Rivard, M J; D'Errico, F; Tsai, J S; Ulin, K; Engler, M J
2002-01-01
The sup 2 sup 5 sup 2 Cf neutron air kerma strength conversion factor (S sub K sub N /m sub C sub f) is a parameter needed to convert the radionuclide mass (mu g) provided by Oak Ridge National Laboratory into neutron air kerma strength required by modern clinical brachytherapy dosimetry formalisms indicated by Task Group No. 43 of the American Association of Physicists in Medicine (AAPM). The impact of currently used or proposed encapsulating materials for sup 2 sup 5 sup 2 Cf brachytherapy sources (Pt/Ir-10%, 316L stainless steel, nitinol, and Zircaloy-2) on S sub K sub N /m sub C sub f was calculated and results were fit to linear equations. Only for substantial encapsulation thicknesses, did S sub K sub N /m sub C sub f decrease, while the impact of source encapsulation composition is increasingly negligible as Z increases. These findings are explained on the basis of the non-relativistic kinematics governing the majority of sup 2 sup 5 sup 2 Cf neutron interactions. Neutron kerma and energy spectra resul...
International Nuclear Information System (INIS)
Derived air concentration (DAC) values for 175 radionuclides* produced at the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source (SNS), but not listed in Appendix A of 10 CFR 835 (01/01/2009 version), are presented. The proposed DAC values, ranging between 1 E-07 (micro)Ci/mL and 2 E-03 (micro)Ci/mL, were calculated in accordance with the recommendations of the International Commission on Radiological Protection (ICRP), and are intended to support an exemption request seeking regulatory relief from the 10 CFR 835, Appendix A, requirement to apply restrictive DACs of 2E-13 (micro)Ci/mL and 4E-11 (micro)Ci/mL and for non-listed alpha and non-alpha-emitting radionuclides, respectively.
International Nuclear Information System (INIS)
Comparison between calculated and measured shielding ratios is made for a polyethylene lined positioned steel box positioned 400 metres from a source of neutron and gamma radiation. The source was suspended outdoors at an altitude of 14 metres above the ground plane. VCS, a compilation of radiation transport codes including MORSE and DOT, was used to calculate the spectral data inside the lined box. The comparison shows fair-to-good agreement between experiment calculations for total kerma shielding ratios. (author)
Air Force neutron dosimetry program
International Nuclear Information System (INIS)
Approximately 1000 Air Force personnel are monitored for neutron radiation resulting from various sources at more than thirty worldwide locations. Neutron radiation spanning several orders of magnitude in energy is encountered. The Air Force currently uses albedo thermoluminescent neutron dosimeters for personnel monitoring. The energy dependence of the albedo neutron dosimeter is a current problem and the development of site specific correction factors is ongoing. A summary of data on the energy dependence is presented as well as efforts to develop algorithms for the dosimeter. An overview of current Air Force neutron dosimetry users and needs is also presented
Methods of core neutronic calculation
International Nuclear Information System (INIS)
Core neutronic calculations lead to the determination of geometry, composition, controls systems and to the core exploitation limits in agreement with the expected performances, with safety rules, technological choices and fuel management methods. Neutronic calculations object are described with physics justifications of hypothesis and approximations. A description and a definition of reactivity and power distribution are also given. A panorama of calculation methods used in the conception of fast breeder and pressure water reactors, are described with numerical aspects and general interest considerations related to the field of these methods and to the industrial options chosen. A complete industrial uses panorama of methods derived from the classical or generalized perturbation theory is followed by the qualification and the definition of the validity field of numerical codes.(A.B.). 88 refs., 6 figs
Accuracy of calculation of neutron detection efficiency
International Nuclear Information System (INIS)
The problems of the accuracy for the scintillator spectrometer calculation of neutron recording efficiency value are discussed. The calculation is performed by the method of direct simulation of neutron interaction with the scintillator substance. The preliminary calculations show that a contribution to efficiency of neutron recording in the range of energies of 10 through 50 MeV due to interaction of neutrons with carbon is mostly determined by reactions 12(in n' 2α)4He and 12(n, n' p)11B. The effciency calculation results are given for the cylindrical crystal of stilbene. Measurements of the neutron recording efficiency in the range of energies from 10 MeV indicate a good agreement between the calculation and the experiment
Calculation of neutron kerma in tissues
International Nuclear Information System (INIS)
Neutron kerma of normal and tumor tissues has been calculated using the tissues elemental concentration. A program developed in Math cad contains the kerma factors of C, H, O, N, Na, Mg, P, S, Cl, K, etc. that are in normal and tumor human tissues. Having the elemental composition of any human tissue the neutron kerma can be calculated. The program was tested using the elemental composition of tumor tissues such as sarcoma, melanoma, carcinoma and adenoid cystic, also neutron kerma for adipose and muscle tissue for normal adult was calculated. The results are in agreement with those published in literature. The neutron kerma for water was also calculated because in some dosimetric calculations water is used to describe normal and tumor tissues. From this comparison was found that at larger energies kerma factors are approximately the same, but energies less than 100 eV the differences are large. (Author)
Calculation of neutron kerma in tissues
Energy Technology Data Exchange (ETDEWEB)
Vega C, H.R.; Manzanares A, E. [Unidades Academicas de Estudios Nucleares, Ing. Electrica y Matematicas, Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico)]. E-mail: rvega@cantera.reduaz.mx
2004-07-01
Neutron kerma of normal and tumor tissues has been calculated using the tissues elemental concentration. A program developed in Math cad contains the kerma factors of C, H, O, N, Na, Mg, P, S, Cl, K, etc. that are in normal and tumor human tissues. Having the elemental composition of any human tissue the neutron kerma can be calculated. The program was tested using the elemental composition of tumor tissues such as sarcoma, melanoma, carcinoma and adenoid cystic, also neutron kerma for adipose and muscle tissue for normal adult was calculated. The results are in agreement with those published in literature. The neutron kerma for water was also calculated because in some dosimetric calculations water is used to describe normal and tumor tissues. From this comparison was found that at larger energies kerma factors are approximately the same, but energies less than 100 eV the differences are large. (Author)
FURNACE calculations for JET neutron diagnostics
International Nuclear Information System (INIS)
Neutron transport calculations have been performed for the JET-torus, using the two-dimensional toroidal geometry transport code system FURNACE, to predict the response of the time integrated neutron yield monitors on the variation of the plasma conditions. Calculations have been performed for the full aperture D-shaped and circular plasmas, for DD-operation and for DT-operation. For the neutron source distribution a simple model was used based on plasma-plasma interaction. For the torus rotation symmetry around the main torus axis was assumed. Curves have been produced that give the radial plasma shift as function of the ratio of the foil activations measured. It is shown that these curves are sufficiently accurate for application in the DT-phase. For application in the DD-phase, however, the flux of neutrons backscattered from the massive torus needs to be calculated more accurately. (Auth.)
Neutron shielding calculation for VVER NPP
International Nuclear Information System (INIS)
There are two methods for neutron transport (shielding) calculation used in Energoproject, Prague, the method of discrete ordinates (code TORT-DORT) and the Monte Carlo method (codes MCNP and module within the code SCALE). The task concerning neutron dose rates calculation near casks with VVER spent fuel are presented as an example. Measured neutron dose rates of real loaded C-30 casks for VVER spent fuel assemblies are compared with calculated values in the frame of the international benchmark calculation task. A part of the task realized by the Atomic Energy Research (AER) organization concerning neutron shielding is calculated. The cask C-30 is used in Slovak Jaslovske Bohunice NPP for transport of spent fuel assemblies to the storage facility. The benchmark task has been calculated by the two-dimensional code DORT originated from Oak Ridge National Laboratory. The code solves transport problems using the method of discrete ordinates (SN - method). Calculated neutron dose rates in azimuth and vertical directions show good agreement with the experiment within the range of the measurement errors. In comparison with the other codes the results of DORT are approximately 20% lower. There have been analysed differences between one- and two- dimensional approach and influence of the flux-to-dose rate conversion factors set
Punchthrough calculations for neutrons using CALOR89
International Nuclear Information System (INIS)
Punchthrough calculations for blocks of iron, copper, depleted uranium and lead of thicknesses of 10 and 12 interaction lengths have been completed for incident negative pions of 10 GeV and 100 GeV using the CALOR89 simulation code. The most numerous particles escaping out the back of the blocks are neutrons. The simulations show that there are significantly more neutrons escaping out the back of the lead block than any of the other absorbers, despite neutron production by fission in the depleted uranium. Two effects are held to be primarily responsible for this. First, the proton and neutron shells in lead nuclei are filled, giving lead a low neutron absorption cross section relative to the other absorbers, particularly uranium. Second, the number density of lead is lower than the other absorbers, particularly copper and iron
Multigroup neutron dose calculations for proton therapy
International Nuclear Information System (INIS)
We have developed tools for the preparation of coupled multigroup proton/neutron cross section libraries. Our method is to use NJOY to process evaluated nuclear data files for incident particles below 150 MeV and MCNPX to produce data for higher energies. We modified the XSEX3 program of the MCNPX code system to produce Legendre expansions of scattering matrices generated by sampling the physics models that are comparable to the output of the GROUPR routine of NJOY. Our code combines the low and high energy scattering data with user input stopping powers and energy deposition cross sections that we also calculated using MCNPX. Our code also calculates momentum transfer coefficients for the library and optionally applies an energy straggling model to the scattering cross sections and stopping powers. The motivation was initially for deterministic solution of space radiation shielding calculations using Attila, but noting that proton therapy treatment planning may neglect secondary neutron dose assessments because of difficulty and expense, we have also investigated the feasibility of multi group methods for this application. We have shown that multigroup MCNPX solutions for secondary neutron dose compare well with continuous energy solutions and are obtainable with less than half computational cost. This efficiency comparison neglects the cost of preparing the library data, but this becomes negligible when distributed over many multi group calculations. Our deterministic calculations illustrate recognized obstacles that may have to be overcome before discrete ordinates methods can be efficient alternatives for proton therapy neutron dose calculations
Multigroup neutron dose calculations for proton therapy
Energy Technology Data Exchange (ETDEWEB)
Kelsey Iv, Charles T [Los Alamos National Laboratory; Prinja, Anil K [Los Alamos National Laboratory
2009-01-01
We have developed tools for the preparation of coupled multigroup proton/neutron cross section libraries. Our method is to use NJOY to process evaluated nuclear data files for incident particles below 150 MeV and MCNPX to produce data for higher energies. We modified the XSEX3 program of the MCNPX code system to produce Legendre expansions of scattering matrices generated by sampling the physics models that are comparable to the output of the GROUPR routine of NJOY. Our code combines the low and high energy scattering data with user input stopping powers and energy deposition cross sections that we also calculated using MCNPX. Our code also calculates momentum transfer coefficients for the library and optionally applies an energy straggling model to the scattering cross sections and stopping powers. The motivation was initially for deterministic solution of space radiation shielding calculations using Attila, but noting that proton therapy treatment planning may neglect secondary neutron dose assessments because of difficulty and expense, we have also investigated the feasibility of multi group methods for this application. We have shown that multigroup MCNPX solutions for secondary neutron dose compare well with continuous energy solutions and are obtainable with less than half computational cost. This efficiency comparison neglects the cost of preparing the library data, but this becomes negligible when distributed over many multi group calculations. Our deterministic calculations illustrate recognized obstacles that may have to be overcome before discrete ordinates methods can be efficient alternatives for proton therapy neutron dose calculations.
Developing neutronics calculation tools for MYRRHA
International Nuclear Information System (INIS)
The design of the Accelerator Driven System MYRRHA requires adequate and specialised tools in the field of neutronics calculations. In order to fill the gaps, several PhD programmes were launched. In 2005 three such PhD projects were running. Each of them focuses on different stages in the computation of a core of MYRRHA. The first project Improvements of the spallation reaction model, a collaboration with the University of Liege, deals with the characterisation of the spallation neutron source using the INCL (Intra-Nuclear Cascade of Liege) model. Since at high energies, nuclear data are sparse, calculations rely on models. Especially for spallation reactions that occur at proton energies of several hundreds of MeV, models are the only means to evaluate the spallation source in MYRRHA. The second project 'Neutron transport with anisotropic scattering', a collaboration with the Universite Libre de Bruxelles, works on the development of a neutronics code, CASE-BSM, for systems with highly anisotropic scattering. The presence in large amounts of both lead and bismuth atoms in the MYRRHA core results in a highly anisotropic scattering of the neutrons in the bulk of the coolant. Neglecting this effect has large consequences on both global parameters, like keff, as well as on local parameters, like the neutron flux seen by the vessel. The third project, 'ALEPH: An integrated Monte Carlo bun-up tool', a collaboration with Ghent University, treats the last phase of a core calculation: the depletion of the fuel during irradiation. For an experimental machine like MYRRHA it is of utmost importance to have a fast calculational tool to evaluate the incineration of both isotopes present in the fuel as isotopes present in experimental devices. The main objective is to improve the current quality of the neutronics codes focused on ADS applications and to have this knowledge 'in-house'
A method for tokamak neutronics calculations
International Nuclear Information System (INIS)
This paper presents a new method for neutron transport calculation in tokamak fusion reactors. The computational procedure is based on the solution of the even-parity transport equation in a toroidal geometry. The angular neutron distribution is treated by even-parity spherical harmonic expansion, while the spatial dependence is approximated by using R-function finite elements that are defined for regions of arbitrary geometric shape. In order to test the method, calculation of a simplified tokamak model is carried out. The results are compared with the results from the literature and for the same order of accuracy a reduction of the number of spatial unknowns is shown. (author)
Uncertainty analysis of neutron transport calculation
International Nuclear Information System (INIS)
A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6Li and 7Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)
Evaluated neutron data for thermal reactor calculations
International Nuclear Information System (INIS)
The paper describes a library of evaluated neutron data designed for thermal reactor calculations and other low energy neutron physics applications. The name of the library is KORT (Evaluated Thermal Reactor Constants). The following information is given in KORT: a general characterization of the nucleus (mass, energy of capture and fission reactions, parameters of radioactive decay); partial cross-sections for neutrons of thermal energy, and the number of secondary fission neutrons (estimated errors in the measurements of these quantities are indicated); coefficients defining the deviation of capture and fission cross-sections from the 1/v law in a Maxwellian spectrum; resonance capture and fission integrals and the estimated errors in these quantities (for nuclei with Z>=90); detailed energy dependence of the cross-sections in the 10-4-5 eV region at T=300 K
Neutronic parameters calculations of a CANDU reactor
International Nuclear Information System (INIS)
Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author)
Equivalent-spherical-shield neutron dose calculations
International Nuclear Information System (INIS)
Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab
RA-0 reactor. New neutronic calculations
International Nuclear Information System (INIS)
An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author)
International Nuclear Information System (INIS)
The kerma heat production density, tritum production density, and dose in a lithium-fluoride pile with a deuterium-tritum neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both the UFO and MORSE-CV values agreed with the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source
Advanced Neutronics Tools for BWR Design Calculations
International Nuclear Information System (INIS)
This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)
Reflector modelization for neutronic diffusion calculations
International Nuclear Information System (INIS)
For neutron diffusion calculations in nuclear reactors, it is always difficult to modelize the reflector. There exist different ways to describe the neutrons density in non fissile areas like the reflector, each of them presenting some advantages and difficulties. The first part of this work gives a new reflector problem formulation, replacing the complete diffusion calculation of the reflector by boundary conditions using non-local operators, the Poincare-Steklov ones. They can be used for the eigenvectors and eigenvalues diffusion problem stated on reactive core only. This theoretical treatment of non fissile areas leads, in second part, to a new interpretation of response matrix methods and Green functions methods. These two methods are in fact the main numerical techniques used to treat reflector as boundary conditions, and an other point of view is given by the Poincare-Steklov operators. Then some simple physical cases are studied, giving explicit expressions of the Poincare-Steklov operators, and allowing numerical estimates of the reflector behaviour in a whole core-reflector PWR calculation. Finally, numerical results of Green functions for boundary perturbations illustrate the physical non-locality of the boundary operators. (author). 16 refs., 2 annexes
Advanced neutronics tools for BWR design calculations
International Nuclear Information System (INIS)
This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
Relativistic calculations of coalescing binary neutron stars
Indian Academy of Sciences (India)
Joshua Faber; Phillippe Grandclément; Frederic Rasio
2004-10-01
We have designed and tested a new relativistic Lagrangian hydrodynamics code, which treats gravity in the conformally flat approximation to general relativity. We have tested the resulting code extensively, finding that it performs well for calculations of equilibrium single-star models, collapsing relativistic dust clouds, and quasi-circular orbits of equilibrium solutions. By adding a radiation reaction treatment, we compute the full evolution of a coalescing binary neutron star system. We find that the amount of mass ejected from the system, much less than a per cent, is greatly reduced by the inclusion of relativistic gravitation. The gravity wave energy spectrum shows a clear divergence away from the Newtonian point-mass form, consistent with the form derived from relativistic quasi-equilibrium fluid sequences.
ZZ DLC-14 AIR, Group Constant Library of Secondary Gamma Transport in Air for ANISN Calculation
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: Format: ANISN, DOT, MORSE (FIDO format); Number of groups: 22 neutron / 18 gamma-ray; Nuclides: air; Origin: ENDF/B for neutron cross sections, DLC-4/HPIC for gamma-ray and DLC-12/POPLIB for secondary gamma-ray production. Weighting spectrum: 1/E for neutron cross sections. The basic idea behind the distribution of this ANISN input data is to allow potential users to repeat the ANISN calculations reported in ref. (1). It is felt that it will be more economical to repeat the calculations rather than to distribute the results of the Straker-Gritzner (1) calculations. However, the cross section part of the data can actually be used in DOT or MORSE or any transport code which will accept input cross section in the FIDO format. 2 - Method of solution: The sample input data for ANISN are for a P5, S16 calculation of the transport of neutrons and secondary gamma-rays from a 12.2 to 15 MeV point neutron source in an infinite air medium. The source is actually uniformly distributed in the first interval (500 cm radius) of a spherical medium of air with radius 3005 meters. The problem is set up for calculating various 'detector responses' by means of the 'activity' option available with ANISN. This is accomplished by providing a cross section table for a 'material' which has detector responses in certain table positions. Then the inclusion of appropriate input data for 22$ and 23$ arrays causes the group fluxes to be multiplied by the group response function values to give the desired answer. The neutron detector responses calculated by this sample problem are Henderson tissue dose, Snyder-Neufeld dose, tissue kerma, and air kerma. The gamma-ray response functions calculated are Henderson tissue dose and air kerma. The neutron cross sections were first reduced from point data from ENDF/B to a 104 fine group structure with a modified version of CSP, assuming a 1/E weighting factor. The gamma-ray data were reduced from point data from DLC
Calculations on neutron irradiation damage in reactor materials
International Nuclear Information System (INIS)
Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)
Neutron dosimetry and radiation damage calculations for HFBR
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R.; Ratner, R.T. [Pacific Northwest National Lab., TN (United States)
1998-03-01
Neutron dosimetry measurements have been conducted for various positions of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) in order to measure the neutron flux and energy spectra. Neutron dosimetry results and radiation damage calculations are presented for positions V10, V14, and V15.
Calculating and measuring thermal neutrons exiting from neutron diffractometers collimators
Tafazolee, K
2000-01-01
process, effectiveness of them are studied for the enhancement of the available system. Final conclusion from the simulation process, indicates that the heavy water with the thickness of 50 to 60 cm. is the best moderator for gaining the better thermal neutrons flux for enhancement of P.N.D. in the T.R.R. Powder Neutron Diffractometer y (P.N.D.) is relatively good and practical way for identification of the 3 dimensional construction of materials. In order to exploit the capabilities of this method, in one of the neutron beam of the Tehran Research Reactor (T.R.R.), a collimator embedded inside the concrete wall, direct the neutrons produced in the core reactor towards a monochromator e. Neutrons having been monochromated by 2 nd collimator are then directed towards the sample. Then the pattern of diffracted neutrons from the sample are studied. In order to make the best out of it, neutrons coming to sit on the sample must be of the thermal type. That means the number/amount of thermal neutrons flux in compar...
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)
Development of transient neutron transport calculation code
International Nuclear Information System (INIS)
A transient neutron transport code for time-dependent analyses of neutronics systems, named DOT4-T, has been developed. The code is based on the Discrete Ordinates code DOT4.2, which solves the steady-state neutron transport equation in two dimensions. For the discretization of time variable, a direct method, the fully implicit and unconditionally stable time integration scheme, has been employed. The resulting code has been tested using several one-dimensional and two-dimensional benchmark problems, and the results obtained with DOT4-T shows very satisfactory agreement with the benchmark problem results. (authors)
International Nuclear Information System (INIS)
The effective neutron multiplication factor, Keff, explicitly appears under the neutron albedo theory. An albedo scheme can be used to determine Keff value without an iterative strategy. The albedo theory is illustrated by the endeavor of calculating Keff by using two-group neutron albedo method for spherical reflected cores. (author). 4 refs, 7 tabs
Neutron activitation analysis of an air-dust sample using a high-flux 14 Mev neutron generator
International Nuclear Information System (INIS)
The 14 MeV neutron activation analysis technique is illustrated for multielement analysis of a Milanese air-dust sample. The neutron generator and electronic system, the efficiency and flux calibration, the γ-ray background, the sample preparation and the peak analysis used are described. After careful corrections of all possible interferences and error calculations, the results of 24 elemental concentrations are compared with those of other analytical techniques in the scope of an interlaboratory test. (orig.)
Exploratory calculations for boron capture therapy using epithermal neutron beams
International Nuclear Information System (INIS)
To get an insight into the problems of boron neutron capture therapy of brain tumours, some calculations of the neutron distribution in a spherical human skull have been made with an ANISN program. The energy of the source neutrons was varied from about 1 keV to about 100 keV. Two different neutron group structures were used with corresponding different cross section libraries. For a spherically symmetric irradiation of a skull with radius 10 cm a source neutron energy of about 50 - 100 keV gives a rather flat boron capture rate over a large part of the skull. This shows the advantage of using epithermal neutrons in the treatment of deepseated tumours by the boron neutron capture method. (Auth.)
Shielding calculations for the Gothenburg Pulsed Neutron Generator by the discrete ordinates method
International Nuclear Information System (INIS)
The discrete ordinates method has been used to calculate a proper shield to be placed around the target of the Gothenburg Pulsed Neutron Generator (PUNGGO) to minimize the dose rate outside the laboratory building. Simple calculations for slab of different materials were performed to study the effectiveness of different shielding materials. Final calculations were performed for a spherical geometry approximating the whole experimental hall to include the effect of neutron scattering from the walls and from the air. An ANISN code with a 22-group coupled neutron-gamma cross section library has been used throughout this work. The adequacy of the ANISN code for dose rate calculation has also been tested through some simple benchmark calculations. (Auth.)
Pade approximant calculations for neutron escape probability
International Nuclear Information System (INIS)
The neutron escape probability from a non-multiplying slab containing internal source is defined in terms of a functional relation for the scattering function for the diffuse reflection problem. The Pade approximant technique is used to get numerical results which compare with exact results. (author)
Calculating fusion neutron energy spectra from arbitrary reactant distributions
Eriksson, J.; Conroy, S.; Andersson Sundén, E.; Hellesen, C.
2016-02-01
The Directional Relativistic Spectrum Simulator (DRESS) code can perform Monte-Carlo calculations of reaction product spectra from arbitrary reactant distributions, using fully relativistic kinematics. The code is set up to calculate energy spectra from neutrons and alpha particles produced in the D(d, n)3He and T(d, n)4He fusion reactions, but any two-body reaction can be simulated by including the corresponding cross section. The code has been thoroughly tested. The kinematics calculations have been benchmarked against the kinematics module of the ROOT Data Analysis Framework. Calculated neutron energy spectra have been validated against tabulated fusion reactivities and against an exact analytical expression for the thermonuclear fusion neutron spectrum, with good agreement. The DRESS code will be used as the core of a detailed synthetic diagnostic framework for neutron measurements at the JET and MAST tokamaks.
A code to calculate multigroup constants for fast neutron reactor
International Nuclear Information System (INIS)
KQCS-2 code is a new improved version of KQCS code, which was designed to calculate multigroup constants for fast neutron reactor. The changes and improvements on KQCS are described in this paper. (author)
Quantum Monte Carlo calculations of two neutrons in finite volume
Klos, P.; Lynn, J. E.; Tews, I.; Gandolfi, S.; Gezerlis, A.; Hammer, H. -W.; Hoferichter, M.; Schwenk, A.
2016-01-01
Ab initio calculations provide direct access to the properties of pure neutron systems that are challenging to study experimentally. In addition to their importance for fundamental physics, their properties are required as input for effective field theories of the strong interaction. In this work, we perform auxiliary-field diffusion Monte Carlo calculations of the ground and first excited state of two neutrons in a finite box, considering a simple contact potential as well as chiral effectiv...
Calculated characteristics of subcritical assembly with anisotropic transport of neutrons
Energy Technology Data Exchange (ETDEWEB)
Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I. [Zababakhin Russian Federal Nuclear Center - All-Russian Scientific Researching Institute of Technical Physics (Russian Federation)
2003-07-01
There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5{sup n}. Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)
Neutron batch size optimisation methodology for Monte Carlo criticality calculations
International Nuclear Information System (INIS)
Highlights: • A method is suggested for improving efficiency of MC criticality calculations. • The method optimises the number of neutrons simulated per cycle. • The optimal number of neutrons per cycle depends on allocated computing time. - Abstract: We present a methodology that improves the efficiency of conventional power iteration based Monte Carlo criticality calculations by optimising the number of neutron histories simulated per criticality cycle (the so-called neutron batch size). The chosen neutron batch size affects both the rate of convergence (in computing time) and magnitude of bias in the fission source. Setting a small neutron batch size ensures a rapid simulation of criticality cycles, allowing the fission source to converge fast to its stationary state; however, at the same time, the small neutron batch size introduces a large systematic bias in the fission source. It follows that for a given allocated computing time, there is an optimal neutron batch size that balances these two effects. We approach this problem by studying the error in the cumulative fission source, i.e. the fission source combined over all simulated cycles, as all results are commonly combined over the simulated cycles. We have deduced a simplified formula for the error in the cumulative fission source, taking into account the neutron batch size, the dominance ratio of the system, the error in the initial fission source and the allocated computing time (in the form of the total number of simulated neutron histories). Knowing how the neutron batch size affects the error in the cumulative fission source allows us to find its optimal value. We demonstrate the benefits of the method on a number of numerical test calculations
Calculation methods for neutron radiography spatial resolution
International Nuclear Information System (INIS)
Spatial resolution is an important parameter for neutron radiography facility. In this paper, different methods to define the spatial resolution,such as point spread function (PSF), line spread function (LSF), edge spread function (ESF) and modulation transfer function (MTF), are analyzed and compared. MTF turns out to be the best, as it is derived from the linear system theory in a given frequency domain, and gives the maximum amount of useful information on system signal modulation. (authors)
Development of Neutron and Photon Shielding Calculation System for Workstation (NPSS-W)
International Nuclear Information System (INIS)
In plant designs and safety evaluations of nuclear fuel cycle facilities, it is important to evaluate the direct radiation and the skyshine (air-scattered photon radiation) from facilities reasonably. The Neutron and Photon Shielding Calculation System for Workstation (NPSS-W) was developed. The NPSS-W can carry out the shielding calculations of the photon and the neutron easily and rapidly. The NPSS-W can easily calculate the radiation source intensity by ORIGEN-S and the dose equivalent rate by SN transport calculational codes, which are ANISN and DOT3.5. The NPSS-W consists of five modules, which named CAL1, CAL2, CAL3, CAL4, CAL5). Some kinds of shielding calculational systems are calculated. The user's manual of NPSS-W, the examples of calculations for each module and the output data are appended. (author)
Absorbed neutron doses in air holes of fast neutron fields at the RB reactor
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Different experimental fast neutron fields are created at the RB reactor. The absorbed neutron doses in their air holes are determined on the basis of intermediate and fast neutron spectra measurements. The obtained results are analyzed in connection with application of these fields. (author)
International Nuclear Information System (INIS)
Presented is a formula for the correction calculation at the analysis of oxygen in materials by the neutron activation method. A nomogram is plotted for the calculation of corrections taking into account the oxygen of capsule material and of air being in the internal volume of the capsule due to its incomplete filling. The accuracy of corrections according to nomogram is 2-3x10-4 mass %
Neutron transport calculations of some fast critical assemblies
International Nuclear Information System (INIS)
To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
Monte Carlo calculation for TLD personal neutron dosimeter
International Nuclear Information System (INIS)
The monitor of neutron personal dose to professional worker become more and more important with the development of nuclear industry, nuclear plant and nuclear radiation cure. In this paper, the design and calculation of TLD-albedo personal dosimeter were taken by using MCNP-3B Monte Carlo code. After the present of neutron and photon fluence response, the method to determine the field correction factor was introduced. The calculated result showed that TLD-albedo personal dosimeter could work well for photon with energy: from 33 keV to 1.5 MeV and for neutron with energy from thermo-neutron to 10 MeV, and corresponding energy response error could be less than 30% and 60% respectively. (authors)
Quantum Monte Carlo calculations of neutron-alpha scattering
Nollett, Kenneth M.; Pieper, Steven C.; Wiringa, R. B.; Carlson, J; Hale, G M
2006-01-01
We describe a new method to treat low-energy scattering problems in few-nucleon systems, and we apply it to the five-body case of neutron-alpha scattering. The method allows precise calculations of low-lying resonances and their widths. We find that a good three-nucleon interaction is crucial to obtain an accurate description of neutron-alpha scattering.
Standard curves and formulae for neutron kinetics calculations
International Nuclear Information System (INIS)
The response of the neutron kinetic equations to a wide range of step and ramp additions of reactivity has been evaluated on the PACE 231R analogue computer for two fuels, U235 and Pu239, with a full range of neutron lifetimes. The results are presented in the form of standard curves which may be readily used to assess the 'zero-energy' performance of a reactor at the early stages of a reactor concept. Appendices contain the derivation of several useful expressions associated with neutron kinetics calculations and demonstrate the use of the curves to estimate reactor behaviour during shut-down following trip action. (author)
Evaluation and calculation of neutron transactinide cross-sections
International Nuclear Information System (INIS)
This paper reviews the state of the art of nuclear theory and its application to the evaluation and calculation of neutron reaction cross sections of transactinium isotopes. In particular, the paper describes the current evaluation of the total files of neutron reaction data for 240Pu and 241Pu in the energy range between 10-5 eV and 15 MeV based on a thorough analysis of available experimental data and on the use of modern theoretical concepts, and the work in progress on the evaluation of the total neutron reaction data file for 242Pu and 241Am. (author)
International Nuclear Information System (INIS)
The INDL/F-83 data library is a computerized library of evaluated neutron reaction data which has been assembled from a variety of other evaluated data files and is intended for use in fusion neutronics calculations of the International Tokamak Reactor (INTOR) Project. These data are available on magnetic tape from the IAEA Nuclear Data Section. (author)
International Nuclear Information System (INIS)
The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariances files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs
International Nuclear Information System (INIS)
The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariance files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs
Thermal and neutronic calculation for fast breeder reactor FBR
International Nuclear Information System (INIS)
This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps
Calculation verification of the utilization of LR-0 for reference neutron spectra
International Nuclear Information System (INIS)
Well-defined neutron spectrum is crucial for calibration and testing of detectors for spectrometry and dosimetry purposes. As a possible source of neutrons nuclear reactors can be utilized. In reactor core most of the neutrons are originated from fission and neutron spectra is usually some form of moderated spectra of fast neutrons. The reactor LR-0 is an experimental light-water zero-power pool-type reactor originally designed for research of the VVER type reactor cores, spent-fuel storage lattices and benchmark experiments. The main reactor feature that influences the performance of experiments is the flexible arrangement of the core. Special types of the possible core arrangements on the reactor LR-0 can provide different neutron spectra in special experimental channels. These neutron spectra are modified by inserting different materials around the channel and whole core is driven by standard fuel assemblies. Fast, epithermal or thermal spectra can be simulated using graphite, H2O, D2O insertions, air, Cd foils or fuel with different enrichment. - Highlights: • Original light water reactor spectra can be modified by material insertions. • Calculations of resulted neutron spectra have been done. • Comparison of the calcualted data to possible further utilization and research has been done
Graphical User Interface for Simplified Neutron Transport Calculations
Energy Technology Data Exchange (ETDEWEB)
Schwarz, Randolph; Carter, Leland L
2011-07-18
A number of codes perform simple photon physics calculations. The nuclear industry is lacking in similar tools to perform simplified neutron physics shielding calculations. With the increased importance of performing neutron calculations for homeland security applications and defense nuclear nonproliferation tasks, having an efficient method for performing simple neutron transport calculations becomes increasingly important. Codes such as Monte Carlo N-particle (MCNP) can perform the transport calculations; however, the technical details in setting up, running, and interpreting the required simulations are quite complex and typically go beyond the abilities of most users who need a simple answer to a neutron transport calculation. The work documented in this report resulted in the development of the NucWiz program, which can create an MCNP input file for a set of simple geometries, source, and detector configurations. The user selects source, shield, and tally configurations from a set of pre-defined lists, and the software creates a complete MCNP input file that can be optionally run and the results viewed inside NucWiz.
Accuracy preserving surrogate for neutron transport calculations
International Nuclear Information System (INIS)
Recent advances in reduced order modeling and exact-to-precision generalized perturbation theory are combined in a novel algorithm that constructs a surrogate model for the Boltzmann equation, commonly used in assembly calculations to functionalize the few-group cross-sections in terms of the various assembly types, depletion characteristics, and thermal-hydraulics conditions. First, the algorithm employs reduced order modeling to determine the dominant input parameters, aggregated in the so-called active subspace, using a random sample of first-order derivatives calculated using an adjoint model. Next, exact-to-precision generalized perturbation theory identifies an active subspace for the state solution (i.e., angular flux) and constructs a surrogate model that is parameterized over the active subspace of the input parameters. This approach is shown to significantly reduce computational time needed for the analysis of a large number of model variations, while meeting the user-defined accuracy requirements. Numerical experiments are employed to demonstrate the mechanics and application of the proposed approach to assembly calculations commonly used in reactor physics analysis. (author)
Measurements and calculations of neutron spectra and neutron dose distribution in human phantoms
International Nuclear Information System (INIS)
The measurement and calculation of the radiation field around and in a phantom, with regard to the neutron component and the contaminating gamma radiation, are essential for radiation protection and radiotherapy purposes. The final report includes the development of the simple detector system, automized detector measuring facilities and a computerized evaluating system. The results of the depth dose and neutron spectra experiments and calculations in a human phantom are given
Design calculation of a horizontal thermal neutronic beam for neutron radiography at the Syrian MNSR
International Nuclear Information System (INIS)
The computer code MCNP4C and the ENDF/B-V cross-section library were used to design calculation of a horizontal thermal beam for neutron radiography (NR) at Syrian MNSR and to evaluate the safety of the reactor after installation of the NR facility (NRF). Thermal, epithermal and fast neutron energy ranges were selected as 10.0 keV, respectively. To produce a good neutron beam in terms of intensity and quality, bismuth (Bi) and silicon (Si) were used as photon and neutron filters, respectively. The ratio of L/D of the NRF ranges between 90 and 125. The thermal neutron flux at the beam exit plane can be varied from 1.836 × 105 to 3.057 × 105 n/cm2 s. If such thermal neutron beam would be built into the Syrian MNSR, many scientific applications of the NR would be available. (author)
Fusion--fission neutronics calculations for the laser solenoid
International Nuclear Information System (INIS)
Neutron transport calculations are presented for several laser solenoid blanket configurations containing fast-fission lattices of uranium and thorium. The presence of a small-bore pulsed magnet and a small first-wall radius results in unique neutronics characteristics relative to other fusion concepts. Parametric calculations were completed to determine the effects of increasing the pulsed magnet thickness and of varying other key blanket parameters. Attractive fissile breeding rates could be achieved for blankets with a wide range of energy multiplication under the constraints of a tritium breeding ratio of about unity and a pulsed magnet thickness of about 3 cm
Neutronic calculation for rod drop accident or Angra-1 reactor
International Nuclear Information System (INIS)
The analysis of Final Safety Analysis Report for rod drop was revised using new computational codes and a new methodology with 3 steps. The purpose of this revision is to eliminate operational restrictions imposed by the actual technical specifications. First, the rod drop combinations that cause high negative neutron flux trip are determined. The second step is the thermodynamic simulation of the plant for the rod drop combinations that have not caused trip at first step. The third step is the Departure from Nucleate Boiling Ratio (DNBR) calculation for the moments of maximum power. This paper shows the neutronic calculations for the 3 steps. (author)
Neutronic calculations for Angra-1 steam line break accident
International Nuclear Information System (INIS)
The reduction of boron concentration in the Boron Injection Tank (BIT), to the room temperature solubility level, makes necessary a reanalysis of the steam line break accident of Angra 1 NPP. This paper describes the neutronic calculation related to this reanalysis. The main steps of the work were: review of reactivity parameters used in the accident simulation; search of xenon profiles that cause the most severe core power distribution; calculation of hot channel factors and other neutronic parameters necessary for DNBR determination. The final conclusion, related to the steam line break accident, states the BIT concentration may be reduced to 2000 ppm. (author)
Microscopic calculations and energy expansions for neutron-rich matter
International Nuclear Information System (INIS)
We investigate the properties of asymmetric nuclear matter with two- and three-nucleon interactions based on chiral effective field theory. Focusing on neutron-rich matter, we calculate the energy for different proton fractions and include estimates of the theoretical uncertainty. We use our ab-initio results to test the quadratic expansion around symmetric matter with the symmetry energy term, and confirm its validity for highly asymmetric systems. Our calculated energy densities are in remarkable agreement with an empirical parameterization, developed to interpolate between pure neutron and symmetric nuclear matter. These findings are very useful for astrophysical applications and for developing new equations of state.
Neutronic calculations for a fast assembly by using two-group neutron albedo theory
International Nuclear Information System (INIS)
Under Two-Group Neutron Albedo Theory, the effective neutron multiplication factor, Keff, explicitly appears and therefore it is possible to obtain an explicit form of variation of Keff. A generalization of the two-group albedo theory can be used if a more detailed energy spectrum treatment is required. The two-group neutron albedo theory is well illustrated by the endeavor of calculating the key parameters for a fast assembly. The results obtained from diffusion approach and albedo method calculations have had excellent concordance. (author)
New methods for neutron response calculations with MCNP
International Nuclear Information System (INIS)
MCNP4B was released for international distribution in February, 1997. The author summarized the new MCNP4B features since the release of MCNP4A over three years earlier and compare some results. Then he describes new methods being developed for future code releases. The focus is methods and applications of ex-core neutron response calculations
Calculation of prompt neutron spectra for curium isotopes
Energy Technology Data Exchange (ETDEWEB)
Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.
1997-03-01
With the aim of checking the existing evaluations contained in JENDL-3.2 and providing new evaluations based on a methodology proposed by the author, a series of calculations of prompt neutron spectra have been undertaken for curium isotopes. Some of the evaluations in JENDL-3.2 was found to be unphysically hard and should be revised. (author)
Calculation of neutron flux in the presence of a source
International Nuclear Information System (INIS)
Neutron sources are introduced into the reactors to initiate the chain reaction. For safety reasons, we have to know the distribution and evolution of the flux throughout the startup phase. The flux is calculated iteratively but convergence of the process can slow down arbitrarily as we approach criticality. A calculation method is presented, with a convergence speed which does not depend on the negative reactivity when it is small. (author). 7 refs
Neutron transport calculations using Quasi-Monte Carlo methods
Energy Technology Data Exchange (ETDEWEB)
Moskowitz, B.S.
1997-07-01
This paper examines the use of quasirandom sequences of points in place of pseudorandom points in Monte Carlo neutron transport calculations. For two simple demonstration problems, the root mean square error, computed over a set of repeated runs, is found to be significantly less when quasirandom sequences are used ({open_quotes}Quasi-Monte Carlo Method{close_quotes}) than when a standard Monte Carlo calculation is performed using only pseudorandom points.
Parallel processing of neutron transport in fuel assembly calculation
International Nuclear Information System (INIS)
Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's
Bourva, L C A
1999-01-01
The general purpose neutron-photon-electron Monte Carlo N-Particle code, MCNP sup T sup M , has been used to simulate the neutronic characteristics of the on-site laboratory passive neutron coincidence counter to be installed, under Euratom Safeguards Directorate supervision, at the Sellafield reprocessing plant in Cumbria, UK. This detector is part of a series of nondestructive assay instruments to be installed for the accurate determination of the plutonium content of nuclear materials. The present work focuses on one aspect of this task, namely, the accurate calculation of the coincidence gate utilisation factor. This parameter is an important term in the interpretative model used to analyse the passive neutron coincidence count data acquired using pulse train deconvolution electronics based on the shift register technique. It accounts for the limited proportion of neutrons detected within the time interval for which the electronics gate is open. The Monte Carlo code MCF, presented in this work, represents...
