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)
Determination of self-shielding factor and cadmium ration of foil and cylindrical probe have been done by measurement and calculation. Self-shielding factor was determined by dividing the activity of detector with its Al-alloy. Theoretically, self-shielding factor can be determined by numerical solution of two-dimensional integral equations in FORTRAN. For gold foil and wire, the calculation result are quite close to the measurement. The relative difference between calculation and measurement of activity, self-shielding factor and cadmium ratio are respectively less than 11%, 9% and 4%. It is therefore, the calculation program can be used for calculation of other kinds of activation detectors. The application in neutron flux measurement gives a better result especially for epithermal flux. For neutron spectrum measurement, self-shielding correction can avoid resonance peaks in epithermal region due to absorption by activation detectors. (author)
When a sample is analysed with neutron activation analysis (NAA) neutron self-shielding and gamma self-absorption affect the accuracy. Both effects become even more important when the mass of a sample analysed is changed from small (say, 1 g) to large (say 30 kg). Therefore, corrections have to be carried out. Correction method for neutron self-shielding is considered for a thermal neutron beam irradiating large homogeneous samples for prompt-gamma NAA (PGNAA). The correction method depends on the macroscopic scattering and absorption cross sections of the sample. To avoid doing experiments with samples with different macroscopic scattering and absorption cross sections, the Monte Carlo model MCNP is applied in the development of the correction method. The computational development of the method to determine these cross sections through flux monitoring outside the sample is described. (author)
Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group (1-g) cross sections must be provided in advance. This paper focuses on generating accurate 1-g cross section values that are necessary for evaluation of nuclide densities as a function of burnup. The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires extensive computational efforts. The method presented here is based on the multi-group (MG) approach, in which pre-generated MG sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate 1-g cross sections requires their tabulation against the background cross-section (σ0) to account for the self-shielding effect. However, in previous studies, the model that was used to calculate σ0 was simplified by fixing Bell and Dancoff factors. This work demonstrates that 1-g values calculated under the previous simplified model may not agree with the tallied values. Therefore, the original background cross section model was extended by implicitly accounting for the Dancoff and bell factors. The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented into BGCore code system. The 1-g cross section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement (<0.05%) in the 1-g cross values was observed. The method dose not carry any additional computational burden and it is universally applicable to the analysis of thermal as well as fast reactor systems. (author)
The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li2CO3 and the Li2O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)
The self-shielding factors in activation detectors used in the IPEN/MB-01 reactor
This work aims to obtain self-shielding factors (G) to several activation foils, such as gold, cobalt, scandium, magnesium, uranium, thorium and indium foils used at measurements of the neutron spectrum energy in the IPEN/MB-01 reactor core. The knowledge of the self-shielding factors allows obtaining of precise nuclear reaction rates without the effects of neutron flux depression inside of activation foils. This study is carried out in two parts. First are determined the self-shielding factors for bare foils (without cadmium covered) and after the self-shielding factors to foils with cadmium covered. (author)
Resonance self-shielding near zone interfaces
A practical methodology is developed to treat the resonance self-shielding transition near zone interfaces. Based on the narrow resonance approximation, a space- and energy-dependent self-shielding factor for a single interface system is derived from the integral transport theory. Using the Wigner rational approximation, the self-shielding factor for a fine region near a zone interface is factorized into a linear combination of individual homogeneous and heterogeneous self-shielding factors. The method has been implemented in a widely used cross-section processing code that is based on the Bondarenko f-factor method. The result of the analysis was applied to a fast reactor blanket mock-up to improve the calculations near a converter-blanket interface. Comparisons of the calculation with /sup 238/U capture experimental data measured in the Purdue Fast Breeder Blanket Facility are also discussed
APOLLO2 code self-shielding formalism
This report describes the various self-shielding methods used in the APOLLO2 code for treating one resonant nucleus or a mixture of resonant nuclei. The methods are expounded in chronological order. First of all, the methods dealing with one resonant isotope are explained. Then an original method dealing directly with a resonant mixture is detailed. This new method is also convenient for one resonant nucleus and leads, in that case, to interesting improvements in the self-shielding modeling. (author)
Self-shielding models of MICROX-2 code: Review and updates
Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study
Nasrabadi, M N; Mohammadi, A; Jalali, M
2009-01-01
In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required. PMID:19328700
Nasrabadi, M.N. [Department of Nuclear Engineering, Faculty of Modern Sciences and Technologies, University of Isfahan, Isfahan 81746-73441 (Iran, Islamic Republic of)], E-mail: mnnasrabadi@ast.ui.ac.ir; Mohammadi, A. [Department of Physics, Payame Noor University (PNU), Kohandej, Isfahan (Iran, Islamic Republic of); Jalali, M. [Isfahan Nuclear Science and Technology Research Institute (NSTRT), Reactor and Accelerators Research and Development School, Atomic Energy Organization of Iran (Iran, Islamic Republic of)
2009-07-15
In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.
Advances in self-shielded accelerators
The use of lead in lieu of concrete for shielding has enabled a significant segment of the electron beam (EB) processing industry to continue to grow. Self-shielded accelerators of 300 kV or less are used in the curing of environmentally-friendly thin-film coatings and in crosslinking extruded polymeric films. New low-voltage accelerator systems have been developed, including very economic modular units, which are expanding market interests. Transportable systems based on placing self-shielded accelerators on vans have found only minor interest for use in environmental remediation, with no commercial success to date. Higher-voltage, around 600 to 800 kV, self-shielded systems have found minimal acceptance in historic markets as in the crosslinking wire insulation and processing of tire components. However, new developments in higher energy, 2.5 MeV, modest current, self-contained systems may find use for in-house sterilization and treatment of products. (author)
Self-shielding clumps in starburst clusters
Palouš, Jan; Ehlerová, Soňa; Tenorio-Tagle, Guillermo
2016-01-01
Young and massive star clusters above a critical mass form thermally unstable clumps reducing locally the temperature and pressure of the hot 10$^{7}$~K cluster wind. The matter reinserted by stars, and mass loaded in interactions with pristine gas and from evaporating circumstellar disks, accumulate on clumps that are ionized with photons produced by massive stars. We discuss if they may become self-shielded when they reach the central part of the cluster, or even before it, during their free fall to the cluster center. Here we explore the importance of heating efficiency of stellar winds.
Effect of Self-Shielding on Burn-Up Calculation of ETRR-2 Reactor
There exist two approaches for burn-up calculation. The first on is to use cell parameters generated using cell calculation code at different degrees of burn-up. The other is to use microscopic cross sections with self-shielding in order to compensate for the variation of spectrum at different degree of burn-up. The effect of using different forms of self-shielding factors on burn-up calculation for ETRR-2 reactor has been determined. The results of the two approaches are inter-compared up to 50% burn-up
URR, Cross-Sections, Self-shielding for Fertile and Fissile Isotopes in Unresolved Region
1 - Description of program or function: URR was developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. 2 - Method of solution: URR uses Monte Carlo techniques to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various self-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 3 - Restrictions on the complexity of the problem: None noted
PAPIN, Cross Section, Self-Shielding Factors for Fertile Isotopes in Unresolved Resonance Region
1 - Description of problem or function: PAPIN calculates cross section probability tables, Bondarenko self-shielding factors and average self-indication ratios for non-fissile isotopes, below the inelastic threshold, on the basis of the ENDF/B prescriptions for the unresolved resonance region. 2 - Method of solution: Monte-Carlo methods are utilized to generate ladders of resonance parameters in the unresolved resonance region, from average parameters and their appropriate distribution functions. The neutron cross-sections are calculated by the single-level Breit-Wigner formalism, with s-, p-, and d-wave contributions. The cross section probability tables are constructed by sampling the Doppler-broadened cross sections. The various self-shielded factors are computed numerically as Lebesgue integrals over the cross section probability tables
Self-shielding Electron Beam Installation for Sterilization
Linac; Laboratory
2002-01-01
China Institute of Atomic Energy (CIAE) has developed a self-shielding electron beam installationfor sterilization as handling letters with anthrax germ or spores which has the least volume and the least
Self-shielding in spatially distributed samples can be very significant when detecting photons below 100 keV. A correction method has been developed for cylindrical samples, typically used in measurements of natural radioactivity with well-type detectors, as is the example of sediment dating. The method calculates the probability of photons to escape the sample, using Monte Carlo techniques with a program written in C. The effects of self-shielding on the angular distribution of photons have been indirectly analyzed as a function of sample geometry, given a fixed detector geometry. The results given by the program have also been used to provide a straightforward way of calculating the self-shielding factors, with uncertainties. Required inputs are the attenuation coefficient of the sample, its radius, and its height. A procedure has also been specified in order to make good use of all the information. To check on the method, measurements of reference material of known activity have been compared with calculated values, obtaining very satisfactory results
Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)
REPOSITORY LAYOUT SUPPORTING DESIGN FEATURE NO.13 - WASTE PACKAGE SELF SHIELDING
The objective of this analysis is to develop a repository layout, for Feature No. 13, that will accommodate self-shielding waste packages (WP) with an areal mass loading of 25 metric tons of uranium per acre (MTU/acre). The scope of this analysis includes determination of the number of emplacement drifts, amount of emplacement drift excavation required, and a preliminary layout for illustrative purposes
Self-shielding in large cross-section neutron absorbers
This study is dealing with finding the effects on neutron regime in several cases, among them a fuel bundle comprised of 16 fuel rods, made of sintered UO2 pellets clad in zircaloy-4 and irradiated in the neutron trap. The variations of the average neutron flux and the effect of self-shielding were studied. Similar calculations were carried out, both theoretically and experimentally for samples of europium oxide. Self-shielding effects were studied, and the variation of the effective multiplication factor was found as function of mass. The isotope generation and depletion code origin was used to compute the radioactivity of fission products from irradiating uranium different enrichments in IRT-5000. The effect of self-shielding on the flux and on the activities were found also. 14 tabs.; 34 figs.; 27 refs
Self shielding in cylindrical fissile sources in the APNea system
Hensley, D.
1997-02-01
In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results.
Resonance self-shielding corrections for activation cross section measurements
The Pade approximations of the Doppler broadening function ψ(θ, x) have been used for the calculations of resonance self-shielding factors used in activation measurements. It is shown that this method of the calculations is effective from the point of view of fastness and accuracy. (author)
Calculation of resonance self-shielding in 235U
The self-shielded fission rates for 235U measured by Czirr have been computed in the 0.464- to 464-eV range with the 235U resolved resonance parameters of the current version of ENDF/B-VI. The main purpose of describing the neutron cross sections of the fissile isotopes with resonance parameters is for the calculation of resonance self-shielding in reactors. Czirr has measured the relative self-shielded fission rate behind absorber thicknesses varying from 0.5 to 19 g/cm2 of uranium. These measurements have repeatedly been used to test the validity of sets of resonance parameters. However, previous calculations of these experiments were limited and inconclusive. An evaluation of the 235U resolved resonance region extending to 2.2 keV was recently completed and incorporated into ENDF/B-VI. The resonance parameters of this evaluation were used to compute the self-shielded fission rates for energy groups 0 to 8 (energies from 0.464 to 464 eV) of the Czirr measurement
Self-shielded electron linear accelerators designed for radiation technologies
Belugin, V. M.; Rozanov, N. E.; Pirozhenko, V. M.
2009-09-01
This paper describes self-shielded high-intensity electron linear accelerators designed for radiation technologies. The specific property of the accelerators is that they do not apply an external magnetic field; acceleration and focusing of electron beams are performed by radio-frequency fields in the accelerating structures. The main characteristics of the accelerators are high current and beam power, but also reliable operation and a long service life. To obtain these characteristics, a number of problems have been solved, including a particular optimization of the accelerator components and the application of a variety of specific means. The paper describes features of the electron beam dynamics, accelerating structure, and radio-frequency power supply. Several compact self-shielded accelerators for radiation sterilization and x-ray cargo inspection have been created. The introduced methods made it possible to obtain a high intensity of the electron beam and good performance of the accelerators.
Unresolved resonance self shielding calculation: causes and importance of discrepancies
To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, we do not know the parameters of each level but only the average parameters. Therefore we simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the X2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, we will survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors. 8 refs
Byoun, T. Y.; Block, R. C.; Semler, T. T.
1972-01-01
A series of average transmission and average self-indication ratio measurements were performed in order to investigate the temperature dependence of the resonance self-shielding effect in the unresolved resonance region of depleted uranium and tantalum. The measurements were carried out at 77 K, 295 K and approximately 1000 K with sample thicknesses varying from approximately 0.1 to 1.0 mean free path. The average resonance parameters as well as the temperature dependence were determined by using an analytical model which directly integrates over the resonance parameter distribution functions.
Measurement of resonance self-shielding factors of neutron capture cross section by 238U
Resonance self-shielding factors fsub(c) of neutron capture cross section by 238U in the 20-100 keV energy range are measured. The method for determining the fsub(c) factor consists in measuring partial transmission and transmission in the total cross section at different 238U filter thickness. The fsub(c) factor values in the 46.5-100 and 21.5-46.5 keV energy ranges are equal to 0.89+-0.03 and 0.81+-0.04, respectively
Resonance Self-Shielding Methodologies in SCALE 6
SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.
Resonance Self-Shielding Methodologies in SCALE 6
Williams, Mark L [ORNL
2011-01-01
SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.
Calculation of resonance self-shielding for 235U from 0 to 2250 eV
Over the years, the evaluated 235U cross sections in the resolved energy range have been extensively revised. A major accomplishment was the first evaluation released to the ENDF/B-VI library. In that evaluation, the low energy range bound was lowered to 10-5 eV, and the upper limit raised to 2,250 eV. Several high-resolution measurements in conjunction with the Bayesian computer code SAMMY were used to perform the analysis of the 235U resonance parameters. SAMMY uses the Reich-Moore formalism, which is adequate for representing neutron cross sections of fissile isotopes, and a generalized least-squares (Bayes) technique for determining the energy-dependence of the neutron cross sections. Recently a re-evaluation of the 235U cross section in the resolved resonance region was completed. This evaluation has undergone integral tests in various laboratories throughout the USA and abroad. The evaluation has been accepted for inclusion in ENDF/B-VI release 5. The intent of this work is to present results of calculations of self-shielded fission rates carried out with these resonance parameters and to compare those fission rates with experimental data. Results of this comparison study provide an assessment of the resonance parameters with respect to the calculation of self-shielded group cross sections
Kim, Jong Woon; Kim, Sang Ji; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of {sup 239}Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis.
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of 239Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis
The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling by Doppler broadened cross-sections. The various self-shielding factors are computer numerically as Lebesgue integrals over the cross-section probability tables
A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)
Formation Mechanism of Inclusion in Self-Shielded Flux Cored Arc Welds
YU Ping; LU Xiao-sheng; PAN Chuan; XUE Jin; LI Zheng-bang
2005-01-01
The formation mechanism of inclusion in welds with different aluminum contents was determined based on thermodynamic equilibrium in self-shielded flux cored arc welds. Inclusions in welds were systematically studied by optical microscopy, scanning microscopy and image analyzer. The results show that the average size and the contamination rate of inclusions in low-aluminum weld are lower than those in high-aluminum weld. Highly faceted AlN inclusions with big size in the high-aluminum weld are more than those in low-aluminum weld. As a result,the low temperature impact toughness of low-aluminum weld is higher than that of high-aluminum weld. Finally,the thermodynamic analysis indicates that thermodynamic result agrees with the experimental data.
The Fortran IV code PAPIN has been developed to calculate cross section probability tables, Bondarenko self-shielding factors and average self-indication ratios for non-fissile isotopes, below the inelastic threshold, on the basis of the ENDF/B prescriptions for the unresolved resonance region. Monte-Carlo methods are utilized to generate ladders of resonance parameters in the unresolved resonance region, from average resonance parameters and their appropriate distribution functions. The neutron cross-sections are calculated by the single level Breit-Wigner (SLBW) formalism, with s, p and d-wave contributions. The cross section probability tables are constructed by sampling the Doppler-broadened cross sections. The various self-shielded factors are computed numerically as Lebesgue integrals over the cross section probability tables. The program PAPIN has been validated through extensive comparisons with several deterministic codes
Munoz-Cobos, J. G.
1981-08-01
A FORTRAN 4 code was developed to calculate cross section probability tables, Bondarenko self-shielding factors, and average self-indication ratios for non-fissile isotopes, below the inelastic threshold, on the basis of prescriptions for the unresolved resonance region. Monte-Carlo methods are utilized to generate ladders of resonance parameters in he unresolved resonance region, from average resonance parameters and their appropriate distribution functions. The neutron cross sections are calculated by the single level Breit-Wigner formalism, with s, p and d-wave contributions. The cross section probability tables are constructed by sampling the Doppler-broadened cross sections. The various self-shielded factors are computed numerically as Lebesgue integrals over the cross section probability tables. The program was validated through extensive comparisons with several deterministic codes.
The gas-cooled, high temperature reactor (HTR) represents a valuable option for the future development of nuclear technology, because of its excellent safety features. One main safety feature is the negative temperature coefficient which is due to the Doppler broadening of the (n,y) resonance absorption cross section. A second important effect is the spatial self-shielding due to the double heterogeneous geometry of a pebble bed reactor. At FZ-Juelich two reactor analysis codes have been developed: VSOP for core design and MGT for transient analysis. Currently an update of the nuclear cross section libraries to ENDF/B-VII.0 of both codes takes place. In order to take the temperature dependency as well as the spatial self-shielding into account the absorption cross sections σ(n,y) for the resonance absorbers like 232Th and 238U have to be provided as function of incident neutron energy, temperature and nuclide concentration. There are two reasons for choosing the Monte-Carlo approach to calculate group wise cross sections. First, the former applied ZUT-DGL code to generate the resonance cross section tables for MGT is so far not able to handle the new resonance description based on Reich-Moore instead of Single-level Breit-Wigner. Second, the rising interest in PuO2 fuel motivated an investigation on the generation of group wise cross sections describing thermal resonances of 240Pu and 242Pu. (orig.)
Self-shielding effect of a single phase liquid xenon detector for direct dark matter search
Minamino, A; Ashie, Y; Hosaka, J; Ishihara, K; Kobayashi, K; Koshio, Y; Mitsuda, C; Moriyama, S; Nakahata, M; Nakajima, Y; Namba, T; Ogawa, H; Sekiya, H; Shiozawa, M; Suzuki, Y; Takeda, A; Takeuchi, Y; Taki, K; Ueshima, K; Ebizuka, Y; Ota, A; Suzuki, S; Hagiwara, H; Hashimoto, Y; Kamada, S; Kikuchi, M; Kobayashi, N; Nagase, T; Nakamura, S; Tomita, K; Uchida, Y; Fukuda, Y; Sato, T; Nishijima, K; Maruyama, T; Motoki, D; Itow, Y; Kim, Y D; Lee, J I; Moon, S H; Lim, K E; Cravens, J P; Smy, M B
2009-01-01
Liquid xenon is a suitable material for a dark matter search. For future large scale experiments, single phase detectors are attractive due to their simple configuration and scalability. However, in order to reduce backgrounds, they need to fully rely on liquid xenon's self-shielding property. A prototype detector was developed at Kamioka Observatory to establish vertex and energy reconstruction methods and to demonstrate the self-shielding power against gamma rays from outside of the detector. Sufficient self-shielding power for future experiments was obtained.
Study on the Processing Method for Resonance Self-shielding Calculations
无
2011-01-01
We investigate a new approach for resonance self-shielding calculations, based on a straightforward subgroup method, used in association with characteristics method. Subgroup method is actually the subdivision of cross section range for resonance energy range.
Self-shielding effect of a single phase liquid xenon detector for direct dark matter search
Minamino, A.; Abe, K.; Ashie, Y.; Hosaka, J.; Ishihara, K; Kobayashi, K; Koshio, Y.; Mitsuda, C.; Moriyama, S.; Nakahata, M.(University of Tokyo, Institute for Cosmic Ray Research, Kamioka Observatory, Kamioka, Japan); Nakajima, Y; Namba, T.; Ogawa, H.; Sekiya, H.; Shiozawa, M
2009-01-01
Liquid xenon is a suitable material for a dark matter search. For future large scale experiments, single phase detectors are attractive due to their simple configuration and scalability. However, in order to reduce backgrounds, they need to fully rely on liquid xenon's self-shielding property. A prototype detector was developed at Kamioka Observatory to establish vertex and energy reconstruction methods and to demonstrate the self-shielding power against gamma rays from outside of the detecto...
Effect of energy self-shielding methods on 238U for criticality safety problems
Multigroup cross sections are obtained by weighting point-wise nuclear data with a flux spectrum. For nuclides having a resonance structure, energy self-shielding calculations are performed to calculate a more detailed flux spectrum. Subsequently, self-shielded multigroup cross sections are generated. Different methods exist for energy self-shielding calculations. Among them are the Bondarenko method, the NJOY flux calculator, and the CENTRM method. The CENTRM method is a more advanced technique that utilizes both multigroup and point-wise cross sections in a one-dimensional transport calculation to solve for a point-wise flux distribution. The method of energy self-shielding is one of the elements in a multigroup cross-section generation that may have a significant impact on the multiplication factor in criticality safety calculations. This paper compares the three aforementioned self-shielding methods applied to 238U. A criticality problem having 23 cases is considered. This system includes water-moderated, low-enriched UO2 fuel rods in square-pitched array, with a thermal flux spectrum. Multiplication factors obtained from transport calculations that use multigroup and continuous energy data are compared. It is observed that multiplication factors calculated with multigroup data containing different self-shielding methods for 238U have less than 500 pcm difference with continuous energy results. (authors)
Shavers, M. R.; Atwell, W.; Cucinotta, F. A.; Badhwar, G. D. (Technical Monitor)
1999-01-01
cell killing from GCR, including patterns of cell killing from single particle tracks. can provide useful information on expected differences between proton and HZE tracks and clinical experiences with photon irradiation. To model effects on cells in the brain, it is important that transport models accurately describe changes in the GCR due to interactions in the cranium and proximate tissues. We describe calculations of the attenuated GCR particle fluxes at three dose-points in the brain and associated patterns of cell killing using biophysical models. The effects of the brain self-shielding and bone-tissue interface of the skull in modulating the GCR environment are considered. For each brain dose-point, the mass distribution in the surrounding 4(pi) solid angle is characterized using the CAM model to trace 512 rays. The CAM model describes the self-shielding by converting the tissue distribution to mass-equivalent aluminum, and nominal values of spacecraft shielding is considered. Particle transport is performed with the proton, neutron, and heavy-ion transport code HZETRN with the nuclear fragmentation model QMSFRG. The distribution of cells killed along the path of individual GCR ions is modeled using in vitro cell inactivation data for cells with varying sensitivity. Monte Carlo simulations of arrays of inactivated cells are considered for protons and heavy ions and used to describe the absolute number of cell killing events of various magnitude in the brain from the GCR. Included are simulations of positions of inactivated cells from stopping heavy ions and nuclear stars produced by high-energy ions most importantly, protons and neutrons.
