Reliability analysis of neutron transport simulation using Monte Carlo method
International Nuclear Information System (INIS)
Souza, Bismarck A. de; Borges, Jose C.
1995-01-01
This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs
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.)
Development and simulation of various methods for neutron activation analysis
International Nuclear Information System (INIS)
Otgooloi, B.
1993-01-01
Simple methods for neutron activation analysis have been developed. The results on the studies of installation for determination of fluorine in fluorite ores directly on the lorry by fast neutron activation analysis have been shown. Nitrogen in organic materials was shown by N 14 and N 15 activation. The description of the new equipment 'FLUORITE' for fluorate factory have been shortly given. Pu and Be isotope in organic materials, including in wheat, was measured. 25 figs, 19 tabs. (Author, Translated by J.U)
Research on neutron noise analysis stochastic simulation method for α calculation
International Nuclear Information System (INIS)
Zhong Bin; Shen Huayun; She Ruogu; Zhu Shengdong; Xiao Gang
2014-01-01
The prompt decay constant α has significant application on the physical design and safety analysis in nuclear facilities. To overcome the difficulty of a value calculation with Monte-Carlo method, and improve the precision, a new method based on the neutron noise analysis technology was presented. This method employs the stochastic simulation and the theory of neutron noise analysis technology. Firstly, the evolution of stochastic neutron was simulated by discrete-events Monte-Carlo method based on the theory of generalized Semi-Markov process, then the neutron noise in detectors was solved from neutron signal. Secondly, the neutron noise analysis methods such as Rossia method, Feynman-α method, zero-probability method, and cross-correlation method were used to calculate a value. All of the parameters used in neutron noise analysis method were calculated based on auto-adaptive arithmetic. The a value from these methods accords with each other, the largest relative deviation is 7.9%, which proves the feasibility of a calculation method based on neutron noise analysis stochastic simulation. (authors)
International Nuclear Information System (INIS)
Muhammad Fakhrurreza; Kusminanto; Y Sardjono
2014-01-01
In this research has made a reflector design to provide beams of Neutron for BNCT with Californium-252 radioactive source. This collimator is useful to obtain optimum epithermal neutron flux with the smallest impurity radiation (thermal neutron, fast neutron, and gamma). The design process is done using Monte Carlo N-Particle simulation version 5 (MCNP5) code to calculate the neutron flux tally form. The chosen reflector design is the reflectors which use material such as BeO ceramic with 13 cm thick. Moderator use sulfur material with the slope angle of the cone is 30°. From the calculation result, it is obtained that Reflector with 1 gram Californium-252 source can produce a neutron output thermal which has thermal neutron specification 2.23189 x 10 9 n/s.cm 2 , epithermal neutron 3.51548 x 10 9 n/s.cm 2 , and fast neutron 4.82241 x 10 9 n/s.cm 2 From the result, it needs additional collimator because the BNCT requirement. (author)
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)
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
BRAND program complex for neutron-physical experiment simulation by the Monte-Carlo method
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.
1984-01-01
Possibilities of the BRAND program complex for neutron and γ-radiation transport simulation by the Monte-Carlo method are described in short. The complex includes the following modules: geometric module, source module, detector module, modules of simulation of a vector of particle motion direction after interaction and a free path. The complex is written in the FORTRAN langauage and realized by the BESM-6 computer
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.
Simulated workplace neutron fields
International Nuclear Information System (INIS)
Lacoste, V.; Taylor, G.; Rottger, S.
2011-01-01
The use of simulated workplace neutron fields, which aim at replicating radiation fields at practical workplaces, is an alternative solution for the calibration of neutron dosemeters. They offer more appropriate calibration coefficients when the mean fluence-to-dose equivalent conversion coefficients of the simulated and practical fields are comparable. Intensive Monte Carlo modelling work has become quite indispensable for the design and/or the characterization of the produced mixed neutron/photon fields, and the use of Bonner sphere systems and proton recoil spectrometers is also mandatory for a reliable experimental determination of the neutron fluence energy distribution over the whole energy range. The establishment of a calibration capability with a simulated workplace neutron field is not an easy task; to date only few facilities are available as standard calibration fields. (authors)
International Nuclear Information System (INIS)
Rogers, J.W.
1980-10-01
Fast neutron spectrum measurements were made in a Light Water Reactor (LWR) Pressure Vessel Simulator (PVS) to provide neutron spectral definition required to appropriately perform and interpret neutron dosimetry measurements related to fast neutron damage in LWR-PV steels. Proton-recoil proportional counter methods using hydrogen and methane gas-filled detectors were applied to obtain the proton spectra from which the neutron spectra were derived. Cylindrical and spherical geometry detectors were used to cover the neutron energy range between 50 keV and 2 MeV. Results show that the neutron spectra shift in energy distribution toward lower energy between the front and back of a PVS. The relative neutron flux densities increase in this energy range with increasing thickness of the steel. Neutron spectrum fine structure shapes and changes are observed. These results should assist in the generation of more accurate effective cross sections and fluences for use in LWR-PV fast neutron dosimetry and materials damage analyses
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.
Shahmohammadi Beni, Mehrdad; Krstic, Dragana; Nikezic, Dragoslav; Yu, Kwan Ngok
2016-09-01
Many studies on biological effects of neutrons involve dose responses of neutrons, which rely on accurately determined absorbed doses in the irradiated cells or living organisms. Absorbed doses are difficult to measure, and are commonly surrogated with doses measured using separate detectors. The present work describes the determination of doses absorbed in the cell layer underneath a medium column (D A ) and the doses absorbed in an ionization chamber (D E ) from neutrons through computer simulations using the MCNP-5 code, and the subsequent determination of the conversion coefficients R (= D A /D E ). It was found that R in general decreased with increase in the medium thickness, which was due to elastic and inelastic scattering. For 2-MeV neutrons, conspicuous bulges in R values were observed at medium thicknesses of about 500, 1500, 2500 and 4000 μm, and these were attributed to carbon, oxygen and nitrogen nuclei, and were reflections of spikes in neutron interaction cross sections with these nuclei. For 0.1-MeV neutrons, no conspicuous bulges in R were observed (except one at ~2000 μm that was due to photon interactions), which was explained by the absence of prominent spikes in the interaction cross-sections with these nuclei for neutron energies <0.1 MeV. The ratio R could be increased by ~50% for small medium thickness if the incident neutron energy was reduced from 2 MeV to 0.1 MeV. As such, the absorbed doses in cells (D A ) would vary with the incident neutron energies, even when the absorbed doses shown on the detector were the same. © The Author 2016. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.
Methods of neutron spectrometry
International Nuclear Information System (INIS)
Doerschel, B.
1981-01-01
The different methods of neutron spectrometry are based on the direct measurement of neutron velocity or on the use of suitable energy-dependent interaction processes. In the latter case the measuring effect of a detector is connected with the searched neutron spectrum by an integral equation. The solution needs suitable unfolding procedures. The most important methods of neutron spectrometry are the time-of-flight method, the crystal spectrometry, the neutron spectrometry by use of elastic collisions with hydrogen nuclei, and neutron spectrometry with the aid of nuclear reactions, especially of the neutron-induced activation. The advantages and disadvantages of these methods are contrasted considering the resolution, the measurable energy range, the sensitivity, and the experimental and computational efforts. (author)
Simulation of neutron transport process, photons and charged particles within the Monte Carlo method
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.; Artamonov, S.N.; Bolonkina, G.V.; Lomtev, V.L.; Pupko, S.V.
1991-01-01
Description is given to the program system BRAND designed for the accurate solution of non-stationary transport equation of neutrons, photons and charged particles in the conditions of real three-dimensional geometry. An extensive set of local and non-local estimates provides an opportunity of calculating a great set of linear functionals normally being of interest in the calculation of reactors, radiation protection and experiment simulation. The process of particle interaction with substance is simulated on the basis of individual non-group data on each isotope of the composition. 24 refs
Energy Technology Data Exchange (ETDEWEB)
Morillon, B.
1996-12-31
With most of the traditional and contemporary techniques, it is still impossible to solve the transport equation if one takes into account a fully detailed geometry and if one studies precisely the interactions between particles and matters. Only the Monte Carlo method offers such a possibility. However with significant attenuation, the natural simulation remains inefficient: it becomes necessary to use biasing techniques where the solution of the adjoint transport equation is essential. The Monte Carlo code Tripoli has been using such techniques successfully for a long time with different approximate adjoint solutions: these methods require from the user to find out some parameters. If this parameters are not optimal or nearly optimal, the biases simulations may bring about small figures of merit. This paper presents a description of the most important biasing techniques of the Monte Carlo code Tripoli ; then we show how to calculate the importance function for general geometry with multigroup cases. We present a completely automatic biasing technique where the parameters of the biased simulation are deduced from the solution of the adjoint transport equation calculated by collision probabilities. In this study we shall estimate the importance function through collision probabilities method and we shall evaluate its possibilities thanks to a Monte Carlo calculation. We compare different biased simulations with the importance function calculated by collision probabilities for one-group and multigroup problems. We have run simulations with new biasing method for one-group transport problems with isotropic shocks and for multigroup problems with anisotropic shocks. The results show that for the one-group and homogeneous geometry transport problems the method is quite optimal without splitting and russian roulette technique but for the multigroup and heterogeneous X-Y geometry ones the figures of merit are higher if we add splitting and russian roulette technique.
Energy Technology Data Exchange (ETDEWEB)
Saha Ray, S., E-mail: santanusaharay@yahoo.com; Patra, A.
2014-10-15
Highlights: • A stationary transport equation has been solved using the technique of Haar wavelet collocation method. • This paper intends to provide the great utility of Haar wavelets to nuclear science problem. • In the present paper, two-dimensional Haar wavelets are applied. • The proposed method is mathematically very simple, easy and fast. - Abstract: In this paper the numerical solution for the fractional order stationary neutron transport equation is presented using Haar wavelet Collocation Method (HWCM). Haar wavelet collocation method is efficient and powerful in solving wide class of linear and nonlinear differential equations. This paper intends to provide an application of Haar wavelets to nuclear science problems. This paper describes the application of Haar wavelets for the numerical solution of fractional order stationary neutron transport equation in homogeneous medium with isotropic scattering. The proposed method is mathematically very simple, easy and fast. To demonstrate about the efficiency and applicability of the method, two test problems are discussed.
Energy Technology Data Exchange (ETDEWEB)
James, Jonathan A. [Open University, Materials Engineering, Walton Hall, Milton Keynes, Buckinghamshire MK7 6AA (United Kingdom)]. E-mail: j.a.j.james@open.ac.uk; Edwards, Lyndon [Open University, Materials Engineering, Walton Hall, Milton Keynes, Buckinghamshire MK7 6AA (United Kingdom)
2007-02-11
Neutron stress measurements require specimens of complex geometry to be speedily and accurately positioned and oriented with respect to the neutron beam. Recognition that a majority of the specimen positioning systems in use at strain scanning facilities are effectively serial robot manipulators, suggests that the methods of serial robot kinematic modelling may be applied to advantage. The adoption of robotics methods provides a simple and reliable framework for controlling positioning systems of arbitrary geometry and complexity. In addition the numerical solution of the inverse kinematic problem is facilitated, allowing specimens to be automatically positioned and orientated so that pre-determined strain components are measured. It is also shown that, given sufficient degrees of freedom, a secondary characteristic of the measurement position such as the measurement count time may be simultaneously optimised.
Validation of the simulator neutronics model
International Nuclear Information System (INIS)
Gregory, M.V.
1984-01-01
The neutronics model in the SRP reactor training simulator computes the variation with time of the neutron population in the reactor core. The power output of a reactor is directly proportional to the neutron population, thus in a very real sense the neutronics model determines the response of the simulator. The geometrical complexity of the reactor control system in SRP reactors requires the neutronics model to provide a detailed, 3D representation of the reactor core. Existing simulator technology does not allow such a detailed representation to run in real-time in a minicomputer environment, thus an entirely different approach to the problem was required. A prompt jump method has been developed in answer to this need
Neutron source multiplication method
International Nuclear Information System (INIS)
Clayton, E.D.
1985-01-01
Extensive use has been made of neutron source multiplication in thousands of measurements of critical masses and configurations and in subcritical neutron-multiplication measurements in situ that provide data for criticality prevention and control in nuclear materials operations. There is continuing interest in developing reliable methods for monitoring the reactivity, or k/sub eff/, of plant operations, but the required measurements are difficult to carry out and interpret on the far subcritical configurations usually encountered. The relationship between neutron multiplication and reactivity is briefly discussed and data presented to illustrate problems associated with the absolute measurement of neutron multiplication and reactivity in subcritical systems. A number of curves of inverse multiplication have been selected from a variety of experiments showing variations observed in multiplication during the course of critical and subcritical experiments where different methods of reactivity addition were used, with different neutron source detector position locations. Concern is raised regarding the meaning and interpretation of k/sub eff/ as might be measured in a far subcritical system because of the modal effects and spectrum differences that exist between the subcritical and critical systems. Because of this, the calculation of k/sub eff/ identical with unity for the critical assembly, although necessary, may not be sufficient to assure safety margins in calculations pertaining to far subcritical systems. Further study is needed on the interpretation and meaning of k/sub eff/ in the far subcritical system
Methods for absorbing neutrons
Guillen, Donna P [Idaho Falls, ID; Longhurst, Glen R [Idaho Falls, ID; Porter, Douglas L [Idaho Falls, ID; Parry, James R [Idaho Falls, ID
2012-07-24
A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cmK. Apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.
Energy Technology Data Exchange (ETDEWEB)
Falcao, R.C.; Facure, A. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Santini, E.S. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Silva, A.X. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear
2005-07-01
In this work, the MCNP code is used to simulate the transport of neutrons in a room of radiotherapy, whose shieldings are designed according to the method of skyshine (scattering in the atmosphere). The simulations are compared with the results obtained from empirically established expressions, which are normally used for designing the ceilings of the rooms facilities, ensuring that dose rates (neutrons + photons) around them do not exceed the maximum limits allowed by the standards of the CNEN. Good agreement is observed between the doses calculated according to these expressions and those obtained through simulation by Monte Carlo in the case of rooms without ceiling, and an overestimate of the calculations by a factor 2 or 3 in relation to the simulations, in the case of rooms with ceiling.
Mathematical methods in neutronics
International Nuclear Information System (INIS)
Planchard, J.
1995-01-01
This book presents the mathematical theory of nuclear reactors. It applies to engineers in neutronics and applied mathematicians. After a recall of the elementary notions of neutronics and of diffusion-type partial derivative equations, the theory of reactors criticality calculation is described. (J.S.)
International Nuclear Information System (INIS)
Yamamoto, Toshihiro
2014-01-01
Highlights: • The cross power spectral density in ADS has correlated and uncorrelated components. • A frequency domain Monte Carlo method to calculate the uncorrelated one is developed. • The method solves the Fourier transformed transport equation. • The method uses complex-valued weights to solve the equation. • The new method reproduces well the CPSDs calculated with time domain MC method. - Abstract: In an accelerator driven system (ADS), pulsed spallation neutrons are injected at a constant frequency. The cross power spectral density (CPSD), which can be used for monitoring the subcriticality of the ADS, is composed of the correlated and uncorrelated components. The uncorrelated component is described by a series of the Dirac delta functions that occur at the integer multiples of the pulse repetition frequency. In the present paper, a Monte Carlo method to solve the Fourier transformed neutron transport equation with a periodically pulsed neutron source term has been developed to obtain the CPSD in ADSs. Since the Fourier transformed flux is a complex-valued quantity, the Monte Carlo method introduces complex-valued weights to solve the Fourier transformed equation. The Monte Carlo algorithm used in this paper is similar to the one that was developed by the author of this paper to calculate the neutron noise caused by cross section perturbations. The newly-developed Monte Carlo algorithm is benchmarked to the conventional time domain Monte Carlo simulation technique. The CPSDs are obtained both with the newly-developed frequency domain Monte Carlo method and the conventional time domain Monte Carlo method for a one-dimensional infinite slab. The CPSDs obtained with the frequency domain Monte Carlo method agree well with those with the time domain method. The higher order mode effects on the CPSD in an ADS with a periodically pulsed neutron source are discussed
International Nuclear Information System (INIS)
Husin Wagiran; Wan Mohd Nasir Wan Kadir
1997-01-01
In neutron scattering processes, the effect of multiple scattering is to cause an effective increase in the measured cross-sections due to increase on the probability of neutron scattering interactions in the sample. Analysis of how the effective cross-section varies with thickness is very complicated due to complicated sample geometries and the variations of scattering cross-section with energy. Monte Carlo method is one of the possible method for treating the multiple scattering processes in the extended sample. In this method a lot of approximations have to be made and the accurate data of microscopic cross-sections are needed at various angles. In the present work, a Monte Carlo simulation programme suitable for a small computer was developed. The programme was capable to predict the number of neutrons scattered from various thickness of aluminium samples at all possible angles between 00 to 36011 with 100 increments. In order to make the the programme not too complicated and capable of being run on microcomputer with reasonable time, the calculations was done in two dimension coordinate system. The number of neutrons predicted from this model show in good agreement with previous experimental results
Binary neutron star merger simulations
Energy Technology Data Exchange (ETDEWEB)
Bruegmann, Bernd [Jena Univ. (Germany)
2016-11-01
Our research focuses on the numerical tools necessary to solve Einstein's equations. In recent years we have been particularly interested in spacetimes consisting of two neutron stars in the final stages of their evolution. Because of the emission of gravitational radiation, the objects are driven together to merge; the emitted gravitational wave signal is visualized. This emitted gravitational radiation carries energy and momentum away from the system and contains information about the system. Late last year the Laser Interferometer Gravitational-wave Observatory (LIGO) began searches for these gravitational wave signals at a sensitivity at which detections are expected. Although such systems can radiate a significant amount of their total mass-energy in gravitational waves, the gravitational wave signals one expects to receive on Earth are not strong, since sources of gravitational waves are often many millions of light years away. Therefore one needs accurate templates for the radiation one expects from such systems in order to be able to extract them out of the detector's noise. Although analytical models exist for compact binary systems when the constituents are well separated, we need numerical simulation to investigate the last orbits before merger to obtain accurate templates and validate analytical approximations. Due to the strong nonlinearity of the equations and the large separation of length scales, these simulations are computationally demanding and need to be run on large supercomputers. When matter is present the computational cost as compared to pure black hole (vacuum) simulations increases even more due to the additional matter fields. But also more interesting astrophysical phenomena can happen. In fact, there is the possibility for a strong electromagnetic signal from the merger (e.g., a short gamma-ray burst or lower-energy electromagnetic signatures from the ejecta) and significant neutrino emission. Additionally, we can expect that
International Nuclear Information System (INIS)
Titt, U.; Newhauser, W. D.
2005-01-01
Proton therapy facilities are shielded to limit the amount of secondary radiation to which patients, occupational workers and members of the general public are exposed. The most commonly applied shielding design methods for proton therapy facilities comprise semi-empirical and analytical methods to estimate the neutron dose equivalent. This study compares the results of these methods with a detailed simulation of a proton therapy facility by using the Monte Carlo technique. A comparison of neutron dose equivalent values predicted by the various methods reveals the superior accuracy of the Monte Carlo predictions in locations where the calculations converge. However, the reliability of the overall shielding design increases if simulation results, for which solutions have not converged, e.g. owing to too few particle histories, can be excluded, and deterministic models are being used at these locations. Criteria to accept or reject Monte Carlo calculations in such complex structures are not well understood. An optimum rejection criterion would allow all converging solutions of Monte Carlo simulation to be taken into account, and reject all solutions with uncertainties larger than the design safety margins. In this study, the optimum rejection criterion of 10% was found. The mean ratio was 26, 62% of all receptor locations showed a ratio between 0.9 and 10, and 92% were between 1 and 100. (authors)
Energy Technology Data Exchange (ETDEWEB)
Kim, Je Hyun; Shim, Chang Ho [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of); Kim, Sung Hyun [Nuclear Fuel Cycle Waste Treatment Research Division, Research Reactor Institute, Kyoto University, Osaka (Japan); Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo [Ionizing Radiation Center, Nuclear Fuel Cycle Waste Treatment Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho [Ionizing Radiation Center, Korea Research Institute of Standards and Science, Daejeon (Korea, Republic of)
2016-12-15
For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.
International Nuclear Information System (INIS)
Kim, Je Hyun; Shim, Chang Ho; Kim, Sung Hyun; Choe, Jung Hun; Cho, In Hak; Park, Hwan Seo; Park, Hyun Seo; Kim, Jung Ho; Kim, Yoon Ho
2016-01-01
For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers
International Nuclear Information System (INIS)
Roche, C.T.; Strauss, M.G.; Brenner, R.
1984-01-01
The spatial linearity and resolution of Anger-type neutron-position scintillation detectors are studied using a semi-empirical model. Detector optics with either an air gap or optical grease between the scintillator and the dispersive light guide are considered. Three signal processing methods which truncate signals from PMT's distant from the scintillation are compared with the linear resistive weighting method. Air gap optics yields a 15% improvement in spatial resolution and 50% reduction in differential and integral nonlinearity relative to grease coupled optics, using linear processing. Using signal truncation instead of linear processing improves the resolution 15-20% for the air gap and 20-30% for the grease coupling case. Thus, the initial discrepancy in the resolution between the two optics nearly vanished, however the linearity of the grease coupled system is still significantly poorer
Monte Carlo simulations of neutron scattering instruments
International Nuclear Information System (INIS)
Aestrand, Per-Olof; Copenhagen Univ.; Lefmann, K.; Nielsen, K.
2001-01-01
A Monte Carlo simulation is an important computational tool used in many areas of science and engineering. The use of Monte Carlo techniques for simulating neutron scattering instruments is discussed. The basic ideas, techniques and approximations are presented. Since the construction of a neutron scattering instrument is very expensive, Monte Carlo software used for design of instruments have to be validated and tested extensively. The McStas software was designed with these aspects in mind and some of the basic principles of the McStas software will be discussed. Finally, some future prospects are discussed for using Monte Carlo simulations in optimizing neutron scattering experiments. (R.P.)
The “neutron channel design”—A method for gaining the desired neutrons
Directory of Open Access Journals (Sweden)
G. Hu
2016-12-01
Full Text Available The neutrons with desired parameters can be obtained after initial neutrons penetrating various structure and component of the material. A novel method, the “neutron channel design”, is proposed in this investigation for gaining the desired neutrons. It is established by employing genetic algorithm (GA combining with Monte Carlo software. This method is verified by obtaining 0.01eV to 1.0eV neutrons from the Compact Accelerator-driven Neutron Source (CANS. One layer polyethylene (PE moderator was designed and installed behind the beryllium target in CANS. The simulations and the experiment for detection the neutrons were carried out. The neutron spectrum at 500cm from the PE moderator was simulated by MCNP and PHITS software. The counts of 0.01eV to 1.0eV neutrons were simulated by MCNP and detected by the thermal neutron detector in the experiment. These data were compared and analyzed. Then this method is researched on designing the complex structure of PE and the composite material consisting of PE, lead and zirconium dioxide.
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
Simulation of a complete inelastic neutron scattering experiment
DEFF Research Database (Denmark)
Edwards, H.; Lefmann, K.; Lake, B.
2002-01-01
A simulation of an inelastic neutron scattering experiment on the high-temperature superconductor La2-xSrxCuO4 is presented. The complete experiment, including sample, is simulated using an interface between the experiment control program and the simulation software package (McStas) and is compared...... with the experimental data. Simulating the entire experiment is an attractive alternative to the usual method of convoluting the model cross section with the resolution function, especially if the resolution function is nontrivial....
Simulating Makrofol as a detector for neutron-induced recoils
International Nuclear Information System (INIS)
Zhang, G.; Becker, F.; Urban, M.; Xuan, Y.
2011-01-01
The response of solid-state nuclear track detector is extremely dependent on incident angles of neutrons, which determine the angular distribution of secondary particles. In this paper, the authors present a method to investigate the angular response of Makrofol detectors. Using the C++-based Monte-Carlo tool-kit Geant4 in combination with SRIM and our MATLAB codes, we simulated the angular response of Makrofol. The simulations were based on the restricted energy loss model, and the concept of energy threshold and critical angle. Experiments were carried out with 252 Cf neutrons to verify the simulation results. (authors)
Neutron lifetime well logging methods and apparatus
International Nuclear Information System (INIS)
Paap, H.J.; Pitts, R.W.
1974-01-01
A method for investigating the earth formations surrounding a well borehole, comprising the steps of: continuously generating high energy neutrons in the borehole and bombarding the surrounding media with such neutrons to develop a cloud of thermal neutrons therein; modulating the intensity of said high energy neutrons harmonically as a function of time in order to intensity modulate said cloud of thermal neutrons as a function of time; and measuring a time-dependant thermal neutron characteristic of said intensity modulated cloud of thermal neutrons
International Nuclear Information System (INIS)
Wu, Jian-Chun; Feng, Qi-Jie; Liu, Xian-Kun; Zhan, Chang-Yong; Zou, Yu; Liu, Yao-Guang
2016-01-01
A multiscale sequence of simulation should be used to predict properties of materials under irradiation. Binary collision theory and molecular dynamics (MDs) method are commonly used to characterize the displacement cascades induced by neutrons in a material. In order to reduce the clock time spent for the MD simulation of damages induced by high-energy primary knock-on atoms (PKAs), the damage zones were split into sub-cascade according to the sub-cascade formation criteria. Two well-known codes, Geant4 and TRIM, were used to simulate high-energy PKA-induced cascades in B_4C and then produce the secondary knock-on atom (SKA) energy spectrum. It has been found that both high-energy primary knock-on B and C atoms move a long range in the boron carbide. These atoms produce sub-cascades at the tip of trajectory. The energy received by most of the SKAs is <10 keV, which can be used as input to reduce the clock time spent for MD simulation.
Efficiency simulation of long neutron counter
International Nuclear Information System (INIS)
Hu Qingyuan; Li Bojun; Zhang De; Guo Hongsheng; Wang Dong; Yang Gaozhao; Si Fenni; Liu Jian
2008-01-01
In order to achieve the high efficiency and uniform sensitivity for neutrons with widely different energies, the efficiency of long boron trifluoride proportional counter imbedded in polyethylene moderator was simulated by MCNP code. The result shows that detective efficiency would increase with increasing moderator radius and response curve at higher energy would be ameliorated through adjusting the thickness of front moderator. Also we calculated the relative efficiencies for different energy of a detector whose efficiencies were calibrated on an accelerator. The simulated efficiency for D-D neutrons (2.4 MeV) is 75% of the efficiency for D-T neutrons (14.1 MeV), which is approximately agreed with experimental data, 61%. The validity of the simulated model was proved by the consistent results between calculation and experiment data. (authors)
Neutron imaging integrated circuit and method for detecting neutrons
Nagarkar, Vivek V.; More, Mitali J.
2017-12-05
The present disclosure provides a neutron imaging detector and a method for detecting neutrons. In one example, a method includes providing a neutron imaging detector including plurality of memory cells and a conversion layer on the memory cells, setting one or more of the memory cells to a first charge state, positioning the neutron imaging detector in a neutron environment for a predetermined time period, and reading a state change at one of the memory cells, and measuring a charge state change at one of the plurality of memory cells from the first charge state to a second charge state less than the first charge state, where the charge state change indicates detection of neutrons at said one of the memory cells.
The neutron instrument simulation package, NISP
International Nuclear Information System (INIS)
Seeger, P.A.; Daemen, L.L.
2004-01-01
The Neutron Instrument Simulation Package (NISP) performs complete source-to-detector simulations of neutron instruments, including neutrons that do not follow the expected path. The original user interface (MC( ) Web) is a web-based application, http://strider.lansce.lanl.gov/NISP/Welcome.html. This report describes in detail the newer standalone Windows version, NISP( ) Win. Instruments are assembled from menu-selected elements, including neutron sources, collimation and transport elements, samples, analyzers, and detectors. Magnetic field regions may also be specified for the propagation of polarized neutrons including spin precession. Either interface writes a geometry file that is used as input to the Monte Carlo engine (MC( ) Run) in the user's computer. Both the interface and the engine rely on a subroutine library, MCLIB. The package is completely open source. New features include capillary optics, temperature dependence of Al and Be, revised source files for ISIS, and visualization of neutron trajectories at run time. Also, a single-crystal sample type has been successfully imported from McStas (with more generalized geometry), demonstrating the capability of including algorithms from other sources, and NISP( ) Win may render the instrument in a virtual reality file. Results are shown for two instruments under development.
An improved fast neutron radiography quantitative measurement method
International Nuclear Information System (INIS)
Matsubayashi, Masahito; Hibiki, Takashi; Mishima, Kaichiro; Yoshii, Koji; Okamoto, Koji
2004-01-01
The validity of a fast neutron radiography quantification method, the Σ-scaling method, which was originally proposed for thermal neutron radiography was examined with Monte Carlo calculations and experiments conducted at the YAYOI fast neutron source reactor. Water and copper were selected as comparative samples for a thermal neutron radiography case and a dense object, respectively. Although different characteristics on effective macroscopic cross-sections were implied by the simulation, the Σ-scaled experimental results with the fission neutron spectrum cross-sections were well fitted to the measurements for both the water and copper samples. This indicates that the Σ-scaling method could be successfully adopted for quantitative measurements in fast neutron radiography
Methods for Probing Magnetic Films with Neutrons
Kozhevnikov, S. V.; Ott, F.; Radu, F.
2018-03-01
We review various methods in the investigation of magnetic films with neutrons, including those based on the effects of Larmor precession, Zeeman spatial splitting of the beam, neutron spin resonance, and polarized neutron channeling. The underlying principles, examples of the investigated systems, specific features, applications, and perspectives of these methods are discussed.
Specific Heat Capacity of Alloy 690 for Simulating Neutron Irradiation
International Nuclear Information System (INIS)
Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Joo, Young Sun; Ahn, Sang Bok; Park, Jin Seok; Lee, Won Jae; Ryu, Woo Seok
2011-01-01
The KAERI(Korea Atomic Energy Research Institute) is developing new type of nuclear reactor, so called 'SMART'(System Integrated Modular Advanced Reactor) which has many features of small power and system integrated modular type. Alloy 690 was selected as the candidate material for the heat exchanger tube of the steam generator of SMART. The SMART R and D is now facing the stage of engineering verification and approval of standard design to apply to DEMO reactors. Therefore, the material performance under the relevant environment is required to be evaluated. The important material performance issues are mechanical properties i.e. (fracture toughness, tensile and hardness) and thermal properties i.e. (thermal diffusivity, specific heat capacity and thermal conductivity) for which the engineering database is necessary to design a steam generator. However, the neutron post irradiation characteristics of the alloy 690 are barely known. As a result, PIE(Post Irradiation Examination) of thermal properties are planed and performed successfully. But specific heat capacity measurement is not performed because of not having proper test system for irradiated materials. Therefore in order to verify the effect of neutron irradiation for alloy 690, simulation method is adopted. In general, high energy neutron bombardment in material bring about lattice defects i.e. void, pore and dislocation. Dominant factor to impact to heat capacity is mainly dislocation in material. Therefore, simulation of neutron irradiation is devised by material rolling method in order to make artificial dislocation in alloy 690 as same effect of neutron irradiation. After preparing test specimens, heat capacity measurements are performed and results are compared with rolled materials and un-rolled materials to verify the effect of neutron irradiation simulation. Main interest of simulation is that heat capacity value is changed by neutron irradiation
International Nuclear Information System (INIS)
Behbahani, A.
1983-01-01
An analytical neutron sensor response model and methods for transient response measurements of neutron sensors (compensated ionization chamber), including possible in-situ techniques have been developed and evaluated to meet the provisions of Draft Standard ISA Sd67.06, IEEE 338-1977, and NRC Guide 1.118. One in-situ method requires the perturbation of the high voltage detector (sensor) power supply and measurement of the sensor response. The response to a perturbation of the power supply is related to the response of the sensor to a transient change in neutron flux. Random signal analysis is another in-situ technique to monitor the neutron sensor response. In this method the power spectrum of the inherent fluctuations from the neutron sensor output (current in CIC) are measured and evaluated. Transient response techniques (including in-situ methods) are experimentally and analytically evaluated to identify the mechanisms which may cause degradation in the response of the neutron sensors. The objective of the experimental evaluation was to correlate the measured response time using transient radiation flux changes and power supply perturbation. The objective of the analytical model of the different sensor response was to predict response time and degradation mechanisms
Monte Carlo simulation of neutron scattering instruments
International Nuclear Information System (INIS)
Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.
1998-01-01
A code package consisting of the Monte Carlo Library MCLIB, the executing code MC RUN, the web application MC Web, and various ancillary codes is proposed as an open standard for simulation of neutron scattering instruments. The architecture of the package includes structures to define surfaces, regions, and optical elements contained in regions. A particle is defined by its vector position and velocity, its time of flight, its mass and charge, and a polarization vector. The MC RUN code handles neutron transport and bookkeeping, while the action on the neutron within any region is computed using algorithms that may be deterministic, probabilistic, or a combination. Complete versatility is possible because the existing library may be supplemented by any procedures a user is able to code. Some examples are shown
Some neutron measurements with simulated ING targets
International Nuclear Information System (INIS)
Walker, J.
1966-01-01
Thermal neutron fluxes in the vicinity of a simulated Intense Neutron Generator target have been measured using Mn and Au foils, and a small BF 3 detector. The target was a Pb cylinder either 4-inch or 8-inch in diameter with a 1.2 g Ra-Be neutron source at its centre. This was centrally mounted in a 5' diam. x 5' high tank which was filled with either H 2 O or D 2 O moderator. Various gaps and absorbing annuli were placed around the target, and air-filled aluminum 'beam tubes' were mounted radially or tangentially from the target to simulate typical ING conditions. The measured thermal neutron fluxes were less than calculated at all radii. The single-age computation clearly gives large errors at large radii, but the multi-energy approach seems to give a useful indication of the thermal flux distribution in spite of the extreme simplicity of the model. The fall in measured fluxes at small radii in both D 2 O and H 2 O is most likely caused by absorption in the target material which is not allowed for in the computational model. (author)
Some neutron measurements with simulated ING targets
Energy Technology Data Exchange (ETDEWEB)
Walker, J
1966-07-01
Thermal neutron fluxes in the vicinity of a simulated Intense Neutron Generator target have been measured using Mn and Au foils, and a small BF{sub 3} detector. The target was a Pb cylinder either 4-inch or 8-inch in diameter with a 1.2 g Ra-Be neutron source at its centre. This was centrally mounted in a 5' diam. x 5' high tank which was filled with either H{sub 2}O or D{sub 2}O moderator. Various gaps and absorbing annuli were placed around the target, and air-filled aluminum 'beam tubes' were mounted radially or tangentially from the target to simulate typical ING conditions. The measured thermal neutron fluxes were less than calculated at all radii. The single-age computation clearly gives large errors at large radii, but the multi-energy approach seems to give a useful indication of the thermal flux distribution in spite of the extreme simplicity of the model. The fall in measured fluxes at small radii in both D{sub 2}O and H{sub 2}O is most likely caused by absorption in the target material which is not allowed for in the computational model. (author)
International Nuclear Information System (INIS)
Zuo Guangxia; He Bin; Xu Peng; Qiu Xiaolin; Ma Wenyan; Li Sufen
2012-01-01
In neutrons detection, it is important to discriminate the neutron signals from the gamma-ray background. In this article, simulation of neutrons and gamma pulse signals is developed based on the LabVIEW platform. Two digital algorithms of the charge comparison method and the pulse duration time method are realized using 10000 simulation signals. Experimental results show that neutron and gamma pulse signals can be discriminated by the two methods, and the pulse duration time method is better than the charge comparison method. (authors)
Dosimetry methods in boron neutron capture therapy
Energy Technology Data Exchange (ETDEWEB)
Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D. [Universita degli Studi di Milano, Department of Physics, Via Festa del Patrono 7, 20122 Milano (Italy); Agosteo, S.; Barcaglioni, L. [Istituto Nazionale di Fisica Nucleare, Milano (Italy); Campi, F.; Garlati, L. [Politecnico di Milano, Energy Department, Piazza Leonardo Da Vinci 32, 20133 Milano (Italy); De Errico, F. [Universita degli Studi di Pisa, Department of Civil and Industrial Engineering, Lungamo Pacinotti 43, 56126 Pisa (Italy); Borroni, M.; Carrara, M. [Fondazione IRCCS Istituto Nazionale Tumori, Medical Physics Unit, Via Venezian 1, 20133 Milano (Italy); Burian, J.; Klupak, V.; Viererbl, L.; Marek, M. [Research Centre Rez, Department of Neutron Physics, 250-68 Husinec-Rez (Czech Republic)
2014-08-15
Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)
Dosimetry methods in boron neutron capture therapy
International Nuclear Information System (INIS)
Gambarini, G.; Artuso, E.; Felisi, M.; Regazzoni, V.; Giove, D.; Agosteo, S.; Barcaglioni, L.; Campi, F.; Garlati, L.; De Errico, F.; Borroni, M.; Carrara, M.; Burian, J.; Klupak, V.; Viererbl, L.; Marek, M.
2014-08-01
Dosimetry studies have been carried out at thermal and epithermal columns of Lvr-15 research reactor for investigating the spatial distribution of gamma dose, fast neutron dose and thermal neutron fluence. Two different dosimetry methods, both based on solid state detectors, have been studied and applied and the accuracy and consistency of the results have been inspected. One method is based on Fricke gel dosimeters that are dilute water solutions and have good tissue equivalence for neutrons and also for all the secondary radiations produced by neutron interactions in tissue or water phantoms. Fricke gel dosimeters give the possibility of separating the various dose contributions, i.e. the gamma dose, the fast neutron dose and the dose due to charged particles generated during thermal neutron reactions by isotopes having high cross section, like 10-B. From this last dose, thermal neutron fluence can be obtained by means of the kerma factor. The second method is based on thermoluminescence dosimeters. In particular, the developed method draw advantage from the different heights of the peaks of the glow curve of such phosphors when irradiated with photons or with thermal neutrons. The results show that satisfactory results can be obtained with simple methods, in spite of the complexity of the subject. However, the more suitable dosimeters and principally their utilization and analysis modalities are different for the various neutron beams, mainly depending on the relative intensities of the three components of the neutron field, in particular are different for thermal and epithermal columns. (Author)
Beam neutron energy optimization for boron neutron capture therapy using monte Carlo method
International Nuclear Information System (INIS)
Pazirandeh, A.; Shekarian, E.
2006-01-01
In last two decades the optimal neutron energy for the treatment of deep seated tumors in boron neutron capture therapy in view of neutron physics and chemical compounds of boron carrier has been under thorough study. Although neutron absorption cross section of boron is high (3836b), the treatment of deep seated tumors such as glioblastoma multiform requires beam of neutrons of higher energy that can penetrate deeply into the brain and thermalized in the proximity of the tumor. Dosage from recoil proton associated with fast neutrons however poses some constraints on maximum neutron energy that can be used in the treatment. For this reason neutrons in the epithermal energy range of 10eV-10keV are generally to be the most appropriate. The simulation carried out by Monte Carlo methods using MCBNCT and MCNP4C codes along with the cross section library in 290 groups extracted from ENDF/B6 main library. The ptimal neutron energy for deep seated tumors depends on the sue and depth of tumor. Our estimated optimized energy for the tumor of 5cm wide and 1-2cm thick stands at 5cm depth is in the range of 3-5keV
A method for neutron dosimetry in ultrahigh flux environments
International Nuclear Information System (INIS)
Ougouag, A.M.; Wemple, C.A.; Rogers, J.W.
1996-01-01
A method for neutron dosimetry in ultrahigh flux environments is developed, and devices embodying it are proposed and simulated using a Monte Carlo code. The new approach no longer assumes a linear relationship between the fluence and the activity of the nuclides formed by irradiation. It accounts for depletion of the original ''foil'' material and for decay and depletion of the formed nuclides. In facilities where very high fluences are possible, the fluences inferred by activity measurements may be ambiguous. A method for resolving these ambiguities is also proposed and simulated. The new method and proposed devices should make possible the use of materials not traditionally considered desirable for neutron activation dosimetry
The MCLIB library: Monte Carlo simulation of neutron scattering instruments
Energy Technology Data Exchange (ETDEWEB)
Seeger, P.A.
1995-09-01
Monte Carlo is a method to integrate over a large number of variables. Random numbers are used to select a value for each variable, and the integrand is evaluated. The process is repeated a large number of times and the resulting values are averaged. For a neutron transport problem, first select a neutron from the source distribution, and project it through the instrument using either deterministic or probabilistic algorithms to describe its interaction whenever it hits something, and then (if it hits the detector) tally it in a histogram representing where and when it was detected. This is intended to simulate the process of running an actual experiment (but it is much slower). This report describes the philosophy and structure of MCLIB, a Fortran library of Monte Carlo subroutines which has been developed for design of neutron scattering instruments. A pair of programs (LQDGEOM and MC{_}RUN) which use the library are shown as an example.
The MCLIB library: Monte Carlo simulation of neutron scattering instruments
International Nuclear Information System (INIS)
Seeger, P.A.
1995-01-01
Monte Carlo is a method to integrate over a large number of variables. Random numbers are used to select a value for each variable, and the integrand is evaluated. The process is repeated a large number of times and the resulting values are averaged. For a neutron transport problem, first select a neutron from the source distribution, and project it through the instrument using either deterministic or probabilistic algorithms to describe its interaction whenever it hits something, and then (if it hits the detector) tally it in a histogram representing where and when it was detected. This is intended to simulate the process of running an actual experiment (but it is much slower). This report describes the philosophy and structure of MCLIB, a Fortran library of Monte Carlo subroutines which has been developed for design of neutron scattering instruments. A pair of programs (LQDGEOM and MC RUN) which use the library are shown as an example
Monte Carlo method in neutron activation analysis
International Nuclear Information System (INIS)
Majerle, M.; Krasa, A.; Svoboda, O.; Wagner, V.; Adam, J.; Peetermans, S.; Slama, O.; Stegajlov, V.I.; Tsupko-Sitnikov, V.M.
2009-01-01
Neutron activation detectors are a useful technique for the neutron flux measurements in spallation experiments. The study of the usefulness and the accuracy of this method at similar experiments was performed with the help of Monte Carlo codes MCNPX and FLUKA
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.
Thermal neutron shield and method of manufacture
Brindza, Paul Daniel; Metzger, Bert Clayton
2013-05-28
A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.
A prestorage method to measure neutron transmission of ultracold neutron guides
International Nuclear Information System (INIS)
Blau, B.; Daum, M.; Fertl, M.; Geltenbort, P.; Göltl, L.; Henneck, R.; Kirch, K.; Knecht, A.; Lauss, B.; Schmidt-Wellenburg, P.; Zsigmond, G.
2016-01-01
There are worldwide efforts to search for physics beyond the Standard Model of particle physics. Precision experiments using ultracold neutrons (UCN) require very high intensities of UCN. Efficient transport of UCN from the production volume to the experiment is therefore of great importance. We have developed a method using prestored UCN in order to quantify UCN transmission in tubular guides. This method simulates the final installation at the Paul Scherrer Institute's UCN source where neutrons are stored in an intermediate storage vessel serving three experimental ports. This method allowed us to qualify UCN guides for their intended use and compare their properties.
Novel applications of fast neutron interrogation methods
International Nuclear Information System (INIS)
Gozani, Tsahi
1994-01-01
The development of non-intrusive inspection methods for contraband consisting primarily of carbon, nitrogen, oxygen, and hydrogen requires the use of fast neutrons. While most elements can be sufficiently well detected by the thermal neutron capture process, some important ones, e.g., carbon and in particular oxygen, cannot be detected by this process. Fortunately, fast neutrons, with energies above the threshold for inelastic scattering, stimulate relatively strong and specific gamma ray lines from these elements. The main lines are: 6.13 for O, 4.43 for C, and 5.11, 2.31 and 1.64 MeV for N. Accelerator-generated neutrons in the energy range of 7 to 15 MeV are being considered as interrogating radiations in a variety of non-intrusive inspection systems for contraband, from explosives to drugs and from coal to smuggled, dutiable goods. In some applications, mostly for inspection of small items such as luggage, the decision process involves a rudimentary imaging, akin to emission tomography, to obtain the localized concentration of various elements. This technique is called FNA - Fast Neutron Analysis. While this approach offers improvements over the TNA (Thermal Neutron Analysis), it is not applicable to large objects such as shipping containers and trucks. For these challenging applications, a collimated beam of neutrons is rastered along the height of the moving object. In addition, the neutrons are generated in very narrow nanosecond pulses. The point of their interaction inside the object is determined by the time of flight (TOF) method, that is measuring the time elapsed from the neutron generation to the time of detection of the stimulated gamma rays. This technique, called PFNA (Pulsed Fast Neutron Analysis), thus directly provides the elemental, and by inference, the chemical composition of the material at every volume element (voxel) of the object. The various neutron-based techniques are briefly described below. ((orig.))
Simulation of Neutron Backscattering applied to organic material detection
International Nuclear Information System (INIS)
Forero, N. C.; Cruz, A. H.; Cristancho, F.
2007-01-01
The Neutron Backscattering technique is tested when performing the task of localizing hydrogenated explosives hidden in soil. Detector system, landmine, soil and neutron source are simulated with Geant4 in order to obtain the number of neutrons detected when several parameters like mine composition, relative position mine-source and soil moisture are varied
Direct Discrete Method for Neutronic Calculations
International Nuclear Information System (INIS)
Vosoughi, Naser; Akbar Salehi, Ali; Shahriari, Majid
2002-01-01
The objective of this paper is to introduce a new direct method for neutronic calculations. This method which is named Direct Discrete Method, is simpler than the neutron Transport equation and also more compatible with physical meaning of problems. This method is based on physic of problem and with meshing of the desired geometry, writing the balance equation for each mesh intervals and with notice to the conjunction between these mesh intervals, produce the final discrete equations series without production of neutron transport differential equation and mandatory passing from differential equation bridge. We have produced neutron discrete equations for a cylindrical shape with two boundary conditions in one group energy. The correction of the results from this method are tested with MCNP-4B code execution. (authors)
Neutronics methods for thermal radiative transfer
International Nuclear Information System (INIS)
Larsen, E.W.
1988-01-01
The equations of thermal radiative transfer are time discretized in a semi-implicit manner, yielding a linear transport problem for each time step. The governing equation in this problem has the form of a neutron transport equation with fission but no scattering. Numerical methods are described, whose origins lie in neutron transport, and that have been successfully adapted to this new problem. Acceleration methods that have been developed specifically for the radiative transfer problem, but may have generalizations applicable in neutronics problems, are also discussed
Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination.
Liu, B; Xu, J; Liu, T; Ouyang, X
2012-10-01
To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a (252)Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D-D neutron generator can create neutrons at up to 10(13) n s(-1) with current technology. All these enable an effective and low-cost method of killing anthrax spores. There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g (252)Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D-D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D-D neutron generator output >10(13) n s(-1) should be attainable in the near future. This indicates that we could use a D-D neutron generator to sterilise anthrax contamination within several seconds.