Calculation of 14 MeV neutron transmission
International Nuclear Information System (INIS)
The possibility of using the 28 group constant system (28-GCS) for calculating the transport of neutrons with initial energy of 14 MeV in thermonuclear reactor blankets is studied. A blanket project suggested by the Oak Ridge National Laboratory is used as a test version to estimate applicability of the 28-GCS. Niobium is used in a blanket as a structural material. A mixture of lithium nuclides is used for tritium production. The results of blanket test calculation and the calculational results obtained using the 28-GCS from the UKNDL library are compared. The numerical 28-group calculation of blonket is carried out by means of the ROZ-6 and ROZ-9 codes but not by the Monte-Carlo method as compared with the test calculation. Time of the blanket calculation on the BESM-6 computer by means of the ROZ-9 code in 2P5 approximation using the 28-GCS amounts to 10 min. It is noted that to create effective codes for the numerical blanket calculation different calculational grids are necessary for different energy grups. The calculations carried out have shown the possibility of using the 28-group library of cross sections for the numerical solution of the neutron transport equation in estimating analysis of blankets
Coupled neutron and photon cross sections for transport calculations
International Nuclear Information System (INIS)
A compact set of multigroup cross sections and transfer tables for use in neutron and photon transport calculations was prepared from ENDF/B-IV using the NJOY processing system. The library includes prompt and steady-state coupled sets for neutrons and photons in FIDO format, prompt and steady-state fission spectra (chi vectors) for the fissionable isotopes, and a table of useful response functions including heating and gas production. These multigroup constants should be useful for a wide variety of problems where self-shielding is not important. 15 references
Two level calculation of assembly neutronic data libraries
International Nuclear Information System (INIS)
The neutronic modeling of a nuclear reactor core requires 2 steps. The first step that is called transport calculation, is an accurate modeling of each type of assemblies put in a simple configuration. APOLLO2, a French neutronic code is used. This step allows the constitution of assembly data libraries. The second step represents the computing of the whole core by the diffusion theory and by using the data libraries defined in the first step. This work is dedicated to the improvement of the first step by allowing both a 172 group energy meshing and a two-dimension spatial processing. (A.C.)
Exact-to-precision generalized perturbation for neutron transport calculation
International Nuclear Information System (INIS)
This manuscript extends the exact-to-precision generalized perturbation theory (EPGPT), introduced previously, to neutron transport calculation whereby previous developments focused on neutron diffusion calculation only. The EPGPT collectively denotes new developments in generalized perturbation theory (GPT) that place premium on computational efficiency and defendable accuracy in order to render GPT a standard analysis tool in routine design and safety reactor calculations. EPGPT constructs a surrogate model with quantifiable accuracy which can replace the original neutron transport model for subsequent engineering analysis, e.g. functionalization of the homogenized few-group cross sections in terms of various core conditions, sensitivity analysis and uncertainty quantification. This is achieved by reducing the effective dimensionality of the state variable (i.e. neutron angular flux) by projection onto an active subspace. Confining the state variations to the active subspace allows one to construct a small number of what is referred to as the 'active' responses which are solely dependent on the physics model rather than on the responses of interest, the number of input parameters, or the number of points in the state phase space. (authors)
Neutron cross section calculations for fission-product nuclei
International Nuclear Information System (INIS)
To satisfy nuclear data requirements for fission-product nuclei, Hauser-Feshbach statistical calculations with preequilibrium corrections for neutron-induced reactions on isotopes of Se, Kr, Sr, Zr, Mo, Sn, Xe, and Ba between 0.001 and 20 MeV. Spherical neutron optical parameters were determined by simultaneous fits to resonance data and total cross sections. Isospin coefficients appearing in the optical potentials were determined through analysis of the behavior of s- and p-wave strengths as a function of mass for a given Z. Gamma-ray strength functions, determined through fits to stable-isotope capture data, were used in the calculation of capture cross sections and gamma-ray competition to particle emission. The resulting (n,γ), (n,n'), (n,2n), and (n,3n) cross sections, the secondary neutron emission spectra, and angular distributions calculated for 19 fission products will be averaged to provide a resulting ENDF-type fission-product neutronics file. 11 references
Quantum Monte Carlo calculations of two neutrons in finite volume
Klos, P; Tews, I; Gandolfi, S; Gezerlis, A; Hammer, H -W; Hoferichter, M; Schwenk, A
2016-01-01
Ab initio calculations provide direct access to the properties of pure neutron systems that are challenging to study experimentally. In addition to their importance for fundamental physics, their properties are required as input for effective field theories of the strong interaction. In this work, we perform auxiliary-field diffusion Monte Carlo calculations of the ground and first excited state of two neutrons in a finite box, considering a simple contact potential as well as chiral effective field theory interactions. We compare the results against exact diagonalizations and present a detailed analysis of the finite-volume effects, whose understanding is crucial for determining observables from the calculated energies. Using the L\\"uscher formula, we extract the low-energy S-wave scattering parameters from ground- and excited-state energies for different box sizes.
Neutronic calculations of cold neutron intensity in a He chamber for ultra cold neutron production
International Nuclear Information System (INIS)
Neutronic optimization studies were performed to get highest cold neutron intensity in a He-II chamber for ultra cold neutron (UCN) production as a UCN source to be installed at a spallation neutron source. Main components of the system studied were Pb-Bi target shield system, graphite reflector, D2O thermal moderator, D2 cold moderator and He-II UCN source. Effect of the size of these components on cold neutron intensity and on heat deposition was studied under the condition of 600 MeV proton energy and 20 μA proton current. It was found that in the limitation of 1 W heat removal of the He cryostat we would obtain a cold neutron average flux of 7x1011 (n/cm2/sec) in the He chamber. (authors)
A MCNP simulation study of neutronic calculations of spallation targets
Directory of Open Access Journals (Sweden)
Feghhi Seyed Amir Hossein
2013-01-01
Full Text Available The accelerator driven system is an innovative reactor which is being considered as a dedicated high-level waste burner. The function of the spallation target in accelerator driven system is to convert the incident high-energy particle beam to low-energy neutrons. One of the quantities of most interest for practical purposes is the number of neutrons produced per proton in a spallation target. However, this vital value depends not only on the material, but on the size of the target as well, due to the internuclear cascade. The MCNPX 2.4 code can be used for spallation target computation. Some benchmark results have been compared with MCNPX 2.4 simulations to verify the code's potential for calculating various parameters of an accelerator driven system target. Using the computation method, neutron interaction processes such as loss, capture and (n, xn into a spallation target have been studied for W, Ta, Pb, Bi, and LBE spallation targets in different target dimensions. With relative errors less than 10%, the numerical simulation provided by the MCNPX code agrees qualitatively with other simulation results previously carried out, qualifying it for spallation calculations. Among the studied targets, W and Ta targets resulted in a higher neutron spallation yield using lesser target dimensions. Pb, Bi, and LBE spallation targets behave similarly regarding the accessible leaked neutron yield on the outer surface of the spallation target. By use of a thicker target, LBE can compete with both W and Ta targets regarding the neutron yield parameter.
Miniature neutron source reactor burnup calculations using IRBURN code system
International Nuclear Information System (INIS)
Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.
Comparison between measured and calculated neutron spectra in FCA assemblies
International Nuclear Information System (INIS)
The neutron spectra measured in FCA Assembly VI-2, VI-1 and V-2 are discussed, and are compared with the results by calculation. The data were obtained by measurements of proton-recoil counter and double scintillator methods. Calculations were made with cell-program SP-2000 and fine-group cross section library AGRI/2, and the spectra with 1950 groups and broadened 64 and 26 group were derived. The measured spectra in the energy range of 5 keV to 6 MeV were effectively compared with the calculational results, by using C/E values. There are large differences between the measured and the calculated spectra near the 430 keV oxygen and 29 keV iron resonances. The experimental and the calculated central fission rate ratios were also compared. (author)
Directory of Open Access Journals (Sweden)
N. Carjan
2015-07-01
Full Text Available The main properties of the neutrons released during the neck rupture are calculated for U236 in the frame of a dynamical scission model: the angular distribution with respect to the fission axis (on spheres of radii R=30 and 40 fm and at time T=4×10−21 s, the distribution of the average neutron energies (for durations of the neck rupture ΔT=1 and 2×10−22 s and the total neutron multiplicity (for two values of the minimum neck-radius rmin=1.6 and 1.9 fm. They are compared with measurements of prompt fission neutrons during U235(nth,f. The experimental trends are qualitatively reproduced, i.e., the focusing of the neutrons along the fission axis, the preference of emission from the light fragment, the range, slope and average value of the neutron energy-spectrum and the average total neutron multiplicity.
On the neutron fields calculations in the nonhomogeneities
International Nuclear Information System (INIS)
Advantages of the methods for bulk integration with bulk message of data as compared with the more known boundary interaction with boundary message of data are treated. The illustrated example showing the advantages of one method before another is demonstrated. Attention on the information content of bulk sources as compared with surface ones is given. In addition, bulk massage of data is more natural for the calculation of neutron transport
Neutron physics calculation for VVER-1000 absorber element lifetime determination
International Nuclear Information System (INIS)
Absorber element (AE) with compound absorber has been operating in WWER-1000 power units since 1995. AE design meets operating organizations requirements for reliability, service life (to 10 years) and safety functions. Extension of AE service life up to 20 - 30 years by the complex of calculation and experimental work is an important problem of WWER new designs development. The paper deals with the issues related to calculation determination of main factors that influence AE service life limitation - neutron flux and fluence onto absorbing and structural materials during extended service life. (authors)
Neutron physics calculation for WWER-1000 absorber element lifetime determination
International Nuclear Information System (INIS)
Absorber element with compound absorber has been operating in WWER-1000 power units since 1995. AE design meets operating organizations requirements for reliability, service life (to 10 years) and safety functions. Extension of AE service life up to 20 - 30 years by the complex of calculation and experimental work is an important problem of WWER new designs development. The paper deals with the issues related to calculation determination of main factors that influence AE service life limitation - neutron flux and fluence onto absorbing and structural materials during extended service life. (Authors)
Calculation of 239Pu neutron inelastic cross sections
International Nuclear Information System (INIS)
We have calculated cross sections for neutron-induced reactions on 239Pu between 0.001 and 5 MeV, with particular emphasis on inelastic scattering. Coupled-channel and Hauser-Feshbach statistical models were used. Within the coupled-channel calculations we employed neutron optical parameters derived from simultaneous fits to total, elastic, inelastic, and resonance data. The resulting transmission coefficients were used in Hauser-Feshbach statistical calculations having a fission channel based on a double-humped barrier representation. Barrier parameters and transition state enhancements needed to reproduce well the (n,f) cross sections between 0.001 and 5 MeV were in general agreement with those from other published analyses. Calculated compound-nucleus and direct-reaction components for inelastic scattering were combined incoherently, and the resultant cross sections agreed well with the Bruyeres-le-Chatel measurements for scattering from levels occupying the ground state rotational band. Our results are in substantial disagreement with ENDF/B-V values for these levels. We are presently performing DWBA calculations to determine direct-reaction components for states occupying higher-lying vibrational bands
International Nuclear Information System (INIS)
Calculation results of an epithermal neutron source which can be created at the Kyiv Research Reactor (KRR) by means of placing of specially selected moderators, filters, collimators, and shielding into the 10-th horizontal experimental tube (so-called thermal column) are presented. The general Monte-Carlo radiation transport code MCNP4C [1], the Oak Ridge isotope generation code ORIGEN2 [2] and the NJOY99 [3] nuclear data processing system have been used for these calculations
New calculations of the atmospheric cosmic radiation field - Results for neutron spectra
International Nuclear Information System (INIS)
The propagation of primary cosmic rays through the Earth's atmosphere and the energy spectra of the resulting secondary particles have been calculated using the Monte Carlo transport code FLUKA with several novel auxiliary methods. Solar-modulated primary cosmic ray spectra were determined through an analysis of simultaneous proton and helium measurements made on spacecraft or high-altitude balloon flights. Primary protons and helium ions are generated within the rigidity range of 0.5 GV-20 TV, uniform in cos2θ. For a given location, primaries above the effective angle-dependent geomagnetic cut-off rigidity, and re-entrant albedo protons, are transported through the atmosphere. Helium ions are initially transported using a separate transport code called HEAVY to simulate fragmentation. HEAVY interfaces with FLUKA to provide interaction starting points for each nucleon originating from a helium nucleus. Calculated cosmic ray neutron spectra and consequent dosimetric quantities for locations with a wide range of altitude (atmospheric depth) and geomagnetic cut-off are presented and compared with measurements made on a high-altitude aeroplane. Helium ion propagation using HEAVY and inclusion of re-entrant albedo protons with the incident primary spectra significantly improved the agreement of the calculated cosmic ray neutron spectra with measured spectra. These cosmic ray propagation calculations provide the basis for a new atmospheric ionising radiation (AIR) model for air-crew dosimetry, calculation of effects on microelectronics, production of cosmogenic radionuclides and other uses. (authors)
OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels
Energy Technology Data Exchange (ETDEWEB)
Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)
2013-07-01
The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.
Methods of core neutronic calculation; Methodes de calcul neutronique de coeur
Energy Technology Data Exchange (ETDEWEB)
Bruna, G.B.; Guesdon, B. [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)
1996-02-01
Core neutronic calculations lead to the determination of geometry, composition, controls systems and to the core exploitation limits in agreement with the expected performances, with safety rules, technological choices and fuel management methods. Neutronic calculations object are described with physics justifications of hypothesis and approximations. A description and a definition of reactivity and power distribution are also given. A panorama of calculation methods used in the conception of fast breeder and pressure water reactors, are described with numerical aspects and general interest considerations related to the field of these methods and to the industrial options chosen. A complete industrial uses panorama of methods derived from the classical or generalized perturbation theory is followed by the qualification and the definition of the validity field of numerical codes.(A.B.). 88 refs., 6 figs.
Neutronic Calculation Analysis for CN HCCB TBM-Set
Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming
2015-07-01
Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)
Neutronics calculations for the TFTR neutral beam injectors
International Nuclear Information System (INIS)
Estimates, based entirely on one-dimensional transport calculations, of some of the effects of radiation on the operation and maintenance of the neutral beam injector for the Tokamak Fusion Test Reactor (TFTR) to be built at the Plasma Physics Laboratory of Princeton University are presented. Radiation effects due to 14-MeV neutrons produced by D-T reactions in the plasma and due to 2.6-MeV neutrons produced by D-D reactions in the calorimeter and in the charged-deuteron beam dump are considered. The results presented here are intended to indicate potential radiation problems rather than to be an accurate estimate of the magnitude of the actual radiation effects that will exist in the vicinity of the final injectors
Study on calculation methods for the effective delayed neutron fraction
International Nuclear Information System (INIS)
The effective delayed neutron fraction βeff is one of the important neutronic parameters from a view point of a reactor kinetics. Several Monte-Carlo-based methods to estimate βeff have been proposed to date. In order to quantify the accuracy of these methods, we study calculation methods for βeff by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate βeff with various definitions such as the fundamental value β0, the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other. Through the present study, we find the following: The largest difference among the standard definition of βeff , Nauchi's βeff and Meulekamp's βeff is approximately 10%. The fundamental value β0 is quite larger than the others in several cases. For all the cases, Meulekamp's βeff is always higher than Nauchi's βeff. This is because Nauchi's βeff considers the average neutron multiplicity value per fission which is large in the high energy range (1MeV-10MeV), while the definition of Meulekamp's βeff does not include this parameter. Furthermore, we evaluate the multi-generation effect on βeff values and demonstrate that this effect should be considered to obtain the standard definition values of βeff. (author)
TRAWA, LWR Dynamic by Coupled Neutron Diffusion and Thermohydraulics Calculation
International Nuclear Information System (INIS)
1 - Description of problem or function: The purpose of the program is to study reactor dynamics in thermal water-cooled reactors. It treats the core as one or a few axially one-dimensional subregions. The two group neutron diffusion equations are solved simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermo- hydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channels and risers with two- phase flow and of pump lines with incompressible flow. Various transients can be calculated by applying external disturbances. They can affect e.g. on movements of control rods, core inlet hydraulic conditions, system pressure or coefficients of neutronic shape function expansion between subregions. 2 - Method of solution: Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. The same spatial and temporal discretization is used for neutronics and thermohydraulics. 3 - Restrictions on the complexity of the problem: The dimensions of the program variable tables can easily be extended. Now the main dimensions are: 52 axial mesh points in core; 3 subregions; 10 axial regions with different fuel compositions; 7 radial mesh points in fuel rod; 6 delayed neutron groups; 6 coupled legs in pressure balance calculation; No flow reversals are allowed
Cronos 2: a neutronic simulation software for reactor core calculations
International Nuclear Information System (INIS)
The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)
Improvements in the model of neutron calculations for research reactors
International Nuclear Information System (INIS)
Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results are researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements, by one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author)
Improvements in the model of neutron calculations for research reactors
International Nuclear Information System (INIS)
Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results, are being researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements by means of one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author)
Aerosol and air pollution study by neutron activation analysis
International Nuclear Information System (INIS)
Thermal neutron activation analysis technique was used in air pollution and aerosol elemental content and size distribution investigations. Air pollution samples were collected on Whatman 41 paper filters which were activated along with known quantities of standards in a flux of approximately 1013 nxcm-2xs-1. The activity of the samples was measured with a 40 cm3 Ge(Li) detector and analyzed with the computer program JANE, which identified the isotopes and found their quantities by normalization with the standard measurement results. Correlation between the various elements, in particular those belonging to dust from the desert and those considered typical urban air pollution, is investigated. (author)
Calculation of neutron flux and spectrum in the irradiation test capsule at HANARO
Energy Technology Data Exchange (ETDEWEB)
Yang, Seong Woo; Cho, Man Soon; Choo, Kee Nam; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-05-15
The irradiation test capsules were mostly used for the irradiation test in CT and OR5 irradiation hole. Since the neutron fluence is an important factor, fluence monitor(F/M)s were inserted in the irradiation test capsule in order to measure the neutron fluence of test specimen. Not only the good measurement technique but also the calculation data is necessary to accurately evaluate the neutron fluence of irradiated material. Therefore, following factors should be calculated for detailed evaluation of the neutron fluence; Neutron flux and spectrum with the position of control absorber rod(CAR), Neutron flux and spectrum at the candidate F/M irradiated position, Neutron fluence difference between F/M and specimen From this calculation data, the neutron fluence of irradiated specimen and F/M can be predicted. In this paper, the neutron flux and spectrum were calculated for the irradiation capsule. This data can be a basic data of neutron dosimetry for the irradiation test and applied to select the optimum F/M installation position and verify the neutron fluence of the specimen. The neutron flux and spectrum was calculated for irradiation test capsule. The difference of neutron flux and spectrum of the irradiation test capsule in CT and OR5 irradiation hole was observed. Also the spectral averaged cross section was calculated and applied to the fast neutron fluence evaluation. As a result of this evaluation, the good agreement between calculated and measured data was shown.
CANISTER HANDLING FACILITY - VENTILATION AIR CALCULATION
International Nuclear Information System (INIS)
The purpose of this analysis is to establish the preliminary Ventilation Confinement Zone for the Canister Handling Facility (CHF). The results of this document will be used to determine the air quantities for each VCZ that will eventually be reflected in the development of the Ventilation Flow Diagrams. The analyses contained in this document are developed by D and E/Mechanical HVAC and are intended solely for the use of the D and E/Mechanical HVAC in its work regarding Confinement Zoning Analysis for the Canister Handling Facility. Yucca Mountain Project personnel from D and E/Mechanical HVAC should be consulted before use of the analyses for purposes other than those stated herein or used by individuals other than authorized personnel in D and E/Mechanical HVAC
Calculation of dosimetry parameters for fast neutron radiotherapy
International Nuclear Information System (INIS)
A computer simulation of the interactions of 50 MeV d+ on Be and 42 MeV p+ on Be neutron spectra with ICRU muscle tissue and Shonka A-150 tissue equivalent plastic was performed to allow computation of the charged particle spectra that result. Nuclear data were obtained from the Evaluated Nuclear Data File (ENDF) whenever possible and from the Intranuclear Cascade and Evaporation models otherwise. The dosimetry parameters calculated are: the kerma ratio, K/sub A-150//K/sub tissue/; the energy required to form an ion pair, W; and the stopping power ratio, S/sub g//sup W/
THERMAL: A routine designed to calculate neutron thermal scattering
International Nuclear Information System (INIS)
THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy
THERMAL: A routine designed to calculate neutron thermal scattering
Energy Technology Data Exchange (ETDEWEB)
Cullen, D.E.
1995-02-24
THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy.
Calculation of dosimetry parameters for fast neutron radiotherapy
Energy Technology Data Exchange (ETDEWEB)
Wells, A.H.
1978-05-01
A computer simulation of the interactions of 50 MeV d/sup +/ on Be and 42 MeV p/sup +/ on Be neutron spectra with ICRU muscle tissue and Shonka A-150 tissue equivalent plastic was performed to allow computation of the charged particle spectra that result. Nuclear data were obtained from the Evaluated Nuclear Data File (ENDF) whenever possible and from the Intranuclear Cascade and Evaporation models otherwise. The dosimetry parameters calculated are: the kerma ratio, K/sub A-150//K/sub tissue/; the energy required to form an ion pair, W; and the stopping power ratio, S/sub g//sup W/.
Weisskopf-Ewing calculations: neutron-induced reactions
International Nuclear Information System (INIS)
The cross sections of several neutron-induced reactions on 55Mn, sup(54,56)Fe, 59Co, sup(58,60)Ni and sup(63,65)Cu are calculated for energies below 20 MeV using the Weisskopf-Ewing theory and compared with experimental data. The total (n,p) and (n, α) cross sections are generally well fitted, especially when they are dominant channels. At the higher energies the (n,p) cross sections have important contributions from pre-equilibrium processes, and these are fitted using the theory of Feshbach, Kerman and Koonin. (author)
Neutron matter with chiral EFT interactions: Perturbative and first QMC calculations
Tews, I.; Krüger, T.; Gezerlis, A.; Hebeler, K.; Schwenk, A.
2013-01-01
Neutron matter presents a unique system in chiral effective field theory (EFT), because all many-body forces among neutrons are predicted to next-to-next-to-next-to-leading order (N3LO). We discuss perturbative and first Quantum Monte Carlo (QMC) calculations of neutron matter with chiral EFT interactions and their astrophysical impact for the equation of state and neutron stars.
The conceptual calculation for the neutron beam device at Mark 1
International Nuclear Information System (INIS)
The thermal neutron beam device, epithermal neutron beam device and test duct experiment device are designed by using Monte Carlo method at 30 kW Mark 1(-1). The compared calculation for transverse cross section dimension, moderator, reflector and others of neutron filter device are studied in this paper. The three optimized neutron beams including thermal neutron beam, epithermal neutron beam and the beam for measuring blood boron density, whose neutron flux density per reactor power are rather high, are also introduced. The results show that the BNCT neutron beam can be designed by using 30kW -1 reactor. (author)
International Nuclear Information System (INIS)
The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements
Uncertainties in Hauser-Feshbach Neutron Capture Calculations for Astrophysics
International Nuclear Information System (INIS)
The calculation of neutron capture cross sections in a statistical Hauser-Feshbach method has proved successful in numerous astrophysical applications. Of increasing interest is the uncertainty associated with the calculated Maxwellian averaged cross sections (MACS). Aspects of a statistical model that introduce a large amount of uncertainty are the level density model, γ-ray strength function parameter, and the placement of Elow – the cut-off energy below which the Hauser-Feshbach method is not applicable. Utilizing the Los Alamos statistical model code CoH3 we investigate the appropriate treatment of these sources of uncertainty via systematics of nuclei in a local region for which experimental or evaluated data is available. In order to show the impact of uncertainty analysis on nuclear data for astrophysical applications, these new uncertainties will be propagated through the nucleosynthesis code NuGrid
Calculation and analysis of the neutron radiography spatial resolution
International Nuclear Information System (INIS)
Background: Spatial resolution is the key parameter for neutron radiography facility. A model of the integrated system resolution is important when designing or using a system to ensure that the realistic resolution goals can be established and achieved. Purpose: For this resolution modeling analysis we focused on the effects of the geometry effects of L/D, the optical diffusion response of the scintillator and the sampling at the sensor (CCD or CMOS camera) and a formula was derived indicating their functional relationship. Methods: This resolution modeling analysis has been down by theoretic calculations. Then this integrated system resolution model was used as an empirical methodology to verify and optimize the performance of the detection system for real-time neutron radiography at China Advance Research Reactor. Results: The special resolutions at very collimation conditions have been calculation by using this method. And three of important parameters of this resolution model have been discussed to optimize the system performance. Conclusion: These resolution analysis concepts and methods will benefit both the design and the characterization of radiography systems. (authors)
A new method for calculation of an air quality index
Energy Technology Data Exchange (ETDEWEB)
Ilvessalo, P. [Finnish Meteorological Inst., Helsinki (Finland). Air Quality Dept.
1995-12-31
Air quality measurement programs in Finnish towns have expanded during the last few years. As a result of this it is more and more difficult to make use of all the measured concentration data. Citizens of Finnish towns are nowadays taking more of an interest in the air quality of their surroundings. The need to describe air quality in a simplified form has increased. Air quality indices permit the presentation of air quality data in such a way that prevailing conditions are more easily understandable than when using concentration data as such. Using an air quality index always means that some of the information about concentrations of contaminants in the air will be lost. How much information is possible to extract from a single index number depends on the calculation method. A new method for the calculation of an air quality index has been developed. This index always indicates the overstepping of an air quality guideline level. The calculation of this air quality index is performed using the concentrations of all the contaminants measured. The index gives information both about the prevailing air quality and also the short-term trend. It can also warn about the expected exceeding of guidelines due to one or several contaminants. The new index is especially suitable for the real-time monitoring and notification of air quality values. The behaviour of the index was studied using material from a measurement period in the spring of 1994 in Kaepylae, Helsinki. Material from a pre-operational period in the town of Oulu was also available. (author)
Comparison of statistical model calculations for stable isotope neutron capture
Beard, M.; Uberseder, E.; Crowter, R.; Wiescher, M.
2014-09-01
It is a well-observed result that different nuclear input models sensitively affect Hauser-Feshbach (HF) cross-section calculations. Less well-known, however, are the effects on calculations originating from nonmodel aspects, such as experimental data truncation and transmission function energy binning, as well as code-dependent aspects, such as the definition of level-density matching energy and the inclusion of shell correction terms in the level-density parameter. To investigate these aspects, Maxwellian-averaged neutron capture cross sections (MACS) at 30 keV have been calculated using the well-established statistical Hauser-Feshbach model codes talys and non-smoker for approximately 340 nuclei. For the same nuclei, MACS predictions have also been obtained using two new HF codes, cigar and sapphire. Details of these two codes, which have been developed to contain an overlapping set of identically implemented nuclear physics input models, are presented. It is generally accepted that HF calculations are valid to within a factor of 3. It was found that this factor is dependent on both model and nonmodel details, such as the coarseness of the transmission function energy binning and data truncation, as well as variances in details regarding the implementation of level-density parameter, backshift, matching energy, and giant dipole strength function parameters.
Neutron and photon transport calculations in fusion system. 2
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1998-03-01
On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)
Descartes: a new generation system for neutronic calculations
International Nuclear Information System (INIS)
Descartes is a common project between CEA, Framatome and EDF for the development of a new generation system for neutronic calculations. The main objectives which have leaded the design of the platform are the following: - flexible: from best-estimate calculations to industrial design; - open: easy coupling with other disciplines (thermo mechanics, thermal hydraulics); - enlarged scope: criticality, shielding, all types of reactors; - robust: well known behavior in its field of application; - safe: qualified and uncertainties assessment; and - User-friendly: user interface, databases; Descartes is based on the object oriented method using UML design and programmed in C++ and the Python interpreted script language. We will present in this paper the general architecture of the platform and the internal data model used which allows the definition of common exchange structures between solvers and the different modules which can be used either for lattice or core calculations. In a second time we will present a short description of the main solvers implemented within the Descartes platform. We will conclude with some first results of industrial PWR calculations. (author)
Transport calculations for a 14.8 MeV neutron beam in a water phantom
International Nuclear Information System (INIS)
A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented
Computational models for probabilistic neutronic calculation in TADSEA
International Nuclear Information System (INIS)
The Very High Temperature Reactor is one of the main candidates for the next generation of nuclear power plants. In pebble bed reactors, the fuel is contained within graphite pebbles in the form of TRISO particles, which form a randomly packed bed inside a graphite-walled cylindrical cavity. In previous studies, the conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made. The TADSEA is a pebble-bed ADS cooled by helium and moderated by graphite. In order to simulate the TADSEA correctly, the double heterogeneity of the system must be considered. It consists on randomly located pebbles into the core and randomly located TRISO particles into the fuel pebbles. These features are often neglected due to the difficulty to model with MCNP code. The main reason is that there is a limited number of cells and surfaces to be defined. In this paper a computational tool, which allows to get a new geometrical model for fuel pebble to neutronic calculation with MCNPX, was presented. The heterogeneity of system is considered, and also the randomly located TRISO particles inside the pebble. There are also compared several neutronic computational models for TADSEA's fuel pebbles in order to study heterogeneity effects. On the other hand the boundary effect given by the intersection between the pebble surface and the TRISO particles could be significative in the multiplicative properties. A model to study this e ect is also presented. (author)
Monte Carlo calculation of ''skyshine'' neutron dose from ALS [Advanced Light Source
International Nuclear Information System (INIS)
This report discusses the following topics on ''skyshine'' neutron dose from ALS: Sources of radiation; ALS modeling for skyshine calculations; MORSE Monte-Carlo; Implementation of MORSE; Results of skyshine calculations from storage ring; and Comparison of MORSE shielding calculations
Neutron spectral adjustment and radiation damage calculations for reactor dosimetry
International Nuclear Information System (INIS)
Nuclear data needs for retrospective reactor dosimetry, including requested evaluated cross sections for 54Fe(n,γ)55Fe, 62Ni(n,γ)63Ni, and 93Nb(n,γ)94Nb are presented. The latest version of the SPECTER computer code, which calculates dpa, pka atomic recoil spectra, and gas production for 40 elements and selected compounds, has been made available to the IAEA-NDS for potential inclusion in IRDF-2002. A PC version of the STAY'SL computer code, which performs neutron spectral adjustments, has also been made available. The STAY' SL data libraries can be updated with the new IRDF-2002 cross sections and covariances, when these data become available. (a.n.)
Nested element method in multidimensional neutron diffusion calculations
International Nuclear Information System (INIS)
A new numerical method is developed that is particularly efficient in solving the multidimensional neutron diffusion equation in geometrically complex systems. The needs for a generally applicable and fast running computer code have stimulated, here presented, the inroad of a nonclassical (R-function) numerical method into the nuclear field. By using the R-functions, the geometrical components of the diffusion problem are a priory analytically implemented into the approximate solution. The class of functions, to which the approximate solution belongs, is chosen as close to the exact solution class as practically acceptable from the time consumption point of view. That implies a drastic reduction of the number of degrees of freedom, compared to the other methods. Furthermore, the reduced number of degrees of freedom enables calculation of large multidimensional problems on small computers
Nested element method in multidimensional neutron diffusion calculations
International Nuclear Information System (INIS)
A new numerical method is developed that is particularly efficient in solving the multidimensional neutron diffusion equation in geometrically complex systems. The needs for a generally applicable and fast running computer code have stimulated the inroad of a nonclassical (R-function) numerical method into the nuclear field. By using the R-functions, the geometrical components of the diffusion problem are a priori analytically implemented into the approximate solution. The class of functions, to which the approximate solution belongs, is chosen as close to the exact solution class as practically acceptable from the time consumption point of view. That implies a drastic reduction of the number of degrees of freedom, compared to the other methods. Furthermore, the reduced number of degrees of freedom enables calculation of large multidimensional problems on small computers
Key precursor data in aggregate delayed-neutron calculations
International Nuclear Information System (INIS)
The reactivity calculations with the delayed neutron (DN) six-group parameter sets in ENDF/B-VI were reported to give significant underestimates for long period (tens of seconds). The parameter sets were obtained form the summation calculations with ENDF/B-VI fission yields and decay data files. In this paper, we try to identify the precursor data that cause the significant underestimates. Because of the relatively long time scale, we examine the DN activity after infinite irradiation, and find that the summation calculation gives significantly smaller DN activity at about 30 s than the currently used six-group parameter set by Tuttle, although this feature does not looks important for the DN activity after a fission burst. From the time dependence of the DN activity, we find that the fission yields of 88Br, 136Te, and 137I are the most probable sources for the underestimate. Furthermore, in order to achieve the required precision (5%) for the DN activity, it is also necessary to perform precise measurements of their Pn values. (author)
An integrated multi-functional neutronics calculation and analysis code system: VisualBUS
International Nuclear Information System (INIS)
Neutronics calculation and analysis are the bases of reactor physics design, radiation protection, fuel management optimization, nuclear safety analysis, etc. After surveying and evaluating the status and trend of development of neutronics calculation and analysis codes, a network-based integrated multi-functional neutronics calculation and analysis code system has been designed and developed for applications in fusion, fission and various hybrid systems based on the adoption of advanced neutronics calculating approaches and modern computer' software technologies. A series of benchmark tests and applications have shown the maturity and effectiveness of the system. This paper gives a brief overview about main technical features of the system, the benchmark tests and applications. (authors)
International Nuclear Information System (INIS)
The objective of the study is to compare the thermal neutron fluxes at specimen positions of neutron radiography facility calculated by MCNP4C code with the measurement. A model for calculation was developed using details of the reactor core configuration no. 14 and neutron radiography facility installed at the existing research reactor, TRR-1/M1 reactor. Assuming all fresh fuel elements and all control rod out condition, the thermal neutron fluxes at various specimen positions were calculated using MCNP4C code. The calculation are verified by the measurement using foil activation method. Generally, the calculated neutron fluxes are overestimated by 16-20% which is reasonably good agreement and acceptable for the complex system. The discrepancy is expected to the assumption of using fresh fuel elements, all control rod out condition, and also lacks of information in develop a more accurate model for calculation. This study shows the possibility of using the MCNP4C code to verify the thermal neutron fluxes at specimen position and shielding design of the new neutron radiography facility at the new Thai research reactor
Neutron and gamma ray transport calculations in shielding system
Energy Technology Data Exchange (ETDEWEB)
Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)
International Nuclear Information System (INIS)
It is shown that the combination of 3D neutron transport calculations and the results from activation foil measurements at a limited number of locations in a materials testing irradiation experiment can provide information at any position in the experiment for detailed neutron dosimetry and damage analysis. 4 refs
International Nuclear Information System (INIS)
A method of solving the diffusion equation for the th ermal neutron flux in a heterogeneous medium is presented. Perturbation calculation is successfully applied for the cylindrical concentric system after testing this method for the spherical concentric geometry analytically solved by Czubek (1981). The method permits to calculate the t hermal neutron decay constant and the space distribution of the thermal neutron flux in a heterogeneous geom etry. The condition of the constant value of the neutron flux in the inner part of the system has to be m et. This method has an application in the measurement of the thermal neutron absorption cross section, presented by Czubek (1981). (author)
Monte Carlo Calculation Of Thermal And Epithermal Neutron Self-Shielding Factors
International Nuclear Information System (INIS)
Neutron activation measurement is often performed in a reactor neutron spectrum. When the size of the irradiation sample is not small enough and resonance peaks present in the cross section of the sample nuclide, the thermal and resonance self-shielding effects of neutron flux in the sample must be considered for correction. In this work, the Monte Carlo code MCNP-5 has been applied for calculation of the self-shielding factors for several standard samples and neutron monitors that are often used in measurements of thermal neutron capture cross sections and resonance integrals. The results of calculation are tabulated with different sample thickness and different irradiation geometries. (author)
International Nuclear Information System (INIS)
Neutron-transport calculations with the FURNACE(2) program system, in support of the Neutron Diagnostic Group at JET, have been performed since 1980, i.e. since the construction phase of JET. FURNACE(2) is a ray-tracing/multiple-reflection transport program system for toroidal geometries, that orginally was developed for blanket neutronics studies and which then was improved and extended for application to the neutron-diagnostics at JET. (orig./WL)
Calculating the Carbon Footprint from Different Classes of Air Travel
Bofinger, Heinrich; Strand, Jon
2013-01-01
This paper develops a new methodology for calculating the "carbon footprint" of air travel whereby emissions from travel in premium (business and first) classes depend heavily on the average class-specific occupied floor space. Unlike methods currently used for the purpose, the approach properly accounts for the fact that the relative number of passenger seats in economy and premium classe...
Calculation principles of humid air in a reversed Brayton cycle
Energy Technology Data Exchange (ETDEWEB)
Backman, J. [Lappeenranta Univ. of Technology (Finland). Dept. of Energy Technology
1997-12-31
The article presents a calculation method for reversed Brayton cycle that uses humid air as working medium. The reversed Brayton cycle can be employed as an air dryer, a heat pump or a refrigerating machine. In this research the use of humid air as a working fluid has an environmental advantage, as well. In this method especially the expansion process in the turbine is important because of the condensation of the water vapour in the humid air. This physical phenomena can have significant effects on the level of performance of the application. The expansion process differs physically from the compression process, when the water vapour in the humid air begins to condensate. In the thermodynamic equilibrium of the flow, the water vapour pressure in humid air cannot exceed the pressure of saturated water vapour in corresponding temperature. Expansion calculation during operation around the saturation zone is based on a quasistatic expansion, in which the system after the turbine is in thermodynamical equilibrium. The state parameters are at every moment defined by the equation of state, and there is no supercooling in the vapour. Following simplifications are used in the calculations: The system is assumed to be adiabatic. This means that there is no heat transfer to the surroundings. This is a common practice, when the temperature differences are moderate as here; The power of the cooling is omitted. The cooling construction is very dependent on the machine and the distribution of the losses; The flow is assumed to be one-dimensional, steady-state and homogenous. The water vapour condensing in the turbine can cause errors, but the errors are mainly included in the efficiency calculation. (author) 11 refs.
Concise four-vector scheme for neutron transport calculations
International Nuclear Information System (INIS)
An explicit Riemannian geometrical form or the vectorial Neutron Streaming Term is presented. The method applies the full Riemannian technique of general covariance. There are cases when the symmetry of the neutron flux must be smaller than that of the arrangement. However, in coordinate space there are always solutions of the Neutron Transport Equation as symmetric as the arrangement, if the latter's symmetry is at least an affine collineation of the Euclidian 3-space. (author). 7 refs
Tables for simplifying calculations of activities produced by thermal neutrons
Senftle, F.E.; Champion, W.R.
1954-01-01
The method of calculation described is useful for the types of work of which examples are given. It is also useful in making rapid comparison of the activities that might be expected from several different elements. For instance, suppose it is desired to know which of the three elements, cobalt, nickel, or vanadium is, under similar conditions, activated to the greatest extent by thermal neutrons. If reference is made to a cross-section table only, the values may be misleading unless properly interpreted by a suitable comparison of half-lives and abundances. In this table all the variables have been combined and the desired information can be obtained directly from the values of A 3??, the activity produced per gram per second of irradiation, under the stated conditions. Hence, it is easily seen that, under similar circumstances of irradiation, vanadium is most easily activated even though the cross section of one of the cobalt isotopes is nearly five times that of vanadium and the cross section of one of the nickel isotopes is three times that of vanadium. ?? 1954 Societa?? Italiana di Fisica.
International Nuclear Information System (INIS)
This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as keff and ksrc, and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)
Study of calculated and measured time dependent delayed neutron yields
International Nuclear Information System (INIS)
Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of 232U, 237Np, 238Pu, 241Am, /sup 242m/Am, 245Cm, and 249Cf were studied for the first time. The delayed neutron emission from 232Th, 233U, 235U, 238U, 239Pu, 241Pu, and 242Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from 232Th to 252Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables
How Accurately Can We Calculate Neutrons Slowing Down In Water ?
International Nuclear Information System (INIS)
We have compared the results produced by a variety of currently available Monte Carlo neutron transport codes for the relatively simple problem of a fast source of neutrons slowing down and thermalizing in water. Initial comparisons showed rather large differences in the calculated flux; up to 80% differences. By working together we iterated to improve the results by: (1) insuring that all codes were using the same data, (2) improving the models used by the codes, and (3) correcting errors in the codes; no code is perfect. Even after a number of iterations we still found differences, demonstrating that our Monte Carlo and supporting codes are far from perfect; in particularly we found that the often overlooked nuclear data processing codes can be the weakest link in our systems of codes. The results presented here represent the today's state-of-the-art, in the sense that all of the Monte Carlo codes are modern, widely available and used codes. They all use the most up-to-date nuclear data, and the results are very recent, weeks or at most a few months old; these are the results that current users of these codes should expect to obtain from them. As such, the accuracy and limitations of the codes presented here should serve as guidelines to code users in interpreting their results for similar problems. We avoid crystal ball gazing, in the sense that we limit the scope of this report to what is available to code users today, and we avoid predicting future improvements that may or may not actual come to pass. An exception that we make is in presenting results for an improved thermal scattering model currently being testing using advanced versions of NJOY and MCNP that are not currently available to users, but are planned for release in the not too distant future. The other exception is to show comparisons between experimentally measured water cross sections and preliminary ENDF/B-VII thermal scattering law, S(α,β) data; although these data are strictly preliminary
How Accurately Can We Calculate Neutrons Slowing Down In Water ?
