical transport algorithms. In addition, these solutions provide excellent educational tools since the important physical features of neutron transport are still present even though the problems solved are idealized. A secondary motivation, though of equal importance, is the intellectual stimulation and understanding provided by the combination of the analytical, numerical, and computational techniques required to obtain these solutions. Therefore, to further the benchmark development, the added complication of time-dependent cross sections in the one-group transport equation is considered here
Transport coefficients in superfluid neutron stars
Energy Technology Data Exchange (ETDEWEB)
Tolos, Laura [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Frankfurt Institute for Advances Studies. Johann Wolfgang Goethe University, Ruth-Moufang-Str. 1, 60438 Frankfurt am Main (Germany); Manuel, Cristina [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Sarkar, Sreemoyee [Tata Institute of Fundamental Research, Homi Bhaba Road, Mumbai-400005 (India); Tarrus, Jaume [Physik Department, Technische Universität München, D-85748 Garching (Germany)
2016-01-22
We study the shear and bulk viscosity coefficients as well as the thermal conductivity as arising from the collisions among phonons in superfluid neutron stars. We use effective field theory techniques to extract the allowed phonon collisional processes, written as a function of the equation of state and the gap of the system. The shear viscosity due to phonon scattering is compared to calculations of that coming from electron collisions. We also comment on the possible consequences for r-mode damping in superfluid neutron stars. Moreover, we find that phonon collisions give the leading contribution to the bulk viscosities in the core of the neutron stars. We finally obtain a temperature-independent thermal conductivity from phonon collisions and compare it with the electron-muon thermal conductivity in superfluid neutron stars.
Neutron transport from targets to moderators
International Nuclear Information System (INIS)
Taylor, A.D.
1981-06-01
By appropriately choosing parameters such as temperature, decoupler, thickness and effective size it is possible to tailor the moderators of a pulsed spallation neutron source in such a way that the different characteristics regarding time structure and spectral distribution as requested for the different instruments can be met very closely. This enables a unique flexibility in the design of neutron spectrometers to be used at such a source. (author)
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
AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B
Energy Technology Data Exchange (ETDEWEB)
Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.
1976-03-01
AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)
Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program
International Nuclear Information System (INIS)
Moskowitz, B.S.
2000-01-01
This paper describes the modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program. This effort represents a complete 'white sheet of paper' rewrite of the code. In this paper, the motivation driving this project, the design objectives for the new version of the program, and the design choices and their consequences will be discussed. The design itself will also be described, including the important subsystems as well as the key classes within those subsystems
A Monte Carlo Green's function method for three-dimensional neutron transport
International Nuclear Information System (INIS)
Gamino, R.G.; Brown, F.B.; Mendelson, M.R.
1992-01-01
This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution
International Nuclear Information System (INIS)
Koch, K.R.
1985-01-01
A new analysis method specially suited for the inherent difficulties of fusion neutronics was developed to provide detailed studies of the fusion neutron transport physics. These studies should provide a better understanding of the limitations and accuracies of typical fusion neutronics calculations. The new analysis method is based on the direct integration of the integral form of the neutron transport equation and employs a continuous energy formulation with the exact treatment of the energy angle kinematics of the scattering process. In addition, the overall solution is analyzed in terms of uncollided, once-collided, and multi-collided solution components based on a multiple collision treatment. Furthermore, the numerical evaluations of integrals use quadrature schemes that are based on the actual dependencies exhibited in the integrands. The new DITRAN computer code was developed on the Cyber 205 vector supercomputer to implement this direct integration multiple-collision fusion neutronics analysis. Three representative fusion reactor models were devised and the solutions to these problems were studied to provide suitable choices for the numerical quadrature orders as well as the discretized solution grid and to understand the limitations of the new analysis method. As further verification and as a first step in assessing the accuracy of existing fusion-neutronics calculations, solutions obtained using the new analysis method were compared to typical multigroup discrete ordinates calculations
NSPEC - A neutron spectrum code for beam-heated fusion plasmas
International Nuclear Information System (INIS)
Scheffel, J.
1983-06-01
A 3-dimensional computer code is described, which computes neutron spectra due to beam heating of fusion plasmas. Three types of interactions are considered; thermonuclear of plasma-plasma, beam-plasma and beam-beam interactions. Beam deposition is modelled by the NFREYA code. The applied steady state beam distribution as a function of pitch angle and velocity contains the effects of energy diffusion, friction, angular scattering, charge exchange, electric field and source pitch angle distribution. The neutron spectra, generated by Monte-Carlo methods, are computed with respect to given lines of sight. This enables the code to be used for neutron diagnostics. (author)
International Nuclear Information System (INIS)
Dominguez, Dany S.; Oliveira, Francisco B.S.; Barros, Ricardo C.
2003-01-01
We present in this paper a multiplatform computational code to calculate elements of Gauss-Legendre angular quadrature sets of arbitrary order used in slab-geometry discrete ordinates (S N ) formulation of neutron transport equation. In the code, the values can be computed with arbitrary arithmetic precision based on the approach of exact computing floating-point numbers. Calculation routines have been developed in the common language ANSI C using standard compiler gcc and the libraries of the open code GMP (GNU Multi precision Library). The code has a graphical interface in order to facilitate user interaction and numerical results analysis. The code architecture allows it to run on different platforms such as Unix, Linux and Windows. Numerical results and performance measures are also given. (author)
Neutron transport from targets to moderators
International Nuclear Information System (INIS)
Taylor, A.D.
1980-01-01
The title of this meeting is 'Targets for Neutron Beam Spallation Sources', but so far all the emphasis in the talks has been on how to produce the fast neutron flux. I would like to stress that that is just the beginning of the story. What we are required to produce are beams of thermal and epithermal neutrons with time and spectral characteristics tailored to the instrumental requirements. The real source of our neutrons is not uranium arrays or thorium cylinders but a small volume of hydrogenous material, some 10 x 10 x 5 cm 3 . This is really what the whole thing is about - the target produces a copious field of fast neutrons, but if we fail to moderate them with the right energy and time characteristics, we will not match to what is happening downstream. In this talk, I am going to deal specifically with what we have done for SNS to optimise the target-moderator-reflector and decoupler system in this respect. (orig.)
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)
Design of tallying function for general purpose Monte Carlo particle transport code JMCT
International Nuclear Information System (INIS)
Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin
2013-01-01
A new postponed accumulation algorithm was proposed. Based on JCOGIN (J combinatorial geometry Monte Carlo transport infrastructure) framework and the postponed accumulation algorithm, the tallying function of the general purpose Monte Carlo neutron-photon transport code JMCT was improved markedly. JMCT gets a higher tallying efficiency than MCNP 4C by 28% for simple geometry model, and JMCT is faster than MCNP 4C by two orders of magnitude for complicated repeated structure model. The available ability of tallying function for JMCT makes firm foundation for reactor analysis and multi-step burnup calculation. (authors)
SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo
International Nuclear Information System (INIS)
2003-01-01
1 - Nature of physical problem solved: The SAM-CE system comprises two Monte Carlo codes, SAM-F and SAM-A. SAM-F supersedes the forward Monte Carlo code, SAM-C. SAM-A is an adjoint Monte Carlo code designed to calculate the response due to fields of primary and secondary gamma radiation. The SAM-CE system is a FORTRAN Monte Carlo computer code designed to solve the time-dependent neutron and gamma-ray transport equations in complex three-dimensional geometries. SAM-CE is applicable for forward neutron calculations and for forward as well as adjoint primary gamma-ray calculations. In addition, SAM-CE is applicable for the gamma-ray stage of the coupled neutron-secondary gamma ray problem, which may be solved in either the forward or the adjoint mode. Time-dependent fluxes, and flux functionals such as dose, heating, count rates, etc., are calculated as functions of energy, time and position. Multiple scoring regions are permitted and these may be either finite volume regions or point detectors or both. Other scores of interest, e.g., collision and absorption densities, etc., are also made. 2 - Method of solution: A special feature of SAM-CE is its use of the 'combinatorial geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All nuclear interaction cross section data (derived from the ENDF for neutrons and from the UNC-format library for gamma-rays) are tabulated in point energy meshes. The energy meshes for neutrons are internally derived, based on built-in convergence criteria and user- supplied tolerances. Tabulated neutron data for each distinct nuclide are in unique and appropriate energy meshes. Both resolved and unresolved resonance parameters from ENDF data files are treated automatically, and extremely precise and detailed descriptions of cross section behaviour is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux
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.
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)
Motivation for Using Generalized Geometry in the Time Dependent Transport Code TDKENO
Energy Technology Data Exchange (ETDEWEB)
Dustin Popp; Zander Mausolff; Sedat Goluoglu
2016-04-01
We are proposing to use the code, TDKENO, to model TREAT. TDKENO solves the time dependent, three dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the neutron flux is factored into two components – a rapidly varying amplitude equation and a slowly varying shape equation and each is solved separately on different time scales. The shape equation is solved using the 3D Monte Carlo transport code KENO, from Oak Ridge National Laboratory’s SCALE code package. Using the Monte Carlo method to solve the shape equation is still computationally intensive, but the operation is only performed when needed. The amplitude equation is solved deterministically and frequently, so the solution gives an accurate time-dependent solution without having to repeatedly We have modified TDKENO to incorporate KENO-VI so that we may accurately represent the geometries within TREAT. This paper explains the motivation behind using generalized geometry, and provides the results of our modifications. TDKENO uses the Improved Quasi-Static method to accomplish this. In this method, the neutron flux is factored into two components. One component is a purely time-dependent and rapidly varying amplitude function, which is solved deterministically and very frequently (small time steps). The other is a slowly varying flux shape function that weakly depends on time and is only solved when needed (significantly larger time steps).