Energy Technology Data Exchange (ETDEWEB)
Mariani, A.; Passard, C.; Jallu, F. E-mail: fanny.jallu@cea.fr; Toubon, H
2003-11-01
The design of a specific nuclear assay system for a dedicated application begins with a phase of development, which relies on information from the literature or on knowledge resulting from experience, and on specific experimental verifications. The latter ones may require experimental devices which can be restricting in terms of deadline, cost and safety. One way generally chosen to bypass these difficulties is to use simulation codes to study particular aspects. This paper deals with the potentialities offered by the simulation in the case of a passive-active neutron (PAN) assay system for alpha low level waste characterization; this system has been carried out at the Nuclear Measurements Development Laboratory of the French Atomic Energy Commission. Due to the high number of parameters to be taken into account for its development, this is a particularly sophisticated example. Since the PAN assay system, called PROMETHEE (prompt epithermal and thermal interrogation experiment), must have a detection efficiency of more than 20% and preserve a high level of modularity for various applications, an improved version has been studied using the MCNP4 (Monte Carlo N-Particle) transport code. Parameters such as the dimensions of the assay system, of the cavity and of the detection blocks, and the thicknesses of the nuclear materials of neutronic interest have been optimised. Therefore, the number of necessary experiments was reduced.
International Nuclear Information System (INIS)
Mariani, A.; Passard, C.; Jallu, F.; Toubon, H.
2003-01-01
The design of a specific nuclear assay system for a dedicated application begins with a phase of development, which relies on information from the literature or on knowledge resulting from experience, and on specific experimental verifications. The latter ones may require experimental devices which can be restricting in terms of deadline, cost and safety. One way generally chosen to bypass these difficulties is to use simulation codes to study particular aspects. This paper deals with the potentialities offered by the simulation in the case of a passive-active neutron (PAN) assay system for alpha low level waste characterization; this system has been carried out at the Nuclear Measurements Development Laboratory of the French Atomic Energy Commission. Due to the high number of parameters to be taken into account for its development, this is a particularly sophisticated example. Since the PAN assay system, called PROMETHEE (prompt epithermal and thermal interrogation experiment), must have a detection efficiency of more than 20% and preserve a high level of modularity for various applications, an improved version has been studied using the MCNP4 (Monte Carlo N-Particle) transport code. Parameters such as the dimensions of the assay system, of the cavity and of the detection blocks, and the thicknesses of the nuclear materials of neutronic interest have been optimised. Therefore, the number of necessary experiments was reduced
International Nuclear Information System (INIS)
Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei
1999-01-01
In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments
Simulation of neutron transport equation using parallel Monte Carlo for deep penetration problems
International Nuclear Information System (INIS)
Bekar, K. K.; Tombakoglu, M.; Soekmen, C. N.
2001-01-01
Neutron transport equation is simulated using parallel Monte Carlo method for deep penetration neutron transport problem. Monte Carlo simulation is parallelized by using three different techniques; direct parallelization, domain decomposition and domain decomposition with load balancing, which are used with PVM (Parallel Virtual Machine) software on LAN (Local Area Network). The results of parallel simulation are given for various model problems. The performances of the parallelization techniques are compared with each other. Moreover, the effects of variance reduction techniques on parallelization are discussed
Particle-transport simulation with the Monte Carlo method
International Nuclear Information System (INIS)
Carter, L.L.; Cashwell, E.D.
1975-01-01
Attention is focused on the application of the Monte Carlo method to particle transport problems, with emphasis on neutron and photon transport. Topics covered include sampling methods, mathematical prescriptions for simulating particle transport, mechanics of simulating particle transport, neutron transport, and photon transport. A literature survey of 204 references is included. (GMT)
A primary simulation for residual stress neutron diffractometer
International Nuclear Information System (INIS)
Wang Shuying; Liu Lijuan; Sun Liangwei
2012-01-01
At present, neutron diffraction method is the unique and nondestructive method that can directly measure the residual stress distribution in deep materials and engineering components. It has an important application in engineering. A simulation of the flux at the position of the sample table was reported, and the resolution of the residual stress instrument was computed at the same time. The effect of the parameters of the second collimator on the flux at the sample position and the resolution of the instrument have been analyzed. The result indicated that the second collimator empress much on the neutron flux and the instrument resolution is well when the sample's diffraction angle is less than 120°. (authors)
Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.
2018-03-01
In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.
International Nuclear Information System (INIS)
Zhang Feng; Hou Shuang; Jin Xiuyun
2010-01-01
The process of neutron interaction induced by D-T pulsed neutron generator and 241 Am-Be source was simulated by using Monte Carlo method. It is concluded that the thermal neutron count descend exponentially as the spacing increasing. The smaller porosity was, the smaller the differences between the two sources were. When the porosity reached 40%, the ratio of thermal neutron count generated by D-T pulsed neutron source was much larger than that generated by 241 Am-Be neutron source, and its distribution range was wider. The near spacing selected was 20-30 cm, and that of far spacing was about 60-70 cm. The detection depth by using D-T pulsed neutron source was almost unchanged under condition of the same sapcing, and the sensitivity of measurement to the formation porosity decreases. The results showed that it can not only guarantee the statistic of count, but also improve detection sensitivity and depth at the same time of increasing spacing. Therefore, 241 Am-Be neutron source can be replaced by D-T neutron tube in LWD tool. (authors)
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.)
Optimization on Measurement Method for Neutron Moisture Meter
International Nuclear Information System (INIS)
Gong Yalin; Wu Zhiqiang; Li Yanfeng; Wang Wei; Song Qingfeng; Liu Hui; Wei Xiaoyun; Zhao Zhonghua
2010-01-01
When the water in the measured material is nonuniformity, the measured results of the neutron moisture meter in the field may have errors, so the measured errors of the moisture meter associated with the water nonuniformity in material were simulated by Monte Carlo method. A new measurement method of moisture meter named 'transmission plus scatter' was put forward. The experiment results show that the new measurement method can reduce the error even if the water in the material is nonuniformity. (authors)
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.
Large subcriticality measurement by pulsed neutron method
International Nuclear Information System (INIS)
Yamane, Y.; Yoshida, A.; Nishina, K.; Kobayashi, K.; Kanda, K.
1985-01-01
To establish the method determining large subcriticalities in the field of nuclear criticality safety, the authors performed pulsed neutron experiments using the Kyoto University Critical Assembly (KUCA) at Research Reactor Institute, Kyoto University and the Cockcroft-Walton type accelerator attached to the assembly. The area-ratio method proposed by Sjoestrand was employed to evaluate subcriticalities from neutron decay curves measured. This method has the shortcomings that the neutron component due to a decay of delayed neutrons remarkably decreases as the subcriticality of an objective increases. To overcome the shortcoming, the authors increased the frequency of pulsed neutron generation. The integral-version of the area-ratio method proposed by Kosaly and Fisher was employed in addition in order to remove a contamination of spatial higher modes from the decay curve. The latter becomes significant as subcriticality increases. The largest subcriticality determined in the present experiments was 125.4 dollars, which was equal to 0.5111 in a multiplication factor. The calculational values evaluated by the computer code KENO-IV with 137 energy groups based on the Monte Carlo method agreed well with those experimental values
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)
Neutron optics using transverse field neutron spin echo method
International Nuclear Information System (INIS)
Achiwa, Norio; Hino, Masahiro; Yamauchi, Yoshihiro; Takakura, Hiroyuki; Tasaki, Seiji; Akiyoshi, Tsunekazu; Ebisawa, Toru.
1993-01-01
A neutron spin echo (NSE) spectrometer with perpendicular magnetic field to the neutron scattering plane, using an iron yoke type electro-magnet has been developed. A combination of cold neutron guider, supermirror neutron polarizer of double reflection type and supermirror neutron analyser was adopted for the spectrometer. The first application of the NSE spectrometer to neutron optics by passing Larmor precessing neutrons through gas, solid and liquid materials of several different lengths which are inserted in one of the precession field have been examined. Preliminary NSE spectra of this sample geometry are discussed. (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....
Energy Technology Data Exchange (ETDEWEB)
Tairan, Liang [School of Physics and Electronic Information Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Zhiduo, Li [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Wen, Yin, E-mail: wenyin@aphy.iphy.ac.cn [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Fei, Shen [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Quanzhi, Yu [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Tianjiao, Liang [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China)
2017-07-11
The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm{sup 2}/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.
Cosmic-ray neutron simulations and measurements in Taiwan
International Nuclear Information System (INIS)
Chen, Wei-Lin; Jiang, Shiang-Huei; Sheu, Rong-Jiun
2014-01-01
This study used simulations of galactic cosmic ray in the atmosphere to investigate the neutron background environment in Taiwan, emphasising its altitude dependence and spectrum variation near interfaces. The calculated results were analysed and compared with two measurements. The first measurement was a mobile neutron survey from sea level up to 3275 m in altitude conducted using a car-mounted high-sensitivity neutron detector. The second was a previous measured result focusing on the changes in neutron spectra near air/ground and air/water interfaces. The attenuation length of cosmic-ray neutrons in the lower atmosphere was estimated to be 163 g cm -2 in Taiwan. Cosmic-ray neutron spectra vary with altitude and especially near interfaces. The determined spectra near the air/ground and air/water interfaces agree well with measurements for neutrons below 10 MeV. However, the high-energy portion of spectra was observed to be much higher than our previous estimation. Because high-energy neutrons contribute substantially to a dose evaluation, revising the annual sea-level effective dose from cosmic-ray neutrons at ground level in Taiwan to 35 μSv, which corresponds to a neutron flux of 5.30 x 10 -3 n cm -2 s -1 , was suggested. The cosmic-ray neutron background in Taiwan was studied using the FLUKA simulations and field measurements. A new measurement was performed using a car-mounted high-efficiency neutron detector, re-coding real-time neutron counting rates from sea level up to 3275 m. The attenuation of cosmic-ray neutrons in the lower atmosphere exhibited an effective attenuation length of 163 g cm -2 . The calculated neutron counting rates over predicted the measurements by ∼32 %, which leaded to a correction factor for the FLUKA-calculated cosmic-ray neutrons in the lower atmosphere in Taiwan. In addition, a previous measurement regarding neutron spectrum variation near the air/ground and air/water interfaces was re-evaluated. The results showed that the
International Nuclear Information System (INIS)
Artem’ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.
2016-01-01
We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10 -7 –10 -3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder
Artem'ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.
2016-01-01
We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2-5 nm and for neutron energies 3 × 10-7-10-3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.
Energy Technology Data Exchange (ETDEWEB)
Artem’ev, V. A., E-mail: niitm@inbox.ru [Research Institute of Materials Technology (Russian Federation); Nezvanov, A. Yu. [Moscow State Industrial University (Russian Federation); Nesvizhevsky, V. V. [Institut Max von Laue—Paul Langevin (France)
2016-01-15
We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10{sup -7}–10{sup -3} eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.
Accurate simulation of neutrons in less than one minute Pt. 1: Acceptance diagram shading
International Nuclear Information System (INIS)
Bentley, Phillip M.; Andersen, Ken H.
2009-01-01
We describe an algorithm for modelling neutron guides based on an extension of acceptance diagram methods: neutron acceptance diagram shading (NADS), using a computer graphical approach. It has a similar black-box, modular strategy to Monte-Carlo simulations, but often outperforms Monte-Carlo by orders of magnitude in terms of calculation speed without sacrificing significant detail. NADS is expected to increase the productivity in the design stage of neutron instrument development, and facilitate the widespread application of advanced, modern artificial-intelligence-based optimisation algorithms for the design of neutron instrumentation.
Neutron activation analysis: principle and methods
International Nuclear Information System (INIS)
Reddy, A.V.R.; Acharya, R.
2006-01-01
Neutron activation analysis (NAA) is a powerful isotope specific nuclear analytical technique for simultaneous determination of elemental composition of major, minor and trace elements in diverse matrices. The technique is capable of yielding high analytical sensitivity and low detection limits (ppm to ppb). Due to high penetration power of neutrons and gamma rays, NAA experiences negligible matrix effects in the samples of different origins. Depending on the sample matrix and element of interest NAA technique is used non-destructively, known as instrumental neutron activation analysis (INAA), or through chemical NAA methods. The present article describes principle of NAA, different methods and gives a overview some applications in the fields like environment, biology, geology, material sciences, nuclear technology and forensic sciences. (author)
Simulations for the neutron detector TETRA with MCNP
International Nuclear Information System (INIS)
Testov, D.; Kuznetsova, E.; Wilson, Jh.
2013-01-01
To study the nuclear structure of β-delayed neutron precursors at ALTO ISOL-facility at IPN (Orsay), the high efficiency 4π neutron detector TETRA with 3 He filled counters built at JINR (Dubna) was modified. The MCNP simulations to optimize the future configuration were necessary. The details of the calculations and the major results obtained are discussed
Genetic Algorithms: A New Method for Neutron Beam Spectral Characterization
International Nuclear Information System (INIS)
David W. Freeman
2000-01-01
A revolutionary new concept for solving the neutron spectrum unfolding problem using genetic algorithms (GAs) has recently been introduced. GAs are part of a new field of evolutionary solution techniques that mimic living systems with computer-simulated chromosome solutions that mate, mutate, and evolve to create improved solutions. The original motivation for the research was to improve spectral characterization of neutron beams associated with boron neutron capture therapy (BNCT). The GA unfolding technique has been successfully applied to problems with moderate energy resolution (up to 47 energy groups). Initial research indicates that the GA unfolding technique may well be superior to popular unfolding methods in common use. Research now under way at Kansas State University is focused on optimizing the unfolding algorithm and expanding its energy resolution to unfold detailed beam spectra based on multiple foil measurements. Indications are that the final code will significantly outperform current, state-of-the-art codes in use by the scientific community
Simulation of Resistive Plate Chamber sensitivity to neutrons
Energy Technology Data Exchange (ETDEWEB)
Altieri, S. E-mail: saverio.altieri@pv.infn.it; Belli, G.; Bruno, G.; Merlo, M.; Ratti, S.P.; Riccardi, C.; Torre, P.; Vitulo, P.; Abbrescia, M.; Colaleo, A.; Iaselli, G.; Loddo, F.; Maggi, M.; Marangelli, B.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Romano, F
2001-04-01
The Resistive Plate Chambers (RPCs) sensitivity to neutrons has been simulated using GEANT code with MICAP and FLUKA interfaces. The calculations have been performed as a function of the neutrons energy in the range 0.02 eV-1 GeV. To evaluate the response of the detector in the LHC background environment, the neutron energy spectrum expected in the CMS muon barrel has been taken into account; a hit rate due to neutrons of about 0.6 Hz cm{sup -2} has been estimated for a 250x250 cm{sup 2} RPC in the RB1 station.
The delayed neutron method of uranium analysis
International Nuclear Information System (INIS)
Wall, T.
1989-01-01
The technique of delayed neutron analysis (DNA) is discussed. The DNA rig installed on the MOATA reactor, the assay standards and the types of samples which have been assayed are described. Of the total sample throughput of about 55,000 units since the uranium analysis service began, some 78% has been concerned with analysis of uranium ore samples derived from mining and exploration. Delayed neutron analysis provides a high sensitivity, low cost uranium analysis method for both uranium exploration and other applications. It is particularly suitable for analysis of large batch samples and for non-destructive analysis over a wide range of matrices. 8 refs., 4 figs., 3 tabs
Method for manufacture of neutron absorbing articles
International Nuclear Information System (INIS)
Owens, D.
1980-01-01
A one-step curing method for the manufacture of a neutron absorbing article which comprises irreversibly curing, in desired article form, a form-retaining mixture of boron carbide particles, curable phenolic resin in solid state and in particula te form and a minor proportion of a liquid medium, which boils at a temperature below 200*c., at an elevated temperature so as to obtain bonding of the irreversibly cured phenolic polymer resulting to the boron carbide particles and production of the neutron absorbing article in desired form
Evaluation methods for neutron cross section standards
International Nuclear Information System (INIS)
Bhat, M.R.
1980-01-01
Methods used to evaluate the neutron cross section standards are reviewed and their relative merits, assessed. These include phase-shift analysis, R-matrix fit, and a number of other methods by Poenitz, Bhat, Kon'shin and the Bayesian or generalized least-squares procedures. The problems involved in adopting these methods for future cross section standards evaluations are considered, and the prospects for their use, discussed. 115 references, 5 figures, 3 tables
Developing an interface between MCNP and McStas for simulation of neutron moderators
DEFF Research Database (Denmark)
Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik
2012-01-01
Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....
Optimization of the SNS magnetism reflectometer neutron-guide optics using Monte Carlo simulations
Klose, F
2002-01-01
The magnetism reflectometer at the spallation neutron source SNS will employ advanced neutron optics to achieve high data rate, improved resolution, and extended dynamic range. Optical components utilized will include a multi-channel polygonal curved bender and a tapered neutron-focusing guide section. The results of a neutron beam interacting with these devices are rather complex. Additional complexity arises due to the spectral/time-emission profile of the moderator and non-perfect neutron optical coatings. While analytic formulae for the individual components provide some design guidelines, a realistic performance assessment of the whole instrument can only be achieved by advanced simulation methods. In this contribution, we present guide optics optimizations for the magnetism reflectometer using Monte Carlo simulations. We compare different instrument configurations and calculate the resulting data rates. (orig.)
SIMULATED 8 MeV NEUTRON RESPONSE FUNCTIONS OF A THIN SILICON NEUTRON SENSOR.
Takada, Masashi; Matsumoto, Tetsuro; Masuda, Akihiko; Nunomiya, Tomoya; Aoyama, Kei; Nakamura, Takashi
2017-12-22
Neutron response functions of a thin silicon neutron sensor are simulated using PHITS2 and MCNP6 codes for an 8 MeV neutron beam at angles of incidence of 0°, 30° and 60°. The contributions of alpha particles created from the 28Si(n,α)25Mg reaction and the silicon nuclei scattered elastically by neutrons in the silicon sensor have not been well reproduced using the MCNP6 code. The 8 MeV neutron response functions simulated using the PHITS2 code with an accurate event generator mode are in good agreement with experimental results and include the contributions of the alpha particles and silicon nuclei. © The Author(s) 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.
Simulation for sodium-24 production using cyclic neutron activation analysis
International Nuclear Information System (INIS)
Ahamed, O. M. H.
2012-04-01
The cyclic neutron activation analysis is a method for elemental analysis which is preferred to use short-lived radio-nuclides. In recent years this method became a new application for radioisotope production especially in low power research reactors. In this study instrumental cyclic neutron activation analysis was used for 2 4N a production using 2 3N a (n, γ) 2 4N a reaction. The simplified westcott convention method is used neutron activation analysis in a research reactor. The method takes into account all corrections that can affect that yield created. In this work a model was devolving for calculations through this simplified Westcott convention method by using C++ program and selecting good parameters that can produce the expected activity. The simulation is used for the theoretical yield calculated has been validated by data from the 1 77L u production used for theoretical yield which it gives the approached result had been obtained by FORTRAN 90 from literature (8). The results were achieved for expected activity at full power (1 x 10 1 2n cm -2 s - 1 ) and half power (5 x 10 11 cm -2 s -1 ) for research reactor MNSR. The activity at full power was equal to about twice the activity at half power ( 49±7, 24.9 ± MBq/g), respectively. The irradiation parameters selected were irradiation time 4 min and decay time 12 min. Sample weight was 50 mg at 12 numbers of cycles, when the K-factor was equal to 1.74. This work is considered as first step for production of 2 4N a which can use such parameters experimentally. It is then possible to compare the expected activity with measured activity. (Author)
Measurement and simulation of thermal neutron flux distribution in the RTP core
Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.
2018-01-01
The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.
The Application of Neutron Transport Green's Functions to Threat Scenario Simulation
Thoreson, Gregory G.; Schneider, Erich A.; Armstrong, Hirotatsu; van der Hoeven, Christopher A.
2015-02-01
Radiation detectors provide deterrence and defense against nuclear smuggling attempts by scanning vehicles, ships, and pedestrians for radioactive material. Understanding detector performance is crucial to developing novel technologies, architectures, and alarm algorithms. Detection can be modeled through radiation transport simulations; however, modeling a spanning set of threat scenarios over the full transport phase-space is computationally challenging. Previous research has demonstrated Green's functions can simulate photon detector signals by decomposing the scenario space into independently simulated submodels. This paper presents decomposition methods for neutron and time-dependent transport. As a result, neutron detector signals produced from full forward transport simulations can be efficiently reconstructed by sequential application of submodel response functions.
Examination of zeolites by neutron reflection method
International Nuclear Information System (INIS)
Szegedi, S.; Varadi, M.; Boedy, Z.T.; Vas, L.
1991-01-01
Neutron reflection method has been used for the determination of zeolite content in minerals. The basis of this measurement is to observe the large difference between the water content of zeolite and that of other mineralic parts of the sample. The method suggested can be used in a zeolite mine for measuring the zeolite content continuously and controlling the quality of the end products. (author) 5 refs.; 3 figs.; 3 tabs
Recombination methods for boron neutron capture therapy dosimetry
International Nuclear Information System (INIS)
Golnik, N.; Tulik, P.; Zielczynski, M.
2003-01-01
The radiation effects of boron neutron capture therapy (BNCT) are associated with four-dose-compartment radiation field - boron dose (from 10 B(n,α) 7 Li) reaction), proton dose from 14 N(n,p) 14 C reaction, neutron dose (mainly fast and epithermal neutrons) and gamma-ray dose (external and from capture reaction 1 H(n,γ) 2 D). Because of this the relation between the absorbed dose and the biological effects is very complex and all the above mentioned absorbed dose components should be determined. From this point of view, the recombination chambers can be very useful instruments for characterization of the BNCT beams. They can be used for determination of gamma and high-LET dose components for the characterization of radiation quality of mixed radiation fields by recombination microdosimetric method (RMM). In present work, a graphite high-pressure recombination chamber filled with nitrogen, 10 BF 3 and tissue equivalent gas was used for studies on application of RMM for BNCT dosimetry. The use of these gases or their mixtures opens a possibility to design a recombination chamber for determination of the dose fractions due to gamma radiation, fast neutrons, neutron capture on nitrogen and high LET particles from (n, 10 B) reaction in simulated tissue with different content of 10 B. (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
Simulation of merging neutron stars in full general relativity
International Nuclear Information System (INIS)
Shibata, M.
2001-01-01
We have performed 3D numerical simulations for merger of equal mass binary neutron stars in full general relativity. We adopt a Γ-law equation of state in the form P = (Γ - 1)ρε where P, ρ, ε and Γ are the pressure, rest mass density, specific internal energy, and the adiabatic constant. As initial conditions, we adopt models of irrotational binary neutron stars in a quasiequilibrium state. Simulations have been carried out for a wide range of Γ and compactness of neutron stars, paying particular attention to the final product and gravitational waves. We find that the final product depends sensitively on the initial compactness of the neutron stars: In a merger between sufficiently compact neutron stars, a black hole is formed in a dynamical timescale. As the compactness is decreased, the formation timescale becomes longer and longer. It is also found that a differentially rotating massive neutron star is formed instead of a black hole for less compact binary cases. In the case of black hole formation, the disk mass around the black hole appears to be very small; less than 1% of the total rest mass. It is indicated that waveforms of high-frequency gravitational waves after merger depend strongly on the compactness of neutron stars before the merger. We point out importance of detecting such gravitational waves of high frequency to constrain the maximum allowed mass of neutron stars. (author)
Comparison of Monte Carlo method and deterministic method for neutron transport calculation
International Nuclear Information System (INIS)
Mori, Takamasa; Nakagawa, Masayuki
1987-01-01
The report outlines major features of the Monte Carlo method by citing various applications of the method and techniques used for Monte Carlo codes. Major areas of its application include analysis of measurements on fast critical assemblies, nuclear fusion reactor neutronics analysis, criticality safety analysis, evaluation by VIM code, and calculation for shielding. Major techniques used for Monte Carlo codes include the random walk method, geometric expression method (combinatorial geometry, 1, 2, 4-th degree surface and lattice geometry), nuclear data expression, evaluation method (track length, collision, analog (absorption), surface crossing, point), and dispersion reduction (Russian roulette, splitting, exponential transform, importance sampling, corrected sampling). Major features of the Monte Carlo method are as follows: 1) neutron source distribution and systems of complex geometry can be simulated accurately, 2) physical quantities such as neutron flux in a place, on a surface or at a point can be evaluated, and 3) calculation requires less time. (Nogami, K.)
Simulation of Light Collection for Neutron Electrical Dipole Moment measurement
Ji, Pan; nEDM Collaboration
2017-09-01
nEDM (Neutron Electrical Dipole moment) measurement addresses a critical topic in particle physics and Standard Model, that is CPT violation in neutron electrical dipole moment if detected in which the Time reversal violation is connected to the matter/antimatter imparity of the universe. The neutron electric dipole moment was first measured in 1950 by Smith, Purcell, and Ramsey at the Oak Ridge Reactor - the first intense neutron source. This measurement showed that the neutron was very nearly round (to better than one part in a million). The goal of the nEDM experiment is to further improve the precision of this measurement by another factor of 100. The signal from the experiment is detected by collecting the photons generated when neutron beams were captured by liquid helium 3. The Geant4 simulation project that I participate simulates the process of light collection to improve the design for higher capture efficiency. The simulated geometry includes light source, reflector, wavelength shifting fibers, wavelength shifting TPB and acrylic as in real experiment. The UV photons exiting from Helium go through two wavelength-shifting processes in TPB and fibers to be finally captured. Oak Ridge National Laboratory Neutron Electric Dipole Moment measurement project.
Maucec, M.; Rigollet, C.
The performance of a detection system based on the pulsed fast/thermal neutron analysis technique was assessed using Monte Carlo simulations. The aim was to develop and implement simulation methods, to support and advance the data analysis techniques of the characteristic gamma-ray spectra,
Method for determining thermal neutron decay times of earth formations
International Nuclear Information System (INIS)
Arnold, D.M.
1976-01-01
A method is disclosed for measuring the thermal neutron decay time of earth formations in the vicinity of a well borehole. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different intensity modulation frequencies. The tangents of the relative phase angles of the fast neutrons and the resulting thermal neutrons at each of the three frequencies of modulation are measured. First and second approximations to the earth formation thermal neutron decay time are derived from the three tangent measurements. These approximations are then combined to derive a value for the true earth formation thermal neutron decay time
Special methods used in neutron dosimetry
International Nuclear Information System (INIS)
Mas, P.
1975-01-01
The methods used in radiation reactor dosimetry which do not depend on fission reaction nor foil activation are reviewed. These other techniques are especially the following: the different types of self-powered detectors, with fast or slow response, their characteristics of noise and temperature effect, their practical uses: the damage detectors, considering the variations of their physical properties (resistivity, density, ...) and their use for characterizing the neutron spectra; the little loop with circulating fluids (air, nitrogen, helium, water) [fr
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)
GEM simulation methods development
International Nuclear Information System (INIS)
Tikhonov, V.; Veenhof, R.
2002-01-01
A review of methods used in the simulation of processes in gas electron multipliers (GEMs) and in the accurate calculation of detector characteristics is presented. Such detector characteristics as effective gas gain, transparency, charge collection and losses have been calculated and optimized for a number of GEM geometries and compared with experiment. A method and a new special program for calculations of detector macro-characteristics such as signal response in a real detector readout structure, and spatial and time resolution of detectors have been developed and used for detector optimization. A detailed development of signal induction on readout electrodes and electronics characteristics are included in the new program. A method for the simulation of charging-up effects in GEM detectors is described. All methods show good agreement with experiment
Experimental methods of effective delayed neutron fraction
International Nuclear Information System (INIS)
Yamaye, Yoshihiro
1995-01-01
The defining principle and examples of β eff measurement method: the substitutional method, Cf neutron source method, Bennett method, the coupling coefficient method and Nelson method were introduced and surveyed. Measurement errors and C/E value of the substitutional, Cf ray source and Bennett method were of the order of 3%, 5% and 3 - 6% and 0.903 - 0.965, 1.85 and 1.019 - 1.165, respectably. Evaluation of the absolute value is so hard that β eff measurement belongs to the difficult experiment. The dependence on nuclear calculation in decreasing order is the substitutional, Cf ray source, Bennett, the coupling coefficient and Nelson number method. If good substitute materials were selected, the substitutional method has possibility to determine β eff by small correction value or independent on calculation. (S.Y.)
Loomis, E N; Grim, G P; Wilde, C; Wilson, D C; Morgan, G; Wilke, M; Tregillis, I; Merrill, F; Clark, D; Finch, J; Fittinghoff, D; Bower, D
2010-10-01
Development of analysis techniques for neutron imaging at the National Ignition Facility is an important and difficult task for the detailed understanding of high-neutron yield inertial confinement fusion implosions. Once developed, these methods must provide accurate images of the hot and cold fuels so that information about the implosion, such as symmetry and areal density, can be extracted. One method under development involves the numerical inversion of the pinhole image using knowledge of neutron transport through the pinhole aperture from Monte Carlo simulations. In this article we present results of source reconstructions based on simulated images that test the methods effectiveness with regard to pinhole misalignment.
Monte Carlo simulation of neutron scattering instruments
International Nuclear Information System (INIS)
Seeger, P.A.
1995-01-01
A library of Monte Carlo subroutines has been developed for the purpose of design of neutron scattering instruments. Using small-angle scattering as an example, the philosophy and structure of the library are described and the programs are used to compare instruments at continuous wave (CW) and long-pulse spallation source (LPSS) neutron facilities. The Monte Carlo results give a count-rate gain of a factor between 2 and 4 using time-of-flight analysis. This is comparable to scaling arguments based on the ratio of wavelength bandwidth to resolution width
Simulation of neutron background for DINO experiment
International Nuclear Information System (INIS)
Meghna, K.K.; Bhattacharjee, Pijushpani; Bhattacharya, Satyaki
2017-01-01
Various cosmological observations such as rotation curve of galaxies, gravitational lensing etc. establish the existence of a non-luminous matter known as Dark Matter which constitutes about 27% of the matter content of the universe. Despite the evidence for the existence of dark matter, its constituents are still unknown. In underground laboratories, neutrons can be generated mainly by spontaneous fission of U and radiogenic processes, such as by U / Th (α;n) reactions on the rock materials and by cosmogenic processes, such as interaction of cosmic ray muons with rock and shielding materials. We have estimated the flux of both the cosmogenic and the radiogenic neutrons for Jaduguda laboratory facility
Analysis of Neutron Flux Using Monte Carlo Methods
International Nuclear Information System (INIS)
Picha, Roppon
2007-08-01
Full text: The energy profile of neutrons from a fission reactor core and a neutron irradiation setup are simulated. The neutron doses deposited inside casings of aluminum, cadmium, and tantalum are studied via MCNP simulations to estimate the doses received by materials with different types of shielding. It is found that the difference in dose reduction between cadmium and tantalum is most pronounced at the thermal energy region
Simulations Of Neutron Beam Optic For Neutron Radiography Collimator Using Ray Tracing Methodology
International Nuclear Information System (INIS)
Norfarizan Mohd Said; Muhammad Rawi Mohamed Zin
2014-01-01
Ray- tracing is a technique for simulating the performance of neutron instruments. McStas, the open-source software package based on a meta-language, is a tool for carrying out ray-tracing simulations. The program has been successfully applied in investigating neutron guide design, flux optimization and other related areas with high complexity and precision. The aim of this paper is to discuss the implementation of ray-tracing technique with McStas for simulating the performance of neutron collimation system developed for imaging system of TRIGA RTP reactor. The code for the simulation was developed and the results are presented. The analysis of the performance is reported and discussed. (author)
International Nuclear Information System (INIS)
Ahn, Seong-Kyu; Lee, Tae-Hoon; Shin, Hee-Sung; Kim, Ho-Dong
2009-01-01
An active neutron coincidence counter using a neutron generator as an interrogation source has been suggested. Because of the high energy of the interrogation neutron source, 2.5 MeV, the induced fission rate is strongly affected by the moderator design. MCNPX simulation has been performed to evaluate the performance achieved with these moderators. The side- and bottom-moderator are significantly important to thermalize neutrons to induce fission. Based on the simulation results, the moderators are designed to be adapted to the experimental system. Their preliminary performance has been tested by using natural uranium oxide powder samples. For a sample of up to 3.5 kg, which contains 21.7 g of 235 U, 2.64 cps/g- 235 U coincidence events have been measured. Mean background error was 9.57 cps and the resultant coincidence error was 13.8 cps. The experimental result shows the current status of an active counting using a neutron generator which still has some challenges to overcome. However, the controllability of an interrogation source makes this system more applicable for a variety of combinations with other non-destructive methods like a passive coincidence counting especially under a harsh environment such as a hot cell. More precise experimental setup and tests with higher enriched samples will be followed to develop a system to apply it to an active measurement for the safeguards of a spent fuel treatment process.
A pulsed neutron Ramsey's method
Energy Technology Data Exchange (ETDEWEB)
Masuda, Y. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan)]. E-mail: yasuhiro.masuda@kek.jp; Ino, T. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Jeong, S.C. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Muto, S. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Skoy, V. [Joint Institute for Nuclear Reasearch, 141980 Dubna (Russian Federation); Watanabe, Y. [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan)
2005-02-15
A Ramsey's method with pulsed neutrons is proposed. A Ramsey signal, which is a neutron spin rotation about a static magnetic field for a time interval between two separated oscillatory fields, is observed as a function of a neutron time of flight (TOF) in this method. The neutron spin rotation or the RF oscillation is used as a clock of the neutron velocity measurement which ranges from cold to epithermal neutron energies. This method together with the TOF measurement can be used for neutron inelastic scattering experiments. In addition, this method can be applied to the measurement of magnetic and pseudomagnetic fields in matter, and also to neutron spin manipulation for spin dependent scattering.
Homotopy analysis method for neutron diffusion calculations
International Nuclear Information System (INIS)
Cavdar, S.
2009-01-01
The Homotopy Analysis Method (HAM), proposed in 1992 by Shi Jun Liao and has been developed since then, is based on a fundamental concept in differential geometry and topology, the homotopy. It has proved useful for problems involving algebraic, linear/non-linear, ordinary/partial differential and differential-integral equations being an analytic, recursive method that provides a series sum solution. It has the advantage of offering a certain freedom for the choice of its arguments such as the initial guess, the auxiliary linear operator and the convergence control parameter, and it allows us to effectively control the rate and region of convergence of the series solution. HAM is applied for the fixed source neutron diffusion equation in this work, which is a part of our research motivated by the question of whether methods for solving the neutron diffusion equation that yield straightforward expressions but able to provide a solution of reasonable accuracy exist such that we could avoid analytic methods that are widely used but either fail to solve the problem or provide solutions through many intricate expressions that are likely to contain mistakes or numerical methods that require powerful computational resources and advanced programming skills due to their very nature or intricate mathematical fundamentals. Fourier basis are employed for expressing the initial guess due to the structure of the problem and its boundary conditions. We present the results in comparison with other widely used methods of Adomian Decomposition and Variable Separation.
Method and apparatus for generating neutrons
International Nuclear Information System (INIS)
Cranberg, L.
1978-01-01
An apparatus and method for generating high-energy neutrons are disclosed. Neutron emissive target material is deposited on one or more surfaces on a rotatable, hollow, toroidal target support. The surfaces are bombarded by beams of ions of generally rectangular cross section, so that when the bombarded surfaces are viewed end-wise, a compact, generally square source of neutrons is provided, such as is required for collimation. A combination of molecular and atomic ions emitted from at least one conventional accelerator are passed through a magnetic field for the purpose of separating the ions into one homogeneous group of atomic and one homogeneous group of molecular ions before the ions are allowed to impinge on the target surfaces. One accelerator directs ions to each target surface as the target rotates. Coolant is directed through a cavity within the toroidal support for the purpose of cooling the target support and target material. A refrigerated surface is placed in close proximity to the target surface to condense vapors which might prove harmful to the target and for thermally cooling said target
Neutron absorbers, and the production method
International Nuclear Information System (INIS)
Kayano, Hideo; Yajima, Seishi; Oono, Hironori.
1979-01-01
Purpose: To integrally sinter a metal powder and a metal network material thereby to obtain a material having a high neutron absorbing function, an excellent corrosion resistance and an excellent oxidation resistance. Method: An element having a high neutron absorbing function, such as Gd, or a compound thereof and a powder of a metal having excellent corrosion resistance, oxidation resistance and ductility, such as Fe, Cr or the like are uniformly mixed with each other. In a case where a substance having a neutron absorbing function is a hydroxide an organic complex or the like, it is formed into a gel-like substance and mixed uniformly with the metal powder, the gel-like substance being pasted, and covered on the surface of the metal powder and dried. Then, the mixture or the dry coated material is extended and the metal network material having excellent corrosion resistance, oxidation resistance and ductility is covered or interposed or between at least one layer of upper, intermediate or lower layers of said laminated material, and thereafter is subjected to cold or hot rolling, and then sintered and furthermore rolled, if necessary, the thus treated material being burned in vacuum or a non-oxidizing atmosphere. (Kamimura, M.)
The neutronic method for measuring soil moisture
International Nuclear Information System (INIS)
Couchat, Ph.
1967-01-01
The three group diffusion theory being chosen as the most adequate method for determining the response of the neutron soil moisture probe, a mathematical model is worked out using a numerical calculation programme with Fortran IV coding. This model is fitted to the experimental conditions by determining the effect of different parameters of measuring device: channel, fast neutron source, detector, as also the soil behaviour under neutron irradiation: absorbers, chemical binding of elements. The adequacy of the model is tested by fitting a line through the image points corresponding to the couples of experimental and theoretical values, for seven media having different chemical composition: sand, alumina, line stone, dolomite, kaolin, sandy loam, calcareous clay. The model chosen gives a good expression of the dry density influence and allows α, β, γ and δ constants to be calculated for a definite soil according to the following relation which gives the count rate of the soil moisture probe: N = (α ρ s +β) H v +γ ρ s + δ. (author) [fr
Method of manufacturing neutron protecting materials
Energy Technology Data Exchange (ETDEWEB)
Kakibana, Hidetake; Okamoto, Masazane; Fujii, Yasuhiko; Koguchi, Noboru; Takesute, Morihito; Miyamatsu, Tokuhisa
1985-06-03
Purpose: To manufacture neutron protecting materials which are highly flexible and can be shaped with ease at a good workability. Method: In this invention, natural lithium, natural boron such as Li-6 or B-10 or enriched isotope thereof with a great neutron absorption cross section is fixed to fibers. As a specific example, lumps of copolymer fibers are fabricated into weave sheets in a carding machine and applied with needle punching to prepare felt-like products. They are conditioned to OH or H type, which are respectively immersed in saturated aqueous boric acid or 1M-aqueous solution of lithium hydroxide and then dewatered and dried. As a result, boric acid type anion exchange fibers and lithium type cation exchange fibers can be obtained from the former and the latter respectively. In this way, blankets or cloths which are light in weight, flexible and have high neutron absorbing performance can be shaped. They are also in good fitting contact to a human body. (Kamimura, M.).
Software for simulation and design of neutron scattering instrumentation
DEFF Research Database (Denmark)
Bertelsen, Mads
designed using the software. The Union components uses a new approach to simulation of samples in McStas. The properties of a sample are split into geometrical and material, simplifying user input, and allowing the construction of complicated geometries such as sample environments. Multiple scattering...... from conventional choices. Simulation of neutron scattering instrumentation is used when designing instrumentation, but also to understand instrumental effects on the measured scattering data. The Monte Carlo ray-tracing package McStas is among the most popular, capable of simulating the path of each...... neutron through the instrument using an easy to learn language. The subject of the defended thesis is contributions to the McStas language in the form of the software package guide_bot and the Union components.The guide_bot package simplifies the process of optimizing neutron guides by writing the Mc...
Energy Technology Data Exchange (ETDEWEB)
Gheisari, R., E-mail: gheisari@pgu.ac.ir [Physics Department, Persian Gulf University, Bushehr 75169 (Iran, Islamic Republic of); Nuclear Energy Research Center, Persian Gulf University, Bushehr 75169 (Iran, Islamic Republic of); Firoozabadi, M. M.; Mohammadi, H. [Department of Physics, University of Birjand, Birjand 97175 (Iran, Islamic Republic of)
2014-01-15
A new idea to calculate ultracold neutron (UCN) production by using Monte Carlo simulation method to calculate the cold neutron (CN) flux and an analytical approach to calculate the UCN production from the simulated CN flux was given. A super-thermal source (UCN source) was modeled based on an arrangement of D{sub 2}O and solid D{sub 2} (sD{sub 2}). The D{sub 2}O was investigated as the neutron moderator, and sD{sub 2} as the converter. In order to determine the required parameters, a two-dimensional (2D) neutron balance equation written in Matlab was combined with the MCNPX simulation code. The 2D neutron-transport equation in cylindrical (ρ − z) geometry was considered for 330 neutron energy groups in the sD{sub 2}. The 2D balance equation for UCN and CN was solved using simulated CN flux as boundary value. The UCN source dimensions were calculated for the development of the next UCN source. In the optimal condition, the UCN flux and the UCN production rate (averaged over the sD{sub 2} volume) equal to 6.79 × 10{sup 6} cm{sup −2}s{sup −1} and 2.20 ×10{sup 5} cm{sup −3}s{sup −1}, respectively.
Time interval approach to the pulsed neutron logging method
International Nuclear Information System (INIS)
Zhao Jingwu; Su Weining
1994-01-01
The time interval of neighbouring neutrons emitted from a steady state neutron source can be treated as that from a time-dependent neutron source. In the rock space, the neutron flux is given by the neutron diffusion equation and is composed of an infinite terms. Each term s composed of two die-away curves. The delay action is discussed and used to measure the time interval with only one detector in the experiment. Nuclear reactions with the time distribution due to different types of radiations observed in the neutron well-logging methods are presented with a view to getting the rock nuclear parameters from the time interval technique
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)
Determination Of Simulated Pellet To Pellet Gap Using Neutron Radiography
International Nuclear Information System (INIS)
Kusnowo, A.
1996-01-01
The defect on the irradiated fuel element could be detected using neutron radiography. The defect could occurred in pellet to pellet gap, cladding, or even cladding to pellet gap. An investigations has been performed to detect pellet to pellet gap defect that might occur in an irradiated fuel element. An Al foil of 0,1; 0,2; 0,3; und 0,4 mm was inserted between pellets to simulate various pellet to pellet gap. The neutron radiography used had power of 700 kW. The result showed that this simulation represented well enough problems that irradiated fuel element may experience
A simulation report for the neutron guide development at HANARO
International Nuclear Information System (INIS)
Cho, S. J.; Soo, J. Y.; Seong, B. S.; Lee, C. H.; Kim, H. R.
2006-04-01
Lately, a demand of the measurement technique on atomic scale has been exceedingly increased over the whole field of the basic and technical science such as biotechnology, nano-technology, solid state physics, solid chemistry etc. Therefore a project called 'infrastructure construction for cold neutron research and utilization technique development' was launched in KAERI in July 2003, in order to raise a domestic basic science with an international level and elevate a international competitiveness for the bio-, nano- and informatics technology area through a wide contribution in a material structure research field. In order to accomplish this project until 2008, some important developments were launched at a same time such as a cold neutron source which shifts neutrons from a short wavelength range to a long wavelength, a system driving part for a smooth operation of a cold neutron source, and a neutron guide tube to be able to send neutrons to spectrometers located over a long distance. The guide simulation should be preferentially performed for an effective use of expensive neutron to meet the requirements such as wavelengths and type of instruments, experimental space, interferences with other instruments. The objective of this study was to decide guide shape, dimension, amount, curvature and instrument layout
Parameters of explosives detection through tagged neutron method
Energy Technology Data Exchange (ETDEWEB)
Bagdasaryan, Kh.E.; Batyaev, V.F.; Belichenko, S.G., E-mail: consul757@mail.ru; Bestaev, R.R.; Gavryuchenkov, A.V.; Karetnikov, M.D.
2015-06-01
The potentialities of tagged neutron method (TNM) for explosives detection are examined on the basis of an idealized geometrical model. The model includes ING-27 14 MeV neutron generator with a built-in α-detector, a LYSO γ-detector and samples of material to be identified of approximately 0.3 kg each: explosives imitators (trinitrotoluene - TNT, tetryl, RDX and ammonium nitrate), legal materials (sugar, water, silk and polyethylene). The samples were unshielded or shielded by a paper layer of various thicknesses. The experimental data were interpreted by numerical simulation using a Poisson distribution of signals with the statistical parameters defined experimentally. The detection parameters were obtained by a pattern classification theory and a Bayes classifier.
Monte Carlo simulation of mixed neutron-gamma radiation fields and dosimetry devices
International Nuclear Information System (INIS)
Zhang, Guoqing
2011-01-01
Monte Carlo methods based on random sampling are widely used in different fields for the capability of solving problems with a large number of coupled degrees of freedom. In this work, Monte Carlos methods are successfully applied for the simulation of the mixed neutron-gamma field in an interim storage facility and neutron dosimeters of different types. Details are discussed in two parts: In the first part, the method of simulating an interim storage facility loaded with CASTORs is presented. The size of a CASTOR is rather large (several meters) and the CASTOR wall is very thick (tens of centimeters). Obtaining the results of dose rates outside a CASTOR with reasonable errors costs usually hours or even days. For the simulation of a large amount of CASTORs in an interim storage facility, it needs weeks or even months to finish a calculation. Variance reduction techniques were used to reduce the calculation time and to achieve reasonable relative errors. Source clones were applied to avoid unnecessary repeated calculations. In addition, the simulations were performed on a cluster system. With the calculation techniques discussed above, the efficiencies of calculations can be improved evidently. In the second part, the methods of simulating the response of neutron dosimeters are presented. An Alnor albedo dosimeter was modelled in MCNP, and it has been simulated in the facility to calculate the calibration factor to get the evaluated response to a Cf-252 source. The angular response of Makrofol detectors to fast neutrons has also been investigated. As a kind of SSNTD, Makrofol can detect fast neutrons by recording the neutron induced heavy charged recoils. To obtain the information of charged recoils, general-purpose Monte Carlo codes were used for transporting incident neutrons. The response of Makrofol to fast neutrons is dependent on several factors. Based on the parameters which affect the track revealing, the formation of visible tracks was determined. For
Monte Carlo simulation of mixed neutron-gamma radiation fields and dosimetry devices
Energy Technology Data Exchange (ETDEWEB)
Zhang, Guoqing
2011-12-22
Monte Carlo methods based on random sampling are widely used in different fields for the capability of solving problems with a large number of coupled degrees of freedom. In this work, Monte Carlos methods are successfully applied for the simulation of the mixed neutron-gamma field in an interim storage facility and neutron dosimeters of different types. Details are discussed in two parts: In the first part, the method of simulating an interim storage facility loaded with CASTORs is presented. The size of a CASTOR is rather large (several meters) and the CASTOR wall is very thick (tens of centimeters). Obtaining the results of dose rates outside a CASTOR with reasonable errors costs usually hours or even days. For the simulation of a large amount of CASTORs in an interim storage facility, it needs weeks or even months to finish a calculation. Variance reduction techniques were used to reduce the calculation time and to achieve reasonable relative errors. Source clones were applied to avoid unnecessary repeated calculations. In addition, the simulations were performed on a cluster system. With the calculation techniques discussed above, the efficiencies of calculations can be improved evidently. In the second part, the methods of simulating the response of neutron dosimeters are presented. An Alnor albedo dosimeter was modelled in MCNP, and it has been simulated in the facility to calculate the calibration factor to get the evaluated response to a Cf-252 source. The angular response of Makrofol detectors to fast neutrons has also been investigated. As a kind of SSNTD, Makrofol can detect fast neutrons by recording the neutron induced heavy charged recoils. To obtain the information of charged recoils, general-purpose Monte Carlo codes were used for transporting incident neutrons. The response of Makrofol to fast neutrons is dependent on several factors. Based on the parameters which affect the track revealing, the formation of visible tracks was determined. For
Research on neutron source multiplication method in nuclear critical safety
International Nuclear Information System (INIS)
Zhu Qingfu; Shi Yongqian; Hu Dingsheng
2005-01-01
The paper concerns in the neutron source multiplication method research in nuclear critical safety. Based on the neutron diffusion equation with external neutron source the effective sub-critical multiplication factor k s is deduced, and k s is different to the effective neutron multiplication factor k eff in the case of sub-critical system with external neutron source. The verification experiment on the sub-critical system indicates that the parameter measured with neutron source multiplication method is k s , and k s is related to the external neutron source position in sub-critical system and external neutron source spectrum. The relation between k s and k eff and the effect of them on nuclear critical safety is discussed. (author)
Monte Carlo simulation of fast neutron scattering experiments including DD-breakup neutrons
Energy Technology Data Exchange (ETDEWEB)
Schmidt, D.; Siebert, B.R.L.