Energy Technology Data Exchange (ETDEWEB)
Cullen, D E; Blomquist, R; Greene, M; Lent, E; MacFarlane, R; McKinley, S; Plechaty, E; Sublet, J C
2006-03-30
We have compared the results produced by a variety of currently available Monte Carlo neutron transport codes for the relatively simple problem of a fast source of neutrons slowing down and thermalizing in water. Initial comparisons showed rather large differences in the calculated flux; up to 80% differences. By working together we iterated to improve the results by: (1) insuring that all codes were using the same data, (2) improving the models used by the codes, and (3) correcting errors in the codes; no code is perfect. Even after a number of iterations we still found differences, demonstrating that our Monte Carlo and supporting codes are far from perfect; in particularly we found that the often overlooked nuclear data processing codes can be the weakest link in our systems of codes. The results presented here represent the today's state-of-the-art, in the sense that all of the Monte Carlo codes are modern, widely available and used codes. They all use the most up-to-date nuclear data, and the results are very recent, weeks or at most a few months old; these are the results that current users of these codes should expect to obtain from them. As such, the accuracy and limitations of the codes presented here should serve as guidelines to code users in interpreting their results for similar problems. We avoid crystal ball gazing, in the sense that we limit the scope of this report to what is available to code users today, and we avoid predicting future improvements that may or may not actual come to pass. An exception that we make is in presenting results for an improved thermal scattering model currently being testing using advanced versions of NJOY and MCNP that are not currently available to users, but are planned for release in the not too distant future. The other exception is to show comparisons between experimentally measured water cross sections and preliminary ENDF/B-VII thermal scattering law, S({alpha},{beta}) data; although these data are strictly
Measurement of neutron fields experienced in commercial air flights
International Nuclear Information System (INIS)
Recently, the International Commission on Radiological Protection (ICRP) published new recommendations on radiation protection (ICRP 60), based on the reanalysis of the atomic bomb survivor data and other epidemiological studies. To reflect these new risk estimates, the regulatory agency in Canada, the Atomic Energy Control Board (AECB), has proposed to reduce the annual stochastic dose limit from 50 to 20 mSv for an atomic radiation worker and from 5 to 1 mSv for the general public. These annual doses are expected to be comparable to those received by commercial air crews. Measurement of the neutron component of the high-altitude, radiation field is most difficult and, up until very recently, required sophisticated electronic equipment. With the development of the bubbler detector - a passive, direct-reading, and accurate neutron monitor - routine measurements of these fields are now possible. This paper reports preliminary results from a study in which bubble detectors are routinely worn by ten Air Canada pilots for a period of 1 yr
Monte Carlo calculations of neutron thermalization in a heterogeneous system
International Nuclear Information System (INIS)
The slowing down of neutrons in a heterogeneous system (a slab geometry) of uranium and heavy water has been investigated by Monte Carlo methods. Effects on the neutron spectrum due to the thermal motions of the scattering and absorbing atoms are taken into account. It has been assumed that the speed distribution of the moderator atoms are Maxwell-Boltzmann in character
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment
International Nuclear Information System (INIS)
Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)
Energy Technology Data Exchange (ETDEWEB)
Artem’ev, V. A., E-mail: niitm@inbox.ru [Research Institute of Materials Technology (Russian Federation); Nezvanov, A. Yu. [Moscow State Industrial University (Russian Federation); Nesvizhevsky, V. V. [Institut Max von Laue—Paul Langevin (France)
2016-01-15
We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10{sup -7}–10{sup -3} eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.
Calculation of neutron spectra on typical irradiation location of the CFBR-II reactor
International Nuclear Information System (INIS)
Neutron energy spectra were simulated by the MCNP code. The neutron energy spectra and corresponding average energy of off-coupling box, irradiation channel and outer surface of the off-coupling cover were calculated. The results indicate that about 90% neutrons are in the energy range of 0.05-3 MeV. The average neutron energy of off-coupling box and irradiation channel present 'S' shape along distance, and space asymmetry must be considered. The average neutron energy above off-coupling cover's 45 degree woof fluctuates slightly and it is an appropriate irradiation area. (authors)
A Preliminary Assessment of Radiation and Air Activation for the Neutron Science Facility in RAON
International Nuclear Information System (INIS)
The works will stay in the DAQ room during an operation for about 1 month. In order to test the characteristics of the detector, the workers are also possible to access the TOF hall after a shutdown. Therefore, the shielding analysis of the NSF is required to meet the above purpose. In view of this, we performed the calculation of the shielding concrete thickness required for a target room by using MCNPX code with a neutron source obtained from Institute for Basic Science (IBS). In addition, the dose distribution and air activation for the entire space in NSF were evaluated using MCNPX and SP-FISPACT 2010 codes. We have performed the shielding calculation with the neutron source produced from the C(d,n) reactions. The concrete thickness was evaluated for all directions of the target room, and it was confirmed by performing the calculation of dose distribution to the entire space. However, the dose rate for the beam line was high. The radioactivity of radionuclides at TOF hall do not exceeded the air concentration and release limits
International Nuclear Information System (INIS)
The method to calculate the response function of spherical BF3 proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF3 proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within ±10%. (author)
Air quality along motorways. Measuring and modelling calculations
International Nuclear Information System (INIS)
This report describes the air quality along Koege Bugt motorway, one of the most trafficked sections in Denmark. A number of measurements have been carried out along Koege Bugt motorway at Greve for a three-month period in the autumn of 2003. For the first time in Denmark, NOx were measured with high time dissolution from different distances of the motorway. Furthermore, a number of meteorological parameters were measured in order to map local meteorological conditions. An air quality model describing dispersal and conversion has been made on the basis of the OML model. The OML model is modified in order to take traffic-made turbulence into consideration. The model has been evaluated through comparisons between measurements and simulated calculations. Furthermore, simulated calculations for the year 2003 has been made for comparison with extreme values. (BA)
Calculation of the dynamic air flow resistivity of fibre materials
DEFF Research Database (Denmark)
Tarnow, Viggo
1997-01-01
The acoustic attenuation of acoustic fiber materials is mainly determined by the dynamic resistivity to an oscillating air flow. The dynamic resistance is calculated for a model with geometry close to the geometry of real fibre material. The model constists of parallel cylinders placed randomly....... Two case are treated: flow perpendicular to the cylinder axes, and flow parallel to the axes. In each case two new approximate procedures were used. In the first procedure, one solves the equation of flow in a Voronoi cell around the fiber, and averages over the distribution of the Voronoi cells.......The second procedure is an extension to oscillating air flow of the Brinkman self-consistent procedure for dc flow. The procedures are valid for volume concentrations of cylinders less than 0.1. The calculations show that for the density of fibers of interest for acoustic fibre materials the simple self...
International Nuclear Information System (INIS)
Analysis is made for the effect of mathematical model accuracy of the system concerned on the calculation results using the BRAND program system. Consideration is given to the impact of the following factors: accuracy of neutron source energy-angular characteristics description, various degrees of system geometry approximation, adequacy of Monte-Carlo method estimation to a real physical neutron detector. The calculation results analysis is made on the basis of the experiments on leakage neutron spectra measurement in spherical lead assemblies with the 14 MeV-neutron source in the centre. 4 refs.; 2 figs.; 10 tabs
Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations
Energy Technology Data Exchange (ETDEWEB)
Zare, Nafiseh [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of); Fadaei, Amir Hosein, E-mail: Fadaei_amir@aut.ac.i [Faculty of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnique), Hafez Street, Tehran (Iran, Islamic Republic of); Rahgoshay, Mohammad [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of); Fadaei, Mohammad Mehdi [Department of Electrical Engineering, Faculty of Engineering, Central Tehran Branch, Islamic Azad University, Punak Square, Tehran (Iran, Islamic Republic of); Kia, Shabnam [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of)
2010-11-15
Research highlights: {yields} Reactor behavior affects from reciprocal effects between neutronic and thermo-hydraulic. {yields} Reliable reactor analysis requires coupling of neutronic and thermal-hydraulic calculation. {yields} Iterative process can be used to perform neutronic and thermal-hydraulic calculation. - Abstract: In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.
Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations
International Nuclear Information System (INIS)
Research highlights: → Reactor behavior affects from reciprocal effects between neutronic and thermo-hydraulic. → Reliable reactor analysis requires coupling of neutronic and thermal-hydraulic calculation. → Iterative process can be used to perform neutronic and thermal-hydraulic calculation. - Abstract: In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.
Experimental and calculated calibration of ionization chambers with air circulation
Peetermans, A
1972-01-01
The reports describes the method followed in order to calibrate the different ionization chambers with air circulation, used by the 'Health Physics Group'. The calculations agree more precisely with isotopes cited previously (/sup 11/C, /sup 13/N, /sup 15/O, /sup 41 /Ar, /sup 14/O, /sup 38/Cl) as well as for /sup 85/Kr, /sup 133/Xe, /sup 14/C and tritium which are used for the experimental standardisation of different chambers.
Calculations of neutron penetration through graphite medium with Monte Carlo code MCNP
International Nuclear Information System (INIS)
Experiments for fast neutron penetration through graphite are analysed with the continuous energy Monte Carlo code MCNP. Reaction rates and energy spectra obtained with the MCNP are compared with measured values and calculated ones with McBEND code. And validity of penetration calculation with the MCNP is comfirmed. In addition, it is revealed that the MCNP code using Weight-Window method is well applicable to calculations of neutron penetration through graphite up to 70 cm in depth. (author)
International Nuclear Information System (INIS)
An investigation has been carried out concerning the transmission of thermal and fast neutrons in air filled annular ducts through laminated Fe-D2O shields. Measurements have been made with annular air gaps of 0.5, 1.0, 1.5 and 2.0 cm, at a duct length of half a meter. The neutron fluxes were determined with a foil activation technique. The thermal flux was theoretically and experimentally divided into three components, a streaming, a leakage and an albedo component. The fast flux was similarly divided into a streaming component and a 'leakage' component. A calculational model to predict the components was then developed and fitted, to the data obtained by experiments. The model reported here for prediction of neutron attenuation in ducted configurations may be applied to straight annular ducts of arbitrary dimensions and material configurations but is especially designed for the problems met with in short ducts
A Neutron Burst Associated with an Extensive Air Shower?
Alves, Mauro; Martin, Inacio; Shkevov, Rumen; Gusev, Anatoly; De Abreu, Alessandro
2016-07-01
A portable and compact system based on a He-3 tube (LND, USA; model 25311) with an area of approximately 250 cm² and is used to record neutron count rates at ground level in the energy range of 0.025 eV to 10 MeV, in São José dos Campos, SP, Brazil (23° 12' 45" S, 45° 52' 00" W; altitude, 660m). The detector, power supply, digitizer and other hardware are housed in an air-conditioned room. The detector power supply and digitizer are not connected to the main electricity network; a high-capacity 12-V battery is used to power the detector and digitizer. Neutron counts are accumulated at 1-minute intervals continuously. The data are stored in a PC for further analysis. In February 8, 2015, at 12 h 22 min (local time) during a period of fair weather with minimal cloud cover (shower that occurred over the detector.
International Nuclear Information System (INIS)
The tritium production density, kerma heat production density, dose and certain integral values of scalar neutron spectra in bare and graphite-reflected lithium-fluoride piles irradiated with D-T neutrons were evaluated from the pulse height distribution of a miniature NE213 neutron spectrometer with UFO data processing code, and compared with the values calculated with MORSE-CV Monte Carlo code. (author). 8 refs.; 1 fig.; 2 tabs
International Nuclear Information System (INIS)
The MGPRAKTINETs computer code for the BESM-6 computer intended for calculation of zone average trmal neutron group fluxes and functionals is described. The neutron spatial-energy distribution in a multizone cyllindrically-symmetric reactor cell is calculated by the operator splitting method. For the solution of the spatial part of the problem the method of surface pseudosources (Gsub(N)-approximation) in approximation of plane derivatives from the energy neutron current is employed. The energy part of the problem is solved in a multigroup approximation. Computer code efficiency has been demonstrated by calculation of two-zone cells with internal and external sources of the cell with on additional absorber and RBMK cell with reduction of the latter to cylindrical geometry. It is shown that the approximation of plane derivatives of neutron energy current allows calculating reactor cell characteristics with a sufficient for design calculations accuracy
AIRDIF, Neutron and Gamma Doses from Nuclear Explosion by 2-D Air Diffusion
International Nuclear Information System (INIS)
1 - Description of problem or function: AIRDIF is a two-dimensional atmospheric radiation diffusion code designed to calculate neutron and gamma doses in the environment of a nuclear explosion. It calculates radiation fluxes in one-dimensional homogeneous air, or two-dimensional variable density air. The results are limited by the assumptions inherent in diffusion theory: the region of interest must be large compared to the radiation mean free path, the spatial flux gradients must not be steep, flux varies linearly with the cosine of the direction angle. The code requires as input data neutron and gamma source spectra, coupled neutron-gamma multigroup cross sections, and, for two- dimensional problems, a set of mass integral scaling (MIS) coefficients. These latter are calculated from an AIRDIF output flux file for a one-dimensional problem by the auxiliary program MISFIT, using a least squares fitting technique to Murphy's radiation transmission equation. MISFIT can also be used to calculate one- dimensional MIS doses. The MIS coefficients and doses can be input to AIRDIF, in two- dimensional mode to calculate 2-D fluxes, doses and K-factors (the ratio of 2-D to 1-D dose). Alternatively the 2-D doses and K-factors may be computed using the output 2-D flux file of a previous AIRDIF run using the auxiliary program DOSCOMP. 2 - Method of solution: Un-collided particle flux is determined from an analytic expression describing exponential attenuation with distance. Diffusion theory is used for the flux, using un-collided flux as a source term. A central collided differencing technique is used to reduce the diffusion equation to a matrix equation, which is solved by the Successive Line Over-relaxation (SLOR) method. Total flux is calculated as the sum of collided and un-collided components. To maintain a mesh interval which has the same relationship to mean free path at all heights, an expanding non-orthogonal coordinate system is used. In homogeneous air this system
Calculation of Multisphere Neutron Spectrometer Response Functions in Energy Range up to 20 MeV
Martinkovic, J
2005-01-01
Multisphere neutron spectrometer is a basic instrument of neutron measurements in the scattered radiation field at charged-particles accelerators for radiation protection and dosimetry purposes. The precise calculation of the spectrometer response functions is a necessary condition of the propriety of neutron spectra unfolding. The results of the response functions calculation for the JINR spectrometer with LiI(Eu) detector (a set of 6 homogeneous and 1 heterogeneous moderators, "bare" detector within cadmium cover and without it) at two geometries of the spectrometer irradiation - in uniform monodirectional and uniform isotropic neutron fields - are given. The calculation was carried out by the code MCNP in the neutron energy range 10$^{-8}$-20 MeV.
Calculation of neutron cross sections on isotopes of yttrium and zirconium
International Nuclear Information System (INIS)
Multistep Hauser-Feshbach calculations with preequilibrium corrections were made for neutron-induced reactions on yttrium and zirconium isotopes between 0.001 and 20 MeV. Recently new neutron cross-section data have been measured for unstable isotopes of these elements. These data, along with results from charged-particle simulation of neutron reactions, provide unique opportunities under which to test nuclear-model techniques and parameters in this mass region. A complete and consistent analysis of varied neutron reaction types using input parameters determined independently from additional neutron and charged-particle data. The overall agreement between calculations and a wide variety of experimental results available for these nuclei leads to increased confidence in calculated cross sections made where data are incomplete or lacking. 75 references
Calculation of multisphere neutron spectrometer response functions in energy range up to 20 MeV
International Nuclear Information System (INIS)
Multisphere neutron spectrometer is a basic instrument of neutron measurements in the scattered radiation field at charged-particles accelerators for radiation protection and dosimetry purposes. The precise calculation of the spectrometer response functions is a necessary condition of the propriety of neutron spectra unfolding. The results of the response functions calculation for the JINR spectrometer with LiI(Eu) detector (a set of 6 homogeneous and 1 heterogeneous moderators, 'bare' detector within cadmium cover and without it) at two geometries of the spectrometer irradiation - in uniform monodirectional and uniform isotropic neutron fields - are given. The calculation was carried out by the code MCNP in the neutron energy range 10-8 - 20 MeV
Calculation of anisotropy factors for 241Am-Be neutron sources
International Nuclear Information System (INIS)
The authors calculated anisotropy factors for 241Am-Be neutron sources used for calibration of neutron-measuring devices for radiation protection purpose. In this calculation, we created a calculation model composed of following three steps: (1) calculation of α-particle spectrum at the surface of spherical cluster of AmO2, (2) calculation of neutron yield in a thick beryllium target and of neutron spectrum produced by 8Be (α,n) reactions; and (3) calculation of angular fluence distribution of neutrons emerging from two different encapsulation types of 241Am-Be neutron sources. This computation was made by combining an in-house code using the 9Be(α,n) cross section data library (JENDL/AN-2005) and the Monte Carlo code MCNP-4C. As a result, anisotropy factors in the direction perpendicular to the source capsule axis were evaluated to be 1.030 and 1.039 for 241Am-Be in a standard Amersham X3 capsule and X4 capsule, respectively. These values are in reasonable close agreement with the published experimental data. If the support structures are included in the simulation, the anisotropy factors for these neutron sources increase by about 10%. (author)
Improvement of neutron dose calculation algorithm using panasonic UD-809P type albedo TLD
International Nuclear Information System (INIS)
Panasonic UD-809P type albedo TLD mounted on a water phantom were used to measure neutron personal dose equivalent in a Korean nuclear power plant. From the measured TL readings, personal dose equivalents from thermal, epithermal and fast neutrons were evaluated by using a method adopted in a neutron dose calculation algorithm for Panasonic UD-809P type albedo TLD, which was recommended in a Panasonic TLD System User's Manual. The results showed that personal dose equivalent for fast neutrons could not be adequately evaluated in a field with high thermal neutron fraction. This seems to be related to the incomplete incidence of albedo thermal neutrons to the TLD. In order to calculate the personal dose equivalent from fast neutrons in the field condition to be encountered in a nuclear power plant, new method for the neutron dose calculation algorithm were suggested. For a known energy spectrum, it is very easy and simple to use this method for the evaluation of neutron personal dose equivalent
A general dimensional neutron diffusion calculation code: ADC
International Nuclear Information System (INIS)
A FORTRAN computer program ADC is developed for the FACOM 230-75 computer to be capable of solving eigenvalue problems of neutron diffusion equation in one, two and three spatial dimensions. The available coordinate systems are orthogonal (X), (X,Y), (X,Y,Z) and cylindrical (R,Z), (R,THETA), (R,THETA,Z). The outer boundary condition for the neutron flux can be chosen to be symmetric, zero flux or log-derivative condition. The present program can be used also for obtaining the adjoint flux. (author)
Development of Library Processing System for Neutron Transport Calculation
Energy Technology Data Exchange (ETDEWEB)
Song, J. S.; Park, S. Y.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)
2008-12-15
A system for library generation was developed for the lattice neutron transport program for pressurized water reactor core analysis. The system extracts multi energy group nuclear data for requested nuclides from ENDF/B whose data are based on continuous energy, generates hydrogen equivalent factor and resonance integral table as functions of temperature and background cross section for resonance nuclides, generates subgroup data for the lattice program to treat resonance exactly as possible, and generates multi-group neutron library file including nuclide depletion data for use of the lattice program.
Theoretical calculation of a complete set of the neutron reaction data for natural tin
International Nuclear Information System (INIS)
The interaction data of neutron with natural Tin (Nnn) have been calculated by means of the optical model (OPM), the Hauser-Feshbach Theory (HFT) and the evaporation model including the Pre-equilibrium emission (PEM) in the incident neutron energy range between 1-20 MeV. Comparing with experimental values, good agreement have been obtained
Calculation of Prompt Fission Neutron Spectra for ~(235)U (n,f)
Institute of Scientific and Technical Information of China (English)
无
2011-01-01
The prompt fission neutron spectra for neutron-induced fission of 235U at En<5 MeV are calculated using the nuclear evaporation theory with a semi-empirical model, in which the non-constant temperature and the constant temperature related to the Fermi gas model
CAREM 25: actual status of the core neutronic design. Calculation line
International Nuclear Information System (INIS)
This work follows the one titled 'Criteria for the CAREM 25 reactor core design. Neutronic aspects' presented at this congress, gives in detail the typical values regarding the core defined at this point. Besides, the neutronic calculation line used for the CAREM 25 reactor design is presented. (Author)
Transport calculation of neutron flux distribution in reflector of PW reactor
International Nuclear Information System (INIS)
Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)
One group neutron flux at a point in a cylindrical reactor cell calculated by Monte Carlo
International Nuclear Information System (INIS)
Mean values of the neutron flux over material regions and the neutron flux at space points in a cylindrical annular cell (one group model) have been calculated by Monte Carlo. The results are compared with those obtained by an improved collision probability method (author)
Precise measurement and calculation of 238U neutron transmissions
International Nuclear Information System (INIS)
The total neutron cross section of 238U has been measured above 0.5 eV in precise transmission experiments and results are compared with ENDF/B-IV. Emphasis has been on measuring transmissions through thick samples in order to obtain accurate total cross sections in the potential-resonance interference regions between resonances. 4 figures, 1 table
Systems for neutronic, thermohydraulic and shielding calculation in personal computers
International Nuclear Information System (INIS)
The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author)
Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor
International Nuclear Information System (INIS)
The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)
The adaptation of methods in multilayer optics for the calculation of specular neutron reflection
International Nuclear Information System (INIS)
The adaptation of standard methods in multilayer optics to the calculation of specular neutron reflection is described. Their application is illustrated with examples which include a glass optical flat and a deuterated Langmuir-Blodgett film. (author)
Calculation of neutron transport in plane geometry by invariant imbedding method
International Nuclear Information System (INIS)
A practical combination of invariant imbedding and transfer matrix methods was displayed in this paper. A very simple scheme for neutron transport analysis was obtained for slab materials and some results of numerical calculations are presented. (author)
International Nuclear Information System (INIS)
The characteristics of the epithermal neutron beam at BMRR were measured, calculated, and reported by R.G. Fairchild. This beam has already been used for animal irradiations. The authors anticipate that it will be used for clinical trials. Thermal and epithermal neutron flux densities distributions, and dose rate distributions, as a function of depth were measured in a lucite dog-head phantom. Monte Carlo calculations were performed and compared with the measured values
Full Core, Heterogeneous, Time Dependent Neutron Transport Calculations with the 3D Code DeCART
Hursin, Mathieu
2010-01-01
The current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, bu...
Monte Carlo calculation of fast neutron spectra inside a lead hollow cylinder
International Nuclear Information System (INIS)
A simulation study was carried out in order to investigate the enhancement of the fast neutron spectra in a sample irradiated near the core of a pool type nuclear reactor. The irradiation device consisted of a lead hollow cylinder where a thin aluminium sample holder containing the sample to be irradiated is placed. Calculations were performed considering water or beryllium+water between the simulated fission neutron source and the irradiation device; the thickness of the wall of the lead cylinder was varied up to 10 cm. The fast neutron spectra in the sample position were calculated and the average fast neutron energy, the total neutron fluence, φ (E>0.1 MeV), the fast neutron fluence, Φ (E>1 MeV), and the conversion factor, C=φ/Φ, were determined. It was found that the lead cylinder surrounding the sample induces an appreciable enhancement of the fast neutron spectra. This effect can be described using a gain factor defined, at each point, as the ratio between the fast neutron fluences with and without the irradiation device, G=Φ/Φ (t=0). The gain of the neutron fluence increases from 2 to 4.5 when the lead thickness varies between 2 and 10 cm. (orig.)
International Nuclear Information System (INIS)
Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ†) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ†) and importance weighted neutron generation time (Λ†) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ† is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function
MCNP calculations of neutron emission anisotropy caused by the GIT-12 hardware
Directory of Open Access Journals (Sweden)
Šíla Ondřej
2015-06-01
Full Text Available The MCNP6 and MCNPX calculations for the GIT-12 device in Tomsk were performed to determine the influence of the gas-puff hardware on the neutron emission anisotropy and the neutron scattering rate. A monoenergetic 2.45 MeV neutron source and F1 and F6 tallies were declared in the simulation input. A comparison between MCNP results and the measured data was made. Differences between MCNPX and MCNP6 output data were investigated. In the experiment, two nTOF scintillation detectors with the Bicron BC-408 scintillator were used to measure the neutron waveform. Four bubble BD-PND detectors were used to estimate the amount of neutrons in different places around the neutron source.
Calculation of fine neutron spectrum in irradiation holes in fuel region of JRR-3M
International Nuclear Information System (INIS)
The authors have a plan to evaluate TRU neutron cross sections based on the activation experiments by using JRR-3M. Fine neutron spectrum expressed by 107 energy group structure at irradiation holes in fuel region of JRR-3M core, which was utilized to analyze experimental data, was calculated by 2 step calculation. The first step is the whole core calculation taking account of burnup history and control rod pattern, and the second step is the irradiation hole calculation without any homogenization of irradiation hole components by taking into account of the neutron spectrum of surrounding region. Fine neutron spectra calculated by 2 step calculation were compared with the experimental results on reaction rate, both agreed within several percents relatively. In the comparison of absolute values, however, the maximum difference was up to 30 percents in the vicinity of control rods. This originates from the neutron transport effect around control rods. An improvement for the treatment of neutron transport effect is needed to get higher accuracy. (author)
Calculation of neutron detection efficiency for the thick lithium glass using Monte Carlo method
International Nuclear Information System (INIS)
The neutron detector efficiencies of a NE912 (45mm in diameter, 9.55 mm in thickness) and 2 pieces of ST601 (40mm in diameter, 3 and 10 mm in thickness respectively) lithium glasses have been calculated with a Monte Carlo computer code. The energy range in the calculation is 10 keV to 2.0 MeV. The effect of time delayed caused by neutron multiple scattering in the detectors (prompt neutron detection efficiency) has been considered
Design basis neutronics calculations for NRU-LOCA experiments
International Nuclear Information System (INIS)
The report describes the neutronics analysis for the LOCA simulation experiments in the NRU reactor. The experimental program will provide greater understanding of nuclear fuel assembly behavior during the heatup, reflood and quench sequence of a hypothetical LOCA. The decay heat and stored heat, which are the energy source in a LOCA will be simulated by fission heat provided by the NRU reactor. The reactor, the test and test operation are described
[Calculating method for crop water requirement based on air temperature].
Tao, Guo-Tong; Wang, Jing-Lei; Nan, Ji-Qin; Gao, Yang; Chen, Zhi-Fang; Song, Ni
2014-07-01
The importance of accurately estimating crop water requirement for irrigation forecast and agricultural water management has been widely recognized. Although it has been broadly adopted to determine crop evapotranspiration (ETc) via meteorological data and crop coefficient, most of the data in whether forecast are qualitative rather than quantitative except air temperature. Therefore, in this study, how to estimate ETc precisely only using air temperature data in forecast was explored, the accuracy of estimation based on different time scales was also investigated, which was believed to be beneficial to local irrigation forecast as well as optimal management of water and soil resources. Three parameters of Hargreaves equation and two parameters of McClound equation were corrected by using meteorological data of Xinxiang from 1970 to 2010, and Hargreaves equation was selected to calculate reference evapotranspiration (ET0) during the growth period of winter wheat. A model of calculating crop water requirement was developed to predict ETc at time scales of 1, 3, and 7 d intervals through combining Hargreaves equation and crop coefficient model based on air temperature. Results showed that the correlation coefficients between measured and predicted values of ETc reached 0.883 (1 d), 0.933 (3 d), and 0.959 (7 d), respectively. The consistency indexes were 0.94, 0.95 and 0.97, respectively, which showed that forecast error decreased with the increasing time scales. Forecasted accuracy with an error less than 1 mm x d(-1) was more than 80%, and that less than 2 mm x d(-1) was greater than 90%. This study provided sound basis for irrigation forecast and agricultural management in irrigated areas since the forecasted accuracy at each time scale was relatively high. PMID:25345053
Calculation of the main neutron parameters of the IEA-R1 research reactor
International Nuclear Information System (INIS)
The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined
International Nuclear Information System (INIS)
Measurements of neutron pulse time-width and intensity have been carried out on grids of small moderators placed side by side and decoupled by cadmium strips; a moderator concept introduced by the authors through previous publications. Transport calculations are based on the standard reactor code DOT 3.5 with the ENDF-B IV nuclear data library. (orig.)
MC2-2: a code to calculate fast neutron spectra and multigroup cross sections
International Nuclear Information System (INIS)
MC2-2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC2-2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC2-2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC2-2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC2-2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers
3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F
Energy Technology Data Exchange (ETDEWEB)
Herreras, Y; Cabellos, O; Perlado, J M [Instituto de Fusion Nuclear (DENIM)/ETSII/Universidad Politecnica, Madrid (Spain); Lafuente, A; Sordo, F [Universidad Politecnica de Madrid (UPM), Madrid (Spain)], E-mail: yuri@denim.upm.es
2008-05-15
This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.
International Nuclear Information System (INIS)
Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method
A group of neutronics calculations in the MNSR using the MCNP-4C code
International Nuclear Information System (INIS)
The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl4-Al) to 19.75 % LEU fuel (UO2). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ιp l and βeff). (authors)
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
2002-01-01
This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.
Some neutronics calculations for the VVER-1000 reactors using SRAC and MCNP5
International Nuclear Information System (INIS)
This paper presents the results of neutronics calculations using the deterministic and Monte Carlo methods (the SRAC and MCNP5 codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The codes use different methods and different nuclear data. The power distribution in each fuel assembly and k-eff values were calculated for the case of benchmark problem and the results show a good agreement between the SRAC and MCNP5 calculations. Then, typical neutronics parameter of VVER-1000/V392 such as power distribution, infinity multiplication factor (k-inf) for fuel assemblies, effective multiplication factor (k-eff), peaking factor and Doppler coefficient were presented and compared between using SRAC and MCNP5. The aim of the study is to verify the calculation methods and calculation codes as well as to obtain insight into the neutronics characteristics of the VVER- 1000/V392 reactor core. (author)
Calculation of neutron importance function in fissionable assemblies using Monte Carlo method
International Nuclear Information System (INIS)
The purpose of the present work is to develop an efficient solution method to calculate neutron importance function in fissionable assemblies for all criticality conditions, using Monte Carlo Method. The neutron importance function has a well important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating adjoint flux through out solving the Adjoint weighted transport equation with deterministic methods. However, in complex geometries these calculations are very difficult. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on physical concept of neutron importance has been introduced for calculating neutron importance function in sub-critical, critical and supercritical conditions. For this means a computer program has been developed. The results of the method has been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries and their correctness has been approved for all three criticality conditions. Ultimately, the efficiency of the method for complex geometries has been shown by calculation of neutron importance in MNSR research reactor
International Nuclear Information System (INIS)
The application of neutron coincidence counting to the assay of special nuclear material involves a major correction for neutron multiplication. The correction commonly used at present requires an accurate knowledge of the intensity ratio of neutrons from (α,n) reactions to those from spontaneous fission. This paper covers various factors, which need to be evaluated in order to assess their importance, in the calculation of (α,n) neutron production using measured thick target yields. They include: accuracy of (α,n) thick target yield measurements; errors introduced by deriving yields in compounds from the measured yields in the constituents and vice-versa; the likely effect of neglecting the difference of α-particle stopping power between Pu and U on the calculated neutron yield from mixed oxide fuel pellets; the intensity of neutrons produced from 1 to 2% of Al used to alloy plutonium metal; the intensity of neutrons produced in Al, used as canning material, from α-particles escaping from the surface layers of oxide or metal fuel; and neutron production from oxygen in the air spaces of powdered PuO2 prior to sintering. (author)
Neutronic calculations for CANDU thorium systems using Monte Carlo techniques
Saldideh, M.; Shayesteh, M.; Eshghi, M.
2014-08-01
In this paper, we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium (CANDU) reactors. The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction. Four different fuel compositions have been selected for analysis. We have obtained the infinite multiplication factor, k∞, under full power operation of the reactor over 8 years. The neutronic flux distribution in the full core reactor has already been investigated.
Coupled hydro-neutronic calculations for fast burst reactor accidents
International Nuclear Information System (INIS)
Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor
Filtered thermal neutron captured cross-sections measurements and decay heat calculations
International Nuclear Information System (INIS)
Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (Rcd) of 420 and neutron flux (Φth) of 1.6x106 n/cm2.s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51V, 55Mn, 180Hf and 186W by the activation method relative to the standard reaction 197Au(n,g)198Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235U, 238U, 239Pu and 232Th are introduced in this report. (author)
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
International Nuclear Information System (INIS)
In line with the actual requirements and based upon the specific characteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poisoning Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.
Calculation of diffusion coefficients in air-metal thermal plasmas
Energy Technology Data Exchange (ETDEWEB)
Cressault, Y; Gleizes, A [Universite de Toulouse, UPS, INPT, LAPLACE (Laboratoire Plasma et Conversion d' Energie), 118 route de Narbonne, F-31062 Toulouse Cedex 9 (France)
2010-11-03
This paper presents the combined diffusion coefficients of metal vapours (silver, copper and iron) in air thermal plasmas for temperatures ranging from 300 to 30 000 K. The theory used to calculate these coefficients is remembered and validated by comparison with the literature values in several cases such as Ar-He, Ar-Cu and N{sub 2}-O{sub 2} mixtures. The results are discussed showing the influences of the metal concentration, of the vapour nature and of the pressure. The results show rather similar behaviour for the three metals. The maximum values of the combined ordinary diffusion coefficient in the evolution with temperature are obtained for temperature around 10 000 K but this peak is shifted to the highest temperatures when the metal proportion increases. Another result shows that the diffusion coefficient decreases when pressure increases.
International Nuclear Information System (INIS)
Biomonitors were used as part of a pollution study of Buenos Aires city atmosphere under the International Atomic Energy Agency Research Contract ARG 7251, from the Co-ordinated Research Programme on Applied Research on Air Pollution using Nuclear Related Analytical Techniques. Lichens were primarily selected as indicators. Two different approaches were conducted, direct sampling of Parmotrema reticulatum, at a few places and the use of lichen bags, filled with Usnea sulcata from a northern national park, and hung at different sites. Simultaneously, tree bark was tried as biomonitor. Platanus acerifolia and Melia azedarach were selected as candidates, for being the most common trees in the city, but only P. acerifolia was analyzed. All the samples were analyzed using instrumental neutron activation analysis at the Ezeiza Atomic Centre of the National Atomic Energy Commission. RA-3) reactor was used for the irradiations, determining: As, Ba, Br, Ce, Co, Cr, Cs, Eu, Fe, Hf, K, La, Lu, Na, Rb, Sb, Sc, Sm, Ta, Tb, Th, U, Yb and Zn. Concentration values for P. reticulatum compared well with values from literature. For U. sulcata differences were found among the tested sites and also, for some elements an increasing trend with time was observed. Enrichment factors calculated using Sc as reference and Mason's crustal average concentrations showed vehicules and refuse incineration as contributing sources to the aerosol. Tree bark from Buenos Aires and from a smaller city with mainly agricultural activities were analyzed and the results are coincident with those from lichens. This work is the first and preliminar contribution to the study of Buenos Aires aerosol using biomonitors. (author)
International Nuclear Information System (INIS)
The Algerian research reactor (Es-Salam) is a 15 MW heavy water reactor type, operating since 1992. It became essential to characterize the neutron field in the most useful irradiation positions, in order to guarantee the accuracy in the application of k0-neutron activation analysis (k0-NAA). Experimental value of the thermal to epithermal neutron flux ratio (f) and of the deviation of the epithermal neutron spectrum from 1/E shape (α) were determined using different methods. This work focuses the verification of Monte Carlo neutron flux calculation in typical irradiation channel. Comparison of the results for parameter f obtained experimentally and by Monte Carlo simulations shows good agreement in the irradiation channel studied. The difference between both results is about 2.08%. (author)
Neutronic calculations of PARR-1 cores using leu-silicide fuel. [leu (low enriched uranium)
Energy Technology Data Exchange (ETDEWEB)
Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.
1991-08-01
Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing Low Enriched Uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full power operation and the equilibrium cores. The burnup study of the equilibrium core and calculations for discharged fuel inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis.
Ab initio calculations versus polarized neutron diffraction for the spin density of free radicals
International Nuclear Information System (INIS)
The determination of the magnetization distribution using polarized neutron diffraction has played a key role during the last twenty years in the field of molecular magnetism. This distribution can also be obtained by first principle ab initio calculations. Such calculations always rely on approximations and the question that arises is to know whether the obtained results are reliable enough to represent accurately the properties of these molecules. The comparison between polarized neutron experimental results and ab initio calculations has turned to provide stringent tests for these methods. In the resent article a comparison between experimental and theoretical results is made and is illustrated by examples based on magnetic free radicals. (author)
Intercomparison of Monte Carlo and SN sensitivity calculations for a 14 MeV neutron benchmark
International Nuclear Information System (INIS)
An inter-comparison has been performed of probabilistic and deterministic sensitivity calculations with the objective to check and validate the Monte Carlo technique for calculating point detector sensitivities as being implemented in MCSEN, a local version of the MCNP4A code. A suitable 14 MeV neutron benchmark problem on an iron assembly has been considered to this end. Good agreement has been achieved for the calculated individual sensitivity profiles, the uncertainties and the neutron flux spectra as well. It is concluded that the Monte Carlo technique for calculating point detector sensitivities and related uncertainties as being implemented in MCSEN is well qualified for sensitivity and uncertainty analyses of fusion neutronics integral experiments. (orig.)
International Nuclear Information System (INIS)
Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab
Ab initio calculation of the neutron-proton mass difference
CERN. Geneva
2015-01-01
The existence and stability of atoms relies on the fact that neutrons are more massive than protons. The mass difference is only 0.14% of the average and has significant astrophysical and cosmological implications. A slightly smaller or larger value would have led to a dramatically different universe. After an introduction to the problem and to lattice quantum chromodynamics (QCD), I will show how this difference can be computed precisely by carefully accounting for electromagnetic and mass isospin breaking effects in lattice computations. I will also report on results for splittings in the \\Sigma, \\Xi, D and \\Xi_{cc} isospin multiplets, some of which are predictions. The computations are performed in lattice QCD plus QED with four, non-degenerate quark flavors.
Monte Carlo perturbation theory in neutron transport calculations
International Nuclear Information System (INIS)
The need to obtain sensitivities in complicated geometrical configurations has resulted in the development of Monte Carlo sensitivity estimation. A new method has been developed to calculate energy-dependent sensitivities of any number of responses in a single Monte Carlo calculation with a very small time penalty. This estimation typically increases the tracking time per source particle by about 30%. The method of estimation is explained. Sensitivities obtained are compared with those calculated by discrete ordinates methods. Further theoretical developments, such as second-order perturbation theory and application to k/sub eff/ calculations, are discussed. The application of the method to uncertainty analysis and to the analysis of benchmark experiments is illustrated. 5 figures
Calculation of the decay power of fission products considering neutron capture transformation
International Nuclear Information System (INIS)
The decay power of fission products has been calculated taking into consideration the neutron capture transformation of each nuclide and its beta decay. The nuclear data library contains 1114 nuclides of which 144 are stable. Neutron capture transformation is considered for 59 nuclides, 31 of which are stable. The atom number of each nuclide is calculated analytically with code DCHAIN. The effect of neutron capture transformation in the decay power of fission products was examined by varying the neutron spectrum, neutron flux, fissioning nuclide, and irradiation and cooling time. From the results obtained the following were revealed: The effect of neutron capture increases with neutron flux and irradiation time, and it becomes salient beyond 105 sec in cooling time. It is small for less than the 104 sec which is important in the design of ECCS (emergency core cooling system) of a light-water reactor. In this region the decay power changes are small, less than 0.2%, by the neutron capture for the thermal fission of 235U irradiated for one year to thermal neutron flux 3 x 1013 n/cm2/sec. The effect of neutron capture has peaks around cooling time 106 sec and 108 sec; it is negligible beyond 109 sec. The changes in decay power are 2.4%, 10.5% and 0.2% at cooling time 106 sec, 108 sec and 109 sec, respectively, in the above irradiation. Around 106 sec, the change in decay power is mainly from the contributions of 134Cs (17%), sup(148m)Pm(60%) and 148Pm(14%). Around 108 sec 134Cs(98%) alone contributes to the change in decay power. (author)
Dose calculation from a D-D-reaction-based BSA for boron neutron capture synovectomy
International Nuclear Information System (INIS)
Monte Carlo simulations were carried out to calculate dose in a knee phantom from a D-D-reaction-based Beam Shaping Assembly (BSA) for Boron Neutron Capture Synovectomy (BNCS). The BSA consists of a D(d,n)-reaction-based neutron source enclosed inside a polyethylene moderator and graphite reflector. The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield at the knee phantom. Then neutron dose was calculated at various depths in a knee phantom loaded with boron and therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose were determined. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values.
Dose calculation from a D-D-reaction-based BSA for boron neutron capture synovectomy
Energy Technology Data Exchange (ETDEWEB)
Abdalla, Khalid [Department of Physics, Hail University, Hail (Saudi Arabia)], E-mail: khalidafnan@uoh.edu.sa; Naqvi, A.A. [Department of Physics, King Fahd University of Petroleum and Minerals and Center for Applied Physical Sciences, Box No. 1815, Dhahran 31261 (Saudi Arabia)], E-mail: aanaqvi@kfupm.edu.sa; Maalej, N.; Elshahat, B. [Department of Physics, King Fahd University of Petroleum and Minerals and Center for Applied Physical Sciences, Box No. 1815, Dhahran 31261 (Saudi Arabia)
2010-04-15
Monte Carlo simulations were carried out to calculate dose in a knee phantom from a D-D-reaction-based Beam Shaping Assembly (BSA) for Boron Neutron Capture Synovectomy (BNCS). The BSA consists of a D(d,n)-reaction-based neutron source enclosed inside a polyethylene moderator and graphite reflector. The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield at the knee phantom. Then neutron dose was calculated at various depths in a knee phantom loaded with boron and therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose were determined. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values.
Delayed neutron spectra and their uncertainties in fission product summation calculations
Energy Technology Data Exchange (ETDEWEB)
Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)
1997-03-01
Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)
Neutron and gamma-ray streaming calculations for the ETF neutral-beam injectors
International Nuclear Information System (INIS)
The tritium plasma of the Engineering Test Facility (ETF) fusion reactor will be heated and ignited by the injection of neutral deuterium. Since the deuterons must be injected through straight ducts into the plasma, the neutron and secondary gamma radiation produced as a result of the D-T reactions will stream directly into the neutral beam injectors and lead to adverse effects in vital components. The radiation leaking through the injection ports will be comprised of approx. 14 MeV neutrons (from the D-T reactions) plus a low-energy neutron and secondary gamma ray distribution that results from the interactions of the energetic neutrons with the plasma liner and the primary shielding about the torus. In this paper two-dimensional radiation transport calculations carried out to estimate the effects on the injector components of radiation streaming through the injection duct will be described and the results of these calculations will be presented and discussed
Neutron and gamma-ray streaming calculations for the ETF neutral-beam injectors
Energy Technology Data Exchange (ETDEWEB)
Lillie, R.A.; Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.