A novel approach to correct the coded aperture misalignment for fast neutron imaging
Energy Technology Data Exchange (ETDEWEB)
Zhang, F. N.; Hu, H. S., E-mail: huasi-hu@mail.xjtu.edu.cn; Wang, D. M.; Jia, J. [School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Zhang, T. K. [Laser Fusion Research Center, CAEP, Mianyang, 621900 Sichuan (China); Jia, Q. G. [Institute of Applied Physics and Computational Mathematics, Beijing 100094 (China)
2015-12-15
Aperture alignment is crucial for the diagnosis of neutron imaging because it has significant impact on the coding imaging and the understanding of the neutron source. In our previous studies on the neutron imaging system with coded aperture for large field of view, “residual watermark,” certain extra information that overlies reconstructed image and has nothing to do with the source is discovered if the peak normalization is employed in genetic algorithms (GA) to reconstruct the source image. Some studies on basic properties of residual watermark indicate that the residual watermark can characterize coded aperture and can thus be used to determine the location of coded aperture relative to the system axis. In this paper, we have further analyzed the essential conditions for the existence of residual watermark and the requirements of the reconstruction algorithm for the emergence of residual watermark. A gamma coded imaging experiment has been performed to verify the existence of residual watermark. Based on the residual watermark, a correction method for the aperture misalignment has been studied. A multiple linear regression model of the position of coded aperture axis, the position of residual watermark center, and the gray barycenter of neutron source with twenty training samples has been set up. Using the regression model and verification samples, we have found the position of the coded aperture axis relative to the system axis with an accuracy of approximately 20 μm. Conclusively, a novel approach has been established to correct the coded aperture misalignment for fast neutron coded imaging.
Time-dependent anisotropic distributed source capability in transient 3-d transport code tort-TD
International Nuclear Information System (INIS)
Seubert, A.; Pautz, A.; Becker, M.; Dagan, R.
2009-01-01
The transient 3-D discrete ordinates transport code TORT-TD has been extended to account for time-dependent anisotropic distributed external sources. The extension aims at the simulation of the pulsed neutron source in the YALINA-Thermal subcritical assembly. Since feedback effects are not relevant in this zero-power configuration, this offers a unique opportunity to validate the time-dependent neutron kinetics of TORT-TD with experimental data. The extensions made in TORT-TD to incorporate a time-dependent anisotropic external source are described. The steady state of the YALINA-Thermal assembly and its response to an artificial square-wave source pulse sequence have been analysed with TORT-TD using pin-wise homogenised cross sections in 18 prompt energy groups with P 1 scattering order and 8 delayed neutron groups. The results demonstrate the applicability of TORT-TD to subcritical problems with a time-dependent external source. (authors)
Considerations in the design of an improved transportable neutron spectrometer
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.
Optix: A Monte Carlo scintillation light transport code
Energy Technology Data Exchange (ETDEWEB)
Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)
2014-02-11
The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.
International Nuclear Information System (INIS)
Fischer, G.A.
2010-01-01
The PCA Benchmark is analyzed using RAPTOR-M3G, a parallel SN radiation transport code. A variety of mesh structures, angular quadrature sets, cross section treatments, and reactor dosimetry cross sections are presented. The results show that RAPTOR-M3G is generally suitable for PWR neutron dosimetry applications. (authors)
A neutron spectrum unfolding computer code based on artificial neural networks
International Nuclear Information System (INIS)
Ortiz-Rodríguez, J.M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J.M.; Vega-Carrillo, H.R.
2014-01-01
The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding
Parallel processing of Monte Carlo code MCNP for particle transport problem
Energy Technology Data Exchange (ETDEWEB)
Higuchi, Kenji; Kawasaki, Takuji
1996-06-01
It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)
Interfacial and Wall Transport Models for SPACE-CAP Code
International Nuclear Information System (INIS)
Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun
2009-01-01
The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code
Interfacial and Wall Transport Models for SPACE-CAP Code
Energy Technology Data Exchange (ETDEWEB)
Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)
2009-10-15
The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.
Application of the three-dimensional Oak Ridge transport code
International Nuclear Information System (INIS)
Rhoades, W.A.; Childs, R.L.; Emmett, M.B.; Cramer, S.N.
1984-01-01
TORT, a 3-d extension of the DOT discrete ordinates transport code is now in production use for studies of radiation penetration into large concrete and masonry structures. This paper discusses certain features of the new code and shows representative results, including comparisons with Monte Carlo calculations
MORSE-CGT Monte Carlo radiation transport code with the capability of the torus geometric treatment
International Nuclear Information System (INIS)
Deng Li
1990-01-01
The combinatorial geometry package CGT with the capability of the torus geometric treatment is introduced. It is get by developing the combinatorial geometry package CG. The CGT package can be transplanted to those codes which the CG package is being used and makes them also with the capability. The MORSE-CGT code can be used to solve the neutron, gamma-ray or coupled neutron-gamma-ray transport problems and time dependence for both shielding and criticality problems in torus system or system which is produced by arbitrary finite combining torus with torus or other bodies in CG package and it can also be used to design the blanket and compute shielding for TOKAMAK Fusion-Fission Hybrid Reactor
International Nuclear Information System (INIS)
Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.
2007-01-01
In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required
Low-discrepancy point sets in transport codes
Energy Technology Data Exchange (ETDEWEB)
Warnock, T.T.
1985-01-01
A drawback to Monte Carlo methods of computation is its rate of convergence. There are methods of sampling that have a better error estimate than those using random numbers. This paper gives the result of some preliminary experiments with these sampling methods on two neutron transport problems.
The Lattice Boltzmann Method applied to neutron transport
International Nuclear Information System (INIS)
Erasmus, B.; Van Heerden, F. A.
2013-01-01
In this paper the applicability of the Lattice Boltzmann Method to neutron transport is investigated. One of the main features of the Lattice Boltzmann method is the simultaneous discretization of the phase space of the problem, whereby particles are restricted to move on a lattice. An iterative solution of the operator form of the neutron transport equation is presented here, with the first collision source as the starting point of the iteration scheme. A full description of the discretization scheme is given, along with the quadrature set used for the angular discretization. An angular refinement scheme is introduced to increase the angular coverage of the problem phase space and to mitigate lattice ray effects. The method is applied to a model problem to investigate its applicability to neutron transport and the results are compared to a reference solution calculated, using MCNP. (authors)
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)
International Nuclear Information System (INIS)
Bernnat, W.; Keinert, J.; Mattes, M.
2004-01-01
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H 2 O, liquid He, liquid D 2 O, liquid and solid H 2 and D 2 , solid CH 4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S N -transport codes and the Monte Carlo Code MCNP. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)
2004-03-01
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)
Benchmarking NNWSI flow and transport codes: COVE 1 results
International Nuclear Information System (INIS)
Hayden, N.K.
1985-06-01
The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs
A neutron spectrum unfolding computer code based on artificial neural networks
Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.
2014-02-01
The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in
ICPP - a collision probability module for the AUS neutronics code system
International Nuclear Information System (INIS)
Robinson, G.S.
1985-10-01
The isotropic collision probability program (ICPP) is a module of the AUS neutronics code system which calculates first flight collision probabilities for neutrons in one-dimensional geometries and in clusters of rods. Neutron sources, including scattering, are assumed to be isotropic and to be spatially flat within each mesh interval. The module solves the multigroup collision probability equations for eigenvalue or fixed source problems
International Nuclear Information System (INIS)
Ouisloumen, M.
1989-03-01
This research thesis reports an investigation of neutron diffusion by the fuel of a nuclear reactor while taking thermal agitation of diffusing nuclei into account. The author adopted the free gas model to describe the atom vibrations in the crystal network. He reports the elaboration of a diffusion law which gives the neutron energy spectrum after collision. Some particular theoretical cases are then studied in order to explain the observed physical phenomena, and to confirm the obtained results. The Monte Carlo method is also used to validate these results. The author presents the various physical criteria which must be met by an energetic grid in order to minimize errors in reaction rate calculations
Development and preliminary verification of 2-D transport module of radiation shielding code ARES
International Nuclear Information System (INIS)
Zhang Penghe; Chen Yixue; Zhang Bin; Zang Qiyong; Yuan Longjun; Chen Mengteng
2013-01-01
The 2-D transport module of radiation shielding code ARES is two-dimensional neutron and radiation shielding code. The theory model was based on the first-order steady state neutron transport equation, adopting the discrete ordinates method to disperse direction variables. Then a set of differential equations can be obtained and solved with the source iteration method. The 2-D transport module of ARES was capable of calculating k eff and fixed source problem with isotropic or anisotropic scattering in x-y geometry. The theoretical model was briefly introduced and series of benchmark problems were verified in this paper. Compared with the results given by the benchmark, the maximum relative deviation of k eff is 0.09% and the average relative deviation of flux density is about 0.60% in the BWR cells benchmark problem. As for the fixed source problem with isotropic and anisotropic scattering, the results of the 2-D transport module of ARES conform with DORT very well. These numerical results of benchmark problems preliminarily demonstrate that the development process of the 2-D transport module of ARES is right and it is able to provide high precision result. (authors)
CHMTRNS, Non-Equilibrium Chemical Transport Code
International Nuclear Information System (INIS)
Noorishad, J.; Carnahan, C.L.; Benson, L.V.
