Simple and effective method of determining multiplicity distribution law of neutrons emitted by fissionable material with significant self -multiplication effect

International Nuclear Information System (INIS)

Yanjushkin, V.A.

1991-01-01

At developing new methods of non-destructive determination of plutonium full mass in nuclear materials and products being involved in uranium -plutonium fuel cycle by its intrinsic neutron radiation, it may be useful to know not only separate moments but the multiplicity distribution law itself of neutron leaving this material surface using the following as parameters - firstly, unconditional multiplicity distribution laws of neutrons formed in spontaneous and induced fission acts of the given fissionable material corresponding nuclei and unconditional multiplicity distribution law of neutrons caused by (α,n) reactions at light nuclei of some elements which compose this material chemical structure; -secondly, probability of induced fission of this material nuclei by an incident neutron of any nature formed during the previous fissions or(α,n) reactions. An attempt to develop similar theory has been undertaken. Here the author proposes his approach to this problem. The main advantage of this approach, to our mind, consists in its mathematical simplicity and easy realization at the computer. In principle, the given model guarantees any good accuracy at any real value of induced fission probability without limitations dealing with physico-chemical composition of nuclear material

On the solution of a few problems of multiple scattering by Monte Carlo method

International Nuclear Information System (INIS)

Bluet, J.C.

1966-02-01

Three problems of multiple scattering arising from neutron cross sections experiments, are reported here. The common hypothesis are: - Elastic scattering is the only possible process - Angular distributions are isotropic - Losses of particle energy are negligible in successive collisions. In the three cases practical results, corresponding to actual experiments are given. Moreover the results are shown in more general way, using dimensionless variable such as the ratio of geometrical dimensions to neutron mean free path. The FORTRAN codes are given together with to the corresponding flow charts, and lexicons of symbols. First problem: Measurement of sodium capture cross-section. A sodium sample of given geometry is submitted to a neutron flux. Induced activity is then measured by means of a sodium iodide cristal. The distribution of active nuclei in the sample, and the counter efficiency are calculated by Monte-Carlo method taking multiple scattering into account. Second problem: absolute measurement of a neutron flux using a glass scintillator. The scintillator is a use of lithium 6 loaded glass, submitted to neutron flux perpendicular to its plane faces. If the glass thickness is not negligible compared with scattering mean free path λ, the mean path e' of neutrons in the glass is different from the thickness. Monte-Carlo calculation are made to compute this path and a relative correction to efficiency equal to (e' - e)/e. Third problem: study of a neutron collimator. A neutron detector is placed at the bottom of a cylinder surrounded with water. A neutron source is placed on the cylinder axis, in front of the water shield. The number of neutron tracks going directly and indirectly through the water from the source to the detector are counted. (author) [fr

Inventory verification measurements using neutron multiplicity counting

International Nuclear Information System (INIS)

Ensslin, N.; Foster, L.A.; Harker, W.C.; Krick, M.S.; Langner, D.G.

1998-01-01

This paper describes a series of neutron multiplicity measurements of large plutonium samples at the Los Alamos Plutonium Facility. The measurements were corrected for bias caused by neutron energy spectrum shifts and nonuniform multiplication, and are compared with calorimetry/isotopics. The results show that multiplicity counting can increase measurement throughput and yield good verification results for some inventory categories. The authors provide recommendations on the future application of the technique to inventory verification

FB-line neutron multiplicity counter operation manual

International Nuclear Information System (INIS)

Langner, D.G.; Sweet, M.R.; Salazar, S.D.; Kroncke, K.E.

1998-01-01

This manual describes the design features, performance, and operating characteristics for the FB-Line Neutron Multiplicity Counter (FBLNMC). The FBLNMC counts neutron multiplicities to quantitatively assay plutonium in many forms, including impure scrap and waste. Monte Carlo neutronic calculations were used to design the high-efficiency (57%) detector that has 113 3 H tubes in a high-density polyethylene body. The new derandomizer circuit is included in the design to reduce deadtime. The FBLNMC can be applied to plutonium masses in the range from a few tens of grams to 5 kg; both conventional coincidence counting and multiplicity counting can be used as appropriate. This manual gives the performance data and preliminary calibration parameters for the FBLNMC

Matrix-type multiple reciprocity boundary element method for solving three-dimensional two-group neutron diffusion equations

International Nuclear Information System (INIS)

Itagaki, Masafumi; Sahashi, Naoki.

1997-01-01

The multiple reciprocity boundary element method has been applied to three-dimensional two-group neutron diffusion problems. A matrix-type boundary integral equation has been derived to solve the first and the second group neutron diffusion equations simultaneously. The matrix-type fundamental solutions used here satisfy the equation which has a point source term and is adjoint to the neutron diffusion equations. A multiple reciprocity method has been employed to transform the matrix-type domain integral related to the fission source into an equivalent boundary one. The higher order fundamental solutions required for this formulation are composed of a series of two types of analytic functions. The eigenvalue itself is also calculated using only boundary integrals. Three-dimensional test calculations indicate that the present method provides stable and accurate solutions for criticality problems. (author)

Neutron multiplication in lead in the experiments with neutron generators

International Nuclear Information System (INIS)

Markovskij, D.V.

1989-01-01

A calculational analysis of neutron multiplication in lead, including the estimates of multiplication limits for the standard ENDF/BIV data set and the effects of various changes in the data themselves is performed. 10 refs, 5 figs

Estimation of subcriticality by neutron source multiplication method

International Nuclear Information System (INIS)

Sakurai, Kiyoshi; Suzaki, Takenori; Arakawa, Takuya; Naito, Yoshitaka

1995-03-01

Subcritical cores were constructed in a core tank of the TCA by arraying 2.6% enriched UO 2 fuel rods into nxn square lattices of 1.956 cm pitch. Vertical distributions of the neutron count rates for the fifteen subcritical cores (n=17, 16, 14, 11, 8) with different water levels were measured at 5 cm interval with 235 U micro-fission counters at the in-core and out-core positions arranging a 252 C f neutron source at near core center. The continuous energy Monte Carlo code MCNP-4A was used for the calculation of neutron multiplication factors and neutron count rates. In this study, important conclusions are as follows: (1) Differences of neutron multiplication factors resulted from exponential experiment and MCNP-4A are below 1% in most cases. (2) Standard deviations of neutron count rates calculated from MCNP-4A with 500000 histories are 5-8%. The calculated neutron count rates are consistent with the measured one. (author)

The relationship between neutron multiplication and keff

International Nuclear Information System (INIS)

Brewer, R.W.

1995-01-01

In recent years the International Criticality Safety Benchmark Evaluation Project under the sponsorship of the Department of Energy has undertaken the task of evaluating past critical experiments. Many of the experiments involving metals were subcritical with extrapolation to some critical characteristic dimension. The metal experiments were commonly limited to a maximum multiplication of 100 for obvious safety considerations. Also many critical experiments often used subcritical measurements to obtain the critical specifications, e.g. Jezebel used subcritical measurements to assess the magnitude of neutron reflection from the surrounding structures. Therefore, the task of evaluating the experimentally derived critical configuration often involves evaluating the subcritical measurements made by the experimentalist. The purpose of past experiments was to determine critical configurations. Many of the modem computer codes (KENO, MCNP, and ONEDANT) calculate values of k eff . However, the subcritical measurements made during the course of the experiment are usually measurements of the neutron multiplication. To evaluate the subcritical experiments, a link was established between the neutronic theory and the practical application of such when using subcritical measurements to establish the critical characteristic dimension. A more in depth derivation of the relationship between k eff and neutron multiplication will be shown along with comparisons between calculated and measured multiplications

Active neutron multiplicity analysis and Monte Carlo calculations

International Nuclear Information System (INIS)

Krick, M.S.; Ensslin, N.; Langner, D.G.; Miller, M.C.; Siebelist, R.; Stewart, J.E.; Ceo, R.N.; May, P.K.; Collins, L.L. Jr

1994-01-01

Active neutron multiplicity measurements of high-enrichment uranium metal and oxide samples have been made at Los Alamos and Y-12. The data from the measurements of standards at Los Alamos were analyzed to obtain values for neutron multiplication and source-sample coupling. These results are compared to equivalent results obtained from Monte Carlo calculations. An approximate relationship between coupling and multiplication is derived and used to correct doubles rates for multiplication and coupling. The utility of singles counting for uranium samples is also examined

Solution of the multigroup neutron diffusion Eigenvalue problem in slab geometry by modified power method

Energy Technology Data Exchange (ETDEWEB)

Zanette, Rodrigo [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pós-Graduação em Matemática Aplicada; Petersen, Claudio Z.; Tavares, Matheus G., E-mail: rodrigozanette@hotmail.com, E-mail: claudiopetersen@yahoo.com.br, E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Programa de Pós-Graduação em Modelagem Matemática

2017-07-01

We describe in this work the application of the modified power method for solve the multigroup neutron diffusion eigenvalue problem in slab geometry considering two-dimensions for nuclear reactor global calculations. It is well known that criticality calculations can often be best approached by solving eigenvalue problems. The criticality in nuclear reactors physics plays a relevant role since establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve the eigenvalue problem, a modified power method is used to obtain the dominant eigenvalue (effective multiplication factor (K{sub eff})) and its corresponding eigenfunction (scalar neutron flux), which is non-negative in every domain, that is, physically relevant. The innovation of this work is solving the neutron diffusion equation in analytical form for each new iteration of the power method. For solve this problem we propose to apply the Finite Fourier Sine Transform on one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. The inverse Fourier transform is used to reconstruct the solution for the original problem. It is known that the power method is an iterative source method in which is updated by the neutron flux expression of previous iteration. Thus, for each new iteration, the neutron flux expression becomes larger and more complex due to analytical solution what makes propose that it be reconstructed through an polynomial interpolation. The methodology is implemented to solve a homogeneous problem and the results are compared with works presents in the literature. (author)

On the solution of a few problems of multiple scattering by Monte Carlo method; Sur la solution de quelques problemes de diffusions multiples par la methode de Monte-Carlo

Energy Technology Data Exchange (ETDEWEB)

Bluet, J C [Commissariat a l' Energie Atomique, Cadarache (France)

1966-02-01

Three problems of multiple scattering arising from neutron cross sections experiments, are reported here. The common hypothesis are: - Elastic scattering is the only possible process - Angular distributions are isotropic - Losses of particle energy are negligible in successive collisions. In the three cases practical results, corresponding to actual experiments are given. Moreover the results are shown in more general way, using dimensionless variable such as the ratio of geometrical dimensions to neutron mean free path. The FORTRAN codes are given together with to the corresponding flow charts, and lexicons of symbols. First problem: Measurement of sodium capture cross-section. A sodium sample of given geometry is submitted to a neutron flux. Induced activity is then measured by means of a sodium iodide cristal. The distribution of active nuclei in the sample, and the counter efficiency are calculated by Monte-Carlo method taking multiple scattering into account. Second problem: absolute measurement of a neutron flux using a glass scintillator. The scintillator is a use of lithium 6 loaded glass, submitted to neutron flux perpendicular to its plane faces. If the glass thickness is not negligible compared with scattering mean free path {lambda}, the mean path e' of neutrons in the glass is different from the thickness. Monte-Carlo calculation are made to compute this path and a relative correction to efficiency equal to (e' - e)/e. Third problem: study of a neutron collimator. A neutron detector is placed at the bottom of a cylinder surrounded with water. A neutron source is placed on the cylinder axis, in front of the water shield. The number of neutron tracks going directly and indirectly through the water from the source to the detector are counted. (author) [French] On traite dans ce rapport de trois problemes avec les hypotheses communes suivantes: 1.- Le seul processus de collision possible est la diffusion electrique. 2.- La distribution angulaire est

Direct integration multiple collision integral transport analysis method for high energy fusion neutronics

International Nuclear Information System (INIS)

Koch, K.R.

1985-01-01

A new analysis method specially suited for the inherent difficulties of fusion neutronics was developed to provide detailed studies of the fusion neutron transport physics. These studies should provide a better understanding of the limitations and accuracies of typical fusion neutronics calculations. The new analysis method is based on the direct integration of the integral form of the neutron transport equation and employs a continuous energy formulation with the exact treatment of the energy angle kinematics of the scattering process. In addition, the overall solution is analyzed in terms of uncollided, once-collided, and multi-collided solution components based on a multiple collision treatment. Furthermore, the numerical evaluations of integrals use quadrature schemes that are based on the actual dependencies exhibited in the integrands. The new DITRAN computer code was developed on the Cyber 205 vector supercomputer to implement this direct integration multiple-collision fusion neutronics analysis. Three representative fusion reactor models were devised and the solutions to these problems were studied to provide suitable choices for the numerical quadrature orders as well as the discretized solution grid and to understand the limitations of the new analysis method. As further verification and as a first step in assessing the accuracy of existing fusion-neutronics calculations, solutions obtained using the new analysis method were compared to typical multigroup discrete ordinates calculations

A set-up for measuring neutron cross sections and radiation multiplicity from neutron-nucleus interaction

International Nuclear Information System (INIS)

Georgiev, G.P.; Ermakov, V.A.; Grigor'ev, Yu.V.

1988-01-01

A multiplicity detector of the ''Romashka'' type has been used on the 500 m flight part of the IBR-30 pulsed reactor. The detector consists of 16 independent sections with NaJ(Tl) crystals with a total volume of 36 liters. The geometric efficiency of single-ray detection is ∼ 80%. The gamma-ray to neutron detection efficiency ratio is ≥600 for neutrons with energies below 200 keV. This detector allows one to perform neutron capture and fission cross section measurements and to study gamma-ray multiplicity and resonance selfabsorption effects in the 20 eV-200keV neutron energy range

A linear multiple balance method for discrete ordinates neutron transport equations

International Nuclear Information System (INIS)

Park, Chang Je; Cho, Nam Zin

2000-01-01

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

MCRTOF, Multiple Scattering of Resonance Region Neutron in Time of Flight Experiments

International Nuclear Information System (INIS)

Ohkubo, Mako

1984-01-01

1 - Description of program or function: Multiple scattering of neutrons in the resonance energy region impinging on a disk with an arbitrary angle. 2 - Method of solution: The Monte Carlo method is employed to simulate the path of an incident neutron in a medium for which macroscopic cross sections are determined by resonance parameters. By tracing a large number of neutrons, probabilities for capture, transmission, front-face scattering, rear-face scattering and side-face scattering are determined and printed out as function of incident neutron energy. Optionally, the distribution of capture locations in the disk can be printed. The incident neutron energy is swept to fit a situation as encountered in time-of-flight experiments. 3 - Restrictions on the complexity of the problem: The cross section file is constructed from input resonance parameters with a single- level Breit-Wigner formula. The following restrictions and simplifications apply: - The maximum number of resonances is five. - Reactions other than capture and scattering are neglected. - The angular scattering distribution in the center-of-mass system is assumed to be uniform. - Chemical binding effects are neglected

NEUTRON SPECTRUM MEASUREMENTS USING MULTIPLE THRESHOLD DETECTORS

Energy Technology Data Exchange (ETDEWEB)

Gerken, William W.; Duffey, Dick

1963-11-15

From American Nuclear Society Meeting, New York, Nov. 1963. The use of threshold detectors, which simultaneously undergo reactions with thermal neutrons and two or more fast neutron threshold reactions, was applied to measurements of the neutron spectrum in a reactor. A number of different materials were irradiated to determine the most practical ones for use as multiple threshold detectors. These results, as well as counting techniques and corrections, are presented. Some materials used include aluminum, alloys of Al -Ni, aluminum-- nickel oxides, and magesium orthophosphates. (auth)

Solving the uncommon reactor core neutronics problems

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1980-01-01

The common reactor core neutronics problems have fundamental neutron space, energy spectrum solutions. Typically the most positive eigenvalue is associated with an all-positive flux for the pseudo-steady-state condition (k/sub eff/), or the critical state is to be effected by selective adjustment of some variable such as the fuel concentration. With sophistication in reactor analysis has come the demand for solutions of other, uncommon neutronics problems. Importance functionss are needed for sensitivity and uncertainty analyses, as for ratios of intergral reaction rates such as the fuel conversion (breeding) ratio. The dominant higher harmonic solution is needed in stability analysis. Typically the desired neutronics solution must contain negative values to qualify as a higher harmonic or to satisfy a fixed source containing negative values. Both regular and adjoint solutions are of interest as are special integrals of the solutions to support analysis

Neutron multiplicity of fission fragments

Energy Technology Data Exchange (ETDEWEB)

Abdelrahman, Y S [Physics department, mu` rah university Al-Karak, (Jordan)

1995-10-01

The total average neutron multiplicity of the fission fragments produced by the spontaneous fission of {sup 248} Cm has been measured. This measurement has been done by using a new experimental technique. This technique mainly depends on {gamma}-{gamma} coincidence using a very high resolution high purity germanium (HPGe) detector. 2 figs.

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

Science.gov (United States)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-07-01

An extension of the point kinetics model is developed to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. The spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

Effects of (α,n) contaminants and sample multiplication on statistical neutron correlation measurements

International Nuclear Information System (INIS)

Dowdy, E.J.; Hansen, G.E.; Robba, A.A.; Pratt, J.C.

1980-01-01

The complete formalism for the use of statistical neutron fluctuation measurements for the nondestructive assay of fissionable materials has been developed. This formalism includes the effect of detector deadtime, neutron multiplicity, random neutron pulse contributions from (α,n) contaminants in the sample, and the sample multiplication of both fission-related and background neutrons

The isotope density inverse problem in multigroup neutron transport

International Nuclear Information System (INIS)

Zazula, J.M.

1981-01-01

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

Fundamental Problems of Neutron Physics at the Spallation Neutron Source at the ORNL

International Nuclear Information System (INIS)

Gudkov, Vladimir

2008-01-01

We propose to provide theoretical support for the experimental program in fundamental neutron physics at the SNS. This includes the study of neutron properties, neutron beta-decay, parity violation effects and time reversal violation effects. The main purpose of the proposed research is to work on theoretical problems related to experiments which have a high priority at the SNS. Therefore, we will make a complete analysis of beta-decay process including calculations of radiative corrections and recoil corrections for angular correlations for polarized neutron decay, with an accuracy better that is supposed to be achieved in the planning experiments. Based on the results of the calculations, we will provide analysis of sensitivity of angular correlations to be able to search for the possible extensions of the Standard model. Also we will help to plan other experiments to address significant problems of modern physics and will work on their theoretical support.

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

Energy Technology Data Exchange (ETDEWEB)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; Folsom, Micah; Kouzes, Richard; Kukharev, Vladislav; Lintereur, Azaree; Robinson, Sean; Siciliano, Edward; Stave, Sean; Valdez, Patrick

2018-04-01

Past 3He shortages led to investigations into replacement options for neutron detectors in systems that previously used 3He-based technologies. The goal of this research was to investigate the feasibility of a full-scale lithium fluoride with silver activated zinc sulfide (LiF/ZnS) based neutron multiplicity counter. The LiF/ZnS based neutron multiplicity counter (LiNMC) was developed based on an iterative process between modeling and experimental measurements. Each active region of the LiNMC contains five sheets of LiF/ZnS sandwiched between six sheets of wavelength shifting plastic to form neutron detection stacks. The wavelength shifted scintillation light was collected by photomultiplier tubes located on each end of the stacks. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high density polyethylene blocks in the corners to reflect high energy neutrons and capture low energy neutrons. Preliminary calibration with a 252Cf neutron source showed that the LiNMC was able to achieve 36% neutron detection efficiency (ε) and an 11.7 μs neutron die-away time (τ) for a doubles Figure-of-merit (ε2/ τ) of 109. This is the highest doubles Figure-of-merit performance measured to-date for a 3He-free neutron multiplicity counter system. By the end of this project, the LiNMC’s basic components were integrated into a single laboratory scale system capable of proof-of-concept measurements.

Development of a lithium fluoride zinc sulfide based neutron multiplicity counter

Science.gov (United States)

Cowles, Christian; Behling, Spencer; Baldez, Phoenix; Folsom, Micah; Kouzes, Richard; Kukharev, Vladislav; Lintereur, Azaree; Robinson, Sean; Siciliano, Edward; Stave, Sean; Valdez, Patrick

2018-04-01

The feasibility of a full-scale lithium fluoride zinc sulfide (LiF/ZnS) based neutron multiplicity counter has been demonstrated. The counter was constructed of modular neutron detecting stacks that each contain five sheets of LiF/ZnS interleaved between six sheets of wavelength shifting plastic with a photomultiplier tube on each end. Twelve such detector stacks were placed around a sample chamber in a square arrangement with lithiated high-density polyethylene blocks in the corners to reflect high-energy neutrons and capture low-energy neutrons. The final system design was optimized via modeling and small-scale test. Measuring neutrons from a 252Cf source, the counter achieved a 36% neutron detection efficiency (ɛ) and an 11 . 7 μs neutron die-away time (τ) for a doubles figure-of-merit (ɛ2 / τ) of 109. This is the highest doubles figure-of-merit measured to-date for a 3He-free neutron multiplicity counter.

Definition of neutron multiplication in a reception capacity of radioactive waste shop

International Nuclear Information System (INIS)

Dulin, V.A.; Dulin, V.V.; Pavlova, O.N.

2006-01-01

To determine neutrons multiplication the measurements and calculations of spatial distributions of neutron counting and absolute fission rates in a reception capacity of IPPE radioactive waste shop have been carried out and analyzed. A content of fissionable medium was unknown. The approach developed has allowed implementing a calculation analysis of the experimental data on determination of the most probable spatial distributions of basic parameters of the fissionable medium of unknown content. It has allowed determining the neutrons multiplication factor in a reception capacity of a tank No. 17. It has been found that the value of neutrons multiplication factor in a tank is 1.07 ± 0.03. The developed measurement method and calculation analysis used for experimental data also can be applied in other cases when the multiplication medium content is unknown [ru

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

International Nuclear Information System (INIS)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-01-01

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If the detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.

The effect of albedo neutrons on the neutron multiplication of small plutonium oxide samples in a PNCC chamber

CERN Document Server

Bourva, L C A; Weaver, D R

2002-01-01

This paper describes how to evaluate the effect of neutrons reflected from parts of a passive neutron coincidence chamber on the neutron leakage self-multiplication, M sub L , of a fissile sample. It is shown that albedo neutrons contribute, in the case of small plutonium bearing samples, to a significant part of M sub L , and that their effect has to be taken into account in the relationship between the measured coincidence count rates and the sup 2 sup 4 sup 0 Pu effective mass of the sample. A simple one-interaction model has been used to write the balance of neutron gains and losses in the material when exposed to the re-entrant neutron flux. The energy and intensity profiles of the re-entrant flux have been parameterised using Monte Carlo MCNP sup T sup M calculations. This technique has been implemented for the On Site Laboratory neutron/gamma counter within the existing MEPL 1.0 code for the determination of the neutron leakage self-multiplication. Benchmark tests of the resulting MEPL 2.0 code with MC...

Non-destructive isotopic uranium assay by multiple delayed neutron measurements

International Nuclear Information System (INIS)

Papadopoulos, N.N.; Tsagas, N.F.

1991-01-01

The high accuracy and precision required in nuclear safeguards measurements can be achieved by an improved neutron activation technique based on multiple delayed fission neutron counting under various experimental conditions. For the necessary ultrahigh counting statistics required, cyclic activation of multiple subsamples has been applied. The home-made automated flexible analytical system with neutron flux and spectrum differentiation by irradiation position adjustment and cadmium screening, permits the non-destructive determination of the U235 abundance and the total U element concentration needed in nuclear safeguards sample analysis, with a high throughout and a low operational cost. Careful experimental optimization led to considerable improvement of the results

Measurement of neutron multiplication in Pb by Mn foils

International Nuclear Information System (INIS)

Chen Yuan; Liu Rong; Guo Haiping; Jiang Wenmian; Shen Jian

1994-01-01

The Leakage neutron multiplication in bulk lead has been measured using the total absorption detector and relative method. The polyethylene sphere of 138 cm in diameter is used as the moderator and total absorption detector. The measured results from 55 Mn foils and 6 Li glass are compared. The neutron multiplication is 1.74 with the lead shell of 23.1 cm thick. The measured result is consistent with the calculated one with ANISN code and ENDF/B-6 evaluated data within the experimental error. (4 figs., 3 tabs.)

Expected precision of neutron multiplicity measurements of waste drums

International Nuclear Information System (INIS)

Ensslin, N.; Krick, M.S.; Menlove, H.O.

1995-01-01

DOE facilities are beginning to apply passive neutron multiplicity counting techniques to the assay of plutonium scrap and residues. There is also considerable interest in applying this new measurement technique to 208-liter waste drums. The additional information available from multiplicity counting could flag the presence of shielding materials or improve assay accuracy by correcting for matrix effects such as (α,n) induced fission or detector efficiency variations. The potential for multiplicity analysis of waste drums, and the importance of better detector design, can be estimated by calculating the expected assay precision using a Figure of Merit code for assay variance. This paper reports results obtained as a function of waste drum content and detector characteristics. We find that multiplicity analysis of waste drums is feasible if a high-efficiency neutron counter is used. However, results are significantly poorer if the multiplicity analysis must be used to solve for detection efficiency

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

Energy Technology Data Exchange (ETDEWEB)

Costa da Silva, Adilson; Carvalho da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil); Senra Martinez, Aquilino, E-mail: aquilino@lmp.ufrj.br [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil)

2011-07-15

Highlights: > We proposed a new neutron diffusion hybrid equation with external neutron source. > A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. > 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

International Nuclear Information System (INIS)

Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

2011-01-01

Highlights: → We proposed a new neutron diffusion hybrid equation with external neutron source. → A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. → 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

Energy dependence of the neutron multiplicity P/sub nu/ in fast neutron induced fission of /sup 235,238/U and 239Pu

International Nuclear Information System (INIS)

Zucker, M.S.; Holden, N.E.

1986-01-01

Certain applications require knowledge of the higher moments of the neutron multiplicity probability. It can be shown that the second factorial moment is proportional to the fission rate in the sample, and that the third factorial moment can be of use in disentangling spontaneous fission from induced fission. Using a source of unpublished work in which neutron multiplicities were derived for the fast neutron induced fission of U-235, U-238, and Pu-239, the multiplicity probability has been calculated as a function of neutron energy for the energy range 0 to 10 MeV

Solving the uncommon nuclear reactor core neutronics problems

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1983-01-01

Calculational procedures have been implemented for solving importance and higher harmonic neutronics problems. Solutions are obtained routinely to support analysis of reactor core performance, treating up to three space coordinates with the multigroup diffusion theory approximation to neutron transport. The techniques used and some of the calculational difficulties are discussed

Determination of shell correction energies at saddle point using pre-scission neutron multiplicities

International Nuclear Information System (INIS)

Golda, K.S.; Saxena, A.; Mittal, V.K.; Mahata, K.; Sugathan, P.; Jhingan, A.; Singh, V.; Sandal, R.; Goyal, S.; Gehlot, J.; Dhal, A.; Behera, B.R.; Bhowmik, R.K.; Kailas, S.

2013-01-01

Pre-scission neutron multiplicities have been measured for 12 C + 194, 198 Pt systems at matching excitation energies at near Coulomb barrier region. Statistical model analysis with a modified fission barrier and level density prescription have been carried out to fit the measured pre-scission neutron multiplicities and the available evaporation residue and fission cross sections simultaneously to constrain statistical model parameters. Simultaneous fitting of the pre-scission neutron multiplicities and cross section data requires shell correction at the saddle point

Neutron multiplicity for neutron induced fission of 235U, 238U, and 239Pu as a function of neutron energy

International Nuclear Information System (INIS)

Zucker, M.S.; Holden, N.E.

1986-01-01

Recent development in the theory and practice of neutron correlation (''coincidence'') counting require knowledge of the higher factorial moments of the P/sub ν/ distribution (the probability that (ν) neutrons are emitted in a fission) for the case where the fission is induced by bombarding neutrons of more than thermal energies. In contrast to the situation with spontaneous and thermal neutron induced fission, where with a few exceptions the P/sub ν/ is reasonably well known, in the fast neutron energy region, almost no information is available concerning the multiplicity beyond the average value, [ν], even for the most important nuclides. The reason for this is the difficulty of such experiments, with consequent statistically poor and physically inconsistent results

Seminar Neutronika-2012. Neutron-physical problems of nuclear-power engineering. Program and abstracts

International Nuclear Information System (INIS)

2012-01-01

On October, 30 - November, 2 in State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky a seminar Neutron-physical problems of nuclear power engineering - Neutronika-2012 took place. On the seminar the following problems were discussed: justification of neutron-physical characteristics of reactor facilities and innovation projects; constant support of neutron-physical calculations of nuclear power installations; numerical simulation during solving reactor physics problems; simulation of neutron-physical processes in reactor facilities by Monte Carlo method; development and verification of programs for reactor facilities neutron-physical calculations; algorithms and programs for solving nonstationary problems of neutron-physical calculation of nuclear reactors; analysis of integral and reactor experiments, experimental database; justification of nuclear and radiation safety of fuel cycle [ru

Reduction of bias in neutron multiplicity assay using a weighted point model

Energy Technology Data Exchange (ETDEWEB)

Geist, W. H. (William H.); Krick, M. S. (Merlyn S.); Mayo, D. R. (Douglas R.)

2004-01-01

Accurate assay of most common plutonium samples was the development goal for the nondestructive assay technique of neutron multiplicity counting. Over the past 20 years the technique has been proven for relatively pure oxides and small metal items. Unfortunately, the technique results in large biases when assaying large metal items. Limiting assumptions, such as unifoh multiplication, in the point model used to derive the multiplicity equations causes these biases for large dense items. A weighted point model has been developed to overcome some of the limitations in the standard point model. Weighting factors are detemiined from Monte Carlo calculations using the MCNPX code. Monte Carlo calculations give the dependence of the weighting factors on sample mass and geometry, and simulated assays using Monte Carlo give the theoretical accuracy of the weighted-point-model assay. Measured multiplicity data evaluated with both the standard and weighted point models are compared to reference values to give the experimental accuracy of the assay. Initial results show significant promise for the weighted point model in reducing or eliminating biases in the neutron multiplicity assay of metal items. The negative biases observed in the assay of plutonium metal samples are caused by variations in the neutron multiplication for neutrons originating in various locations in the sample. The bias depends on the mass and shape of the sample and depends on the amount and energy distribution of the ({alpha},n) neutrons in the sample. When the standard point model is used, this variable-multiplication bias overestimates the multiplication and alpha values of the sample, and underestimates the plutonium mass. The weighted point model potentially can provide assay accuracy of {approx}2% (1 {sigma}) for cylindrical plutonium metal samples < 4 kg with {alpha} < 1 without knowing the exact shape of the samples, provided that the ({alpha},n) source is uniformly distributed throughout the

Using anisotropies in prompt fission neutron coincidences to assess the neutron multiplication of highly multiplying subcritical plutonium assemblies

Energy Technology Data Exchange (ETDEWEB)

Mueller, J.M., E-mail: jonathan_mueller@ncsu.edu; Mattingly, J.

2016-07-21

There is a significant and well-known anisotropy between the prompt neutrons emitted from a single fission event; these neutrons are most likely to be observed at angles near 0° or 180° relative to each other. However, the propagation of this anisotropy through different generations of a fission chain reaction has not been previously studied. We have measured this anisotropy in neutron–neutron coincidences from a subcritical highly-multiplying assembly of plutonium metal. The assembly was a 4.5 kg α-phase plutonium metal sphere composed of 94% {sup 239}Pu and 6% {sup 240}Pu by mass. Data were collected using two EJ-309 liquid scintillators and two EJ-299 plastic scintillators. The angular distribution of neutron–neutron coincidences was measured at 90° and 180° and found to be largely isotropic. Simulations were performed using MCNPX-PoliMi of similar plutonium metal spheres of varying sizes and a correlation between the neutron multiplication of the assembly and the anisotropy of neutron–neutron coincidences was observed. In principle, this correlation could be used to assess the neutron multiplication of an unknown assembly.

Reducing neutron multiplicity counting bias for plutonium warhead authentication

Energy Technology Data Exchange (ETDEWEB)

Goettsche, Malte

2015-06-05

Confidence in future nuclear arms control agreements could be enhanced by direct verification of warheads. It would include warhead authentication. This is the assessment based on measurements whether a declaration that a specific item is a nuclear warhead is true. An information barrier can be used to protect sensitive information during measurements. It could for example show whether attributes such as a fissile mass exceeding a threshold are met without indicating detailed measurement results. Neutron multiplicity measurements would be able to assess a plutonium fissile mass attribute if it were possible to show that their bias is low. Plutonium measurements have been conducted with the He-3 based Passive Scrap Multiplicity Counter. The measurement data has been used as a reference to test the capacity of the Monte Carlo code MCNPX-PoliMi to simulate neutron multiplicity measurements. The simulation results with their uncertainties are in agreement with the experimental results. It is essential to use cross-sections which include neutron scattering with the detector's polyethylene molecular structure. Further MCNPX-PoliMi simulations have been conducted in order to study bias that occurs when measuring samples with large plutonium masses such as warheads. Simulation results of solid and hollow metal spheres up to 6000 g show that the masses are underpredicted by as much as 20%. The main source of this bias has been identified in the false assumption that the neutron multiplication does not depend on the position where a spontaneous fission event occurred. The multiplication refers to the total number of neutrons leaking a sample after a primary spontaneous fission event, taking induced fission into consideration. The correction of the analysis has been derived and implemented in a MATLAB code. It depends on four geometry-dependent correction coefficients. When the sample configuration is fully known, these can be exactly determined and remove this type of

Applications of neutrons for laboratory and industrial activation analysis problems

International Nuclear Information System (INIS)

Szabo, Elek; Bakos, Laszlo

1986-01-01

This chapter presents some particular applications and case studies of neutrons in activation analysis for research and industrial development purposes. The reactor neutrons have been applied in Hungarian laboratories for semiconductor research, for analysis of geological (lunar) samples, and for a special comparator measurement of samples. Some industrial applications of neutron generator and sealed sources for analytical problems are presented. Finally, prompt neutron activation analysis is outlined briefly. (R.P.)

Measurements of 14 MeV neutron multiplication in spherical beryllium shells

International Nuclear Information System (INIS)

Moellendorff, U. von; Alevra, A.V.; Giese, H.; Kappler, F.; Klein, H.; Klein, H.; Tayama, R.

1995-01-01

New results of spherical-shell transmission measurements with 14MeV neutrons on pure beryllium shells up to 17cm thick are reported. The total leakage neutron multiplications were measured using a Bonner sphere system. Independently, the leakage neutron spectra were measured over the entire energy range, 15MeV to thermal energies, by proton-recoil and time-of-flight methods. The total leakage multiplications are in excellent agreement with three-dimensional Monte Carlo calculations using beryllium nuclear data based on the Young and Stewart evaluation. The leakage in the evaporation energy window confirms the Be(n,2n) cross-section of the Young and Stewart evaluation rather than that used in the ENDF/B-VI library. At energies below 1keV, a surplus of leakage neutrons over the calculation is found for smaller beryllium thicknesses. (orig.)

Non-destructive assay of fissile materials by detection and multiplicity analysis of spontaneous neutrons

International Nuclear Information System (INIS)

Prosdocimi, A.

1979-01-01

A method for determining the absolute reaction rate of nuclear events giving rise to neutron emission, according to their neutron multiplicity, is proposed. A typical application is the measurement of the (α, n) and spontaneous fission rates in a fissile material sample, particularly of Pu oxide composition. An analysis of random and correlated neutron pulses is carried out on the basis of sequential order without requiring any time interval analysis, then the primary nuclear events are sorted versus their neutron multiplicity. Suitable theoretical relationships enable to derive the absolute (α, n) and SF reaction rates when the physical parameters of the neutron detector and the multiplicity spectrumm of pulses are known. A typical device is described and the results of experiments leading to Pu-239 and Pu-240 assay are given

Variable dead time counters. 1 - theoretical responses and the effects of neutron multiplication

International Nuclear Information System (INIS)

Lees, E.W.; Hooton, B.W.

1978-10-01

A theoretical expression is derived for calculating the response of any variable dead time counter (VDC) used in the passive assay of plutonium by neutron counting of the natural spontaneous fission activity. The effects of neutron multiplication in the sample arising from interactions of the original spontaneous fission neutrons is shown to modify the linear relationship between VDC signal and Pu mass. Numerical examples are shown for the Euratom VDC and a systematic investigation of the various factors affecting neutron multiplication is reported. Limited comparisons between the calculations and experimental data indicate provisional validity of the calculations. (author)

Multiplicity and energy of neutrons from {sup 233}U(n{sub th},f) fission fragments

Energy Technology Data Exchange (ETDEWEB)

Nishio, Katsuhisa; Kimura, Itsuro; Nakagome, Yoshihiro [Kyoto Univ. (Japan)

1998-03-01

The correlation between fission fragments and prompt neutrons from the reaction {sup 233}U(n{sub th},f) was measured with improved accuracy. The results determined the neutron multiplicity and emission energy as a function of fragment mass and total kinetic energy. The average energy as a function of fragment mass followed a nearly symmetric distribution centered about the equal mass-split and formed a remarkable contrast with the saw-tooth distribution of the average neutron multiplicity. The neutron multiplicity from the specified fragment decreases linearly with total kinetic energy, and the slope of multiplicity with kinetic energy had the minimum value at about 130 u. The level density parameter versus mass determined from the neutron data showed a saw-tooth structure with the pronounced minimum at about 128 and generally followed the formula by Gilbert and Cameron, suggesting that the neutron emission process was very much affected by the shell-effect of the fission fragment. (author)

Study of gamma ray multiplicity spectra for radiative capture of neutrons in 113,115In

International Nuclear Information System (INIS)

Georgiev, G.P.; Fajkov-Stanchik, Kh.; Grigor'ev, Yu.V.; Muradyan, G.V.; Yaneva, N.B.