1981-01-01
The tritium plasma of the Engineering Test Facility (ETF) fusion reactor will be heated and ignited by the injection of neutral deuterium. Since the deuterons must be injected through straight ducts into the plasma, the neutron and secondary gamma radiation produced as a result of the D-T reactions will stream directly into the neutral beam injectors and lead to adverse effects in vital components. The radiation leaking through the injection ports will be comprised of approx. 14 MeV neutrons (from the D-T reactions) plus a low-energy neutron and secondary gamma ray distribution that results from the interactions of the energetic neutrons with the plasma liner and the primary shielding about the torus. In this paper two-dimensional radiation transport calculations carried out to estimate the effects on the injector components of radiation streaming through the injection duct will be described and the results of these calculations will be presented and discussed.
Combining neutron and X-ray imaging to study air and water behaviour in the soil macropores
Snehota, Michal; Sobotkova, Martina; Jelinkova, Vladimira; Kaestner, Anders
2016-04-01
Infiltration of water and gas trapping in soil macropores were investigated on intact sample of coarse sandy loam soil (Cambisol series) taken from the B horizon by combined X-ray and neutron tomography imaging. The soil under study is known for the occurrence of the preferential flow, in which a majority of the water flux is conducted through small, highly conductive, fraction of the soil volume. Experiment performed in the NEUTRA beamline of Paul Scherrer Institut consisted of two infiltration episodes during which a layer of heavy and light water mixture was maintained on the sample surface created a ponding boundary condition. The initial state of the sample was recorded by one X-ray and two neutron scans prior to the first infiltration. Another 20 neutron tomograms were acquired during the following 25 hours of the experiment. Fine co-registration of the reconstructed X-ray and neutron tomograms was performed. Then, bi-variate histograms helped to identify the thresholds that were subsequently used for segmentation of the macropores from the X-ray tomograms. The segmented regions served as a binary mask for calculating the water volume using the neutron tomograms. Volume of water and subsequently the average water content in the macropore system were calculated. Results then quantitatively show the extent of the water content reduction in the macropores during the second infiltration that was caused by enhanced air trapping in the wet soil.
International Nuclear Information System (INIS)
Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n)4He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed. The measured and calculated neutron energy spectra obtained for the attenuation experiments are in excellent agreement for shield compositions and thicknesses up to 412 g/cm2 thick. The calculated gamma-ray spectra agree with the measured data to within 15% for the slabs containing stainless steel and borated polyethylene and within a factor of 5 when Hevimet is included in the shield composition. The calculated neutron spectra obtained for the streaming experiments are in good agreement with the measured data for the on-axis detector position. For the off-axis detector locations, the calculations overestimate the measurements by as much as factor of 5 depending on detector location. (author)
Calculation of the energy dependent efficiency of gridded 3He fast neutron ionization chambers
International Nuclear Information System (INIS)
The relative efficiency function for total energy events in a 3He fast neutron ionization chamber has been calculated with a Monte Carlo approach. It is shown that the efficiency function applicable to a point isotropic source located near the surface of the spectrometer differs significantly from that obtained in standard calibration procedures using neutrons from the 7Li(p,n)7Be reaction for Esub(n) > 1.5 MeV. (orig.)
Neutron dosimetry and damage calculations for the ATR-A1 irradiation
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R.; Ratner, R.T. [Pacific Northwest National Lab., Richland, WA (United States)
1998-09-01
Neutron fluence measurements and radiation damage calculations are reported for the collaborative US/Japan ATR-A1 irradiation in the Advanced Test Reactor (ATR) at Idaho National Engineering Laboratory (INEL). The maximum total neutron fluence at midplane was 9.4 {times} 10{sup 21} n/cm{sup 2} (5.5 {times} 10{sup 21} n/cm{sup 2} above 0.1 MeV), resulting in about 4.6 dpa in vanadium.
Neutron dosimetry and damage calculations for the HFIR-JP-23 irradiations
International Nuclear Information System (INIS)
Neutron fluence measurements and radiation damage calculations are reported for the joint U.S. Japanese experiment JP-23, which was conducted in target position G6 of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum neutron fluence at midplanes was 4.4E+22 n/cm2 resulting in about 9.0 dpa in type 316 stainless steel
Neutron dosimetry and damage calculations for the HFIR-JP-23 irradiations
International Nuclear Information System (INIS)
Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment JP-23, which was conducted in target position G6 of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum neutron fluence at midplane was 4.4E+22 n/cm2 resulting in about 9.0 dpa in type 316 stainless steel
Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data
International Nuclear Information System (INIS)
Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author)
Neutron dosimetry and damage calculations for the HFIR-JP-23 irradiations
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R.; Ratner, R.T. [Pacific Northwest National Lab., Richland, WA (United States)
1996-10-01
Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment JP-23, which was conducted in target position G6 of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum neutron fluence at midplane was 4.4E+22 n/cm{sup 2} resulting in about 9.0 dpa in type 316 stainless steel.
Calculated neutron KERMA factors based on the LLNL ENDL data file. Volume 27
International Nuclear Information System (INIS)
Neutron KERMA factors calculated from the LLNL ENDL data file are tabulated for 15 composite materials and for the isotopes or elements in the ENDL file from Z = 1 to Z = 29. The incident neutron energies range from 1.882 x 10-5 to 20. MeV for the composite materials and from 1.30 x 10-9 to 20. MeV for the isotopes and elements
Chen, Yong-Jing; Min, Jia; Liu, Ting-Jin; Shu, Neng-Chuan
2013-01-01
The prompt fission neutron spectra for neutron-induced fission of 233U for low energy neutrons (below 6 MeV) are calculated using the nuclear evaporation theory with a semi-empirical method, in which the partition of the total excitation energy between the fission fragments for the nth+233U fission reactions are determined with the available experimental and evaluation data. The calculated prompt fission neutron spectra agree well with the experimental data. The proportions of high- energy ou...
Demonstration of core neutronic calculation for research and training reactors via SCALE4.4
International Nuclear Information System (INIS)
In this work, full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 system. KENOV.a module of SCALE4.4 system is utilized for full core neutronic analysis. The ORIGEN-S module is also coupled with the KENOV.a module to perform burnup dependent core analyses. Results of control rod worths for 1st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. (author)
International Nuclear Information System (INIS)
At the research reactor WWR-M during the long period, the study of neutron cross sections for nuclei, important as for nuclear physics investigations, so as for applied purposes have been fulfilled. Applied purposes include, among others, the production of radioactive isotopes for practical use. This paper covers the results of radioisotope program development, based on the neutron fluxes in the reactor core, and also the formation of the specific neutron data library for nuclear data support of radioisotope accumulation calculations at reactor
Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor
International Nuclear Information System (INIS)
A detailed MCNP model of the Advanced Neutron Source Reactor has been developed. All reactor components inside the reflector tank were included, and all components were highly segmented. Neutron and photon multigroup flux spectra have been calculated for each segment in the model, and thermal-to-fast neutron flux ratios were determined for each component segment. Axial profiles of the spectra are provided for all components of the reactor. Individual segment statistical uncertainties were limited wherever possible, and the group fluxes for all important reflector components have a standard deviation below 10%
Realistic shell-model calculations for neutron deficient Sn isotopes
Energy Technology Data Exchange (ETDEWEB)
Andreozzi, F.; Coraggio, L.; Covello, A.; Gargano, A.; Kuo, T.T.; Li, Z.B.; Porrino, A. [Dipartimento di Scienze Fisiche, Universita di Napoli Federico II]|[Istituto Nazionale di Fisica Nucleare, Mostra d`Oltremare, Pad. 20, 80125 Napoli (Italy)]|[Department of Physics, SUNY, Stony Brook, New York 11794 (United States)
1996-10-01
We have performed shell-model calculations for {sup 102,103,104,105}Sn using two realistic effective interactions derived from the Bonn A and Paris nucleon-nucleon potentials, respectively. From the comparison of the calculated spectra of {sup 104}Sn and {sup 105}Sn with the experimental ones it turns out that the best agreement is obtained with the weaker tensor force potential (Bonn A). This agreement appears to be significantly better than for other nuclear regions, such as the {ital sd} shell, and thus encourages use of modern realistic potentials in shell-model calculations for medium- and heavy-mass nuclei. In addition, it supports confidence in our predictions of the spectra of the hitherto unknown isotopes {sup 102}Sn and {sup 103}Sn. {copyright} {ital 1996 The American Physical Society.}
Realistic shell-model calculations for neutron deficient Sn isotopes
International Nuclear Information System (INIS)
We have performed shell-model calculations for 102,103,104,105Sn using two realistic effective interactions derived from the Bonn A and Paris nucleon-nucleon potentials, respectively. From the comparison of the calculated spectra of 104Sn and 105Sn with the experimental ones it turns out that the best agreement is obtained with the weaker tensor force potential (Bonn A). This agreement appears to be significantly better than for other nuclear regions, such as the sd shell, and thus encourages use of modern realistic potentials in shell-model calculations for medium- and heavy-mass nuclei. In addition, it supports confidence in our predictions of the spectra of the hitherto unknown isotopes 102Sn and 103Sn. copyright 1996 The American Physical Society
Development of an effective delayed neutron fraction calculation code, BETA-K
Energy Technology Data Exchange (ETDEWEB)
Kim, Taek Kyum; Song, Hoon; Kim, Young Il; Kim, Young In; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-08-01
BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method (NEM), has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction({beta}{sub eff}), neutron lifetime(l{sub eff}), fission spectrum ({chi}-bar) and fission yield data({nu}) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356 {mu}sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER. (author). 9 refs., 6 figs., 12 tabs.
Development of an effective delayed neutron fraction calculation code for hexagonal core
International Nuclear Information System (INIS)
BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method(NEM) of hexagonal geometric core, has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction(betaeff) and neutron lifetime(leff) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356μ sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER
HEINBE; the calculation program for helium production in beryllium under neutron irradiation
International Nuclear Information System (INIS)
HEINBE is a program on personal computer for calculating helium production in beryllium under neutron irradiation. The program can also calculate the tritium production in beryllium. Considering many nuclear reactions and their multi-step reactions, helium and tritium productions in beryllium materials irradiated at fusion reactor or fission reactor may be calculated with high accuracy. The calculation method, user's manual, calculated examples and comparison with experimental data were described. This report also describes a neutronics simulation method to generate additional data on swelling of beryllium, 3,000-15,000 appm helium range, for end-of-life of the proposed design for fusion blanket of the ITER. The calculation results indicate that helium production for beryllium sample doped lithium by 50 days irradiation in the fission reactor, such as the JMTR, could be achieved to 2,000-8,000 appm. (author)
CRONOS: A modular computational system for neutronic core calculations
International Nuclear Information System (INIS)
The CRONOS code has been designed to provide all the computational means needed for Pressurized Water Reactor calculations, including design, fuel management, follow up and accidents. CRONOS allows steady state, kinetic and transient multigroup calculations of power distribution taking into account the thermal-hydraulic feedback effects. All this can be done without any limitation on any parameter (energy groups, meshes...). The code solves either the diffusion equation or the even parity transport equation with isotropic scattering and sources. Different geometries are available such as 1, 2 or 3 dimensions cartesian geometries, 2 or 3D hexagonal geometries and cylindrical geometries. The numerical method is based on the finite difference or finite element methods. CRONOS 2 has been written with the constant will of optimizing its portability. Presently, it is running on very different computers such as IBM 3090, CRAY 1, CRAY 2, SUN 4, MIPS RS2030 or IBM RS6000. A special data structure is used in order to improve vectorization. CRONOS is based on a modular structure that allows a great flexibility of use. It is implemented in the SAPHYR system which includes assembly calculation code (APOLLO), and thermal-hydraulic core calculation code (FLICA IV). A special object oriented language, named GIBIANE, and a common tool library have been developed to chain the various computation modules of those codes. (author). 11 refs, 1 fig., 5 tabs
Neutron reflectivity measurement of polymer monolayer and brush at the air/water interface
International Nuclear Information System (INIS)
We have been studied on amphiphilic polymer monolayer structure at the air/water interface by X-ray and neutron reflectometry. By complemently use of X-ray and neutron reflectometry, we have found (1) the existence of carpet layer in ionic polymer brush in monolayer system and (2) characteristic structural change in polymer/subphase interface. Furthermore, interesting experiment on small ion distribution was carried out by NR with contrast variation method. With our experimental examples, characteristic points in the neutron reflectivity measurement at the air/water interface and further possibility in this research area are discussed. (author)
Comparisons of Measured and Calculated Neutron Fluxes in Laminated iron and Heavy Water
International Nuclear Information System (INIS)
Measurements of neutron fluxes have been performed in configurations depicting the regions extending radially and axially outwards from the core of a PHWR reactor in order to test the accuracy of the available methods in shield design on thin alternating laminae of Fe and D2O. A 'dry' experimental set-up was constructed, i.e. the D2O was contained in flat tanks made of Al. The first set of measurements was performed through solid Fe and D2O layers, and only the results of these experiments are described in this report. The set-up allowed measurements also in a mock-up of a reactor top penetrated by D2O or air-filled channels (to be reported later). The results are compared to fluxes calculated by the British 18-group removal-diffusion method and by the NRN method developed at AE. The results show that the values predicted may be expected to be within a factor of 2 from the true values in most cases. The predicted relative flux distributions follow the observed ones with a very good accuracy in spite of the apparent misuse of diffusion theory for the thin regions in question. Finally, it is shown that the predicted change in the fast spectrum while penetrating these set-ups should be confirmable with certain threshold detectors
International Nuclear Information System (INIS)
This task involved the calculation of neutron and proton radii of cesium isotopes. The author has written a computer code that calculates radii according to two models: Myers 1983 and FRDM 1992. Results of calculations in both these models for both cesium and francium isotopes are attached as figures. He is currently interpreting these results in collaboration with D. Vieira and J.R. Nix, and they expect to use the computer code for further studies of nuclear radii
Benchmark calculations of neutron dose rates at transport and storage casks
International Nuclear Information System (INIS)
The application of numerical calculations methods for demonstration of sufficient radiation shielding of radioactive waste transport and storage casks requires a validation based on appropriate measurements of gamma and neutron sources. The results of the comparison of measured data and calculations using the Monte Carlo program MCNP show deviations dependent on the loading of the cask within the standard deviation which is dominated by the measuring method. Considering the neutrons scattered at the salt MCNP (in case of disposal in the salt) tends to underestimate the nominal values, but still within the double standard deviation. This accuracy is not reached with MAVRIC. Based on AHE (active handling experiments) data benchmark calculations were performed that can be used as reference value. The total accuracy results from the accuracy of the source term and the measurement of the neutron dose rate with a deviation of 15%.
Program POD. A computer code to calculate cross sections for neutron-induced nuclear reactions
International Nuclear Information System (INIS)
A computer code, POD, was developed for neutron-induced nuclear data evaluations. This program is based on four theoretical models, (1) the optical model to calculate shape-elastic scattering and reaction cross sections, (2) the distorted wave Born approximation to calculate neutron inelastic scattering cross sections, (3) the preequilibrium model, and (4) the multi-step statistical model. With this program, cross sections can be calculated for reactions (n, γ), (n, n'), (n, p), (n, α), (n, d), (n, t), (n, 3He), (n, 2n), (n, np), (n, nα), (n, nd), and (n, 3n) in the neutron energy range above the resonance region to 20 MeV. The computational methods and input parameters are explained in this report, with sample inputs and outputs. (author)
Calculation of neutron die-away times in a large-vehicle portal monitor
International Nuclear Information System (INIS)
Monte Carlo methods have been used to calculate neutron die-away times in a large-vehicle portal monitor. These calculations were performed to investigate the adequacy of using neutron die-away time measurements to detect the clandestine movement of shielded nuclear materials. The geometry consisted of a large tunnel lined with He3 proportional counters. The time behavior of the (n,p) capture reaction in these counters was calculated when the tunnel contained a number of different tractor-trailer load configurations. Neutron die-away times obtained from weighted least squares fits to these data were compared. The change in neutron die-away time due to the replacement of cargo in a fully loaded truck with a spherical shell containing 240 kg of borated polyethylene was calculated to be less than 3%. This result together with the overall behavior of neutron die-away time versus mass inside the tunnel strongly suggested that measurements of this type will not provide a reliable means of detecting shielded nuclear materials in a large vehicle. 5 figures, 4 tables
International Nuclear Information System (INIS)
Neutron elastic scattering cross section measurements have been going on for a long period at the Studsvik Van de Graaff laboratory. The cross sections of a range of elements have been investigated in the energy interval 1.5 to 8 MeV. The experimental data have been compared with cross sections calculated with the optical model when using a local nuclear potential
Calculation of the Inelastic Scattering of Neutrons from Polyethylene and Water
International Nuclear Information System (INIS)
A model for the calculation of the scattering of thermal neutrons from chemical system was proposed by Nelkin. This model considered the actual dynamics of the scattering system as composed of a set of oscillatory motions, each describable by a Hamiltonian which commuted with each of the others. It was then possible to express the differential scattering cross-section in closed form. This model has been used to calculate the scattering of neutrons by water. Some care must be taken in performing the numerical integration over angle and energy. The scattering model has been extended to the calculation of neutron scattering from polyethylene CnH2n. Analogous levels of polyethylene can be noted at 0.089 eV, 0.182 eV, 0.354 eV, and 0.533 eV. The differential and total cross-sections have been calculated for the scattering and the latter has been seen to be in reasonable agreement with experiment at room temperature. Scattering kernels have been calculated for a number of temperatures and where possible the results have been compared with experiment. In addition, neutron flux spectra and diffusion lengths have been calculated using the equations of reactor physics. Comparison of these Results with experimental data indicates that such integral measurements are indicative of at least the gross features of the scattering system and should be analysed in conduction with the detailed differential cross-section results. (author)
Subregions approach to boundary element neutron diffusion calculations
International Nuclear Information System (INIS)
Full text: The boundary element method (BEM) is a relatively new numerical method for the numerical solution of partial differential equations (PDE). BEM is based on the idea of converting the governing PDE with constant coefficients for a homogeneous region to a boundary integral equation (BIE) which contains unknowns only on the boundary of that region. A boundary element mesh is introduced over the boundary of the homogeneous region and the solution function and its normal derivative is assumed to have a polynomial dependence (constant, linear, quadratic...) over each boundary element. When the BIE is required to be satisfied at each node of the boundary element mesh, a linear system of dimension equal to the number of nodes on the boundary element mesh is obtained; but the number of unknowns is twice the number of equations since the nodal value of both the solution function and its normal derivative appear as unknowns. If the system consists of just one homogeneous region, half of the unknowns are eliminated by boundary conditions and the number of unknowns becomes equal to the number of equations and the linear system can be uniquely solved. When the system consists of more than one homogeneous region, the equations belonging to each region are assembled and the number of unknowns and equations are made equal by application of the continuity of the solution function and its normal derivative. In this work, we investigated a novel approach: a system consisting of one homogeneous region is divided into subregions and each subregion is treated as if it were a separate homogeneous region. This approach naturally increases the dimension of the resulting linear system, but its effect on the accuracy of the solution is a question that requires investigation. We used this subregions approach in the constant BEM solution of the 2-D neutron diffusion equation and investigated its effect on accuracy in terms of the multiplication eigenvalue and flux distribution by
Calculation of the angular distribution of delay times in neutron scattering on 58Ni nuclei
International Nuclear Information System (INIS)
Angular distributions of average delay times and time variances are calculated for resonance-neutron scattering on 58Ni nuclei at neutron energies in the range E = 600−700 keV. The effect of the energy spectrum and polarization of the beam on the scattering-process time is discussed. The angular dependence of the time law is also considered for the decay of an intermediate compound nuclear system. It is shown that the results of stationary and nonstationary calculations are in good agreement.
Calculation of the angular distribution of delay times in neutron scattering on {sup 58}Ni nuclei
Energy Technology Data Exchange (ETDEWEB)
Prokopets, G. A., E-mail: gaprok@uos.net.ua [National University of Kyiv-Mohyla Academy (Ukraine)
2011-05-15
Angular distributions of average delay times and time variances are calculated for resonance-neutron scattering on {sup 58}Ni nuclei at neutron energies in the range E = 600-700 keV. The effect of the energy spectrum and polarization of the beam on the scattering-process time is discussed. The angular dependence of the time law is also considered for the decay of an intermediate compound nuclear system. It is shown that the results of stationary and nonstationary calculations are in good agreement.
Shell model calculations for neutron rich nuclei with A=35-41
International Nuclear Information System (INIS)
Shell model calculations are presented for neutron-rich nuclei in the mass region A=35-41 using two new interactions. The usefulness and reliability of the interactions are evaluated with particular emphasis on their predictions for netron-rich isotopes. The calculations are performed in a 0(h/2π)ω basis space with active protons in the 1d5/2, 2s1/2 and 1d3/2 orbitals and active neutrons in the 1f7/2 and 2p3/2 orbitals
Monte Carlo simulation in the reaction rate's calculation with neutron-activation method
International Nuclear Information System (INIS)
With MCNP/4B code, the influence of cut-off energy, flux tallies, nuclear databases and perturbation on the reaction rate's calculation with neutron-activation method are analysed. When the effective reaction threshold is chosen as the cut-off energy, calculation time is considerably reduced and yet the results are not changed. Comparing calculations with cell tallies (F4) with those performed with detector tallies (F5), the counting efficiency of cell tallies is higher and the results are slightly higher, but still credible. With different nuclear databases, calculated results can be different. The perturbation among the detectors doesn't effect on the calculated results. (authors)
International Nuclear Information System (INIS)
The main objective of this work is to create a neutronic calculations system for the SILOE-SILOETTE reactors, adaptable to other types of plate reactors. The author presents the methodology and the development of the APOLLO 1D (99 gr.) calculations for the creation of cross sections libraries. After a recall of the Discrete Ordinate Method (DOT), the method accuracy is studied in order to optimize the spatial discretization of the calculations; calculations of DOT 3.5 and of SILOETTE core are conducted and their convergence and costs are examined. DOT calculations of SILOETTE and experimental tests results are then compared
Calculation of air supply rates for nonunidirectional airflow cleanrooms
Whyte, W; Whyte, W.M.; Eaton, T; Lenegan, N.
2014-01-01
This article describes a method for estimating the air supply rate required in non-unidirectional airflow cleanrooms to obtain a required concentration of airborne particles and microbe-carrying particles. The variables considered are: surface deposition, emission rates of airborne contamination from personnel and machinery, filter removal efficiency, effectiveness of cleanroom garments, effectiveness of air supply distribution, and the contribution of filtered air from clean air ...
International Nuclear Information System (INIS)
The features and the algorithm of the program to calculate adjoint neutron cross sections on the basis of the continuous energy neutron cross sections as well as energy and angular distributions are described. The calculated adjoint cross sections are intended for Monte Carlo investigation of the nonuniform adjoint Boltzmann equation. 16 refs
Relativistic collision rate calculations for electron-air interactions
International Nuclear Information System (INIS)
The most recent data available on differential cross sections for electron-air interactions are used to calculate the avalanche, momentum transfer, and energy loss rates that enter into the fluid equations. Data for the important elastic, inelastic, and ionizing processes are generally available out to electron energies of 1--10 kev. Prescriptions for extending these cross sections to the relativistic regime are presented. The angular dependence of the cross sections is included where data is available as is the doubly differential cross section for ionizing collisions. The collision rates are computed by taking moments of the Boltzmann collision integrals with the assumption that the electron momentum distribution function is given by the Juettner distribution function which satisfies the relativistic H- theorem and which reduces to the familiar Maxwellian velocity distribution in the nonrelativistic regime. The distribution function is parameterized in terms of the electron density, mean momentum, and thermal energy and the rates are therefore computed on a two-dimensional grid as a function of mean kinetic energy and thermal energy
International Nuclear Information System (INIS)
Thermal-neutron research reactors are currently the most common source of neutron beams for both research and clinical trials of neutron capture therapy (NCT). Neutron spectra suitable for NCT are typically produced either by beam filtering or spectrum shifting techniques. However, fast-neutron reactors are also being considered for NCT application as it is recognized that they may allow for improved beam quality. TAPIRO is a low power, high flux, highly enriched (93.5% 235U) fast reactor. The power is 5 kW and the maximum neutron flux in the core is 3x1012 cm-2.s-1. Both a thermal and an epithermal column have been designed and constructed, aimed at dosimetry and animal experiments. The configurations of the columns have been designed by means of Monte Carlo calculations. The columns have been characterized by means of measurements performed with activation techniques and thermoluminescence and gel dosimeters. Experimental results have shown good consistency with calculations. Moreover, they have confirmed the good quality of the beams obtainable with such a reactor. An epithermal column for clinical trials of patients with brain gliomas has been designed and is under construction. The treatment planning figures-of-merit in an anthropomorphic phantom look very satisfactory. (author)
Overview of the neutronics calculation system for the HANARO
International Nuclear Information System (INIS)
KAERI established the HANAFMS (HANARO Nuclear Analyses and Fuel Management System) for the in-core fuel management. The major components of the HANAFMS are the WIMS-KAERI and VENTURE. And several auxiliary codes such as REGAV-K, WIMPAK, MAPHEX, HEXSHUF, are supporting the system. The HANARO have carried out various kinds of reactor physics tests and experiments for 11 years. To support those experimental activities and to ensure the safe operation of the HANARO, the core follow up calculation is always performed along with the cycle operation using HANAFMS. (author)
Microscopic Calculations of Vortex-Nucleus Interaction in the Neutron Star Crust
Sekizawa, Kazuyuki; Magierski, Piotr; Bulgac, Aurel; Forbes, Michael McNeil
2016-01-01
We investigate the dynamics of a quantized vortex and a nuclear impurity immersed in a neutron superfluid within a fully microscopic time-dependent three-dimensional approach. The magnitude and even the sign of the force between the quantized vortex and the nuclear impurity have been a matter of debate for over four decades. We determine that the vortex and the impurity repel at neutron densities, 0.014 fm$^{-3}$ and 0.031 fm$^{-3}$, which are relevant to the neutron star crust and the origin of glitches, while previous calculations have concluded that the force changes its sign between these two densities and predicted contradictory signs. The magnitude of the force increases with the density of neutron superfluid, while the magnitude of the pairing gap decreases in this density range.
The neutron 'thunder' accompanying large extensive air showers
Erlykin, A. D.
2006-01-01
The bulk of neutrons which appear with long delays in neutron monitors nearby the EAS core (~'neutron thunder'~) are produced by high energy EAS hadrons hitting the monitors. This conclusion raises an important problem of the interaction of EAS with the ground, the stuff of the detectors and their environment. Such interaction can give an additional contribution to the signal in the EAS detectors at {\\em km}-long distances from the large EAS core after a few $\\mu s$ behind the EAS front.
Capote, R.; Carjan, N.; Chiba, S.
2016-02-01
The multiplicities of scission neutrons νs c are calculated for series of U, Pu, Cm, and Cf isotopes assuming a sudden transition between two different nuclear configurations (αi→αf ): one just before the neck rupture and one immediately after the disappearance of the neck. This calculation requires only the knowledge of the corresponding two sets of neutron eigenstates. The nuclear shapes around the scission point are described in terms of Cassinian ovals with only two parameters: α (that positions the shape with respect to the zero-neck shape) and α1 (that defines the mass asymmetry). Based on these shapes, a neutron mean field of the Woods-Saxon type is constructed using two prescriptions to calculate the distance to the nuclear surface. The accent in the present work is put on the dependence of νs c on the neutron number Nf of the fissioning nucleus and on the mass asymmetry AL/AH of the primary fission fragments. The relative dependence of these multiplicities, averaged over the mass yields, , are finally compared with existing experimental data on prompt fission neutrons .
International Nuclear Information System (INIS)
Highlights: • Performance estimation of nuclear-data benchmark was investigated. • Point detector contribution played a benchmark role not only to the neutron producing the detector contribution but also equally to all the upstream transport neutrons. • New functions were defined to give how well the contribution could be interpreted for benchmarking. • Benchmark performance could be evaluated only by a forward Monte Carlo calculation. -- Abstract: The author's group has been investigating how the performance estimation of nuclear-data benchmark using experiment and its analysis by Monte Carlo code should be carried out especially at 14 MeV. We have recently found that a detector contribution played a benchmark role not only to the neutron producing the detector contribution but also equally to all the upstream neutrons during the neutron history. This result would propose that the benchmark performance could be evaluated only by a forward Monte Carlo calculation. In this study, we thus defined new functions to give how well the contribution could be utilized for benchmarking using the point detector, and described that it was deeply related to the newly introduced “partial adjoint contribution”. By preparing these functions before benchmark experiments, one could know beforehand how well and for which nuclear data the experiment results could do benchmarking in forward Monte Carlo calculations
Institute of Scientific and Technical Information of China (English)
Shuqing HAO; Hongwei HUANG; Kun YIN
2007-01-01
By simplifying the characters in the air reverse circulation bit interior fluid field, the authors used air dynamics and fluid mechanics to calculate the air distribution in the bit and obtained an equation of flow distribution with a unique resolution. This study will provide help for making certain the bit parameters of the bit structure effectively and study the air reverse circulation bit interior fluid field character deeply.
Neutron and gamma ray calculation for Hiroshima-type atomic bomb
Energy Technology Data Exchange (ETDEWEB)
Hoshi, Masaharu; Endo, Satoru; Takada, Jun [Hiroshima Univ. (Japan). Research Inst. for Radiation Biology and Medicine; Iwatani, Kazuo; Oka, Takamitsu; Shizuma, Kiyoshi; Fujita, Shoichiro; Hasai, Hiromi
1998-03-01
We looked at the radiation dose of Hiroshima and Nagasaki atomic bomb again in 1986. We gave it the name of ``Dosimetry System 1986`` (DS86). We and other groups have measured the expose dose since 1986. Now, the difference between data of {sup 152}Eu and the calculation result on the basis of DS86 was found. To investigate the reason, we carried out the calculations of neutron transport and neutron absorption gamma ray for Hiroshima atomic bomb by MCNP3A and MCNP4A code. The problems caused by fast neutron {sup 32}P from sulfur in insulator of pole. To correct the difference, we investigated many models and found agreement of all data within 1 km. (S.Y.)
Talys calculations for evaluation of neutron-induced single-event upset cross sections
International Nuclear Information System (INIS)
The computer code TALYS has been used to calculate interactions between cosmic-ray neutrons and silicon nuclei with the goal to describe single-event upset (SEU) cross sections in microelectronics devices. Calculations for the Si(n,X) reaction extend over an energy range of 2 to 200 MeV. The obtained energy spectra of the resulting residuals and light-ions have been integrated using several different critical charges as SEU threshold. It is found that the SEU cross section seems largely to be dominated by 28Si recoils from elastic scattering. Furthermore, the shape of the SEU cross section as a function of the energy of the incoming neutron changes drastically with decreasing critical charge. The results presented in this report stress the importance of performing studies at mono-energetic neutron beams to advance the understanding of the underlying mechanisms causing SEUs
Calculation of the reactor neutron time of flight spectrum by convolution technique
International Nuclear Information System (INIS)
It is a very complex and time-consuming process to simulate the nuclear reactor neutron spectrum from the reactor core to the export channel by applying a Monte Carlo program. This paper presents a new method to calculate the neutron spectrum by using the convolution technique which considers the channel transportation as a linear system and the transportation scattering as the response function. It also applies Monte Carlo Neutron and Photon Transport Code (MCNP) to simulate the response function numerically. With the application of convolution technique to calculate the spectrum distribution from the core to the channel, the process is then much more convenient only with the simple numerical integral numeration. This saves computer time and reduces some trouble in re-writing of the MCNP program. (general)
Calculation of prompt fission neutron spectra for 235U(n,f)
Institute of Scientific and Technical Information of China (English)
CHEN Yong-Jing; JIA Min; TAO Xi; QIAN Jing; LIU Ting-Jin; SHU Neng-Chuan
2012-01-01
The prompt fission neutron spectra for the neutron-induced fission of 235U at En ＜ 5 MeV are calculated using nuclear evaporation theory with a semi-empirical model,in which the nonconstant and constant temperatures related to the Fermi gas model are taken into account. The calculated prompt fission neutron spectra reproduce the experimental data well.For the n(thermal)+235U reaction,the average nuclear temperature of the fission fragment,and the probability distribution of the nuclear temperature,are discussed and compared with the Los Alamos model.The energy carried away by γ rays emitted from each fragment is also obtained and the results are in good agreement with the existing experimental data.
Calculation of the reactor neutron time of flight spectrum by convolution technique
Institute of Scientific and Technical Information of China (English)
Cheng Jin-Xing; Ouyang Xiao-Ping; Zheng Yi; Zhang An-Hui; Ouyang Mao-Jie
2008-01-01
It is a very complex and tlme-consuming process to simulate the nuclear reactor neutron spectrum from the reactor core to the export channel by applying a Monte Carlo program. This paper presents a new method to calculate the neutron spectrum by using the convolution technique which considers the channel transportation as a linear system and the transportation scattering as the response function. It also applies Monte Carlo Neutron and Photon Transport Code (MCNP) to simulate the response function numerically. With the application of convolution technique to calculate thespectrum distribution from the core to the channel, the process is then much more convenient only with the simple numerical integral numeration. This saves computer time and reduces some trouble in re-writing of the MCNP program.
Neutron and gamma ray calculation for Hiroshima-type atomic bomb
International Nuclear Information System (INIS)
We looked at the radiation dose of Hiroshima and Nagasaki atomic bomb again in 1986. We gave it the name of ''Dosimetry System 1986'' (DS86). We and other groups have measured the expose dose since 1986. Now, the difference between data of 152Eu and the calculation result on the basis of DS86 was found. To investigate the reason, we carried out the calculations of neutron transport and neutron absorption gamma ray for Hiroshima atomic bomb by MCNP3A and MCNP4A code. The problems caused by fast neutron 32P from sulfur in insulator of pole. To correct the difference, we investigated many models and found agreement of all data within 1 km. (S.Y.)
Calculation of effective delayed neutron fraction with modified library of Monte Carlo code
International Nuclear Information System (INIS)
Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data
International Nuclear Information System (INIS)
Irradiation Experimental Area of TechnoFusion will emulate the extreme irradiation fusion conditions in materials by means of three ion accelerators: one used for self-implanting heavy ions (Fe, Si, C,...) to emulate the displacement damage induced by fusion neutrons and the other two for light ions (H and He) to emulate the transmutation induced by fusion neutrons. This Laboratory will play an essential role in the selection of functional materials for DEMO reactor since it will allow reproducing the effects of neutron radiation on fusion materials. Ion irradiation produces little or no residual radioactivity, allowing handling of samples without the need for special precautions. Currently, two different methods are used to calculate the primary displacement damage by neutron irradiation or by ion irradiation. On one hand, the displacement damage doses induced by neutrons are calculated considering the NRT model based on the electronic screening theory of Linhard. This methodology is commonly used since 1975. On the other hand, for experimental research community the SRIM code is commonly used to calculate the primary displacement damage dose induced by ion irradiation. Therefore, both methodologies of primary displacement damage calculation have nothing in common. However, if we want to design ion irradiation experiments capable to emulate the neutron fusion effect in materials, it is necessary to develop comparable methodologies of damage calculation for both kinds of radiation. It would allow us to define better the ion irradiation parameters (Ion, current, Ion energy, dose, etc) required to emulate a specific neutron irradiation environment. Therefore, our main objective was to find the way to calculate the primary displacement damage induced by neutron irradiation and by ion irradiation starting from the same point, that is, the PKA spectrum. In order to emulate the neutron irradiation that would prevail under fusion conditions, two approaches are contemplated: a) on
International Nuclear Information System (INIS)
The neutron generation time Λ plays an important role in the reactor kinetics. However, it is not straightforward nor standard in most continuous energy Monte Carlo codes which are able to calculate the prompt neutron lifetime lp directly. The difference between Λ and lp are sometimes very apparent. As very few delayed neutrons are produced in the reactor, they have little influence on Λ. Thus on the assumption that no delayed neutrons are produced in the system, the prompt kinetics equations for critical system and subcritical system with an external source are proposed. And then the equations are applied to calculating Λ with pulsed neutron technique using Monte Carlo. Only one fission neutron source is simulated with Monte Carlo in critical system while two neutron sources, including a fission source and an external source, are simulated for subcritical system. Calculations are performed on both critical benchmarks and subcritical system with an external source and the results are consistent with the reference values. (author)
Model-Independent Calculation of Radiative Neutron Capture on Lithium-7
Rupak, Gautam; Higa, Renato
2011-01-01
The radiative neutron capture on lithium-7 is calculated model independently using a low-energy halo effective field theory. The cross section is expressed in terms of scattering parameters directly related to the S-matrix elements. It depends on the poorly known p-wave effective range parameter r(1
The background cross section method for calculating the epithermal neutron spectra
International Nuclear Information System (INIS)
We have developed a new methodology to the multigroup constants calculations, for thermal and fast reactors. The method to obtain the constants is extremely fast and simple, and it avoid repeated computations of the detailed neutron spectrum for different cell configurations (composition, geometry and temperature). (author)
Energy Technology Data Exchange (ETDEWEB)
Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)
1998-03-01
Application programs for personal computers are developed to calculate the decay heat power and delayed neutron activity from fission products. The main programs can be used in any computers from personal computers to main frames because their sources are written in Fortran. These programs have user friendly interfaces to be used easily not only for research activities but also for educational purposes. (author)
International Nuclear Information System (INIS)
This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (Mnr) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperature coefficient of reactivity at different temperatures and it's average value in a range of temperature directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment. (author)
International Nuclear Information System (INIS)
Characteristics of neutrons generated from the lithium target bombarded with high energetic deuterons of 10-40 MeV have been calculated to determine the specification for the neutron irradiation material test facility (ESNIT) planned at Japan Atomic Energy Research Institute. The simple nuclear reaction model was applied to estimation of neutron flux distribution and energy spectrum and the results showed an agreement with the reported experiment within a factor of 2. The present calculation gives the basic spectrum data for estimation of damage parameters in test samples to evaluate the high energy neutron effect on them. (author)
Calculation of neutron flux in PUSPATI TRIGA MARK II reactor using Monte-Carlo n-particle approach
International Nuclear Information System (INIS)
A Monte Carlo simulation of neutron flux at the TRIGA MARK II PUSPATI (RTP) nuclear research reactor at Agensi Nuklear Malaysia was carried out using the MCNP5 program. The objective of the work is to simulate the neutron flux inside the reactor core. Calculations of neutron flux for fast and thermal neutron were carried out under the conditions in which the control rod was either fully withdrawn from or fully inserted into the reactor. (Author)
Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel
2014-01-01
High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately...
Calculation of the Chilling Requirement for Air Conditioning in the Excavation Roadway
Directory of Open Access Journals (Sweden)
Yueping Qin
2015-10-01
Full Text Available To effectively improve the climate conditions of the excavation roadway in coal mine, the calculation of the chilling requirement taking air conditioning measures is extremely necessary. The temperature field of the surrounding rock with moving boundary in the excavation roadway was numerically simulated by using finite volume method. The unstable heat transfer coefficient between the surrounding rock and air flow was obtained via the previous calculation. According to the coupling effects of the air flow inside and outside air duct, the differential calculation mathematical model of air flow temperature in the excavation roadway was established. The chilling requirement was calculated with the selfdeveloped computer program for forecasting the required cooling capacity of the excavation roadway. A good air conditioning effect had been observed after applying the calculated results to field trial, which indicated that the prediction method and calculation procedure were reliable.
Measured and calculated fission-product poisoning in neutron-irradiated uranium-233
International Nuclear Information System (INIS)
Samples of 233U and of natural thorium have been irradiated in high neutron-flux facilities, in both soft and hard neutron spectra, and for both short and long exposure times. Included are exposures resulting in depletions of more than 90 percent of the 233U in the fissile material and burnups of more than 30,000 MWd/MT in the fertile material. Fission-product poison cross sections in two energy groups (thermal and epithermal) exhibit differences between measurement and calculation that are believed to be attributable to a lack of adequate information on important fission products in the literature. Experimental results for transient absorbers in irradiated 233U give at least 20,000 b for the neutron absorption resonance integral of 149Pm. This is a factor of 15 higher than that obtained by a 1/v extrapolation of the thermal cross sections. For transient 135Xe, the measured absorption is 7.5 percent higher than that calculated using ENDF/B-IV data. Information is also provided concerning such matters as fission yields and neutron absorption of neodymium isotopes, the existence of significant transient fission-product poisons other than 135Xe and 149Sm, and the shielding of 233U by 232Th. Such shielding suggests the need for a change in the energy dependence of the 232Th thermal-neutron cross section
Calculations to support JET neutron yield calibration: Modelling of the JET remote handling system
Energy Technology Data Exchange (ETDEWEB)
Snoj, Luka, E-mail: luka.snoj@ijs.si [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-MHEST Association, Reactor Physics Division, Jožef Stefan Institute, Jamova Cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-MHEST Association, Reactor Physics Division, Jožef Stefan Institute, Jamova Cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, OXON (United Kingdom); Conroy, Sean [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-VR Association, Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Meredith, Lewis [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, OXON (United Kingdom)
2013-08-15
Highlights: ► We model JET remote handling system in MCNP. ► We examine the effect of JET remote handling system on neutron monitor response. ► The integral effect of JET RH system on neutron monitors is less than 5%. -- Abstract: After the coated CFC wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) transition in 2010–2011, confirmation of the neutron yield calibration will be ensured by direct measurements using a calibrated {sup 252}Cf neutron source deployed by the in-vessel remote handling boom and Mascot manipulator inside the JET vacuum vessel. Neutronic calculations are required to calculate the effects of the JET remote handling (RH) system on the neutron monitors. We developed a simplified geometrical computational model of the JET remote handling system in MCNP. In parallel we developed a script that translates the RH movement data to transformations of individual geometrical parts of the RH model in MCNP. After that a benchmarking of the model was performed to verify and validate the accordance of the target positions of source and RH system with the ones from our model. In the last phase we placed the JET RH system in the simplified MCNP model of the JET tokamak and studied its effect on neutron monitor response for some example source positions and boom configurations. As the correction factors due to presence of the JET RH system can potentially be significant in cases when the boom is blocking a port close to the detector under investigation, we have chosen boom configurations so that this is avoided in the vast majority of the source locations. Examples are given.