1998-01-01
1 - Description of program or function: CHMTRNS simulates solute transport for steady one-dimensional fluid flow by convection and diffusion or dispersion in a saturated porous medium based on the assumption of local chemical equilibrium. The chemical interactions included in the model are aqueous-phase complexation, solid-phase ion exchange of bare ions and complexes using the surface complexation model, and precipitation or dissolution of solids. The program can simulate the kinetic dissolution or precipitation for calcite and silica as well as irreversible dissolution of glass. Thermodynamic parameters are temperature dependent and are coupled to a companion heat transport simulator; thus, the effects of transient temperature conditions can be considered. Options for oxidation-reduction (redox) and C-13 fractionation as well as non-isothermal conditions are included. 2 - Method of solution: The governing equations for both reactive chemical and heat transport are discretized in time and space. For heat transport, the Crank-Nicolson approximation is used in conjunction with a LU decomposition and backward substitution solution procedure. To deal with the strong nonlinearity of the chemical transport equations, a generalized Newton-Raphson method is used
Some neutronics and thermal-hydraulics codes for reactor analysis using personal computers
International Nuclear Information System (INIS)
Woodruff, W.L.
1990-01-01
Some neutronics and thermal-hydraulics codes formerly available only for main frame computers may now be run on personal computers. Brief descriptions of the codes are provided. Running times for some of the codes are compared for an assortment of personal and main frame computers. With some limitations in detail, personal computer versions of the codes can be used to solve many problems of interest in reactor analyses at very modest costs. 11 refs., 4 tabs
Energy Technology Data Exchange (ETDEWEB)
Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-05-01
The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)
Benchmark studies of BOUT++ code and TPSMBI code on neutral transport during SMBI
Energy Technology Data Exchange (ETDEWEB)
Wang, Y.H. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Z.H., E-mail: zhwang@swip.ac.cn [Southwestern Institute of Physics, Chengdu 610041 (China); Guo, W., E-mail: wfguo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Ren, Q.L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, A.P.; Xu, M.; Wang, A.K. [Southwestern Institute of Physics, Chengdu 610041 (China); Xiang, N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China)
2017-06-09
SMBI (supersonic molecule beam injection) plays an important role in tokamak plasma fuelling, density control and ELM mitigation in magnetic confinement plasma physics, which has been widely used in many tokamaks. The trans-neut module of BOUT++ code is the only large-scale parallel 3D fluid code used to simulate the SMBI fueling process, while the TPSMBI (transport of supersonic molecule beam injection) code is a recent developed 1D fluid code of SMBI. In order to find a method to increase SMBI fueling efficiency in H-mode plasma, especially for ITER, it is significant to first verify the codes. The benchmark study between the trans-neut module of BOUT++ code and the TPSMBI code on radial transport dynamics of neutral during SMBI has been first successfully achieved in both slab and cylindrical coordinates. The simulation results from the trans-neut module of BOUT++ code and TPSMBI code are consistent very well with each other. Different upwind schemes have been compared to deal with the sharp gradient front region during the inward propagation of SMBI for the code stability. The influence of the WENO3 (weighted essentially non-oscillatory) and the third order upwind schemes on the benchmark results has also been discussed. - Highlights: • A 1D model of SMBI has developed. • Benchmarks of BOUT++ and TPSMBI codes have first been finished. • The influence of the WENO3 and the third order upwind schemes on the benchmark results has also been discussed.
International Nuclear Information System (INIS)
Pautz, A.; Tyobeka, B.; Ivanov, K.
2009-01-01
In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Belicev, P [Vojnotehnicki Inst., Belgrade (Yugoslavia)
1988-07-01
An outline of the problems encountered in the multigroup calculations of the neutron transport in the resonance region is given. The difference between subgroup and multigroup approximation is described briefly. The features of the Monte Carlo code SUBGR are presented. The results of the calculations of the neutron transmission and albedo for infinite iron slabs are given. (author)
Load balancing in highly parallel processing of Monte Carlo code for particle transport
International Nuclear Information System (INIS)
Higuchi, Kenji; Takemiya, Hiroshi; Kawasaki, Takuji
1998-01-01
In parallel processing of Monte Carlo (MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)
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)
International Nuclear Information System (INIS)
Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung
2007-01-01
EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system
Survey of 1 1/2D transport codes
International Nuclear Information System (INIS)
Grad, H.
1978-10-01
A survey is given of a family of classical transport codes, recently termed ''1 1/2D'', which efficiently and accurately follow the evolution of plasma configurations on a long time scale, following coupled changes in plasma shape and topology with transport (but not wave motion). Codes have been constructed and operated (since 1974) which include various combinations of finite beta, general plasma cross-section and aspect, various topologies (Doublet, tearing, reversed-field mirror) including time dependent transitions in topology resulting from external coil variation and plasma transport, with models including (classical) tensor resistivity and heat flow as well as the adiabatic limiting case
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)
International Nuclear Information System (INIS)
Reginatto, M.; Goldhagen, P.
1998-06-01
The problem of analyzing data from a multisphere neutron spectrometer to infer the energy spectrum of the incident neutrons is discussed. The main features of the code MAXED, a computer program developed to apply the maximum entropy principle to the deconvolution (unfolding) of multisphere neutron spectrometer data, are described, and the use of the code is illustrated with an example. A user's guide for the code MAXED is included in an appendix. The code is available from the authors upon request
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.
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.
ACDOS2: a code for neutron-induced activities and dose rates
International Nuclear Information System (INIS)
Ruby, L.; Keney, G.S.; Lagache, J.C.
1981-10-01
In order to anticipate problems from the radioactivation of neutral beam sources as a result of testing, a code has been developed which calculates both the radioactivities produced and the dose rates resulting therefrom. The code ACDOS2 requires neutron source strength and spectral distribution as input, or alternately, the source strength can be calculated internally from an input of neutral beam source parameters. A variety of simple geometries can be specified, and up to 12 times of interest following the shutdown of the neutron source. Radiation attenuating and daughter radioactivities are treated accurately. ACDOS2 is also of use for neutron-induced radioactivation problems involving accelerators, fusion reactors, or fission reactors
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)
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
International Nuclear Information System (INIS)
Stefanovic, D.
1975-09-01
The research work of this contract was oriented towards the study of different methods in neutron transport theory. Authors studied analytical solution of the neutron slowing down transport equation and extension of this solution to include the energy dependence of the anisotropy of neutron scattering. Numerical solution of the fast and resonance transport equation for the case of mixture of scatterers including inelastic effects were also reviewed. They improved the existing formalism for treating the scattering of neutrons on water molecules; Identifying modal analysis as the Galerkin method, general conditions for modal technique applications have been investigated. Inverse problems in transport theory were considered. They obtained the evaluation of an advanced level distribution function, made improvement of the standard formalism for treating the inelastic scattering and development of a cluster nuclear model for this evaluation. Authors studied the neutron transport treatment in space energy groups for criticality calculation of a reactor core, and development of the Monte Carlo sampling scheme from the neutron transport equation
Acceleration of a Monte Carlo radiation transport code
International Nuclear Information System (INIS)
Hochstedler, R.D.; Smith, L.M.
1996-01-01
Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics
Mesh requirements for neutron transport calculations
International Nuclear Information System (INIS)
Askew, J.R.
1967-07-01
Fine-structure calculations are reported for a cylindrical natural uranium-graphite cell using different solution methods (discrete ordinate and collision probability codes) and varying the spatial mesh. It is suggested that of formulations assuming the source constant in a mesh interval the differential approach is generally to be preferred. Due to cancellation between approximations made in the derivation of the finite difference equations and the errors in neglecting source variation, the discrete ordinate code gave a more accurate estimate of fine structure for a given mesh even for unusually coarse representations. (author)
Experimental transport analysis code system in JT-60
International Nuclear Information System (INIS)
Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru
1988-03-01
Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)
Finite element method for solving neutron transport problems
International Nuclear Information System (INIS)
Ferguson, J.M.; Greenbaum, A.
1984-01-01
A finite element method is introduced for solving the neutron transport equations. Our method falls into the category of Petrov-Galerkin solution, since the trial space differs from the test space. The close relationship between this method and the discrete ordinate method is discussed, and the methods are compared for simple test problems
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
ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses
International Nuclear Information System (INIS)
2008-01-01
1 - Description of program or function: The European Reactor Analysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R and D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core
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.
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)
Benchmarking time-dependent neutron problems with Monte Carlo codes
International Nuclear Information System (INIS)
Couet, B.; Loomis, W.A.
1990-01-01
Many nuclear logging tools measure the time dependence of a neutron flux in a geological formation to infer important properties of the formation. The complex geometry of the tool and the borehole within the formation does not permit an exact deterministic modelling of the neutron flux behaviour. While this exact simulation is possible with Monte Carlo methods the computation time does not facilitate quick turnaround of results useful for design and diagnostic purposes. Nonetheless a simple model based on the diffusion-decay equation for the flux of neutrons of a single energy group can be useful in this situation. A combination approach where a Monte Carlo calculation benchmarks a deterministic model in terms of the diffusion constants of the neutrons propagating in the media and their flux depletion rates thus offers the possibility of quick calculation with assurance as to accuracy. We exemplify this approach with the Monte Carlo benchmarking of a logging tool problem, showing standoff and bedding response. (author)
Los Alamos neutral particle transport codes: New and enhanced capabilities
International Nuclear Information System (INIS)
Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Clark, B.A.; Koch, K.R.; Marr, D.R.
1992-01-01
We present new developments in Los Alamos discrete-ordinates transport codes and introduce THREEDANT, the latest in the series of Los Alamos discrete ordinates transport codes. THREEDANT solves the multigroup, neutral-particle transport equation in X-Y-Z and R-Θ-Z geometries. THREEDANT uses computationally efficient algorithms: Diffusion Synthetic Acceleration (DSA) is used to accelerate the convergence of transport iterations, the DSA solution is accelerated using the multigrid technique. THREEDANT runs on a wide range of computers, from scientific workstations to CRAY supercomputers. The algorithms are highly vectorized on CRAY computers. Recently, the THREEDANT transport algorithm was implemented on the massively parallel CM-2 computer, with performance that is comparable to a single-processor CRAY-YMP We present the results of THREEDANT analysis of test problems
PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code
International Nuclear Information System (INIS)
Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.