Hartwig, Tilman; Klessen, Ralf S; Latif, Muhammad A; Volonteri, Marta
2015-01-01
The highest redshift quasars at z>6 have mass estimates of about a billion M$_\\odot$. One of the pathways to their formation includes direct collapse of gas, forming a supermassive star ($\\sim 10^5\\,\\mathrm{M}_\\odot$) precursor of the black hole seed. The conditions for direct collapse are more easily achievable in metal-free haloes, where atomic hydrogen cooling operates and molecular hydrogen (H$_2$) formation is inhibited by a strong external UV flux. Above a certain value of UV flux ($J_{\\rm crit}$), the gas in a halo collapses isothermally at $\\sim10^4$K and provides the conditions for supermassive star formation. However, H$_2$ can self-shield and the effect of photodissociation is reduced. So far, most numerical studies used the local Jeans length to calculate the column densities for self-shielding. We implement an improved method for the determination of column densities in 3D simulations and analyse its effect on the value of $J_{\\rm crit}$. This new method captures the gas geometry and velocity fie...
Zeng Huilin
2014-10-01
Full Text Available In order to realize the automatic welding of pipes in a complex operation environment, an automatic welding system has been developed by use of all-position self-shielded flux cored wires due to their advantages, such as all-position weldability, good detachability, arc's stability, low incomplete fusion, no need for welding protective gas or protection against wind when the wind speed is < 8 m/s. This system consists of a welding carrier, a guide rail, an auto-control system, a welding source, a wire feeder, and so on. Welding experiments with this system were performed on the X-80 pipeline steel to determine proper welding parameters. The welding technique comprises root welding, filling welding and cover welding and their welding parameters were obtained from experimental analysis. On this basis, the mechanical properties tests were carried out on welded joints in this case. Results show that this system can help improve the continuity and stability of the whole welding process and the welded joints' inherent quality, appearance shape, and mechanical performance can all meet the welding criteria for X-80 pipeline steel; with no need for windbreak fences, the overall welding cost will be sharply reduced. Meanwhile, more positive proposals were presented herein for the further research and development of this self-shielded flux core wires.
Calculation of resonance self-shielding for {sup 235}U from 0 to 2250 eV
Leal, L.C.; Larson, N.M.; Derrien, H. [Oak Ridge National Lab., TN (United States); Santos, G.R. [Cidade Univ., Rio de Janeiro (Brazil). Inst. de Engenharia Nuclear
1998-08-01
Over the years, the evaluated {sup 235}U cross sections in the resolved energy range have been extensively revised. A major accomplishment was the first evaluation released to the ENDF/B-VI library. In that evaluation, the low energy range bound was lowered to 10{sup {minus}5} eV, and the upper limit raised to 2,250 eV. Several high-resolution measurements in conjunction with the Bayesian computer code SAMMY were used to perform the analysis of the {sup 235}U resonance parameters. SAMMY uses the Reich-Moore formalism, which is adequate for representing neutron cross sections of fissile isotopes, and a generalized least-squares (Bayes) technique for determining the energy-dependence of the neutron cross sections. Recently a re-evaluation of the {sup 235}U cross section in the resolved resonance region was completed. This evaluation has undergone integral tests in various laboratories throughout the USA and abroad. The evaluation has been accepted for inclusion in ENDF/B-VI release 5. The intent of this work is to present results of calculations of self-shielded fission rates carried out with these resonance parameters and to compare those fission rates with experimental data. Results of this comparison study provide an assessment of the resonance parameters with respect to the calculation of self-shielded group cross sections.
Sohail, Muhammad; Kim, Myunghyun [Kyung Hee Univ., Yongin (Korea, Republic of)
2013-05-15
It has the applicability for the cases of arbitrary geometry or direct whole-core transport calculation. Conventionally in subgroup method the subgroup data is generated without considering resonance interference and is therefore included at the use of subgroup data. A modification in subgroup method to consider resonance interference explicitly in more consistent way has been proposed in this study. Owing to the fact that these self-shielded cross-sections in interference term is also lethargy dependent, it can be converted to subgroup level dependent self-shielded cross-sections. The proposed method is implemented in 3-D whole core transport lattice code nTRACER. More consistent method of resonance interference interaction has shown relatively negligible error in self shielded cross-section. This new interference treatment method is investigated at various temperatures and has shown better results regardless of temperature changes of mixture of resonance isotopes mixture.
Resonance self-shielding in the blanket of a hybrid reactor
Three sets of energy group cross sections were obtained using various approximations for resonance self shielding. The three models used in obtaining the cross sections were: (a) infinitely dilute model, (b) homogeneous-medium resonance self shielding, and (c) heterogeneous-medium resonance self shielding. The effects on the blanket performance of fusion--fission hybrid reactors, and in particular, on the performance of the current reference Westinghouse Demonstration Tokamak Hybrid Reactor blanket, were compared and analyzed for a variety of fuel-coolant combinations. It has been concluded that (1) the infinitely dilute cross sections can be used to produce preliminary crude estimates for beginning-of-life (BOL) only, (2) the resonance absorber finite dilution should be considered for BOL, poorly moderated blankets and well moderated blankets with low fissile material content situations, and (3) the spacial details should be considered in high fissile content, well moderated blanket situations
Verification of effectiveness of borated water shield for a cyclotron type self-shielded
The technological advances in positron emission tomography (PET) in conventional clinic imaging have led to a steady increase in the number of cyclotrons worldwide. Most of these cyclotrons are being used to produce 18F-FDG, either for themselves as for the distribution to other centers that have PET. For there to be safety in radiological facilities, the cyclotron intended for medical purposes can be classified in category I and category II, ie, self-shielded or non-shielded (bunker). Therefore, the aim of this work is to verify the effectiveness of borated water shield built for a cyclotron accelerator-type Self-shielded PETtrace 860. Mixtures of water borated occurred in accordance with the manufacturer’s specifications, as well as the results of the radiometric survey in the vicinity of the self-shielding of the cyclotron in the conditions established by the manufacturer showed that radiation levels were below the limits. (author)
Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes
As part of the self-shielding model used in the APOLLO2 lattice code, probability tables are required to compute self-shielded cross sections for coarse energy groups (typically with 99 or 172 groups). This paper describes the replacement of the multiband tables (typically with 51 subgroups) with moment-based tables in release 2.5 of APOLLO2. An improved Ribon method is proposed to compute moment-based probability tables, allowing important savings in CPU resources while maintaining the accuracy of the self-shielding algorithm. Finally, a validation is presented where the absorption rates obtained with each of these techniques are compared with exact values obtained using a fine-group elastic slowing-down calculation in the resolved energy domain. Other results, relative to the Rowland's benchmark and to three assembly production cases, are also presented
Self-shielding in the NET fusion reactor blanket and effects on uncertainty calculations
In this report the results are presented of an analysis of a NET iron/water inboard shielding blanket using energy self-shielded cross sections. Coupled (n,γ) transport calculations have been performed in an S8P8 approximation with the code ANISN using cross sections in the 121-group GAM-II structure. Basic cross sections were obtained from the 217-group MAT175 library, which is based on JEF-1 and EFF-1. Energy self-shielding was taken into account using the Bondarenko method. The results of this analysis are compared with those obtained in report ECN-C--90-034, in which a similar analysis had been presented using infinite dilute cross sections. It is shown that the effect of energy self-shielding on the neutron flux in the coils of the NET design is considerable, whereas the γ-flux is hardly influenced. Also the effect of using energy self-shielded cross sections in sensitivity and uncertainty analyses was studied. In these analyses the sensitivity of the total nuclear heating in the innermost interval of the inboard coils to the total cross sections of Fe, Cr and Ni has been studied. The analyses have been performed using an ECN-modified version of the code SENSIT. Due to the effect of self-shielding not only the value of the response parameter changes (the total nuclear heating in the coil increases with 13%), but also the associated relative uncertainty (the relative uncertainty due to uncertainties in the total cross sections of Fe, Cr and Ni decreases with 8%; the absolute value of the uncertainty increases however). The conclusion is that for reliable calculations of the nuclear heating the effects of self-shielding should be taken into account; for the uncertainty estimates this is less important. (author). 15 refs.; 15 figs.; 6 tabs
Multi-Determinant Wave-functions in Quantum Monte Carlo
Morales, Miguel A [Lawrence Livermore National Laboratory (LLNL); Mcminis, Jeremy [University of Illinois, Urbana-Champaign; Clark, Bryan K. [Princeton University; Kim, Jeongnim [ORNL; Scuseria, Gustavo E [Rice University
2012-01-01
Quantum Monte Carlo methods have received considerable attention over the last decades due to the great promise they have for the direct solution to the many-body Schrodinger equation for electronic systems. Thanks to a low scaling with number of particles, they present one of the best alternatives in the accurate study of large systems and solid state calculations. In spite of such promise, the method has not become popular in the quantum chemistry community, mainly due to the lack of control over the fixed-node error which can be large in many cases. In this article we present the application of large multi-determinant expansions in quantum Monte Carlo, studying its performance with first row dimers and the 55 molecules of the G1 test set. We demonstrate the potential of the wave-function to systematically reduce the fixed-node error in the calculations, achieving chemical accuracy in almost all cases studied. When compared to traditional methods in quantum chemistry, the results show a marked improvement over most methods including MP2, CCSD(T) and DFT with various functionals; in fact the only method able to produce better results is the explicitly-correlated CCSD(T) method with a large basis set. With recent developments in trial wave functions and algorithmic improvements in Quantum Monte Carlo, we are quickly approaching a time where the method can become the standard in the study of large molecular systems and solids.
SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation
Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2016-06-06
The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.
Improvement of accuracy of resonance self-shielding calculation based on subgroup method
Based on the neutron self-shielding calculation code SGMOC, which is the combination of subgroup method and characteristics method that developed by ourselves, we studied the two techniques to improve the SGMOC calculation accuracy. The numerical results prove that both techniques have the capability to improve the resonance self-shielding calculation accuracy. The resonance interference effect treatment which uses a new method to obtain the conditional probabilities, has a correction effect about 20∼230 pcm. When the impact of the resonance scattering is considered, the correction effect is about 100 Dcm. When utilizing the above two techniques simultaneously, the correction effect is about 30∼270 pcm. (authors)
Evaluation of resonance self-shielding factors for 238U in the unresolved resonance region
On the basis of a theoretical model of identical equidistant resonances for the energy dependence of cross-sections in the unresolved resonance region, the authors have parametrized the values of the resonance self-shielding factors and their Doppler increments for 238U. They have proposed a method by which the Doppler increments of the self-shielding factors can be calculated from simple analytical formulae by redetermination of the model parameters. Analysing the experimental data on direct and capture transmissions in the unresolved resonance region, they demonstrate the possibility of describing those data as a whole and of deriving from them the cross-section group functionals. (author)
A Direct Iteration Method using Resonance Integral Table for the Self-Shielding Calculations
In this paper, a direct iteration method using the resonance integral table is introduced for the self-shielding calculations. The basic purpose of this paper is to show the possibility that the HELIOS subgroup method can be replaced with this method. This method doesn't use the subgroup data but only the resonance integral tables given in library. The basic idea of this method is to use the Bondarenko's iteration in order to obtain the self-shielded effective cross sections with the background cross sections which are calculated by the heterogeneous transport calculation. This method is implemented in the KARMA lattice calculation code and tested
A Direct Iteration Method using Resonance Integral Table for the Self-Shielding Calculations
Hong, Ser Gi; Kim, Kang Seog; Song, Jae Seung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2009-10-15
In this paper, a direct iteration method using the resonance integral table is introduced for the self-shielding calculations. The basic purpose of this paper is to show the possibility that the HELIOS subgroup method can be replaced with this method. This method doesn't use the subgroup data but only the resonance integral tables given in library. The basic idea of this method is to use the Bondarenko's iteration in order to obtain the self-shielded effective cross sections with the background cross sections which are calculated by the heterogeneous transport calculation. This method is implemented in the KARMA lattice calculation code and tested.
License Application Design Selection Feature Report: Waste Package Self Shielding Design Feature 13
In the Viability Assessment (VA) reference design, handling of waste packages (WPs) in the emplacement drifts is performed remotely, and human access to the drifts is precluded when WPs are present. This report will investigate the feasibility of using a self-shielded WP design to reduce the radiation levels in the emplacement drifts to a point that, when coupled with ventilation, will create an acceptable environment for human access. This provides the benefit of allowing human entry to emplacement drifts to perform maintenance on ground support and instrumentation, and carry out performance confirmation activities. More direct human control of WP handling and emplacement operations would also be possible. However, these potential benefits must be weighed against the cost of implementation, and potential impacts on pre- and post-closure performance of the repository and WPs. The first section of this report will provide background information on previous investigations of the self-shielded WP design feature, summarize the objective and scope of this document, and provide quality assurance and software information. A shielding performance and cost study that includes several candidate shield materials will then be performed in the subsequent section to allow selection of two self-shielded WP design options for further evaluation. Finally, the remaining sections will evaluate the impacts of the two WP self-shielding options on the repository design, operations, safety, cost, and long-term performance of the WPs with respect to the VA reference design
Experimental evaluation of epithermal neutron self-shielding for 96Zr and 98Mo
In a previous work we experimentally tested some neutron self-shielding calculations methods for thermal absorbers, from which the semi-empirical 'sigmoid method' gave the most accurate results. In this work we aim at evaluating the accuracy of this method on the epithermal self-shielding phenomena as compared to the analytical 'MatSSF method'. Metallic foils of Zr and Mo were compactly stacked together into small cylinders (or disks) of different thickness, allowing for up to 20 % epithermal self-shielding when irradiated on two channels of the BR1 reactor. A 2 % relative difference between calculated and experimental self-shielding factors was obtained from the MatSSF method when a perpendicular source-sample axial configuration was assumed, while the isotropic or the co-axial configuration alternatives gave up to 10 % relative differences. On the other hand, the sigmoid method gave relative differences of up to 6 % that can be reduced to just 2 % by applying the 'effective' epithermal absorption cross-sections for 98Mo and 96Zr proposed in this work. (author)
Advanced resonance self-shielding method for gray resonance treatment in lattice physics code GALAXY
A new resonance self-shielding method based on the equivalence theory is developed for general application to the lattice physics calculations. The present scope includes commercial light water reactor (LWR) design applications which require both calculation accuracy and calculation speed. In order to develop the new method, all the calculation processes from cross-section library preparation to effective cross-section generation are reviewed and reframed by adopting the current enhanced methodologies for lattice calculations. The new method is composed of the following four key methods: (1) cross-section library generation method with a polynomial hyperbolic tangent formulation, (2) resonance self-shielding method based on the multi-term rational approximation for general lattice geometry and gray resonance absorbers, (3) spatially dependent gray resonance self-shielding method for generation of intra-pellet power profile and (4) integrated reaction rate preservation method between the multi-group and the ultra-fine-group calculations. From the various verifications and validations, applicability of the present resonance treatment is totally confirmed. As a result, the new resonance self-shielding method is established, not only by extension of a past concentrated effort in the reactor physics research field, but also by unification of newly developed unique and challenging techniques for practical application to the lattice physics calculations. (author)
T. Downar
2009-03-31
The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system.
Characterization and dosimetry of a practical X-ray alternative to self-shielded gamma irradiators
Mehta, Kishor; Parker, Andrew
2011-01-01
The Insect Pest Control Laboratory of the Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture recently purchased an X-ray irradiator as part of their programme to develop the sterile insect technique (SIT). It is a self-contained type with a maximum X-ray beam energy of 150 keV using a newly developed 4 π X-ray tube to provide a very uniform dose to the product. This paper describes the results of our characterization study, which includes determination of dose rate in the centre of a canister as well as establishing absorbed dose distribution in the canister. The irradiation geometry consists of five canisters rotating around an X-ray tube—the volume of each canister being 3.5 l. The dose rate at the maximum allowed power of the tube (about 6.75 kW) in the centre of a canister filled with insects (or a simulated product) is about 14 Gy min -1. The dose uniformity ratio is about 1.3. The dose rate was measured using a Farmer type 0.18-cm 3 ionization chamber calibrated at the relevant low photon energies. Routine absorbed dose measurement and absorbed dose mapping can be performed using a Gafchromic® film dosimetry system. The radiation response of Gafchromic film is almost independent of X-ray energy in the range 100-150 keV, but is very sensitive to the surrounding material with which it is in immediate contact. It is important, therefore, to ensure that all absorbed dose measurements are performed under identical conditions to those used for the calibration of the dosimetry system. Our study indicates that this X-ray irradiator provides a practical alternative to self-shielded gamma irradiators for SIT programmes. Food and Agriculture Organization/International Atomic Energy Agency.
Characterization and dosimetry of a practical X-ray alternative to self-shielded gamma irradiators
Mehta, Kishor, E-mail: mehta@aon.a [Insect Pest Control Laboratory, Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture, International Atomic Energy Agency, Vienna (Austria); Parker, Andrew [Insect Pest Control Laboratory, Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture, International Atomic Energy Agency, Vienna (Austria)
2011-01-15
The Insect Pest Control Laboratory of the Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture recently purchased an X-ray irradiator as part of their programme to develop the sterile insect technique (SIT). It is a self-contained type with a maximum X-ray beam energy of 150 keV using a newly developed 4{pi} X-ray tube to provide a very uniform dose to the product. This paper describes the results of our characterization study, which includes determination of dose rate in the centre of a canister as well as establishing absorbed dose distribution in the canister. The irradiation geometry consists of five canisters rotating around an X-ray tube-the volume of each canister being 3.5 l. The dose rate at the maximum allowed power of the tube (about 6.75 kW) in the centre of a canister filled with insects (or a simulated product) is about 14 Gy min{sup -1}. The dose uniformity ratio is about 1.3. The dose rate was measured using a Farmer type 0.18-cm{sup 3} ionization chamber calibrated at the relevant low photon energies. Routine absorbed dose measurement and absorbed dose mapping can be performed using a Gafchromic (registered) film dosimetry system. The radiation response of Gafchromic film is almost independent of X-ray energy in the range 100-150 keV, but is very sensitive to the surrounding material with which it is in immediate contact. It is important, therefore, to ensure that all absorbed dose measurements are performed under identical conditions to those used for the calibration of the dosimetry system. Our study indicates that this X-ray irradiator provides a practical alternative to self-shielded gamma irradiators for SIT programmes.
Characterization and dosimetry of a practical X-ray alternative to self-shielded gamma irradiators
The Insect Pest Control Laboratory of the Joint FAO/IAEA Division of Nuclear Techniques in Food and Agriculture recently purchased an X-ray irradiator as part of their programme to develop the sterile insect technique (SIT). It is a self-contained type with a maximum X-ray beam energy of 150 keV using a newly developed 4π X-ray tube to provide a very uniform dose to the product. This paper describes the results of our characterization study, which includes determination of dose rate in the centre of a canister as well as establishing absorbed dose distribution in the canister. The irradiation geometry consists of five canisters rotating around an X-ray tube-the volume of each canister being 3.5 l. The dose rate at the maximum allowed power of the tube (about 6.75 kW) in the centre of a canister filled with insects (or a simulated product) is about 14 Gy min-1. The dose uniformity ratio is about 1.3. The dose rate was measured using a Farmer type 0.18-cm3 ionization chamber calibrated at the relevant low photon energies. Routine absorbed dose measurement and absorbed dose mapping can be performed using a Gafchromic (registered) film dosimetry system. The radiation response of Gafchromic film is almost independent of X-ray energy in the range 100-150 keV, but is very sensitive to the surrounding material with which it is in immediate contact. It is important, therefore, to ensure that all absorbed dose measurements are performed under identical conditions to those used for the calibration of the dosimetry system. Our study indicates that this X-ray irradiator provides a practical alternative to self-shielded gamma irradiators for SIT programmes.
The up-scattering treatment in the fine-structure self-shielding method in APOLLO3®
The use of the exact elastic scattering in resonance domain introduces the neutron up-scattering which must be taken into account in the deterministic transport code. We present the newly implemented up-scattering treatment in the fine-structure self-shielding method of APOLLO3®. Two pin cell calculations have been carried out in order to evaluate the impact of the up-scattering treatment. The results are compared to those obtained by the Monte Carlo code TRIPOLI-4® with its newly implemented DBRC model. The comparison of k-eff values on the examples of single cell calculations shows a very good agreement between the APOLLO3® up-scattering treatment and the TRIPOLI-4® DBRC model, which is less than 30 pcm for UOX fuel and less than 110 pcm for MOX. Also, the differential effects of asymptotic versus exact kernel produced by APOLLO3® compared to TRIPOLI-4®, do not exceed 20 pcm for the UOX cell and 40 pcm for the MOX cell. A detailed comparison of the U238 absorption rates shows clearly the influences of the first four big resonances of U238 to the calculation results. (author)
A Monte Carlo simulation technique to determine the optimal portfolio
Hassan Ghodrati
2014-03-01
Full Text Available During the past few years, there have been several studies for portfolio management. One of the primary concerns on any stock market is to detect the risk associated with various assets. One of the recognized methods in order to measure, to forecast, and to manage the existing risk is associated with Value at Risk (VaR, which draws much attention by financial institutions in recent years. VaR is a method for recognizing and evaluating of risk, which uses the standard statistical techniques and the method has been used in other fields, increasingly. The present study has measured the value at risk of 26 companies from chemical industry in Tehran Stock Exchange over the period 2009-2011 using the simulation technique of Monte Carlo with 95% confidence level. The used variability in the present study has been the daily return resulted from the stock daily price change. Moreover, the weight of optimal investment has been determined using a hybrid model called Markowitz and Winker model in each determined stocks. The results showed that the maximum loss would not exceed from 1259432 Rials at 95% confidence level in future day.
Combination of self-shielded and gas-shielded flux-cored arc welding
Lian, Atle Korsnes
2011-01-01
This master thesis have consisted of experimental and theoretical studies of the change in microstructure and mechanical properties in intermixed weld metal from self-shielded and gas-shielded flux-cored welding wires. The main objective of the present thesis has been to do detailed metallographic analysis on different weld metal combinations, and find out and give an explanation why satisfying values were achieved or not achieved.The report is divided into four parts. Part one consists of re...