Energy Technology Data Exchange (ETDEWEB)
Smith, L.A.; Gallmeier, F.X. [Oak Ridge Institute for Science and Energy, TN (United States); Gehin, J.C. [Oak Ridge National Lab., TN (United States)] [and others
1995-05-01
The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are {approx} 13%, while the average differences are < 8%.
International Nuclear Information System (INIS)
Sensitivity and uncertainty calculations methods of neutronics parameters in pressurized light water reactors have been developed. The sensitivity is composed of three terms; the first is the sensitivity of cell-averaged multi-group cross-sections relative to multi-group infinite dilution cross-sections, the second is the sensitivity of assembly averaged few-group macroscopic cross-sections relative to cell-averaged multi-group cross-sections, and the third is the sensitivity of neutronics parameters in PWR cores relative to few-group macroscopic cross-sections. Combining the three sensitivities, the sensitivity of neutronics parameters in PWR cores relative to multi-group infinite dilution cross-sections is obtained. The discussion of this method will be presented in two papers; the present paper is part I, where the theory and some numerical results for typical pin cells, fuel assemblies and a simple PWR core are shown. The present method gives us multi-group sensitivities for individual nuclides in each reaction type, and wide ranges of applications are possible to the fields such as cross-section adjustment and uncertainty reduction. (author)
Single event upsets calculated from new ENDF/B-VI proton and neutron data up to 150 MeV
Energy Technology Data Exchange (ETDEWEB)
Chadwick, M.B. [Los Alamos National Lab., NM (United States). Theoretical Div.; Normand, E. [Boeing Military Aircraft and Missile Systems, Seattle, WA (United States)
1999-06-01
Single-event upsets (SEU) in microelectronics are calculated from newly-developed silicon nuclear reaction recoil data that extend up to 150 MeV, for incident protons and neutrons. Calculated SEU cross sections are compared with measured data.
Energy Technology Data Exchange (ETDEWEB)
Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R., E-mail: rosanemribeiro@oi.com.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2014-07-01
Fluence to dose equivalent conversion coefficients provide the basis for the calculation of area and personal monitors. Recently, the ICRP has started a revision of these coefficients, including new Monte Carlo codes for benchmarking. So far, little information is available about neutron transport below 10 MeV in tissue-equivalent (TE) material performed with Monte Carlo GEANT4 code. The objective of this work is to calculate neutron fluence to personal dose equivalent conversion coefficients, H{sub p} (10)/Φ, with GEANT4 code. The incidence of monoenergetic neutrons was simulated as an expanded and aligned field, with energies ranging between thermal neutrons to 10 MeV on the ICRU slab of dimension 30 x 30 x 15 cm{sup 3}, composed of 76.2% of oxygen, 10.1% of hydrogen, 11.1% of carbon and 2.6% of nitrogen. For all incident energy, a cylindrical sensitive volume is placed at a depth of 10 mm, in the largest surface of the slab (30 x 30 cm{sup 2}). Physic process are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles which reach the sensitive volume. Results obtained are thus compared with values published in ICRP 74. Neutron fluence in the sensitive volume was calculated for benchmarking. The Monte Carlo GEANT4 code was found to be appropriate to calculate neutron doses at energies below 10 MeV correctly. (author)
International Nuclear Information System (INIS)
To check the data of carbon material reflecting neutrons, the distribution of 238U fission reaction rates induced by D-T fusion neutrons reflected by carbon material was measured by using the small depleted uranium fission chamber and the capturing detector. For comparison, 238U fission rates without carbon material was measured too. The combined standard uncertainty of 238U fission reaction rate is 5.1%-6.4%. The measured results are consistent with the calculated ones with MCNP/4A code and ENDF/B-IV library data in the range of the error
Neutron dosimetry and damage calculations for the HFIR-MFE-200J-1 irradiation
International Nuclear Information System (INIS)
Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment MFE-200-J-, which was conducted in the removable beryllium (RB) position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum neutron fluence at midplane was 4.1 x 1022 n/cm2 (1.9 x 1022 n/cm2 above 0.1 MeV), resulting in about 12 dpa and 28 appm helium in type 316 stainless steel
Results of coupled channels calculations for the neutrons cross sections of a set of actinide nuclei
International Nuclear Information System (INIS)
This report gathers recents results of neutrons interactions with the following actinide nuclei: 230Th, 232Th, 234U, 238U, 242Pu, 246Cm and 252Cf from the use of the coupled channels optical model. Tabulations of the following quantities are given in Annexe: total, direct elastic and inelastic scattering (integrated and differential), and compound nucleus formation cross sections; ground state generalized transmission coefficients needed to calculate the cross sections of partial compound nucleus processes. This work was carried out within the framework of the IAEA-NDS Coordinated Research Programme on the Intercomparison of Actinide Neutron Cross Section Evaluations
Neutron dosimetry and damage calculation for the JP-10, 11, 13, and 16 experiments in HFIR
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R.; Ratner, R.T.
1996-04-01
Neutron fluence measurements and radiation damage calculations are reported for the joint U.S./Japanese experiments JP-10, 11, 13, and 16 in the target of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL). These experiments were irradiated at 85 MW for 238.5 EFPD. The maximum fast neutron fluence >0.1 MeV was about 2.1E + 22 n/cm{sup 2} for all of the experiments resulting in about 17.3 dpa in 316 stainless steel.
Neutron dosimetry and damage calculations for the JP-17, 18 and 19 experiments in HFIR
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R.; Baldwin, C.A.
1996-04-01
Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiments JP-17, 18, and 19 in the target of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). These experiments were irradiated at 85 MW for two cycles resulting in 43.55 EFPD for JP-17 and 42.06 EFPD for JP-18 and 19. The maximum fast neutron fluence > 0.1 MeV was about 3.7E + 21 n/cm{sup 2} for all three irradiations, resulting in about 3 dpa in 316 stainless steel.
Neutron dosimetry and damage calculations for the HFIR-MFE-200J-1 irradiation
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Baldwin, C.A. [Oak Ridge National Lab., TN (United States)
1998-03-01
Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment MFE-200-J-, which was conducted in the removable beryllium (RB) position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum neutron fluence at midplane was 4.1 {times} 10{sup 22} n/cm{sup 2} (1.9 {times} 10{sup 22} n/cm{sup 2} above 0.1 MeV), resulting in about 12 dpa and 28 appm helium in type 316 stainless steel.
Neutron dosimetry and damage calculations for the HFIR-JP-9, -12, and -15 irradiations
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Baldwin, C.A. [Oak Ridge National Lab., TN (United States)
1998-03-01
Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiments JP-9, -12, and -15. These experiments were conducted in target positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) for a period of nearly four years. The maximum neutron fluence at midplane was 2.6 {times} 10{sup 23} n/cm{sup 2} (7.1 {times} 10{sup 22} n/cm{sup 2} above 0.1 MeV), resulting in about 60 dpa and 3900 appm helium in type 316 stainless steel.
Neutron dosimetry and damage calculations for the HFIR-JP-20 irradiation
Energy Technology Data Exchange (ETDEWEB)
Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Baldwin, C.A. [Oak Ridge National Lab., TN (United States)
1998-03-01
Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment JP-20, which was conducted in a target position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum total neutron fluence at midplane was 4.2 {times} 10{sup 22} n/cm{sup 2} (1.0 {times} 10{sup 22} n/cm{sup 2} above 0.1 MeV), resulting in about 8.4 dpa and 388 appm helium in type 316 stainless steel.
Dekker, C. M.; Sliggers, C. J.
To spur on quality assurance for models that calculate air pollution, quality criteria for such models have been formulated. By satisfying these criteria the developers of these models and producers of the software packages in this field can assure and account for the quality of their products. In this way critics and users of such (computer) models can gain a clear understanding of the quality of the model. Quality criteria have been formulated for the development of mathematical models, for their programming—including user-friendliness, and for the after-sales service, which is part of the distribution of such software packages. The criteria have been introduced into national and international frameworks to obtain standardization.
Total #betta#-ray spectra and isomeric ratio calculations in fast neutron radiative capture
International Nuclear Information System (INIS)
Estimates of total #betta#-ray spectra and isomeric ratios have been attempted in the framework of the optical and statistical models for neutron radiative capture. The role of optical model, of Brink-Axel and Weisskopf assumptions as well as of the most important parameters have been investigated. The results of these calculations satisfactorily agree with experimental information in all cases considered. Spectra calculations were used in relative neutron capture measurements for correction of systematic uncertainties due to non-linear efficiency of the Moxon-Rae detector. Data on the calculated total #betta#-ray spectra and the corresponding integrated cross sections are shown for the isotopes investigated. The impact of parameters involved is discussed
Lahti, G. P.; Mueller, R. A.
1973-01-01
Measurements of MeV neutron were made at the surface of a lithium hydride and depleted uranium shielded reactor. Four shield configurations were considered: these were assembled progressively with cylindrical shells of 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, and 3-centimeter-thick depleted uranium. Measurements were made with a NE-218 scintillation spectrometer; proton pulse height distributions were differentiated to obtain neutron spectra. Calculations were made using the two-dimensional discrete ordinates code DOT and ENDF/B (version 3) cross sections. Good agreement between measured and calculated spectral shape was observed. Absolute measured and calculated fluxes were within 50 percent of one another; observed discrepancies in absolute flux may be due to cross section errors.
MAMONT program for neutron field calculation by the Monte Carlo method
International Nuclear Information System (INIS)
The MAMONT program (MAthematical MOdelling of Neutron Trajectories) designed for three-dimensional calculation of neutron transport by analogue and nonanalogue Monte Carlo methods in the range of energies from 15 MeV to the thermal ones is described. The program is written in FORTRAN and is realized at the BESM-6 computer. Group constants of the library modulus are compiled of the ENDL-83, ENDF/B-4 and JENDL-2 files. The possibility of calculation for the layer spherical, cylindrical and rectangular configurations is envisaged. Accumulation and averaging of slowing-down kinetics functionals (averaged logarithmic losses of energy, time of slowing- down, free paths, the number of collisions, age), diffusion parameters, leakage spectra and fluxes as well as formation of separate isotopes over zones are realized in the process of calculation. 16 tabs
CABRI Reactor: The fast neutron Hodoscope Calibration curves calculation with MORET
Bernard, Franck; Chevalier, Vincent; Venanzi, Damiano
2014-06-01
This poster presents the Hodoscope calibration curves calculation with 3D Monte Carlo code MORET. The fast neutron hodoscope is a facility of the CABRI research reactor at Cadarache (FRANCE). This hodoscope is designed to measure the fuel motion during a RIA in a pressurized water reactor. The fuel motion is measured by counting fast fission neutrons emerging from the test fuel placed in an experimental loop functioning like a Pressurized Water Reactor (T=300°C and P=155 bar), at the center of the CABRI core. The detection system of the hodoscope measures a signal which is a function of the fuel motion. The calibration curves allow then to convert the signal in a fuel mass. In order to calculate these curves, we have developed a method based on a Monte Carlo calculation code.
Using of discrete ordinate method in the spallation target neutronics and shielding calculations
International Nuclear Information System (INIS)
A discrete ordinate algorithm for coupled charged/neutral particle transport calculations in 2D pencil beam problems is developed. It is based on the use of the second order of accuracy adaptive WDD (AWDD) scheme for approximation both the continuous slowing down (CSD) and streaming terms of the charged particle transport equation in z geometry, and a suitable algorithm for treatment of the extended uncollided flux from an initially monodirectional beam of charged particles with given radial distribution. The developed algorithm is implemented in the 2D transport code KASKAD-S-1.5 and is applied to the high-energy coupled proton-pion-neutron-photon transport calculations. The multigroup cross-section library SADCO-2 for nucleon-meson cascade calculations coupled with standard neutron and gamma-ray cross-section libraries below 20 MeV is used. Some numerical examples are given.(author)
Lattice EFT calculation of thermal properties of low-density neutron matter
International Nuclear Information System (INIS)
Thermal properties of low-density neutron matter are investigated by lattice calculation with nuclear effective field theory without pions up to the next-to-leading order. The 1S0 pairing gap is extracted near zero temperature at low densities. We find that the pairing gap is smaller than the BCS approximation with the conventional NN potentials, but not as small as those by various many-body calculations beyond BCS approximation. Our result is consistent with the recent Green's Function Monte Carlo calculation within the statistical errors. The critical temperature of the normal-to-superfluid phase transition and the pairing temperature scale are also extracted at low densities, and the phase diagram is given. We find that the physics of low-density neutron matter is clearly identified as being BCS-BEC crossover.
NXDC-neutron and x-ray diffraction code for crystal structures calculations
International Nuclear Information System (INIS)
A computer program NXDC for the calculations of neutron diffraction and x-ray diffraction intensities is reported. The program is very flexible and allows the intensity of a reflection with a given Miller indices to be calculated if the unit cell and its contents are specified together with the equipement used Neutrons or X-rays-and if necessary introducing temperature and absorption factors corrections. For the refinement of crystal structures provision is made for the comparison of the calculated intensities and the intergrated intensities observed from the diffraction diagrams using the least-squares analysis to obtain the reliability factor R. The program is written in FORTRAN Iv and is very suitable for minicomputers
International Nuclear Information System (INIS)
This report documents the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations''. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.
CDFMC: a program that calculates the fixed neutron source distribution for a BWR using Monte Carlo
International Nuclear Information System (INIS)
The three-dimensional neutron flux calculation using the synthesis method, it requires of the determination of the neutron flux in two two-dimensional configurations as well as in an unidimensional one. Most of the standard guides for the neutron flux calculation or fluences in the vessel of a nuclear reactor, make special emphasis in the appropriate calculation of the fixed neutron source that should be provided to the used transport code, with the purpose of finding sufficiently approximated flux values. The reactor core assemblies configuration is based on X Y geometry, however the considered problem is solved in R θ geometry for what is necessary to make an appropriate mapping to find the source term associated to the R θ intervals starting from a source distribution in rectangular coordinates. To develop the CDFMC computer program (Source Distribution calculation using Monte Carlo), it was necessary to develop a theory of independent mapping to those that have been in the literature. The method of meshes overlapping here used, is based on a technique of random points generation, commonly well-known as Monte Carlo technique. Although the 'randomness' of this technique it implies considering errors in the calculations, it is well known that when increasing the number of points randomly generated to measure an area or some other quantity of interest, the precision of the method increases. In the particular case of the CDFMC computer program, the developed technique reaches a good general behavior when it is used a considerably high number of points (bigger or equal to a hundred thousand), with what makes sure errors in the calculations of the order of 1%. (Author)
International Nuclear Information System (INIS)
Neutron and gamma-ray spectra resulting from the interactions of approx. 14-MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree within 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra is also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE
International Nuclear Information System (INIS)
Neutron and gamma-ray energy spectra resulting from the interactions of approx. 14 MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree witin 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra are also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE
Radiation transport in earth for neutron and gamma ray point sources above an air-ground interface
International Nuclear Information System (INIS)
Two-dimensional discrete ordinates methods were used to calculate the instantaneous dose rate in silicon and neutron and gamma ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 meters) above an air--ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth--concrete model containing 0.9 meters of borated concrete beginning 0.5 meters below the earth's surface to obtain fluence distributions to a depth of 3.0 meters. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon and, in all cases, the secondary gamma ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths greater than 0.6 meter. 4 figures, 4 tables
Radiation transport in earth for neutron and gamma ray point sources above an air-ground interface
Energy Technology Data Exchange (ETDEWEB)
Lillie, R.A.; Santoro, R.T.
1979-03-01
Two-dimensional discrete ordinates methods were used to calculate the instantaneous dose rate in silicon and neutron and gamma ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 meters) above an air--ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth--concrete model containing 0.9 meters of borated concrete beginning 0.5 meters below the earth's surface to obtain fluence distributions to a depth of 3.0 meters. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon and, in all cases, the secondary gamma ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths greater than 0.6 meter. 4 figures, 4 tables.
Radiation transport in earth for neutron and gamma-ray point sources above an air-ground interface
International Nuclear Information System (INIS)
Two-dimensional discrete-ordinates methods have been used to calculate the instantaneous dose rate in silicon and neutron and gamma-ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 m) above an air-ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth-concrete model containing 0.9 m of borated concrete beginning 0.5 m below the earth's surface to obtain fluence distributions to a depth of 3.0 m. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon, and in all cases, the secondary gamma-ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths > 0.6 m
Radiation transport in earth for neutron and gamma-ray point sources above an air-ground interface
Energy Technology Data Exchange (ETDEWEB)
Lillie, R.A.; Santoro, R.T.
1980-01-01
Two-dimensional discrete-ordinates methods have been used to calculate the instantaneous dose rate in silicon and neutron and gamma-ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 m) above an air-ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth-concrete model containing 0.9 m of borated concrete beginning 0.5 m below the earth's surface to obtain fluence distributions to a depth of 3.0 m. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon, and in all cases, the secondary gamma-ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths > 0.6 m.
HERMET: cell neutronic calculation code for MTR (materials testing reactors) fuels
International Nuclear Information System (INIS)
The HERMET neutronic calculation code was developed for resolution of systems, at a cell calculation level in one-dimensional plain geometry (MTR), preserving its heterogeneous character with or without reflecting boundary conditions and reducing the cost as regards time and machine-memory. This code also includes the burn-up calculation which may be performed with the critical spectra B0, B1 or the one improved by leakages corresponding to the buckling given by the user. The burn-up scheme may be carried out by a transport equation with intermediate stages without flux reevaluation or by a predictor-corrector scheme. (Author)
Neutronic calculations for the new fuel configuration of the ETRR-1 research reactor
International Nuclear Information System (INIS)
Neutronic calculations were performed for the new loading configuration of the ETRR-1 research reactor. The MCNP three dimensions Monte Carlo code and the two dimensions CITATION code are used to model the reactor. The power and thermal flux distributions in the reactor core are calculated. The power peak factor and the effect of control rod insertion on both flux and power profiles in the reactor core are determined and analyzed. The partial and total control rods worth are calculated. It was found that the difference between MCNP and CITATION in power distributions is 4 to 8% and for thermal flux ranges between 3 to 14%. (orig.)
Neutronic calculation for cobalt irradiation devices and test loop of etrr2 research reactor
International Nuclear Information System (INIS)
MCNP monte Carlo code were used to model ETRR-2 research reactor. The model were used to simulate the irradiation facilities of the reactor. The reactivity worth and neutron flux for cobalt irradiation device and test loop were calculated. The axial thermal flux were also calculated at the six irradiation water channels around the reactor core. The results of the present model were compared both with the experimental measurements for cobalt device and with the design calculations for test loop. Satisfactory agreement were found between the present MCNP results and measurements for both cobalt device and test loop
International Nuclear Information System (INIS)
Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n)4He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed at the Oak Ridge National Laboratory. The facility, NE-213 liquid scintillator detector system, and experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed
Measurement and calculation of the neutron flux distribution in the RP-10 reactor
International Nuclear Information System (INIS)
In this work implementing experimental methods are implemented for easy reproduction for measuring the spatial distribution or thermal neutron flux in the RP-10 reactor core. Using two measuring methods: the passive and the active ones. In the passive method was used the activation technique using foils such as gold, manganese, and indium. These were irradiated in the reactor core and treated through the Westcott's formalism. In the active method was used the Self Powered Neutron Detectors (SPNs) for which was necessary to condition the detectors response for the data acquisition. The knowledge of the spatial distribution of RP-10 reactor neutrons flux will contribute in the understanding of other interesting parameters of reactor physics such as power density, reactivity, buckling, etc.. Wish knowledge is important for reactor operation. Fuel burnup calculations as well as others related to safety. (author)
Neutron matter from chiral two- and three-nucleon calculations up to N$^3$LO
Drischler, C; Hebeler, K; Schwenk, A
2016-01-01
Neutron matter is an ideal laboratory for nuclear interactions derived from chiral effective field theory since all contributions are predicted up to next-to-next-to-next-to-leading order (N$^3$LO) in the chiral expansion. By making use of recent advances in the partial-wave decomposition of three- nucleon (3N) forces, we include for the first time N$^3$LO 3N interactions in many-body perturbation theory (MBPT) up to third order and in self-consistent Green's function theory (SCGF). Using these two complementary many-body frameworks we provide improved predictions for the equation of state of neutron matter at zero temperature and also analyze systematically the many-body convergence for different chiral EFT interactions. Furthermore, we present an extension of the normal-ordering framework to finite temperatures. These developments open the way to improved calculations of neutron-rich matter including estimates of theoretical uncertainties for astrophysical applications.
Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna
Weber, H. W.; Böck, H.; Unfried, E.; Greenwood, L. R.
1986-02-01
In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10 17, 6.1 × 10 16 ( E 0.1 MeV) and 4.0 × 10 16 ( E > 1 MeV) m -2 s -1. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.
FLUKA Calculation of the Neutron Albedo Encountered at Low Earth Orbits
Claret, Arnaud; Combier, Natacha; Ferrari, Alfredo; Laurent, Philippe
2014-01-01
This paper presents Monte-Carlo simulations based on the Fluka code aiming to calculate the contribution of the neutron albedo at a given date and altitude above the Earth chosen by the user. The main input parameters of our model are the solar modulation affecting the spectra of cosmic rays, and the date of the Earth’s geomagnetic fi eld. The results consist in a two-parameter distribution, the neutron energy and the angle to the tangent plane of the sphere containing the orbi t of interest, and are provided by geographical position above the E arth at the chosen altitude. This model can be used to predict the te mporal variation of the neutron fl ux encountered along the orbit, and thus constrain the determination of the instrumental backg round noise of space experiments in low earth orbit.
Energy Technology Data Exchange (ETDEWEB)
Kuijper, J.C.
1992-01-01
The aim of the authors' work was to investigate the static and dynamic properties of a GCFR with oscillating (moving) fuel gas. A simplified schematic diagram of such a GCFR, similar to the concept of Kistemaker (Kis78a), is shown. It consists of a graphite cylinder of, say, 2 m diameter and 10 m length, filled with a mixture of uranium and carbon fluorides (UCF) at high temperature in ionized state, in chemical and thermodynamical equilibrium with the graphite cylinder wall (Kis78a, Kis86, Kle87). The cylindrical gas space is divided into an active 'core' region, surrounded by an effective (thick) neutron reflector, and a so-called 'expander' region, surrounded by a much less effective (thinner or with neutron poison) neutron reflector. In operation, part of the fuel gas oscillates back and forth between core and expander region. The investigation requires the study of neutron statics, neutron kinetics, reactor gas thermodynamics and gas dynamics, resulting in a combined calculational model, containing these aspects. In order to achieve this the authors followed a step-by-step approach.
Neutron Flux and Activation Calculations for a High Current Deuteron Accelerator
Coniglio, Angela; Sandri, Sandro
2005-01-01
Neutron analysis of the first Neutral Beam (NB) for the International Thermonuclear Experimental Reactor (ITER) was performed to provide the basis for the study of the following main aspects: personnel safety during normal operation and maintenance, radiation shielding design, transportability of the NB components in the European countries. The first ITER NB is a medium energy light particle accelerator. In the scenario considered for the calculation the accelerated particles are negative deuterium ions with maximum energy of 1 MeV. The average beam current is 13.3 A. To assess neutron transport in the ITER NB structure a mathematical model of the components geometry was implemented into MCNP computer code (MCNP version 4c2. "Monte Carlo N-Particle Transport Code System." RSICC Computer Code Collection. June 2001). The neutron source definition was outlined considering both D-D and D-T neutron production. FISPACT code (R.A. Forrest, FISPACT-2003. EURATOM/UKAEA Fusion, December 2002) was used to assess neutron...
International Nuclear Information System (INIS)
The calculations given in the paper are intended to explain the delayed neutron emission (energy spectra and emission probabilities Pn) which follows the β- disintegration of the precursors produced by fission. The probability of β- transition, the level density ω (E, J) of the emitter and the competition (β-, γ) and (β-, n) de-energizations are analysed for each precursor studied. All the possible channels open to the process of neutron emission on grounds of energy considerations (Qβ-, Bn) are taken into account through the introduction of the spin and parity selection rules at each stage of the sequence: precursor, emitter, final nucleus. The results of the calculations are compared with the known experimental measurements of the neutron energy spectra and probabilities Pn. The precursors 87Br, 88Br, 137I and 93-97Rb were selected for this examination. This comparison shows in particular that the structure of experimental energy spectra can be well reproduced by the calculations given in the paper. Moreover, it emerges that the spectra calculated are very sensitive to the choice of the spins of the precursor and the final nucleus. (author)
EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2
International Nuclear Information System (INIS)
The neutronics and the transmutation properties of the IAEA ADS benchmark 3.2 setup, the 'Yalina' experiment or ISTC project B-70, have been studied through an extensive amount of 3-D Monte Carlo calculations at CERN. The simulations were performed with the state-of-the-art computer code package EA-MC, developed at CERN. The calculational approach is outlined and the results are presented in accordance with the guidelines given in the benchmark description. A variety of experimental conditions and parameters are examined; three different fuel rod configurations and three types of neutron sources are applied to the system. Reactivity change effects introduced by removal of fuel rods in both central and peripheral positions are also computed. Irradiation samples located in a total of 8 geometrical positions are examined. Calculations of capture reaction rates in 129I, 237Np and 243Am samples and of fission reaction rates in 235U, 237Np and 243Am samples are presented. Simulated neutron flux densities and energy spectra as well as spectral indices inside experimental channels are also given according to benchmark specifications. Two different nuclear data libraries, JAR-95 and JENDL-3.2, are applied for the calculations
Measurement and calculation of neutron leakage from a medical electron accelerator
International Nuclear Information System (INIS)
The leakage neutron spectra and dose equivalent were systematically measured in the irradiation field, treatment room, maze, and outside the shielding door at the microtron medical electron accelerator facility of the National Cancer Center, Tokyo. For these measurements, we used two types of multimoderator neutron spectrometers (Bonner spheres containing indium activation detectors and 3He detector), an aluminum activation detector, and a commercially available neutron rem counter. The measured results were compared with the combined calculation of the one-dimensional ANISN and two-dimensional DOT3.5 discrete ordinates transport codes. The calculation was performed by using a measured source spectrum in the irradiation field and by computer modeling of the maze entrance. The calculation indicated good agreement in spectral shape and agreement with experiment within a factor of 2 in absolute dose-equivalent values. This transport calculation was systematically repeated for different geometrical and material parameters, and simple analytical formulas and their parameters applicable for shielding design of a medical electron accelerator facility were obtained in general form
232Th and 238U neutron emission cross section calculations and analysis of experimental data
International Nuclear Information System (INIS)
In this study, pre-equilibrium neutron-emission spectra produced by (n,xn) reactions on nuclei 232Th and 238U have been calculated. Angle-integrated cross sections in neutron induced reactions on targets 232Th and 238U have been calculated at the bombarding energies up to 18 MeV. We have investigated multiple pre-equilibrium matrix element constant from internal transition for 232Th (n,xn) neutron emission spectra. In the calculations, the geometry dependent hybrid model and the cascade exciton model including the effects of pre-equilibrium have been used. In addition, we have described how multiple pre-equilibrium emissions can be included in the Feshbach-Kerman-Koonin (FKK) fully quantum-mechanical theory. By analyzing (n,xn) reaction on 232Th and 238U, with the incident energy from 2 Me V to 18 Me V, the importance of multiple pre-equilibrium emission can be seen cleady. All calculated results have been compared with experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other
40 CFR 86.166-12 - Method for calculating emissions due to air conditioning leakage.
2010-07-01
... to air conditioning leakage. 86.166-12 Section 86.166-12 Protection of Environment ENVIRONMENTAL... for calculating emissions due to air conditioning leakage. This section describes procedures used to determine a refrigerant leakage rate in grams per year from vehicle-based air conditioning units....
Neutronic and thermal calculation of blanket for high power operating condition of fusion reactor
International Nuclear Information System (INIS)
Internal (breeding region) structures of ceramic breeder blanket to accommodate high power operating conditions such as a DEMO reactor have been investigated. The conditions considered here are the maximum neutron wall load of 2.8 MW/m2 at outboard midplane corresponding to a fusion power of 3.0 GW and the coolant temperature of 200 degrees C. Structure of a blanket is based on the layered pebble bed concept, which has been proposed by Japan since the ITER CDA. Lithium oxide with 50% enriched 6Li is used in a shape of small spherical pebbles which are filled in a 316SS can avoid its compatibility issue with Be. Beryllium around the breeder can is filled also in a shape of spherical pebbles which works not only as a neutron multiplier but also as a thermal resistant layer to maintain breeder temperature for effective in-situ tritium recovery. Diameters and packing fractions of both pebbles are ≤ 1 mm and 65%, respectively. A layer of block Be between cooling panels is introduced as a neutron multiplier (not as the thermal resistant layer) to enhance tritium breeding performance. Inlet temperature of water coolant is 200 degrees C to meet the high temperature conditioning requirement to the first wall which is one of walls of the blanket vessel. Neutronics calculations have been carried out by one-dimensional transport code, and thermal calculations have also been carried out by one-dimensional slab code
International Nuclear Information System (INIS)
Full text: The neutron-capture therapy with use of gadolinium-containing pharmacological preparations is one of perspective and not enough investigated directions of application of neutron irradiation in medicine. At definition of the absorbed dose of neutron-capture therapy one of important questions is definition of concentration gadolinium and pharmacokinetics in irradiated tumour. In the given study has been investigated pharmacokinetics of gadolinium-containing preparation 'Magnevist' at intratumoral injection in inoculated tumours of sarcoma C180 at mice. For 'Magnevist' detection its property of radioopacity has been used. In experiments to mice with inoculated tumours C180 the various doses of 'Magnevist' (0.1, 0.2, 0.3 and 0.4 ml) were injected into tumour centre. X-ray images were made before 'Magnevist' injection (control) and after preparation injection every 5 minutes within one hour. It has been shown that at dose 0.1 ml 'Magnevist' eliminated from tumour within 10 minutes. At higher doses of preparation more slow elimination of 'Magnevist' from injection site was observed. Obtained results allow with sufficient accuracy to calculate the time of presence of optimum concentration of 'Magnevist' in tumour at intratumoral injection. It in turn gives the chance to calculate precisely the absorbed dose at irradiation by beam of epi-thermal neutrons. (author)
International Nuclear Information System (INIS)
The thermal neutron flux distribution in a water phantom caused by narrow beam are separately calculated for scattered and non-scattered neutrons. Two-dimensional transport code TWOTRAN in the SRAC code system is used for the calculation of scattered neutrons. The calculated results are compared with experimental ones measured by using indium (n, γ) reactions at the Nickel-Mirror Neutron Guide-Tube (NM NGT) attached to the KUR (Kyoto University Research Reactor, 5MW). Breakdown of flux distribution into two components and their depth dependency are clearly observed. Comparison is made between experiments and calculations of the total neutron flux distribution for the incident axis in the water phantom. A good agreement is obtained. Investigation is made of the contours of neutron intensities in 5 cm and 9 cm phantoms. They show similar intensity contours up to 15 mm radius from the beam incident surface. Calculation is also made of the axial neutron flux distributions in phantoms with different diameters. Very similar distributions are seen near the incident surface. (Nogami, K.)
International Nuclear Information System (INIS)
Well-verified models for elastic and inelastic scattering of fast neutrons (and protons) from 12C and 16O are translated into a useful form for transport, dosimetry, and scintillator response calculations. The information presented here is complementary to published calculations of secondary-particle production by incident neutrons in the energy region 15 ≤ En ≤ 60 MeV. Tables are given of Legendre expansion coefficients derived from fits to experimental cross sections at incident neutron energies En = 18-26 MeV and from fits to model predictions for 30- to 60-MeV neutrons. 11 refs., 8 tabs
Energy Technology Data Exchange (ETDEWEB)
Ohta, Masayuki, E-mail: ohta.masayuki@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)
2013-10-15
In order to examine a basic performance of the TRIPOLI code, two types of analyses were carried out with TRIPOLI-4.4 and MCNP5-1.40; one is a simple model calculation and the other is an analysis of iron fusion neutronics experiments with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA). In the simple model calculation, we adopted a sphere of 0.5 m in radius with a 20 MeV neutron source in the center and calculated leakage neutron spectra from the sphere. We also analyzed in situ and Time-of-Flight (TOF) experiments for iron at JAEA/FNS. For the in situ experiment, neutron spectra and reaction rates for dosimetry reactions were calculated for several points inside the assembly. For the TOF experiment, angular neutron leakage spectra from the assembly were calculated. Results with TRIPOLI were comparable to those with MCNP in most calculations, but a difference between TRIPOLI and MCNP calculation results, probably caused by inadequate treatment of inelastic scattering data in TRIPOLI, appears in some calculations.
Calculation of the IAEA ADS neutronics benchmark (stage-1) (2D discrete coordinate method)
International Nuclear Information System (INIS)
To study the neutronics for the ADS system, a set of computation software based on discrete ordinate method is selected and established. The set is tested through an IAEA benchmark. In the test process, the understanding and using of this software set are improved. The benchmark is analyzed. The calculations include the effective multiplication factor keff , the required strength of the spallation neutron source for 1.5 GW thermal power, the distribution of power density and the spectrum index, and the void effect at the beginning of life, BOL; the spatial and time-dependent density distribution of various nuclides (actinides and fission products) for burn-up process. The results are given in figures and tables and are consistent with calculations made abroad. The conclusion is that this software set can be applied to the optimization of design study for the ADS system
dpa calculations for neutron irradiated amorphous Fe40Ni40B20 ribbons
International Nuclear Information System (INIS)
Calculations of the displacement per atom have been performed for an amorphous metal Fe40Ni40B20, which has been exposed to incore reactor irradiation. Up to 99% of the radiation induced displacements result from the nuclear reaction 10B(n,α)7Li + 2.79 MeV, which is initiated by thermal neutrons. The number of primary knock-on atoms generated by α- and Li-particles and the size of cascades have also been calculated. A fluence of thermal neutrons of 1 x 1019 nsub(th)/cm2 is found to produce a damage in amorphous Fe40Ni40B20 which corresponds to 0.65 dpa. (orig.)
Calculated neutron-induced cross sections for 53Cr from 1 to 20 MeV
International Nuclear Information System (INIS)
Neutron-induced cross sections of 53Cr have been calculated in the energy regions from 1 to 20 MeV. The quantities obtained are the cross sections for the reactions (n,n'γ), (n,2n), (n,np), (n,nα), (n,pγ), (n,pn), (n,αγ), (n,αn), (n,d), (n,t), (n,3He), and (n,γ), as well as the spectra of emitted neutrons, protons, alpha particles, and gamma rays. The precompound process was included above 5 MeV in addition to the compound process. For the inelastic scattering, the contribution of the direct interaction was calculated with DWBA. 36 refs., 23 figs., 11 tabs
Calculation and measurement of neutron flux in internal parts of the VVER-1000 mock-up
International Nuclear Information System (INIS)
Highlights: • Measurement of reaction rates in reactor baffle simulator. • Comparison of calculated and measured reaction rates in reactor baffle. • Measurement of fast neutron spectra in reactor baffle cooling channel. • Determination of 3He attenuation in lateral reflector model. - Abstract: The radiation situation in the reactor’s internal stainless steel parts is an important parameter during reactor operation. They are the most radiation-stressed structures because they are very close to the fuel. Knowledge of neutron flux distribution is important both for estimation of radiation-induced swelling of internal parts, radiation heating, and internal part activation. This paper aims to compare the experimental and calculation reaction rate distribution in a VVER-1000 mock-up placed in reactor LR-0
International Nuclear Information System (INIS)
the availability of burn-up data is an essential first step in any systematic approach to the enhancement of safety, economics and performance of research reactors. A computer program has been designed to solve the system of equations describing the depletion, decay and production of uranium, plutonium and transplutonium nuclides. monte Carlo code was used to calculate the effective one group microscopic cross sections averaged over ETRR-1 fuel cell. the compositions of actinide isotopes, burn-up and neutrons emission rate have been calculated as a function of irradiation time and cooling time. results indicate that the amount of plutonium produced and neutrons emission rate are strongly dependent on the fuel burn-up
Neutron and photon shielding benchmark calculations by MCNP on the LR-0 experimental facility.
Hordósy, G
2005-01-01
In the framework of the REDOS project, the space-energy distribution of the neutron and photon flux has been calculated over the pressure vessel simulator thickness of the LR-0 experimental reactor, Rez, Czech Republic. The results calculated by the Monte Carlo code MCNP4C are compared with the measurements performed in the Nuclear Research Institute, Rez. The spectra have been measured at the barrel, in front of, inside and behind the pressure vessel in different configurations. The neutron measurements were performed in the energy range 0.1-10 MeV. This work has been done in the frame of the 5th Frame Work Programme of the European Community 1998-2002. PMID:16604591
International Nuclear Information System (INIS)
The importance of accounting for resonance self-screening effects in multigroup cross sections when calculating fast reactors and neutron shields is considered. Formulae for averaging cross sections over resonance features with the account of anisotropy for scattering with large energy losses are derived. The model calculations of neutron fluxes have been performed for a U-H mixture (rhosub(H)/rhosub(U)=0.1), a U-Fe-H mixture and for the latter with rhosub(5)/rhosub(Fe)=0.01-0.5. It is concluded that in hydrogen-containing reactors the effect may be significant if the core contains iron in large quantities. The cross section averaging is considered for 3 systems: the KBR-2 critical assembly, spherical model of a large breeder, critical sphere of UO2 with 30% enrichment. The scattering anisotropy changes the multiplication factors of the first two systems by about 0.3%
Analysis and evaluation of critical experiments for validation of neutron transport calculations
International Nuclear Information System (INIS)
The calculation schemes, computational codes and nuclear data used in neutronic design require validation to obtain reliable results. In the nuclear criticality safety field this reliability also translates into a higher level of safety in procedures involving fissile material. The International Criticality Safety Benchmark Evaluation Project is an OECD/NEA activity led by the United States, in which participants from over 20 countries evaluate and publish criticality safety benchmarks. The product of this project is a set of benchmark experiment evaluations that are published annually in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. With the recent participation of Argentina, this information is now available for use by the neutron calculation and criticality safety groups in Argentina. This work presents the methodology used for the evaluation of experimental data, some results obtained by the application of these methods, and some examples of the data available in the Handbook.
Power and neutron flux calculation for the PUSPATI TRIGA Reactor using MCNP
International Nuclear Information System (INIS)
The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the calculation of neutron flux and power distribution in PUSPATI TRIGA REACTOR (RTP) 14th core configuration. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core and fuels. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data as well as S (α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed. (author)
Neutron spectra calculation in material in order to compute irradiation damage
International Nuclear Information System (INIS)
This short presentation will be on neutron spectra calculation methods in order to compute the damage rate formation in irradiated structure. Three computation schemes are used in the French C.E.A.: (1) 3-dimensional calculations using the line of sight attenuation method (MERCURE IV code), the removal cross section being obtained from an adjustment on a 1-dimensional transport calculation with the discrete ordinate code ANISN; (2) 2-dimensional calculations using the discrete ordinates method (DOT 3.5 code), 20 to 30 group library obtained by collapsing the 100 group a library on fluxes computed by ANISN; (3) 3-dimensional calculations using the Monte Carlo method (TRIPOLI system). The cross sections which originally came from UKNDL 73 and ENDF/B3 are now processed from ENDF B IV. (author)
International Nuclear Information System (INIS)
Highlights: ► We have extended the KAERI library generation system to include gamma cross section generation capability. ► A gamma transport/diffusion calculation module has been implemented in KARMA 1.2. ► The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. - Abstract: KAERI has developed a lattice transport calculation code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and its library generation system. Recently, the library generation system has been extended to include a gamma cross section generation capability and a gamma transport/diffusion calculation module has been implemented in KARMA 1.2. The method of characteristics for the neutron transport calculation to estimate eigenvalue has been utilized to predict gamma flux distribution and energy deposition. In addition, the coarse mesh finite difference method with diffusion approximation has also been utilized to estimate gamma flux distribution and energy depositions for each coarse mesh with homogenized pins as a computationally efficient alternative. This paper describes the procedure to generate neutron induced gamma production and gamma cross section data, and the methods to predict gamma flux distribution, gamma energy deposition and gamma smeared pin power distribution. The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. And it is noted that gamma smeared power distributions predicted by coarse mesh diffusion calculation are very accurate compared to the results of transport calculation
International Nuclear Information System (INIS)
Efigie is a program written in Fortran V which can calculate the concentration of radionuclides produced by neutron irradiation of a target made of either a single isotope or several isotopes. The program includes optimization criteria that can be applied when the goal is the production of a single nuclide. The effect of a cooling time before chemical processing of the target is also accounted for.(author)
Fast nodal core-wise green's function method for neutron diffusion calculations
International Nuclear Information System (INIS)
A fast nodal core-wise Green's function method for neutron diffusion calculations was developed. A new idea of building core-wise Green's function library was proposed, and the computer code CGFM was encoded. It was qualified by some benchmark problems and the Qinshan Nuclear Power Plant problem. The numerical results demonstrated that this method is 10 times faster than nodal Green's function method (NGFM) with idea precision
International Nuclear Information System (INIS)
This report contains the texts of the invited presentations (20) delivered at the third Research Co-ordination Meeting of the Co-ordinated Research Programme on Methods for the Calculation of Neutron Nuclear Data for Structural Materials of Fast and Fusion Reactors. The meeting was held at the IAEA Headquarters, Vienna, Austria, from 20 to 22 June 1990. A separate abstract was prepared for each of these presentations. Refs, figs and tabs
Fission Product Decay Heat Calculations for Neutron Fission of 232Th
Son, P. N.; Hai, N. X.