1997-08-01

Neutron radiative capture measurements were performed for the enriched isotopes 113 In and 115 In on the neutron spectrometer at the Neutron Physics Laboratory of the Joint Institute for Nuclear Research employing the gamma ray multiplicity technique and using a ''Romashka'' multi-sectional 4p detector on the 500 m time base of the IBR-30 booster. The gamma multiplicity spectra of resolved resonances were obtained for the 20-500 eV energy range. The mean gamma ray multiplicity was determined for each resonance. The dependence of the ratio S of the low-energy coincidence multiplicity spectrum to the high-energy coincidence multiplicity spectrum on resonance energy exhibits a non-statistical structure. This structure was found to correlate with the local neutron strength function. (author). 10 refs, 6 figs, 2 tabs

Fast solution of neutron diffusion problem by reduced basis finite element method

International Nuclear Information System (INIS)

Chunyu, Zhang; Gong, Chen

2018-01-01

Highlights: •An extremely efficient method is proposed to solve the neutron diffusion equation with varying the cross sections. •Three orders of speedup is achieved for IAEA benchmark problems. •The method may open a new possibility of efficient high-fidelity modeling of large scale problems in nuclear engineering. -- Abstract: For the important applications which need carry out many times of neutron diffusion calculations such as the fuel depletion analysis and the neutronics-thermohydraulics coupling analysis, fast and accurate solutions of the neutron diffusion equation are demanding but necessary. In the present work, the certified reduced basis finite element method is proposed and implemented to solve the generalized eigenvalue problems of neutron diffusion with variable cross sections. The order reduced model is built upon high-fidelity finite element approximations during the offline stage. During the online stage, both the k eff and the spatical distribution of neutron flux can be obtained very efficiently for any given set of cross sections. Numerical tests show that a speedup of around 1100 is achieved for the IAEA two-dimensional PWR benchmark problem and a speedup of around 3400 is achieved for the three-dimensional counterpart with the fission cross-sections, the absorption cross-sections and the scattering cross-sections treated as parameters.

The MARVEL assembly for neutron multiplication

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester; Mathew T. Kinlaw

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (keff=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication.

The MARVEL assembly for neutron multiplication.

Science.gov (United States)

Chichester, David L; Kinlaw, Mathew T

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (k(eff)=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication. Copyright © 2013 Elsevier Ltd. All rights reserved.

Calculation of contribution of multiple interactions and efficiency of neutron detectors

International Nuclear Information System (INIS)

Androsenko, A.A.; Androsenko, P.A.; Kazakov, L.E.; Kononov, V.N.; Poletaev, E.D.

1986-01-01

Results of calculation of multiple neutron interactions contribution to efficiency of detectors with 6 Li glass and 10 B plate in the energy range of 0.01-1 MeV are given. The calculation was performed by the Monte-Carlo method using BRAND program complex. It is shown that a correction value for multiple neutron interaction in 6 Li glass of 1 mm thickness constitutes 4.5 % at energy of up to 100 keV and at higher energies has a complex energy dependence reaching 25 % at 440 keV

Thermal-neutron multiple scattering: critical double scattering

International Nuclear Information System (INIS)

Holm, W.A.

1976-01-01

A quantum mechanical formulation for multiple scattering of thermal-neutrons from macroscopic targets is presented and applied to single and double scattering. Critical nuclear scattering from liquids and critical magnetic scattering from ferromagnets are treated in detail in the quasielastic approximation for target systems slightly above their critical points. Numerical estimates are made of the double scattering contribution to the critical magnetic cross section using relevant parameters from actual experiments performed on various ferromagnets. The effect is to alter the usual Lorentzian line shape dependence on neutron wave vector transfer. Comparison with corresponding deviations in line shape resulting from the use of Fisher's modified form of the Ornstein-Zernike spin correlations within the framework of single scattering theory leads to values for the critical exponent eta of the modified correlations which reproduce the effect of double scattering. In addition, it is shown that by restricting the range of applicability of the multiple scattering theory from the outset to critical scattering, Glauber's high energy approximation can be used to provide a much simpler and more powerful description of multiple scattering effects. When sufficiently close to the critical point, it provides a closed form expression for the differential cross section which includes all orders of scattering and has the same form as the single scattering cross section with a modified exponent for the wave vector transfer

MPACT Fast Neutron Multiplicity System Prototype Development

Energy Technology Data Exchange (ETDEWEB)

D.L. Chichester; S.A. Pozzi; J.L. Dolan; M.T. Kinlaw; S.J. Thompson; A.C. Kaplan; M. Flaska; A. Enqvist; J.T. Johnson; S.M. Watson

2013-09-01

This document serves as both an FY2103 End-of-Year and End-of-Project report on efforts that resulted in the design of a prototype fast neutron multiplicity counter leveraged upon the findings of previous project efforts. The prototype design includes 32 liquid scintillator detectors with cubic volumes 7.62 cm in dimension configured into 4 stacked rings of 8 detectors. Detector signal collection for the system is handled with a pair of Struck Innovative Systeme 16-channel digitizers controlled by in-house developed software with built-in multiplicity analysis algorithms. Initial testing and familiarization of the currently obtained prototype components is underway, however full prototype construction is required for further optimization. Monte Carlo models of the prototype system were performed to estimate die-away and efficiency values. Analysis of these models resulted in the development of a software package capable of determining the effects of nearest-neighbor rejection methods for elimination of detector cross talk. A parameter study was performed using previously developed analytical methods for the estimation of assay mass variance for use as a figure-of-merit for system performance. A software package was developed to automate these calculations and ensure accuracy. The results of the parameter study show that the prototype fast neutron multiplicity counter design is very nearly optimized under the restraints of the parameter space.

Measurements of fusion neutron multiplication in spherical beryllium shells

International Nuclear Information System (INIS)

Giese, H.; Kappler, F.; Tayama, R.; Moellendorff, U. von; Alevra, A.; Klein, H.

1996-01-01

New results of spherical-shell transmission measurements with 14-MeV neutrons on pure beryllium shells up to 17 cm thick are reported. The spectral flux above 3 MeV was measured using a liquid scintillation detector. At 17 cm thickness, also the total neutron multiplication was measured using a Bonner sphere system. The results agree well with calculations using beryllium nuclear data from the EFF-1 or the ENDF/B-Vi library. (author). 23 refs, 4 figs, 1 tab

A neutron multiplicity analysis method for uranium samples with liquid scintillators

Energy Technology Data Exchange (ETDEWEB)

Zhou, Hao, E-mail: zhouhao_ciae@126.com [China Institute of Atomic Energy, P.O.BOX 275-8, Beijing 102413 (China); Lin, Hongtao [Xi' an Reasearch Institute of High-tech, Xi' an, Shaanxi 710025 (China); Liu, Guorong; Li, Jinghuai; Liang, Qinglei; Zhao, Yonggang [China Institute of Atomic Energy, P.O.BOX 275-8, Beijing 102413 (China)

2015-10-11

A new neutron multiplicity analysis method for uranium samples with liquid scintillators is introduced. An active well-type fast neutron multiplicity counter has been built, which consists of four BC501A liquid scintillators, a n/γdiscrimination module MPD-4, a multi-stop time to digital convertor MCS6A, and two Am–Li sources. A mathematical model is built to symbolize the detection processes of fission neutrons. Based on this model, equations in the form of R=F*P*Q*T could be achieved, where F indicates the induced fission rate by interrogation sources, P indicates the transfer matrix determined by multiplication process, Q indicates the transfer matrix determined by detection efficiency, T indicates the transfer matrix determined by signal recording process and crosstalk in the counter. Unknown parameters about the item are determined by the solutions of the equations. A {sup 252}Cf source and some low enriched uranium items have been measured. The feasibility of the method is proven by its application to the data analysis of the experiments.

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

International Nuclear Information System (INIS)

Sohn, Hee Dong

2012-02-01

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

Energy Technology Data Exchange (ETDEWEB)

Sohn, Hee Dong

2012-02-15

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1975-10-01

The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level

Observation of multiple Bragg reflections of neutrons in bent perfect crystals

Czech Academy of Sciences Publication Activity Database

Mikula, Pavol; Vrána, Miroslav; Šaroun, Jan; Seong, B. S.; Moon, MK.

2011-01-01

Roč. 634, č. 1 (2011), S108-S111 ISSN 0168-9002. [International Workshop on Neutron Optics. Grenoble, 17.03.2010-19.03.2010] R&D Projects: GA ČR GAP204/10/0654 Institutional research plan: CEZ:AV0Z10480505 Keywords : Neutron diffraction * Bent perfect crystal * Multiple reflections Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.207, year: 2011

Improving neutron multiplicity counting for the spatial dependence of multiplication: Results for spherical plutonium samples

Energy Technology Data Exchange (ETDEWEB)

Göttsche, Malte, E-mail: malte.goettsche@physik.uni-hamburg.de; Kirchner, Gerald

2015-10-21

The fissile mass deduced from a neutron multiplicity counting measurement of high mass dense items is underestimated if the spatial dependence of the multiplication is not taken into account. It is shown that an appropriate physics-based correction successfully removes the bias. It depends on four correction coefficients which can only be exactly determined if the sample geometry and composition are known. In some cases, for example in warhead authentication, available information on the sample will be very limited. MCNPX-PoliMi simulations have been performed to obtain the correction coefficients for a range of spherical plutonium metal geometries, with and without polyethylene reflection placed around the spheres. For hollow spheres, the analysis shows that the correction coefficients can be approximated with high accuracy as a function of the sphere's thickness depending only slightly on the radius. If the thickness remains unknown, less accurate estimates of the correction coefficients can be obtained from the neutron multiplication. The influence of isotopic composition is limited. The correction coefficients become somewhat smaller when reflection is present.

To the problem of the coherence length of neutrons

International Nuclear Information System (INIS)

Varga, P.

1992-11-01

The challenge of the high accuracy of certain optical measurements, the long coherence length of light provokes one to search for possibilities to enlarge the neutron coherence length. A proposal is made to achieve this by using a five or a four plate Bonse-Hart interferometer. A further problem is, whether the neutron beam is composed of wave packets or of overlapping independent monochromatic waves; it is considered that the former is more likely. (author) 12 refs.; 3 figs

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)

Theoretical aspects and experimental of neutronic interaction of multiplying media; Aspects theoriques et experimentaux de l'interaction neutronique entre milieux multiplicateurs de neutrons

Energy Technology Data Exchange (ETDEWEB)

Mougniot, J C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

A theoretical study of neutronic interaction of multiplying media is presented. The use of the surface multiplication constant and of the effective multiplication constant is considered. Three classical methods of interaction calculations are studied in parallel and the application of the Keff method to problems of nuclear safety is discussed. (authors) [French] Une etude theorique de l'interaction neutronique entre milieux multiplicateurs de neutrons est presentee. L'utilisation du coefficient de multiplication de surface et du coefficient de multiplication effectif est envisagee. Trois methodes classiques de calcul d'interaction sont etudiees parallelement et l'adaptation de la methode du Keff, aux problemes de securite nucleaire est ensuite discutee. (auteurs)

Fuel assembly inspection by three-dimensional neutron radiography

International Nuclear Information System (INIS)

Lapinski, N.P.; Reimann, K.J.; Berger, H.

1979-01-01

Radiographic inspection of complex objects such as fuel subassemblies often presents problems because superimposition of images at different depths in the object complicates interpretation. One method for obtaining and displaying three-dimensional neutron radiographic images in multiple-film laminagraphy; a series of radiographs generated at different angular orientations are superimposed to provide focussed images of any object plane. In the present work multiple-film neutron laminagraphs were generated using direct and indirect exposure techniques, with neutrons in thermal, epithermal, and fast energy ranges

Fission neutrons experiments, evaluation, modeling and open problems

CERN Document Server

Kornilov, Nikolay

2014-01-01

Although the fission of heavy nuclei was discovered over 75 years ago, many problems and questions still remain to be addressed and answered. The reader will be presented with an old, but persistent problem of this field: The contradiction between Prompt Fission Neutron (PFN) spectra measured with differential (microscopic) experiments and integral (macroscopic and benchmark) experiments (the Micro-Macro problem). The difference in average energy is rather small ~3% but it is stable and we cannot explain the difference due to experimental uncertainties. Can we measure the PFN spectrum with hig

Multiple small-angle neutron scattering studies of anisotropic materials

CERN Document Server

Allen, A J; Long, G G; Ilavsky, J

2002-01-01

Building on previous work that considered spherical scatterers and randomly oriented spheroidal scatterers, we describe a multiple small-angle neutron scattering (MSANS) analysis for nonrandomly oriented spheroids. We illustrate this with studies of the multi-component void morphologies found in plasma-spray thermal barrier coatings. (orig.)

Neutron magnetic multiple diffraction in a natural magnetite crystal

International Nuclear Information System (INIS)

Mazzocchi, V.L.; Parente, C.B.R.

1988-09-01

Neutron multiple diffraction has been employed in the study of the magnetism in magnetite (Fe 3 O 4 ). Magnetite has a crystallographic structure of an inverted spinel with tetrahedral A sites occupied solely by trivalent Fe 3+ ions and octahedral B sites occupied both by divalent Fe 2+ ions and the remaining Fe 3+ ions in random distribution. At room temperature magnetite is a Neel A-B ferrimagnet where the ions on the A, B sites are coupled antiferromagneticaly. This coupling disappears at T sup c approx. or approx.= 580 0 C. Employing a natural single crystal of magnetite experimental neutron multiple diffraction patterns were obtained for the primary reflection 111 at room temperature and 703 0 C. This reflection is almost entirely magnetic in origin resulting in 'Aufhellung' patterns below T c and mixed 'Aufhellung-Umweganregung' patterns above T c . Theoretical patterns were calculated employing the iterative method for the approximation of intensities by a Taylor series and compared to the experimental results. (author) [pt

Neutron induced radiation damage

International Nuclear Information System (INIS)

Williams, M.M.R.

1977-01-01

We derive a general expression for the number of displaced atoms of type j caused by a primary knock-on of type i. The Kinchin-Pease model is used, but considerably generalised to allow for realistic atomic potentials. Two cases are considered in detail: the single particle problem causing a cascade and the neutron initiated problem which leads to multiple subcascades. Numerical results have been obtained for a variety of scattering laws. An important conclusion is that neutron initiated damage is much more severe than atom-initiated damage and leads to the number of displaced atoms being a factor of (A+1) 2 /4A larger than the single primary knock-on theory predicts. A is the ratio of the atomic mass to the neutron mass. The importance of this result to the theory of neutron sputtering is explained. (orig.) [de

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.)

Benchmarking time-dependent neutron problems with Monte Carlo codes

International Nuclear Information System (INIS)

Couet, B.; Loomis, W.A.

1990-01-01

Many nuclear logging tools measure the time dependence of a neutron flux in a geological formation to infer important properties of the formation. The complex geometry of the tool and the borehole within the formation does not permit an exact deterministic modelling of the neutron flux behaviour. While this exact simulation is possible with Monte Carlo methods the computation time does not facilitate quick turnaround of results useful for design and diagnostic purposes. Nonetheless a simple model based on the diffusion-decay equation for the flux of neutrons of a single energy group can be useful in this situation. A combination approach where a Monte Carlo calculation benchmarks a deterministic model in terms of the diffusion constants of the neutrons propagating in the media and their flux depletion rates thus offers the possibility of quick calculation with assurance as to accuracy. We exemplify this approach with the Monte Carlo benchmarking of a logging tool problem, showing standoff and bedding response. (author)

Neutron multiplicities as a measure for scission time scales and reaction violences

International Nuclear Information System (INIS)

Knoche, K.; Scobel, W.; Sprute, L.

1991-01-01

We discuss the temporal evolution of the fusion-fission reactions 32 S + 197 Au, 232 Th measured for 838 MeV projectiles by means of the neutron clock method. The results confirm existent precision lifetime versus fissility data. The total neutron multiplicity as a measure of the initial excitation energy E * is compared with the folding angle method. (author). 13 refs, 8 figs

The multiple disk chopper neutron time-of-flight spectrometer at NIST

International Nuclear Information System (INIS)

Altorfer, F.B.; Cook, J.C.; Copley, J.R.D.

1995-01-01

A highly versatile multiple disk chopper neutron time-of-flight spectrometer is being installed at the Cold Neutron Research Facility of the National institute of Standards and Technology. This new instrument will fill an important gap in the portfolio of neutron inelastic scattering spectrometers in North America. It will be used for a wide variety of experiments such as studies of magnetic and vibrational excitations, tunneling spectroscopy, and quasielastic neutron scattering investigations of local and translational diffusion. The instrument uses disk choppers to monochromate and pulse the incident beam, and the energy changes of scattered neutrons are determined from their times-of-flight to a large array of detectors. The disks and the guide have been designed to make the instrument readily adaptable to the specific performance requirements of experimenters. The authors present important aspects of the design, as well as estimated values of the flux at the sample and the energy resolution for elastic scattering. The instrument should be operational in 1996

A Monte Carlo evaluation of analytical multiple scattering corrections for unpolarised neutron scattering and polarisation analysis data

International Nuclear Information System (INIS)

Mayers, J.; Cywinski, R.

1985-03-01

Some of the approximations commonly used for the analytical estimation of multiple scattering corrections to thermal neutron elastic scattering data from cylindrical and plane slab samples have been tested using a Monte Carlo program. It is shown that the approximations are accurate for a wide range of sample geometries and scattering cross-sections. Neutron polarisation analysis provides the most stringent test of multiple scattering calculations as multiply scattered neutrons may be redistributed not only geometrically but also between the spin flip and non spin flip scattering channels. A very simple analytical technique for correcting for multiple scattering in neutron polarisation analysis has been tested using the Monte Carlo program and has been shown to work remarkably well in most circumstances. (author)

Principles and problems in neutron nuclear data evaluation

International Nuclear Information System (INIS)

Schmidt, J.J.

1967-01-01

The history of neutron nuclear data evaluation is briefly summarized. The physical problems involved in nuclear data evaluation, such as discrepancies and inconsistencies between different experimental data sets and gaps in experimental information, are discussed. The discrepancies in the capture cross-section data for molybdenum and iron are chosen to illustrate the great difficulties in systematizing and automatizing the evaluation process. The technical problems of data evaluation, such as computer storage and the establishment of nuclear data files, are not discussed. (author)

The spectral element approach for the solution of neutron transport problems

International Nuclear Information System (INIS)

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

2011-01-01

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

Finite element based composite solution for neutron transport problems

International Nuclear Information System (INIS)

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

1995-01-01

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

RDANN a new methodology to solve the neutron spectra unfolding problem

Energy Technology Data Exchange (ETDEWEB)

Ortiz R, J.M.; Martinez B, M.R.; Vega C, H.R. [UAZ, Av. Ramon Lopez Velarde No. 801, 98000 Zacatecas (Mexico)

2006-07-01

The optimization processes known as Taguchi method and DOE methodology are applied to the design, training and testing of Artificial Neural Networks in the neutron spectrometry field, which offer potential benefits in the evaluation of the behavior of the net as well as the ability to examine the interaction of the weights and neurons inside the same one. In this work, the Robust Design of Artificial Neural Networks methodology is used to solve the neutron spectra unfolding problem, designing, training and testing an ANN using a set of 187 neutron spectra compiled by the International Atomic Energy Agency, to obtain the better neutron spectra unfolded from the Bonner spheres spectrometer's count rates. (Author)

RDANN a new methodology to solve the neutron spectra unfolding problem

International Nuclear Information System (INIS)

Ortiz R, J.M.; Martinez B, M.R.; Vega C, H.R.

2006-01-01

The optimization processes known as Taguchi method and DOE methodology are applied to the design, training and testing of Artificial Neural Networks in the neutron spectrometry field, which offer potential benefits in the evaluation of the behavior of the net as well as the ability to examine the interaction of the weights and neurons inside the same one. In this work, the Robust Design of Artificial Neural Networks methodology is used to solve the neutron spectra unfolding problem, designing, training and testing an ANN using a set of 187 neutron spectra compiled by the International Atomic Energy Agency, to obtain the better neutron spectra unfolded from the Bonner spheres spectrometer's count rates. (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

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

On an analytical evaluation of the flux and dominant eigenvalue problem for the steady state multi-group multi-layer neutron diffusion equation

Energy Technology Data Exchange (ETDEWEB)

Ceolin, Celina; Schramm, Marcelo; Bodmann, Bardo Ernst Josef; Vilhena, Marco Tullio Mena Barreto de [Universidade Federal do Rio Grande do Sul, Porto Alegre (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Bogado Leite, Sergio de Queiroz [Comissao Nacional de Energia Nuclear, Rio de Janeiro (Brazil)

2014-11-15

In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on an expansion in Taylor Series resulting in an analytical expression. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations.

Development of pulse neutron coal analyzer

International Nuclear Information System (INIS)

Jing Shiwie; Gu Deshan; Qiao Shuang; Liu Yuren; Liu Linmao; Jing Shiwei

2005-01-01

This article introduced the development of pulsed neutron coal analyzer by pulse fast-thermal neutron analysis technology in the Radiation Technology Institute of Northeast Normal University. The 14 MeV pulse neutron generator and bismuth germanate detector and 4096 multichannel analyzer were applied in this system. The multiple linear regression method employed to process data solved the interferential problem of multiple elements. The prototype (model MZ-MKFY) had been applied in Changshan and Jilin power plant for about a year. The results of measuring the main parameters of coal such as low caloric power, whole total water, ash content, volatile content, and sulfur content, with precision acceptable to the coal industry, are presented

Surrogate 239Pu(n, fxn) and 241Pu(n, fxn) average fission-neutron-multiplicity measurements

Energy Technology Data Exchange (ETDEWEB)

Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Alan, B. S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Akindele, O. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Casperson, R. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hughes, R. O. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fisher, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-09-26

We have constructed a new neutron-charged-particle detector array called NeutronSTARS. It has been described extensively in LLNL-TR-703909 [1] and Akindele et al [2]. We have used this new neutron-charged-particle array to measure the 241Pu and 239Pu fissionneutron multiplicity as a function of equivalent incident-neutron energy from 100 keV to 20 MeV. The experimental approach, detector array, data analysis, and results are summarized in the following sections.

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

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1981-01-01

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

Measuring method for effective neutron multiplication factor upon containing irradiated fuel assembly

International Nuclear Information System (INIS)

Ueda, Makoto; Mitsuhashi, Ishi; Sasaki, Tomoharu.

1993-01-01

A portion of irradiated fuel assemblies at a place where a reactivity effect is high, that is, at a place where neutron importance is high is replaced with standard fuel assemblies having a known composition to measure neutron fluxes at each of the places. An effective composition at the periphery of the standard fuel assemblies is determined by utilizing a calibration curve determined separately based on the composition and neutron flux values of the standard assemblies. By using the calibration curve determined separately based on this composition and the known composition of the standard fuel assemblies, an effective neutron multiplication factor for the fuel containing portion containing the irradiated fuel assemblies is recognized. Then, subcriticality is ensured and critical safety upon containing the fuel assemblies can be secured quantitatively. (N.H.)

Solid-State Neutron Multiplicity Counting System Using Commercial Off-the-Shelf Semiconductor Detectors

Energy Technology Data Exchange (ETDEWEB)

Rozhdestvenskyy, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-08-09

This work iterates on the first demonstration of a solid-state neutron multiplicity counting system developed at Lawrence Livermore National Laboratory by using commercial off-the-shelf detectors. The system was demonstrated to determine the mass of a californium-252 neutron source within 20% error requiring only one-hour measurement time with 20 cm² of active detector area.

Neutron targets of Moscow meson facility status, problems, prospects

Energy Technology Data Exchange (ETDEWEB)

Sidorkin, S.; Koptelov, E.; Perekrestenko, A.; Stavissky, Y.; Trushkin, V.; Sobolevsky, N. [Institute for Nuclear Research RAS, 60-th October Anniversary Prospect, Moscow (Russian Federation)

2001-03-01

The status, problems and possible perspectives of target complexes of the Moscow meson factory is described in the report. The results of test proton beam session to neutron source are analysed. Some technical features of targets and expected modes in the nearest sessions are stated. (author)

Radiation problems expected for the German spallation neutron source

International Nuclear Information System (INIS)

Goebel, K.

1981-01-01

The German project for the construction of a Spallation Neutron Source with high proton beam power (5.5 MW) will have to cope with a number of radiation problems. The present report describes these problems and proposes solutions for keeping exposures for the staff and release of activity and radiation into the environment as low as reasonably achievable. It is shown that the strict requirements of the German radiation protection regulations can be met. The main problem will be the exposure of maintenance personnel to remanent gamma radiation, as is the case at existing proton accelerators. Closed ventilation and cooling systems will reduce the release of (mainly short-lived) activity to acceptable levels. Shielding requirements for different sections are discussed, and it is demonstrated by calculations and extrapolations from experiments that fence-post doses well below 150 mrem/y can be obtained at distances of the order of 100 metres from the principal source points. The radiation protection system proposed for the Spallation Neutron Source is discussed, in particular the needs for monitor systems and a central radiation protection data base and alarm system. (orig.)

The hydrogen anomaly problem in neutron Compton scattering

Science.gov (United States)

Karlsson, Erik B.

2018-03-01

Neutron Compton scattering (also called ‘deep inelastic scattering of neutrons’, DINS) is a method used to study momentum distributions of light atoms in solids and liquids. It has been employed extensively since the start-up of intense pulsed neutron sources about 25 years ago. The information lies primarily in the width and shape of the Compton profile and not in the absolute intensity of the Compton peaks. It was therefore not immediately recognized that the relative intensities of Compton peaks arising from scattering on different isotopes did not always agree with values expected from standard neutron cross-section tables. The discrepancies were particularly large for scattering on protons, a phenomenon that became known as ‘the hydrogen anomaly problem’. The present paper is a review of the discovery, experimental tests to prove or disprove the existence of the hydrogen anomaly and discussions concerning its origin. It covers a twenty-year-long history of experimentation, theoretical treatments and discussions. The problem is of fundamental interest, since it involves quantum phenomena on the subfemtosecond time scale, which are not visible in conventional thermal neutron scattering but are important in Compton scattering where neutrons have two orders of magnitude times higher energy. Different H-containing systems show different cross-section deficiencies and when the scattering processes are followed on the femtosecond time scale the cross-section losses disappear on different characteristic time scales for each H-environment. The last section of this review reproduces results from published papers based on quantum interference in scattering on identical particles (proton or deuteron pairs or clusters), which have given a quantitative theoretical explanation both regarding the H-cross-section reduction and its time dependence. Some new explanations are added and the concluding chapter summarizes the conditions for observing the specific quantum

Manual for the Epithermal Neutron Multiplicity Detector (ENMC) for Measurement of Impure MOX and Plutonium Samples

International Nuclear Information System (INIS)

Menlove, H. O.; Rael, C. D.; Kroncke, K. E.; DeAguero, K. J.

2004-01-01

We have designed a high-efficiency neutron detector for passive neutron coincidence and multiplicity counting of dirty scrap and bulk samples of plutonium. The counter will be used for the measurement of impure plutonium samples at the JNC MOX fabrication facility in Japan. The counter can also be used to create working standards from bulk process MOX. The detector uses advanced design "3He tubes to increase the efficiency and to shorten the neutron die-away time. The efficiency is 64% and the die-away time is 19.1 ?s. The Epithermal Neutron Multiplicity Counter (ENMC) is designed for high-precision measurements of bulk plutonium samples with diameters of less than 200 mm. The average neutron energy from the sample can be measured using the ratio of the inner ring of He-3 tubes to the outer ring. This report describes the hardware, performance, and calibration for the ENMC.

Experimental Assessment of a New Passive Neutron Multiplication Counter for Partial Defect Verification of LWR Fuel Assemblies

International Nuclear Information System (INIS)

LaFleur, A.; Menlove, H.; Park, S.-H.; Lee, S. K.; Oh, J.-M.; Kim, H.-D.

2015-01-01

The development of non-destructive assay (NDA) capabilities to improve partial defect verification of spent fuel assemblies is needed to improve the timely detection of the diversion of significant quantities of fissile material. This NDA capability is important to the implementation of integrated safeguards for spent fuel verification by the International Atomic Energy Agency (IAEA) and would improve deterrence of possible diversions by increasing the risk of early detection. A new NDA technique called Passive Neutron Multiplication Counter (PNMC) is currently being developed at Los Alamos National Laboratory (LANL) to improve safeguards measurements of LightWater Reactor (LWR) fuel assemblies. The PNMC uses the ratio of the fast-neutron emission rate to the thermalneutron emission rate to quantify the neutron multiplication of the item. The fast neutrons versus thermal neutrons are measured using fission chambers (FC) that have differential shielding to isolate fast and thermal energies. The fast-neutron emission rate is directly proportional to the neutron multiplication in the spent fuel assembly; whereas, the thermalneutron leakage is suppressed by the fissile material absorption in the assembly. These FCs are already implemented in the basic Self-Interrogation Neutron Resonance Densitometry (SINRD) detector package. Experimental measurements of fresh and spent PWR fuel assemblies were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using a hybrid PNMC and SINRD detector. The results from these measurements provides valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. (author)

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

Virtual sampling in variational processing of Monte Carlo simulation in a deep neutron penetration problem

International Nuclear Information System (INIS)

Allagi, Mabruk O.; Lewins, Jeffery D.

1999-01-01

In a further study of virtually processed Monte Carlo estimates in neutron transport, a shielding problem has been studied. The use of virtual sampling to estimate the importance function at a certain point in the phase space depends on the presence of neutrons from the real source at that point. But in deep penetration problems, not many neutrons will reach regions far away from the source. In order to overcome this problem, two suggestions are considered: (1) virtual sampling is used as far as the real neutrons can reach, then fictitious sampling is introduced for the remaining regions, distributed in all the regions, or (2) only one fictitious source is placed where the real neutrons almost terminate and then virtual sampling is used in the same way as for the real source. Variational processing is again found to improve the Monte Carlo estimates, being best when using one fictitious source in the far regions with virtual sampling (option 2). When fictitious sources are used to estimate the importances in regions far away from the source, some optimization has to be performed for the proportion of fictitious to real sources, weighted against accuracy and computational costs. It has been found in this study that the optimum number of cells to be treated by fictitious sampling is problem dependent, but as a rule of thumb, fictitious sampling should be employed in regions where the number of neutrons from the real source fall below a specified limit for good statistics

FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1980-01-01

1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can

Measurements and applications of neutron multiple scattering in resonance region

International Nuclear Information System (INIS)

Ohkubo, Makio

1977-02-01

Capture yield of neutrons impinging on a thick material is complicated due to self-shielding and multiple scattering, especially in the resonance region. When the incident neutron energy is equal to a resonance energy of the material, capture probability of the neutron increases with sample thickness and reaches a saturation value P sub(CO). There is a simple relation between P sub(CO) and GAMMA sub(n)/GAMMA and the recoil energy by the Monte-Carlo calculation. To examine validity of the relation, P sub(CO) was measured for 19 resonances in 12 nuclides with thick samples, using a JAERI linac time-of-flight spectrometer with Moxon-Rae type gamma ray detector and transmission type neutron flux monitor. Results of the measurements confirmed the validity. With this relation, the GAMMA sub(n)/GAMMA or GAMMA sub(γ)/GAMMA value can be obtained from the measured P sub(CO), and also the level spins be determined by combining the transmission data. Because of the definition of P sub(CO), determination of the resonance parameters is not sensitive to the sample thickness as far as it is sufficiently thick. (auth.)

Criticality problems in energy dependent neutron transport theory

International Nuclear Information System (INIS)

Victory, H.D. Jr.

1979-01-01

The criticality problem is considered for energy dependent neutron transport in an isotropically scattering, homogeneous slab. Under a positivity assumption on the scattering kernel, an expression can be found relating the thickness of the slab to a parameter characterizing production by fission. This is accomplished by exploiting the Perron-Frobenius-Jentsch characterization of positive operators (i.e. those leaving invariant a normal, reproducing cone in a Banach space). It is pointed out that those techniques work for classes of multigroup problems were the Case singular eigenfunction approach is not as feasible as in the one-group theory, which is also analyzed

The single-collision thermalization approximation for application to cold neutron moderation problems

International Nuclear Information System (INIS)

Ritenour, R.L.

1989-01-01

The single collision thermalization (SCT) approximation models the thermalization process by assuming that neutrons attain a thermalized distribution with only a single collision within the moderating material, independent of the neutron's incident energy. The physical intuition on which this approximation is based is that the salient properties of neutron thermalization are accounted for in the first collision, and the effects of subsequent collisions tend to average out statistically. The independence of the neutron incident and outscattering energy leads to variable separability in the scattering kernel and, thus, significant simplification of the neutron thermalization problem. The approximation also addresses detailed balance and neutron conservation concerns. All of the tests performed on the SCT approximation yielded excellent results. The significance of the SCT approximation is that it greatly simplifies thermalization calculations for CNS design. Preliminary investigations with cases involving strong absorbers also indicates that this approximation may have broader applicability, as in the upgrading of the thermalization codes

Application of direct discrete method (DDM) to multigroup neutron transport problems

International Nuclear Information System (INIS)

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

2003-01-01

The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)

Positive solution of a time and energy dependent neutron transport problem

International Nuclear Information System (INIS)

Pao, C.V.

1975-01-01

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

On solution to the problem of reactor kinetics with delayed neutrons by Monte Carlo method

International Nuclear Information System (INIS)

Kyncl, Jan

2013-07-01

The initial value problem is addressed for the neutron transport equation and for the system of equations that describe the behaviour of emitters of delayed neutrons. Examination of the solution to this problem is based on several main assumptions concerning the behaviour of macroscopic effective cross-sections describing the reaction of the neutron with the medium, the temperature of medium and the remaining parameters of the equations. Formulation of these assumptions is adequately general and is in agreement with the properties of all known models of the physical quantities involved. Among others, the assumptions admit dependence of the macroscopic effective cross-sections and temperature on spatial coordinates and time that can be arbitrary to a great extent. The problem starts from a set of integro-differential equations. This problem is first transposed into the equivalent problem of solving a linear integral equation for neutron flux. This integral equation is solved by the method of successive iterations and its uniqueness is demonstrated. Numeric solution to the integral equation by Monte Carlo method consists in finding a functional of the exact solution. For this, a random process is set up and some random variables are proposed. Then it is demonstrated that each of these variables is an unbiased estimator of that functional. (author)

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

International Nuclear Information System (INIS)

Sahni, D.C.

1989-01-01

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

The importance of anisotropic scattering in high energy neutron transport problems

International Nuclear Information System (INIS)

Prillinger, G.; Mattes, M.

1984-01-01

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

Computational methods for the nuclear and neutron matter problems: Final report

International Nuclear Information System (INIS)

Kalos, M.H.; Chen, J.M.C.

1988-01-01

This paper discusses the following topics: variational Monte Carlo study of oxygen 16; microscopic calculations of alpha-neutron scattering; exact Monte Carlo treatment of the fermion problem; and random field method

Domain decomposition methods for the mixed dual formulation of the critical neutron diffusion problem; Methodes de decomposition de domaine pour la formulation mixte duale du probleme critique de la diffusion des neutrons

Energy Technology Data Exchange (ETDEWEB)

Guerin, P

2007-12-15

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)

Neutron generators with size scalability, ease of fabrication and multiple ion source functionalities

Science.gov (United States)

Elizondo-Decanini, Juan M

2014-11-18

A neutron generator is provided with a flat, rectilinear geometry and surface mounted metallizations. This construction provides scalability and ease of fabrication, and permits multiple ion source functionalities.

Single-sphere multiple-detector neutron spectrometer. Final report on Phase 1

International Nuclear Information System (INIS)

Sinclair, F.; Stern, I.; Hahn, R.W.; Entine, G.

1987-07-01

To address the problem of accurate, timely estimates of the neutron spectral flux, researchers are developing a monitoring instrument based on a single moderating sphere with a large number of independent sensors. Such a single-sphere spectrometer would allow easy measurement of quality factors. This is made possible by the recent development of a novel digital sensor which detects radiation induced errors in a dynamic random-access memory. During Phase I of the SBIR program, researchers constructed a first prototype of the single-sphere spectrometer, measured its response in a neutron flux from an isotopic Am-Be source in several geometries, and compared these with the results of Monte Carlo simulations of neutron transport. The preliminary results show that the approach is feasible and relatively straightforward

Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

International Nuclear Information System (INIS)

Kim, S.S.; Levine, S.H.

1985-01-01

An analytical/experimental method has been developed to monitor the subcritical reactivity and unfold the k/sub infinity/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. Therefore, it is called the Asymmetric Multiple Position Neutron Source (AMPNS) method. The AMPNS method employs nucleonic codes to analyze the neutron multiplication of a Cf-252 neutron source. An optimization program, GPM, is utilized to unfold the k/sub infinity/ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn State Breazeale TRIGA Reactor (PSBR). A significant result of this study is that it provides a method to monitor the criticality of a damaged core during the recovery period

Research on amplification multiple of source neutron number for ADS

International Nuclear Information System (INIS)

Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qingbiao; Ding Dazhao

1998-01-01

NJOY-91.91 and MILER code systems was applied to process and generate 44 group cross sections in AMPX master library format from CENDL-2 and ENDF/B-6. It is important an ADS (Accelerator-Driven System) assembly spectrum is used as the weighting spectrum for generating multi-group constants. Amplification multiples of source neutron number for several fast assemblies were calculated

Domain decomposition methods for the mixed dual formulation of the critical neutron diffusion problem

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)

An efficient method for generalized linear multiplicative programming problem with multiplicative constraints.

Science.gov (United States)

Zhao, Yingfeng; Liu, Sanyang

2016-01-01

We present a practical branch and bound algorithm for globally solving generalized linear multiplicative programming problem with multiplicative constraints. To solve the problem, a relaxation programming problem which is equivalent to a linear programming is proposed by utilizing a new two-phase relaxation technique. In the algorithm, lower and upper bounds are simultaneously obtained by solving some linear relaxation programming problems. Global convergence has been proved and results of some sample examples and a small random experiment show that the proposed algorithm is feasible and efficient.

Effect of fission dynamics on the spectra and multiplicities of prompt fission neutrons

International Nuclear Information System (INIS)

Nix, J.R.; Madland, D.G.; Sierk, A.J.