BONAMI-S: resonance self-shielding by the Bondarenko method
BONAMI-S is a module of the SCALE system which is used to perform Bondarenko calculations for resonance self-shielding. Cross sections and Bondarenko factor data are input from an AMPX master library. The output is written as an AMPX master library. A wide variety of options is provided for different lattices and cell geometries through the use of Dancoff approximations. A novel interpolational scheme is used which avoids many of the problems of the widely employed Lagrangian schemes
Importance of self-shielding for improving sensitivity coefficients in light water nuclear reactors
Highlights: • A new method has been developed for calculating sensitivity coefficients. • This method is based on the use of infinite dilution cross-sections instead of effective cross-sections. • The change of self-shielding factor due to cross-section perturbation has been considered. • SRAC and SAINT codes are used for calculating improved sensitivities, while MCNP code has been used for verification. - Abstract: In order to perform sensitivity analyzes in light water reactors where self-shielding effect becomes important, a new method has been developed for calculating sensitivity coefficient of core characteristics relative to the infinite dilution cross-sections instead of the effective cross-sections. This method considers the change of the self-shielding factor due to cross-section perturbation for different nuclides and reactions. SRAC and SAINT codes are used to calculate the improved sensitivity; while the accuracy of the present method has been verified by MCNP code and good agreement has been found
Modeling resonance interference by 0-D slowing-down solution with embedded self-shielding method
Liu, Y.; Martin, W. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI, 48109 (United States); Kim, K. S.; Williams, M. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States)
2013-07-01
The resonance integral table based methods employing conventional multigroup structure for the resonance self-shielding calculation have a common difficulty on treating the resonance interference. The problem arises due to the lack of sufficient energy dependence of the resonance cross sections when the calculation is performed in the multigroup structure. To address this, a resonance interference factor model has been proposed to account for the interference effect by comparing the interfered and non-interfered effective cross sections obtained from 0-D homogeneous slowing-down solutions by continuous-energy cross sections. A rigorous homogeneous slowing-down solver is developed with two important features for reducing the calculation time and memory requirement for practical applications. The embedded self-shielding method (ESSM) is chosen as the multigroup resonance self-shielding solver as an integral component of the interference method. The interference method is implemented in the DeCART transport code. Verification results show that the code system provides more accurate effective cross sections and multiplication factors than the conventional interference method for UO{sub 2} and MOX fuel cases. The additional computing time and memory for the interference correction is acceptable for the test problems including a depletion case with 87 isotopes in the fuel region. (authors)
The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM()NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.
A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the Keff, neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.
Zou Jun, E-mail: jzou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)
2010-12-15
A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K{sub eff}, neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.
Effects of CeF3 on properties of self-shielded flux cored wire
Yu Ping; Tian Zhiling; Pan Chuan; Xue Jin
2006-01-01
Effects of CeF3 on properties of self-shielded flux cored wire including welding process, inclusions in weld metal and mechanical properties are systematically studied. Welding smoke and spatter are reduced with the addition of CeF3. The main non-metallic inclusions in weld metal are AlN and Al2 O3. CeF3 can refine non-metallic inclusions and reduce the amount of large size inclusions, which is attributed to the inclusion floating behavior during the solidification of weld metal. The low temperature impact toughness is improved by adding suitable amount of CeF3 in the flux.
Procedures for the verification of the self-shielding of Cyclotron PET trace
According to the literature, shielding can be defined as a physical entity interposed between the ionizing radiation source and an object to be protected so that the level of radiation is reduced in the position where the object is (C hilton et al., 1984). Regarding shielding, cyclotrons can be self-shielded or not. The first type has a heavy armor built around it, while the bunker-type cyclotron must have additional structural shields. Those are required to reduce the radiation levels within safety limits, according to regulatory agencies. Therefore, it is important that the shielding is properly designed and installed, since corrections or additions are generally expensive after the installation is complete. For the same reason, planning should also take into account possible future modifications. For instance, use of higher radiation energies, the increasing of the beam intensity, use of different types of accelerated particles, and an increase in work load (NCRP No. 144, 2005). The objective of this study consists of verifying the effectiveness of shielding of borated water built for a self-shielded cyclotron accelerator PET trace 860.The self-shielding of PET trace cyclotron is composed of eight tanks. Each tank was filled with a mixture of water with 3.5% of boron and locally coated with lead plates and bricks. In the end of the preparation of each mixture the indication of the solution ph was measured to ensure that it was a neutral ph. Shielding verification was performed by using the radiometric survey provided with the condition of an irradiation of 40 uA of protons in a target of H2O18 (98% purity), with a neutron and gamma detector. Measurements were taken at different points around the shielding and through the radiometric survey around the installation. Due to sky shine phenomenon, verification was also performed with gamma and neutron monitors in the condition of simultaneous irradiation of two targets of H2O18 (98% purity) with 50 uA of protons on
BONAMI-S: response self-shielding by the Bondarenko method. Volume 2. Functional studies
BONAMI-S is a module of the SCALE (standardized computer analyses for licensing evaluation) system which is used to perform Bondarenko calculations for resonance self-shielding. Cross sections and Bondarenko factor data are input from an AMPX master library. The output is written as an AMPX master library. A wide variety of options is provided for different lattices and cell geometries through the use of Dancoff approximations. A novel interpolational scheme is used which avoids many of the problems of the widely employed Lagrangian schemes
CREST : a computer program for the calculation of composition dependent self-shielded cross-sections
A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)
ZZ ABBN, 26 Group Cross-Sections and Self Shielding Factors for Fast Reactors
1 - Description of program or function: Format: special format; Number of groups: 26 group X-section and resonance self-shielding factor library. Nuclides: H, D, Li-6, Li-7, Be, B-10, B-11, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, V, Cr, Fe, Ni, Cu, Zr, Nb, Mo, Ta, W, Re, Pb, Bi, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, FP-U-233, FP-U-235, FP-Pu-239. Origin: Multiple experimental sources; Weighting spectrum: yes; 26 group cross section and resonance self-shielding factor library for the following materials: H, D, Li-6, Li-7, Be, B-10, B-11, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, V, Cr, Fe, Ni, Cu, Zr, Nb, Mo, Ta, W, Re, Pb, Bi, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, FP-U-233, FP-U-235, FP-Pu-239. 2 - Restrictions on the complexity of the problem: This group cross section library has been developed for fast and intermediate reactors
Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material
Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153Gd, 160Tb, and 161Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)
Covariances of the self-shielding factor and its temperature gradient for the uranium-238 neutron capture reaction have been evaluated from the resonance parameter covariance matrix and the sensitivity of the self-shielding factor and its temperature gradient to the resonance parameters. The resonance parameters and their covariance matrix for uranium-238 were taken from JENDL-3.3, while the sensitivity coefficients were calculated by varying resonance parameters and temperature. A set of computer code modules has been developed for the calculation of the sensitivity coefficients at numerous resonance levels. The present result shows that the correlation among resonance parameters yields a substantial contribution to the standard deviations of the self-shielding factor and its temperature gradient. In addition to the standard deviations of these quantities, their correlation matrices in the JFS-3 70 group structure are also obtained. (author)
Resolution and intensity in neutron spectrometry determined by Monte Carlo simulation
Dietrich, O.W.
1968-01-01
The Monte Carlo simulation technique was applied to the propagation of Bragg-reflected neutrons in mosaic single crystals. The method proved to be very useful for the determination of resolution and intensity in neutron spectrometers.......The Monte Carlo simulation technique was applied to the propagation of Bragg-reflected neutrons in mosaic single crystals. The method proved to be very useful for the determination of resolution and intensity in neutron spectrometers....
CO Self-Shielding as a Mechanism to Make 16 O-Enriched Solids in the Solar Nebula
Joseph A. Nuth, III; Natasha M. Johnson; Hugh G. M. Hill
2014-01-01
Photochemical self-shielding of CO has been proposed as a mechanism to produce solids observed in the modern, 16 O-depleted solar system. This is distinct from the relatively 16 O-enriched composition of the solar nebula, as demonstrated by the oxygen isotopic composition of the contemporary sun. While supporting the idea that self-shielding can produce local enhancements in 16 O-depleted solids, we argue that complementary enhancements of 16 O-enriched solids can also be produced via C 16 O-...
Self-shielding flex-circuit drift tube, drift tube assembly and method of making
Jones, David Alexander
2016-04-26
The present disclosure is directed to an ion mobility drift tube fabricated using flex-circuit technology in which every other drift electrode is on a different layer of the flex-circuit and each drift electrode partially overlaps the adjacent electrodes on the other layer. This results in a self-shielding effect where the drift electrodes themselves shield the interior of the drift tube from unwanted electro-magnetic noise. In addition, this drift tube can be manufactured with an integral flex-heater for temperature control. This design will significantly improve the noise immunity, size, weight, and power requirements of hand-held ion mobility systems such as those used for explosive detection.
Line Overlap and Self-Shielding of Molecular Hydrogen in Galaxies
Gnedin, Nickolay Y
2014-01-01
The effect of line overlap in the Lyman and Werner bands, often ignored in galactic studies of the atomic-to-molecular transition, greatly enhances molecular hydrogen self-shielding in low metallicity environments, and dominates over dust shielding for metallicities below about 10% solar. We implement that effect in cosmological hydrodynamics simulations with an empirical model, calibrated against the observational data, and provide fitting formulae for the molecular hydrogen fraction as a function of gas density on various spatial scales and in environments with varied dust abundance and interstellar radiation field. We find that line overlap, while important for detailed radiative transfer in the Lyman and Werner bands, has only a minor effect on star formation on galactic scales, which, to a much larger degree, is regulated by stellar feedback.
Investigation of the resonance self-shielding effect in the α-value of 239Pu
Gamma-spectra from 1 to 15 multiplicities were measured on 122 m flight path of the IBR-30 pulsed neutron booster using a 16-section liquid scintillation detector of a total volume of 80 l for two thin metallic radiator-samples of 239Pu in the presence of a filter-samples with four thickness: 0.3, 0.5, 1 and 2.3 mm in the neutron beam and without them. The multiplicity distribution spectra, the average multiplicities and the alpha value α = σγ/σf for 27 resonances in the energy region 7-170 eV and for the energy group in the region 4.65-2150 eV with and without filter-samples of 239Pu in the beam were obtained. The self-shielding effect in the alpha value and the average multiplicity in limits 5-40% of experimental value is observed
Design of a control system for self-shielded irradiators with remote access capability
With self-shielded irradiators like Gamma chambers, and Blood irradiators are being sold by BRIT to customers both within and outside the country, it has become necessary to improve the quality of service without increasing the overheads. The recent advances in the field of communications and information technology can be exploited for improving the quality of service to the customers. A state of the art control system with remote accessibility has been designed for these irradiators enhancing their performance. This will provide an easy access to these units wherever they might be located, through the Internet. With this technology it will now be possible to attend to the needs of the customers, as regards fault rectification, error debugging, system software update, performance testing, data acquisition etc. This will not only reduce the downtime of these irradiators but also reduce the overheads. (author)
Nuclear reactions and self-shielding effects of gamma-ray database for nuclear materials
A database for transmutation and radioactivity of nuclear materials is required for selection and design of materials used in various nuclear reactors. The database based on the FENDL/A-2.0 on the Internet and the additional data collected from several references has been developed in NRIM site of 'Data-Free-Way' on the Internet. Recently, the function predicted self-shielding effect of materials for γ-ray was added to this database. The user interface for this database has been constructed for retrieval of necessary data and for graphical presentation of the relation between the energy spectrum of neutron and neutron capture cross section. It is demonstrated that the possibility of chemical compositional change and radioactivity in a material caused by nuclear reactions can be easily retrieved using a browser such as Netscape or Explorer. (author)
A description on a control system on production line of 200 kV self-shielded electron accelerator coating solidification, including interface module assignment, the methods of mouse operation, chinese display in english DOS is presented. Although it is designed for special use, yet transformation for other purpose is very simple
2013-08-30
... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF STATE Culturally Significant Objects Imported for Exhibition Determinations: ``Venetian Glass by Carlo Scarpa: The... April 15, 2003, I hereby determine that the objects to be included in the exhibition ``Venetian Glass...
Lautenschlager, Gary J.
The parallel analysis method for determining the number of components to retain in a principal components analysis has received a recent resurgence of support and interest. However, researchers and practitioners desiring to use this criterion have been hampered by the required Monte Carlo analyses needed to develop the criteria. Two recent…
Validation of new sub-group algorithms for resonance self-shielding in heterogeneous structures
Within the European cooperation on fast reactors a new cell code, ECCO, is being developed. This code uses the sub-group method to handle the resonance self-shielding in complex heterogeneous structures. The ECCO code uses several different energy group schemes. The fine group library allows a reference calculation, while the broad group library permits faster ECCO runs for design calculations. Both fine and broad group libraries contain infinite dilution cross-sections and sub-group parameters, as it is the case for the library FGL5 used by the MURAL code and for the library CARNAVAL4 used by the HETAIRE code. Libraries for ECCO can be created by processing the JEF evaluated nuclear data files. The CALENDF code generates sub-group parameters by an original method while the code system THEMIS, derived from NJOY, is used to calculate infinite dilution cross-sections and the other nuclear data. The consistency between CALENDF and NJOY/THEMIS for the calculation of shielded cross-sections is sufficiently accurate. Results of calculations using these two libraries give good agreement for cells typical of the fast reactor programme. It will be then acceptable to use broad group calculations in cases where fine group calculations are too costly, for example when treating very complicated geometries
Resonance self-shielding methodology of new neutron transport code STREAM
This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)
Weld metal microstructures of hardfacing deposits produced by self-shielded flux-cored arc welding
The molten pool weld produced during self-shielded flux-cored arc welding (SSFCAW) is protected from gas porosity arising from oxygen and nitrogen by reaction ('killing') of these gases by aluminium. However, residual Al can result in mixed micro-structures of δ-ferrite, martensite and bainite in hardfacing weld metals produced by SSFCAW and therefore, microstructural control can be an issue for hardfacing weld repair. The effect of the residual Al content on weld metal micro-structure has been examined using thermodynamic modeling and dilatometric analysis. It is concluded that the typical Al content of about 1 wt% promotes δ-ferrite formation at the expense of austenite and its martensitic/bainitic product phase(s), thereby compromising the wear resistance of the hardfacing deposit. This paper also demonstrates how the development of a Schaeffler-type diagram for predicting the weld metal micro-structure can provide guidance on weld filler metal design to produce the optimum microstructure for industrial hardfacing applications.
Monte Carlo simulation: tool for the calibration in analytical determination of radionuclides
This work shows how is established the traceability of the analytical determinations using this calibration method. Highlights the advantages offered by Monte Carlo simulation for the application of corrections by differences in chemical composition, density and height of the samples analyzed. Likewise, the results obtained by the LVRA in two exercises organized by the International Agency for Atomic Energy (IAEA) are presented. In these exercises (an intercomparison and a proficiency test) all reported analytical results were obtained based on calibrations in efficiency by Monte Carlo simulation using the DETEFF program
Meric, N; Bor, D
1999-01-01
Scatter fractions have been determined experimentally for lucite, polyethylene, polypropylene, aluminium and copper of varying thicknesses using a polyenergetic broad X-ray beam of 67 kVp. Simulation of the experiment has been carried out by the Monte Carlo technique under the same input conditions. Comparison of the measured and predicted data with each other and with the previously reported values has been given. The Monte Carlo calculations have also been carried out for water, bakelite and bone to examine the dependence of scatter fraction on the density of the scatterer.
The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems
A computational method for calculating multigroup self-shielded cross sections in heterogeneous media containing arbitrary mixtures of resonant isotopes is presented. The method accounts for resonance interference between immixed resonant nuclei as well as for spatial resonance interference between resonant isotopes in different geometrical locations. A general correction is used to generate an intermediary reaction-rate library for resonant isotopic mixtures from a single-isotope, standard preprocessed library. Reaction rates for the heterogeneous fine-structure equation are computed from the intermediary library by invoking an equivalence theorem either on a group basis or using Bell's factors defined on macrogroups. Results are presented for a homogeneous mixture of Uranium oxide as well as for a recycled-fuel PWR cell. A study of the radial dependence of self-shielding for a recycled mixture of Uranium-Plutonium oxide in a PWR cell and in a submoderated cell is also included
Initial experience with an 11 MeV self-shielded medical cyclotron on operation and radiation safety
Pant G; Senthamizhchelvan S
2007-01-01
A self-shielded medical cyclotron (11 MeV) was commissioned at our center, to produce positron emitters, namely, 18 F, 15 O, 13 N and 11 C for positron emission tomography (PET) imaging. Presently the cyclotron has been exclusively used for the production of 18 F - for 18 F-FDG imaging. The operational parameters which influence the yield of 18 F - production were monitored. The radiation levels in the cyclotron and radiochemistry laboratory were also monitored to assess the radiatio...
CO Self-Shielding as a Mechanism to Make 16O-Enriched Solids in the Solar Nebula
Joseph A. Nuth, III
2014-05-01
Full Text Available Photochemical self-shielding of CO has been proposed as a mechanism to produce solids observed in the modern, 16O-depleted solar system. This is distinct from the relatively 16O-enriched composition of the solar nebula, as demonstrated by the oxygen isotopic composition of the contemporary sun. While supporting the idea that self-shielding can produce local enhancements in 16O-depleted solids, we argue that complementary enhancements of 16O-enriched solids can also be produced via C16O-based, Fischer-Tropsch type (FTT catalytic processes that could produce much of the carbonaceous feedstock incorporated into accreting planetesimals. Local enhancements could explain observed 16O enrichment in calcium-aluminum-rich inclusions (CAIs, such as those from the meteorite, Isheyevo (CH/CHb, as well as in chondrules from the meteorite, Acfer 214 (CH3. CO self-shielding results in an overall increase in the 17O and 18O content of nebular solids only to the extent that there is a net loss of C16O from the solar nebula. In contrast, if C16O reacts in the nebula to produce organics and water then the net effect of the self-shielding process will be negligible for the average oxygen isotopic content of nebular solids and other mechanisms must be sought to produce the observed dichotomy between oxygen in the Sun and that in meteorites and the terrestrial planets. This illustrates that the formation and metamorphism of rocks and organics need to be considered in tandem rather than as isolated reaction networks.
Zeng Huilin; Wang Changjiang; Yang Xuemei; Wang Xinsheng; Liu Ran
2014-01-01
In order to realize the automatic welding of pipes in a complex operation environment, an automatic welding system has been developed by use of all-position self-shielded flux cored wires due to their advantages, such as all-position weldability, good detachability, arc's stability, low incomplete fusion, no need for welding protective gas or protection against wind when the wind speed is
Continuing with the domestic 'Burnable Absorbers Research Plan' studies were done to estimate self-shielding effects during Gd2O3 burnup as burnable absorber included in fuel pins of a CAREM geometry. In this way, its burnup was calculated without and with self-shielding. For the second case, were obtained values depending on internal pin radius and the effective one for the homogenized pin. For Gd 157, the burnup corresponding to the first case resulted 52.6 % and of 1.23 % for the effective one. That shows the magnitude of the effects under study. Considering that is necessary to perform one experimental verification, also are presented calculational results for the case to irradiate a pellet containing UO2 (natural) and 8 wt % of Gd2O3, as a function of cooling time, that include: measurable isotopes concentrations, expected activities, and photon spectra for conditions able to be compared with bidimensional calculations with self-shielding. The irradiation time was supposed 30 dpp using RA-3 reactor at 10 MW. (author)
The effect of self-shielding of resonance cross sections on the tritium breeding ratio was investigated for three promising fusion blanket designs with liquid lithium, lithium oxide and lithium-lead breeders. Calculations were performed using ANISN and MCNP transport codes with the ENDF/B-V based nuclear data libraries. It is found that the self-shielding effect cannot be neglected in the blanket design if the blanket is neutron leaky in the case when the blanket is thin or with lower Li-6 enrichment in Li. This may result in an underestimate of the tritium breeding ratio if the cross sections are infinitely diluted. This is due to the resonances in the structure materials in which the absorption cross sections are enhanced in the infinitely diluted case. Thus the effect of self-shielding of resonance cross sections should be considered in neutronics calculations of fusion reactors. It is shown that the MCNP results are better reproduced by those from the transport code with the infinitely diluted library. This is probably due to the weight function used to generate the library and to the number of groups considered. Thus for fusion applications it is recommanded to collapse broad group cross sections with the spectrum obtained from an accurate calculation based on many fine groups. (author)
Aguirre, Eder; David, Mariano; deAlmeida, Carlos E
2016-01-01
This work studies the impact of systematic uncertainties associated to interaction cross sections on depth dose curves determined by Monte Carlo simulations. The corresponding sensitivity factors are quantified by changing cross sections in a given amount and determining the variation in the dose. The influence of total cross sections for all particles, photons and only for Compton scattering is addressed. The PENELOPE code was used in all simulations. It was found that photon cross section sensitivity factors depend on depth. In addition, they are positive and negative for depths below and above an equilibrium depth, respectively. At this depth, sensitivity factors are null. The equilibrium depths found in this work agree very well with the mean free path of the corresponding incident photon energy. Using the sensitivity factors reported here, it is possible to estimate the impact of photon cross section uncertainties on the uncertainty of Monte Carlo-determined depth dose curves.
The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self- indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various shelf-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 6 refs
Hamer, Harold A.; Mayer, John P.; Huston, Wilber B.
1961-01-01
Results of a statistical analysis of horizontal-tail loads on a fighter airplane are presented. The data were obtained from a number of operational training missions with flight at altitudes up to about 50,000 feet and at Mach numbers up to 1.22. The analysis was performed to determine the feasibility of calculating horizontal-tail load from data on the flight conditions and airplane motions. In the analysis the calculated loads are compared with the measured loads for the different types of missions performed. The loads were calculated by two methods: a direct approach and a Monte Carlo technique. The procedures used and some of the problems associated with the data analysis are discussed. frequencies of occurrence of tail loads of given magnitudes are derived from statistical information on the flight quantities. In the direct method, a time history of tail load is calculated from time-history measurements of the flight quantities. The Monte Carlo method could be useful for extending loads information for design of prospective airplanes . For the Monte Carlo method, the The results indicate that the accuracy of loads, regardless of the method used for calculation, is largely dependent on the knowledge of the pertinent airplane aerodynamic characteristics and center-of-gravity location. In addition, reliable Monte Carlo results require an adequate sample of statistical data and a knowledge of the more important statistical dependencies between the various flight conditions and airplane motions.