1976-12-01
Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems
Coupling of 3D neutronics models with the system code ATHLET
International Nuclear Information System (INIS)
Langenbuch, S.; Velkov, K.
1999-01-01
The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)
HADES. A computer code for fast neutron cross section from the Optical Model
International Nuclear Information System (INIS)
Guasp, J.; Navarro, C.
1973-01-01
A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs
International Nuclear Information System (INIS)
Robinson, G.S.
1985-08-01
The calculation of resonance shielding by the subgroup method, as incorporated in the MIRANDA module of the AUS neutronics code system, is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered, AUS results for resonance absorption were high by up to two per cent
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
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
Using MCNP-4C code for design of the thermal neutron beam for neutron radiography at the MNSR
International Nuclear Information System (INIS)
Shaaban, I.
2009-11-01
Studies were carried out for determination of the parameters of a thermal neutron beam at the MNSR reactor (MNSR-30 kW) for neutron radiography in the vertical beam port by using the MCNP-4C (Monte Carlo Neutron - Photon transport). Thermal, epithermal and fast neutron energy ranges were selected as 10 keV respectively. To produce a good neutron beam in terms of intensity and quality, several materials Lead (Pb), Bismuth (Bi), Borated polyethelyene and Alumina Oxide (Al 2 O 3 ) were used as neutron and photon filters. Based on the current design, the L/D of the facility ranges between 125, 110 and 90. The thermal neutron flux at the beam exit is 1.436x10 5 n/cm2 .s ,1.843x10 5 n/cm2 .s and 2.845x10 5 n/cm2 .s respectively, middots with a Cd-ratio of ∼ 2.829, 2.766, 3.191 for the L/D = 125, 110, 90 respectively. The estimated values for gamma doses are 6.705x10 -2 Rem/h and 1.275x10 -1 Rem/h and 2.678x10 -1 Rem/ h with bismuth. The divergent angle of the collimator is 1.348 degree - 2.021 degree. Such neutron beams, if built into the Syrian MNSR reactor, could support the application of NRG in Syria. (author)
Test and validation of the iterative code for the neutrons spectrometry and dosimetry: NSDUAZ
International Nuclear Information System (INIS)
Reyes H, A.; Ortiz R, J. M.; Reyes A, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.
2014-08-01
In this work was realized the test and validation of an iterative code for neutronic spectrometry known as Neutron Spectrometry and Dosimetry of the Universidad Autonoma de Zacatecas (NSDUAZ). This code was designed in a user graph interface, friendly and intuitive in the environment programming of LabVIEW using the iterative algorithm known as SPUNIT. The main characteristics of the program are: the automatic selection of the initial spectrum starting from the neutrons spectra catalog compiled by the International Atomic Energy Agency, the possibility to generate a report in HTML format that shows in graph and numeric way the neutrons flowing and calculates the ambient dose equivalent with base to this. To prove the designed code, the count rates of a spectrometer system of Bonner spheres were used with a detector of 6 LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: 252 Cf and 239 PuBe were used to validate the code, the obtained results were compared against those obtained using the BUNKIUT code. We find that the reconstructed spectra present an error that is inside the limit reported in the literature that oscillates around 15%. Therefore, it was concluded that the designed code presents similar results to those techniques used at the present time. (Author)
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)
Current and anticipated uses of thermalhydraulic and neutronic codes at PSI
Energy Technology Data Exchange (ETDEWEB)
Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G. [Paul Scherrer Institut, Villigen (Switzerland)
1997-07-01
The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together with some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized.
Current and anticipated uses of thermalhydraulic and neutronic codes at PSI
International Nuclear Information System (INIS)
Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G.
1997-01-01
The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together with some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized
Verification of Monte Carlo transport codes by activation experiments
Chetvertkova, Vera
2013-01-01
With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is...
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)
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
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)
Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual
International Nuclear Information System (INIS)
Vergnaud, Th.; Nimal, J.C.; Chiron, M.
2001-01-01
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
Parallel computing solution of Boltzmann neutron transport equation
International Nuclear Information System (INIS)
Ansah-Narh, T.
2010-01-01
The focus of the research was on developing parallel computing algorithm for solving Eigen-values of the Boltzmam Neutron Transport Equation (BNTE) in a slab geometry using multi-grid approach. In response to the problem of slow execution of serial computing when solving large problems, such as BNTE, the study was focused on the design of parallel computing systems which was an evolution of serial computing that used multiple processing elements simultaneously to solve complex physical and mathematical problems. Finite element method (FEM) was used for the spatial discretization scheme, while angular discretization was accomplished by expanding the angular dependence in terms of Legendre polynomials. The eigenvalues representing the multiplication factors in the BNTE were determined by the power method. MATLAB Compiler Version 4.1 (R2009a) was used to compile the MATLAB codes of BNTE. The implemented parallel algorithms were enabled with matlabpool, a Parallel Computing Toolbox function. The option UseParallel was set to 'always' and the default value of the option was 'never'. When those conditions held, the solvers computed estimated gradients in parallel. The parallel computing system was used to handle all the bottlenecks in the matrix generated from the finite element scheme and each domain of the power method generated. The parallel algorithm was implemented on a Symmetric Multi Processor (SMP) cluster machine, which had Intel 32 bit quad-core x 86 processors. Convergence rates and timings for the algorithm on the SMP cluster machine were obtained. Numerical experiments indicated the designed parallel algorithm could reach perfect speedup and had good stability and scalability. (au)
Multidimensional electron-photon transport with standard discrete ordinates codes
International Nuclear Information System (INIS)
Drumm, C.R.
1995-01-01
A method is described for generating electron cross sections that are compatible with standard discrete ordinates codes without modification. There are many advantages of using an established discrete ordinates solver, e.g. immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and man-made radiation environments. The cross sections have been successfully used in the DORT, TWODANT and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electronphoton transport problems
RADTRAN 5: A computer code for transportation risk analysis
International Nuclear Information System (INIS)
Neuhauser, K.S.; Kanipe, F.L.
1991-01-01
RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air
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
Discontinuous nodal schemes applied to the bidimensional neutron transport equation
International Nuclear Information System (INIS)
Delfin L, A.; Valle G, E. Del; Hennart B, J.P.
1996-01-01
In this paper several strong discontinuous nodal schemes are described, starting from the one that has only two interpolation parameters per cell to the one having ten. Their application to the spatial discretization of the neutron transport equation in X-Y geometry is also described, giving, for each one of the nodal schemes, the approximation for the angular neutron flux that includes the set of interpolation parameters and the corresponding polynomial space. Numerical results were obtained for several test problems presenting here the problem with the highest degree of difficulty and their comparison with published results 1,2 . (Author)
Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code
International Nuclear Information System (INIS)
Kim, K.Y.; Kim, K.S.; Kim, H.Y.; Lee, C.C.; Zee, S.Q.
2007-01-01
In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)
A method for solving neutron transport equation
International Nuclear Information System (INIS)
Dimitrijevic, Z.
1993-01-01
The procedure for solving the transport equation by directly integrating for case one-dimensional uniform multigroup medium is shown. The solution is expressed in terms of linear combination of function H n (x,μ), and the coefficient is determined from given conditions. The solution is applied for homogeneous slab of critical thickness. (author)
Development of neutron diffuse scattering analysis code by thin film and multilayer film
International Nuclear Information System (INIS)
Soyama, Kazuhiko
2004-01-01
To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering by thin film, roughness of surface of thin film, correlation function, neutron propagation by thin film, diffuse scattering by DWBA theory, measurement model, SDIFFF (neutron diffuse scattering analysis program by thin film) and simulation results are explained. On neutron diffuse scattering by multilayer film, roughness of multilayer film, principle of diffuse scattering, measurement method and simulation examples by MDIFF (neutron diffuse scattering analysis program by multilayer film) are explained. (S.Y.)To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering
Calculation of neutron spectra for a 252Cf transport cask using ANISN running on a PC
International Nuclear Information System (INIS)
West, L.; Akin, B.P.; Lemley, E.C.
1995-01-01
Neutron spectra have been calculated using the ANISN one-dimensional discrete ordinates code for the case of a 152 Cf source in a transport cask of a particular design. All computations were done on personal computers (PCs) (mostly 486 models) with the ANISN-ORNL (486 version) computer code. With a source of 252 Cf fission neutrons, the neutron flux spectrum in the cask cannot be characterized as open-quotes moderated.close quotes Concern about an appropriate choice for the cross-section data set has led to a comparison, for this application, of three different cross-section libraries: DABL, HILO, and BUGLE-80. Although the cross-section sets were not originally designed for PC use, the libraries have been successfully employed for PC computations. Work with yet another data library, BUGLE-93, is incomplete at this stage. From neutron flux spectra on the surface of the cask, personnel dosimetric quantities (such as dose equivalent) have been determined for the DABL, HILO, and BUGLE-80 ANISN calculations
TIERCE: A code system for particles and radiation transport in thick targets
Energy Technology Data Exchange (ETDEWEB)
Bersillon, O.; Bauge, E.; Borne, F.; Clergeau, J.F.; Collin, M.; Cotten, D.; Delaroche, J.P.; Duarte, H.; Flament, J.L.; Girod, M.; Gosselin, G.; Granier, T.; Hilaire, S.; Morel, P.; Perrier, R.; Romain, P.; Roux, L. [CEA, Bruyeres-le-Chatel (France). Service de Physique Nucleaire
1997-09-01
Over the last few years, a great effort at Bruyeres-le-Chatel has been the development of the TIERCE code system for the transport of particles and radiations in complex geometry. The comparison of calculated results with experimental data, either microscopic (double differential spectra, residual nuclide yield...) or macroscopic (energy deposition, neutron leakage...), shows the need to improve the nuclear reaction models used. We present some new developments concerning data required for the evaporation model in the framework of a microscopic approach. 22 refs., 6 figs.