1985-01-01

With the goal of examining their effect on the spectra and multiplicities of the prompt neutrons emitted in fission, we discuss recent advances in a unified macroscopic-microscopic description of large-amplitude collective nuclear dynamics. The conversion of collective energy into single-particle excitation energy is calculated for a new surface-plus-window dissipation mechanism. By solving the Hamilton equations of motion for initial conditions appropriate to fission, we obtain the average fission-fragment translational kinetic energy and excitation energy. The spectra and multiplicities of the emitted neutrons, which depend critically upon the average excitation energy, are then calculated on the basis of standard nuclear evaporation theory, taking into account the average motion of the fission fragments, the distribution of fission-fragment residual nuclear temperature, the energy dependence of the cross section for the inverse process of compound-nucleus formation, and the possibility of multiple-chance fission. Some illustrative comparisons of our calculations with experimental data are shown

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1977-11-01

The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently

Study of the ferrimagnetic and paramagnetic phases of magnetite measured by multiple neutron diffraction

International Nuclear Information System (INIS)

Mazzocchi, V.L.

1992-01-01

Structural parameters of the ferrimagnetic and paramagnetic phases of magnetite have been refined from neutron multiple diffraction data. Experimental multiple diffraction patterns used in the refinement, were obtained by measuring the 111 primary reflection of a natural single crystal of this compound, at room temperature for the ferrimagnetic phase and 703 0 C for the paramagnetic phase. Corresponding theoretical patterns for both phases have been calculated by the program MULTI which uses the iterative method for the intensity calculations in neutron multiple diffraction. In this method intensities are calculated as Taylor series expansions summed up to a order sufficient for a good approximation. A step by step process has been used in the refinements according to the parameter-shift method. Both isotropic and anisotropic thermal parameters were used in the calculation of the temperature factor. (author)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Neutron excess generation by fusion neutron source for self-consistency of nuclear energy system

International Nuclear Information System (INIS)

Saito, Masaki; Artisyuk, V.; Chmelev, A.

1999-01-01

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities. (author)

Yields of correlated fragment pairs and neutron multiplicity in spontaneous fission of {sup 242}Pu

Energy Technology Data Exchange (ETDEWEB)

Veselsky, M.; Kliman, J.; Morhaccaron, M. [Institute of Physics of Slovak Academy of Sciences, Dubravska 9, 84228 Bratislava (Slovakia); Ramayya, A.V.; Kormicki, J.; Daniel, A.V. [Physics Department, Vanderbilt University, Nashville (United States)] Rasmussen, J.O. [Lawrence Berkeley National Laboratory, Berkeley (United States)] Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore (United States); Daniel, A.V.; Popeko, G.S.; Oganessian, Yu. Ts. [Joint Institute for Nuclear Research, Dubna (Russia)] Greiner, W. [Institut fur Theoretische Physik, J. W. Goethe Universitaet, Frankfurt a. M. (Germany); Aryaeinejad, R. [Idaho National Engineering Laboratory, Idaho Falls (United States)

1998-10-01

Yields of correlated fragment pairs were obtained in spontaneous fission of {sup 242}Pu. Charge, mass and neutron multiplicity distributions of fragment pairs were determined and compared to available data. The yield of cold fission without neutron emission was determined to about 10{percent} for the set of observed correlated fragment pairs. {copyright} {ital 1998 American Institute of Physics.}

Generalization of the Fourier Convergence Analysis in the Neutron Diffusion Eigenvalue Problem

International Nuclear Information System (INIS)

Lee, Hyun Chul; Noh, Jae Man; Joo, Hyung Kook

2005-01-01

Fourier error analysis has been a standard technique for the stability and convergence analysis of linear and nonlinear iterative methods. Lee et al proposed new 2- D/1-D coupling methods and demonstrated several advantages of the new methods by performing a Fourier convergence analysis of the methods as well as two existing methods for a fixed source problem. We demonstrated the Fourier convergence analysis of one of the 2-D/1-D coupling methods applied to a neutron diffusion eigenvalue problem. However, the technique cannot be used directly to analyze the convergence of the other 2-D/1-D coupling methods since some algorithm-specific features were used in our previous study. In this paper we generalized the Fourier convergence analysis technique proposed and analyzed the convergence of the 2-D/1-D coupling methods applied to a neutron diffusion Eigenvalue problem using the generalized technique

Neutron detection and multiplicity counting using a boron-loaded plastic scintillator/bismuth germanate phoswich detector array

International Nuclear Information System (INIS)

Miller, M.C.

1998-03-01

Neutron detection and multiplicity counting has been investigated using a boron-loaded plastic scintillator/bismuth germanate phoswich detector array. Boron-loaded plastic combines neutron moderation (H) and detection ( 10 B) at the molecular level, thereby physically coupling increasing detection efficiency and decreasing die-away time with detector volume. Both of these characteristics address a fundamental limitation of thermal-neutron multiplicity counters, where 3 He proportional counters are embedded in a polyethylene matrix. Separation of the phoswich response into its plastic scintillator and bismuth germanate components was accomplished on a pulse-by-pulse basis using custom integrator and timing circuits. In addition, a custom time-tag module was used to provide a time for each detector event. Analysis of the combined energy and time event stream was performed by calibrating each detector's response and filtering based on the presence of a simultaneous energy deposition corresponding to the 10 B(n,alpha) reaction products in the plastic scintillator (93 keV ee ) and the accompanying neutron-capture gamma ray in the bismuth germanate (478 keV). Time-correlation analysis was subsequently performed on the filtered event stream to obtain shift-register-type singles and doubles count rates. Proof-of-principle measurements were conducted with a variety of gamma-ray and neutron sources including 137 Cs, 54 Mn, AmLi, and 252 Cf. Results of this study indicate that a neutron-capture probability of ∼10% and a die-away time of ∼10 micros are possible with a 4-detector array with a detector volume of 1600 cm 3 . Simulations were performed that indicate neutron-capture probabilities on the order of 50% and die-away times of less than 4 micros are realistically achievable. While further study will be required for practical application of such a detection system, the results obtained in this investigation are encouraging and may lead to a new class of high

Warhead verification as inverse problem: Applications of neutron spectrum unfolding from organic-scintillator measurements

Science.gov (United States)

Lawrence, Chris C.; Febbraro, Michael; Flaska, Marek; Pozzi, Sara A.; Becchetti, F. D.

2016-08-01

Verification of future warhead-dismantlement treaties will require detection of certain warhead attributes without the disclosure of sensitive design information, and this presents an unusual measurement challenge. Neutron spectroscopy—commonly eschewed as an ill-posed inverse problem—may hold special advantages for warhead verification by virtue of its insensitivity to certain neutron-source parameters like plutonium isotopics. In this article, we investigate the usefulness of unfolded neutron spectra obtained from organic-scintillator data for verifying a particular treaty-relevant warhead attribute: the presence of high-explosive and neutron-reflecting materials. Toward this end, several improvements on current unfolding capabilities are demonstrated: deuterated detectors are shown to have superior response-matrix condition to that of standard hydrogen-base scintintillators; a novel data-discretization scheme is proposed which removes important detector nonlinearities; and a technique is described for re-parameterizing the unfolding problem in order to constrain the parameter space of solutions sought, sidestepping the inverse problem altogether. These improvements are demonstrated with trial measurements and verified using accelerator-based time-of-flight calculation of reference spectra. Then, a demonstration is presented in which the elemental compositions of low-Z neutron-attenuating materials are estimated to within 10%. These techniques could have direct application in verifying the presence of high-explosive materials in a neutron-emitting test item, as well as other for treaty verification challenges.

Determination of nuclear friction in strongly damped reactions from prescission neutron multiplicities

NARCIS (Netherlands)

Wilczynski, J; SiwekWilczynska, K; Wilschut, HW

Nonfusion, fissionlike reactions in collisions of four heavy systems (well below the fusion extra-push energy threshold), Mr which Hinde and co-workers had measured the prescission neutron multiplicities, have been analyzed in terms of the deterministic dynamic model of Feldmeier coupled to a

Correction for variable moderation and multiplication effects associated with thermal neutron coincidence counting

International Nuclear Information System (INIS)

Baron, N.

1978-01-01

A correction is described for multiplication and moderation when doing passive thermal neutron coincidence counting nondestructive assay measurements on powder samples of PuO 2 mixed arbitrarily with MgO, SiO 2 , and moderating material. The multiplication correction expression is shown to be approximately separable into the product of two independent terms; F/sub Pu/ which depends on the mass of 240 Pu, and F/sub αn/ which depends on properties of the matrix material. Necessary assumptions for separability are (1) isotopic abundances are constant, and (2) fission cross sections are independent of incident neutron energy: both of which are reasonable for the 8% 240 Pu powder samples considered here. Furthermore since all prompt fission neutrons are expected to have nearly the same energy distributions, variations among different samples can be due only to the moderating properties of the samples. Relative energy distributions are provided by a thermal neutron well counter having two concentric rings of 3 He proportional counters placed symmetrically about the well. Measured outer-to-inner ring ratios raised to an empirically determined power for coincidences, (N/sup I//N/sup O/)/sup Z/, and singles, (T/sup O//T/sup I/)/sup delta/, provide corrections for moderation and F/sub αn/ respectively, and F/sub Pu/ is approximated by M 240 /sup X//M 240 . The exponents are calibration constants determined by a least squares fitting procedure using standards' data. System calibration is greatly simplified using the separability principle. Once appropriate models are established for F/sub Pu/ and F/sub αn/, only a few standards are necessary to determine the calibration constants associated with these terms. Since F/sub Pu/ is expressed as a function of M 240 , correction for multiplication in a subsequent assay demands only a measurement of F/sub αn/

Multiplicity of neutrons from violent heavy-ion collisions: 40Ar + Th and U from 10 to 77 MeV/u

International Nuclear Information System (INIS)

Jahnke, U.; Cramer, B.; Ingold, G.; Schwinn, E.; Charvet, J.L.; Frehaut, J.; Lott, B.; Morjean, M.; Patin, Y.; Pranal, Y.; Uzureau, J.L.; Doubre, H.; Galin, J.; Guerreau, D.; Jiang, D.X.; Pouthas, J.; Sokolov, A.; Gatty, B.; Jacquet, D.; Magnago, C.

1988-12-01

With large 4π scintillator tanks the multiplicity of neutrons released from the most dissipative reactions in 10 to 77 MeV/u 40 Ar induced collisions with Th and U nuclei has been investigated. The central issue is the relation between the fission-fragment folding angle or linear momentum transfer and the multiplicity of evaporated neutrons and charged particles or dissipated energy. Their multiplicity points to a 'soft saturation' of the maximum energy deposit with increasing bombarding energy near 700 MeV of excitation. Unlike the folding-angle distributions, the inclusive neutron multiplicity spectra, which are unbiased by specific decay properties of the intermediate nuclear system, do not show a decline of the most dissipative processes within this range of incident energy. (orig.)

Passive assay of plutonium metal plates using a fast-neutron multiplicity counter

Energy Technology Data Exchange (ETDEWEB)

Di Fulvio, A., E-mail: difulvio@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Shin, T.H.; Jordan, T.; Sosa, C.; Ruch, M.L.; Clarke, S.D. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Pozzi, S.A. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

2017-05-21

We developed a fast-neutron multiplicity counter based on organic scintillators (EJ-309 liquid and stilbene). The system detects correlated photon and neutron multiplets emitted by fission reactions, within a gate time of tens of nanoseconds. The system was used at Idaho National Laboratory to assay a variety of plutonium metal plates. A coincidence counting strategy was used to quantify the {sup 240}Pu effective mass of the samples. Coincident neutrons, detected within a 40-ns coincidence window, show a monotonic trend, increasing with the {sup 240}Pu-effective mass (in this work, we tested the 0.005–0.5 kg range). After calibration, the system estimated the {sup 240}Pu effective mass of an unknown sample ({sup 240}Pu{sub eff} >50 g) with an uncertainty lower than 1% in a 4-min assay time.

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1977-11-01

The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.

THE MULTIPLE CHOICE PROBLEM WITH INTERACTIONS BETWEEN CRITERIA

Directory of Open Access Journals (Sweden)

Luiz Flavio Autran Monteiro Gomes

2015-12-01

Full Text Available ABSTRACT An important problem in Multi-Criteria Decision Analysis arises when one must select at least two alternatives at the same time. This can be denoted as a multiple choice problem. In other words, instead of evaluating each of the alternatives separately, they must be combined into groups of n alternatives, where n = 2. When the multiple choice problem must be solved under multiple criteria, the result is a multi-criteria, multiple choice problem. In this paper, it is shown through examples how this problemcan be tackled on a bipolar scale. The Choquet integral is used in this paper to take care of interactions between criteria. A numerical application example is conducted using data from SEBRAE-RJ, a non-profit private organization that has the mission of promoting competitiveness, sustainable developmentand entrepreneurship in the state of Rio de Janeiro, Brazil. The paper closes with suggestions for future research.

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

International Nuclear Information System (INIS)

Lefvert, T.

1975-11-01

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

Fourier convergence analysis applied to neutron diffusion Eigenvalue problem

International Nuclear Information System (INIS)

Lee, Hyun Chul; Noh, Jae Man; Joo, Hyung Kook

2004-01-01

Fourier error analysis has been a standard technique for the stability and convergence analysis of linear and nonlinear iterative methods. Though the methods can be applied to Eigenvalue problems too, all the Fourier convergence analyses have been performed only for fixed source problems and a Fourier convergence analysis for Eigenvalue problem has never been reported. Lee et al proposed new 2-D/1-D coupling methods and they showed that the new ones are unconditionally stable while one of the two existing ones is unstable at a small mesh size and that the new ones are better than the existing ones in terms of the convergence rate. In this paper the convergence of method A in reference 4 for the diffusion Eigenvalue problem was analyzed by the Fourier analysis. The Fourier convergence analysis presented in this paper is the first one applied to a neutronics eigenvalue problem to the best of our knowledge

Design of auto-control high-voltage control system of pulsed neutron generator

International Nuclear Information System (INIS)

Lv Juntao

2008-01-01

It is difficult to produce multiple anode controlling time sequences under different logging mode for the high-voltage control system of the conventional pulsed neutron generator. It is also difficult realize sequential control among anode high-voltage, filament power supply and target voltage to make neutron yield stable. To these problems, an auto-control high-voltage system of neutron pulsed generator was designed. It not only can achieve anode high-voltage double blast time sequences, which can measure multiple neutron blast time sequences such as Σ, activated spectrum, etc. under inelastic scattering mode, but also can realize neutron generator real-time measurement of multi-state parameters and auto-control such as target voltage pulse width modulation (PWM), filament current, anode current, etc., there by it can produce stable neutron yield and realize stable and accurate measurement of the pulsed neutron full spectral loging tool. (authors)

Three-dimensional multiple reciprocity boundary element method for one-group neutron diffusion eigenvalue computations

International Nuclear Information System (INIS)

Itagaki, Masafumi; Sahashi, Naoki.

1996-01-01

The multiple reciprocity method (MRM) in conjunction with the boundary element method has been employed to solve one-group eigenvalue problems described by the three-dimensional (3-D) neutron diffusion equation. The domain integral related to the fission source is transformed into a series of boundary-only integrals, with the aid of the higher order fundamental solutions based on the spherical and the modified spherical Bessel functions. Since each degree of the higher order fundamental solutions in the 3-D cases has a singularity of order (1/r), the above series of boundary integrals requires additional terms which do not appear in the 2-D MRM formulation. The critical eigenvalue itself can be also described using only boundary integrals. Test calculations show that Wielandt's spectral shift technique guarantees rapid and stable convergence of 3-D MRM computations. (author)

Simulation of neutron multiplicity measurements using Geant4. Open source software for nuclear arms control

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.

Violence of heavy-ion reactions from neutron multiplicity: 11 to 20A MeV /sup 20/Ne+ /sup 238/U

International Nuclear Information System (INIS)

Jahnke, U.; Ingold, G.; Hilscher, D.; Lehmann, M.; Schwinn, E.; Zank, P.

1986-01-01

The suitability of the neutron multiplicity as a gauge for the violence of medium-energy heavy-ion reactions is investigated for the first time. For this purpose the number of neutrons emitted from fission reactions induced by 220-, 290-, and 400-MeV /sup 20/Ne on /sup 238/U is registered event-by-event with a large 4π scintillator tank. It is shown that the neutron multiplicity is indeed closely related to the two quantities characterizing the violence: the induced total intrinsic excitation and the linear momentum transfer

Sensitivity analysis of physical/operational parameters in neutron multiplicity counting

International Nuclear Information System (INIS)

Peerani, P.; Marin Ferrer, M.

2007-01-01

In this paper, we perform a sensitivity study on the influence of various physical and operational parameters on the results of neutron multiplicity counting. The purpose is to have a better understanding of the importance of each component and its contribution to the measurement uncertainty. Then we will be able to determine the optimal conditions for the operational parameters and for detector design and as well to point out weaknesses in the knowledge of critical fundamental nuclear data

Multiple scattering and attenuation corrections in Deep Inelastic Neutron Scattering experiments

International Nuclear Information System (INIS)

Dawidowski, J; Blostein, J J; Granada, J R

2006-01-01

Multiple scattering and attenuation corrections in Deep Inelastic Neutron Scattering experiments are analyzed. The theoretical basis of the method is stated, and a Monte Carlo procedure to perform the calculation is presented. The results are compared with experimental data. The importance of the accuracy in the description of the experimental parameters is tested, and the implications of the present results on the data analysis procedures is examined

Uranium mass and neutron multiplication factor estimates from time-correlation coincidence counts

Energy Technology Data Exchange (ETDEWEB)

Xie, Wenxiong [China Academy of Engineering Physics, Center for Strategic Studies, Beijing 100088 (China); Li, Jiansheng [China Academy of Engineering Physics, Institute of Nuclear Physics and Chemistry, Mianyang 621900 (China); Zhu, Jianyu [China Academy of Engineering Physics, Center for Strategic Studies, Beijing 100088 (China)

2015-10-11

Time-correlation coincidence counts of neutrons are an important means to measure attributes of nuclear material. The main deficiency in the analysis is that an attribute of an unknown component can only be assessed by comparing it with similar known components. There is a lack of a universal method of measurement suitable for the different attributes of the components. This paper presents a new method that uses universal relations to estimate the mass and neutron multiplication factor of any uranium component with known enrichment. Based on numerical simulations and analyses of 64 highly enriched uranium components with different thicknesses and average radii, the relations between mass, multiplication and coincidence spectral features have been obtained by linear regression analysis. To examine the validity of the method in estimating the mass of uranium components with different sizes, shapes, enrichment, and shielding, the features of time-correlation coincidence-count spectra for other objects with similar attributes are simulated. Most of the masses and multiplications for these objects could also be derived by the formulation. Experimental measurements of highly enriched uranium castings have also been used to verify the formulation. The results show that for a well-designed time-dependent coincidence-count measuring system of a uranium attribute, there are a set of relations dependent on the uranium enrichment by which the mass and multiplication of the measured uranium components of any shape and size can be estimated from the features of the source-detector coincidence-count spectrum.

Tritium solid targets for intense D-T neutron production and its related problems

International Nuclear Information System (INIS)

Sumita, Kenji

1988-01-01

This review paper is divided into three parts. Firstly, to attain an intense neutron production rate, the construction of a design with a higher tritium-containing surface and an effective cooling system like a rotating target device are discussed. The maximum attainable intensity based on tritium solid targets shall be estimated regarding planning for future D-T sources. Secondly, on the way to carry out some experiments, an absolute intensity calibration and an angular dependent neutron energy spectrum of the neutron source are essential parameters to analyse the results of the experiments. Sometimes the space dependent neutron spectrum is required as well as the space dependent neutron flux near the targets and irradiation samples. The measurement methods and their examples are reviewed for tritium solid targets. The third part is devoted to discuss the protection to tritium contamination problems due to unavoidable release of tritium gas from targets. Performance and effectiveness of tritium collection systems for intense D-T neutron sources shall be discussed in some examples. Tritium contamination incidents due to the faulted film powder of target surface are also reported in some real incident cases. (author). Abstract only

Can a large neutron excess help solve the baryon loading problem in gamma-Ray burst fireballs?

Science.gov (United States)

Fuller; Pruet; Abazajian

2000-09-25

We point out that the baryon loading problem in gamma-ray burst (GRB) models can be ameliorated if a significant fraction of the baryons which inertially confine the fireball is converted to neutrons. A high neutron fraction can result in a reduced transfer of energy from relativistic light particles in the fireball to baryons. The energy needed to produce the required relativistic flow in the GRB is consequently reduced, in some cases by orders of magnitude. A high neutron-to-proton ratio has been calculated in neutron star-merger fireball environments. Significant neutron excess also could occur near compact objects with high neutrino fluxes.

Characterizations of double pulsing in neutron multiplicity and coincidence counting systems

Energy Technology Data Exchange (ETDEWEB)

Koehler, Katrina E., E-mail: kkoehler@lanl.gov [Los Alamos National Laboratory, P. O. Box 1663, Los Alamos, NM 87545 (United States); Henzl, Vladimir [Los Alamos National Laboratory, P. O. Box 1663, Los Alamos, NM 87545 (United States); Croft, Stephen S. [Oak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN 37831 (United States); Henzlova, Daniela; Santi, Peter A. [Los Alamos National Laboratory, P. O. Box 1663, Los Alamos, NM 87545 (United States)

2016-10-01

Passive neutron coincidence/multiplicity counters are subject to non-ideal behavior, such as double pulsing and dead time. It has been shown in the past that double-pulsing exhibits a distinct signature in a Rossi-alpha distribution, which is not readily noticed using traditional Multiplicity Shift Register analysis. However, it has been assumed that the use of a pre-delay in shift register analysis removes any effects of double pulsing. In this work, we use high-fidelity simulations accompanied by experimental measurements to study the effects of double pulsing on multiplicity rates. By exploiting the information from the double pulsing signature peak observable in the Rossi-alpha distribution, the double pulsing fraction can be determined. Algebraic correction factors for the multiplicity rates in terms of the double pulsing fraction have been developed. We discuss the role of these corrections across a range of scenarios.

Utilizing of computational tools on the modelling of a simplified problem of neutron shielding

Energy Technology Data Exchange (ETDEWEB)

Lessa, Fabio da Silva Rangel; Platt, Gustavo Mendes; Alves Filho, Hermes [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico]. E-mails: fsrlessa@gmail.com; gmplatt@iprj.uerj.br; halves@iprj.uerj.br

2007-07-01

In the current technology level, the investigation of several problems is studied through computational simulations whose results are in general satisfactory and much less expensive than the conventional forms of investigation (e.g., destructive tests, laboratory measures, etc.). Almost all of the modern scientific studies are executed using computational tools, as computers of superior capacity and their systems applications to make complex calculations, algorithmic iterations, etc. Besides the considerable economy in time and in space that the Computational Modelling provides, there is a financial economy to the scientists. The Computational Modelling is a modern methodology of investigation that asks for the theoretical study of the identified phenomena in the problem, a coherent mathematical representation of such phenomena, the generation of a numeric algorithmic system comprehensible for the computer, and finally the analysis of the acquired solution, or still getting use of pre-existent systems that facilitate the visualization of these results (editors of Cartesian graphs, for instance). In this work, was being intended to use many computational tools, implementation of numeric methods and a deterministic model in the study and analysis of a well known and simplified problem of nuclear engineering (the neutron transport), simulating a theoretical problem of neutron shielding with physical-material hypothetical parameters, of neutron flow in each space junction, programmed with Scilab version 4.0. (author)

Utilizing of computational tools on the modelling of a simplified problem of neutron shielding

International Nuclear Information System (INIS)

Lessa, Fabio da Silva Rangel; Platt, Gustavo Mendes; Alves Filho, Hermes

2007-01-01

In the current technology level, the investigation of several problems is studied through computational simulations whose results are in general satisfactory and much less expensive than the conventional forms of investigation (e.g., destructive tests, laboratory measures, etc.). Almost all of the modern scientific studies are executed using computational tools, as computers of superior capacity and their systems applications to make complex calculations, algorithmic iterations, etc. Besides the considerable economy in time and in space that the Computational Modelling provides, there is a financial economy to the scientists. The Computational Modelling is a modern methodology of investigation that asks for the theoretical study of the identified phenomena in the problem, a coherent mathematical representation of such phenomena, the generation of a numeric algorithmic system comprehensible for the computer, and finally the analysis of the acquired solution, or still getting use of pre-existent systems that facilitate the visualization of these results (editors of Cartesian graphs, for instance). In this work, was being intended to use many computational tools, implementation of numeric methods and a deterministic model in the study and analysis of a well known and simplified problem of nuclear engineering (the neutron transport), simulating a theoretical problem of neutron shielding with physical-material hypothetical parameters, of neutron flow in each space junction, programmed with Scilab version 4.0. (author)

neutron multiplicity measurements on 220 l waste drums containing Pu in the range 0.1-1 g 240Pueff with the time interval analysis method

International Nuclear Information System (INIS)

Baeten, P.; Bruggeman, M.; Carchon, R.; De Boeck, W.

1998-01-01

Measurement results are presented for the assay of plutonium in 220 l waste drums containing Pu-masses in the range 0.1-1 g 240 Pu eff obtained with the time interval analysis (TIA) method. TIA is a neutron multiplicity method based on the concept of one- and two-dimensional Rossi-alpha distributions. The main source of measurement bias in neutron multiplicity measurements at low count-rates is the impredictable variation of the high-multiplicity neutron background of spallation neutrons induced by cosmic rays. The TIA-method was therefore equipped with a special background filter, which is designed and optimized to reduce the influence of these spallation neutrons by rejecting the high-multiplicity events. The measurement results, obtained with the background correction filter outlined in this paper, prove the repeatability and validity of the TIA-method and show that multiplicity counting with the TIA-technique is applicable for masses as low as 0.1 g 240 Pu eff even at a detection efficiency of 12%. (orig.)

To the question of definition of fissile material mass and neutron multiplication in deep sub-critical systems

International Nuclear Information System (INIS)

Dulin, V.V.

2006-01-01

A method of determination neutrons multiplication in deep sub-critical multiplying media has been developed. It is based on a modified of Rossi - alpha method. It will consist in use of integral on time (a method of the areas) from correlated parts of distribution and integral in area, independent of time a part of distribution (area of a constant background). It allows to spend the calculated analysis, using the integrated equation on time for a neutrons flux and to not use representation of point kinetic model. A calculation spatially-correlation factor the adjoint (relative the detector count rate) inhomogeneous equation is used. Its calculation takes into account fission both in multiplying media and in a spontaneous neutron source. Measurements with plutonium-steel and uranium-steel blocks, and blocks from uranium and plutonium dioxide of different enrichment are have been carried out. The measured values of neutrons multiplication in a range 1.03-1.82 will be well coordinated to results of calculations. The question on an opportunity of definition of weight of the measured blocks of fissile material is considered [ru

Neutron multiplication and shielding problems in pressurized water reactor spent fuel shipping casks

International Nuclear Information System (INIS)

Devillers, C.; Blum, P.

1977-01-01

To evaluate the degree of accuracy of computational methods used in the shield design of spent fuel shipping casks, comparisons have been made between biological dose-rate calculations and measurements at the surface of a cask carrying three pressurized water reactor fuel assemblies. Neutron dose-rate measurements made with the fuel-carrying region successively wet and dry are also used to derive an experimental value of the k/sub eff/ of the wet fuel assemblies. Results obtained by this method are shown to be consistent with criticality calculations, taking into account fuel depletion

Determining neutron multiplication factor in the infinite system by reactivity dependence on one dimension of the reactor core

International Nuclear Information System (INIS)

Pesic, M.

1975-01-01

The objective of this task was to apply Fermi age theory for determining τ and neutron multiplication factor in infinite medium by measuring reactivity coefficient of heavy water in heterogeneous mixed reactor lattice. Basis of experiment is the measurement of stable reactor period. Measurement of heavy water reactivity coefficient by measuring the stable reactor period is described for chosen overcritical heavy water levels. Calculated values of infinite multiplication factor for measured neutron age data are presented and they are compared to expected theoretical values

Neutron dosimetry program at Mound - problems and solutions

International Nuclear Information System (INIS)

Winegardner, M.K.

1991-01-01

The Mound personnel neutron dosimetry program utilizes TLD albedo technology. The neutron dosimeter design incorporates a two-element spectrometer for site-specific neutron quality determination and empirical application of field neutron calibration factors. Design elements feature two Li(6)F (TLD- 600) chips for neutron detection and one Li(7)F (TLD-700) chip for gamma compensation of the TLD- 600 chips. One TLD-600 chip is Cadmium shielded on the front side of the dosimeter, the other is Cadmium shielded from the back side. Tin filters are placed opposite of the Cadmium shield on each of the TLD-600 chips and on both sides of the TLD-700 chip for symmetrically equivalent gamma absorption characteristics. Neutron quality determination is accomplished by the albedo neutron-to- incident thermal neutron response ratio above the Cadmium cutoff. This front Cadmium shielded-to-back Cadmium shielded response ratio, compensated for the presence of gamma radiation, provides the basis for neutron energy calibration via the albedo response curve

On the problem of bound states of pions and neutrons

International Nuclear Information System (INIS)

Gudima, K.K.; Karnaukhov, V.A.

1992-01-01

The problem of existence of the bound states of negative pions and neutrons has been widely discussed for the last years. It is considered possibilities of the experimental observation of pion-neutron clusters, if they do exist, in nucleus-nucleus collisions. The yields of exotic fragments π -Z n A in the interactions of 12 C and 56 Fe with 208 Pb at the energies from 0.3 to 3.7 GeV per nucleon are calculated. For 40 Ar+ 238 U and 139 La+ 238 U collisions the calculations were performed at the energied of 1.8 GeV and 1.3 GeV per nucleon, respectively. These calculations were performed in the framework of the coalescence mechanism with the differential cross sections for pion and neutron production generated by firestreak model. The differential cross sections for production of π -1 n -2 , π -2 N 2 , π - n 4 , π -4 n 6 , and π -12 n 6 were calculated. It is shown that the use of very heavy projectiles like 56 Fe and 139 La has a great advantage in the experimental search for the exotic clusters. 20 refs.; 8 figs

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

International Nuclear Information System (INIS)

Oeztuerk, Hakan; Guengoer, Sueleyman

2008-01-01

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

Genetic Algorithms for Multiple-Choice Problems

Science.gov (United States)

Aickelin, Uwe

2010-04-01

This thesis investigates the use of problem-specific knowledge to enhance a genetic algorithm approach to multiple-choice optimisation problems.It shows that such information can significantly enhance performance, but that the choice of information and the way it is included are important factors for success.Two multiple-choice problems are considered.The first is constructing a feasible nurse roster that considers as many requests as possible.In the second problem, shops are allocated to locations in a mall subject to constraints and maximising the overall income.Genetic algorithms are chosen for their well-known robustness and ability to solve large and complex discrete optimisation problems.However, a survey of the literature reveals room for further research into generic ways to include constraints into a genetic algorithm framework.Hence, the main theme of this work is to balance feasibility and cost of solutions.In particular, co-operative co-evolution with hierarchical sub-populations, problem structure exploiting repair schemes and indirect genetic algorithms with self-adjusting decoder functions are identified as promising approaches.The research starts by applying standard genetic algorithms to the problems and explaining the failure of such approaches due to epistasis.To overcome this, problem-specific information is added in a variety of ways, some of which are designed to increase the number of feasible solutions found whilst others are intended to improve the quality of such solutions.As well as a theoretical discussion as to the underlying reasons for using each operator,extensive computational experiments are carried out on a variety of data.These show that the indirect approach relies less on problem structure and hence is easier to implement and superior in solution quality.

Effect of double false pulses in calibrated neutron coincidence collar during measuring time-correlated neutrons from PuBe neutron sources

Energy Technology Data Exchange (ETDEWEB)

Nguyen, Tam Cong, E-mail: tam.nguyen.cong@energia.mta.hu; Huszti, Jozsef; Nguyen, Quan Van

2015-09-01

Effect of double false pulses of preamplifiers in neutron coincidence collar was investigated to explain non-parallel shape of calibrated D/S–M{sub Pu} curves of two commercial neutron coincidence collars, JCC-31 and JCC-13. Two curves, which were constructed from D/S ratio (doubles and singles count rate), and Pu content M{sub Pu}, of the same set of secondary standard PuBe neutron sources, should be parallel. Non-parallelism rises doubt about usability of the method based on this curve for determination of Pu content in PuBe neutron sources. We have shown in three steps that the problem originates from double false pulses of preamplifiers in JCC-13. First we used a pulse train diagram for analyzing the non-parallel shape, second we used Rossi-Alpha distribution measured by pulse train recorder developed in our institute and finally, we investigated the effect of inserted noise pulses. This implies a new type of QA test option in traditional multiplicity shift registers for excluding presence of double false pulses.

Absorption and activation techniques in measurements of fast-neutron capture cross sections

International Nuclear Information System (INIS)

Bergqvist, I.

1982-01-01

The absorption and activation methods have been applied for a long time to systematic studies of fast neutron capture cross sections. Both methods are simple in principle but difficult in practice. The simplicity should ensure a wider use of the methods in particular for problems which may be complicated to approach with other methods. The difficulties encountered in absorption measurements are related to multiple scattering and resonance shielding effects. In activation experiments the influence of secondary low-energy neutrons causes the main problems

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

On a possible use of multiple Bragg reflections for high-resolution monochromatization of neutrons

Czech Academy of Sciences Publication Activity Database

Mikula, Pavol; Vrána, Miroslav; Wagner, V.

2004-01-01

Roč. 350, - (2004), e667-e670 ISSN 0921-4526 R&D Projects: GA ČR GA202/03/0891 Keywords : neutron diffraction * multiple reflections * higg-resolution monochromator Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.679, year: 2004

Adjacent-cell Preconditioners for solving optically thick neutron transport problems

International Nuclear Information System (INIS)

Azmy, Y.Y.

1994-01-01

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

Neutron transmission benchmark problems for iron and concrete shields in low, intermediate and high energy proton accelerator facilities

Energy Technology Data Exchange (ETDEWEB)

Nakane, Yoshihiro; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Katsumi [and others

1996-09-01

Benchmark problems were prepared for evaluating the calculation codes and the nuclear data for accelerator shielding design by the Accelerator Shielding Working Group of the Research Committee on Reactor Physics in JAERI. Four benchmark problems: transmission of quasi-monoenergetic neutrons generated by 43 MeV and 68 MeV protons through iron and concrete shields at TIARA of JAERI, neutron fluxes in and around an iron beam stop irradiated by 500 MeV protons at KEK, reaction rate distributions inside a thick concrete shield irradiated by 6.2 GeV protons at LBL, and neutron and hadron fluxes inside an iron beam stop irradiated by 24 GeV protons at CERN are compiled in this document. Calculational configurations and neutron reaction cross section data up to 500 MeV are provided. (author)

Resolvent-Techniques for Multiple Exercise Problems

International Nuclear Information System (INIS)

Christensen, Sören; Lempa, Jukka

2015-01-01

We study optimal multiple stopping of strong Markov processes with random refraction periods. The refraction periods are assumed to be exponentially distributed with a common rate and independent of the underlying dynamics. Our main tool is using the resolvent operator. In the first part, we reduce infinite stopping problems to ordinary ones in a general strong Markov setting. This leads to explicit solutions for wide classes of such problems. Starting from this result, we analyze problems with finitely many exercise rights and explain solution methods for some classes of problems with underlying Lévy and diffusion processes, where the optimal characteristics of the problems can be identified more explicitly. We illustrate the main results with explicit examples

Resolvent-Techniques for Multiple Exercise Problems

Energy Technology Data Exchange (ETDEWEB)

Christensen, Sören, E-mail: christensen@math.uni-kiel.de [Christian–Albrechts-University in Kiel, Mathematical Institute (Germany); Lempa, Jukka, E-mail: jukka.lempa@hioa.no [Oslo and Akershus University College, School of business, Faculty of Social Sciences (Norway)

2015-02-15

We study optimal multiple stopping of strong Markov processes with random refraction periods. The refraction periods are assumed to be exponentially distributed with a common rate and independent of the underlying dynamics. Our main tool is using the resolvent operator. In the first part, we reduce infinite stopping problems to ordinary ones in a general strong Markov setting. This leads to explicit solutions for wide classes of such problems. Starting from this result, we analyze problems with finitely many exercise rights and explain solution methods for some classes of problems with underlying Lévy and diffusion processes, where the optimal characteristics of the problems can be identified more explicitly. We illustrate the main results with explicit examples.

Determination of the hexagonal network parameters of the quartz β using neutron multiple diffraction

International Nuclear Information System (INIS)

Campos, L.C.; Parente, C.B.R.; Mazzocchi, V.L.; Helene, O.

2000-01-01

In this work, neutron multiple diffraction is employed for the determination of the parameters a and c of the β-quartz hexagonal cell. This crystalline phase of silica (SiO 2 ) occurs in temperatures between ca. 846 and 1143 K. A β-quartz neutron multiple diffraction pattern has been used in the determinations. This pattern was obtained with a natural quartz single crystal heated to 1003 K. During the indexing of the pattern it was verified that most of the pairs of secondary reflections, which are responsible for the formation of peaks, could be classified as 'good for the determination of a' or 'good for the determination of c'. With this classification, it became possible to employ an iterative method for the determination of both parameters. After 8 cycles of iteration the values found for the parameters were a = 4.9964 +- 0.0018 and c = 5.46268 +- 0.00052 A. (author)

Nuclear data for neutron emission in the fission process

International Nuclear Information System (INIS)

Ganesan, S.