1993-06-01
The computational simulation of the deuteron breakup in a scattering experiment has been investigated. Experimental breakup spectra measured at 16 deuteron energies and at 7 angles for each energy served as the data base. Analysis of these input data and of the conditions of the scattering experiment made it possible to reduce the input data. The use of one weighted breakup spectrum is sufficient to simulate the scattering spectra at one incident neutron energy. A number of tests were carried out to prove the validity of this result. The simulation of neutron scattering on carbon, including the breakup, was compared with measured spectra. Differences between calculated and measured spectra were for the most part within the experimental uncertainties. Certain significant deviations can be attributed to erroneous scattering cross sections taken from an evaluation and used in the simulation. Scattering on higher-lying states in [sup 12]C can be analyzed by subtracting the simulated breakup-scattering from the experimental spectra. (orig.)
Monte Carlo simulation of fast neutron scattering experiments including DD-breakup neutrons
International Nuclear Information System (INIS)
Schmidt, D.; Siebert, B.R.L.
1993-06-01
The computational simulation of the deuteron breakup in a scattering experiment has been investigated. Experimental breakup spectra measured at 16 deuteron energies and at 7 angles for each energy served as the data base. Analysis of these input data and of the conditions of the scattering experiment made it possible to reduce the input data. The use of one weighted breakup spectrum is sufficient to simulate the scattering spectra at one incident neutron energy. A number of tests were carried out to prove the validity of this result. The simulation of neutron scattering on carbon, including the breakup, was compared with measured spectra. Differences between calculated and measured spectra were for the most part within the experimental uncertainties. Certain significant deviations can be attributed to erroneous scattering cross sections taken from an evaluation and used in the simulation. Scattering on higher-lying states in 12 C can be analyzed by subtracting the simulated breakup-scattering from the experimental spectra. (orig.)
An alternative method for the measurement of neutron flux
Indian Academy of Sciences (India)
A simple and easy method for measuring the neutron flux is presented. This paper deals with the experimental verification of neutron dose rate–flux relationship for a non-dissipative medium. Though the neutron flux cannot be obtained from the dose rate in a dissipative medium, experimental result shows that for ...
Modularized Parallel Neutron Instrument Simulation on the TeraGrid
International Nuclear Information System (INIS)
Chen, Meili; Cobb, John W.; Hagen, Mark E.; Miller, Stephen D.; Lynch, Vickie E.
2007-01-01
In order to build a bridge between the TeraGrid (TG), a national scale cyberinfrastructure resource, and neutron science, the Neutron Science TeraGrid Gateway (NSTG) is focused on introducing productive HPC usage to the neutron science community, primarily the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL). Monte Carlo simulations are used as a powerful tool for instrument design and optimization at SNS. One of the successful efforts of a collaboration team composed of NSTG HPC experts and SNS instrument scientists is the development of a software facility named PSoNI, Parallelizing Simulations of Neutron Instruments. Parallelizing the traditional serial instrument simulation on TeraGrid resources, PSoNI quickly computes full instrument simulation at sufficient statistical levels in instrument de-sign. Upon SNS successful commissioning, to the end of 2007, three out of five commissioned instruments in SNS target station will be available for initial users. Advanced instrument study, proposal feasibility evaluation, and experiment planning are on the immediate schedule of SNS, which pose further requirements such as flexibility and high runtime efficiency on fast instrument simulation. PSoNI has been redesigned to meet the new challenges and a preliminary version is developed on TeraGrid. This paper explores the motivation and goals of the new design, and the improved software structure. Further, it describes the realized new features seen from MPI parallelized McStas running high resolution design simulations of the SEQUOIA and BSS instruments at SNS. A discussion regarding future work, which is targeted to do fast simulation for automated experiment adjustment and comparing models to data in analysis, is also presented
Shielding evaluation of neutron generator hall by Monte Carlo simulations
Energy Technology Data Exchange (ETDEWEB)
Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)
2017-04-01
A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)
Cold neutron radiographic apparatus and method
International Nuclear Information System (INIS)
Larsen, J.E.
1980-01-01
Cold neutron radiography may be improved by matching neutron temperature to the specific material to be analyzed. It is possible to bombard the material with neutrons having the precise average temperature necessary to realize the minimum attenuation coefficient, or to choose a neutron temperature that would increase the attenuation by inclusions, defects, etc., or to choose a neutron temperature that provides a good balance between sample transmission and defect attenuation. Other neutron temperatures might also be chosen for other reasons. This may be done by having a source of neutrons embedded in a moderator material, such as solid methane, and cooling the moderator material to the desired temperature by a cryogenic refrigerator. In another embodiment, neutrons from a nuclear reactor are passed through a moderator cooled by a cryogenic refrigerator. Since the neutron temperature is matched to the material being radiographically inspected, improved contrast and resolution can be obtained through thicker materials than it has heretofore been possible to analyze by cold neutron radiography. More optimum filtering of a neutron beam is also achieved by using a cryogenic refrigerator to cool the neutron beam filter. (auth)
Thermal expansion study of simulated DUPIC fuel using neutron diffraction
International Nuclear Information System (INIS)
Kang, Kweon Ho; Ryu, H. J.; Bae, J. H.; Kim, H. S.; Song, K. C.; Yang, M. S.; Choi, Y. N.; Han, Y. S.; Oh, H. S.
2001-07-01
The lattice parameters of simulated DUPIC fuel and UO2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO2 and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO2. For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO2 and simulated DUPIC fuel are 10.471 ''10-6 and 10.751 ''10-6 K-1, respectively
Polarized Neutron Reflectivity Simulation of Ferromagnet/ Antiferromagnet Thin Films
Energy Technology Data Exchange (ETDEWEB)
Kim, Ki Yeon; Lee, Jeong Soo
2008-02-15
This report investigates the current simulating and fitting programs capable of calculating the polarized neutron reflectivity of the exchange-biased ferromagnet/antiferromagnet magnetic thin films. The adequate programs are selected depending on whether nonspin flip and spin flip reflectivities of magnetic thin films and good user interface are available or not. The exchange-biased systems such as Fe/Cr, Co/CoO, CoFe/IrMn/Py thin films have been simulated successfully with selected programs.
A methodology of neutronic-thermodynamics simulation for fast reactor
International Nuclear Information System (INIS)
Waintraub, M.
1986-01-01
Aiming at a general optimization of the project, controlled fuel depletion and management, this paper develop a neutronic thermodynamics simulator, SIRZ, which besides being sufficiently precise, is also economic. That results in a 75% reduction in CPU time, for a startup calculation, when compared with the same calculation at the CITATION code. The simulation system by perturbation calculations, applied to fast reactors, which produce errors smaller than 1% in all components of the reference state given by the CITATION code was tested. (author)
International Nuclear Information System (INIS)
Sasaki, Toshihiko; Takago, Shigeki
2016-01-01
This paper outlined a stress measurement method using neutrons, and introduced the application examples to stress measurement for metal-based composite materials. In the angular dispersion type measurement using a steady-state reactor type neutron source, the white beams taken out from a nuclear reactor are monochromatized (wavelength λ is a constant value) with a single crystal monochromator and utilized. As an example of measurement, there was the case as follows: the stress of a sintered material which has been put to practical use as valve seat part for automobiles was measured by the neutron method, and the deformation behavior during load was studied. This study performed neutron diffraction measurement using a residual stress analyzer (RESA: Diffractometer for Residual Stress Analysis) installed at JAEA's experimental reactor JRR-3. As a result, it was found that the stress state of the sintered composite material of Fe-Cr and TiN can be predicted with a micromechanics model. A neutron diffraction ring can be obtained using a neutron image plate (IP), where fine powder of gadolinium (Gd) was incorporated into IP for X-rays, and it can be used as an IP reader in the same way as the case of X-rays. A report has been introduced on the examination results of the highly accurate stress measurement by applying the cos α method devised for X-ray stress measurement to neutron diffraction ring. (A.O.)
Simulation of neutron fluxes around the W7-X Stellarator
Energy Technology Data Exchange (ETDEWEB)
Andersson, Jenny
1999-12-01
A new fusion experiment, the WENDELSTEIN 7-X Stellarator (W7-X), will be undertaken in Greifswald in Germany. Measurements of the neutron flux will provide information on fusion reaction rates and possibly also on ion temperatures as function of time. For this purpose moderating neutron counters will be designed, tested, calibrated and eventually used at W7-X. Extensive Monte-Carlo simulations have been performed in order to select the most suitable detector and moderator combination with a flat response function and highest achievable efficiency. Different detector configurations with different moderating materials have been tried out, showing that a 32 cm thick graphite moderating BF{sub 3} -counter gives the desired flat response and sufficient efficiency. Neutron spectra calculations have been made for different torus models and the influence of floor, walls and ceiling (i.e. reactor hall) have been investigated. Presented results suggest that a more detailed torus model significantly reduces the number of neutron counts at the detector. Calculations including the reactor hall indicate a tendency of shifting the neutron spectra towards the thermal region. The main part of the scattered neutrons are back-scattered from the floor. Finally, calculations on the graphite moderating BF{sub 3} -counter in the detailed torus environment were performed in order to assess the absolute response function under the influence of the reactor hall. The results show that the detector count rate will increase by only 5-7 % when the reactor hall is taken into account. With a stellarator generating 10{sup 12} to 10{sup 16} neutrons per second the detector count rate will be 2x10{sup 5} to 2x10{sup 9} neutrons per second.
Simulation of neutron fluxes around the W7-X Stellarator
International Nuclear Information System (INIS)
Andersson, Jenny
1999-12-01
A new fusion experiment, the WENDELSTEIN 7-X Stellarator (W7-X), will be undertaken in Greifswald in Germany. Measurements of the neutron flux will provide information on fusion reaction rates and possibly also on ion temperatures as function of time. For this purpose moderating neutron counters will be designed, tested, calibrated and eventually used at W7-X. Extensive Monte-Carlo simulations have been performed in order to select the most suitable detector and moderator combination with a flat response function and highest achievable efficiency. Different detector configurations with different moderating materials have been tried out, showing that a 32 cm thick graphite moderating BF 3 -counter gives the desired flat response and sufficient efficiency. Neutron spectra calculations have been made for different torus models and the influence of floor, walls and ceiling (i.e. reactor hall) have been investigated. Presented results suggest that a more detailed torus model significantly reduces the number of neutron counts at the detector. Calculations including the reactor hall indicate a tendency of shifting the neutron spectra towards the thermal region. The main part of the scattered neutrons are back-scattered from the floor. Finally, calculations on the graphite moderating BF 3 -counter in the detailed torus environment were performed in order to assess the absolute response function under the influence of the reactor hall. The results show that the detector count rate will increase by only 5-7 % when the reactor hall is taken into account. With a stellarator generating 10 12 to 10 16 neutrons per second the detector count rate will be 2x10 5 to 2x10 9 neutrons per second
International Nuclear Information System (INIS)
Abou-Zaid, A.A.; Pytel, K.
1999-01-01
A non-destructive method of measurement of fissile nuclides content remained in spent fuel from research reactor is presented. The method, called the neutron emission one, is based on counting of fission neutrons emitted from fissile isotopes: 235 U, 239 Pu, 241 Pu. Fissions are induced mainly by neutrons supplied by the external neutron source. Another effects contribute also to the measured neutron population, e. g. source neutrons from penetrating the fuel without being captured and scattered, neutrons (α,n) reactions and from spontaneous fissions of actinides. Complexity of phenomena occurring within the measurement facility required the detailed numerical simulation and experimental studies prior design of ultimate measurement stand. In the previous paper, the results of Monte Carlo simulation on optimisation of measuring stand for neutron emission method were presented. On the basis of those results, the experimental stand for Maria reactor fuel investigation has been designed and manufactured. The present paper, being the continuation of previous one, contains the description of experimental facility and the results of measurements for the fresh fuel (without burnup) and the fuel mock-up (without fissile materials). Although some discrepancies were found between Monte Carlo and experimental results, the main conclusions concerning the optimal geometry of measuring facility have been confirmed. (author)
International Nuclear Information System (INIS)
Ikeda, Hideaki; Takeda, Toshikazu
2001-01-01
A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)
International Nuclear Information System (INIS)
Camargo, Dayana Q. de; Bodmann, Bardo E.J.; Vilhena, Marco T. de; Froehlich, Herberth B.
2011-01-01
In this work we developed a stochastic model to simulate neutron transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using the Monte Carlo method for the propagation of neutrons in different environments. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational time we introduced a variable control volume together with (pseudo-) periodic boundary conditions in order to overcome this problem. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude that there is need to consider the energy dependence and hence defined a spectral effective multiplication factor per Monte Carlo step. (author)
Monte Carlo simulations of neutron-scattering instruments using McStas
DEFF Research Database (Denmark)
Nielsen, K.; Lefmann, K.
2000-01-01
Monte Carlo simulations have become an essential tool for improving the performance of neutron-scattering instruments, since the level of sophistication in the design of instruments is defeating purely analytical methods. The program McStas, being developed at Rise National Laboratory, includes...
Neutron star properties from an NJL model modified to simulate confinement
Energy Technology Data Exchange (ETDEWEB)
Lawley, S. [Special Research Centre for the Subatomic Structure of Matter, University of Adelaide, Adelaide SA 5005 (Australia); Bentz, W. [Department of Physics, School of Science, Tokai University, 117 Kita-Kaname, Hiratsuka 259-1207 (Japan); Thomas, A.W. [Jefferson Lab, 12000 Jefferson Avenue, Newport News, VA 23606 (United States)
2005-04-15
The NJL model has recently been extended with a method to simulate confinement. This leads in mean field approximation to a natural mechanism for the saturation of nuclear matter. We use the model to investigate the equation of state of asymmetric nuclear matter and then use it to compute the properties of neutron stars.
Methods of channeling simulation
International Nuclear Information System (INIS)
Barrett, J.H.
1989-06-01
Many computer simulation programs have been used to interpret experiments almost since the first channeling measurements were made. Certain aspects of these programs are important in how accurately they simulate ions in crystals; among these are the manner in which the structure of the crystal is incorporated, how any quantity of interest is computed, what ion-atom potential is used, how deflections are computed from the potential, incorporation of thermal vibrations of the lattice atoms, correlations of thermal vibrations, and form of stopping power. Other aspects of the programs are included to improve the speed; among these are table lookup, importance sampling, and the multiparameter method. It is desirable for programs to facilitate incorporation of special features of interest in special situations; examples are relaxations and enhanced vibrations of surface atoms, easy substitution of an alternate potential for comparison, change of row directions from layer to layer in strained-layer lattices, and different vibration amplitudes for substitutional solute or impurity atoms. Ways of implementing all of these aspects and features and the consequences of them will be discussed. 30 refs., 3 figs
The experimental method for neutron dose-equivalent detection
International Nuclear Information System (INIS)
Ji Changsong
1992-01-01
A new method, for getting neutron dose-equivalent Cd rode absorption method is described. The method adopts Cd-rode-swarm buck absorption, which greatly improved the neutron sensitivity and simplified the adjustment method. By this method, the author has developed BH3105 model neutron dose equivalent meter, the sensitivity of this instrument reach 10 cps/μSvh -1 . γ-ray depression rate reaches 4000:1, the measurement range is 0.1 μSv/h-10 6 μSv/h. The energy response is good (from thermal neutron-14 MeV neutron), this instrument can be used to measure the dose equivalent of the neutron areas
METHODS OF ASSESSMENT OF THE RELATIVE BIOLOGICAL EFFECTIVENESS OF NEUTRONS IN NEUTRON THERAPY
Directory of Open Access Journals (Sweden)
V. A. Lisin
2017-01-01
Full Text Available The relative biological effectiveness (RBE of fast neutrons is an important factor influencing the quality of neutron therapy therefore, the assessment of RBE is of great importance. Experimental and clinical studies as well as different mathematical and radiobiological models are used for assessing RBE. Research is conducted for neutron sources differing in the method of producing particles, energy and energy spectrum. Purpose: to find and analyze the dose-dependence of fast neutron RBE in neutron therapy using the U-120 cyclotron and NG-12I generator. Material and methods: The optimal method for assessing the relative biological effectiveness of neutrons for neutron therapy was described. To analyze the dependence of the RBE on neutron dose, the multi-target model of cell survival was applied. Results: The dependence of the RBE of neutrons produced from the U-120 cyclotron and NG-120 generator on the dose level was found for a single irradiation of biological objects. It was shown that the function of neutron dose was consistent with similar dependencies found by other authors in the experimental and clinical studies.
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)
Numerical simulation for neutron pinhole imaging in ICF
International Nuclear Information System (INIS)
Chen Faxin; Yang Jianlun; Wen Shuhuai
2005-01-01
Pinhole imaging of the neutron production in laser-driven inertial confinement fusion experiments can provide important information about performance of various capsule designs. In order to get good results in experiments, it is needed to judge performance of various pinhole designs qualitatively or quantitatively before experiment. Calculation of imaging can be simply separated into pinhole imaging and image spectral analysis. In this paper, pinhole imaging is discussed, codes for neutron pinhole imaging and image showing is programed. The codes can be used to provide theoretical foundation for pinhole designing and simulating data for image analysing. (authors)
Simulation report for neutron guide and spectrometer layout at HANARO
International Nuclear Information System (INIS)
Cho, S. J.; Cho, Y. G.; Ryu, J. S.; Seong, B. S.; Lee, C. H.; Shin, J. W.
2006-01-01
A project called 'infrastructure construction for cold neutron research and utilization technique development' was launched in KAERI in July 2003, in order to raise a domestic basic science with an international level and elevate a international competitiveness for the bio-, nano- and informatics technology area. At the end of this project, 3 new instruments and 3 instruments to be moved will be installed in the guide hall. In order to accomplish this project until 2008, guide simulation should be performed for an effective use of expensive neutron to meet the requirements such as wavelengths and type of instruments, experimental space, interferences with other instruments
Directory of Open Access Journals (Sweden)
Mounia Laassiri
Full Text Available For efficient exploitation of research reactors, it is important to discern neutron flux distribution inside the reactor with the best possible precision. For this reason, fission and ionization chambers are used to measure the neutron field. In these arrays, the sequences of the neutron interaction points in the fission chamber can correctly be identified in order to obtain true neutron energies emitted by nuclei of interest. However, together with the neutrons, gamma-rays are also emitted from nuclei and thereby affect neutron spectra. The originality of this study consists in the application of tensor based blind source separation methods to extract independent components from signals recorded at the fission chamber preamplifier’s output. The objective is to achieve software neutron-gamma discrimination using Nonnegative Tensor Factorization tools. For reasons of nuclear safety, we first simulate the neutron flux inside the TRIGA Mark II Reactor using Monte Carlo methods under Geant4 platform linked to Garfield++. Geant4 simulations allow the fission chamber construction whereas linking the model to Garfield++ permits to simulate drift parameters from the ionization of the filling gas, which is not possible otherwise. Keywords: Fission chamber (FC, Geant4, Garfield++, Neutron-gamma discrimination, Nonnegative Tensor Factorization (NTF
International Nuclear Information System (INIS)
Silva, A.X. da; Crispim, V.R.
2002-01-01
This work present the application of a computer package for generating of projection data for neutron computerized tomography, and in second part, discusses an application of neutron tomography, using the projection data obtained by Monte Carlo technique, for the detection and localization of light materials such as those containing hydrogen, concealed by heavy materials such as iron and lead. For tomographic reconstructions of the samples simulated use was made of only six equal projection angles distributed between 0 deg C and 180 deg C, with reconstruction making use of an algorithm (ARIEM), based on the principle of maximum entropy. With the neutron tomography it was possible to detect and locate polyethylene and water hidden by lead and iron (with 1 cm-thick). Thus, it is demonstrated that thermal neutrons tomography is a viable test method which can provide important interior information about test components, so, extremely useful in routine industrial applications.(author)
Monte Carlo simulation of neutron counters for safeguards applications
International Nuclear Information System (INIS)
Looman, Marc; Peerani, Paolo; Tagziria, Hamid
2009-01-01
MCNP-PTA is a new Monte Carlo code for the simulation of neutron counters for nuclear safeguards applications developed at the Joint Research Centre (JRC) in Ispra (Italy). After some preliminary considerations outlining the general aspects involved in the computational modelling of neutron counters, this paper describes the specific details and approximations which make up the basis of the model implemented in the code. One of the major improvements allowed by the use of Monte Carlo simulation is a considerable reduction in both the experimental work and in the reference materials required for the calibration of the instruments. This new approach to the calibration of counters using Monte Carlo simulation techniques is also discussed.
Monte Carlo Simulations of Neutron Oil well Logging Tools
International Nuclear Information System (INIS)
Azcurra, Mario
2002-01-01
Monte Carlo simulations of simple neutron oil well logging tools into typical geological formations are presented.The simulated tools consist of both 14 MeV pulsed and continuous Am-Be neutron sources with time gated and continuous gamma ray detectors respectively.The geological formation consists of pure limestone with 15% absolute porosity in a wide range of oil saturation.The particle transport was performed with the Monte Carlo N-Particle Transport Code System, MCNP-4B.Several gamma ray spectra were obtained at the detector position that allow to perform composition analysis of the formation.In particular, the ratio C/O was analyzed as an indicator of oil saturation.Further calculations are proposed to simulate actual detector responses in order to contribute to understand the relation between the detector response with the formation composition
Monte Carlo simulations of neutron oil well logging tools
International Nuclear Information System (INIS)
Azcurra, Mario O.; Zamonsky, Oscar M.
2003-01-01
Monte Carlo simulations of simple neutron oil well logging tools into typical geological formations are presented. The simulated tools consist of both 14 MeV pulsed and continuous Am-Be neutron sources with time gated and continuous gamma ray detectors respectively. The geological formation consists of pure limestone with 15% absolute porosity in a wide range of oil saturation. The particle transport was performed with the Monte Carlo N-Particle Transport Code System, MCNP-4B. Several gamma ray spectra were obtained at the detector position that allow to perform composition analysis of the formation. In particular, the ratio C/O was analyzed as an indicator of oil saturation. Further calculations are proposed to simulate actual detector responses in order to contribute to understand the relation between the detector response with the formation composition. (author)
Reference neutron radiations. Part 1: Characteristics and methods of production
International Nuclear Information System (INIS)
2001-01-01
ISO 8529 consists of the following parts, under the general title Reference neutron radiations: Part 1: Characteristics and methods of production; Part 2: Calibration fundamentals of radiation protection devices related to the basic quantities characterizing the radiation field; Part 3: Calibration of area and personal dosimeters and determination of response as a function of energy and angle of incidence. This Part 1. of ISO 8529 specifies the reference neutron radiations, in the energy range from thermal up to 20 MeV, for calibrating neutron-measuring devices used for radiation protection purposes and for determining their response as a function of neutron energy. Reference radiations are given for neutron fluence rates of up to 1x10 9 m 2 s-1 , corresponding, at a neutron energy of 1 MeV, to dose-equivalent rates of up to 100 mSv h -1 . This part of ISO 8529 is concerned only with the methods of producing and characterizing the neutron reference radiations. The procedures for applying these radiations for calibrations are described in ISO 8529-2 and ISO 8529-3. The reference radiations specified are the following: neutrons from radionuclide sources, including neutrons from sources in a moderator; neutrons produced by nuclear reactions with charged particles from accelerators; neutrons from reactors. In view of the methods of production and use of them, these reference radiations are divided, for the purposes of this part of ISO 8529, into the following two separate sections. In clause 4, radionuclide neutron sources with wide spectra are specified for the calibration of neutron measuring devices. These sources should be used by laboratories engaged in the routine calibration of neutron-measuring devices, the particular design of which has already been type tested. In clause 5, accelerator-produced monoenergetic neutrons and reactor-produced neutrons with wide or quasi monoenergetic spectra are specified for determining the response of neutron-measuring devices
Domain decomposition methods for the neutron diffusion problem
International Nuclear Information System (INIS)
Guerin, P.; Baudron, A. M.; Lautard, J. J.
2010-01-01
The neutronic simulation of a nuclear reactor core is performed using the neutron transport equation, and leads to an eigenvalue problem in the steady-state case. Among the deterministic resolution methods, simplified transport (SPN) or diffusion approximations are often used. The MINOS solver developed at CEA Saclay uses a mixed dual finite element method for the resolution of these problems. and has shown his efficiency. In order to take into account the heterogeneities of the geometry, a very fine mesh is generally required, and leads to expensive calculations for industrial applications. In order to take advantage of parallel computers, and to reduce the computing time and the local memory requirement, we propose here two domain decomposition methods based on the MINOS solver. The first approach is a component mode synthesis method on overlapping sub-domains: several Eigenmodes solutions of a local problem on each sub-domain are taken as basis functions used for the resolution of the global problem on the whole domain. The second approach is an iterative method based on a non-overlapping domain decomposition with Robin interface conditions. At each iteration, we solve the problem on each sub-domain with the interface conditions given by the solutions on the adjacent sub-domains estimated at the previous iteration. Numerical results on parallel computers are presented for the diffusion model on realistic 2D and 3D cores. (authors)
Monte Carlo simulation of grating-based neutron phase contrast imaging at CPHS
International Nuclear Information System (INIS)
Zhang Ran; Chen Zhiqiang; Huang Zhifeng; Xiao Yongshun; Wang Xuewu; Wie Jie; Loong, C.-K.
2011-01-01
Since the launching of the Compact Pulsed Hadron Source (CPHS) project of Tsinghua University in 2009, works have begun on the design and engineering of an imaging/radiography instrument for the neutron source provided by CPHS. The instrument will perform basic tasks such as transmission imaging and computerized tomography. Additionally, we include in the design the utilization of coded-aperture and grating-based phase contrast methodology, as well as the options of prompt gamma-ray analysis and neutron-energy selective imaging. Previously, we had implemented the hardware and data-analysis software for grating-based X-ray phase contrast imaging. Here, we investigate Geant4-based Monte Carlo simulations of neutron refraction phenomena and then model the grating-based neutron phase contrast imaging system according to the classic-optics-based method. The simulated experimental results of the retrieving phase shift gradient information by five-step phase-stepping approach indicate the feasibility of grating-based neutron phase contrast imaging as an option for the cold neutron imaging instrument at the CPHS.
Neutron spectrum measurement using rise-time discrimination method
International Nuclear Information System (INIS)
Luo Zhiping; Suzuki, C.; Kosako, T.; Ma Jizeng
2009-01-01
PSD method can be used to measure the fast neutron spectrum in n/γ mixed field. A set of assemblies for measuring the pulse height distribution of neutrons is built up,based on a large volume NE213 liquid scintillator and standard NIM circuits,through the rise-time discrimination method. After that,the response matrix is calculated using Monte Carlo method. The energy calibration of the pulse height distribution is accomplished using 60 Co radioisotope. The neutron spectrum of the mono-energetic accelerator neutron source is achieved by unfolding process. Suggestions for further improvement of the system are presented at last. (authors)
Simulation of Thermal Neutron Transport Processes Directly from the Evaluated Nuclear Data Files
Androsenko, P. A.; Malkov, M. R.
The main idea of the method proposed in this paper is to directly extract thetrequired information for Monte-Carlo calculations from nuclear data files. The met od being developed allows to directly utilize the data obtained from libraries and seehs to be the most accurate technique. Direct simulation of neutron scattering in themmal energy range using file 7 ENDF-6 format in terms of code system BRAND has beer achieved. Simulation algorithms have been verified using the criterion x2
M.C. simulation of GEM neutron beam monitor with 10B
International Nuclear Information System (INIS)
Wang Yanfeng; Sun Zhijia; Liu Ben; Zhou Jianrong; Yang Guian; Dong Jing; Xu Hong; Zhou Liang; Huang Guangming; Yang Lei; Li Yi
2010-01-01
The neutron beam monitor based on GEM detector has been carefully studied with the Monte-Carlo method in this article. The simulation framework is including the ANSYS and the Garfield, which was used to compute the electric field of GEM foils and simulate the movement of electrons in gas mixture respectively. The GEM foils' focus and extract coefficients have been obtained. According to the primary results, the performing of the monitor is improved. (authors)
A genetic algorithm based method for neutron spectrum unfolding
International Nuclear Information System (INIS)
Suman, Vitisha; Sarkar, P.K.
2013-03-01
An approach to neutron spectrum unfolding based on a stochastic evolutionary search mechanism - Genetic Algorithm (GA) is presented. It is tested to unfold a set of simulated spectra, the unfolded spectra is compared to the output of a standard code FERDOR. The method was then applied to a set of measured pulse height spectrum of neutrons from the AmBe source as well as of emitted neutrons from Li(p,n) and Ag(C,n) nuclear reactions carried out in the accelerator environment. The unfolded spectra compared to the output of FERDOR show good agreement in the case of AmBe spectra and Li(p,n) spectra. In the case of Ag(C,n) spectra GA method results in some fluctuations. Necessity of carrying out smoothening of the obtained solution is also studied, which leads to approximation of the solution yielding an appropriate solution finally. Few smoothing techniques like second difference smoothing, Monte Carlo averaging, combination of both and gaussian based smoothing methods are also studied. Unfolded results obtained after inclusion of the smoothening criteria are in close agreement with the output obtained from the FERDOR code. The present method is also tested on a set of underdetermined problems, the outputs of which is compared to the unfolded spectra obtained from the FERDOR applied to a completely determined problem, shows a good match. The distribution of the unfolded spectra is also studied. Uncertainty propagation in the unfolded spectra due to the errors present in the measurement as well as the response function is also carried out. The method appears to be promising for unfolding the completely determined as well as underdetermined problems. It also has provisions to carry out the uncertainty analysis. (author)
A multi-group neutron noise simulator for fast reactors
International Nuclear Information System (INIS)
Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe
2013-01-01
Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring
Vagelatos, Nicholas; Steinman, Donald K.; John, Joseph; Young, Jack C.
1981-01-01
A nuclear method and apparatus determines the temperature of a medium by injecting fast neutrons into the medium and detecting returning slow neutrons in three first energy ranges by producing three respective detection signals. The detection signals are combined to produce three derived indicia each systematically related to the population of slow neutrons returning from the medium in a respective one of three second energy ranges, specifically exclusively epithermal neutrons, exclusively substantially all thermal neutrons and exclusively a portion of the thermal neutron spectrum. The derived indicia are compared with calibration indicia similarly systematically related to the population of slow neutrons in the same three second energy ranges returning from similarly irradiated calibration media for which the relationships temperature, neutron absorption cross section and neutron moderating power to such calibration indicia are known. The comparison indicates the temperature at which the calibration indicia correspond to the derived indicia and consequently the temperature of the medium. The neutron absorption cross section and moderating power of the medium can be identified at the same time.
Investigation of propagation algorithms for ray-tracing simulation of polarized neutrons
DEFF Research Database (Denmark)
Bergbäck Knudsen, Erik; Tranum-Rømer, A.; Willendrup, Peter Kjær
2014-01-01
Ray-tracing of polarized neutrons faces a challenge when the neutron propagates through an inhomogeneous magnetic field. This affects simulations of novel instruments using encoding of energy or angle into the neutron spin. We here present a new implementation of propagation of polarized neutrons...
Improved cold neutron radiographic apparatus and method
International Nuclear Information System (INIS)
1981-01-01
An improved cold neutron radiography technique is described in which the neutron temperature is matched to the specific material to be analyzed. In addition to a beam source and detector the apparatus incorporates a cryogenic refrigerator which enables the moderator material to be cooled to a predetermined adjustable temperature below the Bragg edge temperature of the sample. (U.K.)
Method and apparatus for neutron radiation monitoring
International Nuclear Information System (INIS)
Schwarzmann, A.
1985-01-01
A self-calibrated neutron radiation monitor includes a flux responsive element comprised of intrinsic silicon neutron detectors and self-calibration resistors in a single structure. As the resistance of the flux responsive element increases to the value of successive calibration resistors, known increments of flux have been encountered
Design of calibration method in neutron and individual dosimeter
International Nuclear Information System (INIS)
Belkhodia, M.
1984-12-01
Usually albedo dosemeters are calibrated with beam of monoenergetic neutrons. Since neutron energy around neutron sources varies greatly, we applied the calibration method to a mixed field whose energy spectrum lies between 0.025 ev and 10 Mev. The method is based on a mathematical model that deals with the dosimeter response as a function at the neutron energy. The measurements carried out with solid state nuclear track detectors show the dosimeter practical aspect. The albedo dosimeter calibration gave results on good agreement with the international institution recommendations
Albedo's determination by the method of neutron impulse
International Nuclear Information System (INIS)
Flores Calderon, J.E.
1982-01-01
Experiments with non-stationary neutron transport in large cavity moderators (l>>Σsub(tr) -1 ) (where l is the characteristic cavity length and Σsub(tr) -1 the macroscopic transport section of the moderator) led to the method reported in this study which, based on neutron impulses for determining albedo of thermal neutrons, gave a precision greater by an order of magnitude over previous methods. A sufficient time interval after introduction of the neutron flux into the moderator chamber decreased exponentially the decay constant L, which was itself related to albedo by a function called f. Numerical calculations of albedo were assisted. (author)
Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.
2017-10-01
Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.
''In situ'' electronic testing method of a neutron detector performance
International Nuclear Information System (INIS)
Gonzalez, J.M.; Levai, F.
1987-01-01
The method allows detection of any important change in the electrical characteristics of a neutron sensor channel. It checks the response signal produced by an electronic detector circuit when a pulse generator is connected as input signal in the high voltage supply. The electronic circuit compares the detector capacitance value, previously measured, against a reference value, which is adjusted in a window type comparator electronic circuit to detect any important degrading condition of the capacitance value in a detector-cable system. The ''in-situ'' electronic testing method of neutron detector performance has been verified in a laboratory atmosphere to be a potential method to detect any significant change in the capacitance value of a nuclear sensor and its connecting cable, also checking: detector disconnections, cable disconnections, length changes of the connecting cable, electric short-opened circuits in the sensor channel, and any electrical trouble in the detector-connector-cable system. The experimental practices were carried out by simulation of several electric changes in a nuclear sensor-cable system from a linear D.C. channel which measures reactor power during nuclear reactor operation. It was made at the Training Reactor Electronic Laboratory. The results and conclusions obtained at the Laboratory were proved, satisfactorily, in the Electronic Instrumentation of Budapest Technical University Training Reactor, Hungary
Plant model of KIPT neutron source facility simulator
International Nuclear Information System (INIS)
Cao, Yan; Wei, Thomas Y.; Grelle, Austin L.; Gohar, Yousry
2016-01-01
Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine are collaborating on constructing a neutron source facility at KIPT, Kharkov, Ukraine. The facility has 100-kW electron beam driving a subcritical assembly (SCA). The electron beam interacts with a natural uranium target or a tungsten target to generate neutrons, and deposits its power in the target zone. The total fission power generated in SCA is about 300 kW. Two primary cooling loops are designed to remove 100-kW and 300-kW from the target zone and the SCA, respectively. A secondary cooling system is coupled with the primary cooling system to dispose of the generated heat outside the facility buildings to the atmosphere. In addition, the electron accelerator has a low efficiency for generating the electron beam, which uses another secondary cooling loop to remove the generated heat from the accelerator primary cooling loop. One of the main functions the KIPT neutron source facility is to train young nuclear specialists; therefore, ANL has developed the KIPT Neutron Source Facility Simulator for this function. In this simulator, a Plant Control System and a Plant Protection System were developed to perform proper control and to provide automatic protection against unsafe and improper operation of the facility during the steady-state and the transient states using a facility plant model. This report focuses on describing the physics of the plant model and provides several test cases to demonstrate its capabilities. The plant facility model uses the PYTHON script language. It is consistent with the computer language of the plant control system. It is easy to integrate with the simulator without an additional interface, and it is able to simulate the transients of the cooling systems with system control variables changing on real-time.
Plant model of KIPT neutron source facility simulator
Energy Technology Data Exchange (ETDEWEB)
Cao, Yan [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, Thomas Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Grelle, Austin L. [Argonne National Lab. (ANL), Argonne, IL (United States); Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-02-01
Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine are collaborating on constructing a neutron source facility at KIPT, Kharkov, Ukraine. The facility has 100-kW electron beam driving a subcritical assembly (SCA). The electron beam interacts with a natural uranium target or a tungsten target to generate neutrons, and deposits its power in the target zone. The total fission power generated in SCA is about 300 kW. Two primary cooling loops are designed to remove 100-kW and 300-kW from the target zone and the SCA, respectively. A secondary cooling system is coupled with the primary cooling system to dispose of the generated heat outside the facility buildings to the atmosphere. In addition, the electron accelerator has a low efficiency for generating the electron beam, which uses another secondary cooling loop to remove the generated heat from the accelerator primary cooling loop. One of the main functions the KIPT neutron source facility is to train young nuclear specialists; therefore, ANL has developed the KIPT Neutron Source Facility Simulator for this function. In this simulator, a Plant Control System and a Plant Protection System were developed to perform proper control and to provide automatic protection against unsafe and improper operation of the facility during the steady-state and the transient states using a facility plant model. This report focuses on describing the physics of the plant model and provides several test cases to demonstrate its capabilities. The plant facility model uses the PYTHON script language. It is consistent with the computer language of the plant control system. It is easy to integrate with the simulator without an additional interface, and it is able to simulate the transients of the cooling systems with system control variables changing on real-time.
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)
Method and apparatus for epithermal neutron porosity well logging
International Nuclear Information System (INIS)
Hertzog, R.C.; Loomis, W.A.; Wraight, P.
1991-01-01
This patent describes a method for investigating the porosity of a subsurface earth formation surrounding a borehole. It comprises repetitively irradiating the borehole and earth formation with discrete bursts of high energy neutrons from a neutron source, which neutrons interact with nuclei of the materials in the borehole and the formation to produce therein populations of epithermal neutrons; detecting the populations of epithermal neutrons at near and far locations in the borehole spaced apart longitudinally by different distances from the neutron source; generating count signals indicative of the magnitudes of the detected epithermal neutron populations at the respective near and far locations; detecting the decay of the epithermal neutron populations following the neutron bursts at least at one location in the borehole and generating signals representative thereof; deriving from the decay signals a signal indicative of the slowing down time of epithermal neutrons in the formation of the at least one location; and deriving from the near and far count signals and the slowing down time signal a measurement signal representative of the porosity of the formation surrounding the borehole inherently compensated for the effects of tool standoff on the responses of the logging tool
International Nuclear Information System (INIS)
Kawai, T.; Tasaki, S.; Ebisawa, T.; Hino, M.; Yamazaki, D.; Achiwa, N.
1999-01-01
Complete text of publication follows. A non-dispersive method is proposed for measuring the longitudinal coherence length of a neutron using a high frequency cold neutron pulser (hf-CNP) placed between two multilayer spin splitters (MSS) which composes the cold neutron spin interferometer. Two spin eigenstates of a neutron polarized x-y plane are split non-dispersively and longitudinally in time by the hf-CNP which could reflect two components alternatively in time. The reduction of the visibility of interference fringes after being superposed by the second MSS is measured as a function of the frequency of the pulser by TOF method. From the zero visibility point obtained by extrapolation one could obtain the longitudinal coherence length of the neutron. (author)
Method and apparatus for determining the dose value of neutrons
International Nuclear Information System (INIS)
Burgkhardt, B.; Piesch, E.
1976-01-01
A method is provided for determining the dose value of neutrons leaving a body as thermal and intermediate neutrons after having been scattered in the body. A first dose value of thermal and intermediate neutrons is detected on the surface of the body by means of a first detector for neutrons which is shielded against thermal and intermediate neutrons not emerging from the body. A second detector is used to measure a second dose value of the thermal and intermediate neutrons not emerging from the body. A first correction factor based on the first and second values is obtained from a calibration diagram and is applied to the first dose value to determine a first corrected first dose value. 21 Claims, 6 Drawing Figures
International Nuclear Information System (INIS)
Azraf Azman; Shahrir Abdullah; Mohd Rizal Mamat
2011-01-01
The cryogenic system for cooling Beryllium filter utilizing liquid nitrogen was designed, fabricated, tested and installed at SANS instrument of TRIGA MARK II PUSPATI research reactor. A computational fluid dynamics (CFD) modeling was used to predict the cooling performance of the beryllium for optimization of neutron beam resolution and transmission. This paper presents the transient CFD results of temperature distributions via the thermal link to the beryllium and simulation of heat flux. The simulation data are also compared with the experimental results for the cooling time and distribution to the beryllium. (author)
Instrumental neutron activation analysis - a routine method
International Nuclear Information System (INIS)
Bruin, M. de.
1983-01-01
This thesis describes the way in which at IRI instrumental neutron activation analysis (INAA) has been developed into an automated system for routine analysis. The basis of this work are 20 publications describing the development of INAA since 1968. (Auth.)
Method of manufacturing neutron protection materials
Energy Technology Data Exchange (ETDEWEB)
Kakibana, Hidetake; Okamoto, Masazane; Fujii, Yasuhiko; Koguchi, Noboru; Takesute, Morito; Miyamatsu, Tokuhisa
1985-06-22
To obtain protection materials easily moldable, flexible and capable of minimizing the workers' neutron exposure dose, a fine fiberous assembly is prepared by dispersing compounds of atoms having neutron absorbing performance such as Li or B, for example, finely powderous compounds of LiF or /sup 6/LiF into a solution of spinnable polymer, particularly, polyolefin polymer such as polyethylene in CH/sub 2/Cl and then flash spinning them. The fine fibers are fabricated into mat-like material, blankets, cloths and the likes for use in neutron exposure protection. In the case of neutron irradiation therapy, protection materials of reduced weight, flexible and giving preferred contact with human body can be obtained with ease for protecting the regions other than the lesion area.
Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code
Faghihi, F.; Mehdizadeh, S.; Hadad, K.
Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.
Current trends in methods for neutron diffusion calculations
International Nuclear Information System (INIS)
Adams, C.H.
1977-01-01
Current work and trends in the application of neutron diffusion theory to reactor design and analysis are reviewed. Specific topics covered include finite-difference methods, synthesis methods, nodal calculations, finite-elements and perturbation theory
Energy Technology Data Exchange (ETDEWEB)
Fong, G; Kapadia, A [Carl E Ravin Advanced Imaging Laboratories, Durham, North Carolina (United States)
2016-06-15
Purpose: To optimize collimation and shielding for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT). Methods: Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm{sup 3} polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source-side and target-side openings) and aperture placements (13–39 cm from source-side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system. Results: Analysis of primary beam indicated the beam-width is linear to the aperture size. Increasing source-side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings. Conclusion: The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan
International Nuclear Information System (INIS)
Fong, G; Kapadia, A
2016-01-01
Purpose: To optimize collimation and shielding for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT). Methods: Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm"3 polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source-side and target-side openings) and aperture placements (13–39 cm from source-side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system. Results: Analysis of primary beam indicated the beam-width is linear to the aperture size. Increasing source-side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings. Conclusion: The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan time
Method to produce a neutron shielding
International Nuclear Information System (INIS)
Merkle, H.J.
1978-01-01
The neutron shielding for armoured vehicles consists of preshaped plastic plates which are coated on the armoured vehicle walls by conversion of the thermoplast. Suitable plastics or thermoplasts are PVC, PVC acetate, or mixtures of these, into which more than 50% B, B 4 C, or BN is embedded. The colour of the shielding may be determined by the choice of the neutron absorber, e.g. a white colour for BN. The plates are produced using an extruder or calender. (DG) [de
Gravitational waveforms for neutron star binaries from binary black hole simulations
Barkett, Kevin; Scheel, Mark; Haas, Roland; Ott, Christian; Bernuzzi, Sebastiano; Brown, Duncan; Szilagyi, Bela; Kaplan, Jeffrey; Lippuner, Jonas; Muhlberger, Curran; Foucart, Francois; Duez, Matthew
2016-03-01
Gravitational waves from binary neutron star (BNS) and black-hole/neutron star (BHNS) inspirals are primary sources for detection by the Advanced Laser Interferometer Gravitational-Wave Observatory. The tidal forces acting on the neutron stars induce changes in the phase evolution of the gravitational waveform, and these changes can be used to constrain the nuclear equation of state. Current methods of generating BNS and BHNS waveforms rely on either computationally challenging full 3D hydrodynamical simulations or approximate analytic solutions. We introduce a new method for computing inspiral waveforms for BNS/BHNS systems by adding the post-Newtonian (PN) tidal effects to full numerical simulations of binary black holes (BBHs), effectively replacing the non-tidal terms in the PN expansion with BBH results. Comparing a waveform generated with this method against a full hydrodynamical simulation of a BNS inspiral yields a phase difference of < 1 radian over ~ 15 orbits. The numerical phase accuracy required of BNS simulations to measure the accuracy of the method we present here is estimated as a function of the tidal deformability parameter λ.
Simulating Neutronic Core Parameters in a Research and Test Reactor
International Nuclear Information System (INIS)
Selim, H.K.; Amin, E.A.; Koutb, M.E.
2011-01-01
The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload
Deficiency in Monte Carlo simulations of coupled neutron-gamma-ray fields
Maleka, Peane P.; Maucec, Marko; de Meijer, Robert J.
2011-01-01
The deficiency in Monte Carlo simulations of coupled neutron-gamma-ray field was investigated by benchmarking two simulation codes with experimental data. Simulations showed better correspondence with the experimental data for gamma-ray transport only. In simulations, the neutron interactions with
International Nuclear Information System (INIS)
Kroc, T.K.
2009-01-01
No fast neutron therapy facility has been built with optimized beam quality based on a thorough understanding of the neutron spectrum and its resulting biological effectiveness. A study has been initiated to provide the information necessary for such an optimization. Monte Carlo studies will be used to simulate neutron energy spectra and LET spectra. These studies will be bench-marked with data taken at existing fast neutron therapy facilities. Results will also be compared with radiobiological studies to further support beam quality ptimization. These simulations, anchored by this data, will then be used to determine what parameters might be optimized to take full advantage of the unique LET properties of fast neutron beams. This paper will present preliminary work in generating energy and LET spectra for the Fermilab fast neutron therapy facility.