Neutron transport in irradiation loops (IRENE loop)
International Nuclear Information System (INIS)
Sarsam, Maher.
1980-09-01
This thesis is composed of two parts with different aspects. Part one is a technical description of the loop and its main ancillary facilities as well as of the safety and operational regulations. The measurement methods on the model of the ISIS reactor and on the loop in the OSIRIS reactor are described. Part two deals with the possibility of calculating the powers dissipated by each rod of the fuel cluster, using appropriate computer codes, not only in the reflector but also in the core and to suggest a method of calculation [fr
A 3D transport-based core analysis code for research reactors with unstructured geometry
International Nuclear Information System (INIS)
Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao
2013-01-01
Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results
DANTSYS: A diffusion accelerated neutral particle transport code system
International Nuclear Information System (INIS)
Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O'Dell, R.D.; Walters, W.F.
1995-06-01
The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZΘ symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing
DANTSYS: A diffusion accelerated neutral particle transport code system
Energy Technology Data Exchange (ETDEWEB)
Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.
1995-06-01
The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.
SUNF, Simplified UNF Code, Fast Neutron Calculation by Unified Hauser-Feshbach Theory
International Nuclear Information System (INIS)
Zhang Jingshang
2001-01-01
1 - Description of program or function: The SUNF code is the simplified version of UNF code and is based on the unified Hauser-Feshbach and exciton model. SUNF code has been developed for calculations of fast neutron data for structural materials with neutron energies below 20 MeV. Besides elastic scattering channel, the code may handle decay sequence up to (n,3n) reaction, including 14 reaction channels. The energy spectra can be obtained and the output form is in the ENDF/B-6 format, but in file 5 form. For the ENDF-B-6 output, the incident energies are divided into two types: only cross section calculation; and those including neutron energy spectra. 2 - Methods: Gaussian integration is used for all numerical integration. 3 - Restrictions on the complexity of the problem: The incident energies of neutrons are from 1 KeV to 20 MeV. There are two parameters in this code: incident neutron energies number 'NEL'; and the number of discrete levels of residual nuclei for the first particle emissions 'NLV'. The users can set the values of NEL and NLV according to the storage size of the computer used. The number of discrete levels of residual nuclei for the multi-particle emissions is not greater than 20
A user's manual for the three-dimensional Monte Carlo transport code SPARTAN
International Nuclear Information System (INIS)
Bending, R.C.; Heffer, P.J.H.
1975-09-01
SPARTAN is a general-purpose Monte Carlo particle transport code intended for neutron or gamma transport problems in reactor physics, health physics, shielding, and safety studies. The code used a very general geometry system enabling a complex layout to be described and allows the user to obtain physics data from a number of different types of source library. Special tracking and scoring techniques are used to improve the quality of the results obtained. To enable users to run SPARTAN, brief descriptions of the facilities available in the code are given and full details of data input and job control language, as well as examples of complete calculations, are included. It is anticipated that changes may be made to SPARTAN from time to time, particularly in those parts of the code which deal with physics data processing. The load module is identified by a version number and implementation date, and updates of sections of this manual will be issued when significant changes are made to the code. (author)
Monte Carlo simulation of a coded-aperture thermal neutron camera
International Nuclear Information System (INIS)
Dioszegi, I.; Salwen, C.; Forman, L.
2011-01-01
We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)
International Nuclear Information System (INIS)
Takahashi, A.; Rusch, D.
1979-07-01
Some recent neutronics experiments for fusion reactor blankets show that the precise treatment of anisotropic secondary emissions for all types of neutron scattering is needed for neutron transport calculations. In the present work new rigorous methods, i.e. based on non-approximative microscopic neutron balance equations, are applied to treat the anisotropic collision source term in transport equations. The collision source calculation is free from approximations except for the discretization of energy, angle and space variables and includes the rigorous treatment of nonelastic collisions, as far as nuclear data are given. Two methods are presented: first the Ii-method, which relies on existing nuclear data files and then, as an ultimate goal, the I*-method, which aims at the use of future double-differential cross section data, but which is also applicable to the present single-differential data basis to allow a smooth transition to the new data type. An application of the Ii-method is given in the code system NITRAN which employs the Ssub(N)-method to solve the transport equations. Both rigorous methods, the Ii- and the I*-method, are applicable to all radiation transport problems and they can be used also in the Monte-Carlo-method to solve the transport problem. (orig./RW) [de
BRC neutron evaluations of actinides with the TALYS code
International Nuclear Information System (INIS)
Morillon, B.; Romain, P.
2014-01-01
We briefly report here part of the list of problems to overcome in order to build evaluations as predictive as possible for simulation of criticality benchmarks. Dispersive potential, large coupling scheme and neutron inelastic scattering are the most crucial points of this list of problems. Different tools to distinguish differences between evaluations are also presented. (authors)
International Nuclear Information System (INIS)
Bresard, I.; Diop, C.M.; Giancarli, L.; Gervaise, F.
1991-01-01
In the frame of the ITER tokamak project, the streaming of neutrons through pumping ducts up to the properly so called pumping system is studied. The gas evacuation device of the ITER plasma consists of a set of vacuum pumps which are located in a room which is outside the main machine building. These pumps receive the exhaust gas through several pumping ducts with a cross section of about four square meters and a length of about ten meters. Although insensitive to the magnetic field, the 14 MeV neutrons from plasma D-T thermonuclear reactions can penetrate in the divertor and reach the room pumping device by propagation through the bent ducts. Different components of this system, such as the bellows, turbomolecular pumps, etc., are irradiated and that raises radiation problems. In this study we determine, by using 3D Monte Carlo transport code TRIPOLI-2, neutron fluxes, dose rates and heatings due to neutrons which have streamed out the plasma through the bent ducts, at several points of the pumping room. Results show the neutron flux attenuation reachs a factor 10 -5 from plasma chamber to the pumping hall; the neutron heatings are estimated to 1.9x10 -3 W/cm 3 in bellow stainless steel at duct entrance, and 8x10 -7 W/cm 3 in the turbopumping stainless steel structure, inside pumping hall. The neutron fluxes obtained will be used to compute gamma source produced by radiative, inelastic process and gamma rays from formed activation products. Then, the knowledge of gamma source will allow to compute gamma dose rate and heating. The dose rates and heatings obtained will contribute to the definition of the ITER pumping system technical options and to establish pumping hall access conditions, also. (orig.)
Neutron shielding point kernel integral calculation code for personal computer: PKN-pc
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sakamoto, Yukio; Nakane, Yoshihiro; Tomita, Ken-ichi; Kurosawa, Naohiro.
1994-07-01
A personal computer version of PKN code, PKN-pc, has been developed to calculate neutron and secondary gamma-ray 1cm depth dose equivalents in water, ordinary concrete and iron for neutron source. Characteristics of PKN code are, to able to calculate dose equivalents in multi-layer three-dimensional system, which are described with two-dimensional surface, for monoenergetic neutron source from 0.01 to 14.9 MeV, 252 Cf fission and 241 Am-Be neutron source quick and easily. In addition to these features, the PKN-pc is possible to process interactive input and to get graphical system configuration and graphical results easily. (author)
FLUKA A multi-particle transport code (program version 2005)
Ferrari, A; Fassò, A; Ranft, Johannes
2005-01-01
This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner’s guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.
User's manual for the Oak Ridge Tokamak Transport Code
International Nuclear Information System (INIS)
Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.
1977-02-01
A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche
Comparison of Neutron Cross-Sections Using IAEA Nuclear Codes ''ABAREX'' and ''SCAT2''
International Nuclear Information System (INIS)
Myint Myint Moe; Win Sin; Sein Htoon
2004-05-01
Moel calculations can be used to provide nuclear data for applications in science and technology. The energy averaged neutron induced nuclear reaction cross-sections particular for Al-27, Mg-24, Cr-52, Mn-55, Zn-64 and U-238 with neutrons of energy (0.005 to 10 MeV) are calculated using IAEA nuclear codes ''ABAREX'' and ''SCAT2''. The results are compared with those given in ENDF- 3 nuclear data
Transport code and nuclear data in intermediate energy region
Energy Technology Data Exchange (ETDEWEB)
Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.
1998-11-01
We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)
Transport code and nuclear data in intermediate energy region
International Nuclear Information System (INIS)
Hasegawa, Akira; Odama, Naomitsu; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.
1998-01-01
We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)
Directory of Open Access Journals (Sweden)
Wonkyeong Kim
2015-01-01
Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.