1991-11-01

This document contains the proceedings of the IAEA Consultants' Meeting on Nuclear Data for Neutron Emission in the Fission Process, Vienna, 22 - 24 October 1990. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers provide a review of the status of experimental and theoretical data on neutron emission in spontaneous and neutron induced fission with reference to the data needs for reactor applications oriented towards actinide burner studies. The specific topics covered are the following: experimental measurements and theoretical predictions and evaluations of fission neutron energy spectra, average prompt fission neutron multiplicity, correlation in neutron emission from complementary fragments, neutron emission during acceleration of fission fragments, statistical properties of neutron rich nuclei by study of emission spectra of neutrons from the excited fission fragments, integral qualification of nu-bar for the major fissile isotopes, nu-bar total of 239 Pu and 235 U, and related problems. Refs figs and tabs

Fast rigorous numerical method for the solution of the anisotropic neutron transport problem and the NITRAN system for fusion neutronics application. Pt. 1

International Nuclear Information System (INIS)

Takahashi, A.; Rusch, D.

1979-07-01

Some recent neutronics experiments for fusion reactor blankets show that the precise treatment of anisotropic secondary emissions for all types of neutron scattering is needed for neutron transport calculations. In the present work new rigorous methods, i.e. based on non-approximative microscopic neutron balance equations, are applied to treat the anisotropic collision source term in transport equations. The collision source calculation is free from approximations except for the discretization of energy, angle and space variables and includes the rigorous treatment of nonelastic collisions, as far as nuclear data are given. Two methods are presented: first the Ii-method, which relies on existing nuclear data files and then, as an ultimate goal, the I*-method, which aims at the use of future double-differential cross section data, but which is also applicable to the present single-differential data basis to allow a smooth transition to the new data type. An application of the Ii-method is given in the code system NITRAN which employs the Ssub(N)-method to solve the transport equations. Both rigorous methods, the Ii- and the I*-method, are applicable to all radiation transport problems and they can be used also in the Monte-Carlo-method to solve the transport problem. (orig./RW) [de

Detection of low caloric power of coal by pulse fast-thermal neutron analysis

International Nuclear Information System (INIS)

Gu De-shan; Sang Hai-feng; Qiao Shuang; Liu Yu-ren, Liu Lin-mao; Jing Shi-wei; Chinese Academy of Sciences, Changchun

2004-01-01

Analysis method and principle of pulse fast-thermal neutron analysis (PFTNA) are introduced. A system for the measurement of low caloric power of coal by PFTNA is also presented. The 14 MeV pulse neutron generator and BGO detector and 4096 MCA were applied in this system. A multiple linear regression method applied to the data solved the interferential problem of multiple elements. The error of low caloric power between chemical analysis and experiment was less than 0.4 MJ/kg. (author)

Use of probability tables for propagating uncertainties in neutronics

International Nuclear Information System (INIS)

Coste-Delclaux, M.; Diop, C.M.; Lahaye, S.

2017-01-01

Highlights: • Moment-based probability table formalism is described. • Representation by probability tables of any uncertainty distribution is established. • Multiband equations for two kinds of uncertainty propagation problems are solved. • Numerical examples are provided and validated against Monte Carlo simulations. - Abstract: Probability tables are a generic tool that allows representing any random variable whose probability density function is known. In the field of nuclear reactor physics, this tool is currently used to represent the variation of cross-sections versus energy (neutron transport codes TRIPOLI4®, MCNP, APOLLO2, APOLLO3®, ECCO/ERANOS…). In the present article we show how we can propagate uncertainties, thanks to a probability table representation, through two simple physical problems: an eigenvalue problem (neutron multiplication factor) and a depletion problem.

Measurements of {sup 237}Np secondary neutron spectra

Energy Technology Data Exchange (ETDEWEB)

Kornilov, N.V.

1997-03-01

The activities carried out during the first year of the project are summarized. The main problems for Np spectra measurements arise from high intrinsic gamma-ray activity of the sample and admixture of the oxygen and iron nuclei. The inelastically scattered neutrons and the fission neutrons spectra for {sup 237}Np were measured by time-of-flight spectrometer of the IPPE at incident neutron energies {approx_equal}1.5 MeV, and {approx_equal}0.5 MeV. A solid tritium target and a Li-metallic target were used as neutron sources. The neutron scattering on C sample (C(n,n) standard reaction) was measured to normalize the Np data. The experimental data should be simulated by Monte Carlo method to correct the experimental data for oxygen and iron admixture as well as for multiple scattering of the neutrons in the sample. Therefore the response function of the spectrometer, and the neutron energy distribution from the source were investigated in detail. (author)

Polarized neutrons

International Nuclear Information System (INIS)

Williams, W.G.

1988-01-01

The book on 'polarized neutrons' is intended to inform researchers in condensed matter physics and chemistry of the diversity of scientific problems that can be investigated using polarized neutron beams. The contents include chapters on:- neutron polarizers and instrumentation, polarized neutron scattering, neutron polarization analysis experiments and precessing neutron polarization. (U.K.)

The numerical analysis of eigenvalue problem solutions in multigroup neutron diffusion theory

International Nuclear Information System (INIS)

Woznicki, Z.I.

1995-01-01

The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iterations within global iterations. Particular iterative strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 35 figs, 16 tabs

Aborption and activation techniques in measurements of fast-neutron capture cross sections

International Nuclear Information System (INIS)

Bergqvist, I.

1982-01-01

The absorption and activation methods have been applied for a long time to systematic studies of fast neutron capture cross sections. Both methods are simple in principle but difficult in practice. The simplicity should ensure a wider use of the methods in particular for problems which may be complicated to approach with other methods The difficulties encountered in absorption measurements are related to multiple scattering and resonance shielding effects. In activation experiments the influence of secondary low-energy neutrons c causes the main problems. (Author)

Problem-solving with multiple interdependent criteria: better solution to complex problems

International Nuclear Information System (INIS)

Carlsson, C.; Fuller, R.

1996-01-01

We consider multiple objective programming (MOP) problems with additive interdependencies, this is when the states of some chosen objective are attained through supportive or inhibitory feed-backs from several other objectives. MOP problems with independent objectives (when the cause-effect relations between the decision variables and the objectives are completely known) will be treated as special cases of the MOP in which we have interdependent objectives. We illustrate our ideas by a simple three-objective real-life problem

On multiple level-set regularization methods for inverse problems

International Nuclear Information System (INIS)

DeCezaro, A; Leitão, A; Tai, X-C

2009-01-01

We analyze a multiple level-set method for solving inverse problems with piecewise constant solutions. This method corresponds to an iterated Tikhonov method for a particular Tikhonov functional G α based on TV–H 1 penalization. We define generalized minimizers for our Tikhonov functional and establish an existence result. Moreover, we prove convergence and stability results of the proposed Tikhonov method. A multiple level-set algorithm is derived from the first-order optimality conditions for the Tikhonov functional G α , similarly as the iterated Tikhonov method. The proposed multiple level-set method is tested on an inverse potential problem. Numerical experiments show that the method is able to recover multiple objects as well as multiple contrast levels

Gamma spectroscopic studies of the neutron-deficient g-g nucleus 74Kr by means of a neutron multiplicity measurement technique

International Nuclear Information System (INIS)

Roth, J.

1981-01-01

The g-g nucleus 74 Kr was studied by means of the reaction 58 Ni ( 19 F, p2n#betta#) 74 Kr. In order to make gamma spectroscopic studies at neutron deficient nuclei like 74 Kr a neutron multiplicity measurement technique was developed. Beside #betta# single spectra, #betta# excitation functions, #betta#-#betta# coincidences, #betta# angular distributions, and lifetime measurements by means of this technique all measurements in coincidence with up to two neutrons were taken up. From these measurement data an extended term scheme with 17 newly found excited states could be extracted. To all levels spins and parities could be assigned. From the four energetically lowest levels of the yrast cascade the mean lifetimes could be determined. A double backbending in the sequence of the yrast cascade was interpreted as crossing of the g 9/2 bands. The irregularities in the lower part of the yrast band correspond to the shape consistence picture. The results were considered in connection with the systematics of the even krypton isotopes and compared with a two-quasiparticle-plas-rotor model calculation. (HSI)

Software tool for resolution of inverse problems using artificial intelligence techniques: an application in neutron spectrometry

International Nuclear Information System (INIS)

Castaneda M, V. H.; Martinez B, M. R.; Solis S, L. O.; Castaneda M, R.; Leon P, A. A.; Hernandez P, C. F.; Espinoza G, J. G.; Ortiz R, J. M.; Vega C, H. R.; Mendez, R.; Gallego, E.; Sousa L, M. A.

2016-10-01

The Taguchi methodology has proved to be highly efficient to solve inverse problems, in which the values of some parameters of the model must be obtained from the observed data. There are intrinsic mathematical characteristics that make a problem known as inverse. Inverse problems appear in many branches of science, engineering and mathematics. To solve this type of problem, researches have used different techniques. Recently, the use of techniques based on Artificial Intelligence technology is being explored by researches. This paper presents the use of a software tool based on artificial neural networks of generalized regression in the solution of inverse problems with application in high energy physics, specifically in the solution of the problem of neutron spectrometry. To solve this problem we use a software tool developed in the Mat Lab programming environment, which employs a friendly user interface, intuitive and easy to use for the user. This computational tool solves the inverse problem involved in the reconstruction of the neutron spectrum based on measurements made with a Bonner spheres spectrometric system. Introducing this information, the neural network is able to reconstruct the neutron spectrum with high performance and generalization capability. The tool allows that the end user does not require great training or technical knowledge in development and/or use of software, so it facilitates the use of the program for the resolution of inverse problems that are in several areas of knowledge. The techniques of Artificial Intelligence present singular veracity to solve inverse problems, given the characteristics of artificial neural networks and their network topology, therefore, the tool developed has been very useful, since the results generated by the Artificial Neural Network require few time in comparison to other techniques and are correct results comparing them with the actual data of the experiment. (Author)

On efficiently computing multigroup multi-layer neutron reflection and transmission conditions

International Nuclear Information System (INIS)

Abreu, Marcos P. de

2007-01-01

In this article, we present an algorithm for efficient computation of multigroup discrete ordinates neutron reflection and transmission conditions, which replace a multi-layered boundary region in neutron multiplication eigenvalue computations with no spatial truncation error. In contrast to the independent layer-by-layer algorithm considered thus far in our computations, the algorithm here is based on an inductive approach developed by the present author for deriving neutron reflection and transmission conditions for a nonactive boundary region with an arbitrary number of arbitrarily thick layers. With this new algorithm, we were able to increase significantly the computational efficiency of our spectral diamond-spectral Green's function method for solving multigroup neutron multiplication eigenvalue problems with multi-layered boundary regions. We provide comparative results for a two-group reactor core model to illustrate the increased efficiency of our spectral method, and we conclude this article with a number of general remarks. (author)

Development of the hierarchical domain decomposition boundary element method for solving the three-dimensional multiregion neutron diffusion equations

International Nuclear Information System (INIS)

Chiba, Gou; Tsuji, Masashi; Shimazu, Yoichiro

2001-01-01

A hierarchical domain decomposition boundary element method (HDD-BEM) that was developed to solve a two-dimensional neutron diffusion equation has been modified to deal with three-dimensional problems. In the HDD-BEM, the domain is decomposed into homogeneous regions. The boundary conditions on the common inner boundaries between decomposed regions and the neutron multiplication factor are initially assumed. With these assumptions, the neutron diffusion equations defined in decomposed homogeneous regions can be solved respectively by applying the boundary element method. This part corresponds to the 'lower level' calculations. At the 'higher level' calculations, the assumed values, the inner boundary conditions and the neutron multiplication factor, are modified so as to satisfy the continuity conditions for the neutron flux and the neutron currents on the inner boundaries. These procedures of the lower and higher levels are executed alternately and iteratively until the continuity conditions are satisfied within a convergence tolerance. With the hierarchical domain decomposition, it is possible to deal with problems composing a large number of regions, something that has been difficult with the conventional BEM. In this paper, it is showed that a three-dimensional problem even with 722 regions can be solved with a fine accuracy and an acceptable computation time. (author)

The Multiple Pendulum Problem via Maple[R

Science.gov (United States)

Salisbury, K. L.; Knight, D. G.

2002-01-01

The way in which computer algebra systems, such as Maple, have made the study of physical problems of some considerable complexity accessible to mathematicians and scientists with modest computational skills is illustrated by solving the multiple pendulum problem. A solution is obtained for four pendulums with no restriction on the size of the…

The 4π neutron detector CARMEN

Energy Technology Data Exchange (ETDEWEB)

Ledoux, X., E-mail: Xavier.ledoux@ganil.fr [CEA/DAM/DIF, F-91297 Arpajon (France); GANIL, CEA/DRF-CNRS/IN2P3, Caen, F-14076 France (France); Laborie, J.-M.; Pras, P.; Lantuéjoul-Thfoin, I.; Varignon, C. [CEA/DAM/DIF, F-91297 Arpajon (France)

2017-02-01

CARMEN is a 4π neutron detector filled with a gadolinium-loaded liquid scintillator built to measure neutron multiplicity distributions. It is used to study fission and (n,xn) reactions. In addition to neutron multiplicity measurements, CARMEN can be used to measure neutron energy spectra with the time-of-flight technique, thanks to the time properties of the prompt signal. The detector, detection technique and efficiency determination are presented in detail. Two examples are also presented: the measurement of {sup 252}Cf spontaneous fission neutron multiplicity probability distribution and the measurement of the neutron energy spectrum emitted by an Am-Be radioactive source.

Entrance channel systematics of pre-scission neutron multiplicities

International Nuclear Information System (INIS)

Shareef, M.; Prasad, E.; Chatterjee, A.

2016-01-01

Statistical model analysis has been performed for the available neutron multiplicity (ν_p_r_e) data in the literature. Larger ν_p_r_e values for more symmetric reactions have been observed in comparison with asymmetric reactions forming the same compound nucleus, in most cases. A reverse trend has also been noticed in a few cases. A systematic entrance channel dependence of fission timescale is brought out in this work. Fission timescales calculated using the experimental ν_p_r_e values fall into two distinct groups according to the entrance channel mass asymmetry of the reaction with respect to the Businaro-Gallone critical mass asymmetry. The difference in the delay between these two groups ranges between 20 and 100 zs, which is larger than that reported in some cases. (orig.)

Nuclear fusion and neutron processes

International Nuclear Information System (INIS)

Orlov, V.V.; Shatalov, G.E.; Sherstnev, K.E.

1984-01-01

Problems of providing development of the design of an experimental fusion reactor with necessary neutron-physical data are discussed. Isotope composition of spent fuel in the blanket of a hybride fusion reactor (HFR) is given. Neutron balance of the reactor with Li-blanket and neutron balance of the reactor with Pb-multiplier are disclosed. A simplified scheme of neutron and energy balance in the HFR blanket is given. Development and construction of the experimental power reactor is shown to become the nearest problem of the UTS program. Alongside with other complex physical and technical problems solution of this problem requires realization of a wide program of neutron-physical investigations including measurements with required accuracy of neutron cross sections, development of methodical, program and constant basis of neutron calculations and macroscopic experiments on neutron sources

Detector point of view of reactor internal vibrations under Gaussian coloured random forces - the problem of fitting neutron noise experimental data

International Nuclear Information System (INIS)

Arnal, R.S.; Martin, G.V.; Gonzalez, J.L.M.-C.

1988-01-01

This paper studies the local vibrations of reactor components driven by Gaussian coloured and white forces, when nonlinear vibrations arise. We study also the important problem of noise sources, modelization and the noise propagation through the neutron field using the discrete ordinates transport theory. Finally, we study the effect of the neutron field upon the PSD (power spectral density) of the noise source and we analyse the problem of fitting neutron noise experimental data to perform pattern recognition analysis. (author)

Performance characteristics of specified power reactors in multidimensional neutron diffusion problems

International Nuclear Information System (INIS)

Kim, M.G.

1980-01-01

The multigroup neutron diffusion equations with the constraint of specified power distributions are investigated by the application of the straight-line method which can be considered as the limiting case of zero mesh space in the finite difference method. The standard partial differential form of the diffusion equation is reduced to sets of ordinary differential equations and then converted into sets of integral equations by using Green's functions defined on the pseudo straight lines. Coupling of each straight line to the adjacent lines arises from the application of a three-point central difference formula. The interfaces encountered between two regions are taken into account by imposing the continuity conditions for the grown fluxes and net currents with Taylor expansions of internal fluxes at the interface positions. A few sample problems are selected to test the validity of the method. It is found that the proposed method of solution is similar to the finite Fourier sine transform. Numerical results for the solutions obtained by the method of straight lines are compared with the results of the exact analytical solutions for simple geometries. These comparisons indicate that the proposed method is compatible with the analytical method, and in some problems considered the straight-line solutions are much more efficient than the exact solutions. The method is also extended to the reactor kinetics problem by expressing the kinetics parameters in terms of the basis functions which are used to obtain the solutions of the steady-state neutron diffusion equations

The searchlight problem for neutrons in a semi-infinite medium

International Nuclear Information System (INIS)

Ganapol, B.D.

1993-01-01

The solution of the Search Light Problem for monoenergetic neutrons in a semi-infinite medium with isotropic scattering illuminated at the free surface is obtained by several methods at various planes within the medium. The sources considered are a normally-incident pencil beam and an isotropic point source. The analytic solution is effected by a recently developed numerical inversion technique applied to the Fourier-Bessel transform. This transform inversion results from the solution method of Rybicki, where the two-dimensional problem is solved by casting it as a variant of a one-dimensional problem. The numerical inversion process results in a highly accurate solution. Comparisons of the analytic solution with results from Monte Carlo (MCNP) and discrete ordinates transport (DORT) codes show excellent agreement. These comparisons, which are free of any associated data or cross section set dependencies, provide significant evidence of the proper operation of both the transport codes tested

Experimental evaluation of the extended Dytlewski-style dead time correction formalism for neutron multiplicity counting

Science.gov (United States)

Lockhart, M.; Henzlova, D.; Croft, S.; Cutler, T.; Favalli, A.; McGahee, Ch.; Parker, R.

2018-01-01

Over the past few decades, neutron multiplicity counting has played an integral role in Special Nuclear Material (SNM) characterization pertaining to nuclear safeguards. Current neutron multiplicity analysis techniques use singles, doubles, and triples count rates because a methodology to extract and dead time correct higher order count rates (i.e. quads and pents) was not fully developed. This limitation is overcome by the recent extension of a popular dead time correction method developed by Dytlewski. This extended dead time correction algorithm, named Dytlewski-Croft-Favalli(DCF), is detailed in reference Croft and Favalli (2017), which gives an extensive explanation of the theory and implications of this new development. Dead time corrected results can then be used to assay SNM by inverting a set of extended point model equations which as well have only recently been formulated. The current paper discusses and presents the experimental evaluation of practical feasibility of the DCF dead time correction algorithm to demonstrate its performance and applicability in nuclear safeguards applications. In order to test the validity and effectiveness of the dead time correction for quads and pents, 252Cf and SNM sources were measured in high efficiency neutron multiplicity counters at the Los Alamos National Laboratory (LANL) and the count rates were extracted up to the fifth order and corrected for dead time. In order to assess the DCF dead time correction, the corrected data is compared to traditional dead time correction treatment within INCC. The DCF dead time correction is found to provide adequate dead time treatment for broad range of count rates available in practical applications.

Is there an Ay problem in low-energy neutron-proton scattering?

International Nuclear Information System (INIS)

Gross, Franz; Stadler, Alfred

2008-01-01

We calculate Ay in neutron-proton scattering for the interactions models WJC-1 and WJC-2 in the Covariant Spectator Theory. We find that the recent 12 MeV measurements performed at TUNL are in better agreement with our results than with the Nijmegen Phase Shift Analysis of 1993, and after reviewing the low energy data, conclude that there is no Ay problem in low-energy np scattering.

The numerical analysis of eigenvalue problem solutions in the multigroup neutron diffusion theory

International Nuclear Information System (INIS)

Woznicki, Z.I.

1994-01-01

The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iteration within global iterations. Particular interactive strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 32 figs, 15 tabs

The numerical analysis of eigenvalue problem solutions in the multigroup neutron diffusion theory

Energy Technology Data Exchange (ETDEWEB)

Woznicki, Z I [Institute of Atomic Energy, Otwock-Swierk (Poland)

1994-12-31

The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iteration within global iterations. Particular interactive strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 32 figs, 15 tabs.

The Development of Advanced Processing and Analysis Algorithms for Improved Neutron Multiplicity Measurements

International Nuclear Information System (INIS)

Santi, P.; Favalli, A.; Hauck, D.; Henzl, V.; Henzlova, D.; Ianakiev, K.; Iliev, M.; Swinhoe, M.; Croft, S.; Worrall, L.

2015-01-01

One of the most distinctive and informative signatures of special nuclear materials is the emission of correlated neutrons from either spontaneous or induced fission. Because the emission of correlated neutrons is a unique and unmistakable signature of nuclear materials, the ability to effectively detect, process, and analyze these emissions will continue to play a vital role in the non-proliferation, safeguards, and security missions. While currently deployed neutron measurement techniques based on 3He proportional counter technology, such as neutron coincidence and multiplicity counters currently used by the International Atomic Energy Agency, have proven to be effective over the past several decades for a wide range of measurement needs, a number of technical and practical limitations exist in continuing to apply this technique to future measurement needs. In many cases, those limitations exist within the algorithms that are used to process and analyze the detected signals from these counters that were initially developed approximately 20 years ago based on the technology and computing power that was available at that time. Over the past three years, an effort has been undertaken to address the general shortcomings in these algorithms by developing new algorithms that are based on fundamental physics principles that should lead to the development of more sensitive neutron non-destructive assay instrumentation. Through this effort, a number of advancements have been made in correcting incoming data for electronic dead time, connecting the two main types of analysis techniques used to quantify the data (Shift register analysis and Feynman variance to mean analysis), and in the underlying physical model, known as the point model, that is used to interpret the data in terms of the characteristic properties of the item being measured. The current status of the testing and evaluation of these advancements in correlated neutron analysis techniques will be discussed

The problem of the black plate with zero thickness and finite width in neutron transport theory

International Nuclear Information System (INIS)

Benoist, Pierre.

1979-08-01

A black plate with zero thickness, finite width and infinite height, imbedded in an infinite and homogeneous medium which scatters and absorbs neutrons, is considered. The problem is time-independent and the neutrons, which are supposed to have a unique speed, are issued, either from a current at infinity (problem A), or from a uniform source (problem B). It is shown that the Csub(N) method seems to be particularly well suited to the resolution of this 'two-dimensional Milne problem'. A particular interest is attached to the determination of the radius R of the black cylinder leading to the same polar behaviour of the flux at infinity as the plate (criterion 1), or absorbing the same number of neutrons as the plate (criterion 2). In this preliminary report, values of R are calculated in various limit cases: the width of the plate being taken equal to one, l being the mean free path and c the number of secondaries par collision in the outer medium, R is calculated at first in the limit l → 0 (for c = 1) by the theory of Musklelishvili, and then in the limit l → infinity (whatever c is) and c → 0 (whatever l is). In the limit c → 1 (whatever l is), R is shown to be the same in problems A and B and criteria 1 and 2. On the other hand, whatever l and c are; the values of R obtained in the problem A with the criterion 2 and in the problem B with the criterion 1 are shown to be equal. All these results allow henceforth a reasonable interpolation which can be useful in the practice [fr

Environmental protection problems from the standpoint of regeneration of fast neutron reactor fuel

International Nuclear Information System (INIS)

Gedeonov, L.I.; Lazarev, L.N.; Suprunenko, A.N.

The discussion of the problem of environmental protection is based on two principles: a strict observance of legislatively established standards for permissible concentrations of radionuclides in objects of the environment and for dose loads for the population; all possible steps to reduce the contamination to a level justified in practice. Environmental protection steps are considered from the points of view of a systematic analysis. A survey of the environmental protection system near sources of radioactive discharges is given. The basic interactions and feedbacks are indicated. Characteristics differentiating the discharges of the fuel cycle of fast neutron breeder reactors from discharges of the slow neutron cycle are discussed. It is shown that it is necessary to study the overall regional and global interactions of discharges of the atomic power industry. The characteristics of situations at nuclear fuel cycle facilities of fast neutron reactors are discussed. The necessity of additional technical steps to prevent accidents and eliminate their effects if they take place is emphasized

Application of the modified neutron source multiplication method for a measurement of sub-criticality in AGN-201K reactor

International Nuclear Information System (INIS)

Myung-Hyun Kim

2010-01-01

Measurement of sub-criticality is a challenging and required task in nuclear industry both for nuclear criticality safety and physics test in nuclear power plant. A relatively new method named as Modified Neutron Source Multiplication Method (MNSM) was proposed in Japan. This method is an improvement of traditional Neutron Source Multiplication (NSM) Method, in which three correction factors are applied additionally. In this study, MNSM was tested in calculation of rod worth using an educational reactor in Kyung Hee University, AGN-201K. For this study, a revised nuclear data library and a neutron transport code system TRANSX-PARTISN were used for the calculation of correction factors for various control rod positions and source locations. Experiments were designed and performed to enhance errors in NSM from the location effects of source and detectors. MNSM can correct these effects but current results showed not much correction effects. (author)

Study of the α and β phases of quartz by neutron multiple diffraction

International Nuclear Information System (INIS)

Mazzocchi, V.L.

1984-01-01

Crystal structures of α and β phases of quartz are studied, employing neutron multiple diffraction as a method of analysis. Theoretical multiple diffraction patterns in a many-beam case were determined by a computer program which calculates intensities of beams as sums of Taylor's series expansions, retaining terms up to a order n. Experimental 'umweg' and transmitted beam patterns were obtained for the 00.1 primary reflection of α and β phases of quartz. To calculate α - quartz multiple diffraction intensities it was necessary to determine the Dauphine twinning fraction for the crystal after having passed by the β-phase. For the two models of β-quartz a better agreement between experimental and calculated integrated intensities was found for the disordered structure model based on split-half-oxigen positions. (Author) [pt

Parallel preconditioned conjugate gradient algorithm applied to neutron diffusion problem

International Nuclear Information System (INIS)

Majumdar, A.; Martin, W.R.

1992-01-01

Numerical solution of the neutron diffusion problem requires solving a linear system of equations such as Ax = b, where A is an n x n symmetric positive definite (SPD) matrix; x and b are vectors with n components. The preconditioned conjugate gradient (PCG) algorithm is an efficient iterative method for solving such a linear system of equations. In this paper, the authors describe the implementation of a parallel PCG algorithm on a shared memory machine (BBN TC2000) and on a distributed workstation (IBM RS6000) environment created by the parallel virtual machine parallelization software

Multiple scattering effects in fast neutron polarization experiments using high-pressure helium-xenon gas scintillators as analyzers

International Nuclear Information System (INIS)

Tornow, W.; Mertens, G.

1977-01-01

In order to study multiple scattering effects both in the gas and particularly in the solid materials of high-pressure gas scintillators, two asymmetry experiments have been performed by scattering of 15.6 MeV polarized neutrons from helium contained in stainless steel vessels of different wall thicknesses. A monte Carlo computer code taking into account the polarization dependence of the differential scattering cross sections has been written to simulate the experiments and to calculate corrections for multiple scattering on helium, xenon and the gas containment materials. Besides the asymmetries for the various scattering processes involved, the code yields time-of-flight spectra of the scattered neutrons and pulse height spectra of the helium recoil nuclei in the gas scintillator. The agreement between experimental results and Monte Carlo calculations is satisfactory. (Auth.)

Neutronics study on hybrid reactor cooled by helium, water and molten salt

International Nuclear Information System (INIS)

Li Zaixin; Feng Kaiming; Zhang Guoshu; Zheng Guoyao; Zhao Fengchao

2009-01-01

There is no serious magnetohydrodynamics (MHD) problem when helium,water or molten salt of Flibe flows in high magnetic field. Thus helium, water and Flibe were proposed as candidate of coolant for fusion-fission hybrid reactor based on magnetic confinement. The effect on neutronics of hybrid reactor due to coolant was investigated. The analyses of neutron spectra and fuel breeding of blanket with different coolants were performed. Variations of tritium breeding ratio (TBR), blanket energy multiplication (M) and keff with operating time were also studied. MCNP code was used for neutron transport simulation. It is shown that spectra change greatly with different coolants. The blanket with helium exhibits very hard spectrum and good tritium breeding ability. And fission reactions are mainly from fast neutron. The blanket with water has soft spectrum and high energy multiplication factor. However, it needs to improve TBR. The blanket with Flibe has hard spectrum and less energy release. (authors)

Neutron lifetime measurements with a large gravitational trap for ultracold neutrons

Science.gov (United States)

Serebrov, A. P.; Kolomensky, E. A.; Fomin, A. K.; Krasnoshchekova, I. A.; Vassiljev, A. V.; Prudnikov, D. M.; Shoka, I. V.; Chechkin, A. V.; Chaikovskiy, M. E.; Varlamov, V. E.; Ivanov, S. N.; Pirozhkov, A. N.; Geltenbort, P.; Zimmer, O.; Jenke, T.; Van der Grinten, M.; Tucker, M.

2018-05-01

Neutron lifetime is one of the most important physical constants: it determines parameters of the weak interaction and predictions of primordial nucleosynthesis theory. There remains the unsolved problem of a 3.9σ discrepancy between measurements of this lifetime using neutrons in beams and those with stored ultracold neutrons (UCN). In our experiment we measure the lifetime of neutrons trapped by Earth's gravity in an open-topped vessel. Two configurations of the trap geometry are used to change the mean frequency of UCN collisions with the surfaces; this is achieved by plunging an additional surface into the trap without breaking the vacuum. The trap walls are coated with a hydrogen-less fluorine-containing polymer to reduce losses of UCN. The stability of this coating over multiple thermal cycles between 80 and 300 K was tested. At 80 K, the probability of UCN loss due to collisions with the trap walls is just 1.5% of the probability of β decay. The free neutron lifetime is determined by extrapolation to an infinitely large trap with zero collision frequency. The result of these measurements is τn=881.5 ±0 .7stat ±0 .6syst s which is consistent with the conventional value of 880.2 ± 1.0 s presented by the Particle Data Group. Future prospects for this experiment are in further cooling to 10 K, which will lead to an improved accuracy of measurement. In conclusion we present an analysis of currently available data on various measurements of the neutron lifetime.

Study on uranium-water multiplicative means of the (RESUCO-Subcritical experimental reactor of uranium with oxygen) subcritical assembly by pulsed neutron technique

International Nuclear Information System (INIS)

Jesus Barbosa, S. de.

1987-01-01

The effective multiplication factor and the nuclear parameters associated with the variation of (RESUCO- Subcritical Experimental Reactor of Uranium with Oxygen) Subcritical Assembly Configuration, using pulsed neutron technique are analysed. BF3 detectors were used to detect the variation of thermal neutrons in the system, positioned parallelly to fuel elements, and a proton recoil detector was used for monitoring the neutron generation. (M.C.K.) [pt

Sensitivity studies of the neutron multiplicity spectrum in the spallation of Pb targets

International Nuclear Information System (INIS)

Sinha, A.; Garg, S.B.; Srinivasan, M.

1986-01-01

The number of neutrons produced per incident proton in the spallation of Pb targets is of direct relevance to the design of accelerator breeders. The nuclear cascade initiated by high-energy protons in spallation targets is usually described by an intranuclear cascade evaporation (INCE) model. Even though this model describes various average nuclear properties of spallation targets fairly well, differential quantities such as energy spectra, angular spectra etc., are not reproduced within the limits of experimental uncertainty. One of the reasons for this is the uncertainty in the magnitude of the parameters involved in the model, notably the level density parameter Bsub(O) whose magnitude is quoted by different workers to be in the range of 8-20 MeV. The accuracy of Bsub(O) could be improved if we could experimentally determine a quantity which is much more sensitive to Bsub(O) than the average neutron yield. In this paper we discuss one such quantity, namely the neutron multiplicity spectrum (MS). We compute the MS due to the spallation of Pb targets of different sizes at proton energies of 1.5, 1.0 and 0.59 GeV using the Monte Carlo code HETC. It is noticed that for the 1.5 GeV proton case the probability P(ν) for leakage of ν neutrons for ν in the range of 60-65, changes by about 70% when Bsub(O) is varied from 8 to 20 MeV. The corresponding change in the average neutron yield is <20%. It is therefore suggested that an accurate measurement of the MS can serve as a useful tool to narrow down the range of uncertainty in the Bsub(O) parameter. (author)

A study of the 208Pb + 197Au reaction at 29 MeV/u through the associated neutron multiplicity

International Nuclear Information System (INIS)

Bresson, S.

1993-01-01

The investigation of this heavy symmetric system has been carried out through the study of the associated neutron multiplicity. The experimental techniques and data processing are first described, with emphasis on the Orion neutron detector and the hodoscope used to detect the charged reaction products at forward angles. It is shown that the neutron multiplicity is a good measure of the violence of the collision and a good way to characterize the different modes of the reaction. The fission of the quasi-projectile is then characterized and is shown to occur for peripheral collisions. Using simulations, the minimal values of the angular momentum transferred to the quasi-projectile are determined. The results of dynamical calculations using the Landau Vlasov equation are described, which show the importance of angular momentum. It is demonstrated that, at 29 MeV/u, the Pb + Au collision is still governed by deep inelastic reactions in which angular momentum in the exit channel plays an important role

Application of neutron backscatter techniques to level measurement problems

International Nuclear Information System (INIS)

Leonardi-Cattolica, A.M.; McMillan, D.H.; Telfer, A.; Griffin, L.H.; Hunt, R.H.

1982-01-01

We have designed and built portable level detectors and fixed level monitors based on neutron scattering and detection principles. The main components of these devices, which we call neutron backscatter gauges, are a neutron emitting radioisotope, a neutron detector, and a ratemeter. The gauge is a good detector for hydrogen but is much less sensitive to most other materials. This allows level measurements of hydrogen bearing materials, such as hydrocarbons, to be made through the walls of metal vessels. Measurements can be made conveniently through steel walls which are a few inches thick. We have used neutron backscatter gauges in a wide variety of level measurement applications encountered in the petrochemical industry. In a number of cases, the neutron techniques have proven to be superior to conventional level measurement methods, including gamma ray methods

Polycapillary neutron lenses

International Nuclear Information System (INIS)

Mildner, D.F.R.

1997-01-01

The principle of multiple mirror reflection from smooth surfaces at small grazing angles enables the transport and guiding of high intensity slow neutron beams to locations of low background for neutron scattering and absorption experiments and to provide facilities for multiple instruments. Curved guides have been widely used at cold neutron facilities to remove the unwanted radiation (fast neutrons and gamma rays) from the beam without the use of filters. A typical guide has transverse dimensions of 50 mm and, with a radius of curvature of 1 km, transmits wavelengths longer than 5 A. Much tighter curves requires narrower transverse dimensions, otherwise there is little transmission. Typical neutron benders have a number of slots with transverse dimensions of ∼5 mm. Based on the same principle but using a different technology, recent developments in glass polycapillary fibers have produced miniature versions of neutron guides. Fibers with many thousands of channels having sizes of ∼ 10 μm enable beams of long wavelength neutrons (λ > 4 A) to be transmitted efficiently in a radius of curvature as small as a fraction of 1 m. A large collection of these miniature versions of neutron guides can be used to bend the neutron trajectories such that the incident beam can be focused. (author)

Discovery of the neutron (to the fiftieth anniversary of neutron discovery)

International Nuclear Information System (INIS)

Pasechnik, M.V.

1984-01-01

Development of neutron physics in the USSR for the recent 50 years from the moment of neutron discovery is considered. History of neutron discovery is presented in brief. Neutron properties and fundamental problems of physics: electric dipole neutron moment, neutron β-decay, neutron interaction with nuclei and potential of nucleon interaction not conserving spatial parity are discussed. Main aspects of neutron physics application in power engineering, nuclear technology and other branches of science and technique are set forth

Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

Energy Technology Data Exchange (ETDEWEB)

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

2005-09-15

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

Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

International Nuclear Information System (INIS)

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

2005-09-01

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

Effect of N/Z in pre-scission neutron multiplicity for 16,18O+198Pt systems

International Nuclear Information System (INIS)

Sandal, R.; Behera, B.R.; Singh, V.; Kaur, M.; Kumar, A.; Singh, G.; Singh, K.P.; Sugathan, P.; Jhingan, A.; Golda, K.S.; Chatterjee, M.B.; Bhowmik, R.K.; Kalkal, S.; Siwal, D.; Goyel, S.; Mandal, S.; Prasad, E.; Sadhukhan, J.; Pal, S.; Mahta, K.; Saxena, A.

2014-01-01

This paper reports the summary of the experimental results of pre-scission neutron multiplicities from four compound nuclei, namely 210,212,214,216 Rn, and statistical model analysis of the corresponding data. The compound nuclei 210,212,214,216 Rn having N/Z values as 1.441, 1.465, 1.488, 1.511 respectively are populated through the 16,18 O+ 194,198 Pt reactions at excitation energies of 50, 61, 71.7 and 79 MeV. The measured neutron multiplicities are further analyzed with the statistical model of nuclear decay where fission hindrance due to nuclear dissipation is considered. The N/Z dependence of the dissipation strength at lowest excitation energy of the compound nuclei suggests shell closure effects. However, such effects are not observed at higher excitations where the variation of the dissipation strength with N/Z does not show any specific trend. The variation of N/Z in fission time scale is also shown. (authors)

The INEL beryllium multiplication experiment

International Nuclear Information System (INIS)

Smith, J.R.; King, J.J.