Virtual experiments: the ultimate aim of neutron ray-tracing simulations
Czech Academy of Sciences Publication Activity Database
Lefmann, K.; Willendrup, P.K.; Šaroun, Jan
2008-01-01
Roč. 16, 3 & 4 (2008), s. 97-111 ISSN 1023-8166 Institutional research plan: CEZ:AV0Z10480505 Keywords : Monte Carlo simulations * neutron scattering * neutron instrumentation Subject RIV: BM - Solid Matter Physics ; Magnetism
Use of high voltage electron microscope to simulate radiation damage by neutrons
International Nuclear Information System (INIS)
Mayer, R.M.
1976-01-01
The use of the high voltage electron microscope to simulate radiation damage by neutrons is briefly reviewed. This information is important in explaining how alloying affects void formation during neutron irradiation
Energy Technology Data Exchange (ETDEWEB)
Wang, Hu; Zou, Yubin, E-mail: zouyubin@pku.edu.cn; Wen, Weiwei; Lu, Yuanrong; Guo, Zhiyu
2016-07-01
Peking University Neutron Imaging Facility (PKUNIFTY) works on an accelerator–based neutron source with a repetition period of 10 ms and pulse duration of 0.4 ms, which has a rather low Cd ratio. To improve the effective Cd ratio and thus improve the detection capability of the facility, energy-filtering neutron imaging was realized with the intensified CCD camera and time-of-flight (TOF) method. Time structure of the pulsed neutron source was firstly simulated with Geant4, and the simulation result was evaluated with experiment. Both simulation and experiment results indicated that fast neutrons and epithermal neutrons were concentrated in the first 0.8 ms of each pulse period; meanwhile in the period of 0.8–2.0 ms only thermal neutrons existed. Based on this result, neutron images with and without energy filtering were acquired respectively, and it showed that detection capability of PKUNIFTY was improved with setting the exposure interval as 0.8–2.0 ms, especially for materials with strong moderating capability.
Methods for the neutronic design of a Supersara experimental loop
International Nuclear Information System (INIS)
Casali, F.; Cepraga, D.
1982-01-01
This paper describes a method for the neutronic design of experimental loops irradiated in D 2 O experimental reactors, like Essor. The calculation approach concerns the definition of a Weigner-Seitz cell where the loop under examination be subjected to the same neutronic conditions as in the actual reactor
Geant4 simulations of NIST beam neutron lifetime experiment
Valete, Daniel; Crawford, Bret; BL2 Collaboration Collaboration
2017-09-01
A free neutron is unstable and its decay is described by the Standard Model as the transformation of a down quark into an up quark through the weak interaction. Precise measurements of the neutron lifetime test the validity of the theory of the weak interaction and provide useful information for the predictions of the theory of Big Bang nucleosynthesis of the primordial helium abundance in the universe and the number of different types of light neutrinos Nν. The predominant experimental methods for determination of the neutron lifetime are commonly called `beam' and `bottle' methods, and the most recent uses of each method do not agree with each other within their stated uncertainties. An improved experiment of the beam technique, which uses magnetic and electric fields to trap and guide the decay protons of a beam of cold neutrons to a detector, is in progress at the National Institute of Standards and Technology, Gaithersburg, MD with a precision goal of 0.1. I acknowledge the support of the Cross-Diciplinary Institute at Gettysburg College.
Neutron absorbers and methods of forming at least a portion of a neutron absorber
Energy Technology Data Exchange (ETDEWEB)
Guillen, Donna P; Porter, Douglas L; Swank, W David; Erickson, Arnold W
2014-12-02
Methods of forming at least a portion of a neutron absorber include combining a first material and a second material to form a compound, reducing the compound into a plurality of particles, mixing the plurality of particles with a third material, and pressing the mixture of the plurality of particles and the third material. One or more components of neutron absorbers may be formed by such methods. Neutron absorbers may include a composite material including an intermetallic compound comprising hafnium aluminide and a matrix material comprising pure aluminum.
Finite size effects in neutron star and nuclear matter simulations
Energy Technology Data Exchange (ETDEWEB)
Giménez Molinelli, P.A., E-mail: pagm@df.uba.ar; Dorso, C.O.
2015-01-15
In this work we study molecular dynamics simulations of symmetric nuclear and neutron star matter using a semi-classical nucleon interaction model. Our aim is to gain insight on the nature of the so-called “finite size effects”, unavoidable in this kind of simulations, and to understand what they actually affect. To do so, we explore different geometries for the periodic boundary conditions imposed on the simulation cell: cube, hexagonal prism and truncated octahedron. For nuclear matter simulations we show that, at sub-saturation densities and low temperatures, the solutions are non-homogeneous structures reminiscent of the “nuclear pasta” phases expected in neutron star matter simulations, but only one structure per cell and shaped by specific artificial aspects of the simulations—for the same physical conditions (i.e. number density and temperature) different cells yield different solutions. The particular shape of the solution at low enough temperature and a given density can be predicted analytically by surface minimization. We also show that even if this behavior is due to the imposition of periodic boundary conditions on finite systems, this does not mean that it vanishes for very large systems, and it is actually independent of the system size. We conclude that, for nuclear matter simulations, the cells' size sets the only characteristic length scale for the inhomogeneities, and the geometry of the periodic cell determines the shape of those inhomogeneities. To model neutron star matter we add a screened Coulomb interaction between protons, and perform simulations in the three cell geometries. Our simulations indeed produce the well known nuclear pasta, with (in most cases) several structures per cell. However, we find that for systems not too large results are affected by finite size in different ways depending on the geometry of the cell. In particular, at the same certain physical conditions and system size, the hexagonal prism yields a
Neutron spectrum determination by activation method in fast neutron fields at the RB reactor
International Nuclear Information System (INIS)
Sokcic-Kostic, M.; Pesic, M.; Antic, D.
1994-01-01
The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (author)
Neutron spectrum determination by activation method in fast neutron fields at the RB reactors
International Nuclear Information System (INIS)
Sokcic-Kostic, M.S.; Pesic, M.P.; Antic, D.P.
1994-01-01
The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (authors). 7 refs., 3 tabs
International Nuclear Information System (INIS)
Ánchel, F.; Barrachina, T.; Miró, R.; Verdú, G.; Juanas, J.; Macián-Juan, R.
2012-01-01
Highlights: ► Best-estimate codes are affected by the uncertainty in the methods and the models. ► Influence of the uncertainty in the macroscopic cross-sections in a BWR and PWR RIA accidents analysis. ► The fast diffusion coefficient, the scattering cross section and both fission cross sections are the most influential factors. ► The absorption cross sections very little influence. ► Using a normal pdf the results are more “conservative” comparing the power peak reached with uncertainty quantified with a uniform pdf. - Abstract: The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether–Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.
Simulation of processes in a SSNTD exposed by monoenergetic neutrons
Meyer, P.; Fromm, M.; Groetz, J. E.; Torrealba, F.; Chambaudet, A.
1998-03-01
The nuclear track technique is based on the registering of latent track ions in a solid state dielectric material (Solid State Nuclear Track Detector) which will be chemically etched in order to be observed and analysed using microscopic analysis tools. The purpose of this study is to observe how a polymeric detector irradiated by a neutron fluence can explain the intensity of ionizing radiations to which structures, electronic components or living organism are exposed. To do this, it is necessary to take into account on the one hand the production of ionizing particles in the detector, and on the other hand the efficacy of their detection. The model we propose must be able to link the number of etched tracks and their parameters to the neutron fluence. The use of both a Monte Carlo code which determines the number of recoil nuclei and a computed model of etched track parameters calculations is needed. The Monte Carlo code performs a simulation of neutron-nucleus elastic collision in the detector. We compute the number of recoil nuclei produced between the surface and a depth of 20 μm in the detector. The computer code assigns six pseudo-random numbers ( xE, zE, a, b, c, α) to each neutron having an energy EN and an angle of incidence and give the results of calculations. We used statistical tests (the moments of the origin of order n and of the average of order n, the χ2 test, the gap test, the Pearson test, the run-up run-down test and the serial test) to check the quality of our pseudo-random number generator. Finally, we will compare the calculated results of the response of the SSNTD (tracks cm -2) with the experimental ones to validate the simulation.
International Nuclear Information System (INIS)
Daniehl', A.V.; Dushin, V.N.
1987-01-01
The methods for simulation of neutron transport with Z<20 MeV used in the SITHA (simulation transport hadron) program, the original library of group microconstants (175 groups) with subgroup description of resonance range and a set of programs for its creation are described. The results of a number of integral experiments are discussed
New developments in the McStas neutron instrument simulation package
International Nuclear Information System (INIS)
Willendrup, P K; Knudsen, E B; Klinkby, E; Nielsen, T; Farhi, E; Filges, U; Lefmann, K
2014-01-01
The McStas neutron ray-tracing software package is a versatile tool for building accurate simulators of neutron scattering instruments at reactors, short- and long-pulsed spallation sources such as the European Spallation Source. McStas is extensively used for design and optimization of instruments, virtual experiments, data analysis and user training. McStas was founded as a scientific, open-source collaborative code in 1997. This contribution presents the project at its current state and gives an overview of the main new developments in McStas 2.0 (December 2012) and McStas 2.1 (expected fall 2013), including many new components, component parameter uniformisation, partial loss of backward compatibility, updated source brilliance descriptions, developments toward new tools and user interfaces, web interfaces and a new method for estimating beam losses and background from neutron optics.
Development of moderated neutron calibration fields simulating workplaces of MOX fuel facilities
International Nuclear Information System (INIS)
Tsujimura, Norio; Yoshida, Tadayoshi; Takada, Chie
2005-01-01
It is important for the MOX fuel facilities to control neutrons produced by the spontaneous fission of plutonium isotopes and those from (α,n) reactions between 18 O and α particles emitted by 238 Pu. Neutron dose meters should be calibrated for measuring these neutrons. We have developed moderated-neutron calibration fields employing a 252 Cf neutron source and moderators mainly for the characteristics evaluation and the calibration of neutron detectors used in MOX fuel facilities. Neutron energy spectrum can be adjusted by changing the position of the 252 Cf neutron source and combining different moderators to simulate the neutron field of the MOX fuel facility. This performance is realized owing to using an existing neutron irradiation room. (K. Yoshida)
Identification of materials by an advanced neutronic method
International Nuclear Information System (INIS)
El Kanawati, W.
2011-01-01
The EURITRACK inspection system, based on the associated particle technique, aims at detecting explosives and narcotics in cargo containers with 14 MeV neutrons produced by the D(T,γ)n reaction. Alpha particle and neutron are emitted almost back to back. Reactions induced by fast neutrons produce gamma rays which are detected in coincidence with the alpha particle to determine the neutron direction. Neutron time-of-flight allows to determine gamma-ray origin inside the container. Information concerning material composition is obtained by unfolding the gamma spectrum into elemental signatures using a database of elemental spectra (C, O, N, Fe...). Carbon, oxygen, and nitrogen count ratios are converted into chemical proportions to distinguish illicit and benign organic materials. Conversion factors based on Monte Carlo simulations have been calculated and validated experimentally, taking into account neutron slowing down and photon attenuation in cargo materials. Application to the elemental characterisation of radioactive wastes is also studied by numerical simulation, with shields and collimators to limit the background due to waste radiations. (author)
Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar
Energy Technology Data Exchange (ETDEWEB)
Lee, K. M.; Sum, G. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified.
Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar
International Nuclear Information System (INIS)
Lee, K. M.; Sum, G. M.
2016-01-01
Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified
Physico-technical progress in neutron-capture therapy method
International Nuclear Information System (INIS)
Kanda, Keiji; Furubayashi, Toru; Aoki, Kazuhiko
1985-01-01
This paper describes mainly development studies on the determination method of in vivo 10 B for the purpose of employment for neutron capture therapy for malignant melanoma and other tumors. To darify the efficacy of the neutron capture therapy, it is necessary to determine 10 B concentration in the diseased part. This study aimed at in vivo 10 B concention determination in living sample to the level of ppm order with 10 % of analytical error within 1 hour, and these determination conditions were satified by prompt γ-ray (478 keV) determination of 10 B (n, αγ) 7 Li reaction. This method required no sample pretreatment. Further, data normalization by γ-ray of H(n, γ)D reaction permitted no disturbance by sample shape or size. Lower limit of detection of the proposed method was estimated in terms of measuring time and statistical error by the equations of 10 B concentration and error analysis derived by the authors. As for the effect of prompt γ-rays of 23 Na(n, γ) 24 Na and 6 Li(n, γ) 7 Li reactions, it was clarified that the former showed no disturbance but some correction was necessary in case of less than 0.1 g of smaple size owing to the latter reaction. In vivo sample determination showed the proposed method was practical. In this paper some results of phantom experiment for in vivo non-destructive 10 B measurement and related simulation calculation, and examination of effect of (γ, n) reaction in heavy water of biomedical irradiation equipment on radiation quality were also described. (Takagi, S.)
Principles and methods of neutron interferometry
International Nuclear Information System (INIS)
Bonse, U.
1978-01-01
The merits of Angstrom range interferometry with neutrons are briefly outlined. The energy (wavelength) range which is accessible with the triple Laue case (LLL) crystal interferometer is estimated, assuming a neutron source with flux characteristics similar to that of the HFR at Grenoble. It appears that a range in E from roughly 2.3 meV to 8.2eV (lambda approximatly equal to 6A to 0.1A) can be covered with LLL interferometers manufactured with presently available perfect crystals of silicon. Within this range there exists a number of scattering resonances that it seems worth while to investigate interferometrically. The attainable resolution ΔE/E is estimated to be at least 10 -3 for E -2 above. The essentials of zero absorption Bragg diffraction optics of the neutron LLL interferometer are described. Virtues and weaknesses of different LLL geometries are discussed. The influence of geometrical abberrations, strain and position instabilities are surveyed. Aspects of coherent scattering length measurements and of neutron phase topography are discussed
Pulsed neutron method for diffusion, slowing down, and reactivity measurements
International Nuclear Information System (INIS)
Sjoestrand, N.G.
1985-01-01
An outline is given on the principles of the pulsed neutron method for the determination of thermal neutron diffusion parameters, for slowing-down time measurements, and for reactivity determinations. The historical development is sketched from the breakthrough in the middle of the nineteen fifties and the usefulness and limitations of the method are discussed. The importance for the present understanding of neutron slowing-down, thermalization and diffusion are point out. Examples are given of its recent use for e.g. absorption cross section measurements and for the study of the properties of heterogeneous systems
Bioassay method for Uranium in urine by Delay Neutron counting
International Nuclear Information System (INIS)
Suratman; Purwanto; Sukarman-Aminjoyo
1996-01-01
A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 μg
International Nuclear Information System (INIS)
Tian Dongfeng; Ho Yukun; Yang Fujia
2001-01-01
The SWINPC integral experiment on neutron multiplication in bulk beryllium showed that there were marked discrepancies between experimental data and calculated values with the ENDF/B-VI data. The calculated values become higher than experimental ones as the sample thickness increases. Several works had been devoted to find problems existing in the experiment. This paper discusses the neutron reflection effect on the total absorption detector method which was used in the experiment to measure the neutron leakage from samples. One systematic correction is suggested to make the experimental values agree with the calculated ones with the ENDF/B-VI data within experimental errors. (author)
Means and method for controlling the neutron output of a neutron generator tube
International Nuclear Information System (INIS)
Langford, O.M.; Peelman, H.E.
1978-01-01
Means and method are described for energizing and regulating a neutron generator tube having a target, an ion source and a replenisher. It providing a negative high voltage to the target and monitoring the target current. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordence with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current
Means and method for controlling the neutron output of a neutron generator tube
International Nuclear Information System (INIS)
1977-01-01
A means and method for energizing and regulating a neutron generator tube is described. It has a target, an ion source and a replenisher. A negative high voltage is applied to the target and the target current monitored. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordance with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current. (C.F.)
System to detect nuclear materials by active neutron method
International Nuclear Information System (INIS)
Koroev, M.; Korolev, Yu.; Lopatin, Yu.; Filonov, V.
1999-01-01
The report presents the results of the development of the system to detect nuclear materials by active neutron method measuring delayed neutrons. As the neutron source the neutron generator was used. The neutron generator was controlled by the system. The detectors were developed on the base of the helium-3 counters. Each detector consist of 6 counters. Using a number of such detectors it is possible to verify materials stored in different geometry. There is an spectrometric scintillator detector in the system which gives an additional functional ability to the system. The system could be used to estimate the nuclear materials in waste, to detect the unauthorized transfer of the nuclear materials, to estimate the material in tubes [ru
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
Double difference method in deep inelastic neutron scattering on the VESUVIO spectrometer
International Nuclear Information System (INIS)
Andreani, C.; Colognesi, D.; Degiorgi, E.; Filabozzi, A.; Nardone, M.; Pace, E.; Pietropaolo, A.; Senesi, R.
2003-01-01
The principles of the Double Difference (DD) method, applied to the neutron spectrometer VESUVIO, are discussed. VESUVIO, an inverse geometry spectrometer operating at the ISIS pulsed neutron source in the eV energy region, has been specifically designed to measure the single particle dynamical properties in condensed matter. The width of the nuclear resonance of the absorbing filter, used for the neutron energy analysis, provides the most important contribution to the energy resolution of the inverse geometry instruments. In this paper, the DD method, which is based on a linear combination of two measurements recorded with filter foils of the same resonance material but of different thickness, is shown to improve significantly the instrumental energy resolution, as compared with the Single Difference (SD) method. The asymptotic response functions, derived through Monte-Carlo simulations for polycrystalline Pb and ZrH 2 samples, are analysed in both DD and SD methods, and compared with the experimental ones for Pb sample. The response functions have been modelled for two distinct experimental configurations of the VESUVIO spectrometer, employing 6 Li-glass neutron detectors and NaI γ detectors revealing the γ-ray cascade from the (n,γ) reaction, respectively. The DD method appears to be an effective experimental procedure for Deep Inelastic Neutron Scattering measurements on VESUVIO spectrometer, since it reduces the experimental resolution of the instrument in both 6 Li-glass neutron detector and γ detector configurations
Double difference method in deep inelastic neutron scattering on the VESUVIO spectrometer
Andreani, C.; Colognesi, D.; Degiorgi, E.; Filabozzi, A.; Nardone, M.; Pace, E.; Pietropaolo, A.; Senesi, R.
2003-02-01
The principles of the Double Difference (DD) method, applied to the neutron spectrometer VESUVIO, are discussed. VESUVIO, an inverse geometry spectrometer operating at the ISIS pulsed neutron source in the eV energy region, has been specifically designed to measure the single particle dynamical properties in condensed matter. The width of the nuclear resonance of the absorbing filter, used for the neutron energy analysis, provides the most important contribution to the energy resolution of the inverse geometry instruments. In this paper, the DD method, which is based on a linear combination of two measurements recorded with filter foils of the same resonance material but of different thickness, is shown to improve significantly the instrumental energy resolution, as compared with the Single Difference (SD) method. The asymptotic response functions, derived through Monte-Carlo simulations for polycrystalline Pb and ZrH 2 samples, are analysed in both DD and SD methods, and compared with the experimental ones for Pb sample. The response functions have been modelled for two distinct experimental configurations of the VESUVIO spectrometer, employing 6Li-glass neutron detectors and NaI γ detectors revealing the γ-ray cascade from the ( n,γ) reaction, respectively. The DD method appears to be an effective experimental procedure for Deep Inelastic Neutron Scattering measurements on VESUVIO spectrometer, since it reduces the experimental resolution of the instrument in both 6Li-glass neutron detector and γ detector configurations.
Double difference method in deep inelastic neutron scattering on the VESUVIO spectrometer
Energy Technology Data Exchange (ETDEWEB)
Andreani, C.; Colognesi, D.; Degiorgi, E.; Filabozzi, A.; Nardone, M.; Pace, E.; Pietropaolo, A. E-mail: antonino.pietropaolo@roma2.infn.it; Senesi, R
2003-02-01
The principles of the Double Difference (DD) method, applied to the neutron spectrometer VESUVIO, are discussed. VESUVIO, an inverse geometry spectrometer operating at the ISIS pulsed neutron source in the eV energy region, has been specifically designed to measure the single particle dynamical properties in condensed matter. The width of the nuclear resonance of the absorbing filter, used for the neutron energy analysis, provides the most important contribution to the energy resolution of the inverse geometry instruments. In this paper, the DD method, which is based on a linear combination of two measurements recorded with filter foils of the same resonance material but of different thickness, is shown to improve significantly the instrumental energy resolution, as compared with the Single Difference (SD) method. The asymptotic response functions, derived through Monte-Carlo simulations for polycrystalline Pb and ZrH{sub 2} samples, are analysed in both DD and SD methods, and compared with the experimental ones for Pb sample. The response functions have been modelled for two distinct experimental configurations of the VESUVIO spectrometer, employing {sup 6}Li-glass neutron detectors and NaI {gamma} detectors revealing the {gamma}-ray cascade from the (n,{gamma}) reaction, respectively. The DD method appears to be an effective experimental procedure for Deep Inelastic Neutron Scattering measurements on VESUVIO spectrometer, since it reduces the experimental resolution of the instrument in both {sup 6}Li-glass neutron detector and {gamma} detector configurations.
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.
Benchmark of the neutronic model used in Maanshan compact simulator
International Nuclear Information System (INIS)
Hu, C.-H.; Gone, J.-K.; Ko, H.-T.
2004-01-01
The Maanshan compact simulator has adopted a three dimensional kinetic model CONcERT, which was developed by GP International Inc. (GPI) in 1991 for real-time neutronic analysis. Maanshan Nuclear Power Plant utilizes a Westinghouse nuclear steam supply system with three-loop pressurized water reactor. There are 157 fuel assemblies and 52 full-length Rod Cluster Control Assemblies in the reactor core. The control of excess reactivity and power peaking is provided by soluble boron in moderator and burnable absorber rods in fuel assemblies. The neutronic model of CONcERT is based on solving a modified time-dependent two-group diffusion equations coupled to the equations of six-group delayed neutron precursor concentrations. The validation of CONcERT for the Maanshan plant is separated into two groups. The first group compared (1) boron endpoints for different control bank inserted conditions, (2) control rod differential and integral worths and (3) temperature coefficients with the measurements in the Low Power Physical Test (LPPT). The second group compared critical boron concentration and power distribution in high power condition with the measurements. In addition, xenon and samarium equilibrium worths at different power levels as well as the time dependent changes of their worth after the reactor scram are illustrated. (author)
Study of spectral response of a neutron filter. Design of a method to adjust spectra
International Nuclear Information System (INIS)
Colomb-Dolci, F.
1999-02-01
The first part of this thesis describes an experimental method which intends to determine a neutron spectrum in the epithermal range [1 eV -10 keV]. Based on measurements of reaction rates provided by activation foils, it gives flux level in each energy range corresponding to each probe. This method can be used in any reactor location or in a neutron beam. It can determine scepter on eight energy groups, five groups in the epithermal range. The second part of this thesis presents a study of an epithermal neutron beam design, in the frame of Neutron Capture Therapy. A beam tube was specially built to test filters made up of different materials. Its geometry was designed to favour epithermal neutron crossing and to cut thermal and fast neutrons. A code scheme was validated to simulate the device response with a Monte Carlo code. Measurements were made at ISIS reactor and experimental spectra were compared to calculated ones. This validated code scheme was used to simulate different materials usable as shields in the tube. A study of these shields is presented at the end of this thesis. (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
Review of experimental methods for evaluating effective delayed neutron fraction
Energy Technology Data Exchange (ETDEWEB)
Yamane, Yoshihiro [Nagoya Univ. (Japan). School of Engineering
1997-03-01
The International Effective Delayed Neutron Fraction ({beta}{sub eff}) Benchmark Experiments have been carried out at the Fast Critical Assembly of Japan Atomic Energy Research Institute since 1995. Researchers from six countries, namely France, Italy, Russia, U.S.A., Korea, and Japan, participate in this FCA project. Each team makes use of each experimental method, such as Frequency Method, Rossi-{alpha} Method, Nelson Number Method, Cf Neutron Source Method, and Covariance Method. In this report these experimental methods are reviewed. (author)
Simulations of a PSD Plastic Neutron Collar for Assaying Fresh Fuel
Energy Technology Data Exchange (ETDEWEB)
Hausladen, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Newby, Jason [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2016-11-01
The potential performance of a notional active coincidence collar for assaying uranium fuel based on segmented detectors constructed from the new PSD plastic fast organic scintillator with pulse shape discrimination capability was investigated in simulation. Like the International Atomic Energy Agency's present Uranium Neutron Collar for LEU (UNCL), the PSD plastic collar would also function by stimulating fission in the ^{235}U content of the fuel with a moderated ^{241}Am/Li neutron source and detecting instances of induced fission via neutron coincidence counting. In contrast to the moderated detectors of the UNCL, the fast time scale of detection in the scintillator eliminates statistical errors due to accidental coincidences that limit the performance of the UNCL. However, the potential to detect a single neutron multiple times historically has been one of the properties of organic scintillator detectors that has prevented their adoption for international safeguards applications. Consequently, as part of the analysis of simulated data, a method was developed by which true neutron-neutron coincidences can be distinguished from inter-detector scatter that takes advantage of the position and timing resolution of segmented detectors. Then, the performance of the notional simulated coincidence collar was evaluated for assaying a variety of fresh fuels, including some containing burnable poisons and partial defects. In these simulations, particular attention was paid to the analysis of fast mode measurements. In fast mode, a Cd liner is placed inside the collar to shield the fuel from the interrogating source and detector moderators, thereby eliminating the thermal neutron flux that is most sensitive to the presence of burnable poisons that are ubiquitous in modern nuclear fuels. The simulations indicate that the predicted precision of fast mode measurements is similar to what can be achieved by the present UNCL in thermal mode. For example, the
Neutron diffraction utilizing the T-O-F method
Energy Technology Data Exchange (ETDEWEB)
Niimura, N [Tohoku Univ., Sendai (Japan). Lab. of Nuclear Science
1974-12-01
Characteristic features of the TOF (time of flight) neutron diffraction are summarized. In this method, i) all the reciprocal points on the rod passing through the origin in the reciprocal space can be scanned by each burst of white neutrons, ii) it is easy to measure high index reflections at the large scattering angle, iii) each reflection is not affected by the higher-order harmonics, and iv) it is easy to measure the physical properties depending on the neutron wavelength. The pulse neutron generator as well as the data acquisition system in the Laboratory of Nuclear Science of Tohoku University is described. The TOF method seems to be very powerful if it is applied to accurate structure analysis. The data correction methods are discussed. The TOF method is prospective to the study of transient phenomena. In this method one can apply to the crystalline sample an external field pulsed with the same frequency as the neutrons. By using this method, the transient state of the polarization reversal of the ferroelectric NaNO/sub 2/ has been observed. The magnetically pulsed neutron TOF spectrometer is briefly introduced after a review of the chopper history.
A neutronic method to determine low hydrogen concentrations in metals
International Nuclear Information System (INIS)
Bennun, Leonardo; Santisteban, Javier; Diaz-Valdes, J.; Granada, J.R.; Mayer, R.E.
2007-01-01
We propose a method for the non-destructive determination of low hydrogen content in metals. The method is based on measurements of neutron inelastic scattering combined with cadmium filters. Determination is simple and the method would allow to construct a mobile device, to perform the analysis 'in situ'. We give a brief description of the usual methods to determine low hydrogen contents in solids, paying special attention to those methods supported by neutron techniques. We describe the proposed method, calculations to achieve a better sensitivity, and experimental results
Energy Technology Data Exchange (ETDEWEB)
Kim, Sang In; Kim, Bong Hwan; Kim, Jang Lyul; Lee, Jung Il [Health Physics Team, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-12-15
The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a {sup 252}Californium ({sup 252}Cf) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1 - 9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.
International Nuclear Information System (INIS)
Kim, Sang In; Kim, Bong Hwan; Kim, Jang Lyul; Lee, Jung Il
2015-01-01
The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a 252 Californium ( 252 Cf) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1 - 9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered
Development of new methods for studying nanostructures using neutron scattering
International Nuclear Information System (INIS)
Pynn, Roger
2016-01-01
The goal of this project was to develop improved instrumentation for studying the microscopic structures of materials using neutron scattering. Neutron scattering has a number of advantages for studying material structure but suffers from the well-known disadvantage that neutrons' ability to resolve structural details is usually limited by the strength of available neutron sources. We aimed to overcome this disadvantage using a new experimental technique, called Spin Echo Scattering Angle Encoding (SESAME) that makes use of the neutron's magnetism. Our goal was to show that this innovation will allow the country to make better use of the significant investment it has recently made in a new neutron source at Oak Ridge National Laboratory (ORNL) and will lead to increases in scientific knowledge that contribute to the Nation's technological infrastructure and ability to develop advanced materials and technologies. We were successful in demonstrating the technical effectiveness of the new method and established a baseline of knowledge that has allowed ORNL to start a project to implement the method on one of its neutron beam lines.
Generation of organic scintillators response function for fast neutrons using the Monte Carlo method
International Nuclear Information System (INIS)
Mazzaro, A.C.
1979-01-01
A computer program (DALP) in Fortran-4-G language, has been developed using the Monte Carlo method to simulate the experimental techniques leading to the distribution of pulse heights due to monoenergetic neutrons reaching an organic scintillator. The calculation of the pulse height distribution has been done for two different systems: 1) Monoenergetic neutrons from a punctual source reaching the flat face of a cylindrical organic scintillator; 2) Environmental monoenergetic neutrons randomly reaching either the flat or curved face of the cylindrical organic scintillator. The computer program has been developed in order to be applied to the NE-213 liquid organic scintillator, but can be easily adapted to any other kind of organic scintillator. With this program one can determine the pulse height distribution for neutron energies ranging from 15 KeV to 10 MeV. (Author) [pt
Directory of Open Access Journals (Sweden)
О. О. Gritzay
2016-12-01
Full Text Available Development of the technique for determination of the total neutron cross sections from the measurements of sample transmission by filtered neutrons, scattered on hydrogen is described. One of the methods of the transmission determination TH52Cr from the measurements of 52Cr sample, using average energy shift method for filtered neutron beam is presented. Using two methods of the experimental data processing, one of which is presented in this paper (another in [1], there is presented a set of transmissions, obtained for different samples and for different measurement angles. Two methods are fundamentally different; therefore, we can consider the obtained processing results, using these methods as independent. In future, obtained set of transmissions is planned to be used for determination of the parameters E0, Гn and R/ of the resonance 52Cr at the energy of 50 keV.
METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY
Wigner, E.P.; Young, G.J.; Weinberg, A.M.
1961-06-27
A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.
Cronos 2: a neutronic simulation software for reactor core calculations
International Nuclear Information System (INIS)
Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.
1999-01-01
The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)
Numerical relativity simulations of precessing binary neutron star mergers
Dietrich, Tim; Bernuzzi, Sebastiano; Brügmann, Bernd; Ujevic, Maximiliano; Tichy, Wolfgang
2018-03-01
We present the first set of numerical relativity simulations of binary neutron mergers that include spin precession effects and are evolved with multiple resolutions. Our simulations employ consistent initial data in general relativity with different spin configurations and dimensionless spin magnitudes ˜0.1 . They start at a gravitational-wave frequency of ˜392 Hz and cover more than 1 precession period and about 15 orbits up to merger. We discuss the spin precession dynamics by analyzing coordinate trajectories, quasilocal spin measurements, and energetics, by comparing spin aligned, antialigned, and irrotational configurations. Gravitational waveforms from different spin configuration are compared by calculating the mismatch between pairs of waveforms in the late inspiral. We find that precession effects are not distinguishable from nonprecessing configurations with aligned spins for approximately face-on binaries, while the latter are distinguishable from nonspinning configurations. Spin precession effects are instead clearly visible for approximately edge-on binaries. For the parameters considered here, precession does not significantly affect the characteristic postmerger gravitational-wave frequencies nor the mass ejection. Our results pave the way for the modeling of spin precession effects in the gravitational waveform from binary neutron star events.
Energy Technology Data Exchange (ETDEWEB)
Kuett, Moritz
2016-07-07
Nuclear arms control, including nuclear safeguards and verification technologies for nuclear disarmament typically use software as part of many different technological applications. This thesis proposes to use three open source criteria for such software, allowing users and developers to have free access to a program, have access to the full source code and be able to publish modifications for the program. This proposition is presented and analyzed in detail, together with the description of the development of ''Open Neutron Multiplicity Simulation'', an open source software tool to simulate neutron multiplicity measurements. The description includes physical background of the method, details of the developed program and a comprehensive set of validation calculations.
Trojan Horse Method for neutrons-induced reaction studies
Gulino, M.; Asfin Collaboration
2017-09-01
Neutron-induced reactions play an important role in nuclear astrophysics in several scenario, such as primordial Big Bang Nucleosynthesis, Inhomogeneous Big Bang Nucleosynthesis, heavy-element production during the weak component of the s-process, explosive stellar nucleosynthesis. To overcome the experimental problems arising from the production of a neutron beam, the possibility to use the Trojan Horse Method to study neutron-induced reactions has been investigated. The application is of particular interest for reactions involving radioactive nuclei having short lifetime.
Uncertainties related to numerical methods for neutron spectra unfolding
International Nuclear Information System (INIS)
Glodic, S.; Ninkovic, M.; Adarougi, N.A.
1987-10-01
One of the often used techniques for neutron detection in radiation protection utilities is the Bonner multisphere spectrometer. Besides its advantages and universal applicability for evaluating integral parameters of neutron fields in health physics practices, the outstanding problems of the method are data analysis and the accuracy of the results. This paper briefly discusses some numerical problems related to neutron spectra unfolding, such as uncertainty of the response matrix as a source of error, and the possibility of real time data reduction using spectrometers. (author)
Selection method and characterization of neutron monochromator natural crystals
International Nuclear Information System (INIS)
Stasiulevicius, R.; Kastner, G.F.
2000-01-01
Thermal neutrons are important analytical tools for microscopic material probe. These neutrons can be selected by diffraction technique using monocrystal, usually artificial. A crystal selection process was implemented and the characteristics of natural specimens were studied by activation analysis-k 0 method. The representative 120 samples, of which 21 best types, were irradiated in IPR-R1 and measured with a neutron diffractometer at IEA-R1m Brazilian reactors. These results are useful for database build up and ease the choice of appropriate natural crystal, with some advantage options: highest intensity diffracted, enlarging the energy operational interval and optimal performance in special applications. (author)
International Nuclear Information System (INIS)
Ohm, H.
1982-01-01
Using the example of the delayed neutron spectrum of 24 s- 137 I the statistical model is tested in view of its applicability. A computer code was developed which simulates delayed neutron spectra by the Monte Carlo method under the assumption that the transition probabilities of the ν and the neutron decays obey the Porter-Thomas distribution while the distances of the neutron emitting levels are Wigner distribution. Gramow-Teller ν-transitions and simply forbidden ν-transitions from the preceding nucleus to the emitting nucleus were regarded. (orig./HSI) [de
The Neutron-Gamma Pulse Shape Discrimination Method for Neutron Flux Detection in the ITER
International Nuclear Information System (INIS)
Xu Xiufeng; Li Shiping; Cao Hongrui; Yin Zejie; Yuan Guoliang; Yang Qingwei
2013-01-01
The neutron flux monitor (NFM), as a significant diagnostic system in the International Thermonuclear Experimental Reactor (ITER), will play an important role in the readings of a series of key parameters in the fusion reaction process. As the core of the main electronic system of the NFM, the neutron-gamma pulse shape discrimination (n-γ PSD) can distinguish the neutron pulse from the gamma pulse and other disturbing pulses according to the thresholds of the rising time and the amplitude pre-installed on the board, the double timing point CFD method is used to get the rising time of the pulse. The n-γ PSD can provide an accurate neutron count. (magnetically confined plasma)
Improved neutron-gamma discrimination for a 3He neutron detector using subspace learning methods
Wang, C. L.; Funk, L. L.; Riedel, R. A.; Berry, K. D.
2017-05-01
3He gas based neutron Linear-Position-Sensitive Detectors (LPSDs) have been used for many neutron scattering instruments. Traditional Pulse-height Analysis (PHA) for Neutron-Gamma Discrimination (NGD) resulted in the neutron-gamma efficiency ratio (NGD ratio) on the order of 105-106. The NGD ratios of 3He detectors need to be improved for even better scientific results from neutron scattering. Digital Signal Processing (DSP) analyses of waveforms were proposed for obtaining better NGD ratios, based on features extracted from rise-time, pulse amplitude, charge integration, a simplified Wiener filter, and the cross-correlation between individual and template waveforms of neutron and gamma events. Fisher Linear Discriminant Analysis (FLDA) and three Multivariate Analyses (MVAs) of the features were performed. The NGD ratios are improved by about 102-103 times compared with the traditional PHA method. Our results indicate the NGD capabilities of 3He tube detectors can be significantly improved with subspace-learning based methods, which may result in a reduced data-collection time and better data quality for further data reduction.
Steyerl, A.; Malik, S. S.; Desai, A. M.; Kaufman, C.
2009-01-01
We review the diffuse scattering and the loss coefficient in ultracold neutron reflection from slightly rough surfaces, report a surprising reduction in loss coefficient due to roughness, and discuss the possibility of transition from quantum treatment to ray optics. The results are used in a computer simulation of neutron storage in a recent neutron lifetime experiment that re-ported a large discrepancy of neutron lifetime with the current particle data value. Our partial re-analysis suggest...
Nodal deterministic simulation for problems of neutron shielding in multigroup formulation
International Nuclear Information System (INIS)
Baptista, Josue Costa; Heringer, Juan Diego dos Santos; Santos, Luiz Fernando Trindade; Alves Filho, Hermes
2013-01-01
In this paper, we propose the use of some computational tools, with the implementation of numerical methods SGF (Spectral Green's Function), making use of a deterministic model of transport of neutral particles in the study and analysis of a known and simplified problem of nuclear engineering, known in the literature as a problem of neutron shielding, considering the model with two energy groups. These simulations are performed in MatLab platform, version 7.0, and are presented and developed with the help of a Computer Simulator providing a friendly computer application for their utilities
Energy spectra of fast neutrons by nuclear emulsion method
International Nuclear Information System (INIS)
Quaresma, A.A.
1977-01-01
An experimental method which uses nuclear emulsion plates to determine the energy spectrum of fission neutrons is described. By using this technique, we have obtained the energy distribution of neutrons from spontaneous fission of Cf 2 5 2 . The results are in good agreement with whose obtained previously by others authors who have used different detection techniques, and they are consistent with a Maxwellian distribution as expected by Weisskopf's nuclear evaporation theory. (author)
On the calibration methods for neutron moisture gauges
International Nuclear Information System (INIS)
Apostol, I.
1975-01-01
Theoretical and experimental calibration methods for devices using neutron sources to measure the water content in subsurface soil and other samples are investigated. Neutron flux density is evaluated by means of the two and three group diffusion and Fermi age theories. The correction criteria for the calibration curves are presented. The agreement of the theoretical curves with the determined experimental data may be considered as excellent. (author)
International Nuclear Information System (INIS)
Aslam; Prestwich, W.V.; McNeill, F.E.; Waker, A.J.
2006-01-01
The neutron irradiation facility developed at the McMaster University 3 MV Van de Graaff accelerator was employed to assess in vivo elemental content of aluminum and manganese in human hands. These measurements were carried out to monitor the long-term exposure of these potentially toxic trace elements through hand bone levels. The dose equivalent delivered to a patient during irradiation procedure is the limiting factor for IVNAA measurements. This article describes a method to estimate the average radiation dose equivalent delivered to the patient's hand during irradiation. The computational method described in this work augments the dose measurements carried out earlier [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724]. This method employs the Monte Carlo simulation of hand irradiation facility using MCNP4B. Based on the estimated dose equivalents received by the patient hand, the proposed irradiation procedure for the IVNAA measurement of manganese in human hands [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724] with normal (1 ppm) and elevated manganese content can be carried out with a reasonably low dose of 31 mSv to the hand. Sixty-three percent of the total dose equivalent is delivered by non-useful fast group (>10 keV); the filtration of this neutron group from the beam will further decrease the dose equivalent to the patient's hand
Experimental method research on neutron equal dose-equivalent detection
International Nuclear Information System (INIS)
Ji Changsong
1995-10-01
The design principles of neutron dose-equivalent meter for neutron biological equi-effect detection are studied. Two traditional principles 'absorption net principle' and 'multi-detector principle' are discussed, and on the basis of which a new theoretical principle for neutron biological equi-effect detection--'absorption stick principle' has been put forward to place high hope on both increasing neutron sensitivity of this type of meters and overcoming the shortages of the two traditional methods. In accordance with this new principle a brand-new model of neutron dose-equivalent meter BH3105 has been developed. Its neutron sensitivity reaches 10 cps/(μSv·h -1 ), 18∼40 times higher than that of all the same kinds of meters 0.23∼0.56 cps/(μSv·h -1 ), available today at home and abroad and the specifications of the newly developed meter reach or surpass the levels of the same kind of meters. Therefore the new theoretical principle of neutron biological equi-effect detection--'absorption stick principle' is proved to be scientific, advanced and useful by experiments. (3 refs., 3 figs., 2 tabs.)
International Nuclear Information System (INIS)
Coeck, M.; Lacoste, V.; Muller, H.
2005-01-01
Full text: Reliable determination of neutron doses in workplaces is still an issue in the field of radiation protection. The EVIDOS project ('evaluation of individual dosimetry in mixed neutron and photon radiation fields', 5FP supported by the EC) aims to evaluate different methods for individual dosimetry in mixed neutron-photon workplaces in nuclear industry, and focuses on the neutron component. This objective cannot be reached on the basis of investigations in calibration fields only, but requires studies in representative workplaces of the nuclear industry. The VENUS reactor, a zero-power research reactor established by the SCK·CEN, was chosen as one of these workplaces. This paper presents the assessment of the neutron field near the VENUS reactor, particularly in areas near the reactor shielding and in the control room where operators are frequently present during a reactor run. From the neutron spectrum, an evaluation of H*(10) can be made. MCNPX simulations were performed to obtain a reference spectrum at the two areas of interest. Using a k eff calculation the source term was acquired which was subsequently used in a fixed source MCNPX model of the complete shielding geometry of the reactor hall. Reference spectrometry was also performed using a Bonner spheres system. The unfolding spectra were obtained using the NUBAY and GRAVEL codes. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterized spectrum and provides posterior probability distributions for both the set of parameters and a set of integral quantities. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. Bonner spheres data GRAVEL unfolding was also performed using the MCNPX spectra as an initial guess. In this paper the outcome of both calculations and measurements is compared. (author)
Calibration of quantitative neutron radiography method for moisture measurement
International Nuclear Information System (INIS)
Nemec, T.; Jeraj, R.
1999-01-01
Quantitative measurements of moisture and hydrogenous matter in building materials by neutron radiography (NR) are regularly performed at TRIGA Mark II research of 'Jozef Stefan' Institute in Ljubljana. Calibration of quantitative method is performed using standard brick samples with known moisture content and also with a secondary standard, plexiglas step wedge. In general, the contribution of scattered neutrons to the neutron image is not determined explicitly what introduces an error to the measured signal. Influence of scattered neutrons is significant in regions with high gradients of moisture concentrations, where the build up of scattered neutrons causes distortion of the moisture concentration profile. In this paper detailed analysis of validity of our calibration method for different geometrical parameters is presented. The error in the measured hydrogen concentration is evaluated by an experiment and compared with results obtained by Monte Carlo calculation with computer code MCNP 4B. Optimal conditions are determined for quantitative moisture measurements in order to minimize the error due to scattered neutrons. The method is tested on concrete samples with high moisture content.(author)
Influence of Neutron Spectra Unfolding Method on Fast Neutron Dose Determination
International Nuclear Information System (INIS)
Marinkovic, P.
1991-01-01
Full text: Accuracy of knowing the fast neutron spectra has great influence on equivalent dose determination. In usual fast neutron spectrum measurements with scintillation detectors based on proton recoil, the main difficulty is confidence of unfolding method. In former ones variance of obtained result is usually great and negative values are possible too, which does means that we don't now exactly is obtained neutron spectrum real one. The new unfolding method based on Shanon's information theory, which gives non-negative spectrum and relative low variance, is obtained and appropriate numerical code for application in fast neutron spectrometry based on proton recoil is realized. In this method principle of maximum entropy and maximum likelihood are used together. Unknown group density distribution functions, which are considered as desired normalized mean neutron group flux, are constl u cted using only constrain of knowing mean value. Obtained distributions are consistent to available information (counts in NCA from proton recoil), while being maximally noncommittal with respect to all other unknown circumstances. For maximum likelihood principle, distribution functions around mean value of counts in the channels of MCA are taken to be Gauss function shape. Optimal non-negative solution is searched by means of Lagrange parameter method. Nonlinear system of equations, is solved using gradient and Newton iterative algorithm. Error covariance matrix is obtained too. (author)
Comparison of methods of calibration of a neutron probe by gravimetry or neutron-capture model
International Nuclear Information System (INIS)
Vachaud, G.; Royer, J.M.; Cooper, J.D.
1977-01-01
This paper presents a systematic analysis of two methods used for determining calibration curves of neutron probes. The uncertainties resulting from the use of the gravimetric method, with a linear correlation between count rates and water content of soil samples, are considered first. Particular care is given to the determination of errors in the values of water content and count rates, and to the difficulties arising from the choice of the correlation technique. The neutron-calibration curve of the soil was also obtained with a technique based on the determination of neutron thermal adsorption and diffusion constants. The importance of errors associated with this method is also analyzed. Different field examples are then presented. It appears that the neutron-capture technique should be particularly well suited for determining the calibration curve of clay soils or heterogeneous materials for which the gravimetric calibration technique cannot be applied with confidence. On the other hand, it is also shown that for a soil with a very well-defined gravimetric calibration curve, the neutron-capture technique gives results still at least as good as with the former method
Energy Technology Data Exchange (ETDEWEB)
Tumelero, Fernanda; Petersen, Claudio Zen; Goncalves, Glenio Aguiar [Universidade Federal de Pelotas, Capao do Leao, RS (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Schramm, Marcelo [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica
2016-12-15
In this work, we report a solution to solve the Neutron Point Kinetics Equations applying the Polynomial Approach Method. The main idea is to expand the neutron density and delayed neutron precursors as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions and the analytical continuation is used to determine the solutions of the next intervals. A genuine error control is developed based on an analogy with the Rest Theorem. For illustration, we also report simulations for different approaches types (linear, quadratic and cubic). The results obtained by numerical simulations for linear approximation are compared with results in the literature.
Perego, R.C.
2004-01-01
Two novel neutron-based analytical techniques have been treated in this thesis, Neutron Resonance Capture Analysis (NRCA), employing a pulsed neutron source, and Neutron Incoherent Scattering (NIS), making use of a cold neutron source. With the NRCA method isotopes are identified by the
Monte Carlo simulation study of the muon-induced neutron flux at LNGS
International Nuclear Information System (INIS)
Persiani, R.; Garbini, M.; Massoli, F.; Sartorelli, G; Selvi, M.