We present a new, nondestructive, method for determining chemical potentials in Monte Carlo and molecular dynamics simulations. The method estimates a value for the chemical potential such that one has a balance between fictitious successful creation and destruction trials in which the Monte Carlo method is used to determine success or failure of the creation/destruction attempts; we thus call the method a detailed balance method. The method allows one to obtain estimates of the chemical potential for a given species in any closed ensemble simulation; the closed ensemble is paired with a ''natural'' open ensemble for the purpose of obtaining creation and destruction probabilities. We present results for the Lennard-Jones system and also for an embedded atom model of liquid palladium, and compare to previous results in the literature for these two systems. We are able to obtain an accurate estimate of the chemical potential for the Lennard-Jones system at higher densities than reported in the literature
Patient-specific CT dose determination from CT images using Monte Carlo simulations
Liang, Qing
Radiation dose from computed tomography (CT) has become a public concern with the increasing application of CT as a diagnostic modality, which has generated a demand for patient-specific CT dose determinations. This thesis work aims to provide a clinically applicable Monte-Carlo-based CT dose calculation tool based on patient CT images. The source spectrum was simulated based on half-value layer measurements. Analytical calculations along with the measured flux distribution were used to estimate the bowtie-filter geometry. Relative source output at different points in a cylindrical phantom was measured and compared with Monte Carlo simulations to verify the determined spectrum and bowtie-filter geometry. Sensitivity tests were designed with four spectra with the same kVp and different half-value layers, and showed that the relative output at different locations in a phantom is sensitive to different beam qualities. An mAs-to-dose conversion factor was determined with in-air measurements using an Exradin A1SL ionization chamber. Longitudinal dose profiles were measured with thermoluminescent dosimeters (TLDs) and compared with the Monte-Carlo-simulated dose profiles to verify the mAs-to-dose conversion factor. Using only the CT images to perform Monte Carlo simulations would cause dose underestimation due to the lack of a scatter region. This scenario was demonstrated with a cylindrical phantom study. Four different image extrapolation methods from the existing CT images and the Scout images were proposed. The results show that performing image extrapolation beyond the scan region improves the dose calculation accuracy under both step-shoot scan mode and helical scan mode. Two clinical studies were designed and comparisons were performed between the current CT dose metrics and the Monte-Carlo-based organ dose determination techniques proposed in this work. The results showed that the current CT dosimetry failed to show dose differences between patients with the same
Application of Monte Carlo method in determination of secondary characteristic X radiation in XFA
Secondary characteristic radiation is excited by primary radiation from the X-ray tube and by secondary radiation of other elements so that excitations of several orders result. The Monte Carlo method was used to consider all these possibilities and the resulting flux of characteristic radiation was simulated for samples of silicate raw materials. A comparison of the results of these computations with experiments allows to determine the effect of sample preparation on the characteristic radiation flux. (M.D.)
Gonzalez, Jorge A. Carrazana; Ferrera, Eduardo A. Capote; Gomez, Isis M. Fernandez; Castro, Gloria V. Rodriguez; Ricardo, Niury Martinez, E-mail: cphr@cphr.edu.cu [Centro de Proteccion e Higiene de las Radiaciones (CPHR), La Habana (Cuba)
2013-07-01
This work shows how is established the traceability of the analytical determinations using this calibration method. Highlights the advantages offered by Monte Carlo simulation for the application of corrections by differences in chemical composition, density and height of the samples analyzed. Likewise, the results obtained by the LVRA in two exercises organized by the International Agency for Atomic Energy (IAEA) are presented. In these exercises (an intercomparison and a proficiency test) all reported analytical results were obtained based on calibrations in efficiency by Monte Carlo simulation using the DETEFF program.
It is well known that the temperature and background dependent neutron cross-sections are conventionally represented, in a problem-independent multigroup cross-section set, by specifying, for each group and reaction, the unshielded cross-section along with a set of self-shielding factors for various background cross-sections and temperatures. Usually the unshielded group cross-section is assumed to be independent of temperature. The observation presented in this paper, with examples, shows that the unshielded cross-section could significantly depend on temperature, depending on the group boundaries. (author)
A low energy, self-shielded electron beam machine for the surface coating curing of wooden panel and gypseous plate was introduced. The quality control during the electron curing process such as the coating technology in top-layer, the influence of coating-layer thickness on the property of coating surface, and the dose and dose distribution in the radiation field was also discussed. An unsaturated polyester resin, as a main component in our EB coating, was synthesized and compared with several other EB coatings. The properties of coating layer on the surface of wooden panel after curing were characterized
The neutron activation analysis errors caused by the effects of self-shielding of the neutron fluence and gamma quantum self-absorption are considered. These errors depend on the analyzed sample material, radiation parameters, radiation geometry and conditions of induced activity recording. The proper experiment is carried out at samples of different density (straw, grass, corn). All measurements were performed in the same geometry. The deltasub(n, γ) (L) effect of self-shielding and self-absorption in a sample of the L length was estimated by the following expression deltasub(n, γ)(L)=(Asub(0)sup(n, γ)(L)-Asub(sample)sup(n, γ)(L))/sub(0)sup(n, γ)(L) where Asub(0)sup(n, γ)(L) and Asub(sample)sup(n, γ)(L) are the induced gamma radiation fluences integrated along the L capsule length in the capsule with and without filler. The deltasub(n, γ)(L) effect was theoretically estimated. The dependence of this effect on the sample mass experimentally measured and theoretically calculated is given. Taking into account of the given effects with introduction of the proper correction during calculation permits to increase the accuracy of the neutron activation analysis by 5%
Monte Carlo simulation of determining porosity by using dual gamma detectors
Current formation elements spectroscopy logging technology utilize 241Am-Be neutron source and single BGO detector to determine elements contents. It plays an important role in mineral analysis and lithology identification of unconventional oil and gas exploration, but information measured is relatively ld. Measured system based on 241Am-Be neutron and dual detectors can be developed to realize the measurement of elements content as well as determine neutron gamma porosity by using ratio of gamma count between near and far detectors. Calculation model is built by Monte Carlo method to study neutron gamma porosity logging response with different spacing and shields. And it is concluded that measuring neutron gamma have high counts and good statistical property contrasted with measuring thermal neutron, but the sensitivity of porosity decrease. Sensitivity of porosity will increase as the spacing of dual detector increases. Spacing of far and near detectors should be around 62 cm and 35 cm respectively. Gamma counts decrease and neutron gamma porosity sensitivity increase when shield is fixed between neutron and detector. The length of main shield should be greater than 10 cm and associated shielding is about 5 cm. By Monte Carlo Simulation study, the result provides technical support for determining porosity in formation elements spectroscopy logging using 241Am-Be neutron and gamma detectors. (authors)
Monte Carlo simulation in quantitative determination of 137Cs in sand and water samples
To understand the distribution of radionuclides in the high background area, one mainly needs to analyse sand, soil, water and other food stuff samples by gamma-spectroscopy. Due to interaction of photons emitted by these radionuclides within the sample, the underestimation of quantity of radionuclides in the sample cannot be ruled out. To overcome this situation, the Monte Carlo method to determine the effect of multiple scattering in Compton profiles has been extended to take better account of interaction of radiation with environmental samples. In this paper, we present the feasibility of Monte Carlo simulation in determining the absorption and multiple scattering of gamma-rays from 137Cs radionuclides in the sand and water samples. It is seen that only 67 % and 90 % photons escaped from the sand and water respectively, can be detected by nuclear spectroscopy techniques. The high percentage of photoelectric absorption and Compton scattering of photons in these samples warrant the underestimation of quantitative determination of 137Cs in these samples. (author)
Determination of true coincidence correction factors using Monte-Carlo simulation techniques
Chionis Dionysios A.
2014-01-01
Full Text Available Aim of this work is the numerical calculation of the true coincidence correction factors by means of Monte-Carlo simulation techniques. For this purpose, the Monte Carlo computer code PENELOPE was used and the main program PENMAIN was properly modified in order to include the effect of the true coincidence phenomenon. The modified main program that takes into consideration the true coincidence phenomenon was used for the full energy peak efficiency determination of an XtRa Ge detector with relative efficiency 104% and the results obtained for the 1173 keV and 1332 keV photons of 60Co were found consistent with respective experimental ones. The true coincidence correction factors were calculated as the ratio of the full energy peak efficiencies was determined from the original main program PENMAIN and the modified main program PENMAIN. The developed technique was applied for 57Co, 88Y, and 134Cs and for two source-to-detector geometries. The results obtained were compared with true coincidence correction factors calculated from the "TrueCoinc" program and the relative bias was found to be less than 2%, 4%, and 8% for 57Co, 88Y, and 134Cs, respectively.
Mapping systematic errors in helium abundance determinations using Markov Chain Monte Carlo
Monte Carlo techniques have been used to evaluate the statistical and systematic uncertainties in the helium abundances derived from extragalactic H II regions. The helium abundance is sensitive to several physical parameters associated with the H II region. In this work, we introduce Markov Chain Monte Carlo (MCMC) methods to efficiently explore the parameter space and determine the helium abundance, the physical parameters, and the uncertainties derived from observations of metal poor nebulae. Experiments with synthetic data show that the MCMC method is superior to previous implementations (based on flux perturbation) in that it is not affected by biases due to non-physical parameter space. The MCMC analysis allows a detailed exploration of degeneracies, and, in particular, a false minimum that occurs at large values of optical depth in the He I emission lines. We demonstrate that introducing the electron temperature derived from the [O III] emission lines as a prior, in a very conservative manner, produces negligible bias and effectively eliminates the false minima occurring at large optical depth. We perform a frequentist analysis on data from several ''high quality'' systems. Likelihood plots illustrate degeneracies, asymmetries, and limits of the determination. In agreement with previous work, we find relatively large systematic errors, limiting the precision of the primordial helium abundance for currently available spectra
Use of Monte Carlo Methods for determination of isodose curves in brachytherapy
Brachytherapy is a special form of cancer treatment in which the radioactive source is very close to or inside the tumor with the objective of causing the necrosis of the cancerous tissue. The intensity of cell response to the radiation varies according to the tissue type and degree of differentiation. Since the malign cells are less differentiated than the normal ones, they are more sensitive to the radiation. This is the basis for radiotherapy techniques. Institutes that work with the application of high dose rates use sophisticated computer programs to calculate the necessary dose to achieve the necrosis of the tumor and the same time, minimizing the irradiation of tissues and organs of the neighborhood. With knowledge the characteristics of the source and the tumor, it is possible to trace isodose curves with the necessary information for planning the brachytherapy in patients. The objective of this work is, using Monte Carlo techniques, to develop a computer program - the ISODOSE - which allows to determine isodose curves in turn of linear radioactive sources used in brachytherapy. The development of ISODOSE is important because the available commercial programs, in general, are very expensive and practically inaccessible to small clinics. The use of Monte Carlo techniques is viable because they avoid problems inherent to analytic solutions as, for instance , the integration of functions with singularities in its domain. The results of ISODOSE were compared with similar data found in the literature and also with those obtained at the institutes of radiotherapy of the 'Hospital do Cancer do Recife' and of the 'Hospital Portugues do Recife'. ISODOSE presented good performance, mainly, due to the Monte Carlo techniques, that allowed a quite detailed drawing of the isodose curves in turn of linear sources. (author)
Simple formalism for efficient derivatives and multi-determinant expansions in quantum Monte Carlo
Filippi, Claudia; Assaraf, Roland; Moroni, Saverio
2016-05-01
We present a simple and general formalism to compute efficiently the derivatives of a multi-determinant Jastrow-Slater wave function, the local energy, the interatomic forces, and similar quantities needed in quantum Monte Carlo. Through a straightforward manipulation of matrices evaluated on the occupied and virtual orbitals, we obtain an efficiency equivalent to algorithmic differentiation in the computation of the interatomic forces and the optimization of the orbital parameters. Furthermore, for a large multi-determinant expansion, the significant computational gain afforded by a recently introduced table method is here extended to the local value of any one-body operator and to its derivatives, in both all-electron and pseudopotential calculations.
A new improved method has been developed for calculating sensitivity coefficients of neutronics parameters in pressurized water reactor cells relative to infinite dilution cross-sections by taking account of resonance self-shielding effect. In our paper, the IR approximation is used in order to get accurate results in both high and low energy groups. This method is applied to UO2 and MOX fueled PWR cells to calculate sensitivity coefficients and uncertainties of eigenvalue responses. We have verified the improved method by comparing the sensitivities with MCNP code and good agreement is found. For uncertainty, the improved results are compared with TSUNAMI-1D, and demonstrate that the differences are caused by the use of different covariance matrix. (author)
Choi, Chang Heon; Jung, Seongmoon; Choi, Kanghyuk; Son, Kwang-Jae; Lee, Jun Sig; Ye, Sung-Joon
2016-04-01
This study aims to determine the activity of a sealed pure beta-source by measuring the surface dose rate using an extrapolation chamber. A conversion factor (cGy s-1 Bq-1), which was defined as the ratio of surface dose rate to activity, can be calculated by Monte Carlo simulations of the extrapolation chamber measurement. To validate this hypothesis the certified activities of two standard pure beta-sources of Sr/Y-90 and Si/P-32 were compared with those determined by this method. In addition, a sealed test source of Sr/Y-90 was manufactured by the HANARO reactor group of KAERI (Korea Atomic Energy Research Institute) and used to further validate this method. The measured surface dose rates of the Sr/Y-90 and Si/P-32 standard sources were 4.615×10-5 cGy s-1 and 2.259×10-5 cGy s-1, respectively. The calculated conversion factors of the two sources were 1.213×10-8 cGy s-1 Bq-1 and 1.071×10-8 cGy s-1 Bq-1, respectively. Therefore, the activity of the standard Sr/Y-90 source was determined to be 3.995 kBq, which was 2.0% less than the certified value (4.077 kBq). For Si/P-32 the determined activity was 2.102 kBq, which was 6.6% larger than the certified activity (1.971 kBq). The activity of the Sr/Y-90 test source was determined to be 4.166 kBq, while the apparent activity reported by KAERI was 5.803 kBq. This large difference might be due to evaporation and diffusion of the source liquid during preparation and uncertainty in the amount of weighed aliquot of source liquid. The overall uncertainty involved in this method was determined to be 7.3%. We demonstrated that the activity of a sealed pure beta-source could be conveniently determined by complementary combination of measuring the surface dose rate and Monte Carlo simulations.
Monte Carlo determination of the lead equivalent for Syrian Building bricks for diagnostic x-ray
The uncertainty band associated with the transmission curve for 100 k Vp x-ray in lead was determined using Monte Carlo methods and the sensitivity analysis approach. All uncertainty sources (Statistical, systematical and the uncertainties arising from the diversity of x-ray tubes) were taken into account. The transmission of 100 k Vp x-ray in Syrian building bricks was then computed together with the uncertainty associated with it. Finally, the lead equivalent thickness for 10, 15 and 20 cm thick bricks were estimated. The results are in very good agreement with experimental results. This study recommends, as a rule of thumb, to use the lead-equivalent values of 0.5, 0.75 and 1.0 mm for the 10, 15 and 20 cm thick building bricks, respectively. (authors)
A 35 group cross-section set with P3-anisotropic scattering matrices and resonance self-shielding factors has been generated from the basic ENDF/B-IV cross-section Library for 57 reactor elements. This library, called BARC35, is considered to be well suited for the neutronics and safety analysis of fission, fusion and hybrid systems. (author)
Research on ADS, related new fuels and their ability for nuclear waste incineration leads to a revival of interest in nuclear cross-sections of many nuclides in a large energy range. Discrepancies observed between nuclear databases require new measurements in several cases. A complete measurement of such cross-sections including resonance resolution consists in an extensive beam time experiment associated to a long analysis. With a slowing down spectrometer associated to a pulsed neutron source, it is possible to determine a good cross-section profile in an energy range from 0.1 eV to 40 keV by making use of a slowing-down time lead spectrometer associated to a pulsed neutron source. These measurements performed at ISN (Grenoble) with the neutron source GENEPI requires only small quantities of matter (as small as 0.1 g) and about one day of beam by target. We present cross-section profile measurements and an experimental study of the self-shielding effect. A CeF3 scintillator coupled with a photomultiplier detects gamma rays from neutronic capture in the studied target. The neutron flux is also measured with a 233U fission detector and a 3He detector at symmetrical position to the PM in relation to the neutron source. Absolute flux values are given by activation of Au and W foils. The cross-section profiles can then be deduced from the target capture rate and are compared with very detailed MCNP simulations, which reproduce the experimental set-up and provide also capture rates and flux. The method is then applied to 232Th, of main interest for new fuel cycle studies, and is complementary to higher energy measurements made by D. Karamanis et al. (CENBG). Results obtained for three target thicknesses will be compared with simulations based on different data bases. Special attention will be paid to the region of unresolved resonances (>100 eV). (author)
Report on some methods of determining the state of convergence of Monte Carlo risk estimates
The Department of the Environment is developing a methodology for assessing potential sites for the disposal of low and intermediate level radioactive wastes. Computer models are used to simulate the groundwater transport of radioactive materials from a disposal facility back to man. Monte Carlo methods are being employed to conduct a probabilistic risk assessment (pra) of potential sites. The models calculate time histories of annual radiation dose to the critical group population. The annual radiation dose to the critical group in turn specifies the annual individual risk. The distribution of dose is generally highly skewed and many simulation runs are required to predict the level of confidence in the risk estimate i.e. to determine whether the risk estimate is converged. This report describes some statistical methods for determining the state of convergence of the risk estimate. The methods described include the Shapiro-Wilk test, calculation of skewness and kurtosis and normal probability plots. A method for forecasting the number of samples needed before the risk estimate is converged is presented. Three case studies were conducted to examine the performance of some of these techniques. (author)
MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX
1 - Description of problem or function: MINX calculates fine-group averaged infinitely diluted cross sections and self-shielding factors from ENDF/B-IV data. Its primary purpose is to generate a pseudo-composition-independent multigroup library which is input to the SPHINX space-energy collapse program (2) (PSR-0129) through standard CCCC-III (8) interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (5) (NESC0388) and ENDRUN (9) and the high-order group-to-group transfer matrices of SUPERTOG (10) (PSR-0013) and ETOG (11). Fine group energy boundaries, Legendre expansion order, gross spectral shape component (in the Bondarenko flux model), temperatures and dilutions can all be used specifically. 2 - Method of solution: Infinitely dilute, un-broadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the
Unfolding an under-determined neutron spectrum using genetic algorithm based Monte Carlo
Spallation in addition to the other photon-neutron reactions in target materials and different components in accelerators may result in production of huge amount of energetic protons which further leads to the production of neutron and contributes to the main component of the total dose. For dosimetric purposes in accelerator facilities the detector measurements doesn't provide directly the actual neutron flux values but a cumulative picture. To obtain Neutron spectrum from the measured data, response functions of the measuring instrument together with the measurements are used into many unfolding techniques which are frequently used for unfolding the hidden spectral information. Here we discuss a genetic algorithm based unfolding technique which is in the process of development. Genetic Algorithm is a stochastic method based on natural selection, which mimics Darwinian theory of survival of the best. The above said method has been tested to unfold the neutron spectra obtained from a reaction carried out at an accelerator facility, with energetic carbon ions on thick silver target along with its respective neutron response of BC501A liquid scintillation detector. The problem dealt here is under-determined where the number of measurements is less than the required energy bin information. The results so obtained were compared with those obtained using the established unfolding code FERDOR, which unfolds data for completely determined problems. It is seen that the genetic algorithm based solution has a reasonable match with the results of FERDOR, when smoothening carried out by Monte Carlo is taken into consideration. This method appears to be a promising candidate for unfolding neutron spectrum in cases of under-determined and over-determined, where measurements are more. The method also has advantages of flexibility, computational simplicity and works well without need of any initial guess spectrum. (author)
Dinpajooh, Mohammadhasan; Bai, Peng; Allan, Douglas A; Siepmann, J Ilja
2015-09-21
Since the seminal paper by Panagiotopoulos [Mol. Phys. 61, 813 (1997)], the Gibbs ensemble Monte Carlo (GEMC) method has been the most popular particle-based simulation approach for the computation of vapor-liquid phase equilibria. However, the validity of GEMC simulations in the near-critical region has been questioned because rigorous finite-size scaling approaches cannot be applied to simulations with fluctuating volume. Valleau [Mol. Simul. 29, 627 (2003)] has argued that GEMC simulations would lead to a spurious overestimation of the critical temperature. More recently, Patel et al. [J. Chem. Phys. 134, 024101 (2011)] opined that the use of analytical tail corrections would be problematic in the near-critical region. To address these issues, we perform extensive GEMC simulations for Lennard-Jones particles in the near-critical region varying the system size, the overall system density, and the cutoff distance. For a system with N = 5500 particles, potential truncation at 8σ and analytical tail corrections, an extrapolation of GEMC simulation data at temperatures in the range from 1.27 to 1.305 yields T(c) = 1.3128 ± 0.0016, ρ(c) = 0.316 ± 0.004, and p(c) = 0.1274 ± 0.0013 in excellent agreement with the thermodynamic limit determined by Potoff and Panagiotopoulos [J. Chem. Phys. 109, 10914 (1998)] using grand canonical Monte Carlo simulations and finite-size scaling. Critical properties estimated using GEMC simulations with different overall system densities (0.296 ≤ ρ(t) ≤ 0.336) agree to within the statistical uncertainties. For simulations with tail corrections, data obtained using r(cut) = 3.5σ yield T(c) and p(c) that are higher by 0.2% and 1.4% than simulations with r(cut) = 5 and 8σ but still with overlapping 95% confidence intervals. In contrast, GEMC simulations with a truncated and shifted potential show that r(cut) = 8σ is insufficient to obtain accurate results. Additional GEMC simulations for hard-core square-well particles with
Dinpajooh, Mohammadhasan [Department of Chemistry and Chemical Theory Center, University of Minnesota, 207 Pleasant Street SE, Minneapolis, Minnesota 55455 (United States); Bai, Peng; Allan, Douglas A. [Department of Chemical Engineering and Materials Science, University of Minnesota, 421 Washington Avenue SE, Minneapolis, Minnesota 55455 (United States); Siepmann, J. Ilja, E-mail: siepmann@umn.edu [Department of Chemistry and Chemical Theory Center, University of Minnesota, 207 Pleasant Street SE, Minneapolis, Minnesota 55455 (United States); Department of Chemical Engineering and Materials Science, University of Minnesota, 421 Washington Avenue SE, Minneapolis, Minnesota 55455 (United States)
2015-09-21
Since the seminal paper by Panagiotopoulos [Mol. Phys. 61, 813 (1997)], the Gibbs ensemble Monte Carlo (GEMC) method has been the most popular particle-based simulation approach for the computation of vapor–liquid phase equilibria. However, the validity of GEMC simulations in the near-critical region has been questioned because rigorous finite-size scaling approaches cannot be applied to simulations with fluctuating volume. Valleau [Mol. Simul. 29, 627 (2003)] has argued that GEMC simulations would lead to a spurious overestimation of the critical temperature. More recently, Patel et al. [J. Chem. Phys. 134, 024101 (2011)] opined that the use of analytical tail corrections would be problematic in the near-critical region. To address these issues, we perform extensive GEMC simulations for Lennard-Jones particles in the near-critical region varying the system size, the overall system density, and the cutoff distance. For a system with N = 5500 particles, potential truncation at 8σ and analytical tail corrections, an extrapolation of GEMC simulation data at temperatures in the range from 1.27 to 1.305 yields T{sub c} = 1.3128 ± 0.0016, ρ{sub c} = 0.316 ± 0.004, and p{sub c} = 0.1274 ± 0.0013 in excellent agreement with the thermodynamic limit determined by Potoff and Panagiotopoulos [J. Chem. Phys. 109, 10914 (1998)] using grand canonical Monte Carlo simulations and finite-size scaling. Critical properties estimated using GEMC simulations with different overall system densities (0.296 ≤ ρ{sub t} ≤ 0.336) agree to within the statistical uncertainties. For simulations with tail corrections, data obtained using r{sub cut} = 3.5σ yield T{sub c} and p{sub c} that are higher by 0.2% and 1.4% than simulations with r{sub cut} = 5 and 8σ but still with overlapping 95% confidence intervals. In contrast, GEMC simulations with a truncated and shifted potential show that r{sub cut} = 8σ is insufficient to obtain accurate results. Additional GEMC simulations for hard
Since the seminal paper by Panagiotopoulos [Mol. Phys. 61, 813 (1997)], the Gibbs ensemble Monte Carlo (GEMC) method has been the most popular particle-based simulation approach for the computation of vapor–liquid phase equilibria. However, the validity of GEMC simulations in the near-critical region has been questioned because rigorous finite-size scaling approaches cannot be applied to simulations with fluctuating volume. Valleau [Mol. Simul. 29, 627 (2003)] has argued that GEMC simulations would lead to a spurious overestimation of the critical temperature. More recently, Patel et al. [J. Chem. Phys. 134, 024101 (2011)] opined that the use of analytical tail corrections would be problematic in the near-critical region. To address these issues, we perform extensive GEMC simulations for Lennard-Jones particles in the near-critical region varying the system size, the overall system density, and the cutoff distance. For a system with N = 5500 particles, potential truncation at 8σ and analytical tail corrections, an extrapolation of GEMC simulation data at temperatures in the range from 1.27 to 1.305 yields Tc = 1.3128 ± 0.0016, ρc = 0.316 ± 0.004, and pc = 0.1274 ± 0.0013 in excellent agreement with the thermodynamic limit determined by Potoff and Panagiotopoulos [J. Chem. Phys. 109, 10914 (1998)] using grand canonical Monte Carlo simulations and finite-size scaling. Critical properties estimated using GEMC simulations with different overall system densities (0.296 ≤ ρt ≤ 0.336) agree to within the statistical uncertainties. For simulations with tail corrections, data obtained using rcut = 3.5σ yield Tc and pc that are higher by 0.2% and 1.4% than simulations with rcut = 5 and 8σ but still with overlapping 95% confidence intervals. In contrast, GEMC simulations with a truncated and shifted potential show that rcut = 8σ is insufficient to obtain accurate results. Additional GEMC simulations for hard-core square-well particles with various ranges of the
Vieira, Jose Wilson
2001-08-01
Brachytherapy is a special form of cancer treatment in which the radioactive source is very close to or inside the tumor with the objective of causing the necrosis of the cancerous tissue. The intensity of cell response to the radiation varies according to the tissue type and degree of differentiation. Since the malign cells are less differentiated than the normal ones, they are more sensitive to the radiation. This is the basis for radiotherapy techniques. Institutes that work with the application of high dose rates use sophisticated computer programs to calculate the necessary dose to achieve the necrosis of the tumor and the same time, minimizing the irradiation of tissues and organs of the neighborhood. With knowledge the characteristics of the source and the tumor, it is possible to trace isodose curves with the necessary information for planning the brachytherapy in patients. The objective of this work is, using Monte Carlo techniques, to develop a computer program - the ISODOSE - which allows to determine isodose curves in turn of linear radioactive sources used in brachytherapy. The development of ISODOSE is important because the available commercial programs, in general, are very expensive and practically inaccessible to small clinics. The use of Monte Carlo techniques is viable because they avoid problems inherent to analytic solutions as, for instance , the integration of functions with singularities in its domain. The results of ISODOSE were compared with similar data found in the literature and also with those obtained at the institutes of radiotherapy of the 'Hospital do Cancer do Recife' and of the 'Hospital Portugues do Recife'. ISODOSE presented good performance, mainly, due to the Monte Carlo techniques, that allowed a quite detailed drawing of the isodose curves in turn of linear sources. (author)
Chang, Kwo-Ping; Wang, Zhi-Wei; Shiau, An-Cheng
2014-02-01
Monte Carlo (MC) method is a well known calculation algorithm which can accurately assess the dose distribution for radiotherapy. The present study investigated all the possible regions of the depth-dose or lateral profiles which may affect the fitting of the initial parameters (mean energy and the radial intensity (full width at half maximum, FWHM) of the incident electron). EGSnrc-based BEAMnrc codes were used to generate the phase space files (SSD=100 cm, FS=40×40 cm2) for the linac (linear accelerator, Varian 21EX, 6 MV photon mode) and EGSnrc-based DOSXYZnrc code was used to calculate the dose in the region of interest. Interpolation of depth dose curves of pre-set energies was proposed as a preliminary step for optimal energy fit. A good approach for determination of the optimal mean energy is the difference comparison of the PDD curves excluding buildup region, and using D(10) as a normalization method. For FWHM fitting, due to electron disequilibrium and the larger statistical uncertainty, using horn or/and penumbra regions will give inconsistent outcomes at various depths. Difference comparisons should be performed in the flat regions of the off-axis dose profiles at various depths to optimize the FWHM parameter.