1991-03-01

The experiment to measure the multiplication of 14-MeV neutrons in bulk beryllium has been completed. The experiment consists of determining the ratio of 56 Mn activities induced in a large manganese bath by a central 14-MeV neutron source, with and without a beryllium sample surrounding the source. In the manganese bath method a neutron source is placed at the center of a totally-absorbing aqueous solution of MnSo 4 . The capture of neutrons by Mn produces a 56 Mn activity proportional to the emission rate of the source. As applied to the measurement of the multiplication of 14- MeV neutrons in bulk beryllium, the neutron source is a tritium target placed at the end of the drift tube of a small deuteron accelerator. Surrounding the source is a sample chamber. When the sample chamber is empty, the neutrons go directly to the surrounding MnSO 4 solution, and produce a 56 Mn activity proportional to the neutron emission rate. When the chamber contains a beryllium sample, the neutrons first enter the beryllium and multiply through the (n,2n) process. Neutrons escaping from the beryllium enter the bath and produce a 56 Mn activity proportional to the neutron emission rate multiplied by the effective value of the multiplication in bulk beryllium. The ratio of the activities with and without the sample present is proportional to the multiplication value. Detailed calculations of the multiplication and all the systematic effects were made with the Monte Carlo program MCNP, utilizing both the Young and Stewart and the ENDF/B-VI evaluations for beryllium. Both data sets produce multiplication values that are in excellent agreement with the measurements for both raw and corrected values of the multiplication. We conclude that there is not real discrepancy between experimental and calculated values for the multiplication of neutrons in bulk beryllium. 12 figs., 11 tabs., 18 refs

Multiple order reflections in crystal neutron monochromators

International Nuclear Information System (INIS)

Fulfaro, R.

1976-01-01

A study of the higher order reflections in neutron crystal monochromators was made in order to obtain, for the IEA single crystal spectrometer, the operation range of 1,0eV to 0,01eV. Two crystals were studied, an Al(III) near 1,0eV and a Ge(III) in lower energies. For the Ge(III) case the higher order contaminations in the reflected beam were determined using as standard the gold total neutron cross section and performing the crystal reflectivity calculation for several orders of reflection. The knowledge of the contamination for each order as a function of neutron wavelength allows the optimization of the filter thickness in order to avoid higher order neutrons. The Ge(III) crystal was used because its second order reflections are theoretically forbidden, giving an advantage on other crystals, since measurements can be made until 0.02eV directly without filters. In the energy range 0.02 to 0.01eV, order contaminations higher than the second are present, therefore, either quartz filters are employed or calculated corrections are applied to the experimental data. The Al(III) crystal was used in order to estimate the second order contamination effect, in the iridium resonance measurements, at E 0 = 0.654eV. In that region, approximations can be made and it was not necessary to make the crystal reflectivity calculation for the filters thickness optimization. Since only the second order affects the results in that region, tellurium was used for the filtration, because this element has a resonance in the range of neutrons with energy 4E [pt

Deterministic calculation of the effective delayed neutron fraction without using the adjoint neutron flux - 299

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.

2010-01-01

In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)

Aspects of OER and RBE relevant to neutron therapy

International Nuclear Information System (INIS)

Field, S.B.; Hornsey, S.

1979-01-01

This chapter contains information concerning the mechanisms involved in neutron radiotherapy. Early studies on the attempts of using neutrons in radiotherapy are described. The rationale for fast neutron therapy is discussed as well as the relationships between OER and LET. Tissue responses include: repopulation of surviving cells; repair of sublethal damage; and slow repair. These mechanisms are considered separately. The relationships between RBE and dose per fraction for damage to skin, intestine, esophagus, lungs, hemopoietic tissue, and nerve tissue are discussed. Factors governing the effects of fractionation of dose in neutron radiotherapy are presented. Observations on mammalian cells and tissues show a general reduction in RBE with increasing neutron energy. The benefits of using mixed treatments, part with neutrons and the remainder with photons, are discussed. Problems with this approach include uncertainties of how the combination will effect normal tissue, how it effects slow repair, or its potentially lethal damage. Tumor response, as compared with x rays, to single and multiple doses of radiation is described. Clinical results are given

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

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

International Nuclear Information System (INIS)

Takahashi, A.; Rusch, D.

1979-10-01

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

Acceleration of criticality analysis solution convergence by matrix eigenvector for a system with weak neutron interaction

Energy Technology Data Exchange (ETDEWEB)

Nomura, Yasushi; Takada, Tomoyuki; Kuroishi, Takeshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kadotani, Hiroyuki [Shizuoka Sangyo Univ., Iwata, Shizuoka (Japan)

2003-03-01

In the case of Monte Carlo calculation to obtain a neutron multiplication factor for a system of weak neutron interaction, there might be some problems concerning convergence of the solution. Concerning this difficulty in the computer code calculations, theoretical derivation was made from the general neutron transport equation and consideration was given for acceleration of solution convergence by using the matrix eigenvector in this report. Accordingly, matrix eigenvector calculation scheme was incorporated together with procedure to make acceleration of convergence into the continuous energy Monte Carlo code MCNP. Furthermore, effectiveness of acceleration of solution convergence by matrix eigenvector was ascertained with the results obtained by applying to the two OECD/NEA criticality analysis benchmark problems. (author)

Boundary element methods applied to two-dimensional neutron diffusion problems

International Nuclear Information System (INIS)

Itagaki, Masafumi

1985-01-01

The Boundary element method (BEM) has been applied to two-dimensional neutron diffusion problems. The boundary integral equation and its discretized form have been derived. Some numerical techniques have been developed, which can be applied to critical and fixed-source problems including multi-region ones. Two types of test programs have been developed according to whether the 'zero-determinant search' or the 'source iteration' technique is adopted for criticality search. Both programs require only the fluxes and currents on boundaries as the unknown variables. The former allows a reduction in computing time and memory in comparison with the finite element method (FEM). The latter is not always efficient in terms of computing time due to the domain integral related to the inhomogeneous source term; however, this domain integral can be replaced by the equivalent boundary integral for a region with a non-multiplying medium or with a uniform source, resulting in a significant reduction in computing time. The BEM, as well as the FEM, is well suited for solving irregular geometrical problems for which the finite difference method (FDM) is unsuited. The BEM also solves problems with infinite domains, which cannot be solved by the ordinary FEM and FDM. Some simple test calculations are made to compare the BEM with the FEM and FDM, and discussions are made concerning the relative merits of the BEM and problems requiring future solution. (author)

Optimization of constrained multiple-objective reliability problems using evolutionary algorithms

International Nuclear Information System (INIS)

Salazar, Daniel; Rocco, Claudio M.; Galvan, Blas J.

2006-01-01

This paper illustrates the use of multi-objective optimization to solve three types of reliability optimization problems: to find the optimal number of redundant components, find the reliability of components, and determine both their redundancy and reliability. In general, these problems have been formulated as single objective mixed-integer non-linear programming problems with one or several constraints and solved by using mathematical programming techniques or special heuristics. In this work, these problems are reformulated as multiple-objective problems (MOP) and then solved by using a second-generation Multiple-Objective Evolutionary Algorithm (MOEA) that allows handling constraints. The MOEA used in this paper (NSGA-II) demonstrates the ability to identify a set of optimal solutions (Pareto front), which provides the Decision Maker with a complete picture of the optimal solution space. Finally, the advantages of both MOP and MOEA approaches are illustrated by solving four redundancy problems taken from the literature

Optimization of constrained multiple-objective reliability problems using evolutionary algorithms

Energy Technology Data Exchange (ETDEWEB)

Salazar, Daniel [Instituto de Sistemas Inteligentes y Aplicaciones Numericas en Ingenieria (IUSIANI), Division de Computacion Evolutiva y Aplicaciones (CEANI), Universidad de Las Palmas de Gran Canaria, Islas Canarias (Spain) and Facultad de Ingenieria, Universidad Central Venezuela, Caracas (Venezuela)]. E-mail: danielsalazaraponte@gmail.com; Rocco, Claudio M. [Facultad de Ingenieria, Universidad Central Venezuela, Caracas (Venezuela)]. E-mail: crocco@reacciun.ve; Galvan, Blas J. [Instituto de Sistemas Inteligentes y Aplicaciones Numericas en Ingenieria (IUSIANI), Division de Computacion Evolutiva y Aplicaciones (CEANI), Universidad de Las Palmas de Gran Canaria, Islas Canarias (Spain)]. E-mail: bgalvan@step.es

2006-09-15

This paper illustrates the use of multi-objective optimization to solve three types of reliability optimization problems: to find the optimal number of redundant components, find the reliability of components, and determine both their redundancy and reliability. In general, these problems have been formulated as single objective mixed-integer non-linear programming problems with one or several constraints and solved by using mathematical programming techniques or special heuristics. In this work, these problems are reformulated as multiple-objective problems (MOP) and then solved by using a second-generation Multiple-Objective Evolutionary Algorithm (MOEA) that allows handling constraints. The MOEA used in this paper (NSGA-II) demonstrates the ability to identify a set of optimal solutions (Pareto front), which provides the Decision Maker with a complete picture of the optimal solution space. Finally, the advantages of both MOP and MOEA approaches are illustrated by solving four redundancy problems taken from the literature.

Ultracold neutrons

International Nuclear Information System (INIS)

Steenstrup, S.

Briefly surveys recent developments in research work with ultracold neutrons (neutrons of very low velocity, up to 10 m/s at up to 10 -7 eV and 10 -3 K). Slow neutrons can be detected in an ionisation chamber filled with B 10 F 3 . Very slow neutrons can be used for investigations into the dipole moment of neutrons. Neutrons of large wave length have properties similar to those of light. The limit angle for total reflection is governed by the wave length and by the material. Total reflection can be used to filter ultracold neutrons out of the moderator material of a reactor. Total reflection can also be used to store ultracold neutrons but certain problems with storage have not yet been clarified. Slow neutrons can be made to lose speed in a neutron turbine, and come out as ultracold neutrons. A beam of ultracold neutrons could be used in a neutron microscope. (J.S.)

Experiments and Simulations of the Use of Time-Correlated Thermal Neutron Counting to Determine the Multiplication of an Assembly of Highly Enriched Uranium

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester; Mathew T. Kinlaw; Scott M. Watson; Jeffrey M. Kalter; Eric C. Miller; William A. Noonan

2014-11-01

A series of experiments and numerical simulations using thermal-neutron time-correlated measurements has been performed to determine the neutron multiplication, M, of assemblies of highly enriched uranium available at Idaho National Laboratory. The experiments used up to 14.4 kg of highly-enriched uranium, including bare assemblies and assemblies reflected with high-density polyethylene, carbon steel, and tungsten. A small 252Cf source was used to initiate fission chains within the assembly. Both the experiments and the simulations used 6-channel and 8-channel detector systems, each consisting of 3He proportional counters moderated with polyethylene; data was recorded in list mode for analysis. 'True' multiplication values for each assembly were empirically derived using basic neutron production and loss values determined through simulation. A total of one-hundred and sixteen separate measurements were performed using fifty-seven unique measurement scenarios, the multiplication varied from 1.75 to 10.90. This paper presents the results of these comparisons and discusses differences among the various cases.

Verification of a three-dimensional neutronics model based on multi-point kinetics equations for transient problems

Energy Technology Data Exchange (ETDEWEB)

Park, Kyung Seok; Kim, Hyun Dae; Yeom, Choong Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

A computer code for solving the three-dimensional reactor neutronic transient problems utilizing multi-point reactor kinetics equations recently developed has been developed. For evaluating its applicability, the code has been tested with typical 3-D LWR and CANDU reactor transient problems. The performance of the method and code has been compared with the results by fine and coarse meshes computer codes employing the direct methods.

Complex multiplication and lifting problems

CERN Document Server

Chai, Ching-Li; Oort, Frans

2013-01-01

Abelian varieties with complex multiplication lie at the origins of class field theory, and they play a central role in the contemporary theory of Shimura varieties. They are special in characteristic 0 and ubiquitous over finite fields. This book explores the relationship between such abelian varieties over finite fields and over arithmetically interesting fields of characteristic 0 via the study of several natural CM lifting problems which had previously been solved only in special cases. In addition to giving complete solutions to such questions, the authors provide numerous examples to illustrate the general theory and present a detailed treatment of many fundamental results and concepts in the arithmetic of abelian varieties, such as the Main Theorem of Complex Multiplication and its generalizations, the finer aspects of Tate's work on abelian varieties over finite fields, and deformation theory. This book provides an ideal illustration of how modern techniques in arithmetic geometry (such as descent the...

Some principal problems in physics and low-energy neutron physics

International Nuclear Information System (INIS)

Aleksandrov, Yu.A.

2004-01-01

The questions connected with internal particle (e.g. neutron) structure obtained at low-energy neutron physics are discussed. The first question deals with the charge neutron radius E 2 > 1/2 connected with the value of neutron-electron scattering length a ne determined at low neutron energies. At present, the obtained accuracy allows us to speak not only about the value of E 2 > but also on the segmentation of E 2 > into Dirac and Foldy addenda. The sign of the Dirac addendum is connected directly with the fundamental Yukawa theory explaining the origin of nuclear forces. One of the popular experimental values of the Dirac addendum (from a ne =(-1.32±0.03)·10 -16 cm) contradicts the Yukawa theory. The second question also concerns the subject of the structure of the neutron, namely its deformation. The notion of deformation (polarizability) of the nucleon in electromagnetic field was introduced in the mid-1950s. The reasons are given in favor of the opinion that the neutron polarizability was observed for the first time in neutron experiments as far back as 1957, i.e. earlier than proton polarizability was detected (1960). Finally, the third question deals with the search for a magnetic charge of the neutron. A beautiful experiment (Finkelstein, Shull, Zeilinger, 1986) testifying with high accuracy the absence of a magnetic charge of the neutron is discussed. This diffraction experiment was based on the concept of anomalously small effective mass of the neutron providing greatly enhanced sensitivity. The existence of an isolated magnetic charge in the nature would explain the quantization of electric and magnetic charges (Dirac, 1931)

Protective factors associated with fewer multiple problem behaviors among homeless/runaway youth.

Science.gov (United States)

Lightfoot, Marguerita; Stein, Judith A; Tevendale, Heather; Preston, Kathleen

2011-01-01

Although homeless youth exhibit numerous problem behaviors, protective factors that can be targeted and modified by prevention programs to decrease the likelihood of involvement in risky behaviors are less apparent. The current study tested a model of protective factors for multiple problem behavior in a sample of 474 homeless youth (42% girls; 83% minority) ages 12 to 24 years. Higher levels of problem solving and planning skills were strongly related to lower levels of multiple problem behaviors in homeless youth, suggesting both the positive impact of preexisting personal assets of these youth and important programmatic targets for further building their resilience and decreasing problem behaviors. Indirect relationships between the background factors of self-esteem and social support and multiple problem behaviors were significantly mediated through protective skills. The model suggests that helping youth enhance their skills in goal setting, decision making, and self-reliant coping could lessen a variety of problem behaviors commonly found among homeless youth.

Neutron-multiplication measurement instrument

Energy Technology Data Exchange (ETDEWEB)

Nixon, K.V.; Dowdy, E.J.; France, S.W.; Millegan, D.R.; Robba, A.A.

1982-01-01

The Advanced Nuclear Technology Group of the Los Alamos National Laboratory is now using intelligent data-acquisition and analysis instrumentation for determining the multiplication of nuclear material. Earlier instrumentation, such as the large NIM-crate systems, depended on house power and required additional computation to determine multiplication or to estimate error. The portable, battery-powered multiplication measurement unit, with advanced computational power, acquires data, calculates multiplication, and completes error analysis automatically. Thus, the multiplication is determined easily and an available error estimate enables the user to judge the significance of results.

Neutron multiplication measurement instrument

International Nuclear Information System (INIS)

Nixon, K.V.; Dowdy, E.J.; France, S.W.; Millegan, D.R.; Robba, A.A.

1983-01-01

The Advanced Nuclear Technology Group of the Los Alamos National Laboratory is now using intelligent data-acquisition and analysis instrumentation for determining the multiplication of nuclear material. Earlier instrumentation, such as the large NIM-crate systems, depended on house power and required additional computation to determine multiplication or to estimate error. The portable, battery-powered multiplication measurement unit, with advanced computational power, acquires data, calculates multiplication, and completes error analysis automatically. Thus, the multiplication is determined easily and an available error estimate enables the user to judge the significance of results

Neutron-multiplication measurement instrument

International Nuclear Information System (INIS)

Nixon, K.V.; Dowdy, E.J.; France, S.W.; Millegan, D.R.; Robba, A.A.

1982-01-01

The Advanced Nuclear Technology Group of the Los Alamos National Laboratory is now using intelligent data-acquisition and analysis instrumentation for determining the multiplication of nuclear material. Earlier instrumentation, such as the large NIM-crate systems, depended on house power and required additional computation to determine multiplication or to estimate error. The portable, battery-powered multiplication measurement unit, with advanced computational power, acquires data, calculates multiplication, and completes error analysis automatically. Thus, the multiplication is determined easily and an available error estimate enables the user to judge the significance of results

MADNIX a code to calculate prompt fission neutron spectra and average prompt neutron multiplicities

International Nuclear Information System (INIS)

Merchant, A.C.

1986-03-01

A code has been written and tested on the CDC Cyber-170 to calculate the prompt fission neutron spectrum, N(E), as a function of both the fissioning nucleus and its excitation energy. In this note a brief description of the underlying physical principles involved and a detailed explanation of the required input data (together with a sample output for the fission of 235 U induced by 14 MeV neutrons) are presented. Weisskopf's standard nuclear evaporation theory provides the basis for the calculation. Two important refinements are that the distribution of fission-fragment residual nuclear temperature and the cooling of the fragments as neutrons are emitted approximately taken into account, and also the energy dependence of the cross section for the inverse process of compound nucleus formation is included. This approach is then used to calculate the average number of prompt neutrons emitted per fission, v-bar p . At high excitation energies, where fission is still possible after neutron emission, the consequences of the competition between first, second and third chance fission on N(E) and v-bar p are calculated. Excellent agreement with all the examples given in the original work of Madland and Nix is obtained. (author) [pt

Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem

International Nuclear Information System (INIS)

William Charlton

2007-01-01

Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions

Preliminary performance analysis of exponential experimental system for the determination of neutron effective multiplication factor of PWR spent fuel

International Nuclear Information System (INIS)

Shin, Heesung; Lee, Sang-Yun; Ro, Seung-Gy; Seo, Gi-Seok; Kim, Ho-Dong

2002-01-01

An exponential experiment system which is composed of neutron detector, signal analysis system and neutron source, 10 mCi Cf-252 has been installed in the storage pool of PIEF at KAERI in order to experimentally determining neutron effective multiplication factors of PWR spent fuel assemblies. Preliminary functional characteristic tests of the experimental system are performed for C15, J14 and J44 assemblies loaded in the pool. As a result of preliminary tests, the average neutron counts obtained for 3 minutes in the plateau of the C15, J14 and J44 assemblies are about 1900, 3800 and 3200, respectively. A dip of the neutron flux density distribution is noticed in the spacer grid position. Neutron counts at those positions appear to be reduced to about 70 % in comparison to the fuel position. The measured axial neutron distribution shapes are compared with the result for the P14 assembly and Cs-137 gamma scanning data performed in KAERI. It is revealed that the spacer grid position measured is consistent with the design specifications within a 2.3 % error. The exponential decay constants for the C15 assembly were determined to be 0.152 and 0.165 for detector and source scanning, respectively. (author)

DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program

International Nuclear Information System (INIS)

Courtney, J. C.

1987-01-01

1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Solution to the monoenergetic time-dependent neutron transport equation with a time-varying source

International Nuclear Information System (INIS)

Ganapol, B.D.

1986-01-01

Even though fundamental time-dependent neutron transport problems have existed since the inception of neutron transport theory, it has only been recently that a reliable numerical solution to one of the basic problems has been obtained. Experience in generating numerical solutions to time-dependent transport equations has indicated that the multiple collision formulation is the most versatile numerical technique for model problems. The formulation coupled with a moment reconstruction of each collided flux component has led to benchmark-quality (four- to five-digit accuracy) numerical evaluation of the neutron flux in plane infinite geometry for any degree of scattering anisotropy and for both pulsed isotropic and beam sources. As will be shown in this presentation, this solution can serve as a Green's function, thus extending the previous results to more complicated source situations. Here we will be concerned with a time-varying source at the center of an infinite medium. If accurate, such solutions have both pedagogical and practical uses as benchmarks against which other more approximate solutions designed for a wider class of problems can be compared

Criticality problems for slabs and spheres in energy dependent neutron transport theory

International Nuclear Information System (INIS)

Victory, H.D. Jr.

1980-01-01

The steady-state equation for energy-dependent neutron transport in isotropically scattering slabs and spheres is formulated as an integral equation. The Perron-Frobenius-Jentzsch theory of positive operators is used to analyze criticality problems for transport in slab and spherical media consisting of core and reflector. In addition, with an adroit selection of diffusion-like solutions, this theory is used to obtain an expression relating the critical radius of a homogeneous sphere to a parameter characterizing fission production. 21 refs

Isotopic neutron sources for neutron activation analysis

International Nuclear Information System (INIS)

Hoste, J.

1988-06-01

This User's Manual is an attempt to provide for teaching and training purposes, a series of well thought out demonstrative experiments in neutron activation analysis based on the utilization of an isotopic neutron source. In some cases, these ideas can be applied to solve practical analytical problems. 19 refs, figs and tabs

A polarizing neutron periscope for neutron imaging

International Nuclear Information System (INIS)

Schulz, Michael; Boeni, Peter; Calzada, Elbio; Muehlbauer, Martin; Neubauer, Andreas; Schillinger, Burkhard

2009-01-01

Optical neutron polarizers like guides or benders destroy the collimation of a neutron beam due to multiple reflections or scattering. This makes them unsuitable for their use in polarized neutron radiography, because the beam collimation is essential to obtain high spatial resolution. We have developed a neutron polarizer based on the principle of an optical periscope with a zigzag double reflection on two parallel high-m supermirror polarizers. If the supermirrors are perfectly parallel and flat, the beam collimation is left unchanged by such a device. A first proof of concept version of this type of polarizer was built and tested. We expect to achieve a beam polarization of up to 99% with an improved version yet to be built.

Cold neutron interferometry and its application. 2. Coherency and cold neutron spin interferometry

International Nuclear Information System (INIS)

Achiwa, Norio; Ebisawa, Toru

1998-03-01

The second workshop entitled 'Interference studies and cold neutron spin interferometry' was held on 10 and 11 March 1998 at KUR (Kyoto University Research Reactor Institute, Kumatori). Cold neutron spin interferometry is a new field. So it is very important for its development to learn the studies of X-ray and neutron optics which are rapidly developing with long history. In the workshop, the issues related to interference were reviewed such as experimental studies on cold neutron spin interferometry, theoretical and experimental approach on tunneling time, interference experiments by neutrons and its application, interference studies using synchrotron radiation, topics on silicon interferometry and quantum measurement problem and cold neutron interference experiment related to quantum measurement problem. The 8 of the presented papers are indexed individually. (J.P.N.)

Scintillation neutron detector with dynamic threshold

International Nuclear Information System (INIS)

Kornilov, N.; Massey, T.; Grimes, S.

2014-01-01

Scintillation neutron detectors with hydrogen are a common tool for neutron spectroscopy. They provide good time resolution, neutron-gamma discrimination and high efficiency of neutron counting. The real open problems connected with application of these detectors are in the energy range >10 MeV. There are no standard neutron spectra known with high accuracy for this energy range. Therefore, traditional methods for experimental investigation of the efficiency function fail for these neutrons. The Monte Carlo simulation cannot provide reasonable accuracy due to unknown characteristics of the reactions for charged particle production (p, α and so on, light output, reaction cross-sections). The application of fission chamber with fissile material as a neutron detector did not help to solve the problem. We may avoid many problems if we use the traditional neutron detector with non-traditional data analysis. In this report we give main relations, and demonstrate the method for Cf-source. Experimental detector efficiency is compared with MC simulation. (authors)

Child outcomes of home-visiting for families with complex and multiple problems

NARCIS (Netherlands)

van Assen, Arend; Dickscheit, Jana; Post, Wendy; Grietens, Hans

2016-01-01

Introduction Families with complex and multiple problems are faced with an accumulation of problems across multiple areas of life. Furthermore, these families are often considered to be ‘difficult to treat’. Children and teenagers growing up in these families are exposed to an accumulation of risks

Neutronic modelling of the Harwell MTR's: some recent problems

International Nuclear Information System (INIS)

Taylor, N.P.

1984-01-01

Use of the Harwell Materials Testing Reactors for the irradiation of experimental rigs gives rise to a number of requirements for calculations of neutron fluxes. In addition photon fluxes are required for estimates of nuclear heating rates. A range of calculational methods are employed, from simple cell to whole reactor models, and the latter have been extended for preliminary design studies for the next generation of MTR to replace DIDO and PLUTO. The technique used for these various models are described in this note, with emphasis on the areas in which modelling problems are encountered. The applications divide into three distinct areas: calculations concerning rigs irradiated within the reactor core, those for rigs positioned in the D 2 O reflector surrounding the core, and design studies for a replacement reactor. (Auth.)

TLD-300 detectors for separate measurement of total and gamma absorbed dose distributions of single, multiple, and moving-field neutron treatments

International Nuclear Information System (INIS)

Rassow, J.

1984-01-01

Fast neutron therapy requirements, because of the poor depth dose characteristic of present therapeutical sources, are at least as complex in treatment plans as photon therapy. The physical part of the treatment planning is very important; however, it is much more complicated than for photons or electrons owing to the need for: Separation of total and gamma absorbed dose distributions (Dsub(T) and Dsub(G)); and more stringent tissue-equivalence conditions of phantoms than in photon therapy. Therefore, methods of clinical dosimetry for the separate determination of total and gamma absorbed dose distributions in irregularly shaped (inhomogeneous) phantoms are needed. A method using TLD-300 (CaF 2 :Tm) detectors is described, which is able to give an approximate solution of the above-mentioned dosimetric requirements. The two independent doses, Dsub(T) and Dsub(G), can be calculated by an on-line computer analysis of the digitalized glow curve of TLD-300 detectors, irradiated with d(14)+Be neutrons of the cyclotron isocentric neutron therapy facility CIRCE in Essen. Results are presented for depth and lateral absorbed dose distributions (Dsub(T) and Dsub(G)) for fixed neutron beams of different field sizes compared with measurements by standard procedures (TE-TE ionization chamber, GM counter) in an A-150 phantom. The TLD-300 results for multiple and moving-field treatments (with and without wedge filters) in a patient simulating irregularly shaped (inhomogeneous) phantoms, are shown together with computer calculations of these dose distributions. The probable causes for some systematic deviations are discussed, which lead to open problems for further investigations owing to features of the detector material and the evaluation method, but mainly to differences in the composition of phantom materials used for the calculations (standard dose distributions) and TLD-300 measurements. (author)

Neutronics design for lead-bismuth cooled accelerator-driven system for transmutation of minor actinide

International Nuclear Information System (INIS)

Tsujimoto, Kazufumi; Sasa, Toshinobu; Nishihara, Kenji; Oigawa, Hiroyuki; Takano, Hideki

2004-01-01

Neutronics design study was performed for lead-bismuth cooled accelerator-driven system (ADS) to transmute minor actinides. Early study for ADS indicated two problems: a large burnup reactivity swing and a significant peaking factor. To solve these problems, effect of design parameters on neutronics characteristics were searched. The design parameters were initial plutonium loading, buffer region between spallation target and core, and zone fuel loading. Parametric survey calculations were performed considering fuel cycle consisting of burnup and recycle. The results showed that burnup reactivity swing depends on the plutonium fraction in the initial fuel loading, and the lead-bismuth buffer region and the two-zone loading were effective for solving the problems. Moreover, an optimum value for the effective multiplication factor was also evaluated using reactivity coefficients. From the result, the maximum allowable value of the effective multiplication factor for a practical ADS can be set at 0.97. Consequently, a new core concept combining the buffer region and the two-zone loading was proposed base on the results of the parametric survey. (author)

NESTLE: Few-group neutron diffusion equation solver utilizing the nodal expansion method for eigenvalue, adjoint, fixed-source steady-state and transient problems

International Nuclear Information System (INIS)

Turinsky, P.J.; Al-Chalabi, R.M.K.; Engrand, P.; Sarsour, H.N.; Faure, F.X.; Guo, W.

1994-06-01

NESTLE is a FORTRAN77 code that solves the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM). NESTLE can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source steady-state; or external fixed-source. or eigenvalue initiated transient problems. The code name NESTLE originates from the multi-problem solution capability, abbreviating Nodal Eigenvalue, Steady-state, Transient, Le core Evaluator. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two or four energy groups can be utilized, with all energy groups being thermal groups (i.e. upscatter exits) if desired. Core geometries modelled include Cartesian and Hexagonal. Three, two and one dimensional models can be utilized with various symmetries. The non-linear iterative strategy associated with the NEM method is employed. An advantage of the non-linear iterative strategy is that NSTLE can be utilized to solve either the nodal or Finite Difference Method representation of the few-group neutron diffusion equation

The principal component analysis method used with polynomial Chaos expansion to propagate uncertainties through critical transport problems

Energy Technology Data Exchange (ETDEWEB)

Rising, M. E.; Prinja, A. K. [Univ. of New Mexico, Dept. of Chemical and Nuclear Engineering, Albuquerque, NM 87131 (United States)

2012-07-01

A critical neutron transport problem with random material properties is introduced. The total cross section and the average neutron multiplicity are assumed to be uncertain, characterized by the mean and variance with a log-normal distribution. The average neutron multiplicity and the total cross section are assumed to be uncorrected and the material properties for differing materials are also assumed to be uncorrected. The principal component analysis method is used to decompose the covariance matrix into eigenvalues and eigenvectors and then 'realizations' of the material properties can be computed. A simple Monte Carlo brute force sampling of the decomposed covariance matrix is employed to obtain a benchmark result for each test problem. In order to save computational time and to characterize the moments and probability density function of the multiplication factor the polynomial chaos expansion method is employed along with the stochastic collocation method. A Gauss-Hermite quadrature set is convolved into a multidimensional tensor product quadrature set and is successfully used to compute the polynomial chaos expansion coefficients of the multiplication factor. Finally, for a particular critical fuel pin assembly the appropriate number of random variables and polynomial expansion order are investigated. (authors)

American National Standard: for safety in conducting subcritical neutron-multiplication measurements in-situ

International Nuclear Information System (INIS)

1983-01-01

This standard provides safety guidance for conducting subcritical neutron-multiplication measurements where physical protection of personnel against the consequences of a criticality accident is not provided. The objectives of in-situ measurements are either to confirm an adequate safety margin or to improve an estimate of such a margin. The first objective may constitute a test of the criticality safety of a design that is based on calculations. The second may effect improved operating conditions by reducing the uncertainty of safety margins and providing guidance to new designs

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.

Multiple scattering problems in heavy ion elastic recoil detection analysis

International Nuclear Information System (INIS)

Johnston, P.N.; El Bouanani, M.; Stannard, W.B.; Bubb, I.F.; Cohen, D.D.; Dytlewski, N.; Siegele, R.

1998-01-01

A number of groups use Heavy Ion Elastic Recoil Detection Analysis (HIERDA) to study materials science problems. Nevertheless, there is no standard methodology for the analysis of HIERDA spectra. To overcome this deficiency we have been establishing codes for 2-dimensional data analysis. A major problem involves the effects of multiple and plural scattering which are very significant, even for quite thin (∼100 nm) layers of the very heavy elements. To examine the effects of multiple scattering we have made comparisons between the small-angle model of Sigmund et al. and TRIM calculations. (authors)

High-Energy Neutron Backgrounds for Underground Dark Matter Experiments

Energy Technology Data Exchange (ETDEWEB)

Chen, Yu [Syracuse Univ., NY (United States)

2016-01-01

Direct dark matter detection experiments usually have excellent capability to distinguish nuclear recoils, expected interactions with Weakly Interacting Massive Particle (WIMP) dark matter, and electronic recoils, so that they can efficiently reject background events such as gamma-rays and charged particles. However, both WIMPs and neutrons can induce nuclear recoils. Neutrons are then the most crucial background for direct dark matter detection. It is important to understand and account for all sources of neutron backgrounds when claiming a discovery of dark matter detection or reporting limits on the WIMP-nucleon cross section. One type of neutron background that is not well understood is the cosmogenic neutrons from muons interacting with the underground cavern rock and materials surrounding a dark matter detector. The Neutron Multiplicity Meter (NMM) is a water Cherenkov detector capable of measuring the cosmogenic neutron flux at the Soudan Underground Laboratory, which has an overburden of 2090 meters water equivalent. The NMM consists of two 2.2-tonne gadolinium-doped water tanks situated atop a 20-tonne lead target. It detects a high-energy (>~ 50 MeV) neutron via moderation and capture of the multiple secondary neutrons released when the former interacts in the lead target. The multiplicity of secondary neutrons for the high-energy neutron provides a benchmark for comparison to the current Monte Carlo predictions. Combining with the Monte Carlo simulation, the muon-induced high-energy neutron flux above 50 MeV is measured to be (1.3 ± 0.2) ~ 10-9 cm-2s-1, in reasonable agreement with the model prediction. The measured multiplicity spectrum agrees well with that of Monte Carlo simulation for multiplicity below 10, but shows an excess of approximately a factor of three over Monte Carlo prediction for multiplicities ~ 10 - 20. In an effort to reduce neutron backgrounds for the dark matter experiment SuperCDMS SNO- LAB, an active neutron veto was developed

Scram device having a multiplicity of neutron absorbing masses

International Nuclear Information System (INIS)

Giuggio, N.; Noyes, R.C.

1981-01-01

An apparatus is described for holding, releasing, and resetting a multiplicity of neutron-absorbing balls within a safety assembly of a liquid metal reactor. Vertically-hinged trap doors rest on the shoulders of a generally cylindrical release valve which is actuated by either the regular or by the self-actuated scram actuator. The doors and the valve shoulder provide a floor for the balls to be suspended above the reactor core during normal operation. When the actuator displaces the release valve, the doors lose their support and swing downward, permitting the poison balls to drop into the core. In the reset mode of operation, a platform at the bottom of the core is raised to lift the balls and swing the trap doors upward until the balls are above the door hinges. The release valve is reset to support the doors and the platform is lowered to the bottom of the safety assembly

Analytical applications for delayed neutrons

International Nuclear Information System (INIS)

Eccleston, G.W.

1983-01-01

Analytical formulations that describe the time dependence of neutron populations in nuclear materials contain delayed-neutron dependent terms. These terms are important because the delayed neutrons, even though their yields in fission are small, permit control of the fission chain reaction process. Analytical applications that use delayed neutrons range from simple problems that can be solved with the point reactor kinetics equations to complex problems that can only be solved with large codes that couple fluid calculations with the neutron dynamics. Reactor safety codes, such as SIMMER, model transients of the entire reactor core using coupled space-time neutronics and comprehensive thermal-fluid dynamics. Nondestructive delayed-neutron assay instruments are designed and modeled using a three-dimensional continuous-energy Monte Carlo code. Calculations on high-burnup spent fuels and other materials that contain a mix of uranium and plutonium isotopes require accurate and complete information on the delayed-neutron periods, yields, and energy spectra. A continuing need exists for delayed-neutron parameters for all the fissioning isotopes

Conjugate Gradient like methods and their application to fixed source neutron diffusion problems

International Nuclear Information System (INIS)

Suetomi, Eiichi; Sekimoto, Hiroshi

1989-01-01

This paper presents a number of fast iterative methods for solving systems of linear equations appearing in fixed source problems for neutron diffusion. We employed the conjugate gradient and conjugate residual methods. In order to accelerate the conjugate residual method, we proposed the conjugate residual squared method by transforming the residual polynomial of the conjugate residual method. Since the convergence of these methods depends on the spectrum of coefficient matrix, we employed the incomplete Choleski (IC) factorization and the modified IC (MIC) factorization as preconditioners. These methods were applied to some neutron diffusion problems and compared with the successive overrelaxation (SOR) method. The results of these numerical experiments showed superior convergence characteristics of the conjugate gradient like method with MIC factorization to the SOR method, especially for a problem involving void region. The CPU time of the MICCG, MICCR and MICCRS methods showed no great difference. In order to vectorize the conjugate gradient like methods based on (M)IC factorization, the hyperplane method was used and implemented on the vector computers, the HITAC S-820/80 and ETA10-E (one processor mode). Significant decrease of the CPU times was observed on the S-820/80. Since the scaled conjugate gradient (SCG) method can be vectorized with no manipulation, it was also compared with the above methods. It turned out the SCG method was the fastest with respect to the CPU times on the ETA10-E. These results suggest that one should implement suitable algorithm for different vector computers. (author)

Very slow neutrons

International Nuclear Information System (INIS)

Frank, A.

1983-01-01

The history is briefly presented of the research so far of very slow neutrons and their basic properties are explained. The methods are described of obtaining very slow neutrons and the problems of their preservation are discussed. The existence of very slow neutrons makes it possible to perform experiments which may deepen the knowledge of the fundamental properties of neutrons. Their wavelength approximates that of visible radiation. The possibilities and use are discussed of neutron optical systems (neutron microscope) which could be an effective instrument for the study of the detailed arrangement, especially of organic substances. (B.S.)

Observation of Neutron Skyshine from an Accelerator Based Neutron Source

Energy Technology Data Exchange (ETDEWEB)

Franklyn, C. B. [Radiation Science Department, Necsa, PO Box 582, Pretoria 0001 (South Africa)

2011-12-13

A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >10{sup 11} n{center_dot}s{sup -1}. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.

Neutrons and fusion

International Nuclear Information System (INIS)

Maynard, C.W.