Neutronics analysis of Dalat Nuclear Research Reactor by MVP/GMVP code
International Nuclear Information System (INIS)
Nguyen Kien Cuong; Toru Obara
2008-01-01
The paper presents neutronics calculation for Dalat Nuclear Research Reactor (DNRR) to validate MVP/GMVP Code. Beside fresh core calculation, burnt core and burn up distribution were also carried out and compared with experimental data or result obtained from other codes. With complex geometry and operating history like DNRR, burn up calculation by Monte Carlo Method is the better choice owing to the use of exact geometry description and continuous neutron energy in calculation. The discrepancy between calculated data and experimental data is good to compare. By using Monte Carlo method, continuous neutron energy from JENDL3.3 library and combined with burn up calculation, MVP/GMVP Code is a very useful tool for reactor calculation. (author)
Coded moderator approach for fast neutron source detection and localization at standoff
Energy Technology Data Exchange (ETDEWEB)
Littell, Jennifer [Department of Nuclear Engineering, University of Tennessee, 305 Pasqua Engineering Building, Knoxville, TN 37996 (United States); Lukosi, Eric, E-mail: elukosi@utk.edu [Department of Nuclear Engineering, University of Tennessee, 305 Pasqua Engineering Building, Knoxville, TN 37996 (United States); Institute for Nuclear Security, University of Tennessee, 1640 Cumberland Avenue, Knoxville, TN 37996 (United States); Hayward, Jason; Milburn, Robert; Rowan, Allen [Department of Nuclear Engineering, University of Tennessee, 305 Pasqua Engineering Building, Knoxville, TN 37996 (United States)
2015-06-01
Considering the need for directional sensing at standoff for some security applications and scenarios where a neutron source may be shielded by high Z material that nearly eliminates the source gamma flux, this work focuses on investigating the feasibility of using thermal neutron sensitive boron straw detectors for fast neutron source detection and localization. We utilized MCNPX simulations to demonstrate that, through surrounding the boron straw detectors by a HDPE coded moderator, a source-detector orientation-specific response enables potential 1D source localization in a high neutron detection efficiency design. An initial test algorithm has been developed in order to confirm the viability of this detector system's localization capabilities which resulted in identification of a 1 MeV neutron source with a strength equivalent to 8 kg WGPu at 50 m standoff within ±11°.
Image enhancement using MCNP5 code and MATLAB in neutron radiography.
Tharwat, Montaser; Mohamed, Nader; Mongy, T
2014-07-01
This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. Copyright © 2014 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.
1989-03-01
A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)
The OpenMOC method of characteristics neutral particle transport code
International Nuclear Information System (INIS)
Boyd, William; Shaner, Samuel; Li, Lulu; Forget, Benoit; Smith, Kord
2014-01-01
Highlights: • An open source method of characteristics neutron transport code has been developed. • OpenMOC shows nearly perfect scaling on CPUs and 30× speedup on GPUs. • Nonlinear acceleration techniques demonstrate a 40× reduction in source iterations. • OpenMOC uses modern software design principles within a C++ and Python framework. • Validation with respect to the C5G7 and LRA benchmarks is presented. - Abstract: The method of characteristics (MOC) is a numerical integration technique for partial differential equations, and has seen widespread use for reactor physics lattice calculations. The exponential growth in computing power has finally brought the possibility for high-fidelity full core MOC calculations within reach. The OpenMOC code is being developed at the Massachusetts Institute of Technology to investigate algorithmic acceleration techniques and parallel algorithms for MOC. OpenMOC is a free, open source code written using modern software languages such as C/C++ and CUDA with an emphasis on extensible design principles for code developers and an easy to use Python interface for code users. The present work describes the OpenMOC code and illustrates its ability to model large problems accurately and efficiently
Colloid transport code-nuclear user`s manual
Energy Technology Data Exchange (ETDEWEB)
Jain, R. [New Mexico Univ., Albuquerque, NM (United States)
1992-04-03
This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems.
Spatial domain decomposition for neutron transport problems
International Nuclear Information System (INIS)
Yavuz, M.; Larsen, E.W.
1989-01-01
A spatial Domain Decomposition method is proposed for modifying the Source Iteration (SI) and Diffusion Synthetic Acceleration (DSA) algorithms for solving discrete ordinates problems. The method, which consists of subdividing the spatial domain of the problem and performing the transport sweeps independently on each subdomain, has the advantage of being parallelizable because the calculations in each subdomain can be performed on separate processors. In this paper we describe the details of this spatial decomposition and study, by numerical experimentation, the effect of this decomposition on the SI and DSA algorithms. Our results show that the spatial decomposition has little effect on the convergence rates until the subdomains become optically thin (less than about a mean free path in thickness)
Library of neutron cross sections of the Thermos code
International Nuclear Information System (INIS)
Alonso V, G.; Hernandez L, H.
1991-10-01
The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)
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
Deterministic methods to solve the integral transport equation in neutronic
International Nuclear Information System (INIS)
Warin, X.
1993-11-01
We present a synthesis of the methods used to solve the integral transport equation in neutronic. This formulation is above all used to compute solutions in 2D in heterogeneous assemblies. Three kinds of methods are described: - the collision probability method; - the interface current method; - the current coupling collision probability method. These methods don't seem to be the most effective in 3D. (author). 9 figs
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
Complex eigenvalues for neutron transport equation with quadratically anisotropic scattering
International Nuclear Information System (INIS)
Sjoestrand, N.G.
1981-01-01
Complex eigenvalues for the monoenergetic neutron transport equation in the buckling approximation have been calculated for various combinations of linearly and quadratically anisotropic scattering. The results are discussed in terms of the time-dependent case. Tables are given of complex bucklings for real decay constants and of complex decay constants for real bucklings. The results fit nicely into the pattern of real and purely imaginary eigenvalues obtained earlier. (author)
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)
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.)
Whole core neutronics modeling of a TRIGA reactor using integral transport theory
International Nuclear Information System (INIS)
Schwinkendorf, K.N.; Toffer, H.
1990-01-01
An innovative analysis approach for performing whole core reactor physics calculations for TRIGA reactors has been employed recently at the Westinghouse Hanford Company. A deterministic transport theory model with sufficient geometric complexity to evaluate asymmetric loading patterns was used. Calculations of this complexity have been performed in the past using Monte Carlo simulation, such as the MCNP code. However, the Monte Carlo calculations are more difficult to prepare and require more computer time. On the Hanford Site CRAY XMP-18 computer, the new methods required less than one-third of the central processing unit time per calculation as compared to an MCNP calculation using 100,000 neutron histories
Fuel management and core design code systems for pressurized water reactor neutronic calculations
International Nuclear Information System (INIS)
Ahnert, C.; Arayones, J.M.
1985-01-01
A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions
International Nuclear Information System (INIS)
Rastogi, B.P.
1989-01-01
This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)
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
DEFF Research Database (Denmark)
Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael
2015-01-01
The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...
Andreasen, M.; Looms, M. C.; Bogena, H. R.; Desilets, D.; Zreda, M. G.; Sonnenborg, T. O.; Jensen, K. H.
2014-12-01
. Additionally, neutron transport modeling, using the extended version of the Monte Carlo N-Particle Transport Code, was conducted. The responses of the reference condition, different amounts of biomass, soil moisture and canopy interception on the cosmic-ray neutron intensity were simulated and compared to the measurements.
Energy Technology Data Exchange (ETDEWEB)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.
The Initial Atmospheric Transport (IAT) Code: Description and Validation
Energy Technology Data Exchange (ETDEWEB)
Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2015-10-01
The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.
Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET)
International Nuclear Information System (INIS)
Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.
1994-02-01
The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses that individuals may have received from operations at the Hanford Site since 1944. This report deals specifically with the atmospheric transport model, Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). RATCHET is a major rework of the MESOILT2 model used in the first phase of the HEDR Project; only the bookkeeping framework escaped major changes. Changes to the code include (1) significant changes in the representation of atmospheric processes and (2) incorporation of Monte Carlo methods for representing uncertainty in input data, model parameters, and coefficients. To a large extent, the revisions to the model are based on recommendations of a peer working group that met in March 1991. Technical bases for other portions of the atmospheric transport model are addressed in two other documents. This report has three major sections: a description of the model, a user's guide, and a programmer's guide. These sections discuss RATCHET from three different perspectives. The first provides a technical description of the code with emphasis on details such as the representation of the model domain, the data required by the model, and the equations used to make the model calculations. The technical description is followed by a user's guide to the model with emphasis on running the code. The user's guide contains information about the model input and output. The third section is a programmer's guide to the code. It discusses the hardware and software required to run the code. The programmer's guide also discusses program structure and each of the program elements
PHITS: Particle and heavy ion transport code system, version 2.23
International Nuclear Information System (INIS)
Niita, Koji; Matsuda, Norihiro; Iwamoto, Yosuke; Sato, Tatsuhiko; Nakashima, Hiroshi; Sakamoto, Yukio; Iwase, Hiroshi; Sihver, Lembit
2010-10-01
A Particle and Heavy-Ion Transport code System PHITS has been developed under the collaboration of JAEA (Japan Atomic Energy Agency), RIST (Research Organization for Information Science and Technology) and KEK (High Energy Accelerator Research Organization). PHITS can deal with the transport of all particles (nucleons, nuclei, mesons, photons, and electrons) over wide energy ranges, using several nuclear reaction models and nuclear data libraries. Geometrical configuration of the simulation can be set with GG (General Geometry) or CG (Combinatorial Geometry). Various quantities such as heat deposition, track length and production yields can be deduced from the simulation, using implemented estimator functions called 'tally'. The code also has a function to draw 2D and 3D figures of the calculated results as well as the setup geometries, using a code ANGEL. Because of these features, PHITS has been widely used for various purposes such as designs of accelerator shielding, radiation therapy and space exploration. Recently PHITS introduces an event generator for particle transport parts in the low energy region. Thus, PHITS was completely rewritten for the introduction of the event generator for neutron-induced reactions in energy region less than 20 MeV. Furthermore, several new tallis were incorporated for estimation of the relative biological effects. This document provides a manual of the new PHITS. (author)
Energy Technology Data Exchange (ETDEWEB)
Page, R.; Jones, J.R.
1997-07-01
Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.
International Nuclear Information System (INIS)
Page, R.; Jones, J.R.
1997-01-01
Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell 'B' Loss of offsite power fault transient
Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code
International Nuclear Information System (INIS)
Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.; Soldevila, M.