2016-06-01
Precise information on the decay heat from fission products following times after a fission reaction is necessary for safety designs and operations of nuclear-power reactors, fuel storage, transport flasks, and for spent fuel management and processing. In this study, the timing distributions of fission products' concentrations and their integrated decay heat as function of time following a fast neutron fission reaction of 232Th were exactly calculated by the numerical method with using the DHP code.
International Nuclear Information System (INIS)
An improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell has been developed. Expanding the neutron flux and source into a series of even powers of the radius, one' gets a convenient method for integration of the one-energy group integral transport equation. It is shown that it is possible to perform an analytical integration in the x-y plane in one variable and to use the effective Gaussian integration over another one. Choosing a convenient distribution of space points in fuel and moderator the transport matrix calculation and cell reaction rate integration were condensed. On the basis of the proposed method, the computer program DISKRET for the ZUSE-Z 23 K computer has been written. The suitability of the proposed method for the calculation of the thermal-neutron-flux distribution in a reactor cell can be seen from the test results obtained. Compared with the other collision probability methods, the proposed treatment excels with a mathematical simplicity and a faster convergence. (author)
Modeling IR-8 research reactor of RRC KI for precision neutronics calculations
International Nuclear Information System (INIS)
The IR-8 pool type research reactor of RRC KI was commissioned in 1981 for carrying out fundamental and applied researches in various areas of science and technique. MCU-PTR code with the MCUDB50 constants library was created to ensure the safe operation of the reactor and for the calculation support of experimental research. The MCU-PTR code is intended for simulation of neutron transport by means of the Monte Carlo method on the basis of evaluated nuclear data taking into account of changes in the nuclide composition of materials in interaction with neutrons. Full-detailed 3D mathematical models of different states of the IR-8 reactor were created for precision neutronics calculations with use of MCU-PTR code. The code verification for pool and tank type research reactors performed based on IHECSBEP Criticality Benchmark Experiments and experiments carried out at the IR-8 reactor. The MCU-PTR code is currently used for calculations of the IR-8 reactor taking into account of fuel burn-up with HEU or LEU, poisoning of the beryllium reflector and burn-up absorber in CPS rods. (author)
Air pollution studies in Tianjing city using neutron activation analysis techniques
International Nuclear Information System (INIS)
Two sites of airborne sampling from industrial and residential areas were made in Tianjing city during February and June using PM-10 sampler and analyzed by NAA techniques; Comparison of air pollution between urban and rural area in Tianjing city was made using neutron activation analysis techniques and some other data analyzing techniques. (author)
Snehota, Michal; Jelinkova, Vladimira; Sobotkova, Martina; Sacha, Jan; Vontobel, Peter; Hovind, Jan
2015-02-01
Saturated flow in soil with the occurrence of preferential flow often exhibits temporal changes of saturated hydraulic conductivity even during the time scale of a single infiltration event. These effects, observed in a number of experiments done mainly on heterogeneous soils, are often attributed to the changing distribution of water and air in the sample. We have measured the variation of the flow rates during the steady state stage of the constant head ponded infiltration experiment conducted on a packed sample composed of three different grades of sand. The experiment was monitored by quantitative neutron imaging, which provided information about the spatial distribution of water in the sample. Measurements were taken during (i) the initial stages of infiltration by neutron radiography and (ii) during the steady state flow by neutron tomography. A gradual decrease of the hydraulic conductivity has been observed during the first 4 h of the infiltration event. A series of neutron tomography images taken during the quasi-steady state stage showed the trapping of air bubbles in coarser sand. Furthermore, the water content in the coarse sand decreased even more while the water content in the embedded fine sand blocks gradually increased. The experimental results support the hypothesis that the effect of the gradual hydraulic conductivity decrease is caused by entrapped air redistribution and the build up of bubbles in preferential pathways. The trapped air thus restricts the preferential flow pathways and causes lower hydraulic conductivity.
International Nuclear Information System (INIS)
The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared
Development of calculational procedures for the neutron physics design of advanced reactors
International Nuclear Information System (INIS)
The Nuclear Reactor Center Karlsruhe has been involved with the development of Light Water Tight Lattice Reactors (LWTLR) since more than ten years. A considerable amount of thermohydraulic and nuclear physics code development has been performed during this time. The present paper describes the main aspects of the neutron physics calculational tools. From the neutron physics point of view, two different tasks have to be adapted for LWTLR calculations: determination of mean cross sections sets within the hexagonal fuel assemblies (FA); determination of the characteristics of LWR cores with hexagonal FA. All developments for the neutron physics design of LWTLR have been performed within the established system for Fast Breeder Reactor (FBR) calculations at KfK, KAPROS, using a various number of available options for FBR work. The present status of the calculational tools for LWTLR-investigations will be described, especially the features of a newly developed KAPROS procedure ARCOSI: Advanced Reactor COre SImulator, including: Preparation of the ARCOSI library HXSLIB, containing burnup dependent cross section sets for FA with control rod positions containing control rod material or waterholes and with borated water in the moderator region of the pin cells. Also, data for different coolant densities and pin cell temperatures may be processed; Simulation of equilibrium core calculations, including critical reactivity search by waterboration control and simple FA management. Three-dimensional full core calculations are performed with the KAPROS version of the hexagonal nodal code HEXNOD, developed by Wagner, KWU; Powerful interfaces for interactive graphical analysis of results. (author). 37 refs, 12 figs
Determination of neutron flux distribution across the RB reactor with large central air hol
International Nuclear Information System (INIS)
The need for the irradiation of large samples in the fast neutron field was led to design of a strongly heterogeneous core at the RB heavy water reactor. This configuration, operates as the internal fast neutron converter, introduces many difficulties in reactor safety and criticality analysis. In this paper, the collision probability method in two-dimensional r-z geometry, implemented in the VEGA code is applied. The neutron flux calculated by the VEGA code is compared to the results obtained by the MCNP sup T sup M continuous-energy Monte Carlo code and to the measured distribution. Results of VEGA and MCNP codes show good agreement with measured values. (author)
Radiation doses from radiation sources of neutrons and photons by different computer calculation
International Nuclear Information System (INIS)
In the present paper the calculation technique aspects of dose rate from neutron and photon radiation sources are covered with reference both to the basic theoretical modeling of the MERCURE-4, XSDRNPM-S and MCNP-3A codes and from practical point of view performing safety analyses of irradiation risk of two transportation casks. The input data set of these calculations -regarding the CEN 10/200 HLW container and dry PWR spent fuel assemblies shipping cask- is frequently commented as for as connecting points of input data and understanding theoric background are concerned
Scoping calculations within the framework of the neutron physics PWR design procedure SAV90
International Nuclear Information System (INIS)
Scoping calculations within the framework of the neutron physics PWR design procedure SAV90. Reliable forecasts of the performance characteristics of pressurized water reactors over several years are an essential precondition for preoptimizing fuelling plans and reducing fuel costs. If a coupled system of three-dimensional partial differential equations (two-group-diffusion equations) is used to exactly describe the processes in the reactor core, a CYBER-990 computer needs about one hour for one burnup cycle. A simplified two-dimensional scoping calculation model has been developed which, within a short time and at low expense, supplies reliable propositions for judging a fuelling plan. (orig./DG)
Energy Technology Data Exchange (ETDEWEB)
Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-10-01
A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)
ZZ UKCTR-1, Cross-Section Library for Neutron Flux and Neutron Reaction Rates in CTR Calculation
International Nuclear Information System (INIS)
1 - Description of problem or function: Format: ANISN, DOT, MORSE, SWANLAKE; Number of Groups: 46 energy group structure from 14.2 MeV to 1 MeV; Nuclides: Li-6, Li-7, O, Be, Pb, Nb, Fe, Ni, Cr, Zr, V, Ti, H, D, T, C, Al, B-10, B-11, Cu-63, Cu-65, F, Na, K, Mo. Origin: UKNDL; Weighting Spectrum: 1/(Sigma t (E).E) weighting is used for groups 1 to 44 with Maxwellian weighting for the two thermal groups. UKCTR1 is a data library of neutron cross sections for 25 materials in a 46 energy group structure from 14.2 MeV to 1 MeV. It is designed for calculation of neutron fluxes and reaction rates in controlled thermonuclear reactors. The energy group structure is fine at 14 MeV and there are two thermal groups; the lethargy interval width per energy group for decreasing energy is as follows: 0.014, 0.036, 2 x 0.15, 15 x 0.3, 25 x 0.5, 2.935 and 3.091. Reaction cross sections including partial inelastic data are provided for the following materials: Li-6, Li-7, O, Be, Pb, Nb, Fe, Ni, Cr, Zr, V, Ti, H, D, T, C, Al, B-10, B-11, Cu-63, Cu-65, F, Na, K, Mo. 1/(Sigma t (E).E) weighting is used for groups 1 to 44 with Maxwellian weighting for the two thermal groups. Anisotropy of scattering is represented by a P order up to 4 (usually 0 to 4). Data for hydrogen and deuterium both in water and heavy water and in the gaseous state is available. As a supplement, neutron kerma factors are included for each of the nuclides in the library as well as 98 activation cross sections of importance in fusion reactor work. (These 98 activation cross sections have been extracted from the bulk of the UKCTR-I library to be in a more convenient form for programs such as ANISN.) The kerma factors were computed using the code ENBAL2, a revised version of ENBAL, which calculates multigroup kerma factors directly from multigroup cross sections together with reaction Q-values. This approach allows neutron heating calculations to be performed consistently with the flux calculation. 2 - Method of
International Nuclear Information System (INIS)
The tangential horizontal channel of No. 3 of the Dalat Research Reactor has been opened and used during the 1990s. The utilizations of the thermal neutron beam at this channel were the Neutron Radiography and the Prompt Gamma Neutron Activation Analysis method (PGNAA). At present, the neutron beam used for nuclear structure data researches based on the Summing of Amplitude Coincident Pulses system (SACP). Beside, several related research equipments have been set up and operated for the research purposes. A renovation of the neutron channel, therefore, will play an important role in safe and effective utilizations of the neutron beam in fields of nuclear physic training and researches. A new configuration for radiation shielding has been simulated by MCNP code. The calculated results of dose rates for neutron and gamma at working positions are in range of dose rate limit. (author)
A new assembly-level Monte Carlo neutron transport code for reactor physics calculations
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This paper presents a new assembly-level Monte Carlo neutron transport code, specifically intended for diffusion code group-constant generation and other reactor physics calculations. The code is being developed at the Technical Research Centre of Finland (VTT), under the working title 'Probabilistic Scattering Game', or PSG. The PSG code uses a method known as Woodcock tracking to simulate neutron histories. The advantages of the method include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. The main drawback is the inability to calculate reaction rates in optically thin volumes. This narrows the field of application to calculations involving parameters integrated over large volumes. The main features of the PSG code and the Woodcock tracking method are introduced. The code is applied in three example cases, involving infinite lattices of two-dimensional LWR fuel assemblies. Comparison calculations are carried out using MCNP4C and CASMO-4E. The results reveal that the code performs quite well in the calculation cases of this study, especially when compared to MCNP. The PSG code is still under extensive development and there are both flaws in the simulation of the interaction physics and programming errors in the source code. The results presented here, however, seem very encouraging, especially considering the early development stage of the code. (author)
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The scattering of slow neutrons by the molecular systems is considered. A theoretical description of this phenomenon is particularly complicated for hydrogenous media. Granada's Synthetic Model of the neutron scattering is applicable for such cases. A comprehensive unified description of the application of model is presented with respect to the general formalism for calculation of the n-order energy-transfer scattering kernel for thermal neutrons. (author). 10 refs
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A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR
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Johnson, J.O.; Miller, L.F.; Kam, F.B.K.
1981-05-01
A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.
Anisotropic kernel p(μ → μ') for transport calculations of elastically scattered neutrons
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Literature in the area of anisotropic neutron scattering is by no means lacking. Attention, however, is usually devoted to solution of some particular neutron transport problem and the model employed is at best approximate. The present approach to the problem in general is classically exact and may be of some particular value to individuals seeking exact numerical results in transport calculations. For attempts neutrons originally directed toward the unit vector Omega, it attempts the evaluation of p(theta'), defined such that p(theta') d theta' is that fraction of scattered neutrons that emerges in the vicinity of a cone i.e., having been scattered to between angles theta' and theta' + d theta' with the axis of preferred orientation i; Omega makes an angle theta with i. The relative simplicity of the final form of the solution for hydrogen, in spite of the complicated nature of the limits involved, is a trade-off that truly is not necessary. The exact general solution presented here in integral form, has exceedingly simple limits, i.e., 0 ≤ theta' ≤ π regardless of the material involved; but the form of the final solution is extraordinarily complicated
Three-dimensional neutronics shielding analysis and calculation for FEB-E
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Three-dimensional shielding analysis has been performed for the FEB-E design by using the Monte Carlo code MCNP. Based on 143 MW fusion power and non-uniform source neutron sampling of the FEB-E, the average neutron wall loading is 0.481 MW/m2, and the peak outboard and inboard wall loading are 0.861 and 0.864 MW/m2, respectively. According to neutronics shielding analysis and calculation for the inboard and divertor duct regions along the toroidal field (TF) coils, adequate shielding designs have been obtained. Tungsten 17 cm thick in total is used in the inboard shield because of serious constraint of the space. In order to provide adequate protection against serious streaming neutron radiation in the divertor duct regions, total additional shield of 25 cm thick is used in the side of TF coils. The value of total heating in the TF coils is about 4.20 kW for the FEB-E design less than the requirement of 55 kW
Calculation of neutron and gamma-ray emission spectra produced by p + 27Al reactions
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Preliminary calculations of neutron and gamma-ray spectra induced by proton reactions on aluminum have been made to provide data required for shielding design for a proposed proton linear accelerator. The nuclear models used in this study were the preequilibrium and Hauser-Feshbach models as embodied in the GNASH program. This nuclear model code has been used in the past to successfully investigate higher energy (E less than or equal to 50 MeV) neutron and proton interactions with nuclei in the structural materials region. Because this study was of an exploratory nature, we did not attempt to optimize input parameters but instead relied upon global sets, especially for optical parameters. In particular, for neutrons we chose the Wilmore-Hodgson parameter set after confirmation of its suitability through comparison to n+27Al total cross-section data between 0.5 and 60 MeV. Agreement with the data on the level of 5-10% occurred. Comparisons were also made to measured nonelastic data for incident energies between 10 and 60 MeV. Again, there was generally good agreement although there was some tendency to overpredict such data for incident neutron energies below several MeV. For protons we found the Becchetti-Greenlees parameter set reproduced nonelastic data recently measured by McGill et al
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The specific yield functions m(R) for the neutron component of cosmic rays have been calculated on the basis of latitude surveys made on board the scientific ship Akademik Kurchatov' in late 1971 - early 1972. Calculations have been performed using the primary cosmic ray spectrum reconstructed with regard for possible anomalous modulation effects of the Sun. Significant discrepancies (up to 2 orders of magnitude) have been found in m(R) values obtained by different authors. They seem to be due to non-adequacy of the primary spectra to the latitude curves used in calculations. The accuracy of SCR absolute spectrum in the range of R > = 1 GV determined with the help of m(R) does not exceed a factor of 2. However it may be improved by comparing the calculated spectra with direct measurements in R < 1 GV range
PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation
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The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,γ), (n,n'), (n,p), (n,α), (n,d), (n,t), (n,3He), (n,2n), (n,n'p), (n,n'α), (n,n'd), (n,n't), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)
System of constants to calculate neutron transport with energy 10-2-4x108 eV
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A description of the library of nuclear data to calculate neutron transport in the energy range 10-2 eV-4x102 MeV (BND-400) is presented. The library contains a seven-group system of data for neutrons of E>10.5 MeV and a standard 26-group system for neutrons with E10.5 MeV, and those for matching with the file of data for neutrons with E<10.5 MeV are briefly described. In the BND-400 complex there are subroutines, which allow one to calculate the cross sections for neutron interaction with nuclei of matter with the help of various methods and models as well as to calculate group cross sections. It also provides output files in the form convenient for work. A brief instruction for BND-400 explotation on the computer BESM-6 is given
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This PhD Thesis aims to achieve a method for the modelling of the reflector surrounding the core for neutronics core calculations. This method should consider the EPR reactor specificities (steel reflector) and the increased demand in precision. In neutronics core calculations, the reflector can be represented either by albedos boundary conditions (current ratios) or by one or several media, surrounding the core, characterised by homogenized parameters. Those parameters (cross sections and diffusion coefficients) should be obtained using equivalence so that they allow a good reproduction of the reference albedos in a representative situation. During this PhD, such an equivalence method has been developed in the APOLLO-2 code with the minimization of a functional of the differences between the reference albedos and those computed with the equivalent parameters. Because of the positiveness constraints, a local minimization, such as Newton-like methods, is not always possible and we have therefore also implemented a Particle Swarm Optimization Algorithm for more than two energy groups' problems. The parameters obtained have been used in two dimensions EPR core calculations with the CRONOS-2 code for various fuel loadings in two to eight groups diffusion. Those core calculation have been validated against reference Monte-Carlo calculations and against core calculations with albedos boundary conditions. In addition to the increased easiness of utilization, the implemented equivalence method has yielded an improvement of the results for the two groups calculation. With a higher energy groups number, the use of a unique equivalent reflector does not account correctly for the two dimensions effects; a modelling with different reflector meshes has improved the results. The modelling of the reflector by two dimensions albedos boundary conditions is the more suited for the representation of the boundary conditions and, therefore, should the two dimensions albedos calculation
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This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs
Calculation of neutron fluence-to-dose conversion factors for extremities
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The Pacific Northwest Laboratory (PNL) is developing a standard for the performance testing of personnel extremity dosemeters for the US Department of Energy (DOE). Part of this effort requires the calculation of neutron fluence-to-dose conversion factors for finger and wrist/ankle extremities. This study focuses on conversion factors for two types of extremity models: (1) the polymethyl methacrylate (PMMA) phantom (as specified in the draft standard for performance testing of extremity dosemeters) and (2) more realistic extremity models composed of tissue and bone. Calculations for each type of model are based on both bare and D2O-moderated 252Cf sources. The results are then tabulated and compared with whole-body conversion factors. More appropriate average quality factors for the extremity models have also been computed from the energy dependent neutron fluence. Tabulated results show that conversion factors for both types of extremity phantoms are 3 to 28% lower than the corresponding whole-body phantom conversion factors for 252Cf neutron sources. This difference in extremity and whole-body conversion factors is attributable to the proportionally smaller amount of scattering that occurs in the extremity phantoms compared to whole-body phantoms. (author)
Calculation of neutron fluence-to-dose conversion factors for extremities
International Nuclear Information System (INIS)
The Pacific Northwest Laboratory is developing a standard for the performance testing of personnel extremity dosimeters for the US Department of Energy. Part of this effort requires the calculation of neutron fluence-to-dose conversion factors for finger and wrist extremities. This study focuses on conversion factors for two types of extremity models: namely the polymethyl methacrylate (PMMA) phantom (as specified in the draft standard for performance testing of extremity dosimeters) and more realistic extremity models composed of tissue-and-bone. Calculations for each type of model are based on both bare and D2O-moderated 252Cf sources. The results are then tabulated and compared with whole-body conversion factors. More appropriate energy-averaged quality factors for the extremity models have also been computed from the neutron fluence in 50 equally spaced energy bins with energies from 2.53 x 10-8 to 15 MeV. Tabulated results show that conversion factors for both types of extremity phantom are 15 to 30% lower than the corresponcung whole-body phantom conversion factors for 252Cf neutron sources. This difference in extremity and whole-body conversion factors is attributable to the proportionally smaller amount of back-scattering that occurs in the extremity phantoms compared with whole-body phantoms
Benchmark calculations on neutrons streaming through mazes at proton accelerator facilities
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In accelerator shielding designs one of the important issues is to estimate radiation streaming through mazes and ducts. In order to validate the accuracy of the calculation methods concerning such neutron streaming, benchmark analyses were carried out using two kinds of benchmark problems based on past experiments. The analyses showed that the design methods were applicable to neutron streaming calculations of proton accelerator facilities with an uncertainty within a factor of two. In the analyses, relative comparisons were conducted using a radiation source generated by GeV energy protons, and absolute comparisons were conducted using a low-energy neutron source of a few tens of MeV. A radiation streaming experiment was planned and carried out at KEK using a radiation source produced by a thin copper target irradiated by 12 GeV protons. The preliminary experimental analysis is presented below. In addition, the authors propose to compile benchmark problems on radiation streaming for accelerator facilities and to search for possible new streaming experiments at other facilities. (authors)
Neutrons and Gamma-Ray Dose Calculations in Subcritical Reactor Facility Using MCNP
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Ned Xoubi
2016-06-01
Full Text Available In nuclear experimental, training and teaching laboratories such as a subcritical reactor facility, huge measures of external radiation doses could be caused by neutron and gamma radiation. It becomes imperative to place the health and safety of staff and students in the reactor facility under proper scrutiny. The protection of these individuals against ionization radiation is facilitated by expected dose mapping and shielding calculations. A three-dimensional (3D Monte Carlo model was developed to calculate the dose rate from neutrons and gamma, using the ANSI/ANS-6.1.1 and the ICRP-74 flux-to-dose conversion factors. Estimation for the dose was conducted across 39 areas located throughout the reactor hall of the facility and its training platform. It was found that the range of the dose rate magnitude is between 7.50 E−01 μSv/h and 1.96 E−04 μSv/h in normal operation mode. During reactor start-up/shut-down mode, it was observed that a large area of the facility can experience exposure to a significant radiation field. This field ranges from 2.99 E+03 μSv/h to 3.12 E+01 μSv/h. There exists no noticeable disparity between results using the ICRP-74 or ANSI/ANS-6.1.1 flux-to-dose rate conversion factors. It was found that the dose rate due to gamma rays is higher than that of neutrons.
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Sensitivity of principal neutronics characteristic quantities for the neutron cross sections of JAERI Experimental Fusion Reactor (JXFR) has been studied by means of sensitivity analysis method based on linear perturbation theory. The same study was made previously. After publication of the previous results, however, the SWANLAKE code used to calculate sensitivities was found to include error derived during its conversion process. The study was thus repeated with corrected SWANLAKE. The quantities studied are calculational results for the first preliminary design of JXFR such as the (n, p) reaction rates of 58Ni and 54Fe in the outer part of superconducting toroidal field coil (TFC), the copper atomic displacement rate in the inner part of TFC and the tritium production rate in the outer blanket. Though the calculational results do not contradict essentially the results in the former study, the newly calculated sensivitities were found to be more or less different from the previous ones. Therefore, the results and discussion of analysis given in this report are revised, with the values corrected. The errors of the (n, p) reaction rates and the copper displacement rate due to the uncertainties of cross sections were estimated to be about 50 - 70% and 25 - 65%, respectively, taking into account the direct sensitivity of (n, p) reaction cross sections in the former. (author)
System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors
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Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)
2008-07-01
There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)
Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations
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This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO2-Gd2O3 poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)
Spatial distribution of muons in extensive air showers: experiment, calculations
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Investigations regarding the spatial distribution of the muons, possessing an energy of more than 10 GeV, have been carried out in the extensive air showers (EAS) at the sea level. The experiment has been staged at the Moscow State University complex installation. The experimental data are compared with the computations carried out for the various models of the hadronic interactions. At the same time it was supposed that the EAS were generated by the primary protons and that the value of the average transverse momentum of the secondary particles was equal to 0.4 GeV/s. It has been shown that the theoretical results are consistent with the experimental data
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Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)
2005-07-01
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
International Nuclear Information System (INIS)
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U235, U238, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
IRACM : A code system to calculate induced radioactivity produced by ions and neutrons
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It is essential to estimate of radioactivity induced in accelerator components and samples bombarded by energetic ion beams and the secondary neutrons of high-energy accelerator facilities in order to reduce the amount of radioactive wastes and to minimize radiation exposure to personnel. A computer code system IRACM has been developed to estimate product nuclides and induced radioactivity in various radiation environments of accelerator facilities. Nuclide transmutation with incident particles of neutron, proton, deuteron, alpha, 12C, 14N, 16O, 20Ne and 40Ar can be computed for arbitrary multi-layer target system in a one-dimensional geometry. The code system consists of calculation modules and libraries including activation cross sections, decay data and photon emission data. The system can be executed in both FACOM-M780 mainframe and DEC workstations. (author)
Calculation of the ex-core neutron noise induced by fuel vibrations in PWRs
International Nuclear Information System (INIS)
Calculation of the neutron noise induced by fuel assembly vibrations in two pressurized water reactor (PWR) cores has been performed to investigate the effect of cycle burnup on the properties of the ex-core detector noise. Pendular vibrations of individual fuel assemblies were assumed to occur at different locations in the core. The auto power spectra density (APSD) of the ex-core detector noise was evaluated with the assumption of stochastic vibrations along a random two-dimensional trajectory. The results show that no general monotonic variation of APSD was found. The increase of APSD occurs predominantly for peripheral assemblies. Assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core with the more realistic perturbation model, the effect of the peripheral assemblies will dominate and the increase of the amplitude of the ex-core neutron noise with burnup can be confirmed. (author)
International Nuclear Information System (INIS)
The one dimensional discrete ordinates code ANISN-F was used to calculate the thermal neutron flux distribution in water from a Ra-Be neutron source. The calculations were performed in order to investigate the different possibilities of the code as well as to verify the results of the calculations in terms of comparisons to corresponding experimental data. Two different group cross section libraries were used in the calculations and conclusions were drawn on the adequacy of these libraries for a fixed source type calculation. Furthermore, critically calculations were performed for an infinite homogeneous slab of multiplying material using different angular and spatial approximations. The results of these calculations were then compared to the corresponding results previously obtained at this department by a different method and a different code. (author)
Gambarini, G; Artuso, E; Giove, D; Felisi, M; Volpe, L; Barcaglioni, L; Agosteo, S; Garlati, L; Pola, A; Klupak, V; Viererbl, L; Vins, M; Marek, M
2015-12-01
The reliability of Fricke gel dosimeters in form of layers for measurements aimed at the characterization of epithermal neutron beams has been studied. By means of dosimeters of different isotopic composition (standard, containing (10)B or prepared with heavy water) placed against the collimator exit, the spatial distribution of gamma and fast neutron doses and of thermal neutron fluence are attained. In order to investigate the accuracy of the results obtained with in-air measurements, suitable MC simulations have been developed and experimental measurements have been performed utilizing Fricke gel dosimeters, thermoluminescence detectors and activation foils. The studies were related to the epithermal beam designed for BNCT irradiations at the research reactor LVR-15 (Řež). The results of calculation and measurements have revealed good consistency of gamma dose and fast neutron 2D distributions obtained with gel dosimeters in form of layers. In contrast, noticeable modification of thermal neutron fluence is caused by the neutron moderation produced by the dosimeter material. Fricke gel dosimeters in thin cylinders, with diameter not greater than 3mm, have proved to give good results for thermal neutron profiling. For greater accuracy of all results, a better knowledge of the dependence of gel dosimeter sensitivity on radiation LET is needed. PMID:26249744
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Villescas, G.; Corchon, F.
2013-07-01
he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.
International Nuclear Information System (INIS)
PEGASUS, a preequilibrium and evaporation theory code, was developed which calculates 17 neutron reaction cross sections, the particle spectra and the double differential cross sections. The code is suited to a rapid and scoping calculation. Theoretical model and the some results of calculation are presented. (author)
''TIBERE'' a calculational method of the effect of lattice heterogeneity on the neutron leakage
International Nuclear Information System (INIS)
The flux calculation in a heterogeneous assembly in APOLLO and APOLLO-2 codes is based on the collision probability method. On the contrary, the neutron leakage calculation is accomplished for a flux-weighted homogenized assembly, which is a good approximation in the case of a non-voided assembly (PWR in normal conditions). If an assembly contains voided zones (for example as the consequence of LOCA) or if the reactor, in normal operation, contains void or transparent media (Na), the approximation may become much more debatable. The formalism proposed here treats the leakage taking into account the heterogeneity of the medium, the leakage being treated as a type of reaction analogous to others. This formalism is introduced in APOLLO-2 code as a module and uses only tools similar to those used for the calculation of the flux. (author)
Calculated cross sections for neutron induced reactions on 19F and uncertainties of parameters
International Nuclear Information System (INIS)
Nuclear model codes were used to calculate cross sections for neutron-induced reactions on 19F for incident energies from 2 to 20 MeV. The model parameters in the codes were adjusted to best reproduce experimental data and are given in this report. The calculated results are compared to measured data and the evaluated values of ENDF/B-V. The covariance matrix for several of the most sensitive model parameters is given based on the scatter of measured data around the theoretical curves and the long-range correlation error of measured data. The results of these calculations form the basis for the new ENDF/B-VI fluorine evaluation. 44 refs., 64 figs., 14 tabs
Application of MCNP for neutronic calculations at VR-1 training reactor
International Nuclear Information System (INIS)
The paper presents the utilization of the Monte Carlo MCNP transport code for neutronic calculations of the VR-1 training reactor. Zero power light water reactor VR-1 is used mainly for training and partially for research. The reactor core consists of IRT-4M fuel elements with enrichment below 20 % of 235U and other components (e.g. control rods, fuel dummies, dry channels etc.). The reactor offers large variety of the core configurations and every year at least one critical experiment with new original core configuration is performed. The results of the calculations are compared with measured data collected during the last critical experiments performed with various reactor core configurations. A very good agreement between calculations and measurements is observed
Monte Carlo Calculation for Landmine Detection using Prompt Gamma Neutron Activation Analysis
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Park, Seungil; Kim, Seong Bong; Yoo, Suk Jae [Plasma Technology Research Center, Gunsan (Korea, Republic of); Shin, Sung Gyun; Cho, Moohyun [POSTECH, Pohang (Korea, Republic of); Han, Seunghoon; Lim, Byeongok [Samsung Thales, Yongin (Korea, Republic of)
2014-05-15
Identification and demining of landmines are a very important issue for the safety of the people and the economic development. To solve the issue, several methods have been proposed in the past. In Korea, National Fusion Research Institute (NFRI) is developing a landmine detector using prompt gamma neutron activation analysis (PGNAA) as a part of the complex sensor-based landmine detection system. In this paper, the Monte Carlo calculation results for this system are presented. Monte Carlo calculation was carried out for the design of the landmine detector using PGNAA. To consider the soil effect, average soil composition is analyzed and applied to the calculation. This results has been used to determine the specification of the landmine detector.
International Nuclear Information System (INIS)
A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)
Calculation of Ambient (H*(10)) and Personal (Hp(10)) Dose Equivalent from a 252Cf Neutron Source
Energy Technology Data Exchange (ETDEWEB)
Traub, Richard J.
2010-03-26
The purpose of this calculation is to calculate the neutron dose factors for the Sr-Cf-3000 neutron source that is located in the 318 low scatter room (LSR). The dose factors were based on the dose conversion factors published in ICRP-21 Appendix 6, and the Ambient dose equivalent (H*(10)) and Personal dose equivalent (Hp(10)) dose factors published in ICRP Publication 74.
International Nuclear Information System (INIS)
We adapted or used on ES EhVM, operating under the control of OS ES, the currently most common algorithms for calculating neutron spectra from measured reaction rates. These programs, together with the neutron cross-section and spectrum libraries, are part of the computerized information system SAIPS. The present article descibes the basic mathematical concepts used in the algorithms of the SAIPS calculation programs
Monte Carlo calculation of 60Co γ-ray's albedo-dose rate from the air
International Nuclear Information System (INIS)
The Monte Carlo calculation of 60Co γ-ray's albedo-dose rate from the air is reported. A formula is presented with which the relations of the albedo-doserate with some parameters are simulated and fitted
International Nuclear Information System (INIS)
This document represents Phase I of a two-phase project. The entire project consists of determining a series of minimum accidents of concern and their associated neutron and photon leakage spectra that may be used to determine Criticality Accident Alarm compliance with ANSI/ANS-8.3. The inadvertent assembly of a critical mass of material presents a multitude of unknown quantities. Depending on the particular process, one can make an educated guess as to fissile material. In a gaseous diffusion cascade, this material is assumed to be uranyl fluoride. However, educated assumptions cannot be readily made for the other variables. Phase I of this project is determining a bounding minimum accident of concern and its associated neutron and photon leakage spectra. To determine the composition of the bounding minimum accident of concern, work was done to determine the effects of geometry, moderation level, and enrichment on the leakage spectra of a critical assembly. The minimum accident of concern is defined as the accident that may be assumed to deliver the equivalent of an absorbed dose in free air of 20 rad at a distance of 2 meters from the reacting material within 60 seconds. To determine this dose, an analyst makes an assumption and choose an appropriate flux to dose response function. The power level required of a critical assembly to constitute a minimum accident of concern depends heavily on the response function chosen. The first step in determining the leakage spectra was to attempt to isolate the effects of geometry, after which all calculations were conducted on critical spheres. The moderation level and enrichment of the spheres were varied and their leakage spectra calculated. These spectra were then multiplied by three different response functions: the Henderson Flux to Dose conversion factors, the ICRU 44 Kerma in Air, and the MCNP Heating Detector. The power level required to produce a minimum accident of concern was then calculated for each combination
Neutron-deuteron scattering calculations with W-matrix representation of the two-body input
International Nuclear Information System (INIS)
Employing the W-matrix representation of the partial-wave T matrix introduced by Bartnik, Haberzettl, and Sandhas, we show for the example of the Malfliet-Tjon potentials I and III that the single-term separable part of the W-matrix representation, when used as input in three-nucleon neutron-deuteron scattering calculations, is fully capable of reproducing the exact results obtained by Kloet and Tjon. This approximate two-body input not only satisfies the two-body off-shell unitarity relation but, moreover, it also contains a parameter which may be used in optimizing the three-body data. We present numerical evidence that there exists a variational (minimum) principle for the determination of the three-body binding energy which allows one to choose this parameter also in the absence of an exact reference calculation. Our results for neutron-deuteron scattering show that it is precisely this choice of the parameter which provides optimal scattering data. We conclude that the W-matrix approach, despite its simplicity, is a remarkably efficient tool for high-quality three-nucleon calculations. (orig.)
International Nuclear Information System (INIS)
Three papers are brought together as a result of collaboration between the Nuclear Engineering Program of COPPE-UFRJ (Coordination of Engineering Post graduate Programs of the Federal University of Rio de Janeiro) and the Division of Nuclear Studies of the Department of Thermal generation of FURNAS S.A. aiming at the analysis of the neutronic behavior of PWR power reactors' core. Modifications were introduced in the methods of calculation utilized by the LEOPARD code. The results presented in the first two papers refer to the calculation of neutron flux in the homogenized reactor, only the dependence on energy being considered. Physical models and mathematical approximations are utilized as an alternative to those conventionally used in the code for the calculation of thermal and non-thermal flux. Some parameters, such as the average thermal cross sections of some elements showed to be sensible to the modifications introduced, and indicate that it is useful to carry on the study. In the third paper, comments are made on the MND (Mixed Number Density) method of effectuating the thermal average of the diffusion coefficient and of the absorption cross section, for application in the diffusion equation and consequent determination of flux in function of the spacial position. (I.C.R.)
International Nuclear Information System (INIS)
For Neutron Capture Therapy (NCT) applications, many research reactors are presently utilized. Clinical trials are performed in thermal reactors that have been appropriately modified, in order to obtain convenient beams for Becton (Boron Neutron Capture Therapy), by means of proper filtering or spectrum shifting. However, the beam quality obtainable by fast reactors is expected to be better than that of thermal reactor facilities. Tapiro is a low power, high flux, highly enriched (93.5%) 235Uranium fast reactor. The power is 5 kw and the maximum neutron flux in the core is 3.2'1012 cm-2 s-1. A thermal column and an epithermal one have been designed and constructed, aimed at dosimetry and animal experiments. The configurations of the columns have been designed by means of calculations based on Monte Carlo with the codes MCNP4B and MCNPX2.1.5 together with the DSA (Direct Statistical Approach) variance reduction optimisation patch. The columns have been characterized by means of measurements performed with activation techniques and thermoluminescence and gel dosimeters. Experimental results have shown good consistency with calculations. Moreover, they have confirmed the good quality of the beams obtainable with such a reactor. The TAPIRO reactor (a) and the scheme of the epithermal column (b) are shown. To have further confirmation of the quality of the radiation field in the constructed epithermal column, in-phantom absorbed doses have been measured and profiled by means of gel dosimeters, separating the various dose contributions having different biological effects. An epithermal column for human clinical trials has been designed by means of Monte Carlo calculations and the construction is now in progress. A section of this column is shown and beam parameters are reported. It is evident that the beam quality of this column is satisfactory in comparison with the IAEA recommendations. Moreover, such parameters are good if compared with those available at the
International Nuclear Information System (INIS)
The purpose of work is the comparison of results of calculations of effective neutron doses behind concrete shielding by a method Monte-Carlo and by phenomenological method. Data obtained by these two methods agree within factor 2 over considered range of neutron energies and shielding thickness. Comparison of the results shows that difference in shield thickness between calculated by Monte Carlo and phenomenological method is not exceeded half-value layer for neutron effective dose, that is from 10 cm to 30 cm for considered energies and thickness of shields. (authors)
International Nuclear Information System (INIS)
INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU)
Factors affecting neutron measurements and calculations. Part E. Hydrogen content in granite
International Nuclear Information System (INIS)
For evaluation of radiation doses from the atomic bomb at Hiroshima, many systematic measurements have been made of the residual activities of activation products in rocks and concrete. For the Motoyasu Bridge, which is located close to the bomb hypocenter, the depth profile of 152Eu was measured in a granite core (Hasai et al. 1987; Shizuma et al. 1997). In order to reproduce the depth profile of the activities, it is important to calculate the neutron scattering and absorption (Endo et al. 1999). In this section, the first result of hydrogen analysis by proton-proton elastic recoil coincidence spectrometry for the granite samples is described. (author)
Relativistic Hartree-Bogoliubov Calculation of Specific Heat of the Inner Crust of Neutron Stars
Nakano, Takuya; Matsuzaki, Masayuki
2001-01-01
We calculate the specific heat of the inner crust of neutron stars within a local-density approximation to an improved relativistic Hartree-Bogoliubov theory. Non-uniformness of the system enhances the specific heat in particular at low temperatures. The degree of enhancement is similar to that in the spherical phase of Elgar{\\o}y et al. We examine a schematic interpolation between the results of Broglia et al. adopting the Gogny force and ours based on the Lagrangian of the relativistic mean...
Relativistic Hartree-Bogoliubov calculation of specific heat of the inner crust of neutron stars
International Nuclear Information System (INIS)
We calculate the specific heat of the inner crust of neutron stars within a local-density approximation to an improved relativistic Hartree-Bogoliubov theory. Non-uniformness of the system enhances the specific heat in particular at low temperatures. The degree of enhancement is similar to that in the spherical phase of Elgaroey et al. We examine a schematic interpolation between the results of Broglia et al. adopting the Gogny force and ours based on the Lagrangian of the relativistic mean field model. (author)
Inelastic neutron scattering spectra in f-electron compounds first-principles calculations
Divis, M
2002-01-01
A theoretical investigation of the rare earth (R sup 3 sup +) localized 4f electron states and band energy spectrum of RGa sub 2 intermetallics and RNi sub 2 B sub 2 C borocarbides is presented. To calculate the crystal field interaction, we used a parameter-free first-principles method based on density functional theory. It is shown that a reasonable and parameter-free theoretical description of the inelastic neutron scattering can be achieved, as is demonstrated for the particular cases DyGa sub 2 and NdNi sub 2 B sub 2 C. (orig.)
International Nuclear Information System (INIS)
In this dissertation we use the Laplace transform to derive expressions for nonstandard albedo boundary conditions for one and two non-multiplying regions at the ends of one dimensional domains. In practice, the fuel regions of reactor cores are surrounded by reflector regions that reduce neutron leakage. In order to exclude the reflector regions from the calculations, we introduce a reflection coefficient or albedo. We use the present albedo boundary conditions to solve numerically slab-geometry monoenergetic and multigroup diffusion equations using the conventional finite difference method. Numerical results are generated for fixed source and eigenvalue diffusion problems in slab geometry(author)
Neutronic Calculations of TRIGA MARK-II with WIMS Cluster Options
International Nuclear Information System (INIS)
Neutronic calculations for RTP are made by using WIMS by utilizing several techniques. In this study, we explore the cluster options available in WIMS. In order to use this technique, the RTP core are split into several annulus containing both water and fuel. This enables us to determine the average flux at each annulus. This paper will demonstrate the required input card and general procedure for preparing WIMS input using the cluster option. Comparison of flux and multiplication factor between WIMS and experimental data are made and the amount of error estimated. (author)
Döge, Stefan; Müller, Stefan; Morkel, Christoph; Gutsmiedl, Erwin; Geltenbort, Peter; Lauer, Thorsten; Fierlinger, Peter; Petry, Winfried; Böni, Peter
2015-01-01
We present scattering cross sections $\\sigma_\\text{scatt}$ of ultracold neutrons (UCN) in liquid deuterium at T = 20.6 K, as recently measured by means of a transmission experiment. The indispensable thorough raw data treatment procedure is explained. A calculation model for coherent and incoherent scattering in liquid deuterium in the hydrodynamic limit based on appropriate physical concepts is provided and shown to ?t the data well. The applicability of the incoherent approximation for UCN scattering in liquid deuterium was tested and found to deliver acceptable results.