1976-01-01

The production of energy from fusion reactions does not require neutrons in the fundamental sense that they are required in a fission reactor. Nevertheless, the dominant fusion reaction, that between deuterium and tritium, yields a 14 MeV neutron. To contrast a fusion reactor based on this reaction with the fission case, 3 x 10 20 such neutrons produced per gigawatt of power. This is four times as many neutrons as in an equivalent fission reactor and they carry seven times the energy of the fission neutrons. Thus, they dominate the energy recovery problem and create technological problems comparable to the original plasma confinement problem as far as a practical power producing device is concerned. Further contrasts of the fusion and fission cases are presented to establish the general role of neutrons in fusion devices. Details of the energy deposition processes are discussed and those reactions necessary for producing additional tritium are outlined. The relatively high energy flux with its large intensity will activate almost any materials of which the reactor may be composed. This activation is examined from the point of view of decay heat, radiological safety, and long-term storage. In addition, a discussion of the deleterious effects of neutron interactions on materials is given in some detail; this includes the helium and hydrogen producing reactions and displacement rate of the lattice atoms. The various materials that have been proposed for structural purposes, for breeding, reflecting, and moderating neutrons, and for radiation shielding are reviewed from the nuclear standpoint. The specific reactions of interest are taken up for various materials and finally a report is given on the status and prospects of data for fusion studies

Fissile mass estimation by pulsed neutron source interrogation

Energy Technology Data Exchange (ETDEWEB)

Israelashvili, I., E-mail: israelashvili@gmail.com [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Dubi, C.; Ettedgui, H.; Ocherashvili, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Pedersen, B. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Beck, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Roesgen, E.; Crochmore, J.M. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Ridnik, T.; Yaar, I. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel)

2015-06-11

Passive methods for detecting correlated neutrons from spontaneous fissions (e.g. multiplicity and SVM) are widely used for fissile mass estimations. These methods can be used for fissile materials that emit a significant amount of fission neutrons (like plutonium). Active interrogation, in which fissions are induced in the tested material by an external continuous source or by a pulsed neutron source, has the potential advantages of fast measurement, alongside independence of the spontaneous fissions of the tested fissile material, thus enabling uranium measurement. Until recently, using the multiplicity method, for uranium mass estimation, was possible only for active interrogation made with continues neutron source. Pulsed active neutron interrogation measurements were analyzed with techniques, e.g. differential die away analysis (DDA), which ignore or implicitly include the multiplicity effect (self-induced fission chains). Recently, both, the multiplicity and the SVM techniques, were theoretically extended for analyzing active fissile mass measurements, made by a pulsed neutron source. In this study the SVM technique for pulsed neutron source is experimentally examined, for the first time. The measurements were conducted at the PUNITA facility of the Joint Research Centre in Ispra, Italy. First promising results, of mass estimation by the SVM technique using a pulsed neutron source, are presented.

Problems in the neutron dynamics of source-driven systems

International Nuclear Information System (INIS)

Ravetto, P.

2001-01-01

The present paper presents some neutronic features of source-driven neutron multiplying systems, with special regards to dynamics, discussing the validity and limitations of classical methods, developed for systems in the vicinity of criticality. Specific characteristics, such as source dominance and the role of delayed neutron emissions are illustrated. Some dynamic peculiarities of innovative concepts proposed for accelerator-driven systems, such as fluid-fuel, are also discussed. The second portion of the work formulates the quasi-static methods for source-driven systems, evidencing its novel features and presenting some numerical results. (author)

Some Principal Problems in Physics and Low-Energy Neutron Physics

CERN Document Server

Alexandrov, Yu A

2004-01-01

The first question deals with the charge neutron radius $^{1/2}$ connected with the value of neutron-electron scattering length $a_{ne}$ determined at low neutron energies. At present, the obtained accuracy allows us to speak not only about the value of $$ but also on the segmentation of $$ into Dirac and Foldy addenda. The sign of the Dirac addendum is connected directly with the fundamental Yukawa theory explaining the origin of nuclear forces. One of the popular experimental values of the Dirac addendum (from ${a}_{ne} = (-1.32 \\pm 0.03) \\cdot 10^{ - 16}$ cm) contradicts the Yukawa theory. The second question also concerns the subject of the structure of the neutron, namely its deformation. The notion of deformation (polarizability) of the nucleon in electromagnetic field was introduced in the mid-1950s. The reasons are given in favor of the opinion that the neutron polarizability was observed for the first time in neutron experiments as far back as 1957, i.\\,e. earlier than proton polarizability was detec...

Current status of neutron scattering in Thailand

International Nuclear Information System (INIS)

Ampornrat, Pantip

2000-01-01

The neutron scattering experiments in Thailand have been done continuously since the start up of the reactor. In 1977, Thai research reactor was modified into TRIGA MARK III core. After that, the neutron spectrometer was installed again under a development program. Installation of upgrading spectrometer was delayed because of some problems involving the neutron intensity and instruments. However, these problems were solved and the setup is almost completed. The paper reports the current status of neutron spectrometer, the problems and plans for the experiments. (author)

The neutron silicon lens. An update of the thermal neutron lens results

International Nuclear Information System (INIS)

Johnson, M.W.; Daymond, M.R.

2001-01-01

This paper introduces the concept of the Neutron Silicon Lens (NSL) and provides and update on the experimental results achieved to date. The NSL design is a cylindrical neutron lens based on the use of multiple neutron mirrors supported and separated by silicon wafers. Such lenses would have many applications in both the primary and scattered beams on neutron instruments, and would lead to immediate improvements where the sample to be illuminated is small, as in high pressure or engineering strain scanning instruments. (author)

The neutron silicon lens. An update of the thermal neutron lens results

Energy Technology Data Exchange (ETDEWEB)

Johnson, M.W.; Daymond, M.R. [Rutherford Appleton Laboratory, Chilton, Didcot, Oxfordshire (United Kingdom)

2001-03-01

This paper introduces the concept of the Neutron Silicon Lens (NSL) and provides and update on the experimental results achieved to date. The NSL design is a cylindrical neutron lens based on the use of multiple neutron mirrors supported and separated by silicon wafers. Such lenses would have many applications in both the primary and scattered beams on neutron instruments, and would lead to immediate improvements where the sample to be illuminated is small, as in high pressure or engineering strain scanning instruments. (author)

Performance modeling of parallel algorithms for solving neutron diffusion problems

International Nuclear Information System (INIS)

Azmy, Y.Y.; Kirk, B.L.

1995-01-01

Neutron diffusion calculations are the most common computational methods used in the design, analysis, and operation of nuclear reactors and related activities. Here, mathematical performance models are developed for the parallel algorithm used to solve the neutron diffusion equation on message passing and shared memory multiprocessors represented by the Intel iPSC/860 and the Sequent Balance 8000, respectively. The performance models are validated through several test problems, and these models are used to estimate the performance of each of the two considered architectures in situations typical of practical applications, such as fine meshes and a large number of participating processors. While message passing computers are capable of producing speedup, the parallel efficiency deteriorates rapidly as the number of processors increases. Furthermore, the speedup fails to improve appreciably for massively parallel computers so that only small- to medium-sized message passing multiprocessors offer a reasonable platform for this algorithm. In contrast, the performance model for the shared memory architecture predicts very high efficiency over a wide range of number of processors reasonable for this architecture. Furthermore, the model efficiency of the Sequent remains superior to that of the hypercube if its model parameters are adjusted to make its processors as fast as those of the iPSC/860. It is concluded that shared memory computers are better suited for this parallel algorithm than message passing computers

Orion, a high efficiency 4π neutron detector

International Nuclear Information System (INIS)

Crema, E.; Piasecki, E.; Wang, X.M.; Doubre, H.; Galin, J.; Guerreau, D.; Pouthas, J.; Saint-Laurent, F.

1990-01-01

In intermediate energy heavy ion collisions the multiplicity of emitted neutrons is strongly connected to energy dissipation and to impact parameter. We present the 4π detector ORION, a high efficiency liquid scintillator detector which permits to get information on the multiplicity of neutrons measured event-wise and on the spatial distribution of these neutrons [fr

Application of the decoupling scheme on complex neutron-gamma shielding problems

Energy Technology Data Exchange (ETDEWEB)

Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

2000-03-01

Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

Application of neutrons in biology

International Nuclear Information System (INIS)

Cser, L.

1982-01-01

Applications of neutron scattering to determine the structure of biological macromolecules are reviewed. A theoretical and experimental introduction to neutron scattering and its mathematical description is given. The analysis of crystal structure using neutron scattering and the problem of Fourier reconstruction of structure are discussed. Some special problems concerning biological materials are described. The isotope effect of neutron scattering is applied to determine and identify the hydrogen atoms in biological macromolecules. Some examples illustrating the structure determination of amino acids and proteins are given. Mathematical methods of evaluation of small angle neutron scattering experiments and applications to investigate E. coli ribosome are described. New developments and new research trends are also reviewed. (D.Gy.)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Evaluation of Importance of Source Neutrons in Accelerator-Driven System

International Nuclear Information System (INIS)

Kim, Yong Hee; Park, Won Seok

2002-01-01

An importance function of the external spallation neutrons in ADS (Accelerator-Driven System) is defined to characterize the source multiplication in subcritical blanket. For a model ADS problem, the source importance function is evaluated with the TRANSX/TWODANT code system. In order to assess the impact of the power distribution on the importance function, both homogeneous and heterogeneous cores are analyzed and corresponding source multiplications are compared. Also, based on the source importance function, an optimization of the shape of the proton current is performed from the source multiplication point of view. Additionally, the source importance function is compared with the conventional λ-mode adjoint flux, which is used as an importance function of fission neutrons in the critical reactors. Concerning an issue in the ADS design, i.e., difficulty in reducing the fission power unless the proton current is shut off, a study is performed to minimize the source importance, thereby minimizing the fission power, even when the k-eff value of the core is quite high. (authors)

Refinement of the ferri and paramagnetic phases of magnetite measured by neutron multiple diffraction

International Nuclear Information System (INIS)

Mazzochi, V.L.; Parente, C.B.R.

1989-10-01

Structural parameters of the ferri and paramagnetic phases of magnetite have been refined from neutron multiple diffraction data. Experimental patterns were obtained by measuring the III primary reflection of a natural single crystal of this compound, at room temperature for the ferrimagnetic phase and 703 0 C for the paramagnetic phase. Theoretical multiple diffraction patterns for both phases have been calculated by the program MULTI which uses the iterative method. In this method intensities are caluclated as Taylor series expansions summed up to a order sufficient for a good approximation. A step by step process has been used in the refinements similarly to the parameter-shift method. Final values for the discrepancy factor found for the ferri and paramagnetic phases were R = 3.96% and R = 3.46%, respectively. (author) [pt

A Hybrid Genetic Algorithm for the Multiple Crossdocks Problem

Directory of Open Access Journals (Sweden)

Zhaowei Miao

2012-01-01

Full Text Available We study a multiple crossdocks problem with supplier and customer time windows, where any violation of time windows will incur a penalty cost and the flows through the crossdock are constrained by fixed transportation schedules and crossdock capacities. We prove this problem to be NP-hard in the strong sense and therefore focus on developing efficient heuristics. Based on the problem structure, we propose a hybrid genetic algorithm (HGA integrating greedy technique and variable neighborhood search method to solve the problem. Extensive experiments under different scenarios were conducted, and results show that HGA outperforms CPLEX solver, providing solutions in realistic timescales.

A search for solar neutron response in neutron monitor data

International Nuclear Information System (INIS)

Kudela, K.

1990-01-01

The search for an impulsive increase corresponding to a solar neutron response on high-mountain neutron monitors requires control of the stability of the measurement and elimination of other sources of short-time increases of different kinds which are involved in fluctuations of cosmic-ray intensity. For the solar flare of June 3, 1982 the excess of counting rate on the Lomnicky stit neutron monitor is, within a factor or 1.8, equal to that expected from solar neutrons. Superposed epoch analysis of 17 flares with gamma-ray or hard X-ray production gives a slight tendency of an occurring signal in cases of high heliocentric angles, indicating anisotropic production of neutrons on the sun. The low statistical significance of the result indicates that higher temporal resolution, better evaluation of multiplicity, better knowledge of the power spectra of short-term intensity fluctuations on neutron monitors, as well as coordinated measurements of solar gamma-rays and neutrons on satellites, are required. 21 refs

Multiplicity counting from fission detector signals with time delay effects

Science.gov (United States)

Nagy, L.; Pázsit, I.; Pál, L.

2018-03-01

In recent work, we have developed the theory of using the first three auto- and joint central moments of the currents of up to three fission chambers to extract the singles, doubles and triples count rates of traditional multiplicity counting (Pázsit and Pál, 2016; Pázsit et al., 2016). The objective is to elaborate a method for determining the fissile mass, neutron multiplication, and (α, n) neutron emission rate of an unknown assembly of fissile material from the statistics of the fission chamber signals, analogous to the traditional multiplicity counting methods with detectors in the pulse mode. Such a method would be an alternative to He-3 detector systems, which would be free from the dead time problems that would be encountered in high counting rate applications, for example the assay of spent nuclear fuel. A significant restriction of our previous work was that all neutrons born in a source event (spontaneous fission) were assumed to be detected simultaneously, which is not fulfilled in reality. In the present work, this restriction is eliminated, by assuming an independent, identically distributed random time delay for all neutrons arising from one source event. Expressions are derived for the same auto- and joint central moments of the detector current(s) as in the previous case, expressed with the singles, doubles, and triples (S, D and T) count rates. It is shown that if the time-dispersion of neutron detections is of the same order of magnitude as the detector pulse width, as they typically are in measurements of fast neutrons, the multiplicity rates can still be extracted from the moments of the detector current, although with more involved calibration factors. The presented formulae, and hence also the performance of the proposed method, are tested by both analytical models of the time delay as well as with numerical simulations. Methods are suggested also for the modification of the method for large time delay effects (for thermalised neutrons).

Genetic algorithms - A new technique for solving the neutron spectrum unfolding problem

International Nuclear Information System (INIS)

Freeman, David W.; Edwards, D. Ray; Bolon, Albert E.

1999-01-01

A new technique utilizing genetic algorithms has been applied to the Bonner sphere neutron spectrum unfolding problem. Genetic algorithms are part of a relatively new field of 'evolutionary' solution techniques that mimic living systems with computer-simulated 'chromosome' solutions. Solutions mate and mutate to create better solutions. Several benchmark problems, considered representative of radiation protection environments, have been evaluated using the newly developed UMRGA code which implements the genetic algorithm unfolding technique. The results are compared with results from other well-established unfolding codes. The genetic algorithm technique works remarkably well and produces solutions with relatively high spectral qualities. UMRGA appears to be a superior technique in the absence of a priori data - it does not rely on 'lucky' guesses of input spectra. Calculated personnel doses associated with the unfolded spectra match benchmark values within a few percent

Neutron Collar Evolution and Fresh PWR Assembly Measurements with a New Fast Neutron Passive Collar

Energy Technology Data Exchange (ETDEWEB)

Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Geist, William H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Root, Margaret A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Belian, Anthony P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-02

The passive neutron collar approach removes the effect of poison rods when using a 1mm Gd liner. This project sets out to solve the following challenges: BWR fuel assemblies have less mass and less neutron multiplication than PWR; and effective removal of cosmic ray spallation neutron bursts needed via QC tests.

RBE for late spinal cord injury following multiple fractions of neutrons

International Nuclear Information System (INIS)

Geraci, J.P.; Jackson, K.L.; Christensen, G.M.; Thrower, P.D.; Mariano, M.

1978-01-01

Using the length of the time interval between the irradiation of lumbosacral spinal cord of mice with ten fractions of either x rays or neutrons, and the onset of hindquarter paralysis, a fast neutron RBE of 3.5 for spinal cord damage at a neutron dose per fraction of 100 rad has been measured. This RBE for spinal cord injury is significant because it is larger than the RBE being used to calculate treatment doses in neutron radiotherapy

Designing and using multiple-possibility physics problems in physics courses

Science.gov (United States)

Shekoyan, Vazgen

2012-02-01

One important aspect of physics instruction is helping students develop better problem solving expertise. Besides enhancing the content knowledge, problems help students develop different cognitive abilities and skills. This presentation focuses on multiple-possibility problems (alternatively called ill-structured problems). These problems are different from traditional ``end of chapter'' single-possibility problems. They do not have one right answer and thus the student has to examine different possibilities, assumptions and evaluate the outcomes. To solve such problems one has to engage in a cognitive monitoring called epistemic cognition. It is an important part of thinking in real life. Physicists routinely use epistemic cognition when they solve problems. I have explored the instructional value of using such problems in introductory physics courses.

Reflected‑Point‑Reactor Kinetics Model for Neutron Coincidence Counting: Comments on the Equation for the Leakage Self‑Multiplication

International Nuclear Information System (INIS)

Croft, S.; McElroy, RD.; Favalli, A.; Hauck, D.; Henzlova, D.; Henzl, V.; Santi, PA.

2015-01-01

Passive neutron correlation counting is widely used, for example by international inspection agencies, for the non‑destructive assay of spontaneously fissile nuclear materials for nuclear safeguards. The mass of special nuclear material present in an item is usually estimated from the observed neutron counting rates by using equations based on mathematically describing the object as an isolated multiplying point‑like source. Calibration using representative physical standards can often adequately compensate for this theoretical oversimplification through the introduction and use of effective‑interpretational‑model‑parameters meaning that useful assay results are obtained. In this work we extend the point‑model treatment by including a simple reflector around the fissioning material. Specifically we show how the leakage self‑multiplication equation mathematically connects the traditional bare source and the reflected source cases. In doing so we explicitly demonstrate that although the presence of a simple reflector changes the leakage self‑multiplication the traditional bare‑item point model multiplicity equations retain the same mathematical form. Making and explaining this connection is important because it helps to explain and justify the practical success and use of the traditional point‑model equations even when the assumptions used to generate the key functional dependences are violated. We are not aware that this point has been recognized previously.

ISINN-5. 5. International seminar on interaction of neutrons with nuclei. Neutron spectroscopy, nuclear structure, related topics

International Nuclear Information System (INIS)

1997-01-01

The materials submitted at the fifth in a series of annual international seminar on interaction of neutrons with nuclei Neutron Spectroscopy, Nuclear Structure, Related Topics (ISINN-5) are given. The Seminar is organized by the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research and took place in Dubna on May 14-17, 1997. About 130 specialists from Belgium, China, Germany, France, Japan, Korea, Latvia, Netherlands, Ukraine, 7 Russian research institutes and a number of JINR laboratories took part in the Seminar. The scope of the problems discussed is traditionally wide. It includes the problems of violation of fundamental symmetries in the interaction of neutrons with nuclei, the properties of the neutron as the fundamental particle, nonstatistical aspects of the radiation capture of neutrons by nuclei, topical problems of the theory of nucleus, and the fission mechanism of heavy nuclei. The latest results obtained with ultracold neutrons (UCN), in particular, different approaches to understanding of the cause of UCN anomalous leakage through the walls of the trap are considered as well. The wide spectrum of methodological aspects of neutron-aided experiments is also discussed in details

Neutron/gamma dose separation by the multiple-ion-chamber technique

International Nuclear Information System (INIS)

Goetsch, S.J.

1983-01-01

Many mixed n/γ dosimetry systems rely on two dosimeters, one composed of a tissue-equivalent material and the other made from a non-hydrogenous material. The paired chamber technique works well in fields of neutron radiation nearly identical in spectral composition to that in which the dosimeters were calibrated. However, this technique is drastically compromised in phantom due to the degradation of the neutron spectrum. The three-dosimeter technique allows for the fall-off in neutron sensitivity of the two non-hydrogenous dosimeters. Precise and physically meaningful results were obtained with this technique with a D-T source in air and in phantom and with simultaneous D-T neutron and 60 Co gamma ray irradiation in air. The MORSE-CG coupled n/γ three-dimensional Monte Carlo code was employed to calculate neutron and gamma doses in a water phantom. Gamma doses calculated in phantom with this code were generally lower than corresponding ion chamber measurements. This can be explained by the departure of irradiation conditions from ideal narrow-beam geometry. 97 references

Development of neutron detectors for neutron scattering experiments

Energy Technology Data Exchange (ETDEWEB)

Moon, Myungkook; Kim, Jongyul; Kim, Jeong ho; Lee, Suhyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Changhwy [Korea Research Institute of Ships and Ocean Engineering, Daejeon (Korea, Republic of)

2015-10-15

Various kinds of detectors are used in accordance with the experimental purpose, such as zero dimensional detector, 1-D or 2-D position-sensitive detectors. Most of neutron detectors use He-3 gas because of its high neutron sensitivity. Since the He-3 supply shortage took place in early 2010, various He-3 alternative detectors have been developed even for the other neutron application. We have developed a new type alternative detector on the basis of He-3 detector technology. Although B- 10 has less neutron detection efficiency compared with He-3, it can be covered by the use of multiple B-10 layers. In this presentation, we would like to introduce the neutron detectors under development and developed detectors. Various types of detector were successfully developed and result of the technical test performance is promising. Even though the detection efficiency of the B-10 detector lower than He-3 one, the continuous research and development is needed for currently not available He-3.

Solution of Constrained Optimal Control Problems Using Multiple Shooting and ESDIRK Methods

DEFF Research Database (Denmark)

Capolei, Andrea; Jørgensen, John Bagterp

2012-01-01

of this paper is the use of ESDIRK integration methods for solution of the initial value problems and the corresponding sensitivity equations arising in the multiple shooting algorithm. Compared to BDF-methods, ESDIRK-methods are advantageous in multiple shooting algorithms in which restarts and frequent...... algorithm. As we consider stiff systems, implicit solvers with sensitivity computation capabilities for initial value problems must be used in the multiple shooting algorithm. Traditionally, multi-step methods based on the BDF algorithm have been used for such problems. The main novel contribution...... discontinuities on each shooting interval are present. The ESDIRK methods are implemented using an inexact Newton method that reuses the factorization of the iteration matrix for the integration as well as the sensitivity computation. Numerical experiments are provided to demonstrate the algorithm....

Theory of stochastic space-dependent neutron kinetics with a Gaussian parametric excitation

International Nuclear Information System (INIS)

Saito, K.

1980-01-01

Neutron kinetics and statics in a multiplying medium with a statistically fluctuating reactivity are unified and systematically studied by applying the Novikov-Furutsu formula. The parametric or multiplicative noise is spatially distributed and of Gaussian nature with an arbitrary spectral profile. It is found that the noise introduces a new definite production term into the conventional balance equation for the mean neutron number. The term is characterized by the magnitude and the correlation function of the random excitation. Its relaxation phenomena bring forth a non-Markoffian or a memory effect, which is conceptualised by introducing 'pseudo-precursors' or 'pseudo-delayed neutrons'. By using the concept, some typical reactor physical problems are solved; they are (1) reactivity and flux perturbation originating from the random dispersal of core materials and (2) analysis of neutron decay mode and it relaxation constant, and derivation of the corresponding new inhour equation. (author)

The analysis of neutron physical characteristics of fast reactors by means of pulsed experiments

International Nuclear Information System (INIS)

Stumbur, Eh.A.; Milyutina, Z.N.

1992-01-01

Possibility is considered for determination of macroscopic cross sections of homogeneous multiplicating media with fast neutrons. It is shown that by means of the critical size, laplaccian and neutron pulse damping decrement measurement results it is possible to obtain values of almost all cross sections of a medium. The method is tested with systems of metal 235 U and BFS-32 assemblies with the composition, typical for fast power reactors. A suitable algorithm is developed for solving nonstationary asymptotic transport problems. Calculation results are compared with experimental ones. 21 refs.; 2 figs.; 3 tabs

DORT and TORT workshop -- Outline for presentation for performance issues for large problems

International Nuclear Information System (INIS)

Barnett, A.

1998-04-01

This paper addresses the problem of running large TORT programs. The problem being a limited amount of time per job and limited amount of memory and disk space. The solution that the author outlines here is to break up the TORT run in time and space. For the time problem run multiple, sequential, dependent jobs. For the space problem use TORT internal memory conservation features. TORT is a three-dimensional discrete ordinates neutron/photon transport code

The Multiple-Minima Problem in Protein Folding

Science.gov (United States)

Scheraga, Harold A.

1991-10-01

The conformational energy surface of a polypeptide or protein has many local minima, and conventional energy minimization procedures reach only a local minimum (near the starting point of the optimization algorithm) instead of the global minimum (the multiple-minima problem). Several procedures have been developed to surmount this problem, the most promising of which are: (a) build up procedure, (b) optimization of electrostatics, (c) Monte Carlo-plus-energy minimization, (d) electrostatically-driven Monte Carlo, (e) inclusion of distance restraints, (f) adaptive importance-sampling Monte Carlo, (g) relaxation of dimensionality, (h) pattern-recognition, and (i) diffusion equation method. These procedures have been applied to a variety of polypeptide structural problems, and the results of such computations are presented. These include the computation of the structures of open-chain and cyclic peptides, fibrous proteins and globular proteins. Present efforts are being devoted to scaling up these procedures from small polypeptides to proteins, to try to compute the three-dimensional structure of a protein from its amino sequence.

DISCUS, Neutron Single to Double Scattering Ratio in Inelastic Scattering Experiment by Monte-Carlo

International Nuclear Information System (INIS)

Johnson, M.W.

1993-01-01

1 - Description of problem or function: DISCUS calculates the ratio of once-scattered to twice-scattered neutrons detected in an inelastic neutron scattering experiment. DISCUS also calculates the flux of once-scattered neutrons that would have been observed if there were no absorption in the sample and if, once scattered, the neutron would emerge without further re-scattering or absorption. Three types of sample geometry are used: an infinite flat plate, a finite flat plate or a finite length cylinder. (The infinite flat plate is included for comparison with other multiple scattering programs.) The program may be used for any sample for which the scattering law is of the form S(/Q/, omega). 2 - Method of solution: Monte Carlo with importance sampling is used. Neutrons are 'forced' both into useful angular trajectories, and useful energy bins. Biasing of the collision point according to the point of entry of the neutron into the sample is also utilised. The first and second order scattered neutron fluxes are calculated in independent histories. For twice-scattered neutron histories a square distribution in Q-omega space is used to sample the neutron coming from the first scattering event, whilst biasing is used for the second scattering event. (A square distribution is used so as to obtain reasonable inelastic-inelastic statistics.) 3 - Restrictions on the complexity of the problem: Unlimited number of detectors. Max. size of (Q, omega) matrix is 39*149. Max. number of points in momentum space for the scattering cross section is 199

A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy

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)

Neutron crosstalk between liquid scintillators

Energy Technology Data Exchange (ETDEWEB)

Verbeke, J.M., E-mail: verbeke2@llnl.gov; Prasad, M.K., E-mail: prasad1@llnl.gov; Snyderman, N.J., E-mail: snyderman1@llnl.gov

2015-09-11

A method is proposed to quantify the fractions of neutrons scattering between liquid scintillators. Using a spontaneous fission source, this method can be utilized to quickly characterize an array of liquid scintillators in terms of crosstalk. The point model theory due to Feynman is corrected to account for these multiple scatterings. Using spectral information measured by the liquid scintillators, fractions of multiple scattering can be estimated, and mass reconstruction of fissile materials under investigation can be improved. Monte Carlo simulations of mono-energetic neutron sources were performed to estimate neutron crosstalk. A californium source in an array of liquid scintillators was modeled to illustrate the improvement of the mass reconstruction.

Neutron crosstalk between liquid scintillators

International Nuclear Information System (INIS)

Verbeke, J.M.; Prasad, M.K.; Snyderman, N.J.

2015-01-01

A method is proposed to quantify the fractions of neutrons scattering between liquid scintillators. Using a spontaneous fission source, this method can be utilized to quickly characterize an array of liquid scintillators in terms of crosstalk. The point model theory due to Feynman is corrected to account for these multiple scatterings. Using spectral information measured by the liquid scintillators, fractions of multiple scattering can be estimated, and mass reconstruction of fissile materials under investigation can be improved. Monte Carlo simulations of mono-energetic neutron sources were performed to estimate neutron crosstalk. A californium source in an array of liquid scintillators was modeled to illustrate the improvement of the mass reconstruction

Analysis of (n, 2n) multiplication in lead

International Nuclear Information System (INIS)

Segev, M.

1984-01-01

Lead is being considered as a possible amplifier of neutrons for fusion blankets. A simple one-group model of neutron multiplications in Pb is presented. Given the 14 MeV neutron cross section on Pb, the model predicts the multiplication. Given measured multiplications, the model enables the determination of the (n, 2n) and transport cross sections. Required for the model are: P-the collision probability for source neutrons in the Pb body-and W- an average collision probability for non-virgin, non-degraded neutrons. In simple geometries, such as a source in the center of a spherical shell, P and an approximate W can be expressed analytically in terms of shell dimensions and the Pb transport cross section. The model was applied to Takahashi's measured multiplications in Pb shells in order to understand the apparent very high multiplicative power of Pb. The results of the analysis are not consistent with basic energy-balance and cross section magnitude constraints in neutron interaction theory. (author)

Some properties of the neutron monochromatic beams obtained by multiple Bragg reflections realized in bent perfect single crystals

Czech Academy of Sciences Publication Activity Database

Mikula, Pavol; Vrána, Miroslav; Šaroun, Jan; Krejčí, F.; Seong, B. S.; Woo, W.; Furusaka, M.

2013-01-01

Roč. 46, č. 1 (2013), s. 128-134 ISSN 0021-8898 R&D Projects: GA ČR GAP204/10/0654; GA MŠk LM2010011 Institutional support: RVO:61389005 Keywords : multiple reflections * bent perfect crystals * neutron diffraction Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 3.950, year: 2013

Influence of Number Size, Problem Structure and Response Mode on Children's Solutions of Multiplication Word Problems.

Science.gov (United States)

De Corte, E.; And Others

One important finding from recent research on multiplication word problems is that children's performances are strongly affected by the nature of the multiplier (whether it is an integer, decimal larger than 1 or a decimal smaller than 1). On the other hand, the size of the multiplicand has little or no effect on problem difficulty. The aim of the…

Numerical estimates of multiple reaction corrections in neutron cross-section measurements

International Nuclear Information System (INIS)

Magnusson, G.

1979-04-01

A method to evaluate the effect of secondary neutrons in 14-15 MeV neutron cross-section measurements is presented. The emission spectra of secondary neutrons are calculated by means of the preequilibrium and statistical models. An expression for the collision probability in a homogenous body has been utilized in the calculations. (author)

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

International Nuclear Information System (INIS)

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

2017-01-01

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

A New Neutron Multiplicity Counter for the Measurement of Impure Plutonium Metal at Westinghouse Savannah River Site

International Nuclear Information System (INIS)

Baker, L.B.; Faison, D.M.; Langner, D.G.; Sweet, M.R.; Salazar, S.D.; Kroncke, K.E.

1998-07-01

A new neutron multiplicity counter has been designed, fabricated, characterized, and installed for use in the assay of impure plutonium metal buttons from the FB-Line at the Westinghouse Savannah River Site (WSRS). This instrument incorporates the performance characteristics of the Pyrochemical or In-plant Multiplicity Counter with the package size of the Plutonium Scrap Multiplicity Counter. In addition, state-of-the art features such as the de-randomizer circuit and separate ring outputs have been added. The counter consists of 113, 71 cm active length 3He tubes in a polyethylene moderator. Its efficiency for 252Cf is 57.8 percent, the highest of any multiplicity counter to date. Its die-away time is 50.4 ms and its deadtime is 50 ns. In this paper we will present the characterization data for the counter and the results of preliminary metal measurements at WSRS. We will also discuss the new challenges the impure metal buttons from FB-Line are presenting to the multiplicity counting technique

Neutrons for Catalysis: A Workshop on Neutron Scattering Techniques for Studies in Catalysis

International Nuclear Information System (INIS)

Overbury, Steven H.; Coates, Leighton; Herwig, Kenneth W.; Kidder, Michelle

2011-01-01

This report summarizes the Workshop on Neutron Scattering Techniques for Studies in Catalysis, held at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) on September 16 and 17, 2010. The goal of the Workshop was to bring experts in heterogeneous catalysis and biocatalysis together with neutron scattering experimenters to identify ways to attack new problems, especially Grand Challenge problems in catalysis, using neutron scattering. The Workshop locale was motivated by the neutron capabilities at ORNL, including the High Flux Isotope Reactor (HFIR) and the new and developing instrumentation at the SNS. Approximately 90 researchers met for 1 1/2 days with oral presentations and breakout sessions. Oral presentations were divided into five topical sessions aimed at a discussion of Grand Challenge problems in catalysis, dynamics studies, structure characterization, biocatalysis, and computational methods. Eleven internationally known invited experts spoke in these sessions. The Workshop was intended both to educate catalyst experts about the methods and possibilities of neutron methods and to educate the neutron community about the methods and scientific challenges in catalysis. Above all, it was intended to inspire new research ideas among the attendees. All attendees were asked to participate in one or more of three breakout sessions to share ideas and propose new experiments that could be performed using the ORNL neutron facilities. The Workshop was expected to lead to proposals for beam time at either the HFIR or the SNS; therefore, it was expected that each breakout session would identify a few experiments or proof-of-principle experiments and a leader who would pursue a proposal after the Workshop. Also, a refereed review article will be submitted to a prominent journal to present research and ideas illustrating the benefits and possibilities of neutron methods for catalysis research.

Neutrons for Catalysis: A Workshop on Neutron Scattering Techniques for Studies in Catalysis

Energy Technology Data Exchange (ETDEWEB)

Overbury, Steven {Steve} H [ORNL; Coates, Leighton [ORNL; Herwig, Kenneth W [ORNL; Kidder, Michelle [ORNL

2011-10-01

This report summarizes the Workshop on Neutron Scattering Techniques for Studies in Catalysis, held at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) on September 16 and 17, 2010. The goal of the Workshop was to bring experts in heterogeneous catalysis and biocatalysis together with neutron scattering experimenters to identify ways to attack new problems, especially Grand Challenge problems in catalysis, using neutron scattering. The Workshop locale was motivated by the neutron capabilities at ORNL, including the High Flux Isotope Reactor (HFIR) and the new and developing instrumentation at the SNS. Approximately 90 researchers met for 1 1/2 days with oral presentations and breakout sessions. Oral presentations were divided into five topical sessions aimed at a discussion of Grand Challenge problems in catalysis, dynamics studies, structure characterization, biocatalysis, and computational methods. Eleven internationally known invited experts spoke in these sessions. The Workshop was intended both to educate catalyst experts about the methods and possibilities of neutron methods and to educate the neutron community about the methods and scientific challenges in catalysis. Above all, it was intended to inspire new research ideas among the attendees. All attendees were asked to participate in one or more of three breakout sessions to share ideas and propose new experiments that could be performed using the ORNL neutron facilities. The Workshop was expected to lead to proposals for beam time at either the HFIR or the SNS; therefore, it was expected that each breakout session would identify a few experiments or proof-of-principle experiments and a leader who would pursue a proposal after the Workshop. Also, a refereed review article will be submitted to a prominent journal to present research and ideas illustrating the benefits and possibilities of neutron methods for catalysis research.

Neutron Absorbing Ability Variation in Neutron Absorbing Material Caused by the Neutron Irradiation in Spent Fuel Storage Facility

Energy Technology Data Exchange (ETDEWEB)

Sohn, Hee Dong; Han, Seul Gi; Lee, Sang Dong; Kim, Ki Hong; Ryu, Eag Hyang; Park, Hwa Gyu [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

2014-10-15

In spent fuel storage facility like high density spent fuel storage racks and dry storage casks, spent fuels are stored with neutron absorbing materials installed as a part of those facilities, and they are used for absorbing neutrons emitted from spent fuels. Usually structural material with neutron absorbing material of racks and casks are located around spent fuels, so it is irradiated by neutrons for long time. Neutron absorbing ability could be changed by the variation of nuclide composition in neutron absorbing material caused by the irradiation of neutrons. So, neutron absorbing materials are continuously faced with spent fuels with boric acid solution or inert gas environment. Major nuclides in neutron absorbing material are Al{sup 27}, C{sup 12}, B{sup 11}, B{sup 10} and they are changed to numerous other ones as radioactive decay or neutron absorption reaction. The B{sup 10} content in neutron absorbing material dominates the neutron absorbing ability, so, the variation of nuclide composition including the decrease of B{sup 10} content is the critical factor on neutron absorbing ability. In this study, neutron flux in spent fuel, the activation of neutron absorbing material and the variation of nuclide composition are calculated. And, the minimum neutron flux causing the decrease of B{sup 10} content is calculated in spent fuel storage facility. Finally, the variation of neutron multiplication factor is identified according to the one of B{sup 10} content in neutron absorbing material. The minimum neutron flux to impact the neutron absorbing ability is 10{sup 10} order, however, usual neutron flux from spent fuel is 10{sup 8} order. Therefore, even though neutron absorbing material is irradiated for over 40 years, B{sup 10} content is little decreased, so, initial neutron absorbing ability could be kept continuously.

Heuristic for Solving the Multiple Alignment Sequence Problem

Directory of Open Access Journals (Sweden)

Roman Anselmo Mora Gutiérrez

2011-03-01

Full Text Available In this paper we developed a new algorithm for solving the problem of multiple sequence alignment (AM S, which is a hybrid metaheuristic based on harmony search and simulated annealing. The hybrid was validated with the methodology of Julie Thompson. This is a basic algorithm and and results obtained during this stage are encouraging.

Multi-element neutron activation analysis and solution of classification problems using multidimensional statistics

International Nuclear Information System (INIS)

Vaganov, P.A.; Kol'tsov, A.A.; Kulikov, V.D.; Mejer, V.A.

1983-01-01

The multi-element instrumental neutron activation analysis of samples of mountain rocks (sandstones, aleurolites and shales of one of gold deposits) is performed. The spectra of irradiated samples are measured by Ge(Li) detector of the volume of 35 mm 3 . The content of 22 chemical elements is determined in each sample. The results of analysis serve as reliable basis for multi-dimensional statistic information processing, they constitute the basis for the generalized characteristics of rocks which brings about the solution of classification problem for rocks of different deposits

Exact and approximate interior corner problem in neutron diffusion by integral transform methods

International Nuclear Information System (INIS)

Bareiss, E.H.; Chang, K.S.J.; Constatinescu, D.A.

1976-09-01

The mathematical solution of the neutron diffusion equation exhibits singularities in its derivatives at material corners. A mathematical treatment of the nature of these singularities and its impact on coarse network approximation methods in computational work is presented. The mathematical behavior is deduced from Green's functions, based on a generalized theory for two space dimensions, and the resulting systems of integral equations, as well as from the Kontorovich--Lebedev Transform. The effect on numerical calculations is demonstrated for finite difference and finite element methods for a two-region corner problem

Individual neutron dosimetry

International Nuclear Information System (INIS)

Mauricio, C.L.P.