Monte Carlo simulation methods of determining red bone marrow dose from external radiation
Objective: To provide evidence for a more reasonable method of determining red bone marrow dose by analyzing and comparing existing simulation methods. Methods: By utilizing Monte Carlo simulation software MCNPX, the absorbed doses of red hone marrow of Rensselaer Polytechnic Institute (RPI) adult female voxel phantom were calculated through 4 different methods: direct energy deposition.dose response function (DRF), King-Spiers factor method and mass-energy absorption coefficient (MEAC). The radiation sources were defined as infinite plate.sources with the energy ranging from 20 keV to 10 MeV, and 23 sources with different energies were simulated in total. The source was placed right next to the front of the RPI model to achieve a homogeneous anteroposterior radiation scenario. The results of different simulated photon energy sources through different methods were compared. Results: When the photon energy was lower than 100 key, the direct energy deposition method gave the highest result while the MEAC and King-Spiers factor methods showed more reasonable results. When the photon energy was higher than 150 keV taking into account of the higher absorption ability of red bone marrow at higher photon energy, the result of the King-Spiers factor method was larger than those of other methods. Conclusions: The King-Spiers factor method might be the most reasonable method to estimate the red bone marrow dose from external radiation. (authors)
Zhang, Tianli [School of Materials Science and Engineering, Tianjin University, Tianjin 300072 (China); College of Materials Science and Engineering, Beijing University of Technology, Beijing 100124 (China); Department of Materials Science and Engineering, University of Wisconsin, Madison, WI 53706 (United States); Li, Zhuoxin [College of Materials Science and Engineering, Beijing University of Technology, Beijing 100124 (China); Kou, Sindo, E-mail: kou@engr.wisc.edu [Department of Materials Science and Engineering, University of Wisconsin, Madison, WI 53706 (United States); Jing, Hongyang [School of Materials Science and Engineering, Tianjin University, Tianjin 300072 (China); Li, Guodong; Li, Hong [College of Materials Science and Engineering, Beijing University of Technology, Beijing 100124 (China); Jin Kim, Hee [Advanced Joining Research Team, Korea Institute of Industrial Technology, Chanan-si 330-825 (Korea, Republic of)
2015-03-25
The effect of inclusions on the microstructure and toughness of the deposited metals of self-shielded flux cored wires was investigated by optical microscopy, electron microscopy and mechanical testing. The deposited metals of three different wires showed different levels of low temperature impact toughness at −40 °C mainly because of differences in the properties of inclusions. The inclusions formed in the deposited metals as a result of deoxidation caused by the addition of extra Al–Mg alloy and ferromanganese to the flux. The inclusions, spherical in shape, were mixtures of Al{sub 2}O{sub 3} and MgO. Inclusions predominantly Al{sub 2}O{sub 3} and 0.3–0.8 μm in diameter were effective for nucleation of acicular ferrite. However, inclusions predominantly MgO were promoted by increasing Mg in the flux and were more effective than Al{sub 2}O{sub 3} inclusions of the same size. These findings suggest that the control of inclusions can be an effective way to improve the impact toughness of the deposited metal.
Initial experience with an 11 MeV self-shielded medical cyclotron on operation and radiation safety
Pant G
2007-01-01
Full Text Available A self-shielded medical cyclotron (11 MeV was commissioned at our center, to produce positron emitters, namely, 18 F, 15 O, 13 N and 11 C for positron emission tomography (PET imaging. Presently the cyclotron has been exclusively used for the production of 18 F - for 18 F-FDG imaging. The operational parameters which influence the yield of 18 F - production were monitored. The radiation levels in the cyclotron and radiochemistry laboratory were also monitored to assess the radiation safety status in the facility. The target material, 18 O water, is bombarded with proton beam from the cyclotron to produce 18 F - ion that is used for the synthesis of 18 F-FDG. The operational parameters which influence the yield of 18 F - were observed during 292 production runs out of a total of more than 400 runs. The radiation dose levels were also measured in the facility at various locations during cyclotron production runs and in the radiochemistry laboratory during 18 F-FDG syntheses. It was observed that rinsing the target after delivery increased the number of production runs in a given target, as well as resulted in a better correlation between the duration of bombardment and the end of bombardment 18 F - activity with absolutely clean target after being rebuilt. The radiation levels in the cyclotron and radiochemistry laboratory were observed to be well within prescribed limits with safe work practice.
The effect of inclusions on the microstructure and toughness of the deposited metals of self-shielded flux cored wires was investigated by optical microscopy, electron microscopy and mechanical testing. The deposited metals of three different wires showed different levels of low temperature impact toughness at −40 °C mainly because of differences in the properties of inclusions. The inclusions formed in the deposited metals as a result of deoxidation caused by the addition of extra Al–Mg alloy and ferromanganese to the flux. The inclusions, spherical in shape, were mixtures of Al2O3 and MgO. Inclusions predominantly Al2O3 and 0.3–0.8 μm in diameter were effective for nucleation of acicular ferrite. However, inclusions predominantly MgO were promoted by increasing Mg in the flux and were more effective than Al2O3 inclusions of the same size. These findings suggest that the control of inclusions can be an effective way to improve the impact toughness of the deposited metal
Uncertainty of Doppler reactivity is theoretically formulated and then uncertainties of self-shielding factor f(T) and its temperature gradient α due to errors of resonance parameters were evaluated from NJOY output. In the course of evaluation, serious investigation on numerical differentiation method was made for so-called two or three-point model, significance of figure of NJOY output and acceptable digits for computers. The f(T) and α uncertainties based on JENDL-3.3 covariance file were evaluated by using computer code system ERRORF newly developed for present work as a modular code system consisting of NJOY for effective cross section and f-factor calculations, REPCHANGE for parameter change and FTOALPHA for covariance calculation. Sensitivity analysis was made for 235U, 238U, 239Pu and 240Pu based on JAERI Fast set-3 70 group structure. Importance of off-diagonal terms was emphasized in the uncertainty evaluation. Resultant sensitivity coefficients are provided for the uncertainty evaluation of Doppler reactivity. (author)
A 39 neutron group self-shielded cross section library for the Lotus fusion-fission test facility
A 39 neutron group cross section library for fusion fission blanket calculations and especially for the analysis of the LOTUS experiment has been processed using the NJOY system. The library has been generated mostly using the ENDF/B-IV basic files at 296 K. All cross sections were self-shielded using the Bondarenko method. 5 background cross sections, namely 1010, 104, 102, 10 and 1 barns respectively were considered. The tabulated dilution dependent cross sections have been interpolated with the code TRANSX-CTR which is adequate for fusion applications. The fission spectrum of the fissionable material thorium has been collapsed from the fission matrices using the Bondarenko weighting scheme. The correct geometry of the LOTUS blanket and the cell specifications were correctly considered in the interpolation scheme. Some reaction cross sections for dosimetry applications have been included into the library. These base on the more recent ENDF/B-V evaluation. Transport and response edit cross sections have been coupled in the usual way to form P0 - P3 card image tables. Furthermore they have been converted into a binary file suitable to our RSYST computational system. Moreover the cross section card image tables have been reformatted and fitted into a BXSLIB binary library for the LANL-ONEDANT transport module. (Auth.)
Omar, Artur; Benmakhlouf, Hamza; Marteinsdottir, Maria; Bujila, Robert; Nowik, Patrik; Andreo, Pedro
2014-03-01
Complex interventional and diagnostic x-ray angiographic (XA) procedures may yield patient skin doses exceeding the threshold for radiation induced skin injuries. Skin dose is conventionally determined by converting the incident air kerma free-in-air into entrance surface air kerma, a process that requires the use of backscatter factors. Subsequently, the entrance surface air kerma is converted into skin kerma using mass energy-absorption coefficient ratios tissue-to-air, which for the photon energies used in XA is identical to the skin dose. The purpose of this work was to investigate how the cranial bone affects backscatter factors for the dosimetry of interventional neuroradiology procedures. The PENELOPE Monte Carlo system was used to calculate backscatter factors at the entrance surface of a spherical and a cubic water phantom that includes a cranial bone layer. The simulations were performed for different clinical x-ray spectra, field sizes, and thicknesses of the bone layer. The results show a reduction of up to 15% when a cranial bone layer is included in the simulations, compared with conventional backscatter factors calculated for a homogeneous water phantom. The reduction increases for thicker bone layers, softer incident beam qualities, and larger field sizes, indicating that, due to the increased photoelectric crosssection of cranial bone compared to water, the bone layer acts primarily as an absorber of low-energy photons. For neurointerventional radiology procedures, backscatter factors calculated at the entrance surface of a water phantom containing a cranial bone layer increase the accuracy of the skin dose determination.
Background. Dosimetry in radionuclide therapy estimates delivered absorbed doses to tumours and ensures that absorbed dose levels to normal organs are below tolerance levels. One procedure is to determine time-activity curves in volumes-of-interests from which the absorbed dose is estimated using SPECT with appropriate corrections for attenuation, scatter and collimator response. From corrected SPECT images the absorbed energy can be calculated by (a) assuming kinetic energy deposited in the same voxel where particles were emitted, (b) convolve with point-dose kernels or (c) use full Monte Carlo (MC) methods. A question arises which dosimetry method is optimal given the limitations in reconstruction- and quantification procedures. Methods. Dosimetry methods (a) and (c) were evaluated by comparing dose-rate volume histograms (DrVHs) from simulated SPECT of 111In, 177Lu, 131I and Bremsstrahlung from 90Y to match true dose rate images. The study used a voxel-based phantom with different tumours in the liver. SPECT reconstruction was made using an iterative OSEM method and MC dosimetry was performed using a charged-particle EGS4 program that also was used to determined true absorbed dose rate distributions for the same phantom geometry but without camera limitations. Results. The DrVHs obtained from SPECT differed from true DrVH mainly due to limited spatial resolution. MC dosimetry had a marginal effect because the SPECT spatial resolution is in the same order as the energy distribution caused by the electron track ranges. For 131I, full MC dosimetry made a difference due to the additional contribution from high-energy photons. SPECT-based DrVHs differ significantly from true DrVHs unless the tumours are considerable larger than the spatial resolution. Conclusion. It is important to understand limitations in quantitative SPECT images and the reasons for apparent heterogeneities since these have an impact on dose-volume histograms. A MC-based dosimetry calculation from
Coste-Delclaux, M
2006-03-15
This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)
Coefficients of an analytical aerosol forcing equation determined with a Monte-Carlo radiation model
Hassan, Taufiq; Moosmüller, H.; Chung, Chul E.
2015-10-01
Simple analytical equations for global-average direct aerosol radiative forcing are useful to quickly estimate aerosol forcing changes as function of key atmosphere, surface and aerosol parameters. The surface and atmosphere parameters in these analytical equations are the globally uniform atmospheric transmittance and surface albedo, and have so far been estimated from simplified observations under untested assumptions. In the present study, we take the state-of-the-art analytical equation and write the aerosol forcing as a linear function of the single scattering albedo (SSA) and replace the average upscatter fraction with the asymmetry parameter (ASY). Then we determine the surface and atmosphere parameter values of this equation using the output from the global MACR (Monte-Carlo Aerosol Cloud Radiation) model, as well as testing the validity of the equation. The MACR model incorporated spatio-temporally varying observations for surface albedo, cloud optical depth, water vapor, stratosphere column ozone, etc., instead of assuming as in the analytical equation that the atmosphere and surface parameters are globally uniform, and should thus be viewed as providing realistic radiation simulations. The modified analytical equation needs globally uniform aerosol parameters that consist of AOD (Aerosol Optical Depth), SSA, and ASY. The MACR model is run here with the same globally uniform aerosol parameters. The MACR model is also run without cloud to test the cloud effect. In both cloudy and cloud-free runs, the equation fits in the model output well whether SSA or ASY varies. This means the equation is an excellent approximation for the atmospheric radiation. On the other hand, the determined parameter values are somewhat realistic for the cloud-free runs but unrealistic for the cloudy runs. The global atmospheric transmittance, one of the determined parameters, is found to be around 0.74 in case of the cloud-free conditions and around 1.03 with cloud. The surface
Coefficients of an analytical aerosol forcing equation determined with a Monte-Carlo radiation model
Simple analytical equations for global-average direct aerosol radiative forcing are useful to quickly estimate aerosol forcing changes as function of key atmosphere, surface and aerosol parameters. The surface and atmosphere parameters in these analytical equations are the globally uniform atmospheric transmittance and surface albedo, and have so far been estimated from simplified observations under untested assumptions. In the present study, we take the state-of-the-art analytical equation and write the aerosol forcing as a linear function of the single scattering albedo (SSA) and replace the average upscatter fraction with the asymmetry parameter (ASY). Then we determine the surface and atmosphere parameter values of this equation using the output from the global MACR (Monte-Carlo Aerosol Cloud Radiation) model, as well as testing the validity of the equation. The MACR model incorporated spatio-temporally varying observations for surface albedo, cloud optical depth, water vapor, stratosphere column ozone, etc., instead of assuming as in the analytical equation that the atmosphere and surface parameters are globally uniform, and should thus be viewed as providing realistic radiation simulations. The modified analytical equation needs globally uniform aerosol parameters that consist of AOD (Aerosol Optical Depth), SSA, and ASY. The MACR model is run here with the same globally uniform aerosol parameters. The MACR model is also run without cloud to test the cloud effect. In both cloudy and cloud-free runs, the equation fits in the model output well whether SSA or ASY varies. This means the equation is an excellent approximation for the atmospheric radiation. On the other hand, the determined parameter values are somewhat realistic for the cloud-free runs but unrealistic for the cloudy runs. The global atmospheric transmittance, one of the determined parameters, is found to be around 0.74 in case of the cloud-free conditions and around 1.03 with cloud. The surface
Palmer, Grant; Prabhu, Dinesh; Cruden, Brett A.
2013-01-01
The 2013-2022 Decaedal survey for planetary exploration has identified probe missions to Uranus and Saturn as high priorities. This work endeavors to examine the uncertainty for determining aeroheating in such entry environments. Representative entry trajectories are constructed using the TRAJ software. Flowfields at selected points on the trajectories are then computed using the Data Parallel Line Relaxation (DPLR) Computational Fluid Dynamics Code. A Monte Carlo study is performed on the DPLR input parameters to determine the uncertainty in the predicted aeroheating, and correlation coefficients are examined to identify which input parameters show the most influence on the uncertainty. A review of the present best practices for input parameters (e.g. transport coefficient and vibrational relaxation time) is also conducted. It is found that the 2(sigma) - uncertainty for heating on Uranus entry is no more than 2.1%, assuming an equilibrium catalytic wall, with the uncertainty being determined primarily by diffusion and H(sub 2) recombination rate within the boundary layer. However, if the wall is assumed to be partially or non-catalytic, this uncertainty may increase to as large as 18%. The catalytic wall model can contribute over 3x change in heat flux and a 20% variation in film coefficient. Therefore, coupled material response/fluid dynamic models are recommended for this problem. It was also found that much of this variability is artificially suppressed when a constant Schmidt number approach is implemented. Because the boundary layer is reacting, it is necessary to employ self-consistent effective binary diffusion to obtain a correct thermal transport solution. For Saturn entries, the 2(sigma) - uncertainty for convective heating was less than 3.7%. The major uncertainty driver was dependent on shock temperature/velocity, changing from boundary layer thermal conductivity to diffusivity and then to shock layer ionization rate as velocity increases. While
Murata, Isao; Shindo, Ryuichi; Shiozawa, Shusaku [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment
1995-10-01
In case of a shielding analysis of the geometry having thick and complicated structures with a Monte Carlo code, it is a serious problem that it takes too much computer time to obtain results with good statistics. Therefore, it is very important to reduce variances in the calculation. In this study, a method to determine the importance function in 3-dimensional Monte Carlo calculation with geometry splitting with Russian roulette was developed for the shielding analysis of thick and complicated core shielding structures. Only two essential importance ratio curves for one material enable us to determine the importance function easily in the shielding calculation. The validity of this method was confirmed through a simple benchmark calculation. From the comparison with the result obtained by using weight window (W-W), it was shown that the present method can give an accurate result on the same level with W-W method with less trial and errors. And this method was applied to an actual reactor core shielding analysis to confirm its applicability to a 3-dimensional thick and complicated structure. Using this method, the variance reduced calculation can be easily realized with the developed importance determination procedure, especially in case that parameter survey calculations are required in order to determine the shield thickness in a design work of a thick and complicated structure. Accordingly, it became easier to use Monte Carlo method as a powerful tool for a reactor core shielding design. (author).