2011-01-01
Muon-induced neutrons are ultimate background for all the experiments searching for rare events in underground laboratories. Several measurements and simulations were performed concerning the neutron production and propagation but there are disagreements between experimental data and simulations. In this work we present our Monte-Carlo simulation study, based on Geant4, to estimate the muon-induced neutron flux at LNGS. The obtained integral flux of neutrons above 1 MeV is 2.31 x 10 -10 n/cm 2 /s.
Studies and applications of neutron radiography with film methods
International Nuclear Information System (INIS)
Ikeda, Yasushi
1989-01-01
Neutron radiography has been studied with film methods and applied to some industrial applications. The film methods include not only conventional silver-halide emulsion films, such as industrial, medical or soft X-ray ones, but also track-etch films and those for indirect methods. The characteristics of the film methods are analyzed and investigated with using various image converters, such as gadolinium metal foil and evaporation films, or some scintillation converters such as NE426. The sensitivities and MTFs for various sets of films and converters have been obtained, which gives a chart of the correlation between the appropriate exposure and resolving powers for them. From the chart, one can select some proper sets for the purpose and given conditions of neutron radiography facilities. The film methods have been applied to inspect very fine cracks in thick steel blocks and plates. It is also applied to observe nuclear fuel pellets or irradiated nuclear fuel pins. Furthermore, the film method has been used for neutron computed tomography. Very fine Eu-particles in TiO pellets, which diameters are nearly 300 micron, can be reconstructed by the neutron CT. The fine neutron CT will be useful for the inspection of Pu-particles in mixed oxide nuclear fuel pellets for future advance nuclear reactors. (author)
Multi-group transport methods for high-resolution neutron activation analysis
International Nuclear Information System (INIS)
Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.
2009-01-01
The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)
Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code
Haas, Roland; Ott, Christian D.; Szilagyi, Bela; Kaplan, Jeffrey D.; Lippuner, Jonas; Scheel, Mark A.; Barkett, Kevin; Muhlberger, Curran D.; Dietrich, Tim; Duez, Matthew D.; Foucart, Francois; Pfeiffer, Harald P.; Kidder, Lawrence E.; Teukolsky, Saul A.
2016-06-01
We present results on the inspiral, merger, and postmerger evolution of a neutron star-neutron star (NSNS) system. Our results are obtained using the hybrid pseudospectral-finite volume Spectral Einstein Code (SpEC). To test our numerical methods, we evolve an equal-mass system for ≈22 orbits before merger. This waveform is the longest waveform obtained from fully general-relativistic simulations for NSNSs to date. Such long (and accurate) numerical waveforms are required to further improve semianalytical models used in gravitational wave data analysis, for example, the effective one body models. We discuss in detail the improvements to SpEC's ability to simulate NSNS mergers, in particular mesh refined grids to better resolve the merger and postmerger phases. We provide a set of consistency checks and compare our results to NSNS merger simulations with the independent bam code. We find agreement between them, which increases confidence in results obtained with either code. This work paves the way for future studies using long waveforms and more complex microphysical descriptions of neutron star matter in SpEC.
Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A
2014-12-01
An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.
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
Monte-Carlo simulation on the cold neutron guides at CARR
International Nuclear Information System (INIS)
Guo Liping; Wang Hongli; Yang Tonghua; Cheng Zhixu; Liu Yi
2003-01-01
The designs of the two cold neutron guides to be built at China Advanced Research Reactor (CARR) are simulated with Monte-Carlo simulation software VITESS. Various parameters of the guides, e.g. transmission efficiency, neutron flux, divergence, etc., are obtained. (author)
Neutronic characteristics simulation of LMFBR of great size
International Nuclear Information System (INIS)
Kim, Y.C.
1987-09-01
The CONRAD experimental program to be executed on the critical mockup MASURCA in Cadarache and use all the european plutonium stock. The objectives of this program are to reduce the uncertainties on important project parameters such as the reactivity value of control rods, the flux distribution to valid calcul methods and data to use for new LMFBR conception (heterogeneous axial core by example) and to resolve the neutronic control problems for a LMFBR of great size. The present study has permitted to define this program and its physical characteristics [fr
High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation
Energy Technology Data Exchange (ETDEWEB)
Was, Gary; Wirth, Brian; Motta, Athur; Morgan, Dane; Kaoumi, Djamel; Hosemann, Peter; Odette, Robert
2018-04-30
Project Objective: The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiated microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations
Design and simulation of an optimized e-linac based neutron source for BNCT research
International Nuclear Information System (INIS)
Durisi, E.; Alikaniotis, K.; Borla, O.; Bragato, F.; Costa, M.; Giannini, G.; Monti, V.; Visca, L.; Vivaldo, G.; Zanini, A.
2015-01-01
The paper is focused on the study of a novel photo-neutron source for BNCT preclinical research based on medical electron Linacs. Previous studies by the authors already demonstrated the possibility to obtain a mixed thermal and epithermal neutron flux of the order of 10"7 cm"−"2 s"−"1. This paper investigates possible Linac’s modifications and a new photo-converter design to rise the neutron flux above 5 10"7 cm"−"2 s"−"1, also reducing the gamma contamination. - Highlights: • Proposal of a mixed thermal and epithermal (named hyperthermal) neutron source based on medical high energy electron Linac. • Photo-neutron production via Giant Dipole Resonance on high Z materials. • MCNP4B-GN simulations to design the photo-converter geometry maximizing the hyperthermal neutron flux and minimizing the fast neutron and gamma contaminations. Hyperthermal neutron field suitable for BNCT preclinical research.
New methods in plasma simulation
International Nuclear Information System (INIS)
Mason, R.J.
1990-01-01
The development of implicit methods of particle-in-cell (PIC) computer simulation in recent years, and their merger with older hybrid methods have created a new arsenal of simulation techniques for the treatment of complex practical problems in plasma physics. The new implicit hybrid codes are aimed at transitional problems that lie somewhere between the long time scale, high density regime associated with MHD modeling, and the short time scale, low density regime appropriate to PIC particle-in-cell techniques. This transitional regime arises in ICF coronal plasmas, in pulsed power plasma switches, in Z-pinches, and in foil implosions. Here, we outline how such a merger of implicit and hybrid methods has been carried out, specifically in the ANTHEM computer code, and demonstrate the utility of implicit hybrid simulation in applications. 25 refs., 5 figs
International Nuclear Information System (INIS)
Malikova, N.
2005-09-01
Montmorillonite clays in low hydration states, with Na + and Cs + compensating counter ions, are investigated by a combination of microscopic simulation and quasi-elastic neutron scattering to obtain information on the local structure and dynamics of water and ions in the interlayer. At first predictions of simulation into the dynamics of water and ions at elevate temperatures are shown (0 deg C 80 deg C, pertinent for the radioactive waste disposal scenario) Marked difference is observed between the modes of diffusion of the Na + and C + counter ions. In water dynamics, a significant step towards bulk water behaviour is seen on transition from the mono- to bilayer states. Secondly, a detailed comparison between simulation and quasi-elastic neutron scattering (Neutron Spin Echo and Time-of-Flight) regarding ambient temperature water dynamics is presented. Overall, the approaches are found to be in good agreement with each other and limitations of each of the methods are clearly shown. (author)
Neutron absorbing article and method for manufacture of such article
International Nuclear Information System (INIS)
Hortman, M.T.; Mcmurtry, C.H.; Naum, R.G.; Owens, D.P.
1980-01-01
A neutron absorbing article, preferably in long, thin, flat form , suitable for but not necessarily limited to use in storage racks for spent nuclear fuel at locations between volumes of such stored fuel, to absorb neutrons from said spent fuel and prevent uncontrolled nuclear reaction of the spent fuel material, is composed of finely divided boron carbide particles and a solid, irreversibly cured phenolic polymer, forming a continuous matrix about the boron carbide particles, in such proportions that at least 6% of b10 from the boron carbide content is present therein. The described articles withstand thermal cycling from repeated spent fuel insertions and removals, withstand radiation from said spent nuclear fuel over long periods of time without losing desirable neutron absorbing and physical properties, are sufficiently chemically inert to water so as to retain neutron absorbing properties if brought into contact with it, are not galvanically corrodible and are sufficiently flexible so as to withstand operational basis earthquake and safe shutdown earthquake seismic events, without loss of neutron absorbing capability and other desirable properties, when installed in storage racks for spent nuclear fuel. The disclosure also relates to a plurality of such neutron absorbing articles in a storage rack for spent nuclear fuel and to a method for the manufacture of the articles
A Monte-Carlo study of landmines detection by neutron backscattering method
International Nuclear Information System (INIS)
Maucec, M.; De Meijer, R.J.
2000-01-01
The use of Monte-Carlo simulations for modelling a simplified landmine detector system with a 252 Cf- neutron source is presented in this contribution. Different aspects and variety of external conditions, affecting the localisation and identification of a buried suspicious object (such as landmine) have been tested. Results of sensitivity calculations confirm that the landmine detection methods, based on the analysis of the backscattered neutron radiation can be applicable in higher density formations, with the mass fraction of present pore-water <15 %. (author)
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)
Energy Technology Data Exchange (ETDEWEB)
Tumelero, Fernanda, E-mail: fernanda.tumelero@yahoo.com.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana, E-mail: claudiopeteren@yahoo.com.br, E-mail: gleniogoncalves@yahoo.com.br, E-mail: luana-lazzari@hotmail.com [Universidade Federal de Pelotas (DME/UFPEL), Capao do Leao, RS (Brazil). Instituto de Fisica e Matematica
2015-07-01
In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)
International Nuclear Information System (INIS)
Tumelero, Fernanda; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana
2015-01-01
In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)
Review of unfolding methods for neutron flux dosimetry
International Nuclear Information System (INIS)
Stallmann, F.W.; Kam, F.B.K.
1975-01-01
The primary method in reactor dosimetry is the foil activation technique. To translate the activation measurements into neutron fluxes, a special data processing technique called unfolding is needed. Some general observations about the problems and the reliability of this approach to reactor dosimetry are presented. Current unfolding methods are reviewed. 12 references. (auth)
Ciolfi, Riccardo; Kastaun, Wolfgang; Giacomazzo, Bruno; Endrizzi, Andrea; Siegel, Daniel M.; Perna, Rosalba
2017-03-01
Merging binary neutron stars (BNSs) represent the ultimate targets for multimessenger astronomy, being among the most promising sources of gravitational waves (GWs), and, at the same time, likely accompanied by a variety of electromagnetic counterparts across the entire spectrum, possibly including short gamma-ray bursts (SGRBs) and kilonova/macronova transients. Numerical relativity simulations play a central role in the study of these events. In particular, given the importance of magnetic fields, various aspects of this investigation require general relativistic magnetohydrodynamics (GRMHD). So far, most GRMHD simulations focused the attention on BNS mergers leading to the formation of a hypermassive neutron star (NS), which, in turn, collapses within few tens of ms into a black hole surrounded by an accretion disk. However, recent observations suggest that a significant fraction of these systems could form a long-lived NS remnant, which will either collapse on much longer time scales or remain indefinitely stable. Despite the profound implications for the evolution and the emission properties of the system, a detailed investigation of this alternative evolution channel is still missing. Here, we follow this direction and present a first detailed GRMHD study of BNS mergers forming a long-lived NS. We consider magnetized binaries with different mass ratios and equations of state and analyze the structure of the NS remnants, the rotation profiles, the accretion disks, the evolution and amplification of magnetic fields, and the ejection of matter. Moreover, we discuss the connection with the central engine of SGRBs and provide order-of-magnitude estimates for the kilonova/macronova signal. Finally, we study the GW emission, with particular attention to the post-merger phase.
Testing the characteristics of a neutron detector array by Monte-Carlo simulations
International Nuclear Information System (INIS)
Timis, C.; Cruceru, I.; Sandu, M.; Borcea, C.; Buta, A.; Negoita, F.; Angelique, J.C.; Martin, T.; Peter, J.; Grevy, S.; Lienard, E.; Orr, N.A.
1998-01-01
The characteristics of the neutron detector array TONNERRE have been determined experimentally via preliminary tests with a 252 Cf source and by means of simulation using a modified version of the Monte-Carlo program of Cecil et al. Of particular interest is the intrinsic detection efficiency. As it is well known, the neutron detection efficiency for one element of the detector array, depends on the threshold for the light collection (bias) expressed in energy electron equivalent. The experimental efficiencies for five neutron energies and for a bias of 80 KeV ee are presented. The efficiencies for three thresholds and neutron energies between 1-10 MeV are simulated. The neutron energy is determined by TOF over a flight path, s, and the relative energy resolution is given as a function of σ s and σ t (the uncertainties in the flight path), s (uniform as a function of depth) and flight time, t. The mean time resolution was 1.13 ns which gives a TOF resolution of 1.48 ns. That gives a relative energy resolution which increases slowly from 2% at E n =1 MeV to 3.5% at 5 MeV. Position resolution along one module is 12 cm. To help boosting the efficiency, the elements can be arranged in two layers, but that complicates the analysis by enhancing the effects of cross-talk and out-scattering. Cross-talk is the familiar problem of one neutron creating signals in two separate detectors. In out-scattering, a neutron scatters from the non-active part of a detector and is then detected in a different detector with incorrect position and TOF. While methods exist for identifying and eliminating cross-talk events, there are no methods available for identifying out-scattered events. For the case of two layers and a bias of 80 KeV ee, simulated efficiency of two superposed elements versus neutron energy, the out-scattering probability and the probability of cross-talk are presented. The out-scattering probability comes mainly from events when neutrons scatter first on carbon nuclei
Numerical simulation of binary black hole and neutron star mergers
Energy Technology Data Exchange (ETDEWEB)
Kastaun, W.; Rezzolla, L. [Albert Einstein Institut, Potsdam-Golm (Germany)
2016-11-01
One of the last predictions of general relativity that still awaits direct observational confirmation is the existence of gravitational waves. Those fluctuations of the geometry of space and time are expected to travel with the speed of light and are emitted by any accelerating mass. Only the most violent events in the universe, such as mergers of two black holes or neutron stars, produce gravitational waves strong enough to be measured. Even those waves are extremely weak when arriving at Earth, and their detection is a formidable technological challenge. In recent years sufficiently sensitive detectors became operational, such as GEO600, Virgo, and LIGO. They are expected to observe around 40 events per year. To interpret the observational data, theoretical modeling of the sources is a necessity, and requires numerical simulations of the equations of general relativity and relativistic hydrodynamics. Such computations can only be carried out on large scale supercomputers, given that many scenarios need to be simulated, each of which typically occupies hundreds of CPU cores for a week. Our main goal is to predict the gravitational wave signal from the merger of two compact objects. Comparison with future observations will provide important insights into the fundamental forces of nature in regimes that are impossible to recreate in laboratory experiments. The waveforms from binary black hole mergers would allow one to test the correctness of general relativity in previously inaccessible regimes. The signal from binary neutron star mergers will provide input for nuclear physics, because the signal depends strongly on the unknown properties of matter at the ultra high densities inside neutron stars, which cannot be observed in any other astrophysical scenario. Besides mergers, we also want to improve the theoretical models of close encounters between black holes. A gravitational wave detector with even higher sensitivity, the Einstein Telescope, is already in the
International Nuclear Information System (INIS)
Endo, Tomohiro
2011-01-01
In this paper, an alternative definition of a neutron multiplication factor, detected-neutron multiplication factor kdet, is produced for the neutron source multiplication method..(NSM). By using kdet, a search strategy of appropriate detector position for NSM is also proposed. The NSM is one of the practical subcritical measurement techniques, i.e., the NSM does not require any special equipment other than a stationary external neutron source and an ordinary neutron detector. Additionally, the NSM method is based on steady-state analysis, so that this technique is very suitable for quasi real-time measurement. It is noted that the correction factors play important roles in order to accurately estimate subcriticality from the measured neutron count rates. The present paper aims to clarify how to correct the subcriticality measured by the NSM method, the physical meaning of the correction factors, and how to reduce the impact of correction factors by setting a neutron detector at an appropriate detector position
International Nuclear Information System (INIS)
Li, Gang; Xu, Jiayun; Zhang, Jie
2015-01-01
Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am–Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am–Be neutron source was investigated by measuring the count rates obtained through a 3 He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, 3 He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter. - Highlights: • The scope of the affected area for
Method of inspecting Raschig rings by neutron absorption counting
International Nuclear Information System (INIS)
Morris, R.N.; Murri, R.L.; Hume, M.W.
1979-01-01
A neutron counting method for inspecting borosilicate glass Raschig rings and an apparatus designed specifically for this method are discussed. The neutron count ratios for rings of a given thickness show a linear correlation to the boron oxide content of the rings. The count ratio also has a linear relationship to the thickness of rings of a given boron oxide content. Consequently, the experimentally-determined count ratio and physically-measured thickness of Raschig rings can be used to statistically predict their boron oxide content and determine whether or not they meet quality control acceptance criteria
Reactor internals vibration monitoring by neutron noise methods in PWRs
International Nuclear Information System (INIS)
Pazsit, I.; Por, G.; Lux, I.
1983-01-01
Certain elements of PWR cores such as control/fuel rods or cassettes, or other parts of reactor internals, often represent a vibration problem. Early analyses at operating PWR plant revealed that these vibrations can be detected by in-core neutron detectors, opening up the possibility of vibration monitoring and diagnostics by noise methods. Theoretical methods of calculating vibration induced neutron noise and its application to vibration diagnostics are summarized. Experiments to check theoretical conclusions are under way at the Central Research Institute for Physics, Budapest. (author)
International Nuclear Information System (INIS)
Ji Gang; Guo Yong; Luo Yisheng; Zhang Wenzhong
2001-01-01
Objective: To provide useful parameters for neutron radiotherapy, the author presents results of a Monte Carlo simulation study investigating the dosimetric characteristics of linear 252 Cf fission neutron sources. Methods: A 252 Cf fission source and tissue equivalent phantom were modeled. The dose of neutron and gamma radiations were calculated using Monte Carlo Code. Results: The dose of neutron and gamma at several positions for 252 Cf in the phantom made of equivalent materials to water, blood, muscle, skin, bone and lung were calculated. Conclusion: The results by Monte Carlo methods were compared with the data by measurement and references. According to the calculation, the method using water phantom to simulate local tissues such as muscle, blood and skin is reasonable for the calculation and measurements of dose distribution for 252 Cf
Comparison of methods for the detection of gravitational waves from unknown neutron stars
Walsh, S.; Pitkin, M.; Oliver, M.; D'Antonio, S.; Dergachev, V.; Królak, A.; Astone, P.; Bejger, M.; Di Giovanni, M.; Dorosh, O.; Frasca, S.; Leaci, P.; Mastrogiovanni, S.; Miller, A.; Palomba, C.; Papa, M. A.; Piccinni, O. J.; Riles, K.; Sauter, O.; Sintes, A. M.
2016-12-01
Rapidly rotating neutron stars are promising sources of continuous gravitational wave radiation for the LIGO and Virgo interferometers. The majority of neutron stars in our galaxy have not been identified with electromagnetic observations. All-sky searches for isolated neutron stars offer the potential to detect gravitational waves from these unidentified sources. The parameter space of these blind all-sky searches, which also cover a large range of frequencies and frequency derivatives, presents a significant computational challenge. Different methods have been designed to perform these searches within acceptable computational limits. Here we describe the first benchmark in a project to compare the search methods currently available for the detection of unknown isolated neutron stars. The five methods compared here are individually referred to as the PowerFlux, sky Hough, frequency Hough, Einstein@Home, and time domain F -statistic methods. We employ a mock data challenge to compare the ability of each search method to recover signals simulated assuming a standard signal model. We find similar performance among the four quick-look search methods, while the more computationally intensive search method, Einstein@Home, achieves up to a factor of two higher sensitivity. We find that the absence of a second derivative frequency in the search parameter space does not degrade search sensitivity for signals with physically plausible second derivative frequencies. We also report on the parameter estimation accuracy of each search method, and the stability of the sensitivity in frequency and frequency derivative and in the presence of detector noise.
Neutronics methods for transient and safety analysis of fast reactors
Energy Technology Data Exchange (ETDEWEB)
Marchetti, Marco
2017-07-01
neutron flux. Recently, there has been a tendency to abandon twodimensional reactor models in favor of full three-dimensional ones. Only the latter models are adequate to represent phenomena that are inherently three-dimensional. Because of the computational costs attached, three-dimensional accident simulation would become practical only if all the different modules applied for the safety analysis, and therefore the neutronics module as well, are sufficiently fast. In this work, fast neutron solvers are found to reduce the total computational time even by 60%, paving thus the way to three-dimensional modeling, and to a more accurate description of the neutron field and its influence on the results of safety analyses.
Radiation Transport Simulation for Boron Neutron Capture Therapy (BNCT)
Energy Technology Data Exchange (ETDEWEB)
Ziegner, M.; Blaickner, M. [AIT Austrian Institute of Technology GmbH, Health and Environment Department, Molecular Medicine, Muthgasse 11, 1190 Wien (Austria); Ziegner, M.; Khan, R.; Boeck, H. [Vienna University of Technology, Institute of Atomic and Subatomic Physics, Stadionallee 2, 1020 Wien (Austria); Bortolussi, S.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, National Institute of Nuclear Physics (INFN) Pavia Section, Pavia (Italy); Schmitz, T.; Hampel, G. [Nuclear Chemistry, University of Mainz, Fritz Strassmann Weg 2, 55099 Mainz (Germany)
2011-07-01
This work is part of a larger project initiated by the University of Mainz and aiming to use the university's TRIGA reactor to develop a treatment for liver metastases based on Boron Neutron Capture Therapy (BNCT). Diffuse distribution of cancerous cells within the organ makes complete resection difficult and the vicinity to radiosensitive organs impedes external irradiation. Therefore the method of 'autotransplantation', first established at the University of Pavia, is used. The liver is taken out of the body, irradiated in the thermal column of the reactor, therewith purged of metastases and then reimplanted. A highly precise dosimetry system is to be developed by means of measurements at the University of Mainz and computational calculations at the AIT. The stochastic MCNP-5 Monte Carlo-Code, developed by Los Alamos Laboratories, is applied. To verify the calculations of the flux and the absorbed dose in matter a number of measurements are performed irradiating different phantoms and liver sections in a 20cm x 20cm beam tube, which was created by removing graphite blocks from the thermal column of the reactor. The detector material consists of L- {alpha} -alanine pellets which are thought to be the most suitable because of their good tissue equivalence, small size and their wide response range. Another experiment focuses on the determination of the relative biological effectiveness (RBE-factor) of the neutron and photon dose for liver cells. Therefore cell culture plates with the cell medium enriched with {sup 157}Gd and {sup 10}B at different concentrations are irradiated. With regard to the alanine pellets MCNP-5 calculations give stable results. Nevertheless the absorbed dose is underestimated compared to the measurements, a phenomenon already observed in previous works. The cell culture calculations showed the enormous impact of the added isotopes with high thermal neutron cross sections, especially {sup 157}Gd, on the absorbed dose
Neutronic simulation calculations to assess the proliferation resistance of nuclear technologies
International Nuclear Information System (INIS)
Englert, Matthias
2009-01-01
This thesis investigates the proliferation resistance of nuclear technologies on the basis of three case studies. After a brief description of the concept of proliferation resistance the utilized computer codes and methods are presented. The first case study investigates the potential of monolithic fuel for the conversion of one-fuel-element high-flux research reactors from highly enriched to low enriched uranium using the example of the german research reactor FRM-II. The second case study assesses the proliferation potential of future tokamak based fusion reactors by using neutronic simulations of a possible plutonium production. The third example investigates the proliferation potential of spallation neutron sources to produce nuclear weapon relevant material and the proliferation resistance of such facilities. (orig.)
One directional polarized neutron reflectometry with optimized reference layer method
International Nuclear Information System (INIS)
Masoudi, S. Farhad; Jahromi, Saeed S.
2012-01-01
In the past decade, several neutron reflectometry methods for determining the modulus and phase of the complex reflection coefficient of an unknown multilayer thin film have been worked out among which the method of variation of surroundings and reference layers are of highest interest. These methods were later modified for measurement of the polarization of the reflected beam instead of the measurement of the intensities. In their new architecture, these methods not only suffered from the necessity of change of experimental setup but also another difficulty was added to their experimental implementations. This deficiency was related to the limitations of the technology of the neutron reflectometers that could only measure the polarization of the reflected neutrons in the same direction as the polarization of the incident beam. As the instruments are limited, the theory has to be optimized so that the experiment could be performed. In a recent work, we developed the method of variation of surroundings for one directional polarization analysis. In this new work, the method of reference layer with polarization analysis has been optimized to determine the phase and modulus of the unknown film with measurement of the polarization of the reflected neutrons in the same direction as the polarization of the incident beam.
Modeling Taylor series approximations for prompt neutron kinetics with lab view simulations
International Nuclear Information System (INIS)
Adzri, E. P.
2012-09-01
The reactor point kinetics equations have been subjected to intense research in an effort to find simple yet accurate numerical solutions methods. The equations are very stiff numerically, meaning that there is a wide variation in the decay constants, so that using a particular time step in the numerical solution may provide sufficient accuracy for the group, but not for another. Several solutions techniques have been presented on the point kinetics equations with varying degrees of complexity. These include Power Series Solutions, CORE, PCA, Genapol and Taylor series methods. In this research, algorithms were developed based on the first and second order Taylor series expansion and simulated in LabVIEW to solve the Reactor Point Kinetics equations using block diagram nodes implemented within stacked sequences. The algorithms developed were fast,accurate and simple to code. Several reactivity insertions were used to simulate the change in neutron population with time. The LabVIEW- Taylor series solutions were compared with other solution techniques such as Power Series Solutions, CORE, PCA, Genapol and McMahon and Pierson's Taylor series approximation. The results of LabVIEW-Taylor series technique used by McMahon and Pearson The LabVIEW-implemented techniques were found to agree very well with these other methods. At 1x10 -8 s the neutron population was 1.000220 neutrons / cm 3 , at 1 x 10 -2 s it was 2.007681 neutrons / cm 3 and at 1x10 -1 s it was 2.075317 neutrons / cm 3 ; same results reported by Genapol for a fast reactor, it produced good and accurate results and compared very favorably with other methods found in the literature. Using much smaller time steps to the order or 10 -8 s commensurate with fast reactor parameters also produced very satisfactory results, indicating that the LabVIEW-based Taylor series technique is suitable for simulating the kinetics of fast reactors as well as thermal reactors. Algorithms developed that included second order terms
Development of new methods for studying nanostructures using neutron scattering
Energy Technology Data Exchange (ETDEWEB)
Pynn, Roger [Indiana Univ., Bloomington, IN (United States)
2016-03-18
The goal of this project was to develop improved instrumentation for studying the microscopic structures of materials using neutron scattering. Neutron scattering has a number of advantages for studying material structure but suffers from the well-known disadvantage that neutrons’ ability to resolve structural details is usually limited by the strength of available neutron sources. We aimed to overcome this disadvantage using a new experimental technique, called Spin Echo Scattering Angle Encoding (SESAME) that makes use of the neutron’s magnetism. Our goal was to show that this innovation will allow the country to make better use of the significant investment it has recently made in a new neutron source at Oak Ridge National Laboratory (ORNL) and will lead to increases in scientific knowledge that contribute to the Nation’s technological infrastructure and ability to develop advanced materials and technologies. We were successful in demonstrating the technical effectiveness of the new method and established a baseline of knowledge that has allowed ORNL to start a project to implement the method on one of its neutron beam lines.
The influence of neutron activated products in BIXS method for tungsten
International Nuclear Information System (INIS)
Yang, Yang; Peng, Taiping; Chen, Zhilin; Zhang, Zhe; Song, Honghu
2016-01-01
Highlights: • An approach is proposed to assess the influence of neutron activated products in BIXS method for tungsten based on Monte Carlo simulations. • Six principal radioactive neutron activated products (W-181, W-185, W-187, Ta-182, Re-186 and Re-188) with exhaustive decay data were included in the calculation to simulate the real condition. • W-187 strongly affects the tritium measurement shortly after DT experiments and about one month later W-181, W-185 and Ta-182 dominate the effects of neutron activated products. • The intensity of spectrum devoted by NAPs is as much as 10.6 times larger than by tritium in the same region of interest (ROI) directly after reactor shutdown, while it will reduce to 0.83% after 19 months. - Abstract: Tritium β-decay Induced X-ray Spectroscopy (BIXS) is a promising method to obtain tritium retention details both on surface and in bulk of plasma facing materials (PFMs) in magnetic fusion devices. Influence of neutron activated products (NAPs) is one of the primary problems in BIXS, since the energy of Compton scattering photons and characteristic X-rays is close to the range of tritium β-rays. In this paper, Monte Carlo simulation was introduced to evaluate effects of NAPs in BIXS measurements for tungsten. Three steps are included in simulation, neutron flux simulation to reckon the neutron flux in PFMs based on the simplified model of ITER, activated products calculation to compute the distribution of NAPs in PFMs and radiation spectra simulation to calculate the spectra of NAPs in PFMs sample. Six typical nuclides (including W-185, W-181 and Re-188, etc.) produced by neutron irradiation were considered in this study. Finally, influence of NAPs was estimated quantitatively with the assumption that 1% of DT fuel retained in PFMs uniformly. It is found that the intensity of spectrum devoted by NAPs is as much as 10.6 times larger than it by tritium in the same region of interest (ROI) directly after reactor shutdown
The influence of neutron activated products in BIXS method for tungsten
Energy Technology Data Exchange (ETDEWEB)
Yang, Yang; Peng, Taiping, E-mail: yangyanginpc@163.com; Chen, Zhilin; Zhang, Zhe; Song, Honghu
2016-12-15
Highlights: • An approach is proposed to assess the influence of neutron activated products in BIXS method for tungsten based on Monte Carlo simulations. • Six principal radioactive neutron activated products (W-181, W-185, W-187, Ta-182, Re-186 and Re-188) with exhaustive decay data were included in the calculation to simulate the real condition. • W-187 strongly affects the tritium measurement shortly after DT experiments and about one month later W-181, W-185 and Ta-182 dominate the effects of neutron activated products. • The intensity of spectrum devoted by NAPs is as much as 10.6 times larger than by tritium in the same region of interest (ROI) directly after reactor shutdown, while it will reduce to 0.83% after 19 months. - Abstract: Tritium β-decay Induced X-ray Spectroscopy (BIXS) is a promising method to obtain tritium retention details both on surface and in bulk of plasma facing materials (PFMs) in magnetic fusion devices. Influence of neutron activated products (NAPs) is one of the primary problems in BIXS, since the energy of Compton scattering photons and characteristic X-rays is close to the range of tritium β-rays. In this paper, Monte Carlo simulation was introduced to evaluate effects of NAPs in BIXS measurements for tungsten. Three steps are included in simulation, neutron flux simulation to reckon the neutron flux in PFMs based on the simplified model of ITER, activated products calculation to compute the distribution of NAPs in PFMs and radiation spectra simulation to calculate the spectra of NAPs in PFMs sample. Six typical nuclides (including W-185, W-181 and Re-188, etc.) produced by neutron irradiation were considered in this study. Finally, influence of NAPs was estimated quantitatively with the assumption that 1% of DT fuel retained in PFMs uniformly. It is found that the intensity of spectrum devoted by NAPs is as much as 10.6 times larger than it by tritium in the same region of interest (ROI) directly after reactor shutdown
Energy Technology Data Exchange (ETDEWEB)
Askew, J R; Brissenden, R J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)
1963-08-15
This report gives an account of the DSN method for simulating neutron transport, together with methods of solution developed to deal with problems in the physics of thermal reactors, for which previously available computer programmes were unsatisfactory. The methods described are those incorporated in the programmes WINFRITH DSN written in FORTRAN language for the IBM 7090 and STRETCH computers. (author)
Methods for processing of pulsed neutron logging data
International Nuclear Information System (INIS)
Iskenderov, V.G.
1975-01-01
Conditions were examined for calculating the lifetime of thermal neutrons in neutron-neutron logging by selecting optimum values for the time lag. The dispersion and mean square error of the calculated lifetime values for thermal neutrons are evaluated
International Nuclear Information System (INIS)
Camargo, Dayana Queiroz de
2011-01-01
This thesis has developed a stochastic model to simulate the neutrons transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using Monte Carlo method for the propagation of neutrons in different environment. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational processing time introduced the variable control volume along the (pseudo-) periodic boundary conditions in order to overcome this problem. The choice of class physical Monte Carlo is due to the fact that it can decompose into simpler constituents the problem of solve a transport equation. The components may be treated separately, these are the propagation and interaction while respecting the laws of energy conservation and momentum, and the relationships that determine the probability of their interaction. We are aware of the fact that the problem approached in this thesis is far from being comparable to building a nuclear reactor, but this discussion the main target was to develop the Monte Carlo model, implement the code in a computer language that allows extensions of modular way. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude the need to consider the energy dependence, i.e. an spectral effective multiplication factor should be introduced each energy group separately. (author)
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.
Neutron-induced cross-sections via the surrogate method
International Nuclear Information System (INIS)
Boutoux, G.
2011-11-01
The surrogate reaction method is an indirect way of determining neutron-induced cross sections through transfer or inelastic scattering reactions. This method presents the advantage that in some cases the target material is stable or less radioactive than the material required for a neutron-induced measurement. The method is based on the hypothesis that the excited nucleus is a compound nucleus whose decay depends essentially on its excitation energy and on the spin and parity state of the populated compound state. Nevertheless, the spin and parity population differences between the compound-nuclei produced in the neutron and transfer-induced reactions may be different. This work reviews the surrogate method and its validity. Neutron-induced fission cross sections obtained with the surrogate method are in general good agreement. However, it is not yet clear to what extent the surrogate method can be applied to infer radiative capture cross sections. We performed an experiment to determine the gamma decay probabilities for 176 Lu and 173 Yb by using the surrogate reactions 174 Yb( 3 He,pγ) 176 Lu * and 174 Yb( 3 He,αγ) 173 Yb * , respectively, and compare them with the well-known corresponding probabilities obtained in the 175 Lu(n,γ) and 172 Yb(n,γ) reactions. This experiment provides answers to understand why, in the case of gamma-decay, the surrogate method gives significant deviations compared to the corresponding neutron-induced reaction. In this work, we have also assessed whether the surrogate method can be applied to extract capture probabilities in the actinide region. Previous experiments on fission have also been reinterpreted. Thus, this work provides new insights into the surrogate method. This work is organised in the following way: in chapter 1, the theoretical aspects related to the surrogate method will be introduced. The validity of the surrogate method will be investigated by means of statistical model calculations. In chapter 2, a review on
MCViNE – An object oriented Monte Carlo neutron ray tracing simulation package
Energy Technology Data Exchange (ETDEWEB)
Lin, Jiao Y.Y., E-mail: linjiao@ornl.gov [Caltech Center for Advanced Computing Research, California Institute of Technology (United States); Department of Applied Physics and Materials Science, California Institute of Technology (United States); Neutron Data Analysis and Visualization Division, Oak Ridge National Laboratory (United States); Smith, Hillary L. [Department of Applied Physics and Materials Science, California Institute of Technology (United States); Granroth, Garrett E., E-mail: granrothge@ornl.gov [Neutron Data Analysis and Visualization Division, Oak Ridge National Laboratory (United States); Abernathy, Douglas L.; Lumsden, Mark D.; Winn, Barry; Aczel, Adam A. [Quantum Condensed Matter Division, Oak Ridge National Laboratory (United States); Aivazis, Michael [Caltech Center for Advanced Computing Research, California Institute of Technology (United States); Fultz, Brent, E-mail: btf@caltech.edu [Department of Applied Physics and Materials Science, California Institute of Technology (United States)
2016-02-21
MCViNE (Monte-Carlo VIrtual Neutron Experiment) is an open-source Monte Carlo (MC) neutron ray-tracing software for performing computer modeling and simulations that mirror real neutron scattering experiments. We exploited the close similarity between how instrument components are designed and operated and how such components can be modeled in software. For example we used object oriented programming concepts for representing neutron scatterers and detector systems, and recursive algorithms for implementing multiple scattering. Combining these features together in MCViNE allows one to handle sophisticated neutron scattering problems in modern instruments, including, for example, neutron detection by complex detector systems, and single and multiple scattering events in a variety of samples and sample environments. In addition, MCViNE can use simulation components from linear-chain-based MC ray tracing packages which facilitates porting instrument models from those codes. Furthermore it allows for components written solely in Python, which expedites prototyping of new components. These developments have enabled detailed simulations of neutron scattering experiments, with non-trivial samples, for time-of-flight inelastic instruments at the Spallation Neutron Source. Examples of such simulations for powder and single-crystal samples with various scattering kernels, including kernels for phonon and magnon scattering, are presented. With simulations that closely reproduce experimental results, scattering mechanisms can be turned on and off to determine how they contribute to the measured scattering intensities, improving our understanding of the underlying physics.
Energy Technology Data Exchange (ETDEWEB)
Mouatassim, S
1994-07-01
The modular neutron detector is dedicated to the study of heavy ion reaction mechanisms. Monte Carlo simulations are performed for the optimization of the NE213 scintillator cell size and the general geometrical setup for the DEMON multidetector of neutrons with a minimum of cross-talk. Tests are performed with various types of photomultiplier tubes and scintillators. Using high energy neutron beams, more than six different reaction processes were identified with pulse shape discrimination by the charge comparison method. Cross sections were estimated. Light yields of charged particles p, d, t and alpha in the NE213 organic scintillator were analyzed using different theoretical approaches, and the intrinsic efficiency of the DEMON`s modules was measured and compared to Monte Carlo calculations. The DEMON experimental filter was simulated and has been associated with the Gemini physical events generator to study the performance of such a multidetector. Thus, the DEMON response for neutron evaporation of excited nuclei and its influence on energy measurement and temperature determination were studied. The same filter was used to simulate pre- and post-fission emission of neutrons for the fission process of the composite {sup 126}Ba system formed in the {sup 19}F + {sup 107}Ag entrance channel. (from author) 70 figs., 99 refs.
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.
A new method to evaluate neutron spectra for bnct
International Nuclear Information System (INIS)
Martin Hernandez, Guido
2001-01-01
This paper deals with the development of a method to evaluate neutron spectra for BNCT. Physical dose deposition calculations for different neutron energies, ranging from thermal to fast, were performed. A matrix, containing dose for each energy and position in the beam center line was obtained. MCNP 4B and Snyder's head model were used. A simple computer code containing the matrix calculates the dose for each point in the beam center line depending on the input energy spectrum to be evaluated. The output of this program is the dose distribution in the brain and the dose gain, that is the ratio between dose to tumor and maximum dose to healthy tissue maximum
Neutron radiative capture methods for surface elemental analysis
Trombka, J.I.; Senftle, F.; Schmadebeck, R.
1970-01-01
Both an accelerator and a 252Cf neutron source have been used to induce characteristic gamma radiation from extended soil samples. To demonstrate the method, measurements of the neutron-induced radiative capture and activation gamma rays have been made with both Ge(Li) and NaI(Tl) detectors, Because of the possible application to space flight geochemical analysis, it is believed that NaI(Tl) detectors must be used. Analytical procedures have been developed to obtain both qualitative and semiquantitative results from an interpretation of the measured NaI(Tl) pulse-height spectrum. Experiment results and the analytic procedure are presented. ?? 1970.
Phosphorus analysis in milk samples by neutron activation analysis method
International Nuclear Information System (INIS)
Oliveira, R.M. de; Cunha, I.I.L.
1991-01-01
The determination of phosphorus in milk samples by instrumental thermal neutron activation analysis is described. The procedure involves a short irradiation in a nuclear reactor and measurement of the beta radiation emitted by phosphorus - 32 after a suitable decay period. The sources of error were studied and the established method was applied to standard reference materials of known phosphorus content. (author)
Benchmarking the inelastic neutron scattering soil carbon method
The herein described inelastic neutron scattering (INS) method of measuring soil carbon was based on a new procedure for extracting the net carbon signal (NCS) from the measured gamma spectra and determination of the average carbon weight percent (AvgCw%) in the upper soil layer (~8 cm). The NCS ext...
Domain decomposition method for solving the neutron diffusion equation
International Nuclear Information System (INIS)
Coulomb, F.
1989-03-01
The aim of this work is to study methods for solving the neutron diffusion equation; we are interested in methods based on a classical finite element discretization and well suited for use on parallel computers. Domain decomposition methods seem to answer this preoccupation. This study deals with a decomposition of the domain. A theoretical study is carried out for Lagrange finite elements and some examples are given; in the case of mixed dual finite elements, the study is based on examples [fr
Apparatus and method for detecting contraband using fast neutron activation
International Nuclear Information System (INIS)
Gozani, T.; Sawa, Z.P.; Shea, P.M.
1992-01-01
This patent describes a method of detecting contraband within an object under investigation. It comprises: generating a beam of case neutrons; irradiating the object with the beam of fast neutrons, the fast neutrons interacting with atomic nuclei of the elements contained within the object to produce a gamma-ray spectrum having spectral lines characteristic of the elements contained within the object; measuring the spectral lines of the gamma-ray spectrum using a multiplicity of gamma-ray detectors judiciously positioned around the object; detecting the number of neutrons that pass through the object without interacting substantially with atomic nuclei within the object; determining the spatial and density distributions of the atomic nuclei of the elements contained within the object from the measured gamma-ray spectrum obtained from the multiplicity of gamma-ray detectors and the number of neutrons that pass through the object; comparing the measured spatial and density distributions of the atomic nuclei of the elements within the object with known spatial and density distributions of atomic nuclei for elements characteristic of contraband; and determining that contraband is present within the object when the comparison indicates a substantial match
Determination of fast neutrons energy spectra by Monte-Carlo Method
International Nuclear Information System (INIS)
Chetaine, A.
1986-01-01
Two computation codes based on the Monte-Carlo method are established for studying the spectrometry of neutrons with 14 Mev as initial energy. The spectra are determined, on one hand, around a neutron generator Ti-T target and, on the other hand, in a big paraffin cylinder. One code allows to determine the spectrum of neutrons irradiating the sample at various distances from the Ti-T target versus accelerator parameters: high voltage, atomic or molecular nature of deuterons beam, target thickness and materials surrounding the target. The other code determines neutron spectra at various positions inside and outside the 30 x 30 cm paraffin cylinder. The validity of the procedure used in these codes is verified by determining the spectrum of neutrons crossing a big surface, using the procedure in question and using direct simulation method. The biasing procedure used in the two codes permits to have results with good statistics from a reduced number of drawings. 70 figs.; 62 refs.; 1 tab. (author)
Activation method for measuring the neutron spectra parameters. Computer software
International Nuclear Information System (INIS)
Efimov, B.V.; Ionov, V.S.; Konyaev, S.I.; Marin, S.V.
2005-01-01
The description of mathematical statement of a task for definition the spectral characteristics of neutron fields with use developed in RRC KI unified activation detectors (UKD) is resulted. The method of processing of results offered by authors activation measurements and calculation of the parameters used for an estimation of the neutron spectra characteristics is discussed. Features of processing of the experimental data received at measurements of activation with using UKD are considered. Activation detectors UKD contain a little bit specially the picked up isotopes giving at irradiation peaks scale of activity in the common spectrum scale of activity. Computing processing of results of the measurements is applied on definition of spectrum parameters for nuclear reactor installations with thermal and close to such power spectrum of neutrons. The example of the data processing, the measurements received at carrying out at RRC KI research reactor F-1 is resulted [ru
Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility
International Nuclear Information System (INIS)
Beller, D.E.; Ott, K.O.; Terry, W.K.
1987-01-01
A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented
International Nuclear Information System (INIS)
Kurosawa, Tadahiro; Nakamura, Takashi; Suzuki, Takashi; Okano, Yasuhiro; Chisaka, Haruo
1998-01-01
We have embarked upon the development of a new detection method for underground water leakage using a neutron backscattering system. We have estimated the performance capabilities of such a system using Monte Carlo simulation. It is indicated that a leak which results in 40% water content in the surrounding soils could be detected at depths of up to 40 cm from the surface to the center of the source of leakage. This new detection system could be useful as a hydrostat of underground water supply in noisy areas such as Tokyo, in place of presently-used hydrostats which are based on detection of changes in sound
A two-dimensional simulator of the neutronic behaviour of low power fast reactors
International Nuclear Information System (INIS)
Penha, M.A.V.R. da.
1984-01-01
A model to simulate the temporal neutronic behaviour of fast breeder reactors was developed. The effective cross-sections are corrected, whenever the reactor state change; by using linear correlations and interpolation schemes with data contained in a library previously compiled. This methodology was coupled with a simplified spatial neutronic calculation to investigate the temporal behaviour of neutronic parameters such as breeding gain, flux and power. (Author) [pt
A finite element method for neutron transport
International Nuclear Information System (INIS)
Ackroyd, R.T.
1983-01-01
A completely boundary-free maximum principle for the first-order Boltzmann equation is derived from the completely boundary-free maximum principle for the mixed-parity Boltzmann equation. When continuity is imposed on the trial function for directions crossing interfaces the completely boundary-free principle for the first-order Boltzmann equation reduces to a maximum principle previously established directly from first principles and indirectly by the Euler-Lagrange method. Present finite element methods for the first-order Boltzmann equation are based on a weighted-residual method which permits the use of discontinuous trial functions. The new principle for the first-order equation can be used as a basis for finite-element methods with the same freedom from boundary conditions as those based on the weighted-residual method. The extremum principle as the parent of the variationally-derived weighted-residual equations ensures their good behaviour. (author)
252Cf-source-driven neutron noise analysis method
International Nuclear Information System (INIS)
Mihalczo, J.T.; King, W.T.; Blakeman, E.D.
1985-01-01
The 252 Cf-source-driven neutron noise analysis method has been tested in a a wide variety of experiments that have indicated the broad range of applicability of the method. The neutron multiplication factor, k/sub eff/ has been satisfactorily determined for a variety of materials including uranium metal, light water reactor fuel pins, fissile solutions, fuel plates in water, and interacting cylinders. For a uranyl nitrate solution tank which is typical of a fuel processing or reprocessing plant, the k/sub eff/ values were satisfactorily determined for values between 0.92 and 0.5 using a simple point kinetics interpretation of the experimental data. The short measurement times, in several cases as low as 1 min, have shown that the development of this method can lead to a practical subcriticality monitor for many in-plant applications. The further development of the method will require experiments and the development of theoretical methods to predict the experimental observables
International Nuclear Information System (INIS)
Freedman, M.S.; Peshkin, M.; Ringo, G.R.; Dombeck, T.W.