1987-01-01

The most important concepts and development in individual neutron dosimetry are presented, especially the dosimetric properties of the albedo technique. The main problem in albedo dosimetry is to calibrate the dosemeter in the environs of each neutron source. Some of the most used calibration techniques are discussed. The IRD albedo dosemeter used in the routine neutron individual monitoring is described in detail. Its dosimetric properties and calibration methods are discussed. (Author) [pt

RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability

International Nuclear Information System (INIS)

Ondis, L.A. II; Tyburski, L.J.; Moskowitz, B.S.

2000-01-01

The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations

RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability

Energy Technology Data Exchange (ETDEWEB)

Ondis, L.A., II; Tyburski, L.J.; Moskowitz, B.S.

2000-03-01

The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations.

Step by Step: Biology Undergraduates’ Problem-Solving Procedures during Multiple-Choice Assessment

Science.gov (United States)

Prevost, Luanna B.; Lemons, Paula P.

2016-01-01

This study uses the theoretical framework of domain-specific problem solving to explore the procedures students use to solve multiple-choice problems about biology concepts. We designed several multiple-choice problems and administered them on four exams. We trained students to produce written descriptions of how they solved the problem, and this allowed us to systematically investigate their problem-solving procedures. We identified a range of procedures and organized them as domain general, domain specific, or hybrid. We also identified domain-general and domain-specific errors made by students during problem solving. We found that students use domain-general and hybrid procedures more frequently when solving lower-order problems than higher-order problems, while they use domain-specific procedures more frequently when solving higher-order problems. Additionally, the more domain-specific procedures students used, the higher the likelihood that they would answer the problem correctly, up to five procedures. However, if students used just one domain-general procedure, they were as likely to answer the problem correctly as if they had used two to five domain-general procedures. Our findings provide a categorization scheme and framework for additional research on biology problem solving and suggest several important implications for researchers and instructors. PMID:27909021

Describing function theory as applied to thermal and neutronic problems

International Nuclear Information System (INIS)

Nassersharif, B.

1983-01-01

Describing functions have traditionally been used to obtain the solutions of systems of ordinary differential equations. In this work the describing function concept has been extended to include nonlinear, distributed parameter partial differential equations. A three-stage solution algorithm is presented which can be applied to any nonlinear partial differential equation. Two generalized integral transforms were developed as the T-transform for the time domain and the B-transform for the spatial domain. The thermal diffusion describing function (TDDF) is developed for conduction of heat in solids and a general iterative solution along with convergence criteria is presented. The proposed solution method is used to solve the problem of heat transfer in nuclear fuel rods with annular fuel pellets. As a special instance the solid cylindrical fuel pellet is examined. A computer program is written which uses the describing function concept for computing fuel pin temperatures in the radial direction during reactor transients. The second problem investigated was the neutron diffusion equation which is intrinsically different from the first case. Although, for most situations, it can be treated as a linear differential equation, the describing function method is still applicable. A describing function solution is derived for two possible cases: constant diffusion coefficient and variable diffusion coefficient. Two classes of describing functions are defined for each case which portray the leakage and absorption phenomena. For the specific case of a slab reactor criticality problem the comparison between analytical and describing function solutions revealed an excellent agreement

On some one-speed neutron transport problems revisited and reformulated

International Nuclear Information System (INIS)

Williams, M.M.R.

2001-01-01

The solution of a number of one-speed neutron transport problems involving infinite media have been re-considered in the light of a transformation first used by Wallace (Wallace, P.R., 1944a. Boundary Conditions at Thin Absorbing Shells and Plates I. Canadian National Research Council Report MT-34; Wallace, P.R., 1944b. On the Thermal Utilisation of Plates in the Presence of Linear Anisotropic Scattering. Canadian National Research Council Report MT-63). The outcome of this transformation is that the infinite medium problem can be reduced to one in terms of an integral equation involving finite regions only. For example, in the case of an infinitely reflected slab, the infinite reflector is removed and its presence transferred to the kernel of a new integral equation. These kernels turn out to be the point or plane kernels of the corresponding infinite medium problem in the pure reflector material. In this paper the method is extended to slabs with arbitrary anisotropic scattering in slab and reflector; it is also applied to reflected spheres. In this case however, there is a limitation that the total mean free path in sphere and reflector be the same. Finally, we comment on the physical meaning of the standard anisotropic formalism and show that a more realistic eigenvalue exists which is directly related to the isotropic fission source. Some numerical results are given to illustrate our conclusions

Finite element method for neutron diffusion problems in hexagonal geometry

International Nuclear Information System (INIS)

Wei, T.Y.C.; Hansen, K.F.

1975-06-01

The use of the finite element method for solving two-dimensional static neutron diffusion problems in hexagonal reactor configurations is considered. It is investigated as a possible alternative to the low-order finite difference method. Various piecewise polynomial spaces are examined for their use in hexagonal problems. The central questions which arise in the design of these spaces are the degree of incompleteness permissible and the advantages of using a low-order space fine-mesh approach over that of a high-order space coarse-mesh one. There is also the question of the degree of smoothness required. Two schemes for the construction of spaces are described and a number of specific spaces, constructed with the questions outlined above in mind, are presented. They range from a complete non-Lagrangian, non-Hermite quadratic space to an incomplete ninth order space. Results are presented for two-dimensional problems typical of a small high temperature gas-cooled reactor. From the results it is concluded that the space used should at least include the complete linear one. Complete spaces are to be preferred to totally incomplete ones. Once function continuity is imposed any additional degree of smoothness is of secondary importance. For flux shapes typical of the small high temperature gas-cooled reactor the linear space fine-mesh alternative is to be preferred to the perturbation quadratic space coarse-mesh one and the low-order finite difference method is to be preferred over both finite element schemes

Bilevel formulation of a policy design problem considering multiple objectives and incomplete preferences

Science.gov (United States)

Hawthorne, Bryant; Panchal, Jitesh H.

2014-07-01

A bilevel optimization formulation of policy design problems considering multiple objectives and incomplete preferences of the stakeholders is presented. The formulation is presented for Feed-in-Tariff (FIT) policy design for decentralized energy infrastructure. The upper-level problem is the policy designer's problem and the lower-level problem is a Nash equilibrium problem resulting from market interactions. The policy designer has two objectives: maximizing the quantity of energy generated and minimizing policy cost. The stakeholders decide on quantities while maximizing net present value and minimizing capital investment. The Nash equilibrium problem in the presence of incomplete preferences is formulated as a stochastic linear complementarity problem and solved using expected value formulation, expected residual minimization formulation, and the Monte Carlo technique. The primary contributions in this article are the mathematical formulation of the FIT policy, the extension of computational policy design problems to multiple objectives, and the consideration of incomplete preferences of stakeholders for policy design problems.

$\\gamma$-ray energy spectra and multiplicities from the neutron-induced fission of $^{235}$U using STEFF

CERN Document Server

An experiment is proposed to use the STEFF spectrometer at n_TOF to study fragment $\\gamma$-correlations following the neutron-induced fission of $^{235}$U. The STEFF array of 12 NaI detectors will allow measurements of the single $\\gamma$-energy, the $\\gamma$ multiplicity, and the summed $\\gamma$energy distributions as a function of the mass and charge split, and deduced excitation energy in the fission event. These data will be used to study the origin of fission-fragment angular momenta, examining angular distribution eects as a function of incident neutron energy. The principal application of this work is in meeting the NEA high-priority request for improved $\\gamma$ray data from $^{235}$U(n; F). To improve the detection rate and expand the range of detection angles, STEFF will be modied to include two new ssion-fragment detectors each at 45 to the beam direction.

On the possibility of multiple utilization of Bowen's Kale for neutron activation analysis of biological materials

International Nuclear Information System (INIS)

Marinov, V.M.; Lazarova, M.S.; Mihajlov, M.I.; Apostolov, D.

1977-01-01

The results of investigations related to the multiple utilization of Bowen's Kale in developing neutron-activation methods for determining microelements in biological materials carried out in recent years are presented. Bowen's Kale might be used as: (1) experimental material in the development of a method and its verification, i.e. as a test for biological materials; (2) a material where experimental conditions might be optimized; (3) a material for investigating the accuracy, reproducibility and the limit of proof at experimental conditions already defined; (4) a monitor; (5) a multielement volume reference standard for a number of microelements during their simultaneous determination and (6) a standard for verifying the authenticity of the results obtained. In this manner, a reliable criterion for comparison of the potentialities, the accuracy, reproducibility, the limits of proof and the authenticity of the neutron-activation methods of determining microelements in biological materials is introduced. (author)

Neutron radiography in metallurgy

International Nuclear Information System (INIS)

Rant, J.; Ilic, R.

1977-01-01

The review surveys microneutronographic and neutron-induced autoradiographic techniques and their applications in metallurgy. A brief survey of applications of neutron radiography as a method of non-destructive testing to some macroscopic problems in metallurgy is included. (author)

One-dimensional computational modeling on nuclear reactor problems

International Nuclear Information System (INIS)

Alves Filho, Hermes; Baptista, Josue Costa; Trindade, Luiz Fernando Santos; Heringer, Juan Diego dos Santos

2013-01-01

In this article, we present a computational modeling, which gives us a dynamic view of some applications of Nuclear Engineering, specifically in the power distribution and the effective multiplication factor (keff) calculations. We work with one-dimensional problems of deterministic neutron transport theory, with the linearized Boltzmann equation in the discrete ordinates (SN) formulation, independent of time, with isotropic scattering and then built a software (Simulator) for modeling computational problems used in a typical calculations. The program used in the implementation of the simulator was Matlab, version 7.0. (author)

Neutron flux stabilization in the NG-150 neutron generators

International Nuclear Information System (INIS)

Kuz'min, L.E.; Makarov, S.A.; Pronman, I.M.

1986-01-01

Problem of metal tritium target lifetime increase and neutron flux stabilization in the NG-150 neutron generators is studied. Possibility on neutron flux stabilization using the mass analyzer for low-angle (4 deg and 41 deg) mass separation of a beam in thre components, which fall on a target simultaneously, is confirmed experimentally. Basic generator parameters are: accelerating voltage of 150 kV, total beam current on a target of 1.5 mA, beam current density of 0.3-1.6 mA/cm 2 , beam diameter of 8 mm. The initial neutron flux on the targets of 0.73 mg/cm 2 thick constituted 1.1x10 11 ssup(-1). The neutron flux monitoring was accomplished from recoil proton recording by a plastic scintillator. Flux decrease by more than 5% served as a signel for measuring mass analyzer magnetic field providing beam displacement on a target and restoration of the given flux. The NG-150 generator neutron flux stabilization was attained during 2h

Reevaluation of the average prompt neutron emission multiplicity (nubar) values from fission of uranium and transuranium nuclides

International Nuclear Information System (INIS)

Holden, N.E.; Zucker, M.S.

1984-01-01

In response to a need of the safeguards community, we have begun an evaluation effort to upgrade the recommended values of the prompt neutron emission multiplicity distribution, P/sub nu/ and its average value, nubar. This paper will report on progress achieved thus far. The evaluation of the uranium, plutonium, americium and curium nuclide's nubar values will be presented. The recommended values will be given and discussed. 61 references

[Supporting parenting in families with multiple problems].

Science.gov (United States)

Le Foll, Julie

2015-01-01

Supporting parenthood in families with multiple problems is a major early prevention challenge. Indeed, the factors of vulnerability, especially if they mount up, expose the child to an increased risk of a somatic pathology, developmental delays, learning difficulties and maltreatment. In order to limit the impact of these vulnerabilities on the health of mothers and infants, it is essential to act early, to adapt the working framework and to collaborate within a network. Copyright © 2015 Elsevier Masson SAS. All rights reserved.

A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy

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)

Neutron structural biology

Energy Technology Data Exchange (ETDEWEB)

Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Neutron diffraction provides an experimental method of directly locating hydrogen atoms in protein which play important roles in physiological functions. However, there are relatively few examples of neutron crystallography in biology since it takes a lot of time to collect a sufficient number of Bragg reflections due to the low flux of neutrons illuminating the sample. In order to overcome the flux problem, we have successfully developed the neutron IP, where the neutron converter, {sup 6}Li or Gd, was mixed with a photostimulated luminescence material on flexible plastic support. Neutron Laue diffraction 2A data from tetragonal lysozyme were collected for 10 days with neutron imaging plates, and 960 hydrogen atoms in the molecule and 157 bound water molecules were identified. These results explain the proposed hydrolysis mechanism of the sugar by the lysozyme molecule and that lysozyme is less active at pH7.0. (author)

On the problem of monitoring the neutron parameters of the Fast Energy Amplifier

International Nuclear Information System (INIS)

Behringer, K.; Wydler, P.

1998-10-01

The conceptual Fast Energy Amplifier, proposed by Rubbia et al. (1995), consists of a combination of a U-233/Th-232 fuelled fast-neutron subcritical facility with a proton accelerator. An intense beam of 1 GeV protons is injected into liquid lead at the core centre and drives the reactor by producing spallation neutrons. The burst of spallation neutrons produced by a single proton alters the basic neutron statistics which are well known for thermal neutrons in conventional nuclear reactors. A short assessment of standard neutron noise analysis methods is made with respect to monitoring neutron parameter data. (author)

Finding Multiple Optimal Solutions to Optimal Load Distribution Problem in Hydropower Plant

Directory of Open Access Journals (Sweden)

Xinhao Jiang

2012-05-01

Full Text Available Optimal load distribution (OLD among generator units of a hydropower plant is a vital task for hydropower generation scheduling and management. Traditional optimization methods for solving this problem focus on finding a single optimal solution. However, many practical constraints on hydropower plant operation are very difficult, if not impossible, to be modeled, and the optimal solution found by those models might be of limited practical uses. This motivates us to find multiple optimal solutions to the OLD problem, which can provide more flexible choices for decision-making. Based on a special dynamic programming model, we use a modified shortest path algorithm to produce multiple solutions to the problem. It is shown that multiple optimal solutions exist for the case study of China’s Geheyan hydropower plant, and they are valuable for assessing the stability of generator units, showing the potential of reducing occurrence times of units across vibration areas.

Integrated Production-Distribution Scheduling Problem with Multiple Independent Manufacturers

Directory of Open Access Journals (Sweden)

Jianhong Hao

2015-01-01

Full Text Available We consider the nonstandard parts supply chain with a public service platform for machinery integration in China. The platform assigns orders placed by a machinery enterprise to multiple independent manufacturers who produce nonstandard parts and makes production schedule and batch delivery schedule for each manufacturer in a coordinate manner. Each manufacturer has only one plant with parallel machines and is located at a location far away from other manufacturers. Orders are first processed at the plants and then directly shipped from the plants to the enterprise in order to be finished before a given deadline. We study the above integrated production-distribution scheduling problem with multiple manufacturers to maximize a weight sum of the profit of each manufacturer under the constraints that all orders are finished before the deadline and the profit of each manufacturer is not negative. According to the optimal condition analysis, we formulate the problem as a mixed integer programming model and use CPLEX to solve it.

Summary of neutron measurements for the Viking Program

International Nuclear Information System (INIS)

Anderson, M.E.

1975-01-01

The results of neutron measurements for 238 Pu-fueled, 683-W (thermal) capsules fabricated for the Viking Program (Mars Lander) are presented. These results include, for each capsule, the total neutron emission rate and neutron multiplication and, for one capsule, the neutron energy spectrum. A precision long counter was used for the neutron emission rate measurements and a single stilbene crystal for the neutron spectrum measurement. (U.S.)

The Core Problem within a Linear Approximation Problem $AX/approx B$ with Multiple Right-Hand Sides

Czech Academy of Sciences Publication Activity Database

Hnětynková, Iveta; Plešinger, Martin; Strakoš, Z.

2013-01-01

Roč. 34, č. 3 (2013), s. 917-931 ISSN 0895-4798 R&D Projects: GA ČR GA13-06684S Grant - others:GA ČR(CZ) GA201/09/0917; GA MŠk(CZ) EE2.3.09.0155; GA MŠk(CZ) EE2.3.30.0065 Program:GA Institutional support: RVO:67985807 Keywords : total least squares problem * multiple right-hand sides * core problem * linear approximation problem * error-in-variables modeling * orthogonal regression * singular value decomposition Subject RIV: BA - General Mathematics Impact factor: 1.806, year: 2013

Neutronics codes

International Nuclear Information System (INIS)

Buckel, G.

1983-01-01

The objectives are the development, testing and cultivation of reliable, efficient and user-optimized neutron-physical calculation methods and conformity with users' requirements concerning design of power reactors, planning and analysis of experiments necessary for their protection as well as research on physical key problems. A short outline of available computing programmes for the following objectives is given: - Provision of macroscopic group constants, - Calculation of neutron flux distribution in transport theory and diffusion approximation, - Evaluation of neutron flux-distribution, - Execution of disturbance calculations for the determination reactivity coefficients, and - graphical representation of results. (orig./RW) [de

Base neutron noise in PWRs

International Nuclear Information System (INIS)

Kosaly, G.; Albrecht, R.W.; Dailey, D.J.; Fry, D.N.

1981-01-01

Considerable activity has been devoted in recent years to the use of neutron noise for investigation of problems in pressurized-water reactors (PWRs). The investigators have found that neutron noise provides an effective way to monitor reactor internal vibrations such as vertical and lateral core motion; core support barrel and thermal shield shell modes, bending modes of fuel assemblies, and control rod vibrations. However, noise analysts have also concluded that diagnosis of a problem is easier if baseline data for normal plant operation is available. Therefore, the authors have obtained ex-core neutron noise signatures from eight PWRs to determine the similarity of signatures between plants and to build a base of data to determine the sources of neutron noise and thus the potential diagnostic information contained in the data. It is concluded that: (1) ex-core neutron noise contains information about the vibration of components in the pressure vessel; (2) baseline signature acquisition can aid understanding of plant specific vibration frequencies and provide a bases for diagnosis of future problems if they occur; and (3) abnormal core support barrel vibration can most likely be detected over and above the plant-to-plant signature variation observed thus far

Neutronic safety parameters of the BN-600 type reactor with hybrid core. Diffusion and transport approach. R-Z homogeneous media

International Nuclear Information System (INIS)

Cherny, V.; Danilytchev, A.; Korobeinikov, V.; Korobeinikova, L.; Stogov, V.

2000-01-01

The present paper includes the results of neutronic safety calculations of the BN-600 hybrid core benchmark problem. Results presented include: multiplication factors, Doppler coefficients, fuel and structure density coefficients, expansion coefficients, power distribution, beta-effective values, reaction rate distributions

Fraction Multiplication and Division Word Problems Posed by Different Years of Pre-Service Elementary Mathematics Teachers

Directory of Open Access Journals (Sweden)

Tuba Aydogdu Iskenderoglu

2018-04-01

Full Text Available It is important for pre-service teachers to know the conceptual difficulties they have experienced regarding the concepts of multiplication and division in fractions and problem posing is a way to learn these conceptual difficulties. Problem posing is a synthetic activity that fundamentally has multiple answers. The purpose of this study is to analyze the multiplication and division of fractions problems posed by pre-service elementary mathematics teachers and to investigate how the problems posed change according to the year of study the pre-service teachers are in. The study employed developmental research methods. A total of 213 pre-service teachers enrolled in different years of the Elementary Mathematics Teaching program at a state university in Turkey took part in the study. The “Problem Posing Test” was used as the data collecting tool. In this test, there are 3 multiplication and 3 division operations. The data were analyzed using qualitative descriptive analysis. The findings suggest that, regardless of the year, pre-service teachers had more conceptual difficulties in problem posing about the division of fractions than in problem posing about the multiplication of fractions.

Neutrons from Antiproton Irradiation

DEFF Research Database (Denmark)

Bassler, Niels; Holzscheiter, Michael; Petersen, Jørgen B.B.

the neutron spectrum. Additionally, we used a cylindrical polystyrene loaded with several pairs of thermoluminescent detectors containing Lithium-6 and Lithium-7, which effectively detects thermalized neutrons. The obtained results are compared with FLUKA imulations. Results: The results obtained...... spectrum is very low, and does not pose a problem for radiation therapy. However, the contribution from fast neutrons is much more significant. The dose equivalent contribution from neutrons originate from the patient alone and reaches levels which are found in passive moderated proton therapy. The exact...

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.)

A Multiple Period Problem in Distributed Energy Management Systems Considering CO2 Emissions

Science.gov (United States)

Muroda, Yuki; Miyamoto, Toshiyuki; Mori, Kazuyuki; Kitamura, Shoichi; Yamamoto, Takaya

Consider a special district (group) which is composed of multiple companies (agents), and where each agent responds to an energy demand and has a CO2 emission allowance imposed. A distributed energy management system (DEMS) optimizes energy consumption of a group through energy trading in the group. In this paper, we extended the energy distribution decision and optimal planning problem in DEMSs from a single period problem to a multiple periods one. The extension enabled us to consider more realistic constraints such as demand patterns, the start-up cost, and minimum running/outage times of equipment. At first, we extended the market-oriented programming (MOP) method for deciding energy distribution to the multiple periods problem. The bidding strategy of each agent is formulated by a 0-1 mixed non-linear programming problem. Secondly, we proposed decomposing the problem into a set of single period problems in order to solve it faster. In order to decompose the problem, we proposed a CO2 emission allowance distribution method, called an EP method. We confirmed that the proposed method was able to produce solutions whose group costs were close to lower-bound group costs by computational experiments. In addition, we verified that reduction in computational time was achieved without losing the quality of solutions by using the EP method.

The Application of the Weighted k-Partite Graph Problem to the Multiple Alignment for Metabolic Pathways.

Science.gov (United States)

Chen, Wenbin; Hendrix, William; Samatova, Nagiza F

2017-12-01

The problem of aligning multiple metabolic pathways is one of very challenging problems in computational biology. A metabolic pathway consists of three types of entities: reactions, compounds, and enzymes. Based on similarities between enzymes, Tohsato et al. gave an algorithm for aligning multiple metabolic pathways. However, the algorithm given by Tohsato et al. neglects the similarities among reactions, compounds, enzymes, and pathway topology. How to design algorithms for the alignment problem of multiple metabolic pathways based on the similarity of reactions, compounds, and enzymes? It is a difficult computational problem. In this article, we propose an algorithm for the problem of aligning multiple metabolic pathways based on the similarities among reactions, compounds, enzymes, and pathway topology. First, we compute a weight between each pair of like entities in different input pathways based on the entities' similarity score and topological structure using Ay et al.'s methods. We then construct a weighted k-partite graph for the reactions, compounds, and enzymes. We extract a mapping between these entities by solving the maximum-weighted k-partite matching problem by applying a novel heuristic algorithm. By analyzing the alignment results of multiple pathways in different organisms, we show that the alignments found by our algorithm correctly identify common subnetworks among multiple pathways.

Iterative solution of multiple radiation and scattering problems in structural acoustics using the BL-QMR algorithm

Energy Technology Data Exchange (ETDEWEB)

Malhotra, M. [Stanford Univ., CA (United States)

1996-12-31

Finite-element discretizations of time-harmonic acoustic wave problems in exterior domains result in large sparse systems of linear equations with complex symmetric coefficient matrices. In many situations, these matrix problems need to be solved repeatedly for different right-hand sides, but with the same coefficient matrix. For instance, multiple right-hand sides arise in radiation problems due to multiple load cases, and also in scattering problems when multiple angles of incidence of an incoming plane wave need to be considered. In this talk, we discuss the iterative solution of multiple linear systems arising in radiation and scattering problems in structural acoustics by means of a complex symmetric variant of the BL-QMR method. First, we summarize the governing partial differential equations for time-harmonic structural acoustics, the finite-element discretization of these equations, and the resulting complex symmetric matrix problem. Next, we sketch the special version of BL-QMR method that exploits complex symmetry, and we describe the preconditioners we have used in conjunction with BL-QMR. Finally, we report some typical results of our extensive numerical tests to illustrate the typical convergence behavior of BL-QMR method for multiple radiation and scattering problems in structural acoustics, to identify appropriate preconditioners for these problems, and to demonstrate the importance of deflation in block Krylov-subspace methods. Our numerical results show that the multiple systems arising in structural acoustics can be solved very efficiently with the preconditioned BL-QMR method. In fact, for multiple systems with up to 40 and more different right-hand sides we get consistent and significant speed-ups over solving the systems individually.

Influence of neutron scattering and source extent on the measurement of neutron energy spectra at ASDEX

International Nuclear Information System (INIS)

Huebner, K.; Baetzner, R.; Roos, M.; Robouch, B.V.; Ingrosso, L.; Wurz, H.

1987-08-01

The problem of nuclear emulsion measurements at ASDEX is considered. Besides the application of the VINIA-3DAMC software, this needs a description of the plasma neutron source, a model of the ASDEX structure, and calculation of the response of the nuclear emulsion to the incoming spectral neutron fluence. The latter is essential for comparing the numerical results with measurements at ASDEX. To treat this part, the NEPMC software was developed. The aim of the present work is to demonstrate the feasibility, reliability and usefulness of the method. Therefore simplified treatments for the ASDEX model, the plasma neutron source and the track statistics in the NEPMC software were used. Such calculations are of interest not only for nuclear emulsion measurements as well as any other neutron diagnostics, but also for all problems of neutron shielding for other diagnostics. (orig./GG)

Realization of a gamma multiplicity filter and gamma multiplicity measurements

International Nuclear Information System (INIS)

Azgui, F.

1981-12-01

A gamma multiplicity filter for the study of reaction mechanism has been realised. It's composed of six NaI(Tl) counters. The flexibility of the geometry allows many configurations. This set up has been tested with gamma radioactive sources and with the 252 Cf source to resolve problems of gamma-efficiency of the NaI(Tl) counters and the contamination of neutrons in these detectors. A logical electronic unit (Encodeur) has been constructed and the around electronic has been developped. This gamma multiplicity filter has been coupled with a detector of high resolution Ge(Li), and used in two reactions: 12 C + 55 Mn at E( 12 C) = 54 MeV; α + 63 Cu at E(α) = 52 MeV. The dominant process is the fusion-evaporation. The compound nucleus 67 Ga, is formed at the same excitation energy. The values of multiplicities Msub(γ) have been extracted using a program based on the formalism of W.J. Ockels. The fractionalization of the angular momentum is well observed for some residual nuclei ( 63 Zn, 64 Zn, 65 Zn), and for each residual nucleus, the average gamma multiplicity is lower with projectile α than that with projectile 12 C. For the most strongly output channel p2n, an entry point for the 64 Zn has been determined in the reactions. All these observations are in good agreement with those published, in the same region (f-p shell) of nuclei. This set up can be coupled with different central detector as, ''X'', neutrons charged particles detectors, and will be used with the new machine SARA to make a systematic study of transfer of angular momentum to the fragments at 30 MeV/A [fr

Physical particularities of nuclear reactors using heavy moderators of neutrons

Energy Technology Data Exchange (ETDEWEB)

Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

2016-12-15

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

Physical particularities of nuclear reactors using heavy moderators of neutrons

International Nuclear Information System (INIS)

Kulikov, G. G.; Shmelev, A. N.

2016-01-01

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using "2"3"3U as a fissile nuclide and "2"3"2Th and "2"3"1Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

International Nuclear Information System (INIS)

Cacuci, D.G.

1978-01-01

Some aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. In deriving these formulas, use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were greatly aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation, via FORMAC, of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, is one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table

Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

International Nuclear Information System (INIS)

Cacuci, D.G.

1978-04-01

Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table

Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

Energy Technology Data Exchange (ETDEWEB)

Cacuci, D.G.

1978-04-01

Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u/sup -5/. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M/sub 2/(u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table.

MCT: a Monte Carlo code for time-dependent neutron thermalization problems

International Nuclear Information System (INIS)

Cupini, E.; Simonini, R.

1974-01-01

In the Monte Carlo simulation of pulse source experiments, the neutron energy spectrum, spatial distribution and total density may be required for a long time after the pulse. If the assemblies are very small, as often occurs in the cases of interest, sophisticated Monte Carlo techniques must be applied which force neutrons to remain in the system during the time interval investigated. In the MCT code a splitting technique has been applied to neutrons exceeding assigned target times, and we have found that this technique compares very favorably with more usual ones, such as the expected leakage probability, giving large gains in computational time and variance. As an example, satisfactory asymptotic thermal spectra with a neutron attenuation of 10 -5 were quickly obtained. (U.S.)

Fission multipliers for D-D/D-T neutron generators

International Nuclear Information System (INIS)

Lou, T.P.; Vujic, J.L.; Koivunoro, H.; Reijonen, J.; Leung, K.-N.

2003-01-01

A compact D-D/D-T fusion based neutron generator is being designed at the Lawrence Berkeley National Laboratory to have a potential yield of 10 12 D-D n/s and 10 14 D-T n/s. Because of its high neutron yield and compact size (∼20 cm in diameter by 4 cm long), this neutron generator design will be suitable for many applications. However, some applications required higher flux available from nuclear reactors and spallation neutron sources operated with GeV proton beams. In this study, a subcritical fission multiplier with k eff of 0.98 is coupled with the compact neutron generators in order to increase the neutron flux output. We have chosen two applications to show the gain in flux due to the use of fission multipliers--in-core irradiation and out-of-core irradiation. For the in-core irradiation, we have shown that a gain of ∼25 can be achieved in a positron production system using D-T generator. For the out-of-core irradiation, a gain of ∼17 times is obtained in Boron Neutron Capture Therapy (BNCT) using a D-D neutron generator. The total number of fission neutrons generated by a source neutron in a fission multiplier with k eff is ∼50. For the out-of-core irradiation, the theoretical maximum net multiplication is ∼30 due to the absorption of neutrons in the fuel. A discussion of the achievable multiplication and the theoretical multiplication will be presented in this paper

An algorithm to compute a rule for division problems with multiple references

Directory of Open Access Journals (Sweden)

Sánchez Sánchez, Francisca J.

2012-01-01

Full Text Available In this paper we consider an extension of the classic division problem with claims: Thedivision problem with multiple references. Hinojosa et al. (2012 provide a solution for this type of pro-blems. The aim of this work is to extend their results by proposing an algorithm that calculates allocationsbased on these results. All computational details are provided in the paper.

Secondary neutron production from thick Pb target by light particle irradiation

CERN Document Server

Adloff, J C; Debeauvais, M; Fernández, F; Krivopustov, M; Kulakov, B A; Sosnin, A; Zamani, M

1999-01-01

Neutron multiplicities from spallation neutron sources were measured by Solid State Nuclear Track Detectors. Light particles as protons, deuterons and alphas in the GeV range were used on Pb targets. For neutron thermalization the targets were covered by 6 cm paraffin moderator. Neutron multiplicity distributions were studied inside and on the moderator surface. Comparison of SSNTDs results were made for thermal-epithermal neutrons with sup 1 sup 3 sup 9 La activation method as well as with Dubna DCM/CEM code. Discussion including previous sup 1 sup 2 C results are given.

An analytical discrete ordinates solution for a nodal model of a two-dimensional neutron transport problem

International Nuclear Information System (INIS)

Filho, J. F. P.; Barichello, L. B.

2013-01-01

In this work, an analytical discrete ordinates method is used to solve a nodal formulation of a neutron transport problem in x, y-geometry. The proposed approach leads to an important reduction in the order of the associated eigenvalue systems, when combined with the classical level symmetric quadrature scheme. Auxiliary equations are proposed, as usually required for nodal methods, to express the unknown fluxes at the boundary introduced as additional unknowns in the integrated equations. Numerical results, for the problem defined by a two-dimensional region with a spatially constant and isotropically emitting source, are presented and compared with those available in the literature. (authors)

Confinement of ultra-cold neutron in a multiple cusp magnetic field

Energy Technology Data Exchange (ETDEWEB)

Akiyama, Nobumichi; Inoue, Nobuyuki; Nihei, Hitoshi; Kinosita, Ken-ichi [Tokyo Univ. (Japan). Faculty of Engineering

1996-08-01

A new confinement system of ultra-cold neutrons is proposed. The neutron bottle is made of a rectangular vacuum chamber with the size of 40 cm x 40 cm x 30 cm covered with arrays of bar type permanent magnets. The operation of bottle requires neither cooling system nor high electric power supply, and thereby the bottle is appropriate to use in the room which is located in controlled area. The maximum kinetic energy of neutrons confined is 20 neV. Experimental scheme to test the performance of the bottle is described. (author)

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

International Nuclear Information System (INIS)

Bosevski, T.

1971-01-01

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

GINA-A polarized neutron reflectometer at the Budapest Neutron Centre

Energy Technology Data Exchange (ETDEWEB)

Bottyan, L.; Merkel, D. G.; Nagy, B.; Sajti, Sz.; Deak, L.; Endroczi, G. [Wigner RCP, RMKI, H-1525 Budapest, P.O. Box 49 (Hungary); Fuezi, J. [Wigner RCP, SZFKI, H-1525 Budapest, P.O. Box 49 (Hungary); University of Pecs, Pollack Mihaly Faculty of Engineering and Information Technology, H-7602 Pecs, P.O. Box 219 (Hungary); Petrenko, A. V. [Frank Laboratory of Neutron Physics, JINR, Joliot-Curie 6, Dubna, 141980 (Russian Federation); Major, J. [Wigner RCP, RMKI, H-1525 Budapest, P.O. Box 49 (Hungary); Max-Planck-Institut fuer Intelligente Systeme (formerly Max-Planck-Institut fuer Metallforschung), Heisenbergstr. 3, D-70569 Stuttgart (Germany)

2013-01-15

The setup, capabilities, and operation parameters of the neutron reflectometer GINA, the recently installed 'Grazing Incidence Neutron Apparatus' at the Budapest Neutron Centre, are introduced. GINA, a dance-floor-type, constant-energy, angle-dispersive reflectometer is equipped with a 2D position-sensitive detector to study specular and off-specular scattering. Wavelength options between 3.2 and 5.7 A are available for unpolarized and polarized neutrons. Spin polarization and analysis are achieved by magnetized transmission supermirrors and radio-frequency adiabatic spin flippers. As a result of vertical focusing by a five-element pyrolytic graphite monochromator, the reflected intensity from a 20 Multiplication-Sign 20 mm{sup 2} sample has been doubled. GINA is dedicated to studies of magnetic films and heterostructures, but unpolarized options for non-magnetic films, membranes, and other surfaces are also provided. Shortly after its startup, reflectivity values as low as 3 Multiplication-Sign 10{sup -5} have been measured by the instrument. The instrument capabilities are demonstrated by a non-polarized and a polarized reflectivity experiment on a Si wafer and on a magnetic film of [{sup 62}Ni/{sup nat}Ni]{sub 5} isotope-periodic layer composition. The facility is now open for the international user community. Its further development is underway establishing new sample environment options and spin analysis of off-specularly scattered radiation as well as further decreasing the background.

Simultaneous measurement of neutrons and fission fragments of thermal neutron fission of U-233

International Nuclear Information System (INIS)

Itsuro Kimura; Katsuhisa Nishio; Yoshihiro Nakagome

2000-01-01

The multiplicity and the energy of prompt neutrons from the fragments for 233 U(n th , f) were measured as functions of fragment mass and total kinetic energy. Average neutron energy against the fragment mass showed a nearly symmetric distribution about the half mass division with two valleys at 98 and 145 u. The slope of the neutron multiplicity with total kinetic energy depended on the fragment mass and showed the minimum at about 130 u. The obtained neutron data were applied to determine the total excitation energy of the system, and the resulting value in the typical asymmetric fission lied between 22 and 25 MeV. The excitation energy agreed with that determined by subtracting the total kinetic energy from the Q-value within 1 MeV, thus satisfied the energy conservation. In the symmetric fission, where the mass yield was drastically suppresses, the total excitation energy is significantly large and reaches to about 40 MeV, suggesting that fragment pairs are preferentially formed in a compact configuration at the scission point [ru

A NEW HEURISTIC ALGORITHM FOR MULTIPLE TRAVELING SALESMAN PROBLEM

Directory of Open Access Journals (Sweden)

F. NURIYEVA

2017-06-01

Full Text Available The Multiple Traveling Salesman Problem (mTSP is a combinatorial optimization problem in NP-hard class. The mTSP aims to acquire the minimum cost for traveling a given set of cities by assigning each of them to a different salesman in order to create m number of tours. This paper presents a new heuristic algorithm based on the shortest path algorithm to find a solution for the mTSP. The proposed method has been programmed in C language and its performance analysis has been carried out on the library instances. The computational results show the efficiency of this method.

Multiple Problem-Solving Strategies Provide Insight into Students' Understanding of Open-Ended Linear Programming Problems

Science.gov (United States)

Sole, Marla A.

2016-01-01

Open-ended questions that can be solved using different strategies help students learn and integrate content, and provide teachers with greater insights into students' unique capabilities and levels of understanding. This article provides a problem that was modified to allow for multiple approaches. Students tended to employ high-powered, complex,…

Quasi-energy of ultracold neutrons

International Nuclear Information System (INIS)

Frank, A.I.; Nosov, V.G.

1992-01-01

A solution is found to the problem of the propagation of a neutron beam transmitted through a periodically acting high-speed chopper. It is a generalization of the Moshinsky's problem of the evolution of a plane wave in the right half-space after an ideal absorber at the origin of coordinates has been instantaneously removed. The energy spectrum of transmitted neutrons is found to be discrete and corresponding to their quasi-energy. Interference of the states corresponding to different satellite lines leads to a complex spatial pattern with typical beats. A number of experiments with ultracold neutrons are suggested and discussed. 12 refs.; 1 fig

CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident

International Nuclear Information System (INIS)

Bessis, J.

1976-01-01

Nature of physical problem solved: Calculation of flux and neutron spectra in the case of a criticality accident. The method is unsophisticated but fast. The program is divided into two parts: (1) The code CRITIC is based on the Fermi age equation and evaluates the neutron number per fission emitted from a moderate critical system and its energy spectrum. (2) The code NARCISSE uses concrete current albedo, evaluates the product of neutron reflection on walls of the source containment and calculates the resulting flux at any point, and its energy distribution into 21 groups. The results obtained seem satisfactory, if compared with a Monte Carlo program

The secondary neutron sources for generation of particular neutron fluxes

International Nuclear Information System (INIS)

Tracz, G.