In case of a shielding analysis of the geometry having thick and complicated structures with a Monte Carlo code, it is a serious problem that it takes too much computer time to obtain results with good statistics. Therefore, it is very important to reduce variances in the calculation. In this study, a method to determine the importance function in 3-dimensional Monte Carlo calculation with geometry splitting with Russian roulette was developed for the shielding analysis of thick and complicated core shielding structures. Only two essential importance ratio curves for one material enable us to determine the importance function easily in the shielding calculation. The validity of this method was confirmed through a simple benchmark calculation. From the comparison with the result obtained by using weight window (W-W), it was shown that the present method can give an accurate result on the same level with W-W method with less trial and errors. And this method was applied to an actual reactor core shielding analysis to confirm its applicability to a 3-dimensional thick and complicated structure. Using this method, the variance reduced calculation can be easily realized with the developed importance determination procedure, especially in case that parameter survey calculations are required in order to determine the shield thickness in a design work of a thick and complicated structure. Accordingly, it became easier to use Monte Carlo method as a powerful tool for a reactor core shielding design. (author)
EURADOS action for determination of americium in skull measures in vivo and Monte Carlo simulation
From the Group of WG7 internal dosimetry of the EURADOS Organization (European Radiation Dosimetry group, e.V.) which It coordinates CIEMAT, international action for the vivo measurement of americium has been conducted in three mannequins type skull with detectors of Germanium by gamma spectrometry and simulation by Monte Carlo methods. Such action has been raised as two separate exercises, with the participation of institutions in Europe, America and Asia. Other actions similar precede this vivo intercomparison of measurement and modeling Monte Carlo1. The preliminary results and associated findings are presented in this work. The laboratory of the body radioactivity (CRC) of service counter of dosimetry staff internal (DPI) of the CIEMAT, it has been one of the participants in vivo measures exercise. On the other hand part, the Group of numerical dosimetry of CIEMAT is participant of the Monte Carlo2 simulation exercise. (Author)
The four primary radiation measurement systems considered to be necessary for the modern radionuclide measurement laboratory are gas-flow proportional counters, liquid scintillation counters, Si spectrometer systems, and Ge spectrometer systems. A national institute of standards recognized by BIPM (Bureau International des Poids et Mesures) use these measurement systems have some limitation. For a gas-filled detector, the source size should not be larger than that determined by the counter geometry and the measurements are limited to sources with the activities of less than 20,000 Bq. A liquid scintillation detector can only be used to measure an activity of liquid mixture isotope. Therefore it can't be used to measure activities of sealed sources. Further it is quite difficult to measure a radioactivity of beta-isotope accurately due to self-absorption and scattering. In particular an radioactivity generated by a nuclear reactor is approximately calculated by using target material compositions, cross-sections, and neutron flux, which the nuclear reactor. However, the results from this approach involve a high uncertainty. We develop a Monte Carlo applied radioactivity determination for beta sources. The extrapolation ion-chamber was used to measure the surface dose rate for a standard source. The Sr/Y-90 standard source is calibrated by NIST used for this study. There was about 1.7% difference in the reference dose rates measured by the two techniques. These difference and relative errors were comparable to those of other studies. It should be noticed that the area of collecting electrode was nine times larger than that of radiation field. It the diameter of the source is smaller than the area of collecting electrode, the source should be located at a sufficiently large distance from the detector surface so that the radiation field at the detector surface was larger than the area of collecting electrode. In addition, beta particles from the applicator interact
International standards and guidelines for calibrating high-dose dosimetry systems to be used in industrial radiation processing recommend that dose-rate effects on dosimeters be evaluated under conditions of use. This is important when the irradiation relies on high-current electron accelerators, which usually provide very high dose-rates. However, most dosimeter calibration facilities use low-intensity gamma radiation or low-current electron accelerators, which deliver comparatively low dose-rates. Because of issues of thermal conductivity and response, portable calorimeters cannot be practically used with high-current accelerators, where product conveyor speeds under an electron beam can exceed several meters per second and the calorimeter is not suitable for use with product handling systems. As an alternative, Monte Carlo calculations can give theoretical estimates of the absorbed dose in materials with flat or complex configurations such that the results are independent of dose-rate. Monte Carlo results can then be compared to experimental dose determinations to see whether dose-rate effects in the dosimeters are significant. A Monte Carlo code has been used in this study to calculate the absorbed doses in alanine film dosimeters supported by flat sheets of plywood irradiated with electrons using incident energies extending from 1.0 MeV to 10 MeV with beam currents up to 30 mA. The same process conditions have been used for dose determinations with high-current electron beams using low dose-rate gamma calibrated alanine film dosimeters. The close agreement between these calculations and the dosimeter determinations indicates that the response of this type of dosimeter system is independent of the dose-rate, and provides assurance that Monte Carlo calculations can yield results with sufficient accuracy for many industrial applications
Isotope production and Application Division of Bhabha Atomic Research Center developed 32P patch sources for treatment of superficial tumors. Surface dose rate of a newly developed 32P patch source of nominal diameter 25 mm was measured experimentally using standard extrapolation ionization chamber and Gafchromic EBT film. Monte Carlo model of the 32P patch source along with the extrapolation chamber was also developed to estimate the surface dose rates from these sources. The surface dose rates to tissue (cGy/min) measured using extrapolation chamber and radiochromic films are 82.03±4.18 (k=2) and 79.13±2.53 (k=2) respectively. The two values of the surface dose rates measured using the two independent experimental methods are in good agreement to each other within a variation of 3.5%. The surface dose rate to tissue (cGy/min) estimated using the MCNP Monte Carlo code works out to be 77.78±1.16 (k=2). The maximum deviation between the surface dose rates to tissue obtained by Monte Carlo and the extrapolation chamber method is 5.2% whereas the difference between the surface dose rates obtained by radiochromic film measurement and the Monte Carlo simulation is 1.7%. The three values of the surface dose rates of the 32P patch source obtained by three independent methods are in good agreement to one another within the uncertainties associated with their measurements and calculation. This work has demonstrated that MCNP based electron transport simulations are accurate enough for determining the dosimetry parameters of the indigenously developed 32P patch sources for contact brachytherapy applications. - Highlights: • Surface dose rates of 25 mm nominal diameter newly developed 32P patch sources were measured experimentally using extrapolation chamber and Gafchromic EBT2 film. Monte Carlo model of the 32P patch source along with the extrapolation chamber was also developed. • The surface dose rates to tissue (cGy/min) measured using extrapolation chamber and
A new method for the experimental determination of stopping powers based on Bayesian Inference with the Markov chain Monte Carlo (MCMC) algorithm has been devised. This method avoids the difficulties related to thin target preparation. By measuring the RBS spectra for a known material, and using the known underlying physics, the stopping powers are determined by best matching the simulated spectra with the experimental spectra. Using silicon, SiO2 and Al2O3 as test cases, good agreement is obtained between calculated and experimental data. (author)
Different codes were used for Monte Carlo calculations in radiation therapy. In this study, a new Monte Carlo Simulation Program (MCSP) was developed for the effects of the physical parameters of photons emitted from a Siemens Primus clinical linear accelerator (LINAC) on the dose distribution in water. For MCSP, it was written considering interactions of photons with matter. Here, it was taken into account mainly two interactions: The Compton (or incoherent) scattering and photoelectric effect. Photons which come to water phantom surface emitting from a point source were Bremsstrahlung photons. It should be known the energy distributions of these photons for following photons. Bremsstrahlung photons which have 6 MeV (6 MV photon mode) maximum energies were taken into account. In the 6 MV photon mode, the energies of photons were sampled from using Mohan's experimental energy spectrum (Mohan at al 1985). In order to investigate the performance and accuracy of the simulation, measured and calculated (MCSP) percentage depth dose curves and dose profiles were compared. The Monte Carlo results were shown good agreement with experimental measurements.
TRIPOLI-3: a neutron/photon Monte Carlo transport code
The present version of TRIPOLI-3 solves the transport equation for coupled neutron and gamma ray problems in three dimensional geometries by using the Monte Carlo method. This code is devoted both to shielding and criticality problems. The most important feature for particle transport equation solving is the fine treatment of the physical phenomena and sophisticated biasing technics useful for deep penetrations. The code is used either for shielding design studies or for reference and benchmark to validate cross sections. Neutronic studies are essentially cell or small core calculations and criticality problems. TRIPOLI-3 has been used as reference method, for example, for resonance self shielding qualification. (orig.)
Rehman Shakeel U.
2009-01-01
Full Text Available A primary-interaction based Monte Carlo algorithm has been developed for determination of the total efficiency of cylindrical scintillation g-ray detectors. This methodology has been implemented in a Matlab based computer program BPIMC. For point isotropic sources at axial locations with respect to the detector axis, excellent agreement has been found between the predictions of the BPIMC code with the corresponding results obtained by using hybrid Monte Carlo as well as by experimental measurements over a wide range of g-ray energy values. For off-axis located point sources, the comparison of the BPIMC predictions with the corresponding results obtained by direct calculations as well as by conventional Monte Carlo schemes shows good agreement validating the proposed algorithm. Using the BPIMC program, the energy dependent detector efficiency has been found to approach an asymptotic profile by increasing either thickness or diameter of scintillator while keeping the other fixed. The variation of energy dependent total efficiency of a 3'x3' NaI(Tl scintillator with axial distance has been studied using the BPIMC code. About two orders of magnitude change in detector efficiency has been observed for zero to 50 cm variation in the axial distance. For small values of axial separation, a similar large variation has also been observed in total efficiency for 137Cs as well as for 60Co sources by increasing the axial-offset from zero to 50 cm.
Kumar, Sudhir; Srinivasan, P; Sharma, S D; Saxena, Sanjay Kumar; Bakshi, A K; Dash, Ashutosh; Babu, D A R; Sharma, D N
2015-09-01
Isotope production and Application Division of Bhabha Atomic Research Center developed (32)P patch sources for treatment of superficial tumors. Surface dose rate of a newly developed (32)P patch source of nominal diameter 25 mm was measured experimentally using standard extrapolation ionization chamber and Gafchromic EBT film. Monte Carlo model of the (32)P patch source along with the extrapolation chamber was also developed to estimate the surface dose rates from these sources. The surface dose rates to tissue (cGy/min) measured using extrapolation chamber and radiochromic films are 82.03±4.18 (k=2) and 79.13±2.53 (k=2) respectively. The two values of the surface dose rates measured using the two independent experimental methods are in good agreement to each other within a variation of 3.5%. The surface dose rate to tissue (cGy/min) estimated using the MCNP Monte Carlo code works out to be 77.78±1.16 (k=2). The maximum deviation between the surface dose rates to tissue obtained by Monte Carlo and the extrapolation chamber method is 5.2% whereas the difference between the surface dose rates obtained by radiochromic film measurement and the Monte Carlo simulation is 1.7%. The three values of the surface dose rates of the (32)P patch source obtained by three independent methods are in good agreement to one another within the uncertainties associated with their measurements and calculation. This work has demonstrated that MCNP based electron transport simulations are accurate enough for determining the dosimetry parameters of the indigenously developed (32)P patch sources for contact brachytherapy applications. PMID:26086681
An accurate dose calculation in phantom and patient geometries requires an accurate description of the radiation source. Errors in the radiation source description are propagated through the dose calculation. With the emergence of linear accelerators whose dosimetric characteristics are similar to within measurement uncertainty, the same radiation source description can be used as the input to dose calculation for treatment planning at many institutions with the same linear accelerator model. Our goal in the current research was to determine the initial electron fluence above the linear accelerator target for such an accelerator to allow a dose calculation in water to within 1% or 1 mm of the measured data supplied by the manufacturer. The method used for both the radiation source description and the patient transport was Monte Carlo. The linac geometry was input into the Monte Carlo code using the accelerator's manufacturer's specifications. Assumptions about the initial electron source above the target were made based on previous studies. The free parameters derived for the calculations were the mean energy and radial Gaussian width of the initial electron fluence and the target density. A combination of the free parameters yielded an initial electron fluence that, when transported through the linear accelerator and into the phantom, allowed a dose-calculation agreement to the experimental ion chamber data to within the specified criteria at both 6 and 18 MV nominal beam energies, except near the surface, particularly for the 18 MV beam. To save time during Monte Carlo treatment planning, the initial electron fluence was transported through part of the treatment head to a plane between the monitor chambers and the jaws and saved as phase-space files. These files are used for clinical Monte Carlo-based treatment planning and are freely available from the authors
The author gives a scheme for the calculation of the self-shielding factors in the unresolved resonance region using the GRUCON applied program package. This package is especially created to be used in the conversion of evaluated neutron cross-section data, as available in existing data libraries, into multigroup microscopic constants. A detailed description of the formulae and algorithms used in the programs is given. Some typical examples of calculation are considered and the results are compared with those of other authors. The calculation accuracy is better than 2%
This paper concludes our efforts in describing SU(3)-Yang-Mills theories at different couplings/temperatures in terms of effective Polyakov-loop models. The associated effective couplings are determined through an inverse Monte Carlo procedure based on novel Schwinger-Dyson equations that employ the symmetries of the Haar measure. Because of the first-order nature of the phase transition we encounter a fine-tuning problem in reproducing the correct behavior of the Polyakov-loop from the effective models. The problem remains under control as long as the number of effective couplings is sufficiently small
Occurrence of hazardous accident in nuclear power plants and industrial units usually lead to release of radioactive materials and pollutants in environment. These materials and pollutants can be transported to a far downstream by the wind flow. In this paper, we implemented an atmospheric dispersion code to solve the inverse problem. Having received and detected the pollutants in one region, we may estimate the rate and location of the unknown source. For the modeling, one needs a model with ability of atmospheric dispersion calculation. Furthermore, it is required to implement a mathematical approach to infer the source location and the related rates. In this paper the AERMOD software and Bayesian inference along the Markov Chain Monte Carlo have been applied. Implementing, Bayesian approach and Markov Chain Monte Carlo for the aforementioned subject is not a new approach, but the AERMOD model coupled with the said methods is a new and well known regulatory software, and enhances the reliability of outcomes. To evaluate the method, an example is considered by defining pollutants concentration in a specific region and then obtaining the source location and intensity by a direct calculation. The result of the calculation estimates the average source location at a distance of 7km with an accuracy of 5m which is good enough to support the ability of the proposed algorithm.
According to dose parameters calculation formula of seed source recommended by AAPM TG43U1, 125I-103Pd seed source dose parameters calculation formula and a variety of radionuclides composite seed source of dose parameters calculation formula can be obtain. Dose rate constant, radial dose function and anisotropy function of 125I-103Pd composite seed source are calculated by Monte-Carlo method, Empiric equations are obtained for radial dose function and anisotropy function by curve fitting. Comparisons with the relative data recommend by AAPM are performed. For the single source, the deviation of dose rate constant is 0.959 (cGy·h-1·U-1), and with 0.6093% from the AAPM. (authors)
A Monte Carlo model for determination of binary diffusion coefficients in gases
Panarese, A.; Bruno, D.; Colonna, G.; Diomede, P.; Laricchiuta, A.; Longo, S.; Capitelli, M.
2011-06-01
A Monte Carlo method has been developed for the calculation of binary diffusion coefficients in gas mixtures. The method is based on the stochastic solution of the linear Boltzmann equation obtained for the transport of one component in a thermal bath of the second one. Anisotropic scattering is included by calculating the classical deflection angle in binary collisions under isotropic potential. Model results are compared to accurate solutions of the Chapman-Enskog equation in the first and higher orders. We have selected two different cases, H 2 in H 2 and O in O 2, assuming rigid spheres or using a model phenomenological potential. Diffusion coefficients, calculated in the proposed approach, are found in close agreement with Chapman-Enskog results in all the cases considered, the deviations being reduced using higher order approximations.
Efficiency determination of whole-body counter by Monte Carlo method, using a microcomputer
The purpose of this investigation was the development of an analytical microcomputer model to evaluate a whole body counter efficiency. The model is based on a modified Sryder's model. A stretcher type geometry along with the Monte Carlo method and a Synclair type microcomputer were used. Experimental measurements were performed using two phantoms, one as an adult and the other as a 5 year old child. The phantoms were made in acrylic and and 99mTc, 131I and 42K were the radioisotopes utilized. Results showed a close relationship between experimental and predicted data for energies ranging from 250 keV to 2 MeV, but some discrepancies were found for lower energies. (author)
The neutron multidetector consists of 81 detectors, made of 4x4x12 cmc BC-400 crystals mounted on XP2972 phototubes. This detector placed in the forward direction at 138 cm from the target, was used to detect the correlated neutrons in the fusion of Li11 halo nuclei with Si targets. To verify the criterion for selecting the true coincidences against cross-talk ( a spurious effect in which the same neutron is registered by two or more detectors) and to establish the optimal distance between adjacent detectors, the program MENATE ( written by P.Desesquelles, IPN - Orsay) was used to generate Monte Carlo neutrons and their interactions in multidetector. The results were analysed with PAW (from CERN Library). (authors)
Videira, Heber S.; Burkhardt, Guilherme M.; Santos, Ronielly S., E-mail: heber@cyclopet.com.br [Cyclopet Radiofarmacos Ltda., Curitiba, PR (Brazil); Passaro, Bruno M.; Gonzalez, Julia A.; Santos, Josefina; Guimaraes, Maria I.C.C. [Universidade de Sao Paulo (HCFMRP/USP), Sao Paulo, SP (Brazil). Faculdade de Medicina. Hospital das Clinicas; Lenzi, Marcelo K. [Universidade Federal do Parana (UFPR), Curitina (Brazil). Programa de Pos-Graduacao em Engenharia Quimica
2013-04-15
The technological advances in positron emission tomography (PET) in conventional clinic imaging have led to a steady increase in the number of cyclotrons worldwide. Most of these cyclotrons are being used to produce {sup 18}F-FDG, either for themselves as for the distribution to other centers that have PET. For there to be safety in radiological facilities, the cyclotron intended for medical purposes can be classified in category I and category II, ie, self-shielded or non-shielded (bunker). Therefore, the aim of this work is to verify the effectiveness of borated water shield built for a cyclotron accelerator-type Self-shielded PETtrace 860. Mixtures of water borated occurred in accordance with the manufacturer’s specifications, as well as the results of the radiometric survey in the vicinity of the self-shielding of the cyclotron in the conditions established by the manufacturer showed that radiation levels were below the limits. (author)
Portal monitoring radiation detectors are commonly used by steel industries in the probing and detection of radioactivity contamination in scrap metal. These portal monitors typically consist of polystyrene or polyvinyltoluene (PVT) plastic scintillating detectors, one or more photomultiplier tubes (PMT), an electronic circuit, a controller that handles data output and manipulation linking the system to a display or a computer with appropriate software and usually, a light guide. Such a portal used by the steel industry was opened and all principal materials were simulated using a Monte Carlo simulation tool (MCNP4C2). Various source-detector configurations were simulated and validated by comparison with corresponding measurements. Subsequently an experiment with a uniform cargo along with two sets of experiments with different scrap loads and radioactive sources (137Cs, 152Eu) were performed and simulated. Simulated and measured results suggested that the nature of scrap is crucial when simulating scrap load-detector experiments. Using the same simulating configuration, a series of runs were performed in order to estimate minimum alarm activities for 137Cs, 60Co and 192Ir sources for various simulated scrap densities. The minimum alarm activities as well as the positions in which they were recorded are presented and discussed.
Determination of shielding parameters for different types of concretes by Monte Carlo methods
The chose of a suitable concrete composition for a biological reactor shield remain as a research target up to now. In the present study the attempts has been made to estimate the influence of the concrete aggregates on the shielding parameters for three type of ordinary, serpentine and steel magnetite concrete by Monte Carlo N-Particle (MCNP ) transport code. MCNP calculations have been performed in order to obtain the leakage of neutrons, photons and electrons from dry shield. Also the mass attenuation coefficients and the liner attenuation coefficient are estimated for neutron and photon in those energies in range of actual energy which exist out of pressure vessel of power reactor in the cavity for the investigated concretes. The concrete densities ranged from 2.3 to 5.11 g/cm3. These calculations were done in the condition of a typical commercial Pressurized Water Reactor (PWR). The results show that Steel-magnetite concrete, with high density (5.11 g/cm3) and constituents of relatively high atomic number, is an effective shield for both photons and neutrons
McNamara, A L; Heijnis, H; Fierro, D; Reinhard, M I
2012-04-01
A Compton suppressed high-purity germanium (HPGe) detector is well suited to the analysis of low levels of radioactivity in environmental samples. The difference in geometry, density and composition of environmental calibration standards (e.g. soil) can contribute to excessive experimental uncertainty to the measured efficiency curve. Furthermore multiple detectors, like those used in a Compton suppressed system, can add complexities to the calibration process. Monte Carlo simulations can be a powerful complement in calibrating these types of detector systems, provided enough physical information on the system is known. A full detector model using the Geant4 simulation toolkit is presented and the system is modelled in both the suppressed and unsuppressed mode of operation. The full energy peak efficiencies of radionuclides from a standard source sample is calculated and compared to experimental measurements. The experimental results agree relatively well with the simulated values (within ∼5 - 20%). The simulations show that coincidence losses in the Compton suppression system can cause radionuclide specific effects on the detector efficiency, especially in the Compton suppressed mode of the detector. Additionally since low energy photons are more sensitive to small inaccuracies in the computational detector model than high energy photons, large discrepancies may occur at energies lower than ∼100 keV. PMID:22304994
The quality correction factor for used beam and qualities is strongly required for clinical dosimetry by TRS-398 protocol of IAEA. In this study the quality correction factors for a commercial plane-parallel ionization chamber in high energy electron beams were calculated by Monte Carlo code (DOSRZnrc/EGSnrc). In comparison of quality correction factor, the difference between this study and TRS-398 were within 1% in 5-20 MeV. In case of 4MeV the difference was 1.9%. As an independent method of determination of quality correction factor this study can be applied to evaluate values in the protocol or calculate the factor for a new chamber.