Synergism of the method of characteristics and CAD technology for neutron transport calculation
International Nuclear Information System (INIS)
The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, a new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)
Dose calculation for neutrons of thermal to 10 MeV
International Nuclear Information System (INIS)
In the ICRP publication 60 adopted in 1990, the drastic change was proposed regarding the definition of the dose used for radiation protection. The main changes were the introduction of radiation weighting factor, the definition of tissue equivalent dose, the change of tissue weighting factor and the change of the equation for defining radiation quality factor (the equation for Q-L relation). In the exposure to neutrons, all these changes exert influence. In the case of neutrons, the conservativeness of operational quantity in relation to the limit of exposure may break down. Effective dose is defined as the sum of weighted risks of the equivalent doses of 12 organs and tissues and the rest of tissues. Also it has been recommended to take the mass-weighted average value for 10 specified organs and tissues as the equivalent dose. As to the effective dose for neutrons, the calculating method, the effective dose for adults, the comparison of effective dose and effective dose equivalent, and the age dependence of effective dose are explained. It is difficult to directly measure effective dose which is the limiting quantity of exposure. Therefore, ICRU defined operational quantity for area monitoring and individual monitoring. The relation of effective dose with operational quantity is shown. (K.I.)
Calculation of gamma-rays and fast neutrons fluxes with the program Mercure-4
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The program MERCURE-4 evaluates gamma ray or fast neutron attenuation, through laminated or bulky three-dimensionnal shields. The method used is that of line of sight point attenuation kernel, the scattered rays being taken into account by means of build-up factors for γ and removal cross sections for fast neutrons. The integration of the point kernel over the range of sources distributed in space and energy, is performed by the Monte-Carlo method, with an automatic adjustment of the importance functions. Since it is operationnal the program MERCURE-4 has been intensively used for many various problems, for example: - the calculation of gamma heating in reactor cores, control rods and shielding screens, as well as in experimental devices and irradiation loops; - the evaluation of fast neutron fluxes and corresponding damage in structural materials of reactors (vessel steels...); - the estimation of gamma dose rates on nuclear instrumentation in the reactors, around the reactor circuits and around spent fuel shipping casks
Improved calculation of the prompt fission neutron spectrum from the spontaneous fission of 252Cf
International Nuclear Information System (INIS)
An improved calculation is presented for the prompt fission neutron spectrum N(E) from the spontaneous fission of 252Cf. In this calculation the fission-spectrum model of Madland and Nix is used, but with several improvements leading to a physically more accurate representation of the spectrum. Specifically, the contributions to N(E) from the entire fission-fragment mass and charge distributions will be calculated instead of calculating on the basis of a seven-point approximation to the peaks of these distributions as has been done in the past. Therefore, values of the energy release in fission, fission-fragment kinetic energy, nuclear level density, and compound nucleus cross section for the inverse process will be considered on a point-by-point basis over the fragment yield distributions instead of considering averages of these quantities over the peaks of the distributions. Particular attention will be given to the energy-dependent compound nucleus cross sections and to the nuclear level density model. Other refinements to the calculation of N(E) will also be discussed. Results will be presented and compared with earlier calculations of the spectrum and with recent experimental measurements of the spectrum. 9 figs
Implementation of the neutronics model of HEXTRAN/HEXBU-3D into APROS for WWER calculations
International Nuclear Information System (INIS)
A new three-dimensional nodal model for neutronics calculation is currently under implementation into APROS - Advanced PROcess Simulation environment - to conform the increasing accuracy requirements. The new model is based on an advanced nodal code HEXTRAN and its static version HEXBU-3D by VTT, Technical Research Centre of Finland. Currently the new APROS is under a testing programme. Later a systematic validation will be performed. In the first phase, a goal is to obtain a fully validated model for VVER-440 calculations. Thus, all the current test calculations are performed by using Loviisa NPP's VVER-440 model of APROS. In future, the model is planned to be applied for the calculations of VVER-1000 type reactors as well as in rectangular fuel geometry. The paper outlines first the general aspects of the method, and then the current situation of the implementation. Because of the identical model with the models of HEXTRAN and HEXBU-3D, the results in the test calculations are compared to the results of those. In the paper, results of two static test calculations are shown. Currently the model works well already in static analyses. Only minor problems with the control assemblies of VVER-440 type reactor still exist but the reasons are known and will be corrected in near future. Dynamical characteristics of the model are up to now tested only by some empirical tests. (author)
International Nuclear Information System (INIS)
The last 12 years studies about the CABRI, SCARABEE and PHEBUS projects are summarized. It describes the object and the genesis of the cores, the evolution of the core concept and the associated neutronic problems. The calculational scheme used is presented, together with its qualification. The formalism, and the qualification of the different modules of GOLEM are presented. COXYS: module of physical analysis in order to determine the best energetic and spatial mesh for the case of interest. GOLU.B: input data management module. VAREC: calculation module of perturbations due to materials enables to compute perturbed flux and reactivity variation. VARYX: calculation module of geometric perturbations. TRACASYN: module of 3D power shape calculation. Finally TRACASTORE: module of management and graphic exploitation of results. Then, one gives utilization directions for these different modules. Qualification results show that GOLEM is able to analyse the fine physics of many various cases, to calculate by perturbation effects greater than 5000 pcm, to rebuild perturbed flux with margins near 3% for difficult situations, like reactor voiding or spectral or spectral variation in a PWR. Furthermore, 3D hot spots are calculated within margins of a magnitude comparable to experimental ones
Large-scale Monte Carlo neutron transport calculations with thermal hydraulic feedback
International Nuclear Information System (INIS)
Highlights: • Method of internal coupling, based on dynamic material distribution, is presented. • The Wielandt shift method is implemented to accelerate Mote Carlo calculations. • The Uniform Fission Site method is introduced for tallies with large numbers of bins. • The stochastic approximation scheme is used to stabilize coupled code convergence. - Abstract: The Monte Carlo method provides the most accurate description of the particle transport problem. The criticality problem is simulated by following the histories of individual particles without approximating the energy, angle or the coordinate dependence. These calculations are usually done using homogeneous thermal hydraulic conditions. This is a very crude approximation in the general case. In this paper, the method of internal coupling between neutron transport and thermal hydraulics is presented. The method is based on dynamic material distribution, where coordinate dependent temperature and density information is supplied on the fly during the transport calculation. This method does not suffer from the deficiencies characteristic of the external coupling via the input files. In latter case, the geometry is split into multiple cells having distinct temperatures and densities to supply the feedback. The possibility to efficiently simulate large scale geometries at pin-by-pin and subchannel level resolution was investigated. The Wielandt shift method for reducing the dominance ratio of the system and accelerating the fission source convergence was implemented. During the coupled iteration a detailed distribution of the fission heat deposition is required by the thermal hydraulics calculation. Providing reasonable statistical uncertainties for tallies having large numbers of bins, is a complicated task. This problem was resolved by applying the Uniform Fission Site method. Previous investigations showed that the convergence of the coupled neutron transport/thermal hydraulics calculation is limited by
Energy Technology Data Exchange (ETDEWEB)
NONE
2000-05-01
Knowing the quantities of certain substances discharged into the atmosphere is a necessary and fundamental stage in any environmental protection policy to tackle today's problems such as acid rain, the degradation of air quality, global warming and climate change, the depletion of the ozone layer, etc. This quantification, usually known as an 'emission inventory', is built on a set of specific rules which may vary from one inventory to another. This state of affairs presents the enormous disadvantage that the data available are not comparable. At the international level, an attempt at harmonization has been going on for some years between the various international bodies. This work is being pursued in parallel with the improvement of methodologies to estimate discharges from various types of source. To take account of changes in specifications and of improvements in our understanding of phenomena giving rise to atmospheric pollution, the results of inventories of emissions need to be regularly revised, even retrospectively, to maintain a consistent series. CITEPA, which acts as a National Reference Centre, has developed a system of inventories as part of the CORALIE programme with financial help from the French Ministry for Planning and the Environment. (author)
Geng, Changran; Tang, Xiaobin; Guan, Fada; Johns, Jesse; Vasudevan, Latha; Gong, Chunhui; Shu, Diyun; Chen, Da
2016-03-01
The purpose of this study is to verify the feasibility of applying GEANT4 (version 10.01) in neutron dose calculations in radiation protection by comparing the calculation results with MCNP5. The depth dose distributions are investigated in a homogeneous phantom, and the fluence-to-dose conversion coefficients are calculated for different organs in the Chinese hybrid male phantom for neutrons with energy ranging from 1 × 10(-9) to 10 MeV. By comparing the simulation results between GEANT4 and MCNP5, it is shown that using the high-precision (HP) neutron physics list, GEANT4 produces the closest simulation results to MCNP5. However, differences could be observed when the neutron energy is lower than 1 × 10(-6) MeV. Activating the thermal scattering with an S matrix correction in GEANT4 with HP and MCNP5 in thermal energy range can reduce the difference between these two codes. PMID:26156875
International Nuclear Information System (INIS)
The purpose of this study is to verify the feasibility of applying GEANT4 (version 10.01) in neutron dose calculations in radiation protection by comparing the calculation results with MCNP5. The depth dose distributions are investigated in a homogeneous phantom, and the fluence-to-dose conversion coefficients are calculated for different organs in the Chinese hybrid male phantom for neutrons with energy ranging from 1 x 10-9 to 10 MeV. By comparing the simulation results between GEANT4 and MCNP5, it is shown that using the high-precision (HP) neutron physics list, GEANT4 produces the closest simulation results to MCNP5. However, differences could be observed when the neutron energy is lower than 1 x 10-6 MeV. Activating the thermal scattering with an S matrix correction in GEANT4 with HP and MCNP5 in thermal energy range can reduce the difference between these two codes. (authors)
Energy Technology Data Exchange (ETDEWEB)
Faik Ouahab, Z.; Jehouani, A.; Ghassoun, J.; Senhou, N. [EPRA, Departement de Physique, Faculte de Sciences Semlalia, BP. 2390, Universite Cadi Ayyad, Marrakech (Morocco); Groetz, J.E. [Universite de Franche-Comte, Laboratoire de Chimie Physique et Rayonnements Alain Chambaudet, UMR CEA E4, 16, route de Gray, 25030 Besancon Cedex (France)
2010-07-01
Summary of a study of assessment of the probability for neutrons to be guided in a full duct with a square cross section and doubly bent. Two software have been developed, based on the Monte Carlo simulation, to compute the neutron transmission probability at the end of the duct. Results are in good agreement with that obtained with the MCNP-5 code. The neutron flow and probability at the duct end have been determined for different materials and different duct dimensions
Energy Technology Data Exchange (ETDEWEB)
Rassow, J.; Poeller, F.; Meissner, P. (Essen Univ. (Gesamthochschule) (Germany). Abt. fuer Medizinische Strahlenphysik); Steinberg, F. (Essen Univ. (Gesamthochschule) (Germany). Inst. fuer Medizinische Strahlenbiologie)
1993-01-01
Fundamentally different aspects apply to dosage in boron neutron capture therapy (BNCT) compared to that in the case of normal radiotherapy with photons, electrons or heavy particles such as neutrons. The reason is that the latter only requires a knowledge of the stochastic distribution of the absorbed dose within cells, radiation quality and atomic composition of tissue in the regions of interest, whereas for the former the absolute concentration and microscopic distribution of [sup 10]B atoms in inter- and intracellular spaces of tumor and healthy cells is additionally of equal importance. The effects of radiation without [sup 10]B must always be superimposed on those of heavy particles resulting from neutron capture reactions on [sup 10]B atoms. Complex geometrical calculaations are necessary with respect to ranges of the heavy particles smaller than a cell diameter. Apart from the direct effects of radiation without [sup 10]B, the dosage therefore depends on thermal neutron fluence, [sup 10]B concentration, its extreme inhomogeneous macroscopic distribution in the tumor tissue, the cellular localization of the [sup 10]B atoms in the large intercellular space, the cell membrane, within cytoplasm or the cell nucleus, the geometrical probability of hitting the cell nucleus, and that such a hit finally results in a cell killing, and a Poisson statistical enhancement factor, which describes the dose-effect relation for cell survival. The calculations necessary are demonstrated in the case of a normal and a tumor cell type, each with representative cell diameter and nucleus size. It is evident that the microscopic distribution of [sup 10]B atoms is one of the most critical parameters which is still insufficiently known. (orig.).
International Nuclear Information System (INIS)
The calculation of neutron fluxes and fluence into reactor pressure vessel is a regulatory requirement in the stages of the design, operation and plan lifetime extension. Being the reactor vessel a part of the primary circuit, its integrity should be preserved under all operation regimes. The reactor vessel is considered a unique and non-substitutable part of the NPP that undergoes degradation. The main source of the aging comes from the fast neutron damage induced in the steel crystalline lattice. In the case of the WWER-type reactors, the vessel fragilization has been identified as one of the main problems concerning the safety of NPPs. Due to the proximity of the core edge to the vessel inner surface; the vessel steel is exposed to high fast neutron fluence. The effect of this irradiation on the mechanical properties becomes more acute because of the impurities measured in the current Russian steel alloys. In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that on the other hand need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons as well as the efficiency increasing options included in the code are studied in order to identify their impact in the final results for fluxes and fluence in the reactor pressure vessel. As a result an optimal set of parameters is suggested
Codes complex for quick transport 3D neutron calculations of WWER
International Nuclear Information System (INIS)
Complex Surface Values System that capable to carry out the all stages of neutron physical stationary calculations for reactors WWER and PWR is presented. This complex based oneself on using the Surface Values Methods. The approach consists in maximum possible account of specific features in reactor calculations. There is a certain amount of specific features in reactor problems due attention to which is crucial for attaining the main goal-organizing of reactor code.s complexes those provide economical and safe reactors' operating. It is important for the estimation of a code quality to fix of the methodical component but it is necessary to have in view of scale of other ones of result uncertainties. We mention in passing it afterwards. Complex Surface Values System present the approach of building computational reactor models that account for the specifics of reactor problems outlined above. This approach is called the methods of surface values. It utilizes method of surface pseudo-sources for calculating cells within active cores and method of surface harmonics for calculating the whole core or certain sub-assemblies that contain several cells. The part of total complex - the complex Surface Values Lattices - is destined for cells, assembly and sub-assembly calculations of thermal water reactors. The complex Surface Values Lattices use micro constant libraries prepared in the format WIMS. There are libraries in such format based on different initial files of valued data (ENDBF, JEFF, UKNDL, JENDL) on site IAEA. We used these libraries for comparing and choused the one similar to recommended IAEA library. The code Surface Values Core is contained in the total complex Surface Values System besides the complex Surface Values Lattice. This code provide the total scale calculations of reactor's cores. The equations of the Surface Harmonics Method are suitable well for inside reactor core. Other approximations are necessary for neutron description inside reflector
International Nuclear Information System (INIS)
Computer program with FORTRAN has been made for calculating activity correction by neutron incident on the edge of the foil. The calculation based on single on collision theory with the assumption of monoenergetic neutron and isotopic distributed. the combination of simpson rule 1/3 and gauss quadrature were chosen to solve the problems and discrete summation was used as approximation of integration to the whole of energy groups (640 groups). the inputs are dimension and mass of the foil and activation cross-section and spectrum neutron. the output are foil activity and activation correction by neutron incident on the edge of the foil. the calculation results of activity correction by neutron incident on the edges of the gold foils of 0.05383 mm and 1.27 mm thick are about 0.2% and 7.8% and for cobalt foils of 0.1128 mm and 1.128 mm thick are about 0.58% and 6.7% respectively. the discrepancy of foil activation between experiment and calculation are about 1.6% for gold foil of 0.05383 mm thick and 1% for cobalt foil of 0.1128 mm thick. from that results can be concluded that the calculation result quit close to the experiment one and the thicker foil give bigger activation correction by neutrons incident on the edges of the foil
Energy Technology Data Exchange (ETDEWEB)
Gerasimenko, B.F. [V.G. Khlopin Radium Inst., Saint Peterburg (Russian Federation)
1997-03-01
The calculations of integral spectra of prompt neutrons of spontaneous fission of {sup 244}Cm and {sup 246}Cm were carried out. The calculations were done by the Statistical Computer Code Complex SCOFIN applying the Hauser-Feschbach method as applied to the description of the de-excitation of excited fission fragments by means of neutron emission. The emission of dipole gamma-quanta from these fragments was considered as a competing process. The average excitation energy of a fragment was calculated by two-spheroidal model of tangent fragments. The density of levels in an excited fragment was calculated by the Fermi-gas model. The quite satisfactory agreement was reached between theoretical and experimental results obtained in frames of Project measurements. The calculated values of average multiplicities of neutron number were 2,746 for {sup 244}Cm and 2,927 for {sup 246}Cm that was in a good accordance with published experimental figures. (author)
International Nuclear Information System (INIS)
There are several new technological application fields of fast neutrons such as accelerator-driven incineration/ transmutation of the long-lived radioactive nuclear wastes (in particular transuranium nuclides) to short-lived or stable isotopes by secondary spallation neutrons produced by high-intensity, intermediate-energy, charged-particle beams, prolonged planetary space missions, shielding for particle accelerators. Especially, accelerator driven subcritical systems (ADS) can be used for fission energy production and /or nuclear waste transmutation as well as in the intermediate-energy accelerator driven neutron sources, ions and neutrons with energies beyond 20 MeV, the upper limit of exiting data files that produced for fusion and fission applications. In these systems, the neutron scattering cross sections and emission differential data are very important for reactor neutronics calculations. The transition rate calculation involves the introduction of the parameter of mean free path determines the mean free path of the nucleon in the nuclear matter. This parameter allows an increase in mean free path, with simulation of effect, which is not considered in the calculations, such as conservation of parity and angular momentum in intra nuclear transitions. In this study, we have investigated the multiple preequilibrium matrix element constant from internal transition for Uranium, Thorium, (n,xn) neutron emission spectra. The neutron-emission spectra produced by (n,xn) reactions on nuclei of some target (for spallation) have been calculated. In the calculations, we have used the geometry dependent hybrid model and the cascade exciton model including the effects of the preequilibrium. The pre-equilibrium direct effects have been examined by using full exciton model. All calculated results have been compared with the experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other
Neutron activation analysis application to the study of air pollution bio monitors
International Nuclear Information System (INIS)
Full text: This work has been done within the IAEA Research Contract Arg 9929, Research Co-ordinated Programme on Validation and application of plants as bio monitors of trace-element atmospheric pollution, analysed using nuclear and related techniques. Knowledge on air pollution levels and identification of polluted areas and potential emission sources are of increasing concern all over the world. Chemical characterisation of atmospheric aerosol, especially its heavy metal contents, is therefore of great importance and neutron activation analysis is a powerful technique for its determination. The advantages of using bio monitors instead of direct sampling lies not only on its lower cost but also on the possibility of using them to measure and/or evaluate deposition over large areas. The general objective of this project is the use of lichen to evaluate pollution levels in an area of Cordoba province (Argentina) and to establish baseline levels and temporal trends and draw distribution maps of pollutants. Based on lichen distribution maps, two species were selected: Raumalina ecklonii and Usnea amblyoclada. Different tests were done to adjust sample preparation methodologies previous to irradiation. The tests included grinding and drying assays to investigate their influence on the following determination using NAA. Sample grinding with and without the addition of liquid nitrogen was tried and oven-dry and freeze-dry were tried on samples of the two selected species. Elemental determination was done using instrumental Neutron Activation Analysis. Samples were irradiated for 5 hours at the RA-3 reactor of the Ezeiza Atomic Center (thermal flux 3.1013cm-2-2.s-1-1, 4.5 M w), and measured twice with different decay times 86 and 30 days) for the determination of medium and long-lived nuclides. The measurements were done using GeHP detectors (30 % efficiency, resolution 1.9 keV for 6060Co 1332.5 keV peak) coupled to a Canberra Series 85 multichannel analyser
Neutronics calculations relevant to the conversion of research reactors to low-enriched fuel
International Nuclear Information System (INIS)
The new spirit and urgency of converting the remaining research reactors from highly enriched uranium (HEU) to low-enriched fuel, combined with the prospects of new ultra-high-density fuels, provides the main impetus and defines the basic scientific objectives for this thesis. It is predictable that activities to convert existing research reactors will intensify in the near-term future, which in turn would simultaneously increase the need for corresponding neutronics calculations. Here, especially the analysis of the remaining high-flux reactors, which are most difficult to convert due to compact core geometries, may benefit from high-precision simulation tools to adequately set-up and study reactor parameters using complete three-dimensional core models. The scope of the present thesis is to support this process in providing a new computational tool for neutronics calculations (M3O), which is based on standard physics codes, while using the technical computing environment Mathematica as the primary user-interface. The use of such modern environments can be very convenient for a variety of reasons: their analytical capabilities allow for a broad range of calculations and data manipulation, while their interactive graphical user-interface facilitates intensive control of input parameters and interpretation of achieved results. At the same time, Monte Carlo methods play an increasing role in neutron transport and burnup analyses. In M3O, the Monte Carlo code MCNP is employed, which offers the potential for high-precision modeling and analysis. Both major components, Mathematica and MCNP, are also used in an optimization tool developed below and based on the linear programming technique to optimize reactor performance by variation of the fundamental core parameters. The potential (and limits) of monolithic fuels is largely unknown today. Even though the conversion of a large number of medium-flux reactors would be relatively straightforward, the performance of
Air and smear sample calculational tool for Fluor Hanford Radiological control
International Nuclear Information System (INIS)
A spreadsheet calculation tool was developed to automate the calculations performed for determining the concentration of airborne radioactivity and smear counting as outlined in HNF--13536, Section 5.2.7, ''Analyzing Air and Smear Samples''. This document reports on the design and testing of the calculation tool. Radiological Control Technicians (RCTs) will save time and reduce hand written and calculation errors by using an electronic form for documenting and calculating work place air samples. Current expectations are RCTs will perform an air sample and collect the filter or perform a smear for surface contamination. RCTs will then survey the filter for gross alpha and beta/gamma radioactivity and with the gross counts utilize either hand calculation method or a calculator to determine activity on the filter. The electronic form will allow the RCT with a few key strokes to document the individual's name, payroll, gross counts, instrument identifiers; produce an error free record. This productivity gain is realized by the enhanced ability to perform mathematical calculations electronically (reducing errors) and at the same time, documenting the air sample
International Nuclear Information System (INIS)
Two-dimensional radiation transport methods have been used to estimate the effects of neutron and gamma-ray streaming on the performance of the Engineering Test Facility neutral beam injectors. The calculations take into account the spatial, angular, and spectral distributions of the radiation entering the injector duct. The instantaneous nuclear heating rate averaged over the length of the cryopumping panel in the injector is 7.5 X 10-3 MW/m3, which implies a total heat load of 2.2 X 10-4 MW. The instantaneous dose rate to the ion gun insulators was estimated to be 3200 rad/s. The radial dependence of the instantaneous dose equivalent rate in the neutral beam injector duct shield was also calculated
Neutron and gamma ray streaming calculations for the ETF neutral beam injectors
Lillie, R. A.; Santoro, R. T.; Alsmiller, R. G., Jr.; Barnes, J. M.
1981-02-01
Two dimensional radiation transport methods were used to estimate the effects of neutron and gamma ray streaming on the performance of the engineering test facility neutral beam injectors. The calculations take into account the spatial, angular, and spectral distributions of the radiation entering the injector duct. The instantaneous nuclear heating rate averaged over the length of the cryopumping panel in the injector is 7.5 x 10(+3) MW/m(3) which implies a total heat load of 2.2 x 10(+4) MW. The instantaneous dose rate to the ion gun insulators was estimated to be 3200 rad/s. The radial dependence of the instantaneous dose equivalent rate in the neutral beam injector duct shield was also calculated.
Neutron and gamma ray streaming calculations for the ETF neutral beam injectors
International Nuclear Information System (INIS)
Two-dimensional radiation transport methods have been used to estimate the effects of neutron and gamma ray streaming on the performance of the Engineering Test Facility (ETF) neutral beam injectors. The calculations take into account the spatial, angular, and spectral distributions of the radiation entering the injector duct. The instantaneous nuclear heating rate averaged over the length of the cryopumping panel in the injector is 7.5 x 10-3 MW/m3 which implies a total heat load of 2.2 x 10-4 MW. The instantaneous dose rate to the ion gun insulators was estimated to be 3200 rad/s. The radial dependence of the instantaneous dose equivalent rate in the neutral beam injector duct shield was also calculated
Energy Technology Data Exchange (ETDEWEB)
Waldo, R.W.
1980-05-01
Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of /sup 232/U, /sup 237/Np, /sup 238/Pu, /sup 241/Am, /sup 242m/Am, /sup 245/Cm, and /sup 249/Cf were studied for the first time. The delayed neutron emission from /sup 232/Th, /sup 233/U, /sup 235/U, /sup 238/U, /sup 239/Pu, /sup 241/Pu, and /sup 242/Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from /sup 232/Th to /sup 252/Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables.
Energy Technology Data Exchange (ETDEWEB)
Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-05-01
The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)
International Nuclear Information System (INIS)
A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for 3He, H2, or BF3 gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of 3He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the 3He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF3 counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations
Energy Technology Data Exchange (ETDEWEB)
Takeda, N. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Kudo, K. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Toyokawa, H. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Torii, T. [Japan Power Reactor and Nuclear Fuel Development Corporation, Tsuruga Office, Fukui 919-12 (Japan); Hashimoto, M. [Japan Power Reactor and Nuclear Fuel Development Corporation, O-arai Engineering Center, Ibaraki 311-13 (Japan); Sugita, T. [Science System Laboratory, Ibaraki 309-17 (Japan); Dietze, G. [Physikalisch-Technische Bundesanstalt, 38023 Braunschweig (Germany); Yang, X. [China Institute of Atomic Energy (China)
1999-02-11
A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for {sup 3}He, H{sub 2}, or BF{sub 3} gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of {sup 3}He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the {sup 3}He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF{sub 3} counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations.
Monte Carlo transport calculations and analysis for reactor pressure vessel neutron fluence
International Nuclear Information System (INIS)
The application of Monte Carlo methods for reactor pressure vessel (RPV) neutron fluence calculations is examined. As many commercial nuclear light water reactors approach the end of their design lifetime, it is of great consequence that reactor operators and regulators be able to characterize the structural integrity of the RPV accurately for financial reasons, as well as safety reasons, due to the possibility of plant life extensions. The Monte Carlo method, which offers explicit three-dimensional geometric representation and continuous energy and angular simulation, is well suited for this task. A model of the Three Mile Island unit 1 reactor is presented for determination of RPV fluence; Monte Carlo (MCNP) and deterministic (DORT) results are compared for this application; and numerous issues related to performing these calculations are examined. Synthesized three-dimensional deterministic models are observed to produce results that are comparable to those of Monte Carlo methods, provided the two methods utilize the same cross-section libraries. Continuous energy Monte Carlo methods are shown to predict more (15 to 20%) high-energy neutrons in the RPV than deterministic methods
International Nuclear Information System (INIS)
A relatively simple formalism for calculating the average neutron elastic angular distribution dσel/dΩ in the resonance region below several hundred keV is presented. The expression for dσel/dΩ depends mainly on the R-matrix parameters S0, R', S1, and R1∞. Comparisons between calculated and experimental angular distributions are presented for 103Rh, 139La, 232Th, and 238U. A fit to 238U data at 75 keV led to a value of the p-wave strength function of S1=1.81±0.35x10-4. Except for measuring a complete set of individual l=1 resonances, determining the p-wave strength function by fitting low-energy angular distributions is probably more reliable than, or competitive with, other techniques which are available. An analysis of elastic angular distributions as a function of neutron energy is also well suited to a search for intermediate structure in the s- or p-wave strength function. copyright 1997 The American Physical Society
International Nuclear Information System (INIS)
Intense neutron fluxes within fusion reactors that are currently being designed will lead to the activation of structural components, and to assess and minimize this radioactivity, nuclear cross sections are needed for neutrons with energies up to 20 MeV. We describe research performed for the International Atomic Energy Agency (IAEA) Coordinated Research Programme on activation cross sections for fusion reactor technology, which has selected certain high-priority reactions for both experimental and theoretical study. Using statistical model codes, we have investigated excitation function cross sections for radionuclide production in the reactions 94Mo(n,p)94Nb, 109Ag(n,2n)108mAg, 151Eu(n,2n)150m Eu, 153Eu(n,2n)152g+m2Eu, 159Tb(n,2n)158Tb, 187Re(n,2n)186mRe, 179Hf(n,2n)178m2Hf, 193Ir(n,2n)192m2Ir. Using our calculated results for the excitation functions, along with calculations by other groups, the theoretical excitation functions have been normalized to experimental values at 14.5 MeV to produce evaluated excitation functions. These evaluations can be used within radiation transport and nuclide inventory codes to design, and assess the environmental impact of, fusion reactors. 23 refs., 4 figs., 1 tab
International Nuclear Information System (INIS)
The reliability of calculation tools to evaluate and calculate dose rates appearing behind multi-layered shields is important with regard to the certification of transport and storage casks. Actual benchmark databases like SINBAD do not offer such configurations because they were developed for reactor and accelerator purposes. Due to this, a bench-mark-suite based on own experiments that contain dose rates measured in different distances and levels from a transport and storage cask and on a public benchmark to validate Monte-Carlo-transport-codes has been developed. The analysed and summarised experiments include a 60Co point-source located in a cylindrical cask, a 252Cf line-source shielded by iron and polyethylene (PE) and a bare 252Cf source moderated by PE in a concrete-labyrinth with different inserted shielding materials to quantify neutron streaming effects on measured dose rates. In detail not only MCNPTM (version 5.1.6) but also MAVRIC, included in the SCALE 6.1 package, have been compared for photon and neutron transport. Aiming at low deviations between calculation and measurement requires precise source term specification and exact measurements of the dose rates which have been evaluated carefully including known uncertainties. In MAVRIC different source-descriptions with respect to the group-structure of the nuclear data library are analysed for the calculation of gamma dose rates because the energy lines of 60Co can only be modelled in groups. In total the comparison shows that MCNPTM fits very wall to the measurements within up to two standard deviations and that MAVRIC behaves similarly under the prerequisite that the source-model can be optimized. (author)
Directory of Open Access Journals (Sweden)
KOHIO Niéssan
2014-12-01
Full Text Available In this paper we calculate the transport coefficients of plasmas formed by air and water vapor mixtures. The calculation, which assume local thermodynamic equilibrium (LTE are performed in the temperature range from 500 to 12000 K. We use the Gibbs free energy minimization method to determine the equilibrium composition of the plasmas, which is necessary to calculate the transport coefficients. We use the Chapman-Enskog method to calculate the transport coefficients. The results are presented and discussed according to the rate of water vapor. The results of the total thermal conductivity and electrical conductivity show in particular that the increasing of the rate of water vapor in air can be interesting for power cut. This could be improve the performance of plasma during current breaking in air contaminate by the water vapor.
Characterisation of air particulate matter in Klang Valley by neutron activation analysis technique
International Nuclear Information System (INIS)
Air particulate matter is known to affect human health, impairs visibility and can cause climate change. Study on air particulate matter in term of particle size and chemical contents is very important to indicate the quality of air in a sampling area. Information on concentration of important constituents in air particles can be used to identify some of emission sources which contribute to the pollution problem. The data collected may also be, used as a basis to design a strategy in order to overcome the air pollution problem in the area. The study involved sampling of air dust at two stations, one in Bangi and the other in Kuala Lumpur using Gent Stack Sampler units. Each sampler capable of collecting air particle sizes smaller than 2.5 micron (PM 2.5) and between 2.5 - O micron on two different filters simultaneously. The filters were measured for their mass, elemental carbon and elemental concentrations using analytical equipment or techniques including reflectometer and Neutron Activation Analysis. The results of analysis on samples collected in 1997-1998 are discussed. (author)
International Nuclear Information System (INIS)
Measured and calculated neutron and gamma-ray energy spectra resulting from the transport of approx. 14 MeV neutrons through a 0.30-m-thick lithium hydride slab and through a 0.05-m-thick lead slab followed by 0.30 m of lithium hydride are compared. Also reported are comparisons of the measured and calculated neutron energy spectra behind an 0.80-m-thick assembly comprised of stainless steel type 304 and borated polyethylene. The spatial dependence of the gamma-ray energy deposition rate measured using thermoluminescent detectors is compared with calculated data. The calculated data obtained using two-dimensional radiation transport methods and ENDF/B-IV cross section data are in good agreement for all of the experimental configurations
Energy Technology Data Exchange (ETDEWEB)
Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)
2004-03-01
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)
International Nuclear Information System (INIS)
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H2O, liquid He, liquid D2O, liquid and solid H2 and D2, solid CH4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common SN-transport codes and the Monte Carlo Code MCNP. (orig.)
Comparison of big event with calculations of the air shower development
Niwa, M.; Misaki, A.; Matano, T.
1985-01-01
The incidence of high energy hadrons and electron-photons in air showers at various stages of development is calculated. Numerical calculation is used to solve the diffusion equation for a nuclear cascade and analytical calculation for cascade shower induced gamma rays. From these calculations, one can get the longitudinal development of the high energy hadron and electron-photon components, and the energy spectra of these components at various depths of air shower development. The total number of hadrons (N sub H) and electron-photon components (N sub gamma) are related according to stages of the air shower development and primary energy. The relation of the total energy of hadron and electron-photon component above the threshold energy is given. The energy balance between both components is also a useful parameter to study high energy events accompanying air showers. The relation of N sub H and fractional hadronic energy E (sum E sub H sup gamma/sum E sub H sup gamma + Sum E sub gamma) is calculated. This relation is helpful to understand the stage of air shower development(t) and primary energy (E sub p).
International Nuclear Information System (INIS)
According to Russian federal norms and the safety guide of the nuclear regulatory body of Russia, the maximum fast neutron fluence above 0.5 MeV at critical positions of the reactor pressure vessel (RPV) of VVER-type reactors is used for prediction of the RPV lifetime. For the computation of neutron fluences in the RPV near the reactor core midplane level, the three-dimensional (3-D) synthesis method based on two- and one-dimensional SN calculations may be acceptable but needs validation. The present validation analysis was carried out on the basis of neutron transport calculations for a VVER-1000 model by means of the well-known codes DORT (R, Θ- and R, Z geometry) and ANISN (R geometry) using the multigroup library BUGLE-96. The 3-D spatial neutron source distribution, including pin-to-pin power variations and the complex baffle construction, were modeled in detail
International Nuclear Information System (INIS)
The mathematical simulation technique used for calculating the photoneutron yield from thick targets made of different Materials is suggested. Cascade-evaporative nucleus model being a part of the IMITATOR program complex is used for calculations. Three groups of materials are investigated: light-oxygen and aluminium, medium- iron and nickel, heavy,tungsten and lead. Maximum thickness of targets consistuting 10 radiation length is determined on the basis of the experiment and represents the thickness at which in the investigated energy range secondary neutron flux ''saturation'' arises. The dependences of total neutron yield on electron beam energy and target material are obtained. The values of fast neutrons yield from thick targets, their spatial distribution and dependences on the energy of primary electrons and target thickness are determined. Anomalies of photoneutrons yield near magic and double magic nuclei are pointed out. A considerable drop of total yield of fast neutrons with increase of atomic number of target material is noted
Abdushukurov, D A; Muminov, Kh Kh; Chistyakov, D Yu
2007-01-01
We consider the results of modeling of the efficiency of registration of thermal neutrons by the converters, which are made from natural gadolinium and its 157 isotope foils. Efficiency for a case of falling of neutrons under various angles to a plane of converters is calculated. It is shown, that at small angles of falling of neutrons to a plane of converters it is possible to receive the efficiency of registration close to a theoretical limit. Efficiency of the complex converter made of kapton supporting film with gadolinium converters layered on both its sides is considered. All calculations are carried out for four fixed neutron energies, which correspond to the wavelengths of 1, 1.8, 3 and 4 $\\AA$.
Pritychenko, B.; Mughabghab, S.F.
2012-01-01
We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-...
International Nuclear Information System (INIS)
A major task of international nuclear material safeguards consists in the experimental determination of the characteristics of spent nuclear fuel. In this respect, passive neutron techniques take a prominent place. The paper describes both the actinide content and the passive neutron emission of spent WWER-440 fuel for a number of operational conditions. For validation purposes the results have been compared with those independently gained by other authors for the same type of fuel. The calculations are the basis for a true to fact interpretation of the primary measurement signal, i.e. for the inference from recorded neutron radiation regarding the properties of the spent nuclear fuel. (author)
International Nuclear Information System (INIS)
The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)
Recent developments in fast neutron radiography for the interrogation of air cargo containers
International Nuclear Information System (INIS)
There is a worldwide need for improved methods for the scanning of consolidated air cargo for contraband such as illicit drugs and explosives. Ideally, cargo containers must be imaged without unpacking and with scan times of less than a few minutes. Fast neutron radiography techniques are particularly attractive for screening cargo. Neutrons have the required penetration, they interact with matter in a manner complementary to X-rays and they can be used to determine cargo composition. The Commonwealth Science and Industrial Research Organisation (CSIRO) has developed a scanner for fully-loaded air cargo containers. The scanner combines fast (14 MeV) neutron and γ-ray (or X-ray) radiography, using intense radiation sources and custom high-efficiency detector arrays. The ratio of the transmissions of neutrons and X-rays provides a measure of material composition that is much more sensitive than alternative dual high-energy (MeV) X-ray systems. A full-scale prototype scanner was used by Australian Customs Service to screen incoming air cargo at Brisbane International Airport in 2005/6. The trial of the scanner at Brisbane demonstrated the material discrimination capability of the technology and its ability to make hidden organic materials more obvious. Consolidated cargo was scanned in less than two minutes allowing high volumes of cargo to be screened rapidly. CSIRO is working directly with Nuctech Company Limited, Beijing, China to develop and commercialise the next generation in air cargo scanning technology. A commercial version of the airport scanner being developed by Nuctech and CSIRO is expected to be commissioned by January 2009. The commercial scanner combines a 14 MeV fast neutron radiography system with Nuctech's dual-energy X-ray technology that uses a 6 MeV LINAC X-ray source and Binocular Stereoscopic imaging technology. The commercial scanner will have much better spatial resolution than the Brisbane scanner. The improved resolution, combined with
Applied research on air pollution in Korea using instrumental neutron activation analysis
International Nuclear Information System (INIS)
The aim of this study is to support the use of nuclear analytical technique for applied research and monitoring studies on air pollution and to make an interpretation of the data obtained. Trace and toxic elements in airborne particulate matter collected at suburban and rural areas seasonally were analyzed by instrumental neutron activation analysis(INAA). Neutron irradiation for the samples was done at the irradiation hole (thermal neutron flux, 1 X 1013n/cm2·s) of TRIGA MARK-III research reactor in the Korea Atomic Energy Research Institute. For the verification of the analytical method, a standard reference material, NIST SRM-1648 was chosen and analyzed. The accuracy and precision for the analysis of trace elements in the standard samples were inter compared with the certified and reported values. The analytical error was agree within the 10% except few elements. Practically, we used this method to analyze trace and toxic elements in airborne particulate matter collected with the high volume air sampler(PM-10) at two different locations and also confirmed the possibility to use this method as a routine monitoring tool to find out environmental pollution sources
International Nuclear Information System (INIS)
Full text: Biomonitoring is an appropriate tool for the air pollution assessment studies. In this work, lichens and barks have been used as bio-accumulators in several sites in two Moroccan cities (Rabat and Mohammadia). The specific ability of absorbing and accumulating heavy metals and toxic element from the air, their longevity and resistance to the environmental stresses, make those bioindicators suitable for this kind of studies. The Instrumental Neutron Activation Analysis (INAA) is universally accepted as one of the most reliable analytical tools for trace and ultra-trace elements determination. Its use in trace elements atmospheric pollution related studies has been and is still extensive as can be demonstrated by several specific works and detailed reviews. In this work, a preliminary investigation employing lichens, barks and instrumental neutron activation analysis (INAA) was carried out to evaluate the trace elements distribution in six different areas of Rabat and Mohammadia cities characterised by the presence of many industries and heavy traffic. Samples were irradiated with thermal neutrons in a nuclear reactor and the induced activity was counted using high-resolution Germanium-Lithium detectors. More than 30 elements were determined using two modes : short irradiation (1 minute) and long irradiation (17 hours). Accuracy and quality control were assessed using the reference standard material IAEA-336. This was less than 1% for major and about 5 to 10% for traces.