1989-08-01
The use of an ultracold neutron interferometer incorporating an electrostatic accelerator having a strong electric field gradient to accelerate neutrons by their possible electric dipole moment is proposed as a method of measuring the neutron electric dipole moment. The method appears to have the possibility of extending the sensitivity of the measurement by several orders of magnitude, perhaps to 10 -30 e-cm. 9 refs., 3 figs
International Nuclear Information System (INIS)
Peng, Xingjie; Li, Qing; Wang, Kan
2015-01-01
Highlights: • Three methods are utilized separately to unfold the control rod position from the fixed in-core neutron detector measurements. • Fixed in-core neutron detector measurements are simulated by neutronics code SMART. • Numerical results show that all these methods can unfold the control rod position accurately. • Two correction strategies are proposed to correct the simulated fixed in-core detector signals. - Abstract: Nuclear reactor core power distribution on-line monitoring system is very important in core surveillance, and this system should have the ability to indicate some abnormal conditions, such as the unacceptable control rod misalignment. In this study, the methodologies of radial basis function neural network (RBFNN), group method of data handling (GMDH) and Levenberg–Marquardt (LM) algorithm are utilized separately to unfold the control rod position from the fixed in-core neutron detector measurements. For using these methods, a large number of in-core detector signals corresponding to various known rod positions are needed. These data can be generated by an advanced core calculation code. In this study, the neutronics code SMART was used. The simulation results show that all these methods can unfold the control rod position accurately, and the performance comparison shows that the regularized RBFNN performs best. Two correction strategies are proposed to correct the simulated fixed in-core detector signals and improve the rod position monitoring accuracy when there are mismatches between actual physical factors and modeled physical factors
Simulations and developments of the Low Energy Neutron detector Array LENA
International Nuclear Information System (INIS)
Langer, C.; Algora, A.; Couture, A.; Csatlós, M.; Gulyás, J.; Heil, M.; Krasznahorkay, A.; O'Donnell, J.M.; Plag, R.; Reifarth, R.; Stuhl, L.; Sonnabend, K.; Tornyi, T.; Tovesson, F.
2011-01-01
Prototypes of the Low Energy Neutron detector Array (LENA) have been tested and compared with detailed GEANT simulations. LENA will consist of plastic scintillation bars with the dimensions 1000×45×10 mm 3 . The tests have been performed with γ-ray sources and neutrons originating from the neutron-induced fission of 235 U. The simulations agreed very well with the measured response and were therefore used to simulate the response to mono-energetic neutrons with different detection thresholds. LENA will be used to detect low-energy neutrons from (p,n)-type reactions with low momentum transfer foreseen at the R 3 B and EXL setups at FAIR, Darmstadt.
IB: A Monte Carlo simulation tool for neutron scattering instrument design under PVM and MPI
International Nuclear Information System (INIS)
Zhao Jinkui
2011-01-01
Design of modern neutron scattering instruments relies heavily on Monte Carlo simulation tools for optimization. IB is one such tool written in C++ and implemented under Parallel Virtual Machine and the Message Passing Interface. The program was initially written for the design and optimization of the EQ-SANS instrument at the Spallation Neutron Source. One of its features is the ability to group simple instrument components into more complex ones at the user input level, e.g. grouping neutron mirrors into neutron guides and curved benders. The simulation engine manages the grouped components such that neutrons entering a group are properly operated upon by all components, multiple times if needed, before exiting the group. Thus, only a few basic optical modules are needed at the programming level. For simulations that require higher computer speeds, the program can be compiled and run in parallel modes using either the PVM or the MPI architectures.
Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility
International Nuclear Information System (INIS)
Beller, D.E.
1986-01-01
A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies that could affect results are insignificant. Inelastic scattering cross sections are identified as a major source of these discrepancies. The conceptual design of this FBBR analog, the fusion reactor blanket facility (FRBF), is presented. Essential features are a cylindrical geometry and a distributed, cosine-shaped line source of 14-MeV neutrons. This source can be created by sweeping a deuteron beam over an elongated titanium-tritide target. To demonstrate that the design of the FRBF will not contribute significant deviations in experimental results, neutronics analyses were performed: results of comparisons of 2-dimensional to 1-dimensional predictions are reported for two blanket compositions. Expected deviations from 1-D predictions which are due to source anisotropy and blanket asymmetry are minimal. Thus, design of the FRBF allows simple and straightforward interpretation of the experimental results, without a need for coarse 3-D calculations
Simulation of a parallel processor on a serial processor: The neutron diffusion equation
International Nuclear Information System (INIS)
Honeck, H.C.
1981-01-01
Parallel processors could provide the nuclear industry with very high computing power at a very moderate cost. Will we be able to make effective use of this power. This paper explores the use of a very simple parallel processor for solving the neutron diffusion equation to predict power distributions in a nuclear reactor. We first describe a simple parallel processor and estimate its theoretical performance based on the current hardware technology. Next, we show how the parallel processor could be used to solve the neutron diffusion equation. We then present the results of some simulations of a parallel processor run on a serial processor and measure some of the expected inefficiencies. Finally we extrapolate the results to estimate how actual design codes would perform. We find that the standard numerical methods for solving the neutron diffusion equation are still applicable when used on a parallel processor. However, some simple modifications to these methods will be necessary if we are to achieve the full power of these new computers. (orig.) [de
International Nuclear Information System (INIS)
Liu Dekun; Zhang Hongyu; Zhang Lihong; Dong Huan; Gu Deshan
2012-01-01
An important index in assessment of coal quality is low calorific value. Using neutron to analysis coal quality, the more the coal moisture content, especially the increasing of external moisture will reduce the low calorific value. The principle of coal quality analysis by neutron prompt Gamma-ray is introduced. The influence of the gamma count of the carbon element peak with increasing external moisture in coal samples was simulated using MCNP code. And discussed the reasons how external moisture content influence the calorific value. Simulation results indicate that with the increasing of external moisture in the coal samples, the gamma count of the carbon element peak dwindling, and the low calorific value reducing. The conclusion is : using neutrons method to analysis coal quality, the more external moisture content, the larger error of the measurement results of the carbon element, and will influence the calculation accuracy of the low calorific value. (authors)
International Nuclear Information System (INIS)
Mouatassim, S.
1994-07-01
The modular neutron detector is dedicated to the study of heavy ion reaction mechanisms. Monte Carlo simulations are performed for the optimization of the NE213 scintillator cell size and the general geometrical setup for the DEMON multidetector of neutrons with a minimum of cross-talk. Tests are performed with various types of photomultiplier tubes and scintillators. Using high energy neutron beams, more than six different reaction processes were identified with pulse shape discrimination by the charge comparison method. Cross sections were estimated. Light yields of charged particles p, d, t and alpha in the NE213 organic scintillator were analyzed using different theoretical approaches, and the intrinsic efficiency of the DEMON's modules was measured and compared to Monte Carlo calculations. The DEMON experimental filter was simulated and has been associated with the Gemini physical events generator to study the performance of such a multidetector. Thus, the DEMON response for neutron evaporation of excited nuclei and its influence on energy measurement and temperature determination were studied. The same filter was used to simulate pre- and post-fission emission of neutrons for the fission process of the composite 126 Ba system formed in the 19 F + 107 Ag entrance channel. (from author) 70 figs., 99 refs
Simulation analysis of radiation fields inside phantoms for neutron irradiation
International Nuclear Information System (INIS)
Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Y.; Miyahara, N.
2007-01-01
Radiation fields inside phantoms have been calculated for neutron irradiation. Particle and heavy-ion transport code system PHITS was employed for the calculation. Energy and size dependences of neutron dose were analyzed using tissue equivalent spheres of different size. A voxel phantom of mouse was developed based on CT images of an 8-week-old male C3H/HeNs mouse. Deposition energy inside the mouse was calculated for 2- and 10-MeV neutron irradiation. (author)
Ab initio simulations and neutron scattering studies of structure and dynamics in PdH
International Nuclear Information System (INIS)
Totolici, I.E.
2001-07-01
The work presented in this PhD thesis is concerned with the interpretation of the neutron scattering measurements from the palladium hydrogen system by means of ab initio electronic structure calculations. The motivation of performing such calculations was due to recent neutron scattering studies on this system that showed a strong directional dependence to the dynamical structure factor together with a complex dependence on energy. Here we attempt to describe the origin of these features by ab initio simulations of the dynamical structure factor. The method assumes an adiabatic separation of the motion of the proton and palladium atoms. The proton wave functions are calculated by a direct solution of the associated single-particle Schroedinger equation using a plane wave basis set method and a mapping of the adiabatic surface. The Fourier components of the adiabatic potential are obtained from LDA pseudopotential calculations. Using Fermi's golden rule within the Born approximation we were then able to calculate the dynamical structure factor, S(Q,ω), for exciting the proton from its ground state to various excited states as a function of the magnitude and direction of the scattering wave vector. The results are in agreement with the inelastic neutron scattering spectra and allow us to identify the origin of previous inexplicable features, in particular the strong directional dependence to the experimental data. The method was extended to investigate the expansion of the equilibrium lattice constant as a function of the H isotope when the zero-point energy of the proton/deuterium is explicitly taken into account in the relaxation process. The results we obtained predicted a bigger lattice constant for the hydride, as expected. Furthermore, other complex ab initio calculations were carried out in order to describe the origin of the large optic dispersion, seen previously in the coherent neutron scattering data. Our calculated dispersion proved to be in good
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)
Energy Technology Data Exchange (ETDEWEB)
An, So Hyun; Lee, Young Ouk; Lee, Cheol Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Young Seok [National Fusion Research Institute, Daejeon (Korea, Republic of)
2007-10-15
It is essential that neutron cross sections are measured precisely for many areas of research and technique. In Korea, these experiments have been performed in the Pohang Neutron Facility (PNF) with the pulsed neutron facility based on the 100 MeV electron linear accelerator. In PNF, the neutron energy spectra have been measured for different water levels inside the moderator and compared with the results of the MCNPX calculation. The optimum size of the water moderator has been determined on the base of these results. In this study, Monte Carlo simulations for the TOF technique were performed and neutron spectra of neutrons were calculated to predict the measurements.
McStas 1.1: A tool for building neutron Monte Carlo simulations
DEFF Research Database (Denmark)
Lefmann, K.; Nielsen, K.; Tennant, D.A.
2000-01-01
McStas is a project to develop general tools for the creation of simulations of neutron scattering experiments. In this paper, we briefly introduce McStas and describe a particular application of the program: the Monte Carlo calculation of the resolution function of a standard triple-axis neutron...
Simulations of muon-induced neutron flux at large depths underground
International Nuclear Information System (INIS)
Kudryavtsev, V.A.; Spooner, N.J.C.; McMillan, J.E.
2003-01-01
The production of neutrons by cosmic-ray muons at large depths underground is discussed. The most recent versions of the muon propagation code MUSIC, and particle transport code FLUKA are used to evaluate muon and neutron fluxes. The results of simulations are compared with experimental data
Simulation of Rossi-α method with analog Monte-Carlo method
International Nuclear Information System (INIS)
Lu Yuzhao; Xie Qilin; Song Lingli; Liu Hangang
2012-01-01
The analog Monte-Carlo code for simulating Rossi-α method based on Geant4 was developed. The prompt neutron decay constant α of six metal uranium configurations in Oak Ridge National Laboratory were calculated. α was also calculated by Burst-Neutron method and the result was consistent with the result of Rossi-α method. There is the difference between results of analog Monte-Carlo simulation and experiment, and the reasons for the difference is the gaps between uranium layers. The influence of gaps decrease as the sub-criticality deepens. The relative difference between results of analog Monte-Carlo simulation and experiment changes from 19% to 0.19%. (authors)
Beam dynamics simulation of the Spallation Neutron Source linear accelerator
International Nuclear Information System (INIS)
Takeda, H.; Billen, J.H.; Bhatia, T.S.
1998-01-01
The accelerating structure for Spallation Neutron Source (SNS) consists of a radio-frequency-quadrupole-linac (RFQ), a drift-tube-linac (DTL), a coupled-cavity-drift-tube-linac (CCDTL), and a coupled-cavity-linac (CCL). The linac is operated at room temperature. The authors discuss the detailed design of linac which accelerates an H - pulsed beam coming out from RFQ at 2.5 MeV to 1000 MeV. They show a detailed transition from 402.5 MHz DTL with a 4 βλ structure to a CCDTL operated at 805 MHz with a 12 βλ structure. After a discussion of overall feature of the linac, they present an end-to-end particle simulation using the new version of the PARMILA code for a beam starting from the RFQ entrance through the rest of the linac. At 1000 MeV, the beam is transported to a storage ring. The storage ring requires a large (±500-keV) energy spread. This is accomplished by operating the rf-phase in the last section of the linac so the particles are at the unstable fixed point of the separatrix. They present zero-current phase advance, beam size, and beam emittance along the entire linac
Methods and applications in high flux neutron imaging
International Nuclear Information System (INIS)
Ballhausen, H.
2007-01-01
This treatise develops new methods for high flux neutron radiography and high flux neutron tomography and describes some of their applications in actual experiments. Instead of single images, time series can be acquired with short exposure times due to the available high intensity. To best use the increased amount of information, new estimators are proposed, which extract accurate results from the recorded ensembles, even if the individual piece of data is very noisy and in addition severely affected by systematic errors such as an influence of gamma background radiation. The spatial resolution of neutron radiographies, usually limited by beam divergence and inherent resolution of the scintillator, can be significantly increased by scanning the sample with a pinhole-micro-collimator. This technique circumvents any limitations in present detector design and, due to the available high intensity, could be successfully tested. Imaging with scattered neutrons as opposed to conventional total attenuation based imaging determines separately the absorption and scattering cross sections within the sample. For the first time even coherent angle dependent scattering could be visualized space-resolved. New applications of high flux neutron imaging are presented, such as materials engineering experiments on innovative metal joints, time-resolved tomography on multilayer stacks of fuel cells under operation, and others. A new implementation of an algorithm for the algebraic reconstruction of tomography data executes even in case of missing information, such as limited angle tomography, and returns quantitative reconstructions. The setup of the world-leading high flux radiography and tomography facility at the Institut Laue-Langevin is presented. A comprehensive appendix covers the physical and technical foundations of neutron imaging. (orig.)
Industrial applications of neutron physics methods
International Nuclear Information System (INIS)
Gozani, T.
1994-01-01
Three areas where nuclear based techniques have significant are briefly described. These are: Nuclear material control and non-proliferation, on-line elemental analysis of coal and minerals, and non- detection of explosives and other contraband. The nuclear physics and the role of reactor physics methods are highlighted. (author). 5 refs., 10 figs., 5 tabs
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
Žukauskaite, A; Plukiene, R; Plukis, A
2007-01-01
Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.
Inversion methods for analysis of neutron brightness measurements in tokamaks
International Nuclear Information System (INIS)
Gorini, G.; Gottardi, N.
1990-02-01
The problem of determining neutron emissivity from neutron brightness measurements in magnetic fusion plasmas is addressed. In the case of two-dimensional measurements with two orthogonal cameras, a complete, tomographic analysis of the data can in principle be performed. The results depend critically on the accuracy of the measurements and alternative solutions can be sought under the assumption of a known emissivity topology (Generalized Abel Inversion). In this work, neutron brightness data from the JET tokamak have been studied with both methods. We find that with the present experimental uncertainty (levels 10-20%) the Abel inversion method works best, while two-dimensional information cannot in general be deduced. This is confirmed by studies of the error propagation in the inversion using artificial data, which are also presented here. An important application of emissivity profile information is the determination of the plasma deuterium temperature profile, T D (R). Results are presented here from the analysis of JET data and the errors in T D (R) are discussed in some detail. It is found that, for typical JET plasma conditions, the dominant source of uncertainty arises from the high plasma impurity level and the fact that it is poorly known; these problems can be expected to be remedied and neutron brightness measurements would be expected to be very effective (especially in high density plasmas) as a T D (R) diagnostics. (author)
International Nuclear Information System (INIS)
Raoux, A. C.; Loridon, J.; Mariani, A.; Passard, C.
2009-01-01
Active neutron measurements such as the Differential Die-Away (DDA) technique involving pulsed neutron generator, are widely applied to determine the fissile content of waste packages. Unfortunately, the main drawback of such techniques is coming from the lack of knowledge of the waste matrix composition. Thus, the matrix effect correction for the DDA measurement is an essential improvement in the field of fissile material content determination. Different solutions have been developed to compensate the effect of the matrix on the neutron measurement interpretation. In this context, this paper describes an innovative matrix correction method we have developed with the goal of increasing the accuracy of the matrix effect correction and reducing the measurement time. The implementation of this method is based on the analysis of the raw signal with an optimisation algorithm called the simulated annealing algorithm. This algorithm needs a reference data base of Multi-Channel Scaling (MCS) spectra, to fit the raw signal. The construction of the MCS library involves a learning phase to define and acquire the DDA signals. This database has been provided by a set of active signals from experimental matrices (mock-up waste drums of 118 litres) recorded in a specific device dedicated to neutron measurement research and development of the Nuclear Measurement Laboratory of CEA-Cadarache, called PROMETHEE 6. The simulated annealing algorithm is applied to make use of the effect of the matrices on the total active signal of DDA measurement. Furthermore, as this algorithm is directly applied to the raw active signal, it is very useful when active background contributions can not be easily estimated and removed. Most of the cases tested during this work which represents the feasibility phase of the method, are within a 4% agreement interval with the expected experimental value. Moreover, one can notice that without any compensation of the matrix effect, the classical DDA prompt
Energy Technology Data Exchange (ETDEWEB)
Raoux, A. C.; Loridon, J.; Mariani, A.; Passard, C. [French Atomic Energy Commission, DEN, Cadarache, F-3108 Saint-Paul-Lez-Durance (France)
2009-07-01
Active neutron measurements such as the Differential Die-Away (DDA) technique involving pulsed neutron generator, are widely applied to determine the fissile content of waste packages. Unfortunately, the main drawback of such techniques is coming from the lack of knowledge of the waste matrix composition. Thus, the matrix effect correction for the DDA measurement is an essential improvement in the field of fissile material content determination. Different solutions have been developed to compensate the effect of the matrix on the neutron measurement interpretation. In this context, this paper describes an innovative matrix correction method we have developed with the goal of increasing the accuracy of the matrix effect correction and reducing the measurement time. The implementation of this method is based on the analysis of the raw signal with an optimisation algorithm called the simulated annealing algorithm. This algorithm needs a reference data base of Multi-Channel Scaling (MCS) spectra, to fit the raw signal. The construction of the MCS library involves a learning phase to define and acquire the DDA signals. This database has been provided by a set of active signals from experimental matrices (mock-up waste drums of 118 litres) recorded in a specific device dedicated to neutron measurement research and development of the Nuclear Measurement Laboratory of CEA-Cadarache, called PROMETHEE 6. The simulated annealing algorithm is applied to make use of the effect of the matrices on the total active signal of DDA measurement. Furthermore, as this algorithm is directly applied to the raw active signal, it is very useful when active background contributions can not be easily estimated and removed. Most of the cases tested during this work which represents the feasibility phase of the method, are within a 4% agreement interval with the expected experimental value. Moreover, one can notice that without any compensation of the matrix effect, the classical DDA prompt
International Nuclear Information System (INIS)
Mori, Chizuo; Gotoh, Junichi; Uritani, Akira; Miyahara, Hiroshi; Ikeda, Yuichiro; Kasugai, Yoshimi; Kaneko, Junichi
1998-01-01
There are many types of fast neutron spectrometers for plasma temperature diagnostics, 28 Si(n,α) 25 Mg reaction giving the energy resolution of 2.2% for 14 MeV neutrons, the 12 C(n,α) 9 Be reaction giving the resolution of 2.15%. These detectors, however suffer from radiation damage, which demands to exchange the detector to a new one in every a few month depending on the usage. Recoil proton method has also been developed by using liquid scintillator or plastic scintillator, as a neutron-to-proton converter in front of a Si-detector, which is called counter telescope type, giving a resolution of 4.0%. This type of spectrometer can reduce radiation damage by placing Si-detector at outside Neutron beam. The scintillator can measure the lost energy of protons in the converter (i.e. the scintillator) and the measured energy loss can be used for improving the energy resolution. However, the energy resolution of organic scintillator itself is generally not so good. We proposed to use a proportional counter with CH 4 as counting gas and also as a neutron-proton converter, which has far better energy resolution than plastic scintillators, although the time resolution of counting in proportional counters is generally inferior to that in organic scintillation counters. The characteristics of the new spectrometer were experimentally studied and also were simulated with analytical calculation. (author)
International Nuclear Information System (INIS)
Steyerl, A.; Malik, S. S.; Desai, A. M.; Kaufman, C.
2010-01-01
We review the diffuse scattering and the loss coefficient in ultracold neutron reflection from slightly rough surfaces, report a surprising reduction in loss coefficient due to roughness, and discuss the possibility of transition from quantum treatment to ray optics. The results are used in a computer simulation of neutron storage in a recent neutron lifetime experiment that reported a large discrepancy of neutron lifetime with the current particle data value. Our partial reanalysis suggests the possibility of systematic effects that were not included in this publication.
Comparison of Experiment and Simulation of the triple GEM-Based Fast Neutron Detector
International Nuclear Information System (INIS)
Wang Xiao-Dong; Luo Wen; Zhang Jun-Wei; Yang He-Run; Duan Li-Min; Lu Chen-Gui; Hu Rong-Jiang; Hu Bi-Tao; Zhang Chun-Hui; Yang Lei; Zhou Jian-Rong; An Lv-Xing
2015-01-01
A detector for fast neutrons based on a 10 × 10 cm"2 triple gas electron multiplier (GEM) device is developed and tested. A neutron converter, which is a high density polyethylene (HDPE) layer, is combined with the triple GEM detector cathode and placed inside the detector, in the path of the incident neutrons. The detector is tested by obtaining the energy deposition spectrum with an Am Be neutron source in the Institute of Modern Physics (IMP) at Lanzhou. In the present work we report the results of the tests and compare them with those of simulations. The transport of fast neutrons and their interactions with the different materials in the detector are simulated with the GEANT4 code, to understand the experimental results. The detector displays a clear response to the incident fast neutrons. However, an unexpected disagreement in the energy dependence of the response between the simulated and measured spectra is observed. The neutron sources used in our simulation include deuterium-tritium (DT, 14 MeV), deuterium-deuterium (DD, 2.45 MeV), and Am Be sources. The simulation results also show that among the secondary particles generated by the incident neutron, the main contributions to the total energy deposition are from recoil protons induced in hydrogen-rich HDPE or Kapton (GEM material), and activation photons induced by neutron interaction with Ar atoms. Their contributions account for 90% of the total energy deposition. In addition, the dependence of neutron deposited energy spectrum on the composition of the gas mixture is presented. (paper)
Neutron Transport Methods for Accelerator-Driven Systems
International Nuclear Information System (INIS)
Nicholas Tsoulfanidis; Elmer Lewis
2005-01-01
The objective of this project has been to develop computational methods that will enable more effective analysis of Accelerator Driven Systems (ADS). The work is centered at the University of Missouri at Rolla, with a subcontract at Northwestern University, and close cooperation with the Nuclear Engineering Division at Argonne National Laboratory. The work has fallen into three categories. First, the treatment of the source for neutrons originating from the spallation target which drives the neutronics calculations of the ADS. Second, the generalization of the nodal variational method to treat the R-Z geometry configurations frequently needed for scoping calculations in Accelerator Driven Systems. Third, the treatment of void regions within variational nodal methods as needed to treat the accelerator beam tube
A standardized method for beam design in neutron capture therapy
International Nuclear Information System (INIS)
Storr, G.J.: Harrington, B.V.
1993-01-01
A desirable end point for a given beam design for Neutron Capture Therapy (NCT) should be quantitative description of tumour control probability and normal tissue damage. Achieving this goal will ultimately rely on data from NCT human clinical trials. Traditional descriptions of beam designs have used a variety of assessment methods to quantify proposed or installed beam designs. These methods include measurement and calculation of open-quotes free fieldclose quotes parameters, such as neutron and gamma flux intensities and energy spectra, and figures-of-merit in tissue equivalent phantoms. The authors propose here a standardized method for beam design in NCT. This method would allow all proposed and existing NCT beam facilities to be compared equally. The traditional approach to determining a quantitative description of tumour control probability and normal tissue damage in NCT research may be described by the following path: Beam design → dosimetry → macroscopic effects → microscopic effects. Methods exist that allow neutron and gamma fluxes and energy dependence to be calculated and measured to good accuracy. By using this information and intermediate dosimetric quantities such as kerma factors for neutrons and gammas, macroscopic effect (absorbed dose) in geometries of tissue or tissue-equivalent materials can be calculated. After this stage, for NCT the data begins to become more sparse and in some areas ambiguous. Uncertainties in the Relative Biological Effectiveness (RBE) of some NCT dose components means that beam designs based on assumptions considered valid a few years ago may have to be reassessed. A standard method is therefore useful for comparing different NCT facilities
International Nuclear Information System (INIS)
Kobayashi, T.; Sakurai, Y.; Kanda, K.
2001-01-01
The neutron irradiation characteristics of the Heavy Water Neutron Irradiation Facility (HWNIF) at the Kyoto University Research Reactor Institute (KIJRRI) for boron neutron capture therapy (BNCT), is described. The present method of dose measurement and its evaluation at the KURRI, is explained. Especially, the special feature and noticeable matters were expounded for the BNCT with craniotomy, which has been applied at present only in Japan. (author)
Kinetic simulation of neutron production in a deuterium z-pinch
International Nuclear Information System (INIS)
Mostrom, C.; Stygar, William A.; Thoma, Carsten; Welch, Dale Robert; Clark, R.E.; Leeper, Ramon Joe; Rose, David V.
2010-01-01
We have found computationally that, at sufficiently high currents, half of the neutrons produced by a deuterium z pinch are thermonuclear in origin. Early experiments below 1-MA current found that essentially all of the neutrons produced by a deuterium pinch are not thermonuclear, but are initiated by an instability that creates beam-target neutrons. Many subsequent authors have supported this result while others have claimed that pinch neutrons are thermonuclear. To resolve this issue, we have conducted fully kinetic, collisional, and electromagnetic simulations of the complete time evolution of a deuterium pinch. We find that at 1-MA pinch currents, most of the neutrons are, indeed, beam-target in origin. At much higher current, half of the neutrons are thermonuclear and half are beam-target driven by instabilities that produce a power law fall off in the ion energy distribution function at large energy. The implications for fusion energy production with such pinches are discussed.
Method and apparatus for epithermal neutron decay logging
International Nuclear Information System (INIS)
Nelligan, W.B.
1998-01-01
The nature of hydrogenous fluids filling the pore spaces in formations surrounding a well bore are determined by irradiating the formations with bursts of high energy neutrons and using the ratio of time-dependent parameters related to the decay of epithermal neutron populations above two different energy levels to provide values indicative of the kind of fluid present, independent of porosity. The measurement above the higher of these energy levels, appr. 0.4eV, indicates the existence of hydrogenous fluid, water and hydrocarbons in the formations. The lower threshold measurement, above appr. 0.15eV, is indicative of the structure of the molecule in which the hydrogen molecule is bound and of the porosity. A pulsed neutron generator in a sonde irradiates the formations with 14meV neutrons and a pair of detectors, one shielded by cadmium, to establish the 0.4eV threshold energy level and the other, shielded by gadolinium, to establish the 0.15eV threshold energy level. Time related parameters of the count rate information, e.g. the decay constants, provided by each detector are derived. The ratio of the respective parameters is used to identify the type of fluid in the information, independent of porosity. The method and apparatus can determine water saturation, i.e. percentage of water in the formation fluid and can identify specific types of hydrocarbons under 100% hydrocarbon saturation conditions. 8 figs
Energy Technology Data Exchange (ETDEWEB)
Cambou, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1961-02-15
This research aimed at improving fast-neutron electronic detectors based on n-p elastic scattering. The first part concerns proportional counters; careful constructional methods have made it possible to plot mono-energetic neutron spectra in the range 700 keV - 3 MeV with a resolution of 7 per cent. The second part concerns scintillation counters: an organic scintillator and an inorganic scintillator covered with a thin layer of a scattering agent. An exact study of the types of scintillation has made it possible to develop efficient discriminator circuits. Different neutron spectra plotted in the presence of a strong gamma background are presented. The last part deals with the development of form discrimination methods for the study, in the actual beam, of the elastic scattering of 14.58 MeV electrons. With hydrogen, the distribution f ({phi}) of the recoil protons is f({phi}) = 1 + 0.034 cos {phi} + 0.042 cos{sup 2} {phi}. With tritium the scattering is strongly anisotropic; the curve representing the variation of the differential cross-section for the elastic scattering in the centre of mass system is obtained with a target containing 1 cm{sup 3} of tritium. (author) [French] Le travail a porte sur l'amelioration des detecteurs electroniques de neutrons rapides bases sur la diffusion elastique n-p. La premiere partie est relative aux compteurs proportionnels; des methodes soignees de fabrication ont permis des traces de spectres de neutrons monoenergetiques dans le domaine 700 keV - 3 MeV avec une resolution de 7 pour cent. La deuxieme partie est relative au compteur a scintillations; scintillateur organique et scintillateur mineral recouvert d'un diffuseur mince. Une etude precise des formes de scintillations a permis la mise au point de circuits discriminateurs efficaces. Differents spectres de neutrons traces en presence d'un fond gamma intense sont presentes. La derniere partie est relative a la mise en oeuvre des methodes de discrimination de forme pour l
International Nuclear Information System (INIS)
Liu, Shichang; Wang, Guanbo; Wu, Gaochen; Wang, Kan
2015-01-01
Highlights: • DRAGON and DONJON are applied and verified in calculations of research reactors. • Continuous-energy Monte Carlo calculations by RMC are chosen as the references. • “ECCO” option of DRAGON is suitable for the calculations of research reactors. • Manual modifications of cross-sections are not necessary with DRAGON and DONJON. • DRAGON and DONJON agree well with RMC if appropriate treatments are applied. - Abstract: Simulation of the behavior of the plate-type research reactors such as JRR-3M and CARR poses a challenge for traditional neutronics calculation tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity and large leakage of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON and DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic approach. The goal of this research is to examine the capability of the deterministic code system DRAGON and DONJON to reliably simulate the research reactors. The results indicate that the DRAGON and DONJON code system agrees well with the continuous-energy Monte Carlo simulation on both k eff and flux distributions if the appropriate treatments (such as the ECCO option) are applied
About application of the 'rough' net method for decision of the neutron transfer transient equation
International Nuclear Information System (INIS)
Seleznev, E.F.; Tarasenko, V.V.
1995-01-01
Method of the decision of a transient equation of the neutrons transfer is developed, which at preservation of necessary accuracy permits considerably to speed up a finding of the decision up to modeling of processes in reactor in real time. The transient equation of neutrons transfer in one-group diffusion approximation is decided by the finite-difference method. The calculating model of reactor is divided into rather large zones, where the currents on internal borders are away, and on external borders ones are a sum of currents on the borders of small-sized zones. For the decision of an equation in finite-difference kind the numerical scheme 'Time - integrate' is used, which permits to search the decision in a half-explicit kind with rather large temporary step. The decision for density of neutrons flux is determine by the SOR method. Under the conducted preliminary analysis of an algorithm efficiency it is possible to conclude, that the time of the decision on a computer can be reduced in 3 and more times, in depending on 'roughness' of a calculated net in comparison with computation on a complete net. The realized algorithm can be used as for scientific researches, and as neutron-physical block of the simulator. 6 refs., 1 fig
Neutron electric dipole moment in the instanton vacuum: Quenched versus unquenched simulations
International Nuclear Information System (INIS)
Faccioli, P.; Guadagnoli, D.; Simula, S.
2004-01-01
We investigate the role played by the fermionic determinant in the evaluation of the CP-violating neutron electric dipole moment (EDM) adopting the Instanton Liquid Model. Significant differences between quenched and unquenched calculations are found. In the case of unquenched simulations the neutron EDM decreases linearly with the quark mass and is expected to vanish in the chiral limit. On the contrary, within the quenched approximation, the neutron EDM increases as the quark mass decreases and is expected to diverge as 1/m N f in the chiral limit. We argue that such a qualitatively different behavior is a parameter-free, semiclassical prediction and occurs because the neutron EDM is sensitive to the topological structure of the vacuum. The present analysis suggests that quenched and unquenched lattice QCD simulations of the neutron EDM as well as of other observables governed by topology might show up important differences in the quark mass dependence for m q QCD
Measurement and simulation of neutron response function of organic liquid scintillator detector
International Nuclear Information System (INIS)
Gohil, M.; Banerjee, K.; Bhattacharya, S.; Bhattacharya, C.; Kundu, S.; Rana, T.K.; Mukherjee, G.; Meena, J.K.; Pandey, R.; Pai, H.; Ghosh, T.K.; Dey, A.; Mukhopadhyay, S.; Pandit, D.; Pal, S.; Banerjee, S.R.; Bandhopadhyay, T.
2012-01-01
Response functions of monoenergetic neutrons at various energies, corresponding to a measured neutron energy spectrum have been extracted. The experimental response functions for neutron energies in the range of ∼2-20 MeV have been compared with the respective GEANT4 predictions. It has been found that, there is some discrepancy between the experimental and the GEANT4 simulated neutron response functions at lower pulse height regions, which increases with the increase of neutron energy. This might be due to the incompleteness of the physics processes used in the present GEANT4 simulations. In particular, higher order reaction processes which become more significant at higher energies should be properly taken into account in the calculation of response function.
International Nuclear Information System (INIS)
Nenot, M.L.
2003-07-01
Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10 -2 Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)
Nodal spectrum method for solving neutron diffusion equation
International Nuclear Information System (INIS)
Sanchez, D.; Garcia, C. R.; Barros, R. C. de; Milian, D.E.
1999-01-01
Presented here is a new numerical nodal method for solving static multidimensional neutron diffusion equation in rectangular geometry. Our method is based on a spectral analysis of the nodal diffusion equations. These equations are obtained by integrating the diffusion equation in X, Y directions and then considering flat approximations for the current. These flat approximations are the only approximations that are considered in this method, as a result the numerical solutions are completely free from truncation errors. We show numerical results to illustrate the methods accuracy for coarse mesh calculations
Study on calculation methods for the effective delayed neutron fraction
International Nuclear Information System (INIS)
Irwanto, Dwi; Obara, Toru; Chiba, Go; Nagaya, Yasunobu
2011-03-01
The effective delayed neutron fraction β eff is one of the important neutronic parameters from a view point of a reactor kinetics. Several Monte-Carlo-based methods to estimate β eff have been proposed to date. In order to quantify the accuracy of these methods, we study calculation methods for β eff by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate β eff with various definitions such as the fundamental value β 0 , the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other. Through the present study, we find the following: The largest difference among the standard definition of β eff , Nauchi's β eff and Meulekamp's β eff is approximately 10%. The fundamental value β 0 is quite larger than the others in several cases. For all the cases, Meulekamp's β eff is always higher than Nauchi's β eff . This is because Nauchi's β eff considers the average neutron multiplicity value per fission which is large in the high energy range (1MeV-10MeV), while the definition of Meulekamp's β eff does not include this parameter. Furthermore, we evaluate the multi-generation effect on β eff values and demonstrate that this effect should be considered to obtain the standard definition values of β eff . (author)
Neutron stress measurement using image plate and the cos α method
International Nuclear Information System (INIS)
Sasaki, Toshihiko; Hirose, Yukio; Minakawa, Nobuaki; Morii, Yukio
2004-01-01
A new type of the neutron stress measurement method, in which the stress is determined by the cos α method, was studied. A neutron image plate was used in this study for detecting the diffraction image. Steel specimens of 5 mm thickness were used for the tensile stress test. The stresses obtain from the experiment showed one-to-one ratio with respect to the applied stresses, though the misfit stress of about 400 MPa was also observed in the result. A simple correction technique for calibrating raw data was proposed to obtain the true stress. A simulation study was carried out to check the experimental result on the peak positions in the diffraction ring and the stress calculation by the cos α method. (author)
Apparatus and method for the measurement of neutron moderating or absorbing properties of objects
International Nuclear Information System (INIS)
Untermyer, S.I.
1981-01-01
An apparatus and method for measuring the neutron moderating or absorbing properties of objects or materials is disclosed in which a fast neutron source cooperates with a neutron absorbing material which reduces the energy of the fast neutrons by inelastic scattering so that they can be readily thermalized by a moderator. A thermal neutron detector is disposed adjacent the material and serves to detect thermal neutrons emitted by a moderator placed to receive and thermalize the reduced energy neutrons. A material whose absorption is to be measured is placed between a moderator and the detector
A comparison of Nodal methods in neutron diffusion calculations
Energy Technology Data Exchange (ETDEWEB)
Tavron, Barak [Israel Electric Company, Haifa (Israel) Nuclear Engineering Dept. Research and Development Div.
1996-12-01
The nuclear engineering department at IEC uses in the reactor analysis three neutron diffusion codes based on nodal methods. The codes, GNOMERl, ADMARC2 and NOXER3 solve the neutron diffusion equation to obtain flux and power distributions in the core. The resulting flux distributions are used for the furl cycle analysis and for fuel reload optimization. This work presents a comparison of the various nodal methods employed in the above codes. Nodal methods (also called Coarse-mesh methods) have been designed to solve problems that contain relatively coarse areas of homogeneous composition. In the nodal method parts of the equation that present the state in the homogeneous area are solved analytically while, according to various assumptions and continuity requirements, a general solution is sought out. Thus efficiency of the method for this kind of problems, is very high compared with the finite element and finite difference methods. On the other hand, using this method one can get only approximate information about the node vicinity (or coarse-mesh area, usually a feel assembly of a 20 cm size). These characteristics of the nodal method make it suitable for feel cycle analysis and reload optimization. This analysis requires many subsequent calculations of the flux and power distributions for the feel assemblies while there is no need for detailed distribution within the assembly. For obtaining detailed distribution within the assembly methods of power reconstruction may be applied. However homogenization of feel assembly properties, required for the nodal method, may cause difficulties when applied to fuel assemblies with many absorber rods, due to exciting strong neutron properties heterogeneity within the assembly. (author).
The study of neutron transport by oscillation method
International Nuclear Information System (INIS)
Raievski, V.
1959-01-01
The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [fr
Energy Technology Data Exchange (ETDEWEB)
Raievski, V
1959-07-01
The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [French] La methode de modulation est un procede tres general d'etude des proprietes neutroniques des milieux contenant des neutrons thermiques. Le present rapport a pour but de decrire les principales de ces experiences et d'en donner une theorie generale. Cette theorie, exposee dans la premiere partie, est etablie dons le cadre de l'approximation a deux groupes de vitesse qui a prouve son efficacite dons le cas des piles a neutrons thermiques. Le domaine de validite de l'approximation a deux groupes est rappele au debut, ce qui permet de definir avec precision la signification des parametres qui entrent dons la theorie et qui font l'objet de ces mesures. La deuxieme partie decrit les experiences realisees, en particulier celles effectuees au Centre d'Etudes Nucleaires de Saclay ou la methode a ete considerablement developpee. Ces experiences sont interpretees dans le cadre de la theorie generale exposee precedemment. On prouve ainsi l'identite des parametres mesures par cette methode et de ceux figurant
Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center
International Nuclear Information System (INIS)
Newhauser, W.D.; Titt, U.; Dexheimer, D.; Yan, X.; Nill, S.
2002-01-01
The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent
International Nuclear Information System (INIS)
Lee, K. M.; Sun, G. M.
2016-01-01
The purpose of this study is to develop fake gold bar detecting method by using Prompt-gamma activation analysis (PGAA) facility at the Korea Atomic Energy Research Institute (KAERI). PGAA is an established nuclear analytical technique for non-destructive determination of elemental and isotopic compositions. For a preliminary study on detecting fake gold bar, Monte Carlo simulation of neutron transmission in gold bar was conducted and the possibility for detecting fake gold bar was confirmed. Under the gold bullion standard, it guaranteed the government would redeem any amount of currency for its value in gold. After the gold bullion standard ended, gold bars have been the target for investment as ever. But it is well known that fake gold bar exist in the gold market. This cannot be identified easily without performing a testing as it has the same appearance as the pure gold bar. In order to avoid the trading of fake gold bar in the market, they should be monitored thoroughly. Although the transmissivity of cold neutrons are low comparing that of thermal neutrons, the slower neutrons are more apt to be absorbed in a target, and can increase the prompt gamma emission rate. Also the flux of both thermal and cold neutron beam is high enough to activate thick target. If the neutron beam is irradiated on the front and the reverse side of gold bar, all insides of it can be detected
Energy Technology Data Exchange (ETDEWEB)
Lee, K. M.; Sun, G. M. [KAERI, Daejeon (Korea, Republic of)
2016-05-15
The purpose of this study is to develop fake gold bar detecting method by using Prompt-gamma activation analysis (PGAA) facility at the Korea Atomic Energy Research Institute (KAERI). PGAA is an established nuclear analytical technique for non-destructive determination of elemental and isotopic compositions. For a preliminary study on detecting fake gold bar, Monte Carlo simulation of neutron transmission in gold bar was conducted and the possibility for detecting fake gold bar was confirmed. Under the gold bullion standard, it guaranteed the government would redeem any amount of currency for its value in gold. After the gold bullion standard ended, gold bars have been the target for investment as ever. But it is well known that fake gold bar exist in the gold market. This cannot be identified easily without performing a testing as it has the same appearance as the pure gold bar. In order to avoid the trading of fake gold bar in the market, they should be monitored thoroughly. Although the transmissivity of cold neutrons are low comparing that of thermal neutrons, the slower neutrons are more apt to be absorbed in a target, and can increase the prompt gamma emission rate. Also the flux of both thermal and cold neutron beam is high enough to activate thick target. If the neutron beam is irradiated on the front and the reverse side of gold bar, all insides of it can be detected.
Method and equipment for γ background compensation in neutron spectrometry
International Nuclear Information System (INIS)
Holman, M.; Marik, P.
1986-01-01
The compensation of background gamma radiation in neutron spectrometry is based on the measurement of the total energy spectrum of all protons and electrons, and of the energy spectrum of those protons and neutrons which are in coincidence with the discriminating signal derived from the integral of the counting rate distribution by pulse shape. The benefits of the method consist in the possibility of using standard single-parameter apparatus, in considerably smaller demands on the memory capacity and the possibility of a substantially finer division of the spectrum and more accurate compensation of the background than has been the case with methods used so far. A practical application is shown in a block diagram. (J.B.)
Three-group albedo method applied to the diffusion phenomenon with up-scattering of neutrons
International Nuclear Information System (INIS)
Terra, Andre M. Barge Pontes Torres; Silva, Jorge A. Valle da; Cabral, Ronaldo G.
2007-01-01
The main objective of this research is to develop a three-group neutron Albedo algorithm considering the up-scattering of neutrons in order to analyse the diffusion phenomenon in nonmultiplying media. The neutron Albedo method is an analytical method that does not try to solve describing explicit equations for the neutron fluxes. Thus the neutron Albedo methodology is very different from the conventional methodology, as the neutron diffusion theory model. Graphite is analyzed as a model case. One major application is in the determination of the nonleakage probabilities with more understandable results in physical terms than conventional radiation transport method calculations. (author)
A combined neutron scattering and simulation study on bioprotectant systems
Energy Technology Data Exchange (ETDEWEB)
Affouard, F. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Bordat, P. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Descamps, M. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Lerbret, A. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Magazu, S. [Dipartimento di Fisica and INFM, Universita di Messina, P.O. Box 55, I-98166 Messina (Italy); Migliardo, F. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Dipartimento di Fisica and INFM, Universita di Messina, P.O. Box 55, I-98166 Messina (Italy)], E-mail: fmigliardo@unime.it; Ramirez-Cuesta, A.J. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, Didcot (United Kingdom); Telling, M.F.T. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, Didcot (United Kingdom)
2005-10-31
The present work shows quasi elastic neutron scattering, neutron spin echo and inelastic neutron scattering results on a class of bioprotectant systems, such as homologous disaccharides (i.e., trehalose and sucrose)/water solutions, as a function of temperature. The whole set of findings indicates a noticeable 'kosmotrope' character of the disaccharides, and in particular of trehalose, which is able to strongly modify both the structural and dynamical properties of water. This superior capability of trehalose can be linked to its higher bioprotective effectiveness in respect with the other disaccharides.
Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.
2011-01-01
Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.
An analytical method for neutron thermalization calculations in heterogenous reactors
Energy Technology Data Exchange (ETDEWEB)
Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)
1965-07-01
It is well known that the use of the diffusion approximation for stu
A study of environmental polluting factors by neutron activation method
International Nuclear Information System (INIS)
Paunoiu, C.; Doca, C.
2004-01-01
The paper presents: a) some importance factors of the environmental pollution; b) the theoretical aspects of the Neutron Activation Analysis (NAA) used in the study of the environmental pollution; c) the NAA specific hardware and software facilities existing at the Institute for Nuclear Research; d) a direct application of the NAA method in the study of the environmental pollution for Pitesti city by the analysis of some ground and vegetation samples; e) results and conclusions. (authors)
An analytical method for neutron thermalization calculations in heterogenous reactors
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1965-01-01
It is well known that the use of the diffusion approximation for studying neutron thermalization in heterogeneous reactors may result in considerable errors. On the other hand, more exact numerical methods are rather laborious and require the use of large digital computers. In this paper, the use of the diffusion approximation in absorbing media has been avoided, but the treatment remained analytical, thus simplifying practical calculations
The effect of the neutron spectra unfolding method on the fast neutron dose determination
International Nuclear Information System (INIS)
Marinkovic, P.; Zavaljevski, N.
1992-01-01
Based on Shanon's information theory, a new unfolding method which gives non-negative spectrum values and a relatively low variance, is proposed, and a numerical code suitable for application in fast neutron spectroscopy based on proton recoil is developed. The principles of maximum entropy and maximum likelihood are jointly applied. According to the principle of maximum likelihood, the distribution functions around the mean value of the counts in the MCA channels are assumed to be Gaussians. The Lagrange parameter method is applied in the search for an optimal non-negative solution. The nonlinear system of equations is solved using the gradient and Newton iterative algorithms. (orig.)
The effect of the neutron spectra unfolding method on the fast neutron dose determination
International Nuclear Information System (INIS)
Marinkovic, P.; Avdic, S.; Pesic, M.; Zavaljevski, N
1992-09-01
Based on Shanon's information theory, a new unfolding method which gives non-negative spectrum values and a relatively low variance, is proposed, and a numerical code suitable for application in fast neutron spectroscopy based on proton recoil is developed. The principles of maximum entropy and maximum likelihood are jointly applied. According to the principle of maximum likelihood, the distribution functions around the mean value of the counts in the MCA channels are assumed to be Gaussians. The Lagrange parameter method is applied in the search for an optimal non-negative solution. The nonlinear system of equations is solved using the gradient and Newton iterative algorithms. (author)
Nanocrystallite characterization of milled simulated dry process fuel powders by neutron diffraction
International Nuclear Information System (INIS)
Ryu, Ho Jin; Kang, Kwon Ho; Moon, Je Sun; Song, Kee Chan; Choi, Yong Nam
2003-01-01
The nano-scale crystallite sizes of simulated spent fuel powders were measured by the neutron diffraction line broadening method in order to analyze the sintering behavior of the dry process fuel. The mixed U0 2 and fission product oxide powders were dry-milled in an attritor for 30, 60, and 120 min. The diffraction patterns of the powders were obtained by using the high resolution powder diffractometer in the HANARO research reactor. Diffraction line broadening due to crystallite size was measured using various techniques such as the Stokes' deconvolution, profile fitting methods using Cauchy function, Gaussian function, and Voigt function, and the Warren-Averbach method. The r.m.s. strain, stacking fault, twin and dislocation density were measured using the information from the diffraction pattern. The realistic crystallite size can be obtained after separation of the contribution from the non-uniform strain, stacking fault and twin
DUST, Albedo Monte-Carlo Simulation of Neutron Streaming in Multi-legged Square Concrete Ducts
International Nuclear Information System (INIS)
1993-01-01
1 - Description of program or function: DUST simulates the thermal neutron streaming through multi-legged square concrete ducts. 2 - Method of solution: DUST uses the albedo Monte Carlo method. The albedo data used are in the form of empirical formulae based on the measured doubly differential albedo data. Sampling of the reflected polar and azimuthal angles is done by the rejection method. Variance reduction devices such as Russian Roulette are used. 3 - Restrictions on the complexity of the problem: - The albedo data and the subroutines for sampling the reflected polar and azimuthal angles are specific for concrete ducts. The maximum number of legs (as specified by dimension statements) is 5 and the maximum number of dose points is 50. The dose points considered are only in the last leg of the multi-legged duct
Neutron flux calculation by means of Monte Carlo methods
International Nuclear Information System (INIS)
Barz, H.U.; Eichhorn, M.