2003-01-01
In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented
Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code
Energy Technology Data Exchange (ETDEWEB)
Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B; Soldevila, M [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S/SERMA/LEPP), 91 - Gif sur Yvette (France)
2003-07-01
In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented.
One dimensional neutron kinetics in the TRAC-BF1 code
International Nuclear Information System (INIS)
Weaver, W.L. III; Wagner, K.C.
1987-01-01
The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model
Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C
International Nuclear Information System (INIS)
Suman, H.; Kharita, M. H.; Yousef, S.
2008-02-01
In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)
International Nuclear Information System (INIS)
Devillers, C.
1973-01-01
1 - Nature of physical problem solved: The ANISN system treats neutron and gamma transport in one-dimensional plane, spherical and cylinder geometry. The multigroup cross sections prepared by the programs LIANE and SUPERTOG are processed by the program RETTOG, which produces a binary library with Legendre expansions. The binary library can be updated and edited with the program LGR/B. The photon multigroup cross sections are created with the program GAMLEG/A. If the bulk of the data is too large, the program TAPEMA produces a special group-by-group library. The volume sources are calculated from a reduced set of input data and punched in a format suitable for input to ANISN, using the program PRESOU. The program ANISN calculates fluxes by groups, space intervals, angle and any number of reaction rates. The energy and space dependent fluxes are stored on tape and can be reprocessed, edited and plotted with the program ANISEX, which also permits to calculate supplementary reaction rates. The program ANISN can condense cross sections into a reduced number of groups. The ANISN system is used as a reference system for the evaluation of approximation methods (space-diffusion or point- kernel) or for the preparation of multigroup libraries for 2- dimensional transport codes (DOT). In particular it is used for shielding problems with high attenuation in water reactors and fast reactors. 2 - Method of solution: Method of discrete ordinates. The program has been designed to treat deep penetration with detailed calculation of spectrum as function of angle. Tests for pointwise convergence have also been introduced. 3 - Restrictions on the complexity of the problem: The complexity of the problem is limited by the storage size
SHINE-III. Simple code for skyshine dose calculation up to 3 GeV neutrons
Energy Technology Data Exchange (ETDEWEB)
Tsukiyama, Toshihisa; Tayama, Ryuichi; Handa, Hiroyuki [Hitachi Engineering Co. Ltd., Ibaraki (Japan)] [and others
2000-03-01
Skyshine dose at site boundary is considered as one of the most fundamental issues to get approval of constructing nuclear installations. Skyshine conical beam response functions (CBRF) for high energy neutrons up to 3 GeV are obtained using NMTC-JAERI and MCNP code. This CBRF is fitted to the four parameters equation. Simple code named SHINE-III using this equation with updated data is developed. (author)
Development of general-purpose particle and heavy ion transport monte carlo code
International Nuclear Information System (INIS)
Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji
2002-01-01
The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)
ACDOS2: an improved neutron-induced dose rate code
International Nuclear Information System (INIS)
Lagache, J.C.
1981-06-01
To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere
ACDOS2: an improved neutron-induced dose rate code
Energy Technology Data Exchange (ETDEWEB)
Lagache, J.C.
1981-06-01
To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.
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
International Nuclear Information System (INIS)
Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana
2006-01-01
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments
A group of neutronics calculations in the MNSR using the MCNP-4C code
International Nuclear Information System (INIS)
Khattab, K.; Sulieman, I.
2009-11-01
The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)
A computer code package for electron transport Monte Carlo simulation
International Nuclear Information System (INIS)
Popescu, Lucretiu M.
1999-01-01
A computer code package was developed for solving various electron transport problems by Monte Carlo simulation. It is based on condensed history Monte Carlo algorithm. In order to get reliable results over wide ranges of electron energies and target atomic numbers, specific techniques of electron transport were implemented such as: Moliere multiscatter angular distributions, Blunck-Leisegang multiscatter energy distribution, sampling of electron-electron and Bremsstrahlung individual interactions. Path-length and lateral displacement corrections algorithms and the module for computing collision, radiative and total restricted stopping powers and ranges of electrons are also included. Comparisons of simulation results with experimental measurements are finally presented. (author)
A validation study of the BURNUP and associated options of the MONTE CARLO neutronics code MONK5W
International Nuclear Information System (INIS)
Howard, E.A.
1985-11-01
This is a report on the validation of the burnup option of the Monte Carlo Neutronics Code MONK5W, together with the associated facilities which allow for control rod movements and power changes. The validation uses reference solutions produced by the Deterministic Neutronics Code LWR-WIMS for a 2D model which represents a whole reactor calculation with control rod movements. (author)
Documentation for TRACE: an interactive beam-transport code
International Nuclear Information System (INIS)
Crandall, K.R.; Rusthoi, D.P.
1985-01-01
TRACE is an interactive, first-order, beam-dynamics computer program. TRACE includes space-charge forces and mathematical models for a number of beamline elements not commonly found in beam-transport codes, such as permanent-magnet quadrupoles, rf quadrupoles, rf gaps, accelerator columns, and accelerator tanks. TRACE provides an immediate graphic display of calculative results, has a powerful and easy-to-use command procedure, includes eight different types of beam-matching or -fitting capabilities, and contains its own internal HELP package. This report describes the models and equations used for each of the transport elements, the fitting procedures, and the space-charge/emittance calculations, and provides detailed instruction for using the code
Simulation code for the interaction of 14 MeV neutrons on cells
Energy Technology Data Exchange (ETDEWEB)
Nenot, M.L.; Alard, J.P.; Dionet, C.; Arnold, J.; Tchirkov, A.; Meunier, H.; Bodez, V.; Rapp, M.; Verrelle, P
2002-07-01
The structure of the survival curve of melanoma cells irradiated by 14 MeV neutrons displays unusual features at very low dose rate where a marked increase in cell killings at 0.05 Gy is followed by a plateau for survival from 0.1 to 0.32 Gy. In parallel a simulation code was constructed for the interaction of 14 MeV neutrons with cellular cultures. The code describes the interaction of the neutrons with the atomic nuclei of the cellular medium and of the external medium (flask culture and culture medium), and is used to compute the deposited energy into the cell volume. It was found that the large energy transfer events associated with heavy charged recoil can occur and that a large part of the energy deposition events are due to recoil protons emitted from the external medium. It is suggested that such events could partially explain the experimental results. (author)
The neutrons flux density calculations by Monte Carlo code for the double heterogeneity fuel
International Nuclear Information System (INIS)
Gurevich, M.I.; Brizgalov, V.I.
1994-01-01
This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig
International Nuclear Information System (INIS)
Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.
1976-03-01
The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)
Verification of RRC Ki code package for neutronic calculations of WWER core with GD
International Nuclear Information System (INIS)
Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.
2001-01-01
The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)
grmonty: A MONTE CARLO CODE FOR RELATIVISTIC RADIATIVE TRANSPORT
International Nuclear Information System (INIS)
Dolence, Joshua C.; Gammie, Charles F.; Leung, Po Kin; Moscibrodzka, Monika
2009-01-01
We describe a Monte Carlo radiative transport code intended for calculating spectra of hot, optically thin plasmas in full general relativity. The version we describe here is designed to model hot accretion flows in the Kerr metric and therefore incorporates synchrotron emission and absorption, and Compton scattering. The code can be readily generalized, however, to account for other radiative processes and an arbitrary spacetime. We describe a suite of test problems, and demonstrate the expected N -1/2 convergence rate, where N is the number of Monte Carlo samples. Finally, we illustrate the capabilities of the code with a model calculation, a spectrum of the slowly accreting black hole Sgr A* based on data provided by a numerical general relativistic MHD model of the accreting plasma.
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)
Automatic modeling for the monte carlo transport TRIPOLI code
International Nuclear Information System (INIS)
Zhang Junjun; Zeng Qin; Wu Yican; Wang Guozhong; FDS Team
2010-01-01
TRIPOLI, developed by CEA, France, is Monte Carlo particle transport simulation code. It has been widely applied to nuclear physics, shielding design, evaluation of nuclear safety. However, it is time-consuming and error-prone to manually describe the TRIPOLI input file. This paper implemented bi-directional conversion between CAD model and TRIPOLI model. Its feasibility and efficiency have been demonstrated by several benchmarking examples. (authors)
BALDUR: a one-dimensional plasma transport code
International Nuclear Information System (INIS)
Singer, C.E.; Post, D.E.; Mikkelsen, D.R.
1986-07-01
The purpose of BALDUR is to calculate the evolution of plasma parameters in an MHD equilibrium which can be approximated by concentric circular flux surfaces. Transport of up to six species of ionized particles, of electron and ion energy, and of poloidal magnetic flux is computed. A wide variety of source terms are calculated including those due to neutral gas, fusion, and auxiliary heating. The code is primarily designed for modeling tokamak plasmas but could be adapted to other toroidal confinement systems
Parallelization of a Monte Carlo particle transport simulation code
Hadjidoukas, P.; Bousis, C.; Emfietzoglou, D.
2010-05-01
We have developed a high performance version of the Monte Carlo particle transport simulation code MC4. The original application code, developed in Visual Basic for Applications (VBA) for Microsoft Excel, was first rewritten in the C programming language for improving code portability. Several pseudo-random number generators have been also integrated and studied. The new MC4 version was then parallelized for shared and distributed-memory multiprocessor systems using the Message Passing Interface. Two parallel pseudo-random number generator libraries (SPRNG and DCMT) have been seamlessly integrated. The performance speedup of parallel MC4 has been studied on a variety of parallel computing architectures including an Intel Xeon server with 4 dual-core processors, a Sun cluster consisting of 16 nodes of 2 dual-core AMD Opteron processors and a 200 dual-processor HP cluster. For large problem size, which is limited only by the physical memory of the multiprocessor server, the speedup results are almost linear on all systems. We have validated the parallel implementation against the serial VBA and C implementations using the same random number generator. Our experimental results on the transport and energy loss of electrons in a water medium show that the serial and parallel codes are equivalent in accuracy. The present improvements allow for studying of higher particle energies with the use of more accurate physical models, and improve statistics as more particles tracks can be simulated in low response time.
