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...
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.
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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)
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
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)
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
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
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
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
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
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)
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)
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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
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
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
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
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
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)
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
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.)
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
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
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)
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
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
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.
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
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 ?
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 ?
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
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
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
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
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)
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
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
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
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
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
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...
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
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
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
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)
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.
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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)
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
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.)
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
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
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
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
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
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
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
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
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.
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
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
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
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
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
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
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)
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)
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
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.
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)
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)
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
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
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.)
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
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
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
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
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
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
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
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.
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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)
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 ...
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
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
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)