Program GROUPIE reads evaluated data in the ENDF/B format and uses these data to calculate Bondarenko self-shielded cross sections and multiband parameters. To give as much generality as possible, the program allows the user to specify arbitrary energy groups and an arbitrary energy groups and an arbitrary energy-dependent neutron spectrum (weighing function). To guarantee the accuracy of the results, all integrals are performed analytically; in no case is iteration or any approximate form of integration used. The output from this program includes both listings and multiband parameters suitable for use either in a normal multigroup transport calculation or in a multiband transport calculation. A listing of the source deck is available on request
Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes
The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)
Sample Size Determination for Regression Models Using Monte Carlo Methods in R
Beaujean, A. Alexander
2014-01-01
A common question asked by researchers using regression models is, What sample size is needed for my study? While there are formulae to estimate sample sizes, their assumptions are often not met in the collected data. A more realistic approach to sample size determination requires more information such as the model of interest, strength of the…
Field gamma spectrometers are widely used to determine gamma dose rates in sedimentary media. However the most widely used technique-the 'window technique'-is time consuming and introduces important statistical uncertainty in the determination of the radioelement contents, and finally on the gamma dose rate. The threshold technique directly relates the number of counts recorded above certain threshold energy to the gamma dose rate. Recently new experimental measurements further investigated this technique but it has not been tested in various sedimentary media. In this paper, numerical simulations using a specifically designed Geant4 code allow to test the sensitivity of this technique to changes of sediments nature, humidity content and disequilibrium in the U-series. Finally another threshold technique, relating the gamma dose rate to the energy per unit time deposited above another threshold energy, is investigated. It is shown than the latter has a number of advantages compared to the classical techniques. Experimental results testing this approach are presented.
Booth, George H.; Thom, Alex J. W.; Alavi, Ali
2009-08-01
We have developed a new quantum Monte Carlo method for the simulation of correlated many-electron systems in full configuration-interaction (Slater determinant) spaces. The new method is a population dynamics of a set of walkers, and is designed to simulate the underlying imaginary-time Schrödinger equation of the interacting Hamiltonian. The walkers (which carry a positive or negative sign) inhabit Slater determinant space, and evolve according to a simple set of rules which include spawning, death and annihilation processes. We show that this method is capable of converging onto the full configuration-interaction (FCI) energy and wave function of the problem, without any a priori information regarding the nodal structure of the wave function being provided. Walker annihilation is shown to play a key role. The pattern of walker growth exhibits a characteristic plateau once a critical (system-dependent) number of walkers has been reached. At this point, the correlation energy can be measured using two independent methods—a projection formula and a energy shift; agreement between these provides a strong measure of confidence in the accuracy of the computed correlation energies. We have verified the method by performing calculations on systems for which FCI calculations already exist. In addition, we report on a number of new systems, including CO, O2, CH4, and NaH—with FCI spaces ranging from 109 to 1014, whose FCI energies we compute using modest computational resources.
This study explores for the first time the possibility of using an evolutionary algorithm (EA) for the determination of atomistic rates for use in kinetic Monte Carlo (KMC) simulations of anisotropic etching of silicon for the manufacture of microelectromechanical systems (MEMS). Traditionally, KMC rates are determined based on (i) computationally expensive density functional theory (DFT) calculations or, when possible, (ii) a combination of physical insight and a labor-intensive, manual procedure where, e.g., experimental and simulated surface morphologies are visually matched for a collection of surface orientations. Compared to these approaches, the evolutionary KMC method proposed in this study provides a more flexible, autonomous procedure to describe correctly a wide variety of etching conditions. We focus on the use of a functional representation of the atomistic rates, referred to as the removal probability function (RPF). This simplifies the EA search space to just a few parameters and reduces the number of required experimental data points to just a few etch rates. By proposing two alternative RPFs with four and six parameters, respectively, we show that the ability to explain the orientation dependence of the etch rate for a wide variety of etchants increases with the number of parameters and conclude that the six-parameter RPF provides sufficiently good simulations for a wide range of etching conditions, including KOH, KOH+IPA, TMAH and TMAH+Triton at different concentrations and temperatures. By uncovering the relationships between the parameters and the concentration of the etchant, it is possible to extend the simulations to nonmeasured etching conditions. Although the use of an RPF effectively restricts the search to a subspace of the atomistic rates, the present results suggest that physically meaningful KMC rates can probably be determined in the near future by direct comparison of macroscopic experiments and simulations through the use of
Xing, Y.; Gosálvez, M. A.; Sato, K.; Tian, M.; Yi, H.
2012-08-01
This study explores for the first time the possibility of using an evolutionary algorithm (EA) for the determination of atomistic rates for use in kinetic Monte Carlo (KMC) simulations of anisotropic etching of silicon for the manufacture of microelectromechanical systems (MEMS). Traditionally, KMC rates are determined based on (i) computationally expensive density functional theory (DFT) calculations or, when possible, (ii) a combination of physical insight and a labor-intensive, manual procedure where, e.g., experimental and simulated surface morphologies are visually matched for a collection of surface orientations. Compared to these approaches, the evolutionary KMC method proposed in this study provides a more flexible, autonomous procedure to describe correctly a wide variety of etching conditions. We focus on the use of a functional representation of the atomistic rates, referred to as the removal probability function (RPF). This simplifies the EA search space to just a few parameters and reduces the number of required experimental data points to just a few etch rates. By proposing two alternative RPFs with four and six parameters, respectively, we show that the ability to explain the orientation dependence of the etch rate for a wide variety of etchants increases with the number of parameters and conclude that the six-parameter RPF provides sufficiently good simulations for a wide range of etching conditions, including KOH, KOH+IPA, TMAH and TMAH+Triton at different concentrations and temperatures. By uncovering the relationships between the parameters and the concentration of the etchant, it is possible to extend the simulations to nonmeasured etching conditions. Although the use of an RPF effectively restricts the search to a subspace of the atomistic rates, the present results suggest that physically meaningful KMC rates can probably be determined in the near future by direct comparison of macroscopic experiments and simulations through the use of
无
2009-01-01
The magnetic anisotropy field in thin films with in-plane uniaxial anisotropy can be deduced from the VSM magnetization curves measured in magnetic fields of constant magnitudes. This offers a new possibility of applying rotational magnetization curves to determine the firstand second-order anisotropy constant in these films. In this paper we report a theoretical derivation of rotational magnetization curve in hexagonal crystal system with easy-plane anisotropy based on the principle of the minimum total energy. This model is applied to calculate and analyze the rotational magnetization process for magnetic spherical particles with hexagonal easy-plane anisotropy when rotating the external magnetic field in the basal plane. The theoretical calculations are consistent with Monte Carlo simulation results. It is found that to well reproduce experimental curves, the effect of coercive force on the magnetization reversal process should be fully considered when the intensity of the external field is much weaker than that of the anisotropy field. Our research proves that the rotational magnetization curve from VSM measurement provides an effective access to analyze the in-plane anisotropy constant K3 in hexagonal compounds, and the suitable experimental condition to measure K3 is met when the ratio of the magnitude of the external field to that of the anisotropy field is around 0.2.
WANG AiMin; PANG Hua
2009-01-01
The magnetic anisotropy field in thin films with in-plane uniaxial anisotropy can be deduced from the VSM magnetization curves measured in magnetic fields of constant magnitudes. This offers a new possibility of applying rotational magnetization curves to determine the first- and second-order ani-aotropy constant in these films. In this paper we report a theoretical derivation of rotational magnetiza-tion curve in hexagonal crystal system with easy-plane anisotropy based on the principle of the minimum total energy. This model is applied to calculate and analyze the rotational magnetization process for magnetic spherical particles with hexagonal easy-plane anisotropy when rotating the external magnetic field in the basal plane. The theoretical calculations are consistent with Monte Carlo simulation results. It is found that to well reproduce experimental curves, the effect of coercive force on the magnetization reversal process should be fully considered when the intensity of the ex-ternal field is much weaker than that of the anisotropy field. Our research proves that the rotational magnetization curve from VSM measurement provides an effective access to analyze the in-plane anisotropy constant K3 in hexagonal compounds, and the suitable experimental condition to measure K3 is met when the ratio of the magnitude of the external field to that of the anisotropy field is around 0.2.
Kianoush Fathi Vajargah
2014-01-01
The accuracy of Monte Carlo and quasi-Monte Carlo methods decreases in problems of high dimensions. Therefore, the objective of this study was to present an optimum method to increase the accuracy of the answer. As the problem gets larger, the resulting accuracy will be higher. In this respect, this study combined the two previous methods, QMC and MC, and presented a hybrid method with efficiency higher than that of those two methods.
As part of developing its nuclear infrastructure base, the National Science and Technology Center Nuclear (CNSTN) examines the technical feasibility of setting up a new installation of subcritical assembly. Our study focuses on determining the neutron parameters of a nuclear zero power reactor based on Monte Carlo simulation MCNP. The objective of the simulation is to model the installation, determine the effective multiplication factor, and spatial distribution of neutron flux.
Purpose: To investigate beam quality correction factors for the flattening-filter-free (FFF) energies of the TrueBeam™ accelerator based on a dosimetry formalism for small and nonstandard fields. Methods: Three detectors - an Exradin W1 scintillator, Sun Nuclear EDGE diode, and LiF(Mg,Tl) TLD-100 chips - were investigated to determine their applicability as tools to measure quality correction factors for ionization chambers in the small and nonstandard fields of the TrueBeam™. Volume-averaging effects and energy dependence were observed in fields ranging from 1×1 to 40×40 cm2 for 6 MV and 10 MV beam energies using both FFF and flattened beam modes. Correction factors were determined for three ionization chambers: an Exradin A12 Farmer-type chamber, an Exradin A1SL scanning chamber, and an Exradin A26 reference-class microchamber. Beam quality corrections were also obtained using a benchmarked model of the TrueBeam™ created with the BEAMnrc user code of EGSnrc. Results: All three detectors demonstrated measureable energy dependence in the megavoltage range. The EDGE diode was deemed the most appropriate tool for beam quality correction factor measurements due to its low energy dependence and small size; however, alanine will be used in the future to reduce energy dependent effects even further. Measured kQmsr,Q corrections of up to 4% were found for the 6MV FFF and 10 MV FFF beams, corresponding to a discrepancy of up to 3% compared to TG-51-determined dose. Up to a 10% kQclin,Qmsr correction was measured for small fields referenced to a 10×10 cm2 field of the same energy. Much larger corrections were determined using the Monte Carlo model, and these discrepancies require further investigation. Conclusion: Progress has been made toward determining beam quality correction factors for the small and nonstandard fields of the TrueBeam™ accelerator. Further work must be done to ensure greater accuracy in patient treatments with this new modality
王长江; 曾惠林; 杨雪梅; 蒋戎; 刘然
2014-01-01
Aiming at the X80 high strength pipeline steel and related welding technical specification requirements,it adopted independent development eight welding torch internal welder and self-shielded flux-cored wire automatic welder to carry out pipeline all-position welding test,and conducted mechanical properties testing of welded joints. It determined reasonable groove type,welding wire type and matching welding parameters,and formulated welding technology internal welding machine for root welding+self-shielded flux-cored wire automatic welding for fill and cover welding. The test results proved that the welding process is reasonable,simple operation,and high efficiency,the various performance indicators of welded joints all meet the X80 steel grade pipeline engineering welding specification and relevant national standards.%针对X80高强管线钢及相关焊接技术规范要求，应用自主研发的八焊炬内焊机及自保护药芯焊丝自动焊机进行管道全位置焊接试验，并进行焊接接头力学性能测试。确定了合理的坡口形式、焊丝型号及与之匹配的焊接参数，制定了“内焊机根焊+管道自保护药芯焊丝自动焊填充、盖面焊”的焊接工艺。试验结果证明，焊接工艺制定合理，操作简单，效率高，其焊接接头的各项性能指标均满足X80钢级管道工程焊接规范及国家相关标准要求。
To evaluate the ionizing radiation effects on the BFO it is necessary to take into account the dependence of the dose effects on BFO, shielding and self-shielding of different parts of the BFO and physical characteristics of radiations. The evaluations of the BFO shielding were made on the basis of the shielding calculation of 330 points in which BFO were placed. The calculations show that tissue thickness of the shielding BFO varies from 1 to 70 g/cm2. The percentage of thicknesses of 2,5 and 10 g/cm2 is 5,22 and 50 respectively. With the help of these data the proton doses from solar cosmic rays with characteristic rigidities in the range from 60 to 200 MV were calculated. These results show that the effective thickness of BFO shielding depends strongly on the characteristic rigidity. As a result of the irradiation irregularity of different parts of the BFO and the type of the dose effect function, the radiation effects of the cosmic ray protons may be evaluated with the help of the dose value mean with respect to BFO
1 1/2 years of experience with a 10 MeV self-shielded on-line e-beam sterilization system
The Vascular Intervening Group of the Guidant Corporation (Guidant IV) has been operating a self-shielded, 10 MeV 4 kW, electron beam sterilization system since July of 1988. The system was designed, built and installed in a 70 square meter area in an existing Guidant manufacturing facility by Titan Scan Corporation and performance of the system was validated in conformance with 1S0-11137 standards. The goal of this on-site e-beam system was 'just in time' JIT, sterilization, i.e. the ability to manufacture, sterilize and ship, high intrinsic value medical devices in less than 24 hours. The benefits of moving from a long gas sterilization cycle of greater than one week to a JIT process were envisioned to be a) speed to market with innovated new products b) rapid response to customer requirements c) reduced inventory carrying costs and finally manufacturing and quality system efficiency. The ability of Guidant to realize these benefits depended upon the ability of the Guidant VI business units to adapt to the new sterilization modality and functionality and on the overall system reliability. This paper reviews the operating experience to date and the overall system reliability. (author)
This paper describes the process of installation of a self-shielded irradiator category I, model ISOGAMMA LL.Co of 60Co, with a nominal 25 kCi activity, rate of absorbed dose 8 kG/h and 5 L workload. The stages are describe step by step: import, the customs procedure which included the interview with the master of the vessel transporter, the monitoring of the entire process by the head of radiological protection of the importing Center, control of the levels of surface contamination of the shipping container of the sources before the removal of the ship, the supervision of the national regulatory authority and the transportation to the final destination. Details of assembling of the installation and the opening of the container for transportation of supplies is outlined. The action plan previously developed for the case of occurrence of radiological successful events is presented, detailing the phase of the load of radioactive sources by the specialists of the company selling the facility (IZOTOP). Finally describes the setting and implementation of the installation and the procedure of licensing for exploitation
Benmakhlouf, H.; Johansson, J.; Paddick, I.; Andreo, P.
2015-05-01
The measurement of output factors (OF) for the small photon beams generated by Leksell Gamma Knife® (LGK) radiotherapy units is a challenge for the physicist due to the under or over estimation of these factors by a vast majority of the detectors commercially available. Output correction factors, introduced in the international formalism published by Alfonso (2008 Med. Phys. 35 5179-86), standardize the determination of OFs for small photon beams by correcting detector-reading ratios to yield OFs in terms of absorbed-dose ratios. In this work output correction factors for a number of detectors have been determined for LGK Perfexion™ 60Co γ-ray beams by Monte Carlo (MC) calculations and measurements. The calculations were made with the MC system PENELOPE, scoring the energy deposited in the active volume of the detectors and in a small volume of water; the detectors simulated were two silicon diodes, one liquid ionization chamber (LIC), alanine and TLD. The calculated LIC output correction factors were within ± 0.4%, and this was selected as the reference detector for experimental determinations where output correction factors for twelve detectors were measured, normalizing their readings to those of the LIC. The MC-calculated and measured output correction factors for silicon diodes yielded corrections of up to 5% for the smallest LGK collimator size of 4 mm diameter. The air ionization chamber measurements led to extremely large output correction factors, caused by the well-known effect of partial volume averaging. The corrections were up to 7% for the natural diamond detector in the 4 mm collimator, also due to partial volume averaging, and decreased to within about ± 0.6% for the smaller synthetic diamond detector. The LIC, showing the smallest corrections, was used to investigate machine-to-machine output factor differences by performing measurements in four LGK units with different dose rates. These resulted in OFs within ± 0.6% and ± 0
The accuracy of estimation of the self-attenuation correction Cs with the Cutshall transmission method in 210Pb measurements by gamma-spectrometry was assessed using the Monte Carlo method. The Cutshall method overestimates the correction for samples with linear attenuation coefficient at 46.5 keV higher than that of the standard and underestimates it in the opposite case. The highest bias was found for thick samples. Cs,Cuts/Cs ratio grows linearly with sample linear attenuation coefficient. - Highlights: • The Cutshall method enables to determine self-attenuation correction experimentally. • The Cutshall method was validated by Monte Carlo simulations. • The Cutshall method provides systematically inaccurate Cs correction values. • The bias depends on relation between attenuation properties of sample and standard. • The highest bias was found for thick samples
A method for tuning parameters in Monte Carlo generators is described and applied to a specific case. The method works in the following way: each observable is generated several times using different values of the parameters to be tuned. The output is then approximated by some analytic form to describe the dependence of the observables on the parameters. This approximation is used to find the values of the parameter that give the best description of the experimental data. This results in significantly faster fitting compared to an approach in which the generator is called iteratively. As an application, we employ this method to fit the parameters of the unintegrated gluon density used in the Cascade Monte Carlo generator, using inclusive deep inelastic data measured by the H1 Collaboration. We discuss the results of the fit, its limitations, and its strong points. (orig.)
The Monte Carlo method can be used to compute the gamma-ray backscattering albedo. This method was used by Raso to compute the angular differential albedo. Raso's results have been used by Chilton and Huddelston to adjust their well-known albedo formula. Here, an efficient estimator is proposed to compute the double-differential angular and energetic albedo from gamma-ray histories simulated in matter by the three-dimensional Monte Carlo transport code TRIPOLI. A detailed physical albedo analysis could be done in this way. The double-differential angular and energetic gamma-ray albedo is calculated for iron material for initial gamma-ray energies of 8, 3, 1, and 0.5 MeV
A new method to accurately determine gas saturation in the tight gas reservoir using a three-detector pulsed neutron logging tool was proposed. Formation porosity is varied from 2% to 15% to simulate the distribution of thermal neutron under different borehole and formation conditions by using Monte Carlo method. The study result shows that the difference of three detectors counts can be used to determine gas saturation and have higher sensitivity than counting the ratio of different detectors. - Highlights: • A new method proposed to determine gas saturation in tight gas reservoir. • The new method has high sensitivity to determine gas saturation than traditional logging method. • Influence factors of the new method determining gas saturation were studied
We developed a novel Monte Carlo simulation model to investigate the line width dependence of the transport properties of multi-layered graphene nanoribbon (GNR) interconnects with edge roughness. We reported that the line width dependence of carrier mobility decreases significantly as the magnitude of the edge roughness gets smaller, which agrees well with experiments. We also discussed the influence of the inelasticity of edge roughness scatterings, inter-layer tunneling, and line width dependent band structures on the line width of the GNR interconnects. (author)
Mohammad W. Marashdeh; Ibrahim F. Al-Hamarneh; Eid M. Abdel Munem; A.A. Tajuddin; Alawiah Ariffin; Saleh Al-Omari
2015-01-01
Rhizophora spp. wood has the potential to serve as a solid water or tissue equivalent phantom for photon and electron beam dosimetry. In this study, the effective atomic number (Zeff) and effective electron density (Neff) of raw wood and binderless Rhizophora spp. particleboards in four different particle sizes were determined in the 10–60 keV energy region. The mass attenuation coefficients used in the calculations were obtained using the Monte Carlo N-Particle (MCNP5) simulation code. The M...
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
Development of 3d reactor burnup code based on Monte Carlo method and exponential Euler method
Burnup analysis plays a key role in fuel breeding, transmutation and post-processing in nuclear reactor. Burnup codes based on one-dimensional and two-dimensional transport method have difficulties in meeting the accuracy requirements. A three-dimensional burnup analysis code based on Monte Carlo method and Exponential Euler method has been developed. The coupling code combines advantage of Monte Carlo method in complex geometry neutron transport calculation and FISPACT in fast and precise inventory calculation, meanwhile resonance Self-shielding effect in inventory calculation can also be considered. The IAEA benchmark text problem has been adopted for code validation. Good agreements were shown in the comparison with other participants' results. (authors)
Purpose: Output factor determination for small fields (less than 20 mm) presents significant challenges due to ion chamber volume averaging and diode over-response. Measured output factor values between detectors are known to have large deviations as field sizes are decreased. No set standard to resolve this difference in measurement exists. We observed differences between measured output factors of up to 14% using two different detectors. Published Monte Carlo derived correction factors were used to address this challenge and decrease the output factor deviation between detectors. Methods: Output factors for Elekta's linac-based stereotactic cone system were measured using the EDGE detector (Sun Nuclear) and the A16 ion chamber (Standard Imaging). Measurements conditions were 100 cm SSD (source to surface distance) and 1.5 cm depth. Output factors were first normalized to a 10.4 cm × 10.4 cm field size using a daisy-chaining technique to minimize the dependence of field size on detector response. An equation expressing the relation between published Monte Carlo correction factors as a function of field size for each detector was derived. The measured output factors were then multiplied by the calculated correction factors. EBT3 gafchromic film dosimetry was used to independently validate the corrected output factors. Results: Without correction, the deviation in output factors between the EDGE and A16 detectors ranged from 1.3 to 14.8%, depending on cone size. After applying the calculated correction factors, this deviation fell to 0 to 3.4%. Output factors determined with film agree within 3.5% of the corrected output factors. Conclusion: We present a practical approach to applying published Monte Carlo derived correction factors to measured small field output factors for the EDGE and A16 detectors. Using this method, we were able to decrease the percent deviation between both detectors from 14.8% to 3.4% agreement
Yan, Yangqian; Blume, D
2016-06-10
The unitary equal-mass Fermi gas with zero-range interactions constitutes a paradigmatic model system that is relevant to atomic, condensed matter, nuclear, particle, and astrophysics. This work determines the fourth-order virial coefficient b_{4} of such a strongly interacting Fermi gas using a customized ab initio path-integral Monte Carlo (PIMC) algorithm. In contrast to earlier theoretical results, which disagreed on the sign and magnitude of b_{4}, our b_{4} agrees within error bars with the experimentally determined value, thereby resolving an ongoing literature debate. Utilizing a trap regulator, our PIMC approach determines the fourth-order virial coefficient by directly sampling the partition function. An on-the-fly antisymmetrization avoids the Thomas collapse and, combined with the use of the exact two-body zero-range propagator, establishes an efficient general means to treat small Fermi systems with zero-range interactions. PMID:27341213
Yan, Yangqian; Blume, D.