International Nuclear Information System (INIS)
1 - Description of program or function: specified on ORNL-RSIC-25, shielding benchmark problems. - BP-3 (Neutron cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: air; Origin: ENDF/B; Weighting spectrum: 1/E; - BP-6 (neutron and gamma-ray cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: Borated Polyethylene (C-12, H, and B-10); Origin: ENDF/B-II. The cross section data can be used to repeat the Shielding Benchmark Problems 3.0 and 6.0 for testing against the results published in ORNL-RSIC-25. 2 - Method of solution: ZZ-BP-3 neutron cross sections from the CCC-17/05R library were processed into 104 neutron groups using the PSR-9/CSP code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The resulting multigroup cross sections are P5 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE. ZZ-BP-6 neutron and gamma-ray cross sections for 12C, H, and 10B were from ENDF/B-II data. The neutron multigroup cross sections were generated into 104 neutron groups using the PSR-13/SUPERTOG code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The gamma-ray multigroup cross sections were generated using PSR-7/MUG. The neutron-gamma-ray coupling utilized yield data from the DLC-12/POPOP4 library (data sets 010101, 060101, 060301, and 05100201). The neutron-gamma-ray coupled multigroup cross-section set was generated using the SAMPLE COUPLING CODE (ASCC). The multigroup cross sections are in a 22-18 group structure with P3 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE
Chock, Jeffrey Mun Kong
1999-01-01
Blast profiles and two primary methods of determining them were reviewed for use in the creation of a computer program for calculating blast pressures which serves as a design tool to aid engineers or analysts in the study of structures subjected to explosive air blast. These methods were integrated into a computer program, BLAST.F, to generate air blast pressure profiles by one of these two differing methods. These two methods were compared after the creation of the program and can conserv...
Time-resolved Fast Neutron Radiography of Air-water Two-phase Flows
Zboray, Robert; Dangendorf, Volker; Mor, Ilan; Tittelmeier, Kai; Bromberger, Benjamin; Prasser, Horst-Michael
Neutron imaging, in general, is a useful technique for visualizing low-Z materials (such as water or plastics) obscured by high-Z materials. However, when significant amounts of both materials are present and full-bodied samples have to be examined, cold and thermal neutrons rapidly reach their applicability limit as the samples become opaque. In such cases one can benefit from the high penetrating power of fast neutrons. In this work we demonstrate the feasibility of time-resolved, fast neutron radiography of generic air-water two-phase flows in a 1.5 cm thick flow channel with Aluminum walls and rectangular cross section. The experiments have been carried out at the high-intensity, white-beam facility of the Physikalisch-Technische Bundesanstalt, Germany. Exposure times down to 3.33 ms have been achieved at reasonable image quality and acceptable motion artifacts. Different two-phase flow regimes such as bubbly slug and churn flows have been examined. Two-phase flow parameters like the volumetric gas fraction, bubble size and bubble velocities have been measured.
International Nuclear Information System (INIS)
The MUSE project, carried out within the European fifth Framework Program, focuses on the coupling of a sub-critical reactor core with an external neutron source. In the first stage of the project a benchmark has been defined in order to define a reference calculational route, which is able to accurately predict the neutronics behavior in an accelerator driven system. Benchmark calculations will be carried out by several members of the project and the results will be compared, also with experimental results. The contribution of NRG to the project consists of the benchmark calculations and additional work that focuses on the calculation of 3D distributions of reaction yields. This paper discusses the non-conventional methods used to perform the benchmark calculations, including the 3D reaction yield distributions. The 3D distributions calculated for the sub-critical core will be Shown and discussed. With the ORANGE-extension to MCNP it is possible to tally 3D distributions, without adding extra cells and surfaces to the geometry and without a significant slowing down of the calculation. These are major advantages when compared to the conventional way of tallying in the MCNP-code. The distributions show details that can be understood in terms of the expected neutron behavior in the different parts of the geometry. For instance, the results show that: 1) a large number of fast neutrons is found in the fuel regions, 2) the reflector region shows an increased number of slower neutrons and 3) the reaction yield in the shielding region declines steeply. The extension therefore seems a useful tool in generating a better understanding of the behavior of neutrons throughout large and complex geometries like accelerator driven systems, but we also expect to use the extension in a variety of different fields. (authors)
International Nuclear Information System (INIS)
An accurate determination of damage fluence accumulated by reactor pressure vessels (RPV) as a function of time is essential in order to evaluate the vessel integrity for both pressurized thermal shock (PTS) transients and end-of-life considerations. The desired accuracy for neutron exposure parameters such as displacements per atom or fluence (E > 1 MeV) is of the order of 20 to 30%. However, these types of accuracies can only be obtained realistically by validation of nuclear data and calculational methods in benchmark facilities. The purposes of this paper are to review the needs and requirements for benchmark experiments, to discuss the status of current benchmark experiments, to summarize results and conclusions obtained so far, and to suggest areas where further benchmarking is needed
A Multigroup Method for the Calculation of Neutron Fluence with a Source Term
Heinbockel, J. H.; Clowdsley, M. S.
1998-01-01
Current research on the Grant involves the development of a multigroup method for the calculation of low energy evaporation neutron fluences associated with the Boltzmann equation. This research will enable one to predict radiation exposure under a variety of circumstances. Knowledge of radiation exposure in a free-space environment is a necessity for space travel, high altitude space planes and satellite design. This is because certain radiation environments can cause damage to biological and electronic systems involving both short term and long term effects. By having apriori knowledge of the environment one can use prediction techniques to estimate radiation damage to such systems. Appropriate shielding can be designed to protect both humans and electronic systems that are exposed to a known radiation environment. This is the goal of the current research efforts involving the multi-group method and the Green's function approach.
Intermediate states in the calculation of the neutron electric dipole moment
International Nuclear Information System (INIS)
In the Kobayashi-Maskawa model, the neutron electric dipole moment is a high-order effect in the electroweak interactions. We need a perturbative calculation in two steps: (1) generation of CP-violating couplings at the quark level through heavy quark loops, and (2) a perturbative expansion by nearest intermediate states at the bound state level. We examine both aspects for a mechanism proposed by Morel and by Cox et al. where CP violation comes from transition quark electric dipole moments. We retain the lowest lying baryon intermediate states 1/2+(56,0+) and 1/2-(70,1-). For this mechanism we get at most Dsub(n)/e approx. equal to 10-34 cm. (orig.)
Dorval, Eric
2016-01-01
Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient model...
Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes
International Nuclear Information System (INIS)
The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR-R1 reactor are calculated. The idea is to obtain the systematic error for k∞ for this methodology comparing the calculated and the experimental results
International Nuclear Information System (INIS)
Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect
International Nuclear Information System (INIS)
The code PERTURB.D computes the thermal neutron flux perturbation factor, K, due to circular foils located in an isotropic neutron field. The calculation is based on the expression K = G.E.F, where G denotes the neutron self-shielding in the foil, E the edge correction factor and F the flux depression in the diffusing medium surrounding the foil. By comparison with published experimental results is was found that, for 66% (85%) of the sample cases, calculated K-values agree to better than 1% (2%) with the experimental ones. As an application of the code PERTURB.D, tables of K-values for different materials, including Mn, Co, In and Au, for diameters in the range 0.5 to 2 cm, and various practical thicknesses, both in water and graphite, are calculated and presented. (Auth.)
Fast-neutron/gamma-ray radiography scanner for the detection of contraband in air cargo containers
Eberhardt, J.; Liu, Y.; Rainey, S.; Roach, G.; Sowerby, B.; Stevens, R.; Tickner, J.
2006-05-01
There is a worldwide need for efficient inspection of cargo containers at airports, seaports and road border crossings. The main objectives are the detection of contraband such as illicit drugs, explosives and weapons. Due to the large volume of cargo passing through Australia's airports every day, it is critical that any scanning system should be capable of working on unpacked or consolidated cargo, taking at most 1-2 minutes per container. CSIRO has developed a fast-neutron/gamma-ray radiography (FNGR) method for the rapid screening of air freight. By combining radiographs obtained using 14 MeV neutrons and 60Co gamma-rays, high resolution images showing both density and material composition are obtained. A near full-scale prototype scanner has been successfully tested in the laboratory. With the support of the Australian Customs Service, a full-scale scanner has recently been installed and commissioned at Brisbane International Airport.
Energy Technology Data Exchange (ETDEWEB)
MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.
1976-07-04
Indexes to the neutron-induced interaction data in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has two arrangements: isotope (ZA) order and reaction-number order.
Calculation of neutron fluxes and radioactivities in and around the Tokai-1 reactor pressure vessel
International Nuclear Information System (INIS)
This paper describes work that has been performed by NNC Ltd and Fuji Electric for the study of decommissioning of Tokai power station (Tokai-1). The objective was for NNC to provide an independent validation of representative selection of the existing Fuji Electric calculations the results of which were obtained by the methods generally used in Japan based on the discrete ordinate code DOT 3.5, in estimating full power neutron fluxes and reaction rates in components located within the reactor biological shield of the Tokai 1 reactor. The calculational methods and modelling assumptions are described for the four regions in which fluxes and reaction rates were required, namely in regions above the core, regions to the side of the core, regions below the core and regions in the concrete walls of the bio shield gas duct penetrations. NNC has considerable experience in performing similar analyses for UK reactors and the methodology and computer codes employed here are based on experience gained in carrying out such work for AGR, PWR and Magnox reactor types. Thus, much of the component modelling has been achieved using the Monte Carlo code MCBEND supplemented, in the case of the gas duct penetrations, by the iterative kernel albedo code MULTISORD. Above, below and to the side of the core, results were obtained in some detail in nearly all of the structural components. In the case of the bioshild concrete, results were obtained in many regions at various depths and axial heights. Along the gas ducts, results were calculated at the concrete wall surfaces of the penetrations to the point where the total flux had reduced to a level of 103 n/cm2/s, this being the level at which the induced concrete activity can be regarded as negligible. Preliminary calculations were carried out using the duct streaming code MULTISORD in order to establish the approximate location where flux levels dropped to this level. MCBEND was then used to model the geometry in detail up to this point
International Nuclear Information System (INIS)
High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by 241Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy-1 for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy-1 achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production. (author)
International Nuclear Information System (INIS)
We present a model which allows for the calculation of fragment excitation energy, fragment kinetic energies and neutron evaporation in nuclear fission. The model is based on the assumption that, at the end of the fission process, fragments are excited to a temperature which is proportional to the reaction Q-value. Starting from this assumption the distribution functions of fragment excitation can be formulated and the distribution functions for the kinetic energies can be derived by a Monte Carlo method. From the distribution functions for the excitation energy neutron evaporation characteristics are calculated. (author)
International Nuclear Information System (INIS)
With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)
Calculation analysis of Wims/D4-Batan-2DIFF neutron spectrum on RSG-GAS with cadmium ratio
International Nuclear Information System (INIS)
The calculation analysis of WIMS/D4-BATAN-2DIFF neutron spectrum was performed by comparison the calculation result of cadmium ratio with the experiment result on CIP, IP2, IP3 and IP4 irradiation positions of RSG GAS tenth core. The foils of Au, Mn and Co were used for determination of the measured and calculated cadmium ratios. Spectrum calculation was done in 69 energy group with 541 energy group (till 10 MeV) cross section of foil absorption reaction. The difference values between cadmium ratio calculation and experiment result for all cases were in interval of 11.0%-26.3% which are out of measurement deviation range. From these result, it concluded that the use of WIM /D4 in generating group constant is not sufficient to obtain the neutron spectrum, especially for non-fuel region
Energy Technology Data Exchange (ETDEWEB)
Salmon, R.; Hermann, O.W.
1992-10-01
The rate of neutron production from ([alpha], n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ([alpha], n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ([alpha], n) neutron production of each actinide in neutrons per second and the total for the canister. The ([alpha], n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.
ALPHN: A computer program for calculating (α, n) neutron production in canisters of high-level waste
International Nuclear Information System (INIS)
The rate of neutron production from (α, n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the (α, n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the (α, n) neutron production of each actinide in neutrons per second and the total for the canister. The (α, n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown
International Nuclear Information System (INIS)
Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the keff and other parameters are investigated.
Energy Technology Data Exchange (ETDEWEB)
Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)
2011-07-01
Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)
Fast neutron radiography scanner for the detection of contraband in air cargo containers
International Nuclear Information System (INIS)
There is a growing need to rapidly scan bulk air cargo for contraband such as illicit drugs and explosives. The Commonwealth Science and Industrial Research Organisation (CSIRO) have been working with Australian Customs Service to develop a scanner capable of directly scanning airfreight containers in 1-2 minutes without unpacking. The scanner combines fast neutron and gamma-ray radiography to provide high-resolution images that include information on material composition. A full-scale prototype scanner has been successfully tested in the laboratory and a commercial-scale scanner is due to be installed at Brisbane airport in 2005
Advances on the study of air pollution in Cordoba by neutron activation analysis
International Nuclear Information System (INIS)
Air pollution biomonitoring has been carried out in an area of 160.000 km2 by neutron activation analysis of lichen samples (Usnea sp. and Ramalina ecklonii) in the framework of a Co-ordinated Research Programme of the IAEA and an ARCAL Technical Co-operation Project. The samples were irradiated in the RA-3 reactor and after a decay time of 6, 12 and 30 days, 24 elements (As, Ba, Br, Ce, Co, Cr, Cs, Eu, Fe, Hf, La, Lu, Na, Nd, Rb, Sb, Sc, Sm, Ta, Tb, Th, U and Zn) were determined by gamma spectrometry. (author)
Energy Technology Data Exchange (ETDEWEB)
NONE
2005-09-15
The European Commission decided in 2001 an analysis program to reduce the atmospheric emissions. This report presents different limit scenari for France in 2020 (the reference scenari and the MTFR scenari, Maximum Technically Feasible Reduction), optimized scenari calculated by the RAINS model (Regional Air Pollution Information and Simulation), the costs of the scenari calculated with RAINS and the cost-benefit analysis of the strategy CAFE. From the study results, the benefits are higher than the costs, even with the most ambitious scenari. At an european level the emission reduction strategies have no effect on the employment but an impact on the Gross Domestic Product (decrease between 0,04 % and 0,12 % in function of the scenari). (A.L.B.)
Recent advances in neutron tomography
International Nuclear Information System (INIS)
Neutron imaging has been shown to be an excellent imaging tool for many nondestructive evaluation applications. Significantly improved contrast over X-ray images is possible for materials commonly found in engineering assemblies. The major limitations have been the neutron source and detection. A low cost, position sensitive neutron tomography detector system has been designed and built based on an electro-optical detector system using a LiF-ZnS scintillator screen and a cooled charge coupled device. This detector system can be used for neutron radiography as well as two and three-dimensional neutron tomography. Calculated performance of the system predicted near-quantum efficiency for position sensitive neutron detection. Experimental data was recently taken using this system at McClellan Air Force Base, Air Logistics Center, Sacramento, CA. With increased availability of low cost neutron sources and advanced image processing, neutron tomography will become an increasingly important nondestructive imaging method
International Nuclear Information System (INIS)
Calculations of cross sections of neutron induced reactions for 52Cr in 6-20 MeV energy range have been performed using Hauser-Feshbach code developed by the author. The calculations include cross sections of (n,n'), (n.np), (n,2n), (n,p), (n,pn), (n,pγ), (n,α), (n,αγ) and (n, αn) reactions induced in 52Cr. The calculations have been compared with measurements and evaluations. (author). 30 refs, 11 figs, 2 tabs
International Nuclear Information System (INIS)
Highlights: • Calculation of effective delayed neutron fraction in circulating-fuel reactors. • Extension of the Monte Carlo SERPENT-2 code for delayed neutron precursor tracking. • Forward and adjoint multi-group diffusion eigenvalue problems in OpenFOAM. • Analytical approach for βeff calculation in simple geometries and flow conditions. • Good agreement among the three proposed approaches in the MSFR test-case. - Abstract: This paper deals with the calculation of the effective delayed neutron fraction (βeff) in circulating-fuel nuclear reactors. The Molten Salt Fast Reactor is adopted as test case for the comparison of the analytical, deterministic and Monte Carlo methods presented. The Monte Carlo code SERPENT-2 has been extended to allow for delayed neutron precursors drift, according to the fuel velocity field. The forward and adjoint eigenvalue multi-group diffusion problems are implemented and solved adopting the multi-physics tool-kit OpenFOAM, by taking into account the convective and turbulent diffusive terms in the precursors balance. These two approaches show good agreement in the whole range of the MSFR operating conditions. An analytical formula for the circulating-to-static conditions βeff correction factor is also derived under simple hypotheses, which explicitly takes into account the spatial dependence of the neutron importance. Its accuracy is assessed against Monte Carlo and deterministic results. The effects of in-core recirculation vortex and turbulent diffusion are finally analysed and discussed
A priori efficiency calculations for Monte Carlo applications in neutron transport
International Nuclear Information System (INIS)
In this paper a general derivation is given of equations describing the variance of an arbitrary detector response in a Monte Carlo simulation and the average number of collisions a particle will suffer until its history ends. The theory is validated for a simple slab system using the two-direction transport model and for a two-group infinite system, which both allow analytical solutions. Numerical results from the analytical solutions are compared with actual Monte Carlo calculations, showing excellent agreement. These analytical solutions demonstrate the possibilities for optimizing the weight window settings with respect to variance. Using the average number of collisions as a measure for the simulation time a cost function inversely proportional to the usual figure of merit is defined, which allows optimization with respect to overall efficiency of the Monte Carlo calculation. For practical applications it is outlined how the equations for the variance and average number of collisions can be solved using a suitable existing deterministic neutron transport code with adapted number of energy groups and scattering matrices. (author)
Energy Technology Data Exchange (ETDEWEB)
Ghassoun, Jillali; Jehoauni, Abdellatif [Nuclear physics and Techniques Lab., Faculty of Science, Semlalia, Marrakech (Morocco)
2000-01-01
In practice, the estimation of the flux obtained by Fredholm integral equation needs a truncation of the Neuman series. The order N of the truncation must be large in order to get a good estimation. But a large N induces a very large computation time. So the conditional Monte Carlo method is used to reduce time without affecting the estimation quality. In a previous works, in order to have rapid convergence of calculations it was considered only weakly diffusing media so that has permitted to truncate the Neuman series after an order of 20 terms. But in the most practical shields, such as water, graphite and beryllium the scattering probability is high and if we truncate the series at 20 terms we get bad estimation of flux, so it becomes useful to use high orders in order to have good estimation. We suggest two simple techniques based on the conditional Monte Carlo. We have proposed a simple density of sampling the steps for the random walk. Also a modified stretching factor density depending on a biasing parameter which affects the sample vector by stretching or shrinking the original random walk in order to have a chain that ends at a given point of interest. Also we obtained a simple empirical formula which gives the neutron flux for a medium characterized by only their scattering probabilities. The results are compared to the exact analytic solution, we have got a good agreement of results with a good acceleration of convergence calculations. (author)
International Nuclear Information System (INIS)
In practice, the estimation of the flux obtained by Fredholm integral equation needs a truncation of the Neuman series. The order N of the truncation must be large in order to get a good estimation. But a large N induces a very large computation time. So the conditional Monte Carlo method is used to reduce time without affecting the estimation quality. In a previous works, in order to have rapid convergence of calculations it was considered only weakly diffusing media so that has permitted to truncate the Neuman series after an order of 20 terms. But in the most practical shields, such as water, graphite and beryllium the scattering probability is high and if we truncate the series at 20 terms we get bad estimation of flux, so it becomes useful to use high orders in order to have good estimation. We suggest two simple techniques based on the conditional Monte Carlo. We have proposed a simple density of sampling the steps for the random walk. Also a modified stretching factor density depending on a biasing parameter which affects the sample vector by stretching or shrinking the original random walk in order to have a chain that ends at a given point of interest. Also we obtained a simple empirical formula which gives the neutron flux for a medium characterized by only their scattering probabilities. The results are compared to the exact analytic solution, we have got a good agreement of results with a good acceleration of convergence calculations. (author)
Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor
Energy Technology Data Exchange (ETDEWEB)
Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2011-07-01
The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)
Calculating Hurst exponent and neutron monitor data in a single parallel algorithm
Kussainov, A. S.; Kussainov, S. G.
2015-09-01
We implemented an algorithm for simultaneous parallel calculation of the Hurst exponent H and the fractal dimension D for the time series of interest. Parallel programming environment was provided by OpenMPI library installed on three machines networked in the virtual cluster and operated by Debian Wheeze operating system. We applied our program for a comparative analysis of week and a half long, one minute resolution, six channels data from neutron monitor. To ensure a faultless functioning of the written code we applied it to analysis of the random Gaussian noise signal and time series with manually introduced self-affinity features. Both of them have the well-known values of H and D. All results are in good correspondence with each other and supported by the modern theories on signal processing thus confirming the validity of the implemented algorithms. Our code could be used as a standalone tool for the different time series data analysis as well as for the further work on development and optimization of the parallel algorithms for the time series parameters calculations.
PTRAC file utilization for calculation of free-air ionization chamber correction factors by MCNPX
International Nuclear Information System (INIS)
A free-air ionization chamber is used as a standard of photon air-kerma. Several correction factors are applied to the air-kerma value. Correction factors for electron loss: k(loss) and for additional ionization current caused by photon scatter: k(sc), photon fluorescence: k(fl), photon transmission through diaphragm edge k(dtr), and photon scatter from the surface of the diaphragm aperture k(dsc) were determined by the MCNPX code utilizing information stored in Particle Track (PTRAC) output files. Individual steps of the procedure are described and the calculated values of the correction factors are presented. The values are in agreement with the correction factors published in the literature for similar free-air chambers and low-energy photons. (authors)
International Nuclear Information System (INIS)
Molten Salt Reactor (MSR) is the only one using liquid fuel in the six candidate reactors of the Generation IV advanced nuclear power systems with expected remarkable advantages in safety, economics, sustainability, and proliferation resistance. The strong coupling between neutronics and thermal-hydraulics due to fuel movement in the liquid-fuel MSRs induces many new challenges in reactor analyses from the perspective of both theoretical models and solution methods. In this study, the multi-group diffusion theory was adopted to deduce the neutronics model for the liquid-fuel MSRs, in which the salt flow effects on the delayed neutron precursor distributions in space were considered particularly. Since the liquid-fuel salt is a Newton fluid, the single-phase thermal hydraulics model for liquid-fuel MSRs was generally established based on the fundamental laws of the mass, momentum and energy conservation equations as used in the computational fluid dynamic (CFD) method. Since the control equations of the neutronic model can be written in the same form of those solved in the CFD softwares, a neutronics and thermal-hydraulics coupling scheme was proposed and a program was developed based on the FLUENT software by using its user-defined functions and subroutines (UDF and UDS). This program was applied to perform the steady state calculation of the molten salt fast reactor (MSFR), and the main results such as the space distributions of the neutron fluxes, delayed neutron precursors, temperatures, velocities were obtained. The results show that the liquid fuel flow influences the delayed neutron precursors significantly, while slightly affects the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank leading the maximal temperature to about 1200 K at the centre of the vortex, which will be optimized in the future core design. (author)
Neutronic calculations of hexagonal lattice nuclear reactors: Modelling of the CAREM-25 reactor
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This work was carried out in the frame of the Cnea CAREM-25 project (Central Argentina de Elementos Modulares).This project involves the development and construction of an argentinian design nuclear reactor for producing electricity. It's a PWR type (light water moderated and enriched U02 fueled) integrated reactor in an hexagonal lattice.The total power of this prototype is 100 MW thermal. In this frame, the main objective of this work is to consolidate and validate a neutronic line of calculus which can be applied to the CAREM-25 core.At a first analysis at cell level, the different fuel elements were modeled with the Dragon code, obtaining homogenised and condensed cross sections.Then a core level analysis with the Puma code was performed at full power condition and room temperature. A comparison of the obtained results is needed.For this reason, a Monte Carlo analysis (at room temperature) was performed.Also a validation of the Dragon code was carried out on the base of experimental data of WWER type lattices (similars to CAREM).The confidence on the results is then granted and their uncertainties were quantified.The Dragon-Puma line of calculus is then established and the main objective of this work is achieved. A full neutronic analysis should be followed by thermohydraulics calculations in an iterative procedure, and it would be the objective of future works.Finally, a burnup analysis was performed, at cell and core level.The design condition for extraction burnup and fuel cycle duration were verified.
A comparison of measured and calculated values of air kerma rates from 137Cs in soil
Directory of Open Access Journals (Sweden)
V. P. Ramzaev
2016-01-01
Full Text Available In 2010, a study was conducted to determine the air gamma dose rate from 137Cs deposited in soil. The gamma dose rate measurements and soil sampling were performed at 30 reference plots from the south-west districts of the Bryansk region (Russia that had been heavily contaminated as a result of the Chernobyl accident. The 137Cs inventory in the top 20 cm of soil ranged from 260 kBq m–2 to 2800 kBq m–2. Vertical distributions of 137Cs in soil cores (6 samples per a plot were determined after their sectioning into ten horizontal layers of 2 cm thickness. The vertical distributions of 137Cs in soil were employed to calculate air kerma rates, K, using two independent methods proposed by Saito and Jacob [Radiat. Prot. Dosimetry, 1995, Vol. 58, P. 29–45] and Golikov et al. [Contaminated Forests– Recent Developments in Risk Identification and Future Perspective. Kluwer Academic Publishers, 1999. – P. 333–341]. A very good coincidence between the methods was observed (Spearman’s rank coefficient of correlation = 0.952; P<0.01; on average, a difference between the kerma rates calculated with two methods did not exceed 3%. The calculated air kerma rates agreed with the measured dose rates in air very well (Spearman’s coefficient of correlation = 0.952; P<0.01. For large grassland plots (n=19, the measured dose rates were on average 6% less than the calculated kerma rates. The tested methods for calculating the air dose rate from 137Cs in soil can be recommended for practical studies in radiology and radioecology.
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) flexibility with respect to input parameters and output data, including easy communication with another software and (6) suitability for didactical/training purposes. ANGLE frame is also convenient for accommodating other efficiency calculation methods of semi-empirical or absolute type, Monte Carlo for instance. In addition, it is a matter of little effort to extend its existing scope of applicability to further/particular user's needs and/or fields of interest (can be regarded as ''open-ended'' computer code). For reactor neutron flux characterization purposes, ANGLE applicability is related to flux monitor measurements. The emphasis is given to suitable metal foils and alloys which are being irradiated/ activated in characteristic reactor positions so as to give subsequently gamma spectrum information of flux density, shape, etc. (author)
Neutronics calculation of a reactor cold neutron source%反应堆冷中子源中子物理学计算
Institute of Scientific and Technical Information of China (English)
胡春明; 余朝举; 童剑飞
2011-01-01
用MCNP软件计算反应堆冷中子源,慢化剂室内平均中子注量率为6.69× 1013/cm-2.s-1,波长为0.4 nm和0.6 nm的冷中子增益因子～16和32.冷源慢化剂中正仲氢比例对输出的冷中子能谱有较大影响,而在3K范围内慢化剂温度变化对冷中子能谱的影响很小.计算结果表明,冷中子源性能达到基本设计要求.%The construction of a reactor cold neutron source (CNS) will be completed in the near future. To evaluate performance of the CNS, a neutronics calculation using MCNP4C code has been carried out. The results show that the average neutron flux in the moderator is 6.69×1013/cm2·s, and the cold neutron gain factors corresponding to 4-(A) and 6-(A) wavelengths are 16 and 32, respectively. The results also indicate that different ratios of ortho-H2/para-H2 have an obvious impact on cold neutron spectrum in the moderator, but within 3 K of the moderator temperature changes, the spectrum varies slightly.
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Highlights: • We analize the performance of neutron scattering libraries for D and O in D2O for nuclear criticality calculations. • We calculated 65 ICSBEP benchmark cases from 8 heavy water moderated thermal systems using MCNP5. • A significant improvement is found when our library is combined with the ROSFOND-2010 evaluation for deuterium. • In 48 of the 65 benchmark cases we obtained a C/E ratio closer to 1.0. • The percentage of benchmark cases calculated within 1-sigma increases from 42% to 82%, compared to ENDF/B-VII calculations. - Abstract: To improve the evaluations in thermal sublibraries, we developed a set of thermal neutron scattering cross sections (scattering kernels) for the deuterium and oxygen bound in heavy water in the ENDF-6 format. These new libraries are based on molecular dynamics simulations and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we show how the use of this new set of cross sections also improves the calculation of thermal critical systems moderated and/or reflected with heavy water. The use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the deuterium cross sections, results in an improvement of the C/E ratio in 48 out of 65 benchmark cases calculated with the Monte Carlo code MCNP5, in comparison with the existing library based on the ENDF/B-VII evaluation
Smart nanogels at the air/water interface: structural studies by neutron reflectivity
Zielińska, Katarzyna; Sun, Huihui; Campbell, Richard A.; Zarbakhsh, Ali; Resmini, Marina
2016-02-01
The development of effective transdermal drug delivery systems based on nanosized polymers requires a better understanding of the behaviour of such nanomaterials at interfaces. N-Isopropylacrylamide-based nanogels synthesized with different percentages of N,N'-methylenebisacrylamide as cross-linker, ranging from 10 to 30%, were characterized at physiological temperature at the air/water interface, using neutron reflectivity (NR), with isotopic contrast variation, and surface tension measurements; this allowed us to resolve the adsorbed amount and the volume fraction of nanogels at the interface. A large conformational change for the nanogels results in strong deformations at the interface. As the percentage of cross-linker incorporated in the nanogels becomes higher, more rigid matrices are obtained, although less deformed, and the amount of adsorbed nanogels is increased. The data provide the first experimental evidence of structural changes of nanogels as a function of the degree of cross-linking at the air/water interface.The development of effective transdermal drug delivery systems based on nanosized polymers requires a better understanding of the behaviour of such nanomaterials at interfaces. N-Isopropylacrylamide-based nanogels synthesized with different percentages of N,N'-methylenebisacrylamide as cross-linker, ranging from 10 to 30%, were characterized at physiological temperature at the air/water interface, using neutron reflectivity (NR), with isotopic contrast variation, and surface tension measurements; this allowed us to resolve the adsorbed amount and the volume fraction of nanogels at the interface. A large conformational change for the nanogels results in strong deformations at the interface. As the percentage of cross-linker incorporated in the nanogels becomes higher, more rigid matrices are obtained, although less deformed, and the amount of adsorbed nanogels is increased. The data provide the first experimental evidence of structural changes
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The anisotropic scattering influences both the transport and slowing processes of neutrons. Since Practical shields are usually anisotropic scatters, several parameterized, anisotropic scattering kernels were used to present a general class of anisotropies. To study the anisotropic sensitivity of the flux in thick shield medium, the feasibility of track- length distribution biasing for calculations of scalar and angular neutron flux and their sensitivity to anisotropic scattering was investigated. To represent more realistic angular distribution in a parameterized functional form, an exponential angular density for sampling the scattering angle is proposed, also an empirical formula for the choice of optimal parameter for track length biasing depending on the anisotropic scattering is proposed. The calculations are performed for a particle transport model having an exact solution. The results show that this distribution covers also isotropic and the anisotropic scattering case. The anisotropic effect has a great influence on the behavior of neutron distribution particularly in thick shield. (author)
International Nuclear Information System (INIS)
1 - Description: The use of albedo techniques is central to many radiation streaming codes and has been widely used as an alternative to much more expensive transport calculations. Key to the albedo technique is the availability of either a large set of albedo data or, preferably, an empirical formula that approximates the albedo over the range of source energies and incident and exit radiation directions involved in a particular problem. Previously proposed neutron and photon albedo approximating formulas have been based on limited energy-angular ranges, a single reflecting material, old cross section data, and, most important, obsolete fluence-to-dose response functions. This library contains differential neutron dose albedo functions, based on modern cross section and response function data. Newly evaluated parameters are tabulated for several empirical differential dose albedo formulas. The albedos considered are (1) two approximations for photon albedo, (2) a new approximation for the neutron albedo, and (3) the secondary-photon albedo for incident neutrons. Albedo data is provided for four Materials: concrete, iron, lead, and water. Unlike previous compilations of albedo data, modern dosimetric units have been employed. Data are presented for (1) the ambient dose equivalent H*(10 mm) and (2) the effective dose equivalent for anteroposterior (AP) illumination of the ICRP anthropomorphic phantom. 2 - Methods: Monte Carlo code, MCNP, was used to calculate the albedo reflected from thick slabs of various materials. In particular, a homogeneous cylindrical slab surrounded by a vacuum was used. The incident neutrons were modeled by a point mono-directional source positioned just inside the center of the circular scoring (reflecting) surface. This was done to facilitate scoring because all particles crossing the surface must be outgoing (reflected) particles. Slab thickness and radius were sufficiently large (1000 cm) so that negligible numbers of neutrons were
Validation of iron nuclear data for the neutron calculation of nuclear reactors
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The GEN-III and GEN-IV reactors will be equipped with heavy reflectors. However, the existing integral validation of the iron nuclear data in the latest JEFF3 European library in the frame of the neutron calculation of the heavy reflector is very partial: some results exist concerning fast reactors but there is no result corresponding to the LWR heavy reflector. No clear trend on the JEFF3 iron cross sections was brought into evidence up to now for fission reactor calculations. Iron nuclear data were completely re-evaluated in the JEFF3 library. Despite the fact that iron is widely used in the nuclear industry, large uncertainties are still associated with its nuclear data, particularly its inelastic cross section which is very important in the neutron slowing down. A validation of 56Fe nuclear data was performed on the basis of the analysis of integral experiments. Two major critical experiments, the PERLE experiment and the Gas Benchmark, were interpreted with 3D reference Monte-Carlo calculations and the JEFF3.1.1 library. The PERLE experiment was recently performed in the EOLE zero-power facility (CEA Cadarache). This experiment is dedicated to heavy reflector physics in GEN-III light water reactors. It was especially conceived for the validation of iron nuclear data. The Gas Benchmark is representative of a Gas Fast Reactor with a stainless steel reflector (with no fertile blanket) in the MASURCA facility (CEA Cadarache). Radial traverses of reaction rates were measured to characterize flux attenuation at various energies in the reflector. The results of the analysis of both experiments show good agreement between the calculations and the measurements, which is confirmed by the analysis of complementary experiments (ZR-6M, MISTRAL4, CIRANO-ZONA2B). A process of re-estimating the 56Fe nuclear data was implemented on the basis of feedback from these two experiments and the RDN code. This code relies on a non-linear regression method using an iterative technique
DEFF Research Database (Denmark)
Christensen, Rune; Hansen, Heine Anton; Vegge, Tejs
2015-01-01
Density functional theory (DFT) calculations have greatly contributed to the atomic level understanding of electrochemical reactions. However, in some cases, the accuracy can be prohibitively low for a detailed understanding of, e.g. reaction mechanisms. Two cases are examined here, i.e. the...... electrocatalytic reduction of CO2 and metal-air batteries. In theoretical studies of electrocatalytic CO2 reduction, calculated DFT-level enthalpies of reaction for CO2reduction to various products are significantly different from experimental values[1-3]. In theoretical studies of metal-air battery reactions......, systematic errors compared to experiments have also been found in calculation of enthalpies of formation for bulk metal oxide, peroxide and superoxide species[4,5]. It is here demonstrated how the errors, which depend explicitly on the choice of applied exchange-correlation functional, can be identified...
Energy Technology Data Exchange (ETDEWEB)
Fonseca Diaz, Nestor [Universidad Tecnologica de Pereira, Facultad de Ingenieria Mecanica, Pereira (Colombia); University of Liege, Campus du Sart Tilman, Bat: B49, P33, B-4000 Liege (Belgium)
2009-09-15
This article presents the general procedure for uncertainty calculation of net total cooling effect estimation for rating room air conditioners and packaged terminal air conditioners, by means of measurements carried out in a test bench specially designed for this purpose. The uncertainty analysis presented in this work looks for establishing a confidence degree or certainty of experimental results. It is particularly important considering that international standards related to this type of analysis are too ambiguous when treating this subject. The uncertainty analysis is on the other hand an indispensable requirement to international standard ISO 17025 [ISO, 2005. International Standard. 17025. General Requirement to Test and Calibration Laboratories Competences. International Organization for Standardization, Geneva.], which must be applied to obtain the required quality levels according to the Word Trade Organization WTO. (author)
Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium
International Nuclear Information System (INIS)
This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of 252Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of 252Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt (7LiF - BeF2 - ThF4 - UF4) as fuel. For design the keff, neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of 233U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the keff of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)
Mueller, Roland Guenther
1987-06-01
In order to account for subcooled boiling in calculations of neutron physics and thermal hydraulics of light water reactors (where vapor bubbles strongly influence the nuclear chain reaction), a dynamic model is derived from the time-dependent conservation equations. It contains methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. It enables the complete two-phase flow region to be treated consistently. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement is reached. The results from the coupling of the new calculation model with a neutron kinetics program proves its suitability for the steady-state and transient calculation of reactor cores.
VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation
International Nuclear Information System (INIS)
1 - Description of problem or function: VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D and 3-D diffusion calculation, depletion and shut-down features, in- core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (steady state and transient). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of thermal reactors, their fuel cycles, thermal transients, and safety assessment. Besides its use in research and development work for the Gas Cooled High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors, MAGNOX, and RBMK. 2 - Method of solution: The nuclear data for 184 isotopes are contained in two libraries. Fast and epithermal data in a 68 group GAM-I structure have been prepared mainly from ENDF/B-V and JEF-1. Resonance cross section data are given as input. Thermal data in a 30 group THERMOS structure have been collapsed from a 96 group THERMALIZATION (GATHER) library by a relevant neutron energy spectrum generated by the THERMALIZATION code. Graphite scattering matrices are based on the Young phonon spectrum in graphite. The neutron spectrum is calculated by a combination of the GAM and THERMOS codes. They can simultaneously be employed for many core regions differing in temperature, burnup, and fuel element lay-out. The thermal cell code THERMOS has been extended to treat the grain structure of the coated particles inside the fuel elements, and the epithermal GAM code uses modified cross sections for the resonance absorbers prepared from double heterogeneous ZUT-DGL calculations. The diffusion module of the code is CITATION with 2 - 8 energy groups. It provides the neutron
International Nuclear Information System (INIS)
Highlights: • Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, in some ITER construction steels. • The neutron flux density was measured by means of activation foil method and unfolding technique. • Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TALYS-2011. • The activity measurements were done by means of gamma-ray spectrometry. - Abstract: Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, for some ITER construction steels. The activation was conducted in fast neutron irradiation channel of the MARIA research fission reactor (Poland). The dimensions of steel samples were 10 mm × 10 mm × 1 mm and mass was approximately 0.8 g. The neutron flux density was measured by means of activation foil method and unfolding technique; fraction of neutrons above 1 keV was 95%. The activation lasted 242 h and cooling took 100 days; the mean neutron flux density was 2.9E12 n/(cm2 s) (neutrons above 500 keV are 53% of total) whereas total fluency 2.53E18 cm−2. The activity measurements were done by means of gamma-ray spectrometry. Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TENDL-2011 and experimentally determined neutron flux. Measured activity of long-lived gamma emitting radionuclides was, in average, about 6.3 MBq/g 100 days after activation; the dominant radionuclides were 58Co and 54Mn (about 81% and 14% of total activity respectively). The C/E ratio differs for particular radionuclides and is in the range 0.86–0.92 for 51Cr, 0.93–1.21 for 54Mn, 0.77–0.98 for 57Co, 0.91–1.21 for 58Co, 1.17–1.27 for 59Fe, and 1.75–2.44 for 60Co
International Nuclear Information System (INIS)
This paper is the first in a series of publications dedicated to the description of the new Russian multigroup data set BNAB-93. The first part of this series is devoted to the description of the neutron and photon data and their formats, and to their use in calculations. (author). 12 refs, 8 tabs
International Nuclear Information System (INIS)
Boron-loaded scintillators offer the potential for neutron spectrometers with a simplified, peak-shaped response. The Monte Carlo code, MCNP, has been used to calculate the detector characteristics of a scintillator made of a boron-loaded plastic, BC454, for neutrons between 1 and 7 MeV. Comparisons with measurements are made of spectral response for neutron energies between 4 and 6 MeV and of intrinsic efficiencies for neutrons up to 7 MeV. In order to compare the calculated spectra with measured data, enhancements to MCNP were introduced to generate tallies of light output spectra for recoil events terminating in a final capture by 10B. The comparison of measured and calculated spectra shows agreement in response shape, full width at half maximum, and recoil energy deposition. Intrinsic efficiencies measured to 7 MeV are also in agreement with the MCNP calculations. These results validate the code predictions and affirm the value of MCNP as a useful tool for development of sensor concepts based on boron-loaded plastics. (orig.)
International Nuclear Information System (INIS)
The effects of the neutron strength function uncertainties on the calculated values of the self-shielding factors and energy dependence of the total and capture 238U cross-sections in the unresolved resonance region are investigated. (author). 26 refs, 5 figs
International Nuclear Information System (INIS)
The report contains the texts of the 9 invited papers delivered during the Second Research Co-ordination Meeting on ''Methods for the Calculation of Fast Neutron Nuclear Data for Structural Materials and Fast and Fusion Reactors'' held in Vienna during 15-17 February 1988. A separate abstract was prepared for each of these 9 papers. Refs, figs and tabs
Ullmann, J. L.; Krticka, M.; Kawano, T.; Bredeweg, T. A.; Baramsai, B.; Couture, A.; Haight, R. C.; Jandel, M.; Mosby, S.; O'Donnell, J. M.; Rundberg, R. S.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Wu, C. Y.; Chyzh, A.
2015-10-01
Calculations of the neutron-capture cross section at low neutron energies (10 eV through 100's of keV) are very sensitive to the nuclear level density and radiative strength function. These quantities are often poorly known, especially for radioactive targets, and actual measurements of the capture cross section are usually required. An additional constraint on the calculation of the capture cross section is provided by measurements of the cascade gamma spectrum following neutron capture. Recent measurements of 234 , 236 , 238U(n, γ) emission spectra made using the DANCE 4 π BaF2 array at the Los Alamos Neutron Science Center will be presented. Calculations of gamma-ray spectra made using the DICEBOX code and of the capture cross section made using the CoH3 code will also be presented. These techniques may be also useful for calculations of more unstable nuclides. This work was performed with the support of the U.S. Department of Energy, National Nuclear Security Administration by Los Alamos National Security, LLC (Contract DE-AC52-06NA25396) and Lawrence Livermore National Security, LLC (Contract DE-AC52-07NA2734).