2007-07-01

The foregoing paper presents the doctor's thesis entitled '' The secondary neutron sources for generation of particular neutron fluxes ''. Two secondary neutron sources have been designed, which exploit already existing primary sources emitting neutrons of energies different from the desired ones. The first source is devoted to boron-neutron capture therapy (BNCT). The research reactor MARIA at the Institute of Atomic Energy in Swierk (Poland) is the primary source of the reactor thermal neutrons, while the secondary source should supply epithermal neutrons. The other secondary source is the pulsed source of thermal neutrons that uses fast 14 MeV neutrons from a pulsed generator at the Institute of Nuclear Physics PAN in Krakow (Poland). The physical problems to be solved in the two mentioned cases are different. Namely, in order to devise the BNCT source the initial energy of particles ought to be increased, whilst in the other case the fast neutrons have to be moderated. Slowing down of neutrons is relatively easy since these particles lose energy when they scatter in media; the most effective moderators are the materials which contain light elements (mostly hydrogen). In order to increase the energy of neutrons from thermal to epithermal (the BNCT case) the so-called neutron converter should be exploited. It contains a fissile material, 235 U. The thermal neutrons from the reactor cause fission of uranium and fast neutrons are emitted from the converter. Then fissile neutrons of energy of a few MeV are slowed down to the required epithermal energy range. The design of both secondary sources have been conducted by means of Monte Carlo simulations, which have been carried out using the MCNP code. In the case of the secondary pulsed thermal neutron source, some of the calculated results have been verified experimentally. (author)

General remarks on fast neutron reactor physics

International Nuclear Information System (INIS)

Barre, J.Y.

1980-01-01

The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

On Solution of Total Least Squares Problems with Multiple Right-hand Sides

Czech Academy of Sciences Publication Activity Database

Hnětynková, I.; Plešinger, Martin; Strakoš, Zdeněk

2008-01-01

Roč. 8, č. 1 (2008), s. 10815-10816 ISSN 1617-7061 R&D Projects: GA AV ČR IAA100300802 Institutional research plan: CEZ:AV0Z10300504 Keywords : total least squares problem * multiple right-hand sides * linear approximation problem Subject RIV: BA - General Mathematics

Proton impurity in the neutron matter: a nuclear polaron problem

Energy Technology Data Exchange (ETDEWEB)

Kutschera, M [Institute of Nuclear Physics, Cracow (Poland); Wojcik, W [Politechnika Krakowska, Cracow (Poland)

1992-10-01

We study interactions of a proton impurity with density oscillations of the neutron matter in a Debye approximation. The proton-phonon coupling is of the deformation-potential type at long wavelengths. It is weak at low density and increases with the neutron matter density. We calculate the proton`s effective mass perturbatively for a weak coupling, and use a canonical transformation technique for stronger couplings. The proton`s effective mass grows significantly with density, and at higher densities the proton impurity can be localized. This behaviour is similar to that of the polaron in solids. We obtain properties of the localized proton in the strong coupling regime from variational calculations, treating the neutron in the Thomas-Fermi approximation. (author). 14 refs, 8 figs.

Regularization methods for ill-posed problems in multiple Hilbert scales

International Nuclear Information System (INIS)

Mazzieri, Gisela L; Spies, Ruben D

2012-01-01

Several convergence results in Hilbert scales under different source conditions are proved and orders of convergence and optimal orders of convergence are derived. Also, relations between those source conditions are proved. The concept of a multiple Hilbert scale on a product space is introduced, and regularization methods on these scales are defined, both for the case of a single observation and for the case of multiple observations. In the latter case, it is shown how vector-valued regularization functions in these multiple Hilbert scales can be used. In all cases, convergence is proved and orders and optimal orders of convergence are shown. Finally, some potential applications and open problems are discussed. (paper)

Neutron activation analysis: Modelling studies to improve the neutron flux of Americium-Beryllium source

Energy Technology Data Exchange (ETDEWEB)

Didi, Abdessamad; Dadouch, Ahmed; Tajmouati, Jaouad; Bekkouri, Hassane [Advanced Technology and Integration System, Dept. of Physics, Faculty of Science Dhar Mehraz, University Sidi Mohamed Ben Abdellah, Fez (Morocco); Jai, Otman [Laboratory of Radiation and Nuclear Systems, Dept. of Physics, Faculty of Sciences, Tetouan (Morocco)

2017-06-15

Americium–beryllium (Am-Be; n, γ) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

Multiple solutions for inhomogeneous nonlinear elliptic problems arising in astrophyiscs

Directory of Open Access Journals (Sweden)

Marco Calahorrano

2004-04-01

Full Text Available Using variational methods we prove the existence and multiplicity of solutions for some nonlinear inhomogeneous elliptic problems on a bounded domain in $mathbb{R}^n$, with $ngeq 2$ and a smooth boundary, and when the domain is $mathbb{R}_+^n$

Enhanced finite difference scheme for the neutron diffusion equation using the importance function

International Nuclear Information System (INIS)

Vagheian, Mehran; Vosoughi, Naser; Gharib, Morteza

2016-01-01

Highlights: • An enhanced finite difference scheme for the neutron diffusion equation is proposed. • A seven-step algorithm is considered based on the importance function. • Mesh points are distributed through entire reactor core with respect to the importance function. • The results all proved that the proposed algorithm is highly efficient. - Abstract: Mesh point positions in Finite Difference Method (FDM) of discretization for the neutron diffusion equation can remarkably affect the averaged neutron fluxes as well as the effective multiplication factor. In this study, by aid of improving the mesh point positions, an enhanced finite difference scheme for the neutron diffusion equation is proposed based on the neutron importance function. In order to determine the neutron importance function, the adjoint (backward) neutron diffusion calculations are performed in the same procedure as for the forward calculations. Considering the neutron importance function, the mesh points can be improved through the entire reactor core. Accordingly, in regions with greater neutron importance, density of mesh elements is higher than that in regions with less importance. The forward calculations are then performed for both of the uniform and improved non-uniform mesh point distributions and the results (the neutron fluxes along with the corresponding eigenvalues) for the two cases are compared with each other. The results are benchmarked against the reference values (with fine meshes) for Kang and Rod Bundle BWR benchmark problems. These benchmark cases revealed that the improved non-uniform mesh point distribution is highly efficient.

Calibration issues for neutron diagnostics

International Nuclear Information System (INIS)

Sadler, G.J.; Adams, J.M.; Barnes, C.W.

1997-01-01

The performance of diagnostic systems are limited by their weakest constituents, including their calibration issues. Neutron diagnostics are notorious for problems encountered while determining their absolute calibrations, due mainly to the nature of the neutron transport problem. In order to facilitate the determination of an accurate and precise calibration, the diagnostic design should be such as to minimize the scattered neutron flux. ITER will use a comprehensive set of neutron diagnostics--comprising radial and vertical neutron cameras, neutron spectrometers, a neutron activation system and internal and external fission chambers--to provide accurate measurements of fusion power and power densities as a function of time. The calibration of such an important diagnostic system merits careful consideration. Some thoughts have already been given to this subject during the conceptual design phase in relation to the time-integrated neutron activation and time-dependent neutron yield monitors. However, no overall calibration strategy has been worked out so far. This paper represents a first attempt to address this vital issue. Experience gained from present large tokamaks (JET, TFTR and JT60U) and proposals for ITER are reviewed. The need to use a 14-MeV neutron generator as opposed to radioactive sources for in-situ calibration of D-T diagnostics will be stressed. It is clear that the overall absolute determination of fusion power will have to rely on a combination of nuclear measuring techniques, for which the provision of accurate and independent calibrations will constitute an ongoing process as ITER moves from one phase of operation to the next

Neutron multiplicity in deep inelastic collisions: 400 MeV Cu + Au system

International Nuclear Information System (INIS)

Tamain, B.; Chechik, R.; Ruchs, H.; Hanappe, F.; Morjean, M.; Ngo, C.; Peter, J.; Dakowski, M.; Lucas, B.; Mazur, C.; Ribrag, M.; Signarbieux, C.

1979-01-01

The authors have detected in nine different positions of space the neutrons associated with the collision of 63 Cu on 197 Au at 400 MeV bombarding energy. The deep inelastic products were detected at two different angles: close to the gazing angle and 30 0 forwards of it. Their measses were measured using a time-of-flight technique. The neutrons were detected in coincidence with the fragments - the efficiency of the neutron detectors was measured relatively to a 252 Cf source during beam time. The neutron threshold was set at 300 keV. Within an accuracy of 10% all the emitted neutrons are evaporated by the fully accelerated deep inelastic fragments. It is shown that the excitation energy is shared between the fragments in proportion to their masses and that the relaxation time for internal equilibration of the composite system is very short (approximately 10 -22 s). (Auth.)

Neutron flux monitor

International Nuclear Information System (INIS)

Oda, Naotaka.

1993-01-01

The device of the present invention greatly saves an analog processing section such as an analog filter and an analog processing circuit. That is, the device of the present invention comprises (1) a neutron flux detection means for detecting neutron fluxed in the reactor, (2) a digital filter means for dividing signals corresponding to the detected neutron fluxes into predetermined frequency band regions, (3) a calculation processing means for applying a calculation processing corresponding to the frequency band regions to the neutron flux detection signals divided by the digital filter means. With such a constitution, since the neutron detection signals are processed by the digital filter means, the accuracy is improved and the change for the property of the filter is facilitated. Further, when a neutron flux level is obtained, a calculation processing corresponding to the frequency band region can be conducted without the analog processing circuit. Accordingly, maintenance and accuracy are improved by greatly decreasing the number of parts. Further, since problems inherent to the analog circuit are solved, neutron fluxes are monitored at high reliability. (I.S.)

Physics of neutron emission in fission

International Nuclear Information System (INIS)

Lemmel, H.D.

1989-06-01

The document contains the proceedings of the IAEA Consultants' Meeting on the Physics of Neutron Emission in Fission, Mito City (Japan), 24-27 May 1988. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers cover the following topics: Energy dependence of the number of fission neutrons ν-bar (3 papers), multiplicity distribution of fission neutrons (3 papers), competition between neutron and γ-ray emission (4 papers), the fission neutron yield in resonances (2 papers) and the energy spectrum of fission neutrons in experiment (9 papers), theory (4 papers) and evaluation (1 paper). A separate abstract was prepared for each of these papers. Refs, figs and tabs

The application of isogeometric analysis to the neutron diffusion equation for a pincell problem with an analytic benchmark

International Nuclear Information System (INIS)

Hall, S.K.; Eaton, M.D.; Williams, M.M.R.

2012-01-01

Highlights: ► Isogeometric analysis used to obtain solutions to the neutron diffusion equation. ► Exact geometry captured for a circular fuel pin within a square moderator. ► Comparisons are made between the finite element method and isogeometric analysis. ► Error and observed order of convergence found using an analytic solution. -- Abstract: In this paper the neutron diffusion equation is solved using Isogeometric Analysis (IGA), which is an attempt to generalise Finite Element Analysis (FEA) to include exact geometries. In contrast to FEA, the basis functions are rational functions instead of polynomials. These rational functions, called non-uniform rational B-splines, are used to capture both the geometry and approximate the solution. The method of manufactured solutions is used to verify a MatLab implementation of IGA, which is then applied to a pincell problem. This is a circular uranium fuel pin within a square block of graphite moderator. A new method is used to compute an analytic solution to a simplified version of this problem, and is then used to observe the order of convergence of the numerical scheme. Comparisons are made against quadratic finite elements for the pincell problem, and it is found that the disadvantage factor computed using IGA is less accurate. This is due to a cancellation of errors in the FEA solution. A modified pincell problem with vacuum boundary conditions is then considered. IGA is shown to outperform FEA in this situation.

Visual Attention for Solving Multiple-Choice Science Problem: An Eye-Tracking Analysis

Science.gov (United States)

Tsai, Meng-Jung; Hou, Huei-Tse; Lai, Meng-Lung; Liu, Wan-Yi; Yang, Fang-Ying

2012-01-01

This study employed an eye-tracking technique to examine students' visual attention when solving a multiple-choice science problem. Six university students participated in a problem-solving task to predict occurrences of landslide hazards from four images representing four combinations of four factors. Participants' responses and visual attention…

Clinical application of fast neutrons

International Nuclear Information System (INIS)

Battermann, J.J.

1981-01-01

The results of treatments and clinical experiments with neutrons (from a medical d+T neutron generator with an output of 10 12 neutrons per second) are reported and discussed. Data on RBE values are presented after single doses and multiple fractions of neutrons and 60 Co-gamma rays on pulmonary metastases. The results of pilot studies on head and neck tumours, brain tumours and pelvic tumours are discussed. The accuracy of the calculated dose is tested with some in-vivo experiments during neutron irradiation of the pelvis. Estimations of RBE values for tumour control, skin damage and intestinal damage after fractionated neutron therapy are dealt with and the results obtained in treatment of sarcomas are discussed. The preliminary results are given of some clinical trials in Amsterdam. Also some data from other centres are reviewed. From these data some remarks about the future of neutron therapy are made. (Auth.)

On the adequacy of message-passing parallel supercomputers for solving neutron transport problems

International Nuclear Information System (INIS)

Azmy, Y.Y.

1990-01-01

A coarse-grained, static-scheduling parallelization of the standard iterative scheme used for solving the discrete-ordinates approximation of the neutron transport equation is described. The parallel algorithm is based on a decomposition of the angular domain along the discrete ordinates, thus naturally producing a set of completely uncoupled systems of equations in each iteration. Implementation of the parallel code on Intcl's iPSC/2 hypercube, and solutions to test problems are presented as evidence of the high speedup and efficiency of the parallel code. The performance of the parallel code on the iPSC/2 is analyzed, and a model for the CPU time as a function of the problem size (order of angular quadrature) and the number of participating processors is developed and validated against measured CPU times. The performance model is used to speculate on the potential of massively parallel computers for significantly speeding up real-life transport calculations at acceptable efficiencies. We conclude that parallel computers with a few hundred processors are capable of producing large speedups at very high efficiencies in very large three-dimensional problems. 10 refs., 8 figs

Direct Discrete Method for Neutronic Calculations

International Nuclear Information System (INIS)

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

2002-01-01

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

Safety analysis report for the Neutron Multiplier Facility, 329 Building

International Nuclear Information System (INIS)

Rieck, H.G.

1978-09-01

Neutron multiplication is a process wherein the flux of a neutron source such as 252 Cf is enhanced by fission reactions that occur in a subcritical assemblage of fissile material. The multiplication factor of the device depends upon the consequences of neutron reactions with matter and is independent of the initial number of neutrons present. Safe utilization of such a device demands that the fissile material assemblage be maintained in a subcritical state throughout all normal and credibly abnormal conditions. Examples of things that can alter the multiplication factor (and degree of subcriticality) are temperature fluctuations, changes in moderator material such as voiding or composition, addition of fissile materials, and change in assembly configuration. The Neutron Multiplier Facility (NMF) utilizes a multiplier- 252 Cf assembly to produce neutrons for activation analysis of organic and inorganic environmental samples and for on-line mass spectrometry analysis of fission products which diffuse from a stationary fissile target (less than or equal to 4 g fissile material) located in the Neutron Multiplier. The NMF annex to the 329 Building provides close proximity to related counting equipment, and delay between sample irradiation and counting is minimized

The double travelling salesman problem with multiple stacks - Formulation and heuristic solution approaches

DEFF Research Database (Denmark)

Petersen, Hanne Løhmann; Madsen, Oli B.G.

2009-01-01

This paper introduces the double travelling salesman problem with multiple stacks and presents four different metaheuristic approaches to its solution. The double TSP with multiple stacks is concerned with determining the shortest route performing pickups and deliveries in two separated networks...

Neutron importance and the generalized Green function for the conventionally critical reactor with normalized neutron distribution

International Nuclear Information System (INIS)

Khromov, V.V.

1978-01-01

The notion of neutron importance when applied to nuclear reactor statics problems described by time-independent homogeneous equations of neutron transport with provision for normalization of neutron distribution is considered. An equation has been obtained for the function of neutron importance in a conditionally critical reactor with respect to an arbitrary nons linear functional determined for the normalized neutron distribution. Relation between this function and the generalized Green function of the selfconjugated operator of the reactor equation is determined and the formula of small perturbations for the functionals of a conditionally critical reactor is deduced

Variance reduction techniques for 14 MeV neutron streaming problem in rectangular annular bent duct

Energy Technology Data Exchange (ETDEWEB)

Ueki, Kotaro [Ship Research Inst., Mitaka, Tokyo (Japan)

1998-03-01

Monte Carlo method is the powerful technique for solving wide range of radiation transport problems. Its features are that it can solve the Boltzmann`s transport equation almost without approximation, and that the complexity of the systems to be treated rarely becomes a problem. However, the Monte Carlo calculation is always accompanied by statistical errors called variance. In shielding calculation, standard deviation or fractional standard deviation (FSD) is used frequently. The expression of the FSD is shown. Radiation shielding problems are roughly divided into transmission through deep layer and streaming problem. In the streaming problem, the large difference in the weight depending on the history of particles makes the FSD of Monte Carlo calculation worse. The streaming experiment in the 14 MeV neutron rectangular annular bent duct, which is the typical streaming bench mark experiment carried out of the OKTAVIAN of Osaka University, was analyzed by MCNP 4B, and the reduction of variance or FSD was attempted. The experimental system is shown. The analysis model by MCNP 4B, the input data and the results of analysis are reported, and the comparison with the experimental results was examined. (K.I.)

Intermediate neutron spectrum problems and the intermediate neutron spectrum experiment

International Nuclear Information System (INIS)

Jaegers, P.J.; Sanchez, R.G.

1996-01-01

Criticality benchmark data for intermediate energy spectrum systems does not exist. These systems are dominated by scattering and fission events induced by neutrons with energies between 1 eV and 1 MeV. Nuclear data uncertainties have been reported for such systems which can not be resolved without benchmark critical experiments. Intermediate energy spectrum systems have been proposed for the geological disposition of surplus fissile materials. Without the proper benchmarking of the nuclear data in the intermediate energy spectrum, adequate criticality safety margins can not be guaranteed. The Zeus critical experiment now under construction will provide this necessary benchmark data

Neutronics of Laser Fission-Fusion Systems

International Nuclear Information System (INIS)

Velarde, G.

1976-01-01

Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-∞). (Author) 14 refs

Neutronics of Laser Fission-Fusion Systems

Energy Technology Data Exchange (ETDEWEB)

Velarde, G

1976-07-01

Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-{infinity}). (Author) 14 refs.

Neutron spectrum for neutron capture therapy in boron; Espectro de neutrones para terapia por captura de neutrones en boro

Energy Technology Data Exchange (ETDEWEB)

Medina C, D.; Soto B, T. G. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Programa de Doctorado en Ciencias Basicas, 98068 Zacatecas, Zac. (Mexico); Baltazar R, A. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Programa de Doctorado en Ingenieria y Tecnologia Aplicada, 98068 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

2016-10-15

Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with {sup 10}B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the {sup 10}B and produce a nucleus of {sup 7}Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10{sup 9} n/cm{sup 2}-sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

Neutron personal dosimetry: state-of-art

International Nuclear Information System (INIS)

Spurný, František

2005-03-01

State-of-art of the personal neutron dosimetry is presented, analysed and discussed. Particular attention is devoted to the problems of this type of the dosimetry of external exposure for radiation fields at nuclear power plants. A review of general problems of neutron dosimetry is given and the active individual dosimetry methods available and/or in the stage of development are briefly reviewed. Main attention is devoted to the analysis of the methods available for passive individual neutron dosimetry. The characteristics of these dosemeters were studied and are compared: their energy response functions, detection thresholds and the highest detection limits, the linearity of response, the influence of environmental factors, etc. Particular attention is devoted to their behavior in reactor neutron fields. It is concluded that the choice of the neutron personal dosemeter depends largely on the conditions in which the instrument should be used (neutron spectrum, the level of exposure and the exposure rate, etc.). The results obtained with some of these dosemeters during international intercomparisons are also presented. Particular attention is paid to the personal neutron dosimeter developed and routinely used by National Personal Dosimetry Service Ltd. in the Czech Republic. (author)

Numerical analysis for multi-group neutron-diffusion equation using Radial Point Interpolation Method (RPIM)

International Nuclear Information System (INIS)

Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong

2017-01-01

Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and

The problem of resonance self-shielding effect in neutron multigroup calculations

International Nuclear Information System (INIS)

Wang Qingming; Huang Jinghua

1991-01-01

It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

Multiple regression for physiological data analysis: the problem of multicollinearity.

Science.gov (United States)

Slinker, B K; Glantz, S A

1985-07-01

Multiple linear regression, in which several predictor variables are related to a response variable, is a powerful statistical tool for gaining quantitative insight into complex in vivo physiological systems. For these insights to be correct, all predictor variables must be uncorrelated. However, in many physiological experiments the predictor variables cannot be precisely controlled and thus change in parallel (i.e., they are highly correlated). There is a redundancy of information about the response, a situation called multicollinearity, that leads to numerical problems in estimating the parameters in regression equations; the parameters are often of incorrect magnitude or sign or have large standard errors. Although multicollinearity can be avoided with good experimental design, not all interesting physiological questions can be studied without encountering multicollinearity. In these cases various ad hoc procedures have been proposed to mitigate multicollinearity. Although many of these procedures are controversial, they can be helpful in applying multiple linear regression to some physiological problems.

Step by Step: Biology Undergraduates' Problem-Solving Procedures during Multiple-Choice Assessment.

Science.gov (United States)

Prevost, Luanna B; Lemons, Paula P

2016-01-01

This study uses the theoretical framework of domain-specific problem solving to explore the procedures students use to solve multiple-choice problems about biology concepts. We designed several multiple-choice problems and administered them on four exams. We trained students to produce written descriptions of how they solved the problem, and this allowed us to systematically investigate their problem-solving procedures. We identified a range of procedures and organized them as domain general, domain specific, or hybrid. We also identified domain-general and domain-specific errors made by students during problem solving. We found that students use domain-general and hybrid procedures more frequently when solving lower-order problems than higher-order problems, while they use domain-specific procedures more frequently when solving higher-order problems. Additionally, the more domain-specific procedures students used, the higher the likelihood that they would answer the problem correctly, up to five procedures. However, if students used just one domain-general procedure, they were as likely to answer the problem correctly as if they had used two to five domain-general procedures. Our findings provide a categorization scheme and framework for additional research on biology problem solving and suggest several important implications for researchers and instructors. © 2016 L. B. Prevost and P. P. Lemons. CBE—Life Sciences Education © 2016 The American Society for Cell Biology. This article is distributed by The American Society for Cell Biology under license from the author(s). It is available to the public under an Attribution–Noncommercial–Share Alike 3.0 Unported Creative Commons License (http://creativecommons.org/licenses/by-nc-sa/3.0).

ISINN-2. Neutron spectroscopy, nuclear structure and related topics

International Nuclear Information System (INIS)

1994-01-01

The proceedings contain the materials presented at the Second International Seminar on Neutron-Nucleus Interactions (ISINN-2) dealing with the problems of neutron spectroscopy, nuclear structure and related topics. The Seminar took place in Dubna on April 26-28, 1994. Over 120 scientists from Belgium, Bulgaria, Czech Republic, Germany, Holland, Italy, Japan, Latvia, Mexico, Poland, Slovakia, Slovenia, Ukraine, US and about 10 Russian research institutes took part in the Seminar. The main problems discussed are the following: P-odd and P-even angular correlation and T-reversal invariance in neutron reactions, nuclear structure investigations by neutron capture, the mechanism of neutron reactions, nuclear fission processes, as well as neutron data for nuclear astrophysics

A novel multi-item joint replenishment problem considering multiple type discounts.

Directory of Open Access Journals (Sweden)

Ligang Cui

Full Text Available In business replenishment, discount offers of multi-item may either provide different discount schedules with a single discount type, or provide schedules with multiple discount types. The paper investigates the joint effects of multiple discount schemes on the decisions of multi-item joint replenishment. In this paper, a joint replenishment problem (JRP model, considering three discount (all-unit discount, incremental discount, total volume discount offers simultaneously, is constructed to determine the basic cycle time and joint replenishment frequencies of multi-item. To solve the proposed problem, a heuristic algorithm is proposed to find the optimal solutions and the corresponding total cost of the JRP model. Numerical experiment is performed to test the algorithm and the computational results of JRPs under different discount combinations show different significance in the replenishment cost reduction.

Double-layer neutron shield design as neutron shielding application

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

Advances in neutron scattering spectroscopy

International Nuclear Information System (INIS)

White, J.W.

1977-01-01

Some aspects of the application of neutron scattering to problems in polymer science, surface chemistry, and adsorption phenomena, as well as molecular biology, are reviewed. In all these areas, very significant work has been carried out using the medium flux reactors at Harwell, Juelich and Risoe, even without the use of advanced multidetector techniques or of a neutron cold source. A general tendency can also be distinguished in that, for each of these new fields, a distinct preference for colder neutrons rather than thermal neutron beams can be seen. (author)

The Karlsruhe Neutron Transmission Experiment (KANT): Spherical shell transmission measurements with 14 MeV neutrons on beryllium

International Nuclear Information System (INIS)

Moellendorff, U. von; Fischer, U.; Giese, H.; Kappler, F.; Tayama, R.; Wiegner, E.; Klein, H.; Alevra, A.

1996-01-01

This is a set of viewgraphs (no additional text) of a presentation on spherical shell transmission measurements with 14 MeV neutrons on beryllium; the cross for 9 Be(n,2n)2α for the energy range between threshold (1.85 MeV) and 20 MeV neutron energy is measured and the measurement is compared with the literature. Also, neutron leakage multiplication in spherical Be shells with various thicknesses are presented. Figs, tabs

Lunar neutron source function

International Nuclear Information System (INIS)

Kornblum, J.J.

1974-01-01

The search for a quantitative neutron source function for the lunar surface region is justified because it contributes to our understanding of the history of the lunar surface and of nuclear process occurring on the moon since its formation. A knowledge of the neutron source function and neutron flux distribution is important for the interpretation of many experimental measurements. This dissertation uses the available pertinent experimental measurements together with theoretical calculations to obtain an estimate of the lunar neutron source function below 15 MeV. Based upon reasonable assumptions a lunar neutron source function having adjustable parameters is assumed for neutrons below 15 MeV. The lunar neutron source function is composed of several components resulting from the action of cosmic rays with lunar material. A comparison with previous neutron calculations is made and significant differences are discussed. Application of the results to the problem of lunar soil histories is examined using the statistical model for soil development proposed by Fireman. The conclusion is drawn that the moon is losing mass

On the multiple depots vehicle routing problem with heterogeneous fleet capacity and velocity

Science.gov (United States)

Hanum, F.; Hartono, A. P.; Bakhtiar, T.

2018-03-01

This current manuscript concerns with the optimization problem arising in a route determination of products distribution. The problem is formulated in the form of multiple depots and time windowed vehicle routing problem with heterogeneous capacity and velocity of fleet. Model includes a number of constraints such as route continuity, multiple depots availability and serving time in addition to generic constraints. In dealing with the unique feature of heterogeneous velocity, we generate a number of velocity profiles along the road segments, which then converted into traveling-time tables. An illustrative example of rice distribution among villages by bureau of logistics is provided. Exact approach is utilized to determine the optimal solution in term of vehicle routes and starting time of service.

Computation of higher spherical harmonics moments of the angular flux for neutron transport problems in spherical geometry

International Nuclear Information System (INIS)

Sahni, D.C.; Sharma, A.

2000-01-01

The integral form of one-speed, spherically symmetric neutron transport equation with isotropic scattering is considered. Two standard problems are solved using normal mode expansion technique. The expansion coefficients are obtained by solving their singular integral equations. It is shown that these expansion coefficients provide a representation of all spherical harmonics moments of the angular flux as a superposition of Bessel functions. It is seen that large errors occur in the computation of higher moments unless we take certain precautions. The reasons for this phenomenon are explained. They throw some light on the failure of spherical harmonics method in treating spherical geometry problems as observed by Aronsson

Neutron irradiation therapy machine

International Nuclear Information System (INIS)

1980-01-01

Conventional neutron irradiation therapy machines, based on the use of cyclotrons for producing neutron beams, use a superconducting magnet for the cyclotron's magnetic field. This necessitates complex liquid He equipment and presents problems in general hospital use. If conventional magnets are used, the weight of the magnet poles considerably complicates the design of the rotating gantry. Such a therapy machine, gantry and target facilities are described in detail. The use of protons and deuterons to produce the neutron beams is compared and contrasted. (U.K.)

Neutron radiography imaging with 2-dimensional photon counting method and its problems

International Nuclear Information System (INIS)

Ikeda, Y.; Kobayashi, H.; Niwa, T.; Kataoka, T.

1988-01-01

A ultra sensitive neutron imaging system has been deviced with a 2-dimensional photon counting camara (ARGUS 100). The imaging system is composed by a 2-dimensional single photon counting tube and a low background vidicon followed with an image processing unit and frame memories. By using the imaging system, electronic neutron radiography (NTV) has been possible under the neutron flux less than 3 x 10 4 n/cm 2 ·s. (author)

The Relationship Between Problem Size and Fixation Patterns During Addition, Subtraction, Multiplication, and Division

Directory of Open Access Journals (Sweden)

Evan T. Curtis

2016-08-01

Full Text Available Eye-tracking methods have only rarely been used to examine the online cognitive processing that occurs during mental arithmetic on simple arithmetic problems, that is, addition and multiplication problems with single-digit operands (e.g., operands 2 through 9; 2 + 3, 6 x 8 and the inverse subtraction and division problems (e.g., 5 – 3; 48 ÷ 6. Participants (N = 109 solved arithmetic problems from one of the four operations while their eye movements were recorded. We found three unique fixation patterns. During addition and multiplication, participants allocated half of their fixations to the operator and one-quarter to each operand, independent of problem size. The pattern was similar on small subtraction and division problems. However, on large subtraction problems, fixations were distributed approximately evenly across the three stimulus components. On large division problems, over half of the fixations occurred on the left operand, with the rest distributed between the operation sign and the right operand. We discuss the relations between these eye tracking patterns and other research on the differences in processing across arithmetic operations.

Pulsed TRIGA reactor as substitute for long pulse spallation neutron source

International Nuclear Information System (INIS)

Whittemore, W.L.

1999-01-01

those from the LPSS but at a considerably lower cost. Using the well proven and developed TRIGA reactor technology for this application would avoid the many complexities associated with either increasing the power of spallation sources or increasing the pulse length for the LPSS. An increasing problem with the spallation target is the thermal fatigue in the LPSS, a problem avoided in the pulsed TRIGA reactor. A properly designed cold source installed in a D 2 O reflector of the multiple pulsed TRIGA reactor can provide pulsed cold neutrons for neutron guides used in many neutron scattering applications. (author)

Special Features of the Air to Space Neutron Transport Problem

Science.gov (United States)

2017-09-14

an atmosphere model. Radioactive Decay Free neutrons are not stable elementary particles. They decay radioactively with a half- life of around ten...milliseconds to seconds, so that radioactive decay of neutrons is negligible. (The probability of decay in 100 milliseconds with a 10 minute half- life is...the bottom and top of a layer are 1bZ - and bZ respectively. The methods developed here apply to any planet with an atmosphere and an orbiting

Application of the Laplace transform method for the albedo boundary conditions in multigroup neutron diffusion eigenvalue problems in slab geometry

International Nuclear Information System (INIS)

Petersen, Claudio Zen; Vilhena, Marco T.; Barros, Ricardo C.

2009-01-01

In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme. (orig.)

Neutron scattering on equilibrium and nonequilibrium phonons, excitons and polaritons

International Nuclear Information System (INIS)

Broude, V.L.; Sheka, E.F.

1978-01-01

A number of problems of solid-state physics representing interest for neutron spectroscopy of future is considered. The development of the neutron inelastic scattering spectroscopy (neutron spectroscopy of equilibrium phonons) is discussed with application to nuclear dynamics of crystals in the thermodynamic equilibrium. The results of high-flux neutron source experiments on molecular crystals are presented. The advantages of neutron inelastic scattering over optical spectroscopy are discussed. The spectroscopy of quasi-equilibrium and non-equilibrium quasi-particles is discussed. In particular, the neutron scattering on polaritons, excitons in thermal equilibrium and production of light-excitons are considered. The problem of the possibility of such experiments is elucidated

A technique for combining neutron and gamma-ray data into a single assay value

International Nuclear Information System (INIS)

Pickrell, M.M.; Mercer, D.; Sharpe, T.J.

1998-01-01

The authors explored the potentials of using both neutron and gamma-ray measurements on a single item and combining these data into a single assay value. The purpose was to improve assay capability for sample matrices that are difficult to measure. They chose an empirical approach because they wanted to address difficult-to-measure items for which the assay problem is complex. They used the tomographic gamma scanner; a passive, high-efficiency neutron counter with add-a-source and multiplicity; and an active neutron, californium shuffler to obtain measurements. Twenty-four 200-L drums were measured with various matrices using all three machines. The matrices were chosen specifically to spain the difficult-to-measure assay problems for some or all of the instruments. For example, the authors measured a drum filled with concrete and another filled with metal. The data from these measurements were analyzed using the alternating conditional expectation algorithm, which is one of a class of generalized additive models. Other data fusion algorithms are also possible and are being explored. The intent was to find ways to combine the data that would reduce the matrix-induced measurement error

Neutron production in interactions of relativistic protons and deuterons with lead targets

International Nuclear Information System (INIS)

Yurevich, V.I.; Amelin, N.S.; Yakovlev, R.M.; Nikolaev, V.A.; Lyapin, V.G.; Tsvetkov, I.O.

2005-01-01

Results on the neutron double-differential cross sections and yields obtained in the time-of-flight measurements with different lead targets and beams of protons and deuterons at an energy of about 2 GeV are discussed. The neutron spatial-energy distribution for an extended lead target was studied by the threshold detector method in the energy range of protons and deuterons 1-3.7 GeV. A dependence of the mean neutron multiplicity, energy of neutrons, and process of neutron multiplication in lead on the target dimension, and the type and energy of the beam particle is analyzed. (author)

A new neutron counter for fission research

Energy Technology Data Exchange (ETDEWEB)

Laurent, B., E-mail: benoit.laurent@cea.fr [CEA, DAM, DIF, F-91297 Arpajon (France); Granier, T.; Bélier, G.; Chatillon, A.; Martin, J.-F.; Taieb, J. [CEA, DAM, DIF, F-91297 Arpajon (France); Hambsch, F.-J. [EC-JRC Institute for Reference Materials and Measurements (IRMM), Retieseweg, 2440 Geel (Belgium); Tovesson, F.; Laptev, A.B.; Haight, R.C.; Nelson, R.O.; O' Donnell, J.M. [Los Alamos Neutron Science Center, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2014-05-01

A new neutron counter for research experiments on nuclear fission has been developed. This instrument is designed for the detection of prompt fission neutrons within relatively high levels of gamma and neutron background. It is composed of a set of {sup 3}He proportional counters arranged within a block of polyethylene which serves as moderator. The detection properties have been studied by means of Monte Carlo simulations and experiments with radioactive sources. These properties are confirmed by an experiment on neutron-induced fission of {sup 238}U at the WNR facility of the Los Alamos Neutron Science Center during which the mean prompt fission neutron multiplicity, or ν{sup ¯} has been measured from 1 to 20 MeV of incident neutron energy.

ISINN-3. Neutron spectroscopy, nuclear structure, related topics

International Nuclear Information System (INIS)

1995-01-01

The proceedings contain the materials presented at the Third International Seminar on Neutron-Nucleus Interactions (ISINN-3) dealing with the problems of neutron spectroscopy, nuclear structure and related topics. The Seminar took place in Dubna on April 26-28, 1995. Over 100 scientists from Belgium, Bulgaria, Czech Republic, Germany, Japan, Latvia, Mexico, Poland, Slovakia, Ukraine, USA and from more than 10 Russian research institutes took part in the Seminar. The Seminar is dedicated to the memory of the founder of the Neutron Physics Laboratory of JINR, the famous soviet scientist Professor Fedor L. Shapiro, whose 80th anniversary is being observed. The main problems discussed are the following: fundamental interactions and symmetries in neutron-induced reactions, fundamental properties of the neutron, properties of excited nuclei after neutron capture and some other ones. Special emphasis is laid upon γ decay and neutron induced nuclear fission as well as upon the methodical aspects of new experiments

The neutron discovery

International Nuclear Information System (INIS)

Six, J.

1987-01-01

The neutron: who had first the idea, who discovered it, who established its main properties. To these apparently simple questions, multiple answers exist. The progressive discovery of the neutron is a marvellous illustration of some characteristics of the scientific research, where the unforeseen may be combined with the expected. This discovery is replaced in the context of the 1930's scientific effervescence that succeeded the revolutionary introduction of quantum mechanics. This book describes the works of Bothe, the Joliot-Curie and Chadwick which led to the neutron in an unexpected way. A historical analysis allows to give a new interpretation on the hypothesis suggested by the Joliot-Curie. Some texts of these days will help the reader to revive this fascinating story [fr

A multiple objective test assembly approach for exposure control problems in Computerized Adaptive Testing

Directory of Open Access Journals (Sweden)

Theo J.H.M. Eggen

2010-01-01

Full Text Available Overexposure and underexposure of items in the bank are serious problems in operational computerized adaptive testing (CAT systems. These exposure problems might result in item compromise, or point at a waste of investments. The exposure control problem can be viewed as a test assembly problem with multiple objectives. Information in the test has to be maximized, item compromise has to be minimized, and pool usage has to be optimized. In this paper, a multiple objectives method is developed to deal with both types of exposure problems. In this method, exposure control parameters based on observed exposure rates are implemented as weights for the information in the item selection procedure. The method does not need time consuming simulation studies, and it can be implemented conditional on ability level. The method is compared with Sympson Hetter method for exposure control, with the Progressive method and with alphastratified testing. The results show that the method is successful in dealing with both kinds of exposure problems.