1988-01-01
In this report a survey of modern neutron flux calculation procedures by means of Monte Carlo methods is given. Due to the progress in the development of variance reduction techniques and the improvements of computational techniques this method is of increasing importance. The basic ideas in application of Monte Carlo methods are briefly outlined. In more detail various possibilities of non-analog games and estimation procedures are presented, problems in the field of optimizing the variance reduction techniques are discussed. In the last part some important international Monte Carlo codes and own codes of the authors are listed and special applications are described. (author)
Energy Technology Data Exchange (ETDEWEB)
Conway, A; Wang, T; Deo, N; Cheung, C; Nikolic, R
2008-06-24
This work reports numerical simulations of a novel three-dimensionally integrated, {sup 10}boron ({sup 10}B) and silicon p+, intrinsic, n+ (PIN) diode micropillar array for thermal neutron detection. The inter-digitated device structure has a high probability of interaction between the Si PIN pillars and the charged particles (alpha and {sup 7}Li) created from the neutron - {sup 10}B reaction. In this work, the effect of both the 3-D geometry (including pillar diameter, separation and height) and energy loss mechanisms are investigated via simulations to predict the neutron detection efficiency and gamma discrimination of this structure. The simulation results are demonstrated to compare well with the measurement results. This indicates that upon scaling the pillar height, a high efficiency thermal neutron detector is possible.
International Nuclear Information System (INIS)
Li Chunjuan; Liu Yi'na; Zhang Weihua; Wang Zhiqiang
2014-01-01
The manganese bath method for measuring the neutron emission rate of radionuclide sources requires corrections to be made for emitted neutrons which are not captured by manganese nuclei. The Monte Carlo particle transport code MCNP was used to simulate the manganese bath system of the standards for the measurement of neutron source intensity. The correction factors were calculated and the reliability of the model was demonstrated through the key comparison for the radionuclide neutron source emission rate measurements organized by BIPM. The uncertainties in the calculated values were evaluated by considering the sensitivities to the solution density, the density of the radioactive material, the positioning of the source, the radius of the bath, and the interaction cross-sections. A new method for the evaluation of the uncertainties in Monte Carlo calculation was given. (authors)
Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators
International Nuclear Information System (INIS)
Ragusa, J.C.
2001-01-01
In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)
Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators
Energy Technology Data Exchange (ETDEWEB)
Ragusa, J.C. [CEA Saclay, Dept. de Mecanique et de Technologie, 91 - Gif-sur-Yvette (France)
2001-07-01
In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)
Energy Technology Data Exchange (ETDEWEB)
Jacquet, P.
2011-05-23
Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author) [French] Les criteres de surete qui regissent le developpement de coeurs de reacteurs de quatrieme generation implique l'usage d'outils de calcul neutronique performants. Une premiere partie de la these reprend toutes les etapes de modelisation neutronique des reacteurs rapides actuellement d'usage dans le code de reference ECCO. La capacite des modeles a decrire le phenomene d'autoprotection, a representer les fuites neutroniques au niveau d'un reseau d'assemblages combustibles et a generer des sections macroscopiques representatives est appreciee sur le domaine des reacteurs rapides innovants respectant les criteres de quatrieme generation. La deuxieme partie de ce memoire se consacre a la modelisation des coeurs rapides avec reflecteur acier. Ces derniers necessitent le developpement de methodes avancees de condensation et d'homogenisation. Plusieurs methodes sont proposees et confrontees sur un probleme de modelisation typique: le coeur ZONA2B du reacteur maquette MASURCA
Neutron absorbing article and method for manufacture thereof
International Nuclear Information System (INIS)
Forsyth, P.F.; Mcmurtry, C.H.; Naum, R.G.
1980-01-01
A composite, neutron absorbing, coated article, suitable for installation in storage racks for spent nuclear fuel and for other neutron absorbing applications, includes a backing member, preferably of flexible material such as woven fiberglass cloth, a synthetic organic polymeric coating or a plurality of such coatings on the backing member, preferably of cured phenolic resin, such as phenol formaldehyde or trimethylolphenol formaldehyde and boron carbide particles held to the backing member by the cured coating or a plurality of such coatings. Also within the invention is a method for the manufacture of the neutron absorbing coated article and the use of such an article. In a preferred method the backing member is first coated on both sides thereof with a filling coating of thermosettable liquid phenolic resin, which is then partially cured to solid state, one side of the backing member is then coated with a mixture of thermosettable liquid resin and finely divided boron carbide particles and the resin is partially cured to solid state, the other side is coated with a similar mixture, larger boron carbide particles are applied to it and the resin is partially cured to solid state, such side of the article is coated with thermosettable liquid phenolic resin, the resin is partially cured to solid state and such resin, including previously applied partially cured resins, is cured to final cross-linked and permanently set form
Analysis of medicinal plant extracts by neutron activation method
International Nuclear Information System (INIS)
Vaz, Sandra Muntz
1995-01-01
This dissertation has presented the results from analysis of medicinal plant extracts using neutron activation method. Instrumental neutron activation analysis was applied to the determination of the elements Al, Br, Ca, Ce, Cl, Cr, Cs, Fe, K, La, Mg, Mn, Na, Rb, Sb, Sc and Zn in medicinal extracts obtained from Achyrolcline satureoides DC, Casearia sylvestris, Centella asiatica, Citrus aurantium L., Solano lycocarpum, Solidago microglossa, Stryphnondedron barbatiman and Zingiber officinale R. plants. The elements Hg and Se were determined using radiochemical separation by means of retention of Se in HMD inorganic exchanger and solvent extraction of Hg by bismuth diethyl-dithiocarbamate solution. Precision and accuracy of the results have been evaluated by analysing reference materials. The therapeutic action of some elements found in plant extracts analyzed was briefly discussed
DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.
2018-06-01
A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.
MCNPX Simulation Study of STRAW Neutron Detectors - Summary Paper
International Nuclear Information System (INIS)
Mukhopadhyay, Sanjoy; Maurer, Richard; Mitchell, Stephen
2010-01-01
A novel prototype fission meter is being designed at National Security Technologies, LLC, using a thin uniform coating (only 1 micron thick) of 10 B as a neutron converter inside a large array of thin (4 mm diameter) copper tubes. The copper tubes are only 2 mils thick, and each holds the stretched anode wire under tension and high voltage. The tubes are filled with proportional counter gas (a mixture of 90%/10% of Ar/CO 2 ). The tubes operate in proportional counter mode and attract mobile charged particles (α's) created in the nuclear interaction 10 B(n, 4 He) 7 Li. However, a single tube has about 1/7th the sensitivity of a 3 He tube. Modeling is required to determine if enough such tubes could be placed in a neutron detection assembly of the current size to give comparable sensitivity to 3 He. Detectors lined with 10 B lie between 3 He and 10 BF 3 proportional counters and fission chambers in terms of neutron detection efficiency and gamma ray insensitivity. The mean free path of thermal neutrons in 10 B is about 18 (micro)m. It takes about 60 (micro)m of 10 B layer to completely stop thermal neutrons, but the energetic α-particles generated in the reaction have a range of only 3.3 (micro)m in 10 B environment - hence the thin layer of boron coating on the copper tube. The prototype design is shown in Figure 1. It consists of two panels of three staggered rows of 500-mm-long, 4-mm-diameter straws, with 20 in each row, embedded in 30-mm-thick high density polyethylene (HDPE). The project demonstrates a new application of thin neutron and gamma converter technique (1 micron thin 10 B coated copper tube). It exploits fast timing from multiple straw detectors to count multiplicity of both gamma and neutrons from fissioning materials. The objective is to find a near-term replacement of 3 He gas in neutron detection and measurement (with a very large neutron detection area). All the solid-state detectors developed thus far are small and inefficient. The thin size
Energy Technology Data Exchange (ETDEWEB)
Patil, B.J., E-mail: bjp@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India)
2012-01-15
The 6 MeV LINAC based pulsed thermal neutron source has been designed for bulk materials analysis. The design was optimized by varying different parameters of the target and materials for each region using FLUKA code. The optimized design of thermal neutron source gives flux of 3 Multiplication-Sign 10{sup 6}ncm{sup -2}s{sup -1} with more than 80% of thermal neutrons and neutron to gamma ratio was 1 Multiplication-Sign 10{sup 4}ncm{sup -2}mR{sup -1}. The results of prototype experiment and simulation are found to be in good agreement with each other. - Highlights: Black-Right-Pointing-Pointer The optimized 6 eV linear accelerator based thermal neutron source using FLUKA simulation. Black-Right-Pointing-Pointer Beryllium as a photonuclear target and reflector, polyethylene as a filter and shield, graphite as a moderator. Black-Right-Pointing-Pointer Optimized pulsed thermal neutron source gives neutron flux of 3 Multiplication-Sign 10{sup 6} n cm{sup -2} s{sup -1}. Black-Right-Pointing-Pointer Results of the prototype experiment were compared with simulations and are found to be in good agreement. Black-Right-Pointing-Pointer This source can effectively be used for the study of bulk material analysis and activation products.
International Nuclear Information System (INIS)
Jin, Y.; Verghese, K.; Gardner, R.P.
1986-01-01
This paper describes a major part of our efforts to simulate the entire spectral response of the neutron capture prompt gamma-ray analyzer for bulk media (or conveyor belt) samples by the Monte Carlo method. This would allow one to use such a model to augment or, in most cases, essentially replace experiments in the calibration and optimum design of these analyzers. In previous work, we simulated the unscattered gamma-ray intensities, but would like to simulate the entire spectral response as we did with the energy-dispersive x-ray fluorescence analyzers. To accomplish this, one must account for the scattered gamma rays as well as the unscattered and one must have available the detector response function to translate the incident gamma-ray spectrum calculated by the Monte Carlo simulation into the detected pulse-height spectrum. We recently completed our work on the germanium detector response function, and the present paper describes our efforts to simulate the entire spectral response by using it with Monte Carlo predicted unscattered and scattered gamma rays
Production of Molybdenum-99 using Neutron Capture Methods
Energy Technology Data Exchange (ETDEWEB)
Toth, James J; Greenwood, Lawrence R; Soderquist, Chuck Z; Wittman, Richard S; Pierson, Bruce D; Burns, Kimberly A; Lavender, Curt A; Painter, Chad L; Love, Edward F; Wall, Donald E
2011-01-01
Pacific Northwest National Laboratory (PNNL), operated by Battelle, has identified a reference process for the production of molybdenum-99 (99Mo) for use in a chromatographic generator to separate the daughter product, technetium-99m (99mTc). The reference process uses the neutron capture reaction of natural or enriched molybdenum oxide via the reaction 98Mo(n,γ)99Mo. The irradiated molybdenum is dissolved in an alkaline solution, whereby the molybdenum, dissolved as the molybdate anion, is loaded on a proprietary ion exchange material in the chromatographic generator. The approach of this investigation is to provide a systematic collection of technologies to make the neutron capture method for Mo-99 production economically viable. This approach would result in the development of a technetium Tc99m generator and a new type of target. The target is comprised of molybdenum, either natural or enriched, and is tailored to the design of currently operating U.S. research reactors. The systematic collection of technologies requires evaluation of new metallurgical methods to produce the target, evaluation of target geometries tailored to research reactors, and chemical methods to dissolve the irradiated target materials for use in a chromatographic generator. A Technical specification for testing the target and neutron capture method in a research reactor is also required. This report includes identification of research and demonstration activities needed to enable deployment of neutron capture production method, including irradiations of prototypic targets, chemical processing of irradiated targets, and loading and extraction tests of Mo99 and Tc99m on the sorbent material in a prototypic generator design. The prototypical generator design is based on the proprietary method and systems for isotope product generation. The proprietary methods and systems described in this report are clearly delineated with footnotes. Ultimately, the Tc-99m generator solution provided by
Energy Technology Data Exchange (ETDEWEB)
Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)
2011-07-01
Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)
International Nuclear Information System (INIS)
Guerin, P.
2007-12-01
The neutronic simulation of a nuclear reactor core is performed using the neutron transport equation, and leads to an eigenvalue problem in the steady-state case. Among the deterministic resolution methods, diffusion approximation is often used. For this problem, the MINOS solver based on a mixed dual finite element method has shown his efficiency. In order to take advantage of parallel computers, and to reduce the computing time and the local memory requirement, we propose in this dissertation two domain decomposition methods for the resolution of the mixed dual form of the eigenvalue neutron diffusion problem. The first approach is a component mode synthesis method on overlapping sub-domains. Several Eigenmodes solutions of a local problem solved by MINOS on each sub-domain are taken as basis functions used for the resolution of the global problem on the whole domain. The second approach is a modified iterative Schwarz algorithm based on non-overlapping domain decomposition with Robin interface conditions. At each iteration, the problem is solved on each sub domain by MINOS with the interface conditions deduced from the solutions on the adjacent sub-domains at the previous iteration. The iterations allow the simultaneous convergence of the domain decomposition and the eigenvalue problem. We demonstrate the accuracy and the efficiency in parallel of these two methods with numerical results for the diffusion model on realistic 2- and 3-dimensional cores. (author)
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)
Neutron activation analysis of certified samples by the absolute method
Kadem, F.; Belouadah, N.; Idiri, Z.
2015-07-01
The nuclear reactions analysis technique is mainly based on the relative method or the use of activation cross sections. In order to validate nuclear data for the calculated cross section evaluated from systematic studies, we used the neutron activation analysis technique (NAA) to determine the various constituent concentrations of certified samples for animal blood, milk and hay. In this analysis, the absolute method is used. The neutron activation technique involves irradiating the sample and subsequently performing a measurement of the activity of the sample. The fundamental equation of the activation connects several physical parameters including the cross section that is essential for the quantitative determination of the different elements composing the sample without resorting to the use of standard sample. Called the absolute method, it allows a measurement as accurate as the relative method. The results obtained by the absolute method showed that the values are as precise as the relative method requiring the use of standard sample for each element to be quantified.
Nodal methods for problems in fluid mechanics and neutron transport
International Nuclear Information System (INIS)
Azmy, Y.Y.
1985-01-01
A new high-accuracy, coarse-mesh, nodal integral approach is developed for the efficient numerical solution of linear partial differential equations. It is shown that various special cases of this general nodal integral approach correspond to several high efficiency nodal methods developed recently for the numerical solution of neutron diffusion and neutron transport problems. The new approach is extended to the nonlinear Navier-Stokes equations of fluid mechanics; its extension to these equations leads to a new computational method, the nodal integral method which is implemented for the numerical solution of these equations. Application to several test problems demonstrates the superior computational efficiency of this new method over previously developed methods. The solutions obtained for several driven cavity problems are compared with the available experimental data and are shown to be in very good agreement with experiment. Additional comparisons also show that the coarse-mesh, nodal integral method results agree very well with the results of definitive ultra-fine-mesh, finite-difference calculations for the driven cavity problem up to fairly high Reynolds numbers
An alternative method for the measurement of neutron flux
Indian Academy of Sciences (India)
dose rate. We have used a 241AmBe neutron source for neutron irradiation, and the neutron dose rate and count rate were ... neutron sources, e.g., for the characterization of superheated droplet detectors (SDD). [1–6]. The SDD is a .... Grants Commission (UGC) for the financial assistance provided for this work. References.
Simulation of complete neutron scattering experiments: from model systems to liquid germanium
International Nuclear Information System (INIS)
Hugouvieux, V.
2004-11-01
In this thesis, both theoretical and experimental studies of liquids are done. Neutron scattering enables structural and dynamical properties of liquids to be investigated. On the theoretical side, molecular dynamics simulations are of great interest since they give positions and velocities of the atoms and the forces acting on each of them. They also enable spatial and temporal correlations to be computed and these quantities are also available from neutron scattering experiments. Consequently, the comparison can be made between results from molecular dynamics simulations and from neutron scattering experiments, in order to improve our understanding of the structure and dynamics of liquids. However, since extracting reliable data from a neutron scattering experiment is difficult, we propose to simulate the experiment as a whole, including both instrument and sample, in order to gain understanding and to evaluate the impact of the different parasitic contributions (absorption, multiple scattering associated with elastic and inelastic scattering, instrument resolution). This approach, in which the sample is described by its structure and dynamics as computed from molecular dynamics simulations, is presented and tested on isotropic model systems. Then liquid germanium is investigated by inelastic neutron scattering and both classical and ab initio molecular dynamics simulations. This enables us to simulate the experiment we performed and to evaluate the influence of the contributions from the instrument and from the sample on the detected signal. (author)
Ilgner, C
2003-01-01
In view of a beam test of RadFET semiconductor detectors and optically stimulated luminescence (OSL) detectors as on-line dosimeters for radiation monitoring purposes in the caverns of the Large Hadron Collider (LHC) experiments, a simulation on the production of secondary particles from a neutron beam on a polyethylene target was carried out. We describe the yield of recoil protons, scattered neutrons as well as electrons, positrons and photons, when neutrons of an average energy of 20 MeV hit polyethylene targets of several thicknesses. The simulation was carried out using the latest release 5.2 of the GEANT4 detector description and simulation tool, including advanced hadron interaction models.
International Nuclear Information System (INIS)
Nigg, D.W.; Wemple, C.A.; Hartwell, J.K.; Harker, Y.D.; Venhuizen, J.R.; Risler, R.
1997-12-01
A closed-form direct method for unfolding neutron spectra from foil activation data is presented. The method is applied to measurements of the free-field neutron spectrum produced by the proton-cyclotron-based fast-neutron radiotherapy facility at the University of Washington (UW) School of Medicine. The results compare favorably with theoretical expectations based on an a-priori calculational model of the target and neutron beamline configuration of the UW facility
Simulation study of accelerator based quasi-mono-energetic epithermal neutron beams for BNCT.
Adib, M; Habib, N; Bashter, I I; El-Mesiry, M S; Mansy, M S
2016-01-01
Filtered neutron techniques were applied to produce quasi-mono-energetic neutron beams in the energy range of 1.5-7.5 keV at the accelerator port using the generated neutron spectrum from a Li (p, n) Be reaction. A simulation study was performed to characterize the filter components and transmitted beam lines. The feature of the filtered beams is detailed in terms of optimal thickness of the primary and additive components. A computer code named "QMNB-AS" was developed to carry out the required calculations. The filtered neutron beams had high purity and intensity with low contamination from the accompanying thermal, fast neutrons and γ-rays. Copyright © 2015 Elsevier Ltd. All rights reserved.
Simulation study on the cold neutron guides in China advanced research reactor
International Nuclear Information System (INIS)
Guo Liping; Yang Tonghua; Wang Hongli; Sun Kai; Zhao Zhixiang
2003-01-01
The designs of the two cold neutron guides, CNG1 and CNG2, to be built in China advanced research reactor (CARR) are studied with Monte-Carlo simulation technique. The neutron flux density at the exit of the both guides can reach above 1 x10 9 cm -2 ·s -1 under the assumed flux spectrum of the cold neutron source. The transmission efficiency is 50% and 42%, and the maximum divergence is about 2.2 degree and 1.9 degree, respectively for CNG1 and CNG2. Neutron distribution along horizontal direction is quite uniform for both guides, with maximum fluctuation of less than 3%. Gravity can affect neutron distribution along vertical direction considerably
GEANT4 simulation study of a gamma-ray detector for neutron resonance densitometry
International Nuclear Information System (INIS)
Tsuchiya, Harufumi; Harada, Hideo; Koizumi, Mitsuo; Kitatani, Fumito; Takamine, Jun; Kureta, Masatoshi; Iimura, Hideki
2013-01-01
A design study of a gamma-ray detector for neutron resonance densitometry was made with GEANT4. The neutron resonance densitometry, combining neutron resonance transmission analysis and neutron resonance capture analysis, is a non-destructive technique to measure amounts of nuclear materials in melted fuels of the Fukushima Daiichi nuclear power plants. In order to effectively quantify impurities in the melted fuels via prompt gamma-ray measurements, a gamma-ray detector for the neutron resonance densitometry consists of cylindrical and well type LaBr 3 scintillators. The present simulation showed that the proposed gamma-ray detector suffices to clearly detect the gamma rays emitted by 10 B(n, αγ) reaction in a high environmental background due to 137 Cs radioactivity with its Compton edge suppressed at a considerably small level. (author)
International Nuclear Information System (INIS)
Hadad, Kamal; Pirouzmand, Ahmad; Ayoobian, Navid
2008-01-01
This paper describes the application of a multilayer cellular neural network (CNN) to model and solve the time dependent one-speed neutron transport equation in slab geometry. We use a neutron angular flux in terms of the Chebyshev polynomials (T N ) of the first kind and then we attempt to implement the equations in an equivalent electrical circuit. We apply this equivalent circuit to analyze the T N moments equation in a uniform finite slab using Marshak type vacuum boundary condition. The validity of the CNN results is evaluated with numerical solution of the steady state T N moments equations by MATLAB. Steady state, as well as transient simulations, shows a very good comparison between the two methods. We used our CNN model to simulate space-time response of total flux and its moments for various c (where c is the mean number of secondary neutrons per collision). The complete algorithm could be implemented using very large-scale integrated circuit (VLSI) circuitry. The efficiency of the calculation method makes it useful for neutron transport calculations
252Cf-source-driven neutron noise analysis method
International Nuclear Information System (INIS)
Mihalczo, J.T.; King, W.T.; Blakeman, E.D.
1985-01-01
The 252 Cf-source-driven neutron noise analysis method has been tested in a wide variety of experiments that have indicated the broad range of applicability of the method. The neutron multiplication factor k/sub eff/ has been satisfactorily detemined for a variety of materials including uranium metal, light water reactor fuel pins, fissile solutions, fuel plates in water, and interacting cylinders. For a uranyl nitrate solution tank which is typical of a fuel processing or reprocessing plant, the k/sub eff/ values were satisfactorily determined for values between 0.92 and 0.5 using a simple point kinetics interpretation of the experimental data. The short measurement times, in several cases as low as 1 min, have shown that the development of this method can lead to a practical subcriticality monitor for many in-plant applications. The further development of the method will require experiments oriented toward particular applications including dynamic experiments and the development of theoretical methods to predict the experimental observables
Neutron methods for the direct determination of the magnetic induction in thick films
Energy Technology Data Exchange (ETDEWEB)
Kozhevnikov, S.V., E-mail: kozhevn@nf.jinr.ru [Frank Laboratory of Neutron Physics, Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Ott, F. [CEA, IRAMIS, Laboratoire Léon Brillouin, F-91191 Gif sur Yvette (France); CNRS, IRAMIS, Laboratoire Léon Brillouin, F-91191 Gif sur Yvette (France); Radu, F. [Helmholtz-Zentrum Berlin für Materialien und Energie, Albert-Einstein Strasse 15, D-12489 Berlin (Germany)
2016-03-15
We review different neutron methods which allow extracting directly the value of the magnetic induction in thick films: Larmor precession, Zeeman spatial beam-splitting and neutron spin resonance. Resulting parameters obtained by the neutron methods and standard magnetometry technique are presented and compared. The possibilities and specificities of the neutron methods are discussed. - Highlights: • We present neutron methods for investigations of the thick magnetic films. • It is the methods for the direct determination of the magnetic induction. • Magnetic induction in bulk, at single interface and in a single domain. • It is Larmor precession, Zeeman spatial beam-splitting and neutron spin resonance. • These methods are complementary to polarized neutron reflectometry.
Energy Technology Data Exchange (ETDEWEB)
Hosseini, Seyed Abolfazl, E-mail: sahosseini@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Tehran 8639-11365 (Iran, Islamic Republic of); Afrakoti, Iman Esmaili Paeen [Faculty of Engineering & Technology, University of Mazandaran, Pasdaran Street, P.O. Box: 416, Babolsar 47415 (Iran, Islamic Republic of)
2017-04-11
Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The {sup 241}Am-{sup 9}Be and {sup 252}Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions. - Highlights: • The neutron pulse height distribution was simulated using MCNPX-ESUT. • The energy spectrum of the neutron source was unfolded using GMDH. • The energy spectrum of the neutron source was
Žukauskaitėa, A; Plukienė, R; Ridikas, D
2007-01-01
Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.
Benchmarking shielding simulations for an accelerator-driven spallation neutron source
Directory of Open Access Journals (Sweden)
Nataliia Cherkashyna
2015-08-01
Full Text Available The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS, currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the Swiss Spallation Neutron Source (SINQ, at the Paul Scherrer Institute (PSI, Villigen, Switzerland. The measurements were carried out at several positions along the SINQ monolith wall with the neutron dosimeter WENDI-2, which has a well-characterized response up to 5 GeV. The simulations were performed using the Monte-Carlo radiation transport code geant4, and include a complete transport from the proton beam to the measurement locations in a single calculation. An agreement between measurements and simulations is about a factor of 2 for the points where the measured radiation dose is above the background level, which is a satisfactory result for such simulations spanning many energy regimes, different physics processes and transport through several meters of shielding materials. The neutrons contributing to the radiation field emanating from the monolith were confirmed to originate from neutrons with energies above 1 MeV in the target region. The current work validates geant4 as being well suited for deep-shielding calculations at accelerator-based spallation sources. We also extrapolate what the simulated flux levels might imply for short (several tens of meters instruments at ESS.
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
Energy Technology Data Exchange (ETDEWEB)
Anelli, M; Bertolucci, S; Curceanu, C; Giovannella, S; Happacher, F; Iliescu, M; Martini, M; Miscetti, S [Laboratori Nazionali di Frascati, INFN (Italy); Battistoni, G [Sezione INFN di Milano (Italy); Bini, C; Zorzi, G De; Domenico, Adi; Gauzzi, P [Ubiversita degli Studi ' La Sapienza' e Sezine INFN di Roma (Italy); Branchini, P; Micco, B Di; Ngugen, F; Paseri, A [Universita degli di Studi ' Roma Tre' e Sezione INFN di Roma Tre (Italy); Ferrari, A [Fondazione CNAO, Milano (Italy); Prokfiev, A [Svedberg Laboratory, Uppsala University (Sweden); Fiore, S, E-mail: matteo.martino@inf.infn.i
2009-04-01
We have measured the overall detection efficiency of a small prototype of the KLOE PB-scintillation fiber calorimeter to neutrons with kinetic energy range [5,175] MeV. The measurement has been done in a dedicated test beam in the neutron beam facility of the Svedberg Laboratory, TSL Uppsala. The measurements of the neutron detection efficiency of a NE110 scintillator provided a reference calibration. At the lowest trigger threshold, the overall calorimeter efficiency ranges from 28% to 33%. This value largely exceeds the estimated {approx}8% expected if the response were proportional only to the scintillator equivalent thickness. A detailed simulation of the calorimeter and of the TSL beam line has been performed with the FLUKA Monte Carlo code. The simulated response of the detector to neutrons is presented together with the first data to Monte Carlo comparison. The results show an overall neutron efficiency of about 35%. The reasons for such an efficiency enhancement, in comparison with the typical scintillator-based neutron counters, are explained, opening the road to a novel neutron detector.
Fission fragment simulation of fusion neutron radiation effects on bulk mechanical properties
International Nuclear Information System (INIS)
Van Konynenburg, R.A.; Mitchell, J.B.; Guinan, M.W.; Stuart, R.N.; Borg, R.J.
1976-01-01
This research demonstrates the feasibility of using homogeneously-generated fission fragments to simulate high-fluence fusion neutron damage in niobium tensile specimens. This technique makes it possible to measure radiation effects on bulk mechanical properties at high damage states, using conveniently short irradiation times. The primary knock-on spectrum for a fusion reactor is very similar to that produced by fission fragments, and nearly the same ratio of gas atoms to displaced atoms is produced in niobium. The damage from fission fragments is compared to that from fusion neutrons and fission reactor neutrons in terms of experimentally measured yield strength increase, transmission electron microscopy (TEM) observations, and calculated damage energies
Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)
International Nuclear Information System (INIS)
Laureau, A.; Rubiolo, P.R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.
2013-01-01
Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)
Energy Technology Data Exchange (ETDEWEB)
Cesaro, A.; Magazu, V.; Migliardo, F.; Sussich, F.; Vadala, M
2004-07-15
Neutron diffraction measurements combined with H/D substitution have been performed on trehalose aqueous solutions as a function of temperature and concentration by using the SANDALS diffractometer at ISIS Facility (UK). The findings point out a high capability of trehalose to strongly affect the tetrahedral hydrogen bond network of water. The neutron diffraction results are also compared with simulation and experimental data obtained by synchrotron radiation on the phospholipid bilayer membranes (DPPC)/trehalose/H{sub 2}O ternary system.
Neutron activation analysis of maltenes recovered from EUROBITUM simulates
International Nuclear Information System (INIS)
Impens, N
2006-01-01
the bitumen. Moreover, the removal of maltenes and soluble salts results in an important volume reduction of the primary waste. The room temperature method is quite safe, as liquids with high flash points are used, and because volatilisation of radionuclides is avoided. The aim is to study the elemental composition of the non-aqueous secondary waste streams of the room temperature re-treatment method for Eurobitum. It is necessary to assess the purity of maltene solutions recovered from Eurobitum simulates in order to define experimental conditions leading to the free release of the recovered maltene solutions. As the detection limit should be very low and as we only used self-made non-radioactive Eurobitum simulates, NAA is the most suitable technique for this study
Monte Carlo simulation of secondary neutron dose for scanning proton therapy using FLUKA.
Directory of Open Access Journals (Sweden)
Chaeyeong Lee
Full Text Available Proton therapy is a rapidly progressing field for cancer treatment. Globally, many proton therapy facilities are being commissioned or under construction. Secondary neutrons are an important issue during the commissioning process of a proton therapy facility. The purpose of this study is to model and validate scanning nozzles of proton therapy at Samsung Medical Center (SMC by Monte Carlo simulation for beam commissioning. After the commissioning, a secondary neutron ambient dose from proton scanning nozzle (Gantry 1 was simulated and measured. This simulation was performed to evaluate beam properties such as percent depth dose curve, Bragg peak, and distal fall-off, so that they could be verified with measured data. Using the validated beam nozzle, the secondary neutron ambient dose was simulated and then compared with the measured ambient dose from Gantry 1. We calculated secondary neutron dose at several different points. We demonstrated the validity modeling a proton scanning nozzle system to evaluate various parameters using FLUKA. The measured secondary neutron ambient dose showed a similar tendency with the simulation result. This work will increase the knowledge necessary for the development of radiation safety technology in medical particle accelerators.
Modern methods to improve the accuracy in fast neutron dosimetry
International Nuclear Information System (INIS)
Baers, B.; Karnani, H.; Seren, T.
1985-01-01
In order to improve the quality of fast neutron dose estimates at the reactor pressure vessel (PV) some modern methods are presented. In addition to basic principles, some error reduction procedures are also presented based on the combined use of relative measurements, direct sample taking from the pressure vessel and the use of iron and niobium as dosimeters. The influence of large systematic errors could be significantly reduced by carrying out relative measurements. This report also presents the successful use of niobium as a dosimeter by destructive treatment of PV samples. (author)
International intercomparison of neutron spectra evaluating methods using activation detectors
International Nuclear Information System (INIS)
Fischer, A.
1975-06-01
The international intercomparison of neutron spectrum evaluation methods using activation detectors was organized by the IAEA in 1971 - 1972. All of the contributions and the results of a critical evaluation are presented here. The spectra of different contributors are compared to a reference spectrum by means of different integrals and weighting functions. Different cross section sets, foil numbers, energy point systems, guess spectra used by the contributors cause differences in the resulting spectra. The possible ways of separating these effects are also investigated. Suggestions are made for the organization of a new intercomparison on the basis of more uniform input data. (orig.) [de
Numerical method for solving integral equations of neutron transport. II
International Nuclear Information System (INIS)
Loyalka, S.K.; Tsai, R.W.
1975-01-01
In a recent paper it was pointed out that the weakly singular integral equations of neutron transport can be quite conveniently solved by a method based on subtraction of singularity. This previous paper was devoted entirely to the consideration of simple one-dimensional isotropic-scattering and one-group problems. The present paper constitutes interesting extensions of the previous work in that in addition to a typical two-group anisotropic-scattering albedo problem in the slab geometry, the method is also applied to an isotropic-scattering problem in the x-y geometry. These results are compared with discrete S/sub N/ (ANISN or TWOTRAN-II) results, and for the problems considered here, the proposed method is found to be quite effective. Thus, the method appears to hold considerable potential for future applications. (auth)
Neutron activation analysis: A primary method of measurement
Energy Technology Data Exchange (ETDEWEB)
Greenberg, Robert R., E-mail: robert.greenberg@nist.gov [National Institute of Standards and Technology, Gaithersburg, MD 20899-8395 (United States); Bode, Peter, E-mail: p.bode@tudelft.nl [Delft University of Technology, Delft (Netherlands); De Nadai Fernandes, Elisabete A., E-mail: lis@cena.usp.br [Centro de Energia Nuclear na Agricultura, Universidade de Sao Paulo, Piracicaba, SP (Brazil)
2011-03-15
Neutron activation analysis (NAA), based on the comparator method, has the potential to fulfill the requirements of a primary ratio method as defined in 1998 by the Comite Consultatif pour la Quantite de Matiere - Metrologie en Chimie (CCQM, Consultative Committee on Amount of Substance - Metrology in Chemistry). This thesis is evidenced in this paper in three chapters by: demonstration that the method is fully physically and chemically understood; that a measurement equation can be written down in which the values of all parameters have dimensions in SI units and thus having the potential for metrological traceability to these units; that all contributions to uncertainty of measurement can be quantitatively evaluated, underpinning the metrological traceability; and that the performance of NAA in CCQM key-comparisons of trace elements in complex matrices between 2000 and 2007 is similar to the performance of Isotope Dilution Mass Spectrometry (IDMS), which had been formerly designated by the CCQM as a primary ratio method.
International Nuclear Information System (INIS)
Nakacho, Keiji; Kasahara, Norifumi; Tamura, Ryota
2012-01-01
The measuring methods of the residual stress are classified into destructive one and nondestructive one. The inherent strain method (ISM) is destructive one. The neutron diffraction method (NDM) is nondestructive one. But the measurable depth is limited within about 20 mm and the method cannot measure the weld zone, without destruction of the object. So, in this study, the hybrid measuring method has been developed, by combining the ISM and the NDM. The theory of the hybrid method is the same as the ISM. In the analysis, the strains measured by the NDM without destruction are used. This hybrid measuring method is a true nondestructive measuring method for a thick welded joint. The applicability of the hybrid method has been verified by simulation, using a butt welded joint of thick pipes. In the simulation, the reliable order of the strains measured by the present NDM is very important, and was considered as 10 micro. The measurable regions by the present NDM were assumed. Under the above conditions, the data (the residual elastic strains assumed to be measured by the NDM) were made, and used in the ISM. As a result of such simulation, it has been cleared that the estimated residual stress has very high accuracy, if enough data are used. The required number of data is less than the ISM. (author)
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
International Nuclear Information System (INIS)
Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei
2015-01-01
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
Energy Technology Data Exchange (ETDEWEB)
Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)
2015-02-15
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.
Inelastic neutron scattering method in hard coal quality monitoring
International Nuclear Information System (INIS)
Cywicka-Jakiel, T.; Loskiewicz, J.; Tracz, G.
1994-07-01
Nuclear methods in mining industry and power generation plants are nowadays very important especially because of the need for optimization of combustion processes and reduction of environmental pollution. On-line analysis of coal quality not only economic benefits but contribute to environmental protection too. Neutron methods especially inelastic scattering and PGNAA are very useful for analysis of coal quality where calorific valve, ash and moisture content are the most important. Using Pu-Be or Am-Be isotopic sources and measuring carbon 4.43 MeV γ-rays from neutron inelastic scattering: 12 C(n,n'γ) 12 C we can evaluate calorific valve in hard coals with precision better than in PGNAA method. This is mainly because of large cross-section for inelastic scattering and the strong correlation between carbon content and calorific value shown in the paper for different coal basins. The influence of moisture on 4.43 MeV carbon γ-rays in considered in the paper in theoretical and experimental aspects and appropriate formula is introduced. Also the possibilities of determine ash, moisture, Cl, Na and Si in coal are shown. (author). 11 refs, 15 figs
Inelastic neutron scattering method in hard coal quality monitoring
Energy Technology Data Exchange (ETDEWEB)
Cywicka-Jakiel, T.; Loskiewicz, J.; Tracz, G. [Institute of Nuclear Physics, Cracow (Poland)
1994-07-01
Nuclear methods in mining industry and power generation plants are nowadays very important especially because of the need for optimization of combustion processes and reduction of environmental pollution. On-line analysis of coal quality not only economic benefits but contribute to environmental protection too. Neutron methods especially inelastic scattering and PGNAA are very useful for analysis of coal quality where calorific valve, ash and moisture content are the most important. Using Pu-Be or Am-Be isotopic sources and measuring carbon 4.43 MeV {gamma}-rays from neutron inelastic scattering: {sup 12}C(n,n`{gamma}){sup 12}C we can evaluate calorific valve in hard coals with precision better than in PGNAA method. This is mainly because of large cross-section for inelastic scattering and the strong correlation between carbon content and calorific value shown in the paper for different coal basins. The influence of moisture on 4.43 MeV carbon {gamma}-rays in considered in the paper in theoretical and experimental aspects and appropriate formula is introduced. Also the possibilities of determine ash, moisture, Cl, Na and Si in coal are shown. (author). 11 refs, 15 figs.
Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.
Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S
2012-10-01
A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.
GPU-based high performance Monte Carlo simulation in neutron transport
Energy Technology Data Exchange (ETDEWEB)
Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br
2009-07-01
Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)
GPU-based high performance Monte Carlo simulation in neutron transport
International Nuclear Information System (INIS)
Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A.
2009-01-01
Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)
Discrimination methods between neutron and gamma rays for boron loaded plastic scintillators
Normand, S; Haan, S; Louvel, M
2002-01-01
Boron loaded plastic scintillators exhibit interesting properties for neutron detection in nuclear waste management and especially in investigating the amount of fissile materials when enclosed in waste containers. Combining a high thermal neutron efficiency and a low mean neutron lifetime, they are suitable in neutron multiplicity counting. However, due to their high sensitivity to gamma rays, pulse shape discrimination methods need to be developed in order to optimize the passive neutron assay measurement. From the knowledge of their physical properties, it is possible to separate the three kinds of particles that have interacted in the boron loaded plastic scintillator (gamma, fast neutron and thermal neutron). For this purpose, we have developed and compared the two well known discrimination methods (zero crossing and charge comparison) applied for the first time to boron loaded plastic scintillator. The setup for the zero crossing discrimination method and the charge comparison methods is thoroughly expl...
PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups
International Nuclear Information System (INIS)
Olson, A.P.
1964-02-01
PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)
International Nuclear Information System (INIS)
Iwashita, H.; Iwasa, H.; Hiraga, F.; Kamiyama, T.; Kiyanagi, Y.; Suzuki, J.; Shinohara, T.; Oku, T.; Shimizu, H.M.
2009-01-01
A pulsed sextupole electromagnetic lens with suitably controlled time-dependent magnetic field can in principle focus pulsed neutrons at the same focal point over a wide range of wavelength as the lens removes aberrations. However, in fact, it is difficult to focus neutrons over a wide range of wavelength because attenuation of a practical pulsed sextupole electromagnet is faster than an ideal case. We have devised a method of canceling the difference between the practical pulsed sextupole magnetic field and the ideal magnetic field with the use of a permanent sextupole magnet. We performed simulation calculations to investigate the feasibility of this method, and it was shown that focusing wavelength range spread compared with the case using a pulsed magnetic lens only. This result indicates the usefulness of the method.
International Nuclear Information System (INIS)
Fevotte, F.
2008-01-01
At the various stages of a nuclear reactor's life, numerous studies are needed to guaranty the safety and efficiency of the design, analyse the fuel cycle, prepare the dismantlement, and so on. Due to the extreme difficulty to take extensive and accurate measurements in the reactor core, most of these studies are numerical simulations. The complete numerical simulation of a nuclear reactor involves many types of physics: neutronics, thermal hydraulics, materials, control engineering, Among these, the neutron transport simulation is one of the fundamental steps, since it allows computation - among other things - of various fundamental values such as the power density (used in thermal hydraulics computations) or fuel burn-up. The neutron transport simulation is based on the Boltzmann equation, which models the neutron population inside the reactor core. Among the various methods allowing its numerical solution, much interest has been devoted in the past few years to the Method of Characteristics in unstructured meshes (MOC), since it offers a good accuracy and operates in complicated geometries. The aim of this work is to propose improvements of the calculation scheme bound on the two dimensions MOC, in order to decrease the needed resources number. (A.L.B.)
DANTE, Activation Analysis Neutron Spectra Unfolding by Covariance Matrix Method
International Nuclear Information System (INIS)
Petilli, M.
1981-01-01
1 - Description of problem or function: The program evaluates activation measurements of reactor neutron spectra and unfolds the results for dosimetry purposes. Different evaluation options are foreseen: absolute or relative fluxes and different iteration algorithms. 2 - Method of solution: A least-square fit method is used. A correlation between available data and their uncertainties has been introduced by means of flux and activity variance-covariance matrices. Cross sections are assumed to be constant, i.e. with variance-covariance matrix equal to zero. The Lagrange multipliers method has been used for calculating the solution. 3 - Restrictions on the complexity of the problem: 9 activation experiments can be analyzed. 75 energy groups are accepted
Neutron transport and Montecarlo method: analysis and revision
International Nuclear Information System (INIS)
Perlado, J.M.
1982-01-01
The resolution of the neutron transport equation by the Montecarlo method is presented. Coming from an extensive discussion on the best formulation of that equation in order to be treated through the mentioned method, the theoretical bases of the estimator and random-walk generation is extensively explained. The most general expression for the estimators in different physical situations, each with a diverse random-walk, is included in this basical theoretical part. Furthemore, a large revision on the variance reduction methods is made. Its theoretical presentation is claimed to be in connection with the need for each one of them. The use of the adjoint equation, as a part of the importance sampling, Russian Roulette, splitting, exponential transform, conditional and correlated Montecarlo, and one-collision and next-event extimators, are discussed. Finally, come comments in the presentation of the last works on the theoretical prediction of errors in the generation of estimators-random walks are made. (author)
Nested element method in multidimensional neutron diffusion calculations
International Nuclear Information System (INIS)
Altiparmakov, D.V.
1983-01-01
A new numerical method is developed that is particularly efficient in solving the multidimensional neutron diffusion equation in geometrically complex systems. The needs for a generally applicable and fast running computer code have stimulated the inroad of a nonclassical (R-function) numerical method into the nuclear field. By using the R-functions, the geometrical components of the diffusion problem are a priori analytically implemented into the approximate solution. The class of functions, to which the approximate solution belongs, is chosen as close to the exact solution class as practically acceptable from the time consumption point of view. That implies a drastic reduction of the number of degrees of freedom, compared to the other methods. Furthermore, the reduced number of degrees of freedom enables calculation of large multidimensional problems on small computers
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
Digital module for neutron flux measurement by Campbell method
International Nuclear Information System (INIS)
Baratte, G.
1987-02-01
The study reported here concerns a wide range measurement channel for reactor control instrumentation but it may also be useful for specific measurements requiring the Campbell method. A wide range measurement channel allows the processing of the signal issued from a single fission chamber so it's possible to insure control of nuclear reactors in three different running modes: pulse processing, fluctuations and current. The study described in this note includes three parts: - the analogical wide range neutron measurement channel is presented in the first chapter; the fluctuation mode is thoroughly studied; the results of tests and proper limitations of analogical processing are summarized. A theoretical study of the neutron flux measurement by numerical calculation of the fluctuation signal variance is given in the second chapter. The digital module is described in the third chapter; the results of experiments are analysed. The validity of the digital method is proved by means of a practical realisation. The performances obtained with the digital fluctuation test model may be compared with those given by the analogical fluctuation channel which can be used for the control of lower fission rates. The digital module may also be used for any fluctuation measurement where very short response time and broad spectral band of analysis are not strictly necessary [fr
Simultaneous analysis of qualitative parameters of solid fuel using complex neutron gamma method
International Nuclear Information System (INIS)
Dombrovskij, V.P.; Ajtsev, N.I.; Ryashchikov, V.I.; Frolov, V.K.
1983-01-01
A study was made on complex neutron gamma method for simultaneous analysis of carbon content, ash content and humidity of solid fuel according to gamma radiation of inelastic fast neutron scattering and radiation capture of thermal neutrons. Metrological characteristics of pulse and stationary neutron gamma methods for determination of qualitative solid fuel parameters were analyzed, taking coke breeze as an example. Optimal energy ranges of gamma radiation detection (2-8 MeV) were determined. The advantages of using pulse neutron generator for complex analysis of qualitative parameters of solid fuel in large masses were shown
A simple method to calculate the neutron flow through full ducts
International Nuclear Information System (INIS)
Faik Ouahab, Z.; Jehouani, A.; Ghassoun, J.; Senhou, N.; Groetz, J.E.
2010-01-01
Summary of a study of assessment of the probability for neutrons to be guided in a full duct with a square cross section and doubly bent. Two software have been developed, based on the Monte Carlo simulation, to compute the neutron transmission probability at the end of the duct. Results are in good agreement with that obtained with the MCNP-5 code. The neutron flow and probability at the duct end have been determined for different materials and different duct dimensions
Energy Technology Data Exchange (ETDEWEB)
Nenot, M.L
2003-07-15
Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10{sup -2} Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)
Response of a BGO detector to photon and neutron sources simulations and measurements
Vincke, H H; Fabjan, Christian Wolfgang; Otto, T
2002-01-01
In this paper Monte Carlo simulations (FLUKA) and measurements of the response of a BGO detector are reported. %For the measurements different radioactive sources were used to irradiate the BGO crystal. For the measurements three low-energy photon emitters $\\left({}^{60}\\rm{Co},\\right.$ ${}^{54}\\rm{Mn},$ $\\left. {}^{137}\\rm{Cs}\\right)$ were used to irradiate the BGO from various distances and angles. The neutron response was measured with an Am--Be neutron source. Simulations of the experimental irradiations were carried out. Our study can also be considered as a benchmark for FLUKA in terms of its reliability to predict the detector response of a BGO scintillator.