Post-merger evolution of a neutron star-black hole binary with neutrino transport
Foucart, Francois; O'Connor, Evan; Roberts, Luke; Duez, Matthew D.; Haas, Roland; Kidder, Lawrence E.; Ott, Christian D.; Pfeiffer, Harald P.; Scheel, Mark A.; Szilagyi, Bela
2015-06-01
We present a first simulation of the post-merger evolution of a black hole-neutron star binary in full general relativity using an energy-integrated general-relativistic truncated moment formalism for neutrino transport. We describe our implementation of the moment formalism and important tests of our code, before studying the formation phase of an accretion disk after a black hole-neutron star merger. We use as initial data an existing general-relativistic simulation of the merger of a neutron star of mass 1.4 M⊙ with a black hole of mass 7 M⊙ and dimensionless spin χBH=0.8 . Comparing with a simpler leakage scheme for the treatment of the neutrinos, we find noticeable differences in the neutron-to-proton ratio in and around the disk, and in the neutrino luminosity. We find that the electron neutrino luminosity is much lower in the transport simulations, and that both the disk and the disk outflows are less neutron rich. The spatial distribution of the neutrinos is significantly affected by relativistic effects, due to large velocities and curvature in the regions of strongest emission. Over the short time scale evolved, we do not observe purely neutrino-driven outflows. However, a small amount of material (3 ×10-4M⊙ ) is ejected in the polar region during the circularization of the disk. Most of that material is ejected early in the formation of the disk, and is fairly neutron rich (electron fraction Ye˜0.15 - 0.25 ). Through r-process nucleosynthesis, that material should produce high-opacity lanthanides in the polar region, and could thus affect the light curve of radioactively powered electromagnetic transients. We also show that by the end of the simulation, while the bulk of the disk remains neutron rich (Ye˜0.15 - 0.2 and decreasing), its outer layers have a higher electron fraction: 10% of the remaining mass has Ye>0.3 . As that material would be the first to be unbound by disk outflows on longer time scales, and as composition evolution is
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)
The RADionuclide Transport, Removal, and Dose (RADTRAD) code
International Nuclear Information System (INIS)
Miller, L.A.; Chanin, D.I.; Lee, J.
1993-01-01
The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ''Accident Source Terms for Light-Water Nuclear Power Plants.'' The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available for these different removal mechanisms that use, as input, information about the conditions in the plant and predict either a removal coefficient (λ) or decontamination factor. The user may elect to use these models or input a single value for a removal coefficient or decontamination factor
MCT: a Monte Carlo code for time-dependent neutron thermalization problems
International Nuclear Information System (INIS)
Cupini, E.; Simonini, R.
1974-01-01
In the Monte Carlo simulation of pulse source experiments, the neutron energy spectrum, spatial distribution and total density may be required for a long time after the pulse. If the assemblies are very small, as often occurs in the cases of interest, sophisticated Monte Carlo techniques must be applied which force neutrons to remain in the system during the time interval investigated. In the MCT code a splitting technique has been applied to neutrons exceeding assigned target times, and we have found that this technique compares very favorably with more usual ones, such as the expected leakage probability, giving large gains in computational time and variance. As an example, satisfactory asymptotic thermal spectra with a neutron attenuation of 10 -5 were quickly obtained. (U.S.)
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.
Paganini, S
2005-01-01
Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from Galactic cosmic radiation. Crews of future high-speed commercial flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the male adult voxels phantom MAX, developed in the Nuclear Energy Department of Pernambuco Federal University in Brazil, has been coupled with the Monte Carlo simulation code GEANT4. This toolkit, distributed and upgraded from the international scientific community of CERN/Switzerland, simulates thermal to ultrahigh energy neutrons transport and interactions in the matter. The high energy neutrons are pointed as the component that contribute about 70% of the neutron effective dose that represent the 35% to 60% total dose at aircraft altitude. In this research calculations of conversion coefficients from fluence to effective dose are performed for neutrons of energies from 100 MeV ...
International Nuclear Information System (INIS)
Kodeli, I.; Diop, C.M.; Nimal, J.C.
1994-01-01
In the framework of future Boron Neutron Capture Therapy (BNCT) experiments, where cells and animals irradiations are planned at the research reactor of Strasbourg University, the feasibility to obtain a suitable epithermal neutron beam is investigated. The neutron fluence and spectra calculations in the reactor are performed using the 3D Monte Carlo code TRIPOLI-3 and the 2D SN code TWODANT. The preliminary analysis of Al 2 O 3 and Al-Al 2 O 3 filters configurations are carried out in an attempt to optimize the flux characteristics in the beam tube facility. 7 figs., 7 refs
Image enhancement using MCNP5 code and MATLAB in neutron radiography
International Nuclear Information System (INIS)
Tharwat, Montaser; Mohamed, Nader; Mongy, T.
2014-01-01
This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. - Highlights: • This work is applicable for static based film neutron radiography and digital neutron imaging. • MATLAB is a useful tool for imaging enhancement in radiographic film. • Advanced imaging processing is available in the ETRR-2 for imaging processing and data extraction. • The digital imaging system is suitable for complex shapes and sizes, while MATLAB technique is suitable for simple shapes and sizes. • Quantitative measurements are available
International Nuclear Information System (INIS)
Santa Cruz, G.A.
1998-01-01
Full text: A charged particles transport Monte Carlo code, specially designed for the boron neutron capture therapy microdosimetry study was developed. The code allows the use of real tri dimensional problem geometry, using serial microscopy slides from a biological substrate where the 10 B(n, Alpha) 7 Li, 14 N(n,p) 14 C reactions and events can occur. The spatial distribution of sources ( 10 B, 14 N concentrations), regions of interest (where the energy deposition, linear energy transfer and other parameters will be calculated) and other zones (without boron) are obtained from the images. The code is in the benchmarking stage, using geometrically simple cases and experimental data obtained from microdosimetric spectra from TEPC (Tissue Equivalent Proportional Counters) doped with 10 B. It allows to obtain LET spectra discriminated by event classes, chord-length distributions, dose and frequency mean values and visualizations of the spatial energy deposition. A similar version of the code uses bidimensional images from a tissue sample containing a great number of cellular structures. An equivalence between the microdosimetry of a bidimensional case and a tri dimensional one can be done. If the real distribution of 10 B is known, for example by high resolution alpha-track autoradiography, the code can use this information explicitly. (author) [es
International Nuclear Information System (INIS)
Sasamoto, Nobuo; Kotegawa, Hiroshi
1984-11-01
In order to improve computational efficiency of PALLAS code, an accuracy is estimated on the neutron penetration calculation through a concrete shield, using bunched component nuclides of concrete. The calculated fast neutron flux is observed to depend weakly on how the nuclides are bunched. Contrary to this, the calculated thermal neutron fluxes are strongly dependent on the manner of bunching, mainly due to the fact that iron cross section has exceptionally large negative sensitivity to thermal neutron flux. (author)
ABAREX -- A neutron spherical optical-statistical-model code -- A user`s manual
Energy Technology Data Exchange (ETDEWEB)
Smith, A.B. [ed.; Lawson, R.D.
1998-06-01
The contemporary version of the neutron spherical optical-statistical-model code ABAREX is summarized with the objective of providing detailed operational guidance for the user. The physical concepts involved are very briefly outlined. The code is described in some detail and a number of explicit examples are given. With this document one should very quickly become fluent with the use of ABAREX. While the code has operated on a number of computing systems, this version is specifically tailored for the VAX/VMS work station and/or the IBM-compatible personal computer.
Verification of a neutronic code for transient analysis in reactors with Hex-z geometry
Energy Technology Data Exchange (ETDEWEB)
Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)
2012-07-01
Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)
International Nuclear Information System (INIS)
Podlazov, L. N.
1998-01-01
Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions
Analysis of EBR-II neutron and photon physics by multidimensional transport-theory techniques
International Nuclear Information System (INIS)
Jacqmin, R.P.; Finck, P.J.; Palmiotti, G.
1994-01-01
This paper contains a review of the challenges specific to the EBR-II core physics, a description of the methods and techniques which have been developed for addressing these challenges, and the results of some validation studies relative to power-distribution calculations. Numerical tests have shown that the VARIANT nodal code yields eigenvalue and power predictions as accurate as finite difference and discrete ordinates transport codes, at a small fraction of the cost. Comparisons with continuous-energy Monte Carlo results have proven that the errors introduced by the use of the diffusion-theory approximation in the collapsing procedure to obtain broad-group cross sections, kerma factors, and photon-production matrices, have a small impact on the EBR-II neutron/photon power distribution
The OpenMC Monte Carlo particle transport code
International Nuclear Information System (INIS)
Romano, Paul K.; Forget, Benoit
2013-01-01
Highlights: ► An open source Monte Carlo particle transport code, OpenMC, has been developed. ► Solid geometry and continuous-energy physics allow high-fidelity simulations. ► Development has focused on high performance and modern I/O techniques. ► OpenMC is capable of scaling up to hundreds of thousands of processors. ► Results on a variety of benchmark problems agree with MCNP5. -- Abstract: A new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high performance scalable algorithms as well as modern software design practices. The present work describes the methods used in the OpenMC code and demonstrates the performance and accuracy of the code on a variety of problems.