Preliminary experiment of fast neutron imaging with direct-film method
International Nuclear Information System (INIS)
Pei Yuyang; Tang Guoyou; Guo Zhiyu; Zhang Guohui
2005-01-01
A preliminary experiment is conducted with direct-film method under the condition that fast neutron is generated by the reaction of 9 Be(d, n) on the Beijing University 4.5 MV Van de Graaff, whose energy is lower than 7 MeV. Basic characteristics of direct-film neutron radiography system are investigated with the help of samples in different materials, different thickness and holes of different diameter. The fast neutron converter, which is vital for fast neutron imaging, is produced with the materials made in China. The result indicates that fast neutron converter can meet the requirement of fast neutron imaging; further research of fast neutron imaging can be conducted on the accelerator and neutron-generator in China. (authors)
Neutron spectrum determination of d(20)+Be source reaction by the dosimetry foils method
Stefanik, Milan; Bem, Pavel; Majerle, Mitja; Novak, Jan; Simeckova, Eva
2017-11-01
The cyclotron-based fast neutron generator with the thick beryllium target operated at the NPI Rez Fast Neutron Facility is primarily designed for the fast neutron production in the p+Be source reaction at 35 MeV. Besides the proton beam, the isochronous cyclotron U-120M at the NPI provides the deuterons in the energy range of 10-20 MeV. The experiments for neutron field investigation from the deuteron bombardment of thick beryllium target at 20 MeV were performed just recently. For the neutron spectrum measurement of the d(20)+Be source reaction, the dosimetry foils activation method was utilized. Neutron spectrum reconstruction from resulting reaction rates was performed using the SAND-II unfolding code and neutron cross-sections from the EAF-2010 nuclear data library. Obtained high-flux white neutron field from the d(20)+Be source is useful for the intensive irradiation experiments and cross-section data validation.
Neutron contamination of Varian Clinac iX 10 MV photon beam using Monte Carlo simulation
International Nuclear Information System (INIS)
Yani, S; Haryanto, F; Arif, I; Tursinah, R; Rhani, M F; Soh, R C X
2016-01-01
High energy medical accelerators are commonly used in radiotherapy to increase the effectiveness of treatments. As we know neutrons can be emitted from a medical accelerator if there is an incident of X-ray that hits any of its materials. This issue becomes a point of view of many researchers. The neutron contamination has caused many problems such as image resolution and radiation protection for patients and radio oncologists. This study concerns the simulation of neutron contamination emitted from Varian Clinac iX 10 MV using Monte Carlo code system. As neutron production process is very complex, Monte Carlo simulation with MCNPX code system was carried out to study this contamination. The design of this medical accelerator was modelled based on the actual materials and geometry. The maximum energy of photons and neutron in the scoring plane was 10.5 and 2.239 MeV, respectively. The number and energy of the particles produced depend on the depth and distance from beam axis. From these results, it is pointed out that the neutron produced by linac 10 MV photon beam in a typical treatment is not negligible. (paper)
International Nuclear Information System (INIS)
Miklaszewski, R.; Drozdowicz, K.; Wiacek, U.; Dworak, D.; Gribkov, V.
2011-01-01
Recent progress in the development of a single-pulse Nanosecond Impulse Neutron Investigation System (NINIS) intended for interrogation of hidden objects (explosives and other illicit materials) by means of measuring elastically scattered neutrons is presented in this paper. The method is based on the well know fact that nuclide-specific information is present in the scattered neutron field. The method uses very bright neutron pulses having duration of the order of few nanoseconds, generated by a dense plasma focus (DPF) devices filled with a pure deuterium or deuterium-tritium mixture as a working gas. Very short duration of the neutron pulse, its high brightness and mono-chromaticity allow to use the time-of-flight method with bases of about few meters to distinguish signals from neutrons scattered by different elements. Results of the Monte Carlo simulations of the scattered neutron field from several compounds (explosives and everyday use materials) are presented in the paper. The MCNP5 code has been used to get information on the angular and energy distributions of the neutrons scattered by the above mentioned compounds assuming the initial neutron energy equal to 2.45 MeV (D-D). A new input has been elaborated that allows the modelling of not only a spectrum of the neutrons scattered at different angles but also their time history from the moment of generation up to detection. Such an approach allows getting approximate signals as registered by scintillator + photomultiplier probes placed at various distances from the scattering object, demonstrating a principal capability of the method to identify an elemental content of the inspected objects. Preliminary results of the MCNP modelling of the interrogation process of the airport luggage containing several illicit objects are presented as well. (authors)
International Nuclear Information System (INIS)
Khorasani, A.; Mousavi Shalmani, M. A.; Piervali Bieranvand, N.
2004-01-01
The effective use of soil water requires frequent and accurate measurements; the technique should be rapid, reliable, simple, cost effective and non-destructive. In this study, the soil moisture neutron probe , Diviner 2000, a capacitance probe and time domain reflectometry were compared in a field tomato experiment carried out at the FAO/IAEA Agriculture and Biotechnology laboratory at Seibersdorf, Austria. The experiment consists of two irrigation treatment furrow versus drip and two nitrogen levels(100,200 Kg N/ha), with each treatment replicated three times, giving a total of twelve plots. The size of the plot was (3.4 x 5 m ). One Soil Moisture Neutron Probe aluminium access tube, one Diviner PVC access tube and one EnviroSCAN access tube were installed in each plot using the slurry method, 30 cm on each side of the tomato plants. In addition, Time Domain Reflectometry probe were installed vertically to monitor over the growth season from June to September. To analyze the results of this experiment, we compared directly The e soil moisture 0-60 cm obtained by different methods. The soil water deficit was calculated for furrow and drip treatment. There was generally good agreement between the changes in soil water deficit as measured and the simulated ones for soil moisture neutron probe
Energy Technology Data Exchange (ETDEWEB)
Herzkamp, Matthias; Engels, Ralf; Kemmerling, Guenter [ZEA-2, Forschungszentrum Juelich (Germany); Brueckel, Thomas [JCNS, Forschungszentrum Juelich (Germany); Stahl, Achim [III. Physikalisches Institut B, RWTH Aachen (Germany); Waasen, Stefan van [ZEA-2, Forschungszentrum Juelich (Germany); Faculty of Engineering, University of Duisburg-Essen (Germany)
2015-07-01
In neutron scattering experiments it is important to have position sensitive large scale detectors for thermal neutrons. A detector based on a neutron scintillator with wave length shifting fibers is a new kind of such a detector. We present the simulation of the detector based on the microscopic structure of the scintillation material of the mentioned detector. It consists of a converter and a scintillation powder bound in a matrix. The converter in our case is lithium fluoride with enriched lithium 6, to convert thermal neutrons into high energetic alpha and triton particles. The scintillation material is silver doped zinc sulfide. We show that pulse height spectra obtained by these scintillators can be be explained by the simple model of randomly distributed spheres of zinc sulfide and lithium fluoride. With this model, it is possible to optimise the mass ratio of zinc sulfide to lithium fluoride with respect to detection efficiency and/or energy deposition in zinc sulfide.
Measurement and simulation of neutron detection efficiency in lead-scintillating fiber calorimeters
Energy Technology Data Exchange (ETDEWEB)
Anelli, M.; Bertolucci, S. [Laboratori Nazionali di Frascati, INFN (Italy); Bini, C. [Dipartimento di Fisica dell' Universita ' La Sapienza' , Roma (Italy); INFN Sezione di Roma, Roma (Italy); Branchini, P. [INFN Sezione di Roma Tre, Roma (Italy); Curceanu, C. [Laboratori Nazionali di Frascati, INFN (Italy); De Zorzi, G.; Di Domenico, A. [Dipartimento di Fisica dell' Universita ' La Sapienza' , Roma (Italy); INFN Sezione di Roma, Roma (Italy); Di Micco, B. [Dipartimento di Fisica dell' Universita ' Roma Tre' , Roma (Italy); INFN Sezione di Roma Tre, Roma (Italy); Ferrari, A. [Fondazione CNAO, Milano (Italy); Fiore, S.; Gauzzi, P. [Dipartimento di Fisica dell' Universita ' La Sapienza' , Roma (Italy); INFN Sezione di Roma, Roma (Italy); Giovannella, S., E-mail: simona.giovannella@lnf.infn.i [Laboratori Nazionali di Frascati, INFN (Italy); Happacher, F. [Laboratori Nazionali di Frascati, INFN (Italy); Iliescu, M. [Laboratori Nazionali di Frascati, INFN (Italy); IFIN-HH, Bucharest (Romania); Martini, M. [Laboratori Nazionali di Frascati, INFN (Italy); Dipartimento di Energetica dell' Universita ' La Sapienza' , Roma (Italy); Miscetti, S. [Laboratori Nazionali di Frascati, INFN (Italy); Nguyen, F. [Dipartimento di Fisica dell' Universita ' Roma Tre' , Roma (Italy); INFN Sezione di Roma Tre, Roma (Italy); Passeri, A. [INFN Sezione di Roma Tre, Roma (Italy); Prokofiev, A. [Svedberg Laboratory, Uppsala University (Sweden); Sciascia, B. [Laboratori Nazionali di Frascati, INFN (Italy)
2009-12-15
The overall detection efficiency to neutrons of a small prototype of the KLOE lead-scintillating fiber calorimeter has been measured at the neutron beam facility of The Svedberg Laboratory, TSL, Uppsala, in the kinetic energy range [5-175] MeV. The measurement of the neutron detection efficiency of a NE110 scintillator provided a reference calibration. At the lowest trigger threshold, the overall calorimeter efficiency ranges from 30% to 50%. This value largely exceeds the estimated 8-15% expected if the response were proportional only to the scintillator equivalent thickness. A detailed simulation of the calorimeter and of the TSL beam line has been performed with the FLUKA Monte Carlo code. First data-MC comparisons are encouraging and allow to disentangle a neutron halo component in the beam.
Measurement and simulation of neutron detection efficiency in lead-scintillating fiber calorimeters
International Nuclear Information System (INIS)
Anelli, M.; Bertolucci, S.; Bini, C.; Branchini, P.; Curceanu, C.; De Zorzi, G.; Di Domenico, A.; Di Micco, B.; Ferrari, A.; Fiore, S.; Gauzzi, P.; Giovannella, S.; Happacher, F.; Iliescu, M.; Martini, M.; Miscetti, S.; Nguyen, F.; Passeri, A.; Prokofiev, A.; Sciascia, B.
2009-01-01
The overall detection efficiency to neutrons of a small prototype of the KLOE lead-scintillating fiber calorimeter has been measured at the neutron beam facility of The Svedberg Laboratory, TSL, Uppsala, in the kinetic energy range [5-175] MeV. The measurement of the neutron detection efficiency of a NE110 scintillator provided a reference calibration. At the lowest trigger threshold, the overall calorimeter efficiency ranges from 30% to 50%. This value largely exceeds the estimated 8-15% expected if the response were proportional only to the scintillator equivalent thickness. A detailed simulation of the calorimeter and of the TSL beam line has been performed with the FLUKA Monte Carlo code. First data-MC comparisons are encouraging and allow to disentangle a neutron halo component in the beam.
Development of Neutronics Model for ShinKori Unit 1 Simulator
Energy Technology Data Exchange (ETDEWEB)
Hong, JinHyuk; Lee, MyeongSoo; Lee, SeungHo; Suh, JungKwan; Hwang, DoHyun [KEPRI, Daejeon (Korea, Republic of)
2008-05-15
ShinKori-Unit 1 and 2 is being built in the Kori site which will be operated at 2815 MWt of thermal core power. The purpose of this paper is to report on the performance of the developed neutronics model of ShinKori Unit 1 and 2. Also this report includes the convenient tool (XS2R5) for processing the large quantity of information received from the DIT/ROCS model and generating cross-sections. The neutronics model is based on the NESTLE code inserted to RELAP5/MOD3 thermal-hydraulics analysis code which was funded as FY-93 LDRD Project 7201 and is running on the commercial simulator environment tool (the 3KeyMaster{sup TM} of the WSC). As some examples for the verification of the developed neutronics model, some figures are provided. The output of the developed neutronics model is in accord with the Preliminary Safety Analysis Report (PSAR) of the reference plant.
Means and method for controlling the neutron output of a neutron generator tube
International Nuclear Information System (INIS)
Langford, O.M.; Peelman, H.E.
1980-01-01
A gas filled neutron tube in a nuclear well logging tool has a target an ion source voltage and a replenisher connected to ground. A negative high voltage is applied to the target by a power supply also providing a target current corresponding to the neutron output of the neutron generator tube. A constant current source provides a constant current. A network receiving the target current and the constant current provides a portion of the constant current as a replenisher current which is applied to the replenisher in a neutron generating tube. The network controls the magnitude of the replenisher current in accordance with the target current so as to control the neutron output of the neutron generating tube. (auth)
Method for controlling an accelerator-type neutron source, and a pulsed neutron source
International Nuclear Information System (INIS)
Givens, W.W.
1991-01-01
The patent deals with an accelerator-type neutron source which employs a target, an ionization section and a replenisher for supplying accelerator gas. A positive voltage pulse is applied to the ionization section to produce a burst of neutrons. A negative voltage pulse is applied to the ionization section upon the termination of the positive voltage pulse to effect a sharp cut-off to the burst of neutrons. 4 figs
Monte-Carlo simulations of neutron shielding for the ATLAS forward region
Stekl, I; Kovalenko, V E; Vorobel, V; Leroy, C; Piquemal, F; Eschbach, R; Marquet, C
2000-01-01
The effectiveness of different types of neutron shielding for the ATLAS forward region has been studied by means of Monte-Carlo simulations and compared with the results of an experiment performed at the CERN PS. The simulation code is based on GEANT, FLUKA, MICAP and GAMLIB. GAMLIB is a new library including processes with gamma-rays produced in (n, gamma), (n, n'gamma) neutron reactions and is interfaced to the MICAP code. The effectiveness of different types of shielding against neutrons and gamma-rays, composed from different types of material, such as pure polyethylene, borated polyethylene, lithium-filled polyethylene, lead and iron, were compared. The results from Monte-Carlo simulations were compared to the results obtained from the experiment. The simulation results reproduce the experimental data well. This agreement supports the correctness of the simulation code used to describe the generation, spreading and absorption of neutrons (up to thermal energies) and gamma-rays in the shielding materials....
Uranium analysis by neutron induced fissionography method using solid state nuclear track detectors
International Nuclear Information System (INIS)
Akyuez, T.; Tretyakova, S. P.; Guezel, T.; Akyuz, S.
1999-01-01
In this study total twenty samples (eight reference materials and twelve sediment samples) were analysed for their uranium content which is in the range of 1-17 μg/g, by neutron induced fissionography (NIF) method using solid state nuclear track detectors (SSNTDs) in comparison with the results of neutron activation analysis (NAA), delayed neutron counting (DNC) technique or fluorometric method. It is found that NIF method using SSNTDs is very sensitive for analysis of uranium
Uranium analysis by neutron induced fissionography method using solid state nuclear track detectors
Akyuez, T; Guezel, T; Akyuz, S
1999-01-01
In this study total twenty samples (eight reference materials and twelve sediment samples) were analysed for their uranium content which is in the range of 1-17 mu g/g, by neutron induced fissionography (NIF) method using solid state nuclear track detectors (SSNTDs) in comparison with the results of neutron activation analysis (NAA), delayed neutron counting (DNC) technique or fluorometric method. It is found that NIF method using SSNTDs is very sensitive for analysis of uranium.
Method and apparatus for formation logging using position sensitive neutron detectors
International Nuclear Information System (INIS)
Gadken, L.L.
1986-01-01
This patent describes a method for logging earth formations using position sensitive neutron detectors. The method consists of: 1) Irradiation of earth formations in the vicinity of a well borehole with a source of fast neutrons. 2) At four longitudinally spaced distances from the neutron source in the borehole, the epithermal neutron population is detected. Each of the four separate populations is detected in an epithermally sensitive and substantially thermally insensitive portion of the same position sensitive neutron detector. A representative signal from each is then individually generated. 3) First, second, third, and fourth neutron population representative signals are combined. They derive a simultaneous measurement signal. This signal is functionally related to the porosity and also a signal functionally related to a neutron characteristic length of the earth formations in the vicinity of the borehole
Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation
Energy Technology Data Exchange (ETDEWEB)
Suman, Vitisha [Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sarkar, P.K., E-mail: pksarkar02@gmail.com [Manipal Centre for Natural Sciences, Manipal University, Manipal 576104 (India)
2014-02-11
A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra.
Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation
International Nuclear Information System (INIS)
Suman, Vitisha; Sarkar, P.K.
2014-01-01
A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra
Flux weighted method for solution of stiff neutron dynamic equations and its application
International Nuclear Information System (INIS)
Li Huiyun; Jiao Huixian
1987-12-01
To analyze reactivity event for nuclear power plants, it is necessary to solve the neutron dynamic equations, which is a group of typical stiff constant differential equations. Very small time steps could only be adopted when the group of equations is solved by common methods. However, a large time steps might be selected if the Flux Weighted Medthod introduced in this paper is used. Generally, weighted factor θ i1 is set as a constant. Naturally, this treatment method can decrease the accuracy of calculation for the increase of the steadiness of solving the equations. An accurate theoretical formula of 4 x 4 matrix of θ i1 is rigorously derived so that the accuracy of calculation is ensured, as well as the steadiness of solved equations is increased. This method have the advantage over classical Runge-kutta Method and other methods. The time steps could be increased by a factor of 1 ∼ 3 orders of magnitude so as to save a lot of computating time. The programe solving neutron dynamic equation, which is prepared by using Flux Weighted Method, could be sued for real time analog of training simulator, as well as for analysis and computation of reactivity event (including rod jumping out event)
Comparison of dynamic compensation methods for delayed self-powered neutron detector
International Nuclear Information System (INIS)
In, Wang Kee; Kim, Joon Sung; Auh, Geun Sun; Yoon, Tae Young
1993-01-01
Dynamic compensation methods for rhodium self-powered neutron detector have been developed by Banda and Hoppe to compensate for the time delay associated with detector signals. The time delay is due to the decay of the neutron-activated rhodium and results in delayed detector response. Two digital dynamic compensation methods, were compared for step change of neutron flux in this paper. The inverse kinetics method gave slightly better response time and noise gain. However, the inverse kinetics method also showed overshooting of neutron flux for the step change. (Author)
International Nuclear Information System (INIS)
Fiechtner, A.; Boschung, M.; Wernli, C.
2007-01-01
In the framework of the EVIDOS (Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields) project, funded by the EC, measurements with PADC personal neutron dosemeters were carried out at several workplace fields of the nuclear industry and at simulated workplace fields. The measured personal neutron dose equivalents of the PADC personal neutron dosemeter are compared with values that were assessed within the EVIDOS project by other partners. The detection limits for different spectra types are given. In cases were the neutron dose was too low to be measured by the PADC personal neutron dosemeter, the response is estimated by convoluting the responses to monoenergetic neutrons with the dose energy distribution measured within EVIDOS. The advantages and limitations of the PADC personal neutron dosemeter are discussed. (authors)
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)
Method of energy calibration of the TANSY neutron detectors
International Nuclear Information System (INIS)
Hoek, M.; Drozdowicz, K.; Aronsson, D.
1990-03-01
A method to calibrate an array of scintillation neutron detectors, using a γ source, is presented. The count rate is measured as a function of high voltage at a given discrimination level. The obtained distribution is differentiated and a maximum value is determined which corresponds to the voltage at which the gamma peak passes through the discrimination level. By repeating the measurement at different discrimination levels the experimental dependence between the discrimination level and the high voltage is found as a straight line in a log-log diagram. Two calibration parameter for each detector are determined from a fit of these straight lines. A recalculation from the energy of the used γ source to any other energy is then possible and the obtained relation can be used to calculate discrimination levels and high voltages for each detector. Verification procedures are described. (authors)
Virtual experiments: the ultimate aim of neutron ray-tracing simulations
DEFF Research Database (Denmark)
Lefmann, Kim; Willendrup, Peter Kjær; Udby, Linda
2008-01-01
We define a virtual neutron experiment as a complete simulation of an experiment, from source over sample to detector. The virtual experiment (VE) will ideally interface with the instrument control software for the input and with standard data analysis packages for the virtual data output. Virtual...
Simulations of chopper jitter at the LET neutron spectrometer at the ISIS TS2
DEFF Research Database (Denmark)
Klenø, Kaspar Hewitt; Lefmann, Kim; Willendrup, Peter Kjær
2014-01-01
The effect of uncertainty in chopper phasing (jitter) has been investigated for the high-resolution time-of-flight spectrometer LET at the ISIS second target station. The investigation is carried out using virtual experiments, with the neutron simulation package McStas, where the chopper jitter i...
Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety
International Nuclear Information System (INIS)
Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.
1993-01-01
This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized
Energy Technology Data Exchange (ETDEWEB)
Elekes, Z., E-mail: z.elekes@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Aumann, T. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Technische Universität Darmstadt, Darmstadt (Germany); Bemmerer, D. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Boretzky, K. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Caesar, C. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Technische Universität Darmstadt, Darmstadt (Germany); Cowan, T.C. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Technische Universität Dresden, Dresden (Germany); Hehner, J.; Heil, M. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Kempe, M. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Rossi, D. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Röder, M. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Technische Universität Dresden, Dresden (Germany); Simon, H. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Sobiella, M.; Stach, D. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Reinhardt, T. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Technische Universität Dresden, Dresden (Germany); Wagner, A.; Yakorev, D. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Zilges, A. [Universität zu Köln, Köln (Germany); Zuber, K. [Technische Universität Dresden, Dresden (Germany)
2013-02-11
Resistive plate chamber (RPC) prototypes of 2 m length were simulated and built. The experimental tests using a 31 MeV electron beam, discussed in details, showed an efficiency higher than 90% and an excellent time resolution of around σ=100ps. Furthermore, comprehensive simulations were performed by GEANT4 toolkit in order to study the possible use of these RPCs for fast neutron (200 MeV–1 GeV) detection and multi-neutron event identification. The validation of simulation parameters was carried out via a comparison to experimental data. A possible setup for invariant mass spectroscopy of multi-neutron emission is presented and the characteristics are discussed. The results show that the setup has a high detection efficiency. Its capability of determining the momentum of the outgoing neutrons and reconstructing the relative energy between the fragments from nuclear reactions is demonstrated for different scenarios.
Measurement of Neutron Energy Spectrum Emitted by Cf-252 Source Using Time-of-Flight Method
Energy Technology Data Exchange (ETDEWEB)
Lee, Cheol Ho; Son, Jaebum; Kim, Tae Hoon; Lee, Sangmin; Kim, Yong-Kyun [Hanyang University, Seoul (Korea, Republic of)
2016-10-15
The techniques proposed to detect the neutrons usually require the detection of a secondary recoiling nucleus in a scintillator (or other type of detector) to indicate the rare collision of a neutron with a nucleus. This is the same basic technique, in this case detection of a recoil proton that was used by Chadwick in the 1930 s to discover and identify the neutron and determine its mass. It is primary technique still used today for detection of fast neutron, which typically involves the use of a hydrogen based organic plastic or liquid scintillator coupled to a photo-multiplier tube. The light output from such scintillators is a function of the cross section and nuclear kinematics of the n + nucleus collision. With the exception of deuterated scintillators, the scintillator signal does not necessarily produce a distinct peak in the scintillator spectrum directly related to the incident neutron energy. Instead neutron time-of-flight (TOF) often must be utilized to determine the neutron energy, which requires generation of a prompt start signal from the nuclear source emitting the neutrons. This method takes advantage of the high number of prompt gamma rays. The Time-of-Flight method was used to measure neutron energy spectrum emitted by the Cf-252 neutron source. Plastic scintillator that has a superior discrimination ability of neutron and gamma-ray was used as a stop signal detector and liquid scintillator was used as a stat signal detector. In experiment, neutron and gamma-ray spectrum was firstly measured and discriminated using the TOF method. Secondly, neutron energy spectrum was obtained through spectrum analysis. Equation of neutron energy spectrum that was emitted by Cf-252 source using the Gaussian fitting was obtained.
Energy Technology Data Exchange (ETDEWEB)
Hassanzadeh, M. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Feghhi, S.A.H., E-mail: a_feghhi@sbu.ac.ir [Department of Radiation Application, Shahid Beheshti University, G.C., Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of)
2013-09-15
Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ{sup †}) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ{sup †}) and importance weighted neutron generation time (Λ{sup †}) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ{sup †} is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.
International Nuclear Information System (INIS)
Hassanzadeh, M.; Feghhi, S.A.H.; Khalafi, H.
2013-01-01
Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ † ) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ † ) and importance weighted neutron generation time (Λ † ) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ † is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function
Monte Carlo simulation of neutron detection efficiency for NE213 scintillation detector
International Nuclear Information System (INIS)
Xi Yinyin; Song Yushou; Chen Zhiqiang; Yang Kun; Zhangsu Yalatu; Liu Xingquan
2013-01-01
A NE213 liquid scintillation neutron detector was simulated by using the FLUKA code. The light output of the detector was obtained by transforming the secondary particles energy deposition using Birks formula. According to the measurement threshold, detection efficiencies can be calculated by integrating the light output. The light output, central efficiency and the average efficiency as a function of the front surface radius of the detector, were simulated and the results agreed well with experimental results. (authors)
Khezripour, S.; Negarestani, A.; Rezaie, M. R.
2017-08-01
Micromegas detector has recently been used for high-energy neutron (HEN) detection, but the aim of this research is to investigate the response of the Micromegas detector to low-energy neutron (LEN). For this purpose, a Micromegas detector (with air, P10, BF3, 3He and Ar/BF3 mixture) was optimized for the detection of 60 keV neutrons using the MCNP (Monte Carlo N Particle) code. The simulation results show that the optimum thickness of the cathode is 1 mm and the optimum of microgrid location is 100 μm above the anode. The output current of this detector for Ar (3%) + BF3 (97%) mixture is greater than the other ones. This mixture is considered as the appropriate gas for the Micromegas neutron detector providing the output current for 60 keV neutrons at the level of 97.8 nA per neutron. Consecuently, this detector can be introduced as LEN detector.
Development of concept and neutronic calculation method for large LMFBR core
International Nuclear Information System (INIS)
Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.
1991-01-01
Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)
Capture analysis of element content of a substance with other neutron methods
International Nuclear Information System (INIS)
Kurbanov, B.I.
2004-01-01
Full text: Neutron analysis method of determining element composition have found wide range of applications in industry thanks to different types of interaction of neutron with substances /1/. With the aim of widening the range of problems to be solved, on the basis of the device /2/ for determining the element content of substance, possibilities of combining the method based on the use of neutron capture gamma-ray spectrometry with other neutron methods, in particular neutron activation analysis and neutron absorption analysis were studied. In this radionuclide source ( 252 Cf) with the yield of 1,5 x 10 7 neutron/sec is used. By means of using neutron capture gamma radiation spectrometry the possibilities of determining some elements (H, B, N, S etc. ), which are not determined by very widely used method, activation analysis. These elements can be determined by both the semiconductor and scintillation detectors with parameters fitting the manufacturing requirements. And for a number of elements ( B, Cl, Cd, Sm, Gd) very high limits of determination ( up to 10- 5 %) are possible using semiconductor Ge (Li) -detectors with high resolution. Possibility of determination of some 'well' activated elements ( K, Al, Fe, Mn, Ti, Sc etc.) in samples of ore and products of their processing using the neutron-activation analysis. For 1 hour of irradiation on the experimental device quite accurate analytical peak, of these elements are obtained, allowing to determine them qualitatively. However, with decreasing neutron yield of radionuclide source it becomes more difficult to achieve the necessary parameters both in neutron capture and activation analysis. Experimental works on determination of some elements with large cross-sections of capture ( B, Cd, Sm ) by absorption of neutrons in the investigated substance, i.e. using the neutron absorption analysis method with absence of other large capture cross section elements in the samples being studied
A hybrid source-driven method to compute fast neutron fluence in reactor pressure vessel - 017
International Nuclear Information System (INIS)
Ren-Tai, Chiang
2010-01-01
A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribution of the peripheral assemblies in a nuclear reactor obtained from reactor core depletion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U-238 (such as U-234, U-235, U-236, Pu-239 etc) are replaced with U-238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly developed hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors. (authors)
A method for prediction of prompt fission neutron spectra
International Nuclear Information System (INIS)
Grashin, A.F.; Lepeshkin, M.V.
1988-01-01
Three-parameter formula for the prompt-fission-neutron integral spectrum is derived from a thermodynamical model. Two parameters, scission-neutron weight p = 11 % and anisotropy factor for accelerated fragments b = 10 %, are determined from experimental data, the same values being assumed for any type of fission. The thermodynamical theory provides the value of the third parameter, temperature τ, thus prognozing neutron spectrum and average energy with an error about 1 %. (author)
Simulation of a silicon neutron detector coated with TiB2 absorber
International Nuclear Information System (INIS)
Krapohl, D; Nilsson, H-E; Petersson, S; Slavicek, T; Thungström, G; Pospisil, S
2012-01-01
Neutron radiation cannot be directly detected in semiconductor detectors and therefore needs converter layers. Planar clean-room processing can be used in the manufacturing process of semiconductor detectors with metal layers to produce a cost-effective device. We used the Geant4 Monte-Carlo toolkit to simulate the performance of a semiconductor neutron detector. A silicon photo-diode was coated with vapour deposited titanium, aluminium thin films and a titaniumdiboride (TiB 2 ) neutron absorber layer. The neutron capture reaction 10B(n, alpha)7Li is taken advantage of to create charged particles that can be counted. Boron-10 has a natural abundance of about SI 19.8%. The emitted alpha particles are absorbed in the underlying silicon detector. We varied the thickness of the converter layer and ran the simulation with a thermal neutron source in order to find the best efficiency of the TiB 2 converter layer and optimize the clean room process.
Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods
International Nuclear Information System (INIS)
Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail
2016-01-01
In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)
Water flow measurements with the pulsed neutron activation method
International Nuclear Information System (INIS)
Linden, P.
1997-05-01
The objective of this work was to develop and study the feasibility of a flow-meter, based on the pulsed neutron activation method. It is a non-invasive method with good potential regarding accuracy. However, the ultimate accuracy has not been fully investigated before. Two series of flow rate measurements have been performed and analysed. The first series was done under moderately accurate flow calibration conditions to get sufficient confidence in the method and to get indication of the obtainable accuracy. The results were encouraging and further measurements with high accuracy flow calibration were planned. A dedicated loop was designed and built, and it was used with satisfactory performance. Two models have been used for analysis of recorded data; time weighting method and a fit to Taylor diffusion theory. The results show that the accuracy in mean flow velocity obtained from the used analysis models is in the range of 2-4% for Reynolds numbers greater than 10,000. Data recorded from high calibration measurements will also be used for validation of future calculations. 19 refs, 4 figs
A high-resolution neutron spectra unfolding method using the Genetic Algorithm technique
Mukherjee, B
2002-01-01
The Bonner sphere spectrometers (BSS) are commonly used to determine the neutron spectra within various nuclear facilities. Sophisticated mathematical tools are used to unfold the neutron energy distribution from the output data of the BSS. This paper highlights a novel high-resolution neutron spectra-unfolding method using the Genetic Algorithm (GA) technique. The GA imitates the biological evolution process prevailing in the nature to solve complex optimisation problems. The GA method was utilised to evaluate the neutron energy distribution, average energy, fluence and equivalent dose rates at important work places of a DIDO class research reactor and a high-energy superconducting heavy ion cyclotron. The spectrometer was calibrated with a sup 2 sup 4 sup 1 Am/Be (alpha,n) neutron standard source. The results of the GA method agreed satisfactorily with the results obtained by using the well-known BUNKI neutron spectra unfolding code.
Methods for Monte Carlo simulations of biomacromolecules.
Vitalis, Andreas; Pappu, Rohit V
2009-01-01
The state-of-the-art for Monte Carlo (MC) simulations of biomacromolecules is reviewed. Available methodologies for sampling conformational equilibria and associations of biomacromolecules in the canonical ensemble, given a continuum description of the solvent environment, are reviewed. Detailed sections are provided dealing with the choice of degrees of freedom, the efficiencies of MC algorithms and algorithmic peculiarities, as well as the optimization of simple movesets. The issue of introducing correlations into elementary MC moves, and the applicability of such methods to simulations of biomacromolecules is discussed. A brief discussion of multicanonical methods and an overview of recent simulation work highlighting the potential of MC methods are also provided. It is argued that MC simulations, while underutilized biomacromolecular simulation community, hold promise for simulations of complex systems and phenomena that span multiple length scales, especially when used in conjunction with implicit solvation models or other coarse graining strategies.
Lotfalizadeh, F.; Faghihi, R.; Bahadorzadeh, B.; Sina, S.
2017-07-01
Neutron spectrometry using a single-sphere containing dosimeters has been developed recently, as an effective replacement for Bonner sphere spectrometry. The aim of this study is unfolding the neutron energy spectra using GRNN artificial neural network, from the response of thermoluminescence dosimeters, TLDs, located inside a polyethylene sphere. The spectrometer was simulated using MCNP5. TLD-600 and TLD-700 dosimeters were simulated at different positions in all directions. Then the GRNN was used for neutron spectra prediction, using the TLDs' readings. Comparison of spectra predicted by the network with the real spectra, show that the single-sphere dosimeter is an effective instrument in unfolding neutron spectra.
Numerical methods used in simulation
International Nuclear Information System (INIS)
Caseau, Paul; Perrin, Michel; Planchard, Jacques
1978-01-01
The fundamental numerical problem posed by simulation problems is the stability of the resolution diagram. The system of the most used equations is defined, since there is a family of models of increasing complexity with 3, 4 or 5 equations although only models with 3 and 4 equations have been used extensively. After defining what is meant by explicit or implicit, the best established stability results is given for one-dimension problems and then for two-dimension problems. It is shown that two types of discretisation may be defined: four and eight point diagrams (in one or two dimensions) and six and ten point diagrams (in one or two dimensions). To end, some results are given on problems that are not usually treated very much, i.e. non-asymptotic stability and the stability of diagrams based on finite elements [fr
Methods for simulating turbulent phase screen
International Nuclear Information System (INIS)
Zhang Jianzhu; Zhang Feizhou; Wu Yi
2012-01-01
Some methods for simulating turbulent phase screen are summarized, and their characteristics are analyzed by calculating the phase structure function, decomposing phase screens into Zernike polynomials, and simulating laser propagation in the atmosphere. Through analyzing, it is found that, the turbulent high-frequency components are well contained by those phase screens simulated by the FFT method, but the low-frequency components are little contained. The low-frequency components are well contained by screens simulated by Zernike method, but the high-frequency components are not contained enough. The high frequency components contained will be improved by increasing the order of the Zernike polynomial, but they mainly lie in the edge-area. Compared with the two methods above, the fractal method is a better method to simulate turbulent phase screens. According to the radius of the focal spot and the variance of the focal spot jitter, there are limitations in the methods except the fractal method. Combining the FFT and Zernike method or combining the FFT method and self-similar theory to simulate turbulent phase screens is an effective and appropriate way. In general, the fractal method is probably the best way. (authors)
Energy Technology Data Exchange (ETDEWEB)
Englert, Matthias
2009-07-13
This thesis investigates the proliferation resistance of nuclear technologies on the basis of three case studies. After a brief description of the concept of proliferation resistance the utilized computer codes and methods are presented. The first case study investigates the potential of monolithic fuel for the conversion of one-fuel-element high-flux research reactors from highly enriched to low enriched uranium using the example of the german research reactor FRM-II. The second case study assesses the proliferation potential of future tokamak based fusion reactors by using neutronic simulations of a possible plutonium production. The third example investigates the proliferation potential of spallation neutron sources to produce nuclear weapon relevant material and the proliferation resistance of such facilities. (orig.)
Means and method for controlling the neutron output of a neutron generator tube
International Nuclear Information System (INIS)
1980-01-01
A specification is given for an energizing and regulating circuit for a gas filled neutron generator tube consisting of a target, an ion source and a replenisher, the circuit consisting of a power supply to provide a negative high voltage to the target and a target current corresponding to the neutron output of the tube, a constant current source, and control means connected to the power supply and to the constant current source, the control means being responsive to the target current to provide a portion of the constant current to the replenisher substantially to regulate the neutron output of the tube. (author)
International Nuclear Information System (INIS)
Wiacek, U.; Krynicka, E.
2005-02-01
Monte Carlo simulations of the pulsed neutron experiment in two- region systems (two concentric spheres and two coaxial finite cylinders) are presented. The MCNP code is used. Aqueous solutions of H 3 BO 3 or KCl are used in the inner region. The outer region is the moderator of Plexiglas. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances are used. The time-dependent thermal neutron transport is simulated when the inner region has a constant size and the external size of the surrounding outer region is variable. The time decay constant of the thermal neutron flux in the system is found in each simulation. The results of the simulations are compared with results of real pulsed neutron experiments on the corresponding systems. (author)
Ródenas, José
2017-11-01
All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.
Dynamics of crystalline acetanilide: Analysis using neutron scattering and computer simulation
Hayward, R. L.; Middendorf, H. D.; Wanderlingh, U.; Smith, J. C.
1995-04-01
The unusual temperature dependence of several optical spectroscopic vibrational bands in crystalline acetanilide has been interpreted as providing evidence for dynamic localization. Here we examine the vibrational dynamics of crystalline acetanilide over a spectral range of ˜20-4000 cm-1 using incoherent neutron scattering experiments, phonon normal mode calculations and molecular dynamics simulations. A molecular mechanics energy function is parametrized and used to perform the normal mode analyses in the full configurational space of the crystal i.e., including the intramolecular and intermolecular degrees of freedom. One- and multiphonon incoherent inelastic neutron scattering intensities are calculated from harmonic analyses in the first Brillouin zone and compared with the experimental data presented here. Phonon dispersion relations and mean-square atomic displacements are derived from the harmonic model and compared with data derived from coherent inelastic neutron scattering and neutron and x-ray diffraction. To examine the temperature effects on the vibrations the full, anharmonic potential function is used in molecular dynamics simulations of the crystal at 80, 140, and 300 K. Several, but not all, of the spectral features calculated from the molecular dynamics simulations exhibit temperature-dependent behavior in agreement with experiment. The significance of the results for the interpretation of the optical spectroscopic results and possible improvements to the model are discussed.
Leakage monitoring equipment of fuel element by delayed neutron method
International Nuclear Information System (INIS)
Ji Changsong; Zhang Shulan; Zhang Shuheng
1999-01-01
Based on monitoring results of delayed neutrons from reactor first circle water, the leakage of reactor fuel elements is monitored. A monitoring equipment consisted of an array of 3 He proportional counter tubes with 75 s delay has been developed. The neutron detection efficiency of 6.1% is obtained
Activation method for measurement of neutron spectrum parameters
International Nuclear Information System (INIS)
Efimov, B.V.; Demidov, A.M.; Ionov, V.S.; Konjaev, S.I.; Marin, S.V.; Bryzgalov, V.I.
2007-01-01
Experimental researches of spectrum parameters of neutrons at nuclear installations RRC KI are submitted. The installations have different designs of the cores, reflector, parameters and types of fuel elements. Measurements were carried out with use of the technique developed in RRC KI for irradiation resonance detectors UKD. The arrangement of detectors in the cores ensured possibility of measurement of neutron spectra with distinguished values of parameters. The spectrum parameters which are introduced by parametrical representation of a neutrons spectrum in the form corresponding to formalism Westcott. On experimental data were determinate absolute values of density neutron flux (DNF) in thermal and epithermal area of a spectrum (F t , f epi ), empirical dependence of temperature of neutron gas (Tn) on parameter of a rigidity of a spectrum (z), density neutron flux in transitional energy area of the spectrum. Dependences of spectral indexes of nuclides (UDy/UX), included in UKD, from a rigidity z and-or temperatures of neutron gas Tn are obtained.B Tools of mathematical processing of results are used for activation data and estimation of parameters of a spectrum (F t , f epi , z, Tn, UDy/UX). In the paper are presented some results of researches of neutron spectrum parameters of the nuclear installations (Authors)
The Dynamic Method for Time-of-Flight Measurement of Thermal Neutron Spectra from Pulsed Sources
International Nuclear Information System (INIS)
Pepelyshev, Yu.N.; Tulaev, A.B.; Bobrakov, V.F.
1994-01-01
The time-of-flight method for a measurement of thermal neutron spectra in the pulsed neutron sources with high efficiency of neutron registration, more than 10 5 times higher in comparison with traditional one, is described. The main problems connected with the electric current technique for time-of-flight spectra measurement are examined. The methodical errors, problems of a special neutron detector design and other questions are discussed. Some experimental results, spectra from surfaces of the water and solid methane moderators, obtained in the pulsed reactor IBR-2 (Dubna, Russia) are presented. 4 refs., 5 figs
Detector Simulation: Data Treatment and Analysis Methods
Apostolakis, J
2011-01-01
Detector Simulation in 'Data Treatment and Analysis Methods', part of 'Landolt-Börnstein - Group I Elementary Particles, Nuclei and Atoms: Numerical Data and Functional Relationships in Science and Technology, Volume 21B1: Detectors for Particles and Radiation. Part 1: Principles and Methods'. This document is part of Part 1 'Principles and Methods' of Subvolume B 'Detectors for Particles and Radiation' of Volume 21 'Elementary Particles' of Landolt-Börnstein - Group I 'Elementary Particles, Nuclei and Atoms'. It contains the Section '4.1 Detector Simulation' of Chapter '4 Data Treatment and Analysis Methods' with the content: 4.1 Detector Simulation 4.1.1 Overview of simulation 4.1.1.1 Uses of detector simulation 4.1.2 Stages and types of simulation 4.1.2.1 Tools for event generation and detector simulation 4.1.2.2 Level of simulation and computation time 4.1.2.3 Radiation effects and background studies 4.1.3 Components of detector simulation 4.1.3.1 Geometry modeling 4.1.3.2 External fields 4.1.3.3 Intro...
Isogeometric methods for numerical simulation
Bordas, Stéphane
2015-01-01
The book presents the state of the art in isogeometric modeling and shows how the method has advantaged. First an introduction to geometric modeling with NURBS and T-splines is given followed by the implementation into computer software. The implementation in both the FEM and BEM is discussed.
International Nuclear Information System (INIS)
Rana, Y.S.; Mishra, Abhishek; Singh, Tej
2016-06-01
Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)
International Nuclear Information System (INIS)
Yamaguchi, Toshio; Okada, Isao; Ohtaki, Hitoshi; Mikami, Masuhiro; Kawamura, Kazutaka
1986-01-01
Molecular dynamics simulations have been performed for lithium and rubidium nitrate melts at 550 and 600K, respectively, together with X-ray and neutron diffraction experiments. Simple Coulomb pair potentials with Born-type repulsions have been adopted in the simulations with a rigid body model for the nitrate ion. Structure functions derived from the X-ray and neutron experiments are well reproduced by the simulations, from which the three-dimensional cation distribution around the nitrate ion has been revealed. The self-diffusion coefficients, the velocity autocorrelation functions and the self-exchange velocities of lithium, rubidium and nitrate ions have been calculated. Anisotropic motion of nitrate ions has been found and is discussed on the basis of the structure of the melts. (author)
Simulations of Lithium-Based Neutron Coincidence Counter for Gd-Loaded Fuel
Energy Technology Data Exchange (ETDEWEB)
Cowles, Christian C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kouzes, Richard T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Siciliano, Edward R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2014-10-01
The Department of Energy Office of Nuclear Safeguards and Security (NA-241) is supporting the project Lithium-Based Alternative Neutron Detection Technology Coincidence Counting for Gd-loaded Fuels at Pacific Northwest National Laboratory for the development of a lithium-based neutron coincidence counter for nondestructively assaying Gd loaded nuclear fuel. This report provides results from MCNP simulations of a lithium-based coincidence counter for the possible measurement of Gd-loaded nuclear fuel. A comparison of lithium-based simulations and UNCL-II simulations with and without Gd loaded fuel is provided. A lithium-based model, referred to as PLNS3A-R1, showed strong promise for assaying Gd loaded fuel.
International Nuclear Information System (INIS)
Kim, Jong-Yun; Choi, Yong Suk; Park, Yong Joon; Jung, Sung-Hee
2009-01-01
Neutron spectrometry, based on the scattering of high energy fast neutrons from a radioisotope and slowing-down by the light hydrogen atoms, is a useful technique for non-destructive, quantitative measurement of hydrogen content because it has a large measuring volume, and is not affected by temperature, pressure, pH value and color. The most common choice for radioisotope neutron source is 252 Cf or 241 Am-Be. In this study, 252 Cf with a neutron flux of 6.3x10 6 n/s has been used as an attractive neutron source because of its high flux neutron and weak radioactivity. Pulse-height neutron spectra have been obtained by using in-house built radioisotopic neutron spectrometric system equipped with 3 He detector and multi-channel analyzer, including a neutron shield. As a preliminary study, polyethylene block (density of ∼0.947 g/cc and area of 40 cmx25 cm) was used for the determination of hydrogen content by using multivariate calibration models, depending on the thickness of the block. Compared with the results obtained from a simple linear calibration model, partial least-squares regression (PLSR) method offered a better performance in a quantitative data analysis. It also revealed that the PLSR method in a neutron spectrometric system can be promising in the real-time, online monitoring of the powder process to determine the content of any type of molecules containing hydrogen nuclei.
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
International Nuclear Information System (INIS)
Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan
2015-01-01
Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes
International Nuclear Information System (INIS)
Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen
2005-01-01
Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)
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)
Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method
International Nuclear Information System (INIS)
Dunley, Leonardo Souza
2002-01-01
The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron
Research on GPU-accelerated algorithm in 3D finite difference neutron diffusion calculation method
International Nuclear Information System (INIS)
Xu Qi; Yu Ganglin; Wang Kan; Sun Jialong
2014-01-01
In this paper, the adaptability of the neutron diffusion numerical algorithm on GPUs was studied, and a GPU-accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. The IAEA 3D PWR benchmark problem was calculated in the numerical test. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. (authors)
ATHENA2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum
International Nuclear Information System (INIS)
1999-01-01
1 - Description of program or function: ATHENA 2 D was written to simulate a hypothetical water reflood of a highly-damaged light water reactor (such as the Three-Mile-Island Unit-2 after meltdown, with a packed debris bed near the center of the core), but with insufficiently-borated reflood water. A recriticality transient may result because of the potentially more reactive debris bed. ATHENA-2D solves the transient multigroup neutron diffusion equations in (r,z) geometry. Executing in parallel with the transient neutronics, is a single-phase computational fluid dynamics (CFD) model, driven by multichannel thermal hydraulics based on detailed pin models. Numerous PV-Wave procedure files are included on the distribution media, useful for those who already have PV-Wave from Visual Numerics. These procedures are documented in the 'README' files included on the distribution CD. Some reactor lattice computer code such as WIMS-E, CCC-576/WIMSD4, or CCC-656/WIMSD5B is required for the creation of macroscopic cross section libraries, given pin-cell geometries. WIMS-E is a commercial product available from AEA Technologies, England, WIMS is not included on the ATHENA 2 D distribution CD. Several auxiliary routines are included in the package. TFMAX: Utility that searches through ATHENA 2 D binary output to find the maximum fuel temperature over space and time. POST V EL: Utility that searches through ATHENA 2 D binary output to find maximum scalar and flow field values (over space) and outputs normalization factors as a function of time. These results are used to correctly scale animations. CONVT: If executing ATHENA 2 D on a PC under Windows, this utility converts one form of binary output (directly from ATHENA 2 D) to another, which is readable by PV-Wave for Windows (PV-Wave is data animation and visualization software from Visual Numerics, Inc.) CALC M TX: Post-processing utility for calculating the model coefficients for the calculation matrix. 2 - Methods: Both the