2016-06-01
The unitary equal-mass Fermi gas with zero-range interactions constitutes a paradigmatic model system that is relevant to atomic, condensed matter, nuclear, particle, and astrophysics. This work determines the fourth-order virial coefficient b4 of such a strongly interacting Fermi gas using a customized ab initio path-integral Monte Carlo (PIMC) algorithm. In contrast to earlier theoretical results, which disagreed on the sign and magnitude of b4 , our b4 agrees within error bars with the experimentally determined value, thereby resolving an ongoing literature debate. Utilizing a trap regulator, our PIMC approach determines the fourth-order virial coefficient by directly sampling the partition function. An on-the-fly antisymmetrization avoids the Thomas collapse and, combined with the use of the exact two-body zero-range propagator, establishes an efficient general means to treat small Fermi systems with zero-range interactions.
Nowadays, radioactive isotopes are used in many different fields, for instance in industry, energy production, archaeology and mainly in medical applications. In addition, bricks and stones, which are used to build these buildings and our homes, have higher natural radiation levels than other building materials such as wood. In this work, the linear and mass attenuation coefficients of different types building materials, needed for the protection of human health against radiation hazards, were investigated with Monte Carlo particle-transport code (MCNP) technique. Simulations were performed in order to obtain these coefficients at photon energies from 80 keV to 1333 keV for clay, perlite and PP. As should be anticipated, the density and photon energy are the main parameters that affect the mass attenuation coefficient
A new approach based on the Monte Carlo simulation is used to calculate the infinite matrix dose rate correction factors of gamma, beta and internal conversion radiations for 250 μm diameter grains of quartz and TLD500 chips. Here, the dependence of the correction factor on the radiation energy is initially calculated for each type of emitted particle and with this result the correction factors for the 232Th and 238U series and 40K are determined. This analysis is made for dry soil and also for different levels of water content in it. The obtained beta correction factors for quartz are in good agreement with those previously reported. For the TLD500 chip certain differences with previously reported data are found. The analysis of the gamma water correction factor for quartz based on Zimmerman equation shows the correspondence with the similar correction factor for electrons. In the case of TLD500 chip a gamma water correction factor value of 1.0 was found. - Highlights: • A new approach based on Monte Carlo simulation is used to compute infinite matrix dose rate correction factors. • Infinite matrix models with real dimensions were analyzed within 3% uncertainties. • The dependence of grain size attenuation on particle energy is determined. • The same dependence for water correction factors is also analyzed
Viel, Alexandra; Coutinho-Neto, Maurício D.; Manthe, Uwe
2007-01-01
Quantum dynamics calculations of the ground state tunneling splitting and of the zero point energy of malonaldehyde on the full dimensional potential energy surface proposed by Yagi et al. [J. Chem. Phys. 1154, 10647 (2001)] are reported. The exact diffusion Monte Carlo and the projection operator imaginary time spectral evolution methods are used to compute accurate benchmark results for this 21-dimensional ab initio potential energy surface. A tunneling splitting of 25.7±0.3cm-1 is obtained, and the vibrational ground state energy is found to be 15122±4cm-1. Isotopic substitution of the tunneling hydrogen modifies the tunneling splitting down to 3.21±0.09cm-1 and the vibrational ground state energy to 14385±2cm-1. The computed tunneling splittings are slightly higher than the experimental values as expected from the potential energy surface which slightly underestimates the barrier height, and they are slightly lower than the results from the instanton theory obtained using the same potential energy surface.
The analysis of depth-dose distributions in bricks sampled from walls in areas with nuclear waste or accident contamination has the potential of providing information on the energy and source configuration of the γ-radiation that had been incident on the brick. In this study, a brick from a mill facing a shallow water reservoir of the contaminated Techa river in the South Ural region is investigated. Thermoluminescene (TL) methods were used to measure the accumulated dose at several depths in the brick. The accidental external γ-dose is obtained by subtracting the natural radiation background dose from the total accumulated dose. In the first segment of the brick, at a depth of about 1.5 cm, the accident dose was found to be roughly 3.5 Gy. Monte-Carlo simulations of the photon transport form the reservoir bed contaminated with 137Cs were calculated for different depths in the brick. The calculations were made assuming different attenuating water levels. It is found that the depth-dose distribution determined by measurements corresponds to a water level between 20 and 50 cm. The results indicate that the TL measurements combined with Monte-Carlo modelling calculations are highly promising for external γ-dose reconstruction applications. (Author)
OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo
1 - Description of problem or function: OMEGA is a Monte Carlo code for the solution of the stationary neutron transport equation with k-eff as the Eigenvalue. A three-dimensional geometry is permitted consisting of a very general arrangement of three basic shapes (columns with circular, rectangular, or hexagonal cross section with a finite height and different material layers along their axes). The main restriction is that all the basic shapes must have parallel axes. Most real arrangements of fissile material inside and outside a reactor (e.g., in a fuel storage or transport container) can be described without approximation. The main field of application is the estimation of criticality safety. Many years of experience and comparison with reference cases have shown that the code together with the built-in cross section libraries gives reliable results. The following results can be calculated: - the effective multiplication factor k-eff; - the flux distribution; - reaction rates; - spatially and energetically condensed cross sections for later use in a subsequent OMEGA run. A running job may be interrupted and continued later, possibly with an increased number of batches for an improved statistical accuracy. The geometry as well as the k-eff results may be visualized. The use of the code is demonstrated by many illustrating examples. 2 - Method of solution: The Monte Carlo method is used with neutrons starting from an initial source distribution. The histories of a generation (or batch) of neutrons are followed from collision to collision until the histories are terminated by capture, fission, or leakage. For the solution of the Eigenvalue problem, the starting positions of the neutrons for a given generation are determined by the fission points of the preceding generation. The summation of the results starts only after some initial generations when the spatial part of the fission source has converged. At present the code uses the BNAB-78 subgroup library of the
The extraction of petroleum from its geological reservoirs involves the piping of crude oil, water and natural gas from the wells as a multiphase mixture. Knowledge of the mass flow rate of each component is required to control production. In this work, this multiphase flow has been pictured as a coarse-grained component system, water as drops and natural gas as bubbles, both are carried out through the crude oil along the pipeline. According to this picture, a Monte Carlo model for grain size dependent gamma-ray attenuation has been constructed and applied to determine the volume fractions of oil, water and natural gas in pipelines by using gamma-ray transmission technique
The Monte-Carlo refractive index matching (MCRIM) technique was developed to determine the physical properties of heavy inorganic scintillators (HIS) which are difficult to measure experimentally. It was designed as a method for obtaining input parameters for Monte-Carlo (MC) simulations of experimental arrangements incorporating HIS in their setups. The MCRIM technique is used to estimate the intrinsic light yield, the scattering coefficient and the absorption coefficient, herein referred to as indirect measurement properties. The MCRIM technique uses an experiment/MC combination to determine these indirect measurement properties. The MCRIM experimental setup comprises a crystal placed on a photomultiplier tube window with the possibility of introducing materials of different refractive indices in a small gap between the crystal and photomultiplier tube (PMT) window. The dependence of the measured light yield on the refractive index of the material in the gap can only be reproduced by simulations if the correct values of scattering, absorption and intrinsic light yield are used. The experimental setup is designed to minimise the presence of optical components such as unpolished surfaces and non-ideal reflectors, which are difficult to simulate. The MCRIM technique is tested on a 1.03x1.00x0.82 cm3 crystal of CaWO4 which is found to have a scattering coefficient of 0.061±0.005 cm-1, an absorption coefficient of 0.065±0.005 cm-1, and an intrinsic light yield of 22700±1700 photons/MeV
MO-G-BRF-05: Determining Response to Anti-Angiogenic Therapies with Monte Carlo Tumor Modeling
Valentinuzzi, D [Jozef Stefan Institute, Ljubljana (Slovenia); Simoncic, U; Jeraj, R [Jozef Stefan Institute, Ljubljana (Slovenia); University of Wisconsin, Madison, WI (United States); Titz, B [University of Wisconsin, Madison, WI (United States)
2014-06-15
Purpose: Patient response to anti-angiogenic therapies with vascular endothelial growth factor receptor - tyrosine kinase inhibitors (VEGFR TKIs) is heterogeneous. This study investigates key biological characteristics that drive differences in patient response via Monte Carlo computational modeling capable of simulating tumor response to therapy with VEGFR TKI. Methods: VEGFR TKIs potently block receptors, responsible for promoting angiogenesis in tumors. The model incorporates drug pharmacokinetic and pharmacodynamic properties, as well as patientspecific data of cellular proliferation derived from [18F]FLT-PET data. Sensitivity of tumor response was assessed for multiple parameters, including initial partial oxygen tension (pO{sub 2}), cell cycle time, daily vascular growth fraction, and daily vascular regression fraction. Results were benchmarked to clinical data (patient 2 weeks on VEGFR TKI, followed by 1-week drug holiday). The tumor pO{sub 2} was assumed to be uniform. Results: Among the investigated parameters, the simulated proliferation was most sensitive to the initial tumor pO{sub 2}. Initial change of 5 mmHg can already Result in significantly different levels of proliferation. The model reveals that hypoxic tumors (pO{sub 2} ≥ 20 mmHg) show the highest decrease of proliferation, experiencing mean FLT standardized uptake value (SUVmean) decrease for at least 50% at the end of the clinical trial (day 21). Oxygenated tumors (pO{sub 2} 20 mmHg) show a transient SUV decrease (30–50%) at the end of the treatment with VEGFR TKI (day 14) but experience a rapid SUV rebound close to the pre-treatment SUV levels (70–110%) at the time of a drug holiday (day 14–21) - the phenomenon known as a proliferative flare. Conclusion: Model's high sensitivity to initial pO{sub 2} clearly emphasizes the need for experimental assessment of the pretreatment tumor hypoxia status, as it might be predictive of response to antiangiogenic therapies and the occurrence
MO-G-BRF-05: Determining Response to Anti-Angiogenic Therapies with Monte Carlo Tumor Modeling
Purpose: Patient response to anti-angiogenic therapies with vascular endothelial growth factor receptor - tyrosine kinase inhibitors (VEGFR TKIs) is heterogeneous. This study investigates key biological characteristics that drive differences in patient response via Monte Carlo computational modeling capable of simulating tumor response to therapy with VEGFR TKI. Methods: VEGFR TKIs potently block receptors, responsible for promoting angiogenesis in tumors. The model incorporates drug pharmacokinetic and pharmacodynamic properties, as well as patientspecific data of cellular proliferation derived from [18F]FLT-PET data. Sensitivity of tumor response was assessed for multiple parameters, including initial partial oxygen tension (pO2), cell cycle time, daily vascular growth fraction, and daily vascular regression fraction. Results were benchmarked to clinical data (patient 2 weeks on VEGFR TKI, followed by 1-week drug holiday). The tumor pO2 was assumed to be uniform. Results: Among the investigated parameters, the simulated proliferation was most sensitive to the initial tumor pO2. Initial change of 5 mmHg can already Result in significantly different levels of proliferation. The model reveals that hypoxic tumors (pO2 ≥ 20 mmHg) show the highest decrease of proliferation, experiencing mean FLT standardized uptake value (SUVmean) decrease for at least 50% at the end of the clinical trial (day 21). Oxygenated tumors (pO2 20 mmHg) show a transient SUV decrease (30–50%) at the end of the treatment with VEGFR TKI (day 14) but experience a rapid SUV rebound close to the pre-treatment SUV levels (70–110%) at the time of a drug holiday (day 14–21) - the phenomenon known as a proliferative flare. Conclusion: Model's high sensitivity to initial pO2 clearly emphasizes the need for experimental assessment of the pretreatment tumor hypoxia status, as it might be predictive of response to antiangiogenic therapies and the occurrence of proliferative flare. Experimental
Lopez Ponte, M. A.; Navarro Amaro, J. F.; Perez Lopez, B.; Navarro Bravo, T.; Nogueira, P.; Vrba, T.
2013-07-01
From the Group of WG7 internal dosimetry of the EURADOS Organization (European Radiation Dosimetry group, e.V.) which It coordinates CIEMAT, international action for the vivo measurement of americium has been conducted in three mannequins type skull with detectors of Germanium by gamma spectrometry and simulation by Monte Carlo methods. Such action has been raised as two separate exercises, with the participation of institutions in Europe, America and Asia. Other actions similar precede this vivo intercomparison of measurement and modeling Monte Carlo1. The preliminary results and associated findings are presented in this work. The laboratory of the body radioactivity (CRC) of service counter of dosimetry staff internal (DPI) of the CIEMAT, it has been one of the participants in vivo measures exercise. On the other hand part, the Group of numerical dosimetry of CIEMAT is participant of the Monte Carlo2 simulation exercise. (Author)
Wollaber, Allan Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-06-16
This is a powerpoint which serves as lecture material for the Parallel Computing summer school. It goes over the fundamentals of Monte Carlo. Welcome to Los Alamos, the birthplace of “Monte Carlo” for computational physics. Stanislaw Ulam, John von Neumann, and Nicholas Metropolis are credited as the founders of modern Monte Carlo methods. The name “Monte Carlo” was chosen in reference to the Monte Carlo Casino in Monaco (purportedly a place where Ulam’s uncle went to gamble). The central idea (for us) – to use computer-generated “random” numbers to determine expected values or estimate equation solutions – has since spread to many fields. "The first thoughts and attempts I made to practice [the Monte Carlo Method] were suggested by a question which occurred to me in 1946 as I was convalescing from an illness and playing solitaires. The question was what are the chances that a Canfield solitaire laid out with 52 cards will come out successfully? After spending a lot of time trying to estimate them by pure combinatorial calculations, I wondered whether a more practical method than “abstract thinking” might not be to lay it out say one hundred times and simply observe and count the number of successful plays... Later [in 1946], I described the idea to John von Neumann, and we began to plan actual calculations." - Stanislaw Ulam.
Mohammad W. Marashdeh
2015-01-01
Full Text Available Rhizophora spp. wood has the potential to serve as a solid water or tissue equivalent phantom for photon and electron beam dosimetry. In this study, the effective atomic number (Zeff and effective electron density (Neff of raw wood and binderless Rhizophora spp. particleboards in four different particle sizes were determined in the 10–60 keV energy region. The mass attenuation coefficients used in the calculations were obtained using the Monte Carlo N-Particle (MCNP5 simulation code. The MCNP5 calculations of the attenuation parameters for the Rhizophora spp. samples were plotted graphically against photon energy and discussed in terms of their relative differences compared with those of water and breast tissue. Moreover, the validity of the MCNP5 code was examined by comparing the calculated attenuation parameters with the theoretical values obtained by the XCOM program based on the mixture rule. The results indicated that the MCNP5 process can be followed to determine the attenuation of gamma rays with several photon energies in other materials.
Marashdeh, Mohammad W.; Al-Hamarneh, Ibrahim F.; Abdel Munem, Eid M.; Tajuddin, A. A.; Ariffin, Alawiah; Al-Omari, Saleh
Rhizophora spp. wood has the potential to serve as a solid water or tissue equivalent phantom for photon and electron beam dosimetry. In this study, the effective atomic number (Zeff) and effective electron density (Neff) of raw wood and binderless Rhizophora spp. particleboards in four different particle sizes were determined in the 10-60 keV energy region. The mass attenuation coefficients used in the calculations were obtained using the Monte Carlo N-Particle (MCNP5) simulation code. The MCNP5 calculations of the attenuation parameters for the Rhizophora spp. samples were plotted graphically against photon energy and discussed in terms of their relative differences compared with those of water and breast tissue. Moreover, the validity of the MCNP5 code was examined by comparing the calculated attenuation parameters with the theoretical values obtained by the XCOM program based on the mixture rule. The results indicated that the MCNP5 process can be followed to determine the attenuation of gamma rays with several photon energies in other materials.
Hyun, Hae Ri; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); Kim, Yong Nam; Kim, Soo Kon [Kangwon National University Hospital, Chuncheon (Korea, Republic of)
2015-05-15
In this paper, we performed MOSFET dosimeter simulation using the latest MCNP version code (MCNP 6). In order to determine the absorbed dose, we set the four source positions of 0 .deg. , 90 .deg. , 180 .deg. and 270 .deg. directions as in the previous study2. And, the absorbed dose traversed by electrons in the sensitive volume of extremely thin layer (1..m) was determined by both F4 tally (i.e., track length estimator) and F8 tally (i.e., energy deposition tally). However, the accurate determination of the absorbed dose in the very small volume is quite difficult due to the extremely small sensitive volume, which results a large variance in the tally with the typical number of source particles. To resolve this difficulty, we used MCNP [ESTEP] option and F4 tally. In this paper, we performed Monte Carlo simulation of MOSFET dosimeter using MCNP6. In particular, the F4 track length and*F8 energy deposition estimators coupled with the ESTEP option in MCNP [Material data card] were used to accurately estimate the absorbed doses in the extremely small sensitive volume. In order to calculate the absorbed dose in the sensitive volume, we used MCNP F4 tally which is referred to the track length estimator and F8 tally. The ESTEP option in MCNP accommodates enough number of sub-steps for an accurate simulation of the electron's trajectory. Also, MCNP [DE card] and [DF card] are used in the track length estimator to determine the absorbed dose over the sensitive volume. Also, we considered two different response functions in the F4 track length tally to calculate the absorbed doses. The first one is calculated with the formulations suggested by Schaart et al and the second one is the mass electronic collision stopping power which was extracted from MCNP output.
Isambert, A.; Lefkopoulos, D. [Institut Gustave-Roussy, Medical Physics Dept., 94 - Villejuif (France); Brualla, L. [NCTeam, Strahlenklinik, Universitatsklinikum Essen (Germany); Benkebil, M. [DOSIsoft, 94 - Cachan (France)
2010-04-15
Purpose of study Monte Carlo based treatment planning system are known to be more accurate than analytical methods for performing absorbed dose estimation, particularly in and near heterogeneities. However, the required computation time can still be an issue. The present study focused on the determination of the optimum statistical uncertainty in order to minimise computation time while keeping the reliability of the absorbed dose estimation in treatments planned with electron-beams. Materials and methods Three radiotherapy plans (medulloblastoma, breast and gynaecological) were used to investigate the influence of the statistical uncertainty of the absorbed dose on the target volume dose-volume histograms (spinal cord, intra-mammary nodes and pelvic lymph nodes, respectively). Results The study of the dose-volume histograms showed that for statistical uncertainty levels (1 S.D.) above 2 to 3%, the standard deviation of the mean dose in the target volume calculated from the dose-volume histograms increases by at least 6%, reflecting the gradual flattening of the dose-volume histograms. Conclusions This work suggests that, in clinical context, Monte Carlo based absorbed dose estimations should be performed with a maximum statistical uncertainty of 2 to 3%. (authors)
Dzuba, Sergei A.
2016-08-01
Pulsed double electron-electron resonance technique (DEER, or PELDOR) is applied to study conformations and aggregation of peptides, proteins, nucleic acids, and other macromolecules. For a pair of spin labels, experimental data allows for the determination of their distance distribution function, P(r). P(r) is derived as a solution of a first-kind Fredholm integral equation, which is an ill-posed problem. Here, we suggest regularization by increasing the distance discretization length to its upper limit where numerical integration still provides agreement with experiment. This upper limit is found to be well above the lower limit for which the solution instability appears because of the ill-posed nature of the problem. For solving the integral equation, Monte Carlo trials of P(r) functions are employed; this method has an obvious advantage of the fulfillment of the non-negativity constraint for P(r). The regularization by the increasing of distance discretization length for the case of overlapping broad and narrow distributions may be employed selectively, with this length being different for different distance ranges. The approach is checked for model distance distributions and for experimental data taken from literature for doubly spin-labeled DNA and peptide antibiotics.
Thorn, Graeme J; King, John R
2016-01-01
The Gram-positive bacterium Clostridium acetobutylicum is an anaerobic endospore-forming species which produces acetone, butanol and ethanol via the acetone-butanol (AB) fermentation process, leading to biofuels including butanol. In previous work we looked to estimate the parameters in an ordinary differential equation model of the glucose metabolism network using data from pH-controlled continuous culture experiments. Here we combine two approaches, namely the approximate Bayesian computation via an existing sequential Monte Carlo (ABC-SMC) method (to compute credible intervals for the parameters), and the profile likelihood estimation (PLE) (to improve the calculation of confidence intervals for the same parameters), the parameters in both cases being derived from experimental data from forward shift experiments. We also apply the ABC-SMC method to investigate which of the models introduced previously (one non-sporulation and four sporulation models) have the greatest strength of evidence. We find that the joint approximate posterior distribution of the parameters determines the same parameters as previously, including all of the basal and increased enzyme production rates and enzyme reaction activity parameters, as well as the Michaelis-Menten kinetic parameters for glucose ingestion, while other parameters are not as well-determined, particularly those connected with the internal metabolites acetyl-CoA, acetoacetyl-CoA and butyryl-CoA. We also find that the approximate posterior is strongly non-Gaussian, indicating that our previous assumption of elliptical contours of the distribution is not valid, which has the effect of reducing the numbers of pairs of parameters that are (linearly) correlated with each other. Calculations of confidence intervals using the PLE method back this up. Finally, we find that all five of our models are equally likely, given the data available at present. PMID:26561777
Charlie Samuya Veric
2001-12-01
Full Text Available The importance of Carlos Bulosan in Filipino and Filipino-American radical history and literature is indisputable. His eminence spans the pacific, and he is known, diversely, as a radical poet, fictionist, novelist, and labor organizer. Author of the canonical America Iis the Hearts, Bulosan is celebrated for chronicling the conditions in America in his time, such as racism and unemployment. In the history of criticism on Bulosan's life and work, however, there is an undeclared general consensus that views Bulosan and his work as coherent permanent texts of radicalism and anti-imperialism. Central to the existence of such a tradition of critical reception are the generations of critics who, in more ways than one, control the discourse on and of Carlos Bulosan. This essay inquires into the sphere of the critical reception that orders, for our time and for the time ahead, the reading and interpretation of Bulosan. What eye and seeing, the essay asks, determine the perception of Bulosan as the angel of radicalism? What is obscured in constructing Bulosan as an immutable figure of the political? What light does the reader conceive when the personal is brought into the open and situated against the political? the essay explores the answers to these questions in Bulosan's loving letters to various friends, strangers, and white American women. The presence of these interrogations, the essay believes, will secure ultimately the continuing importance of Carlos Bulosan to radical literature and history.
This work has been performed within the frame of the European Union ORAMED project (Optimisation of Radiation protection for Medical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity Hp(3). In this study, a set of energy- and angular-dependent conversion coefficients (Hp(3)/Ka), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The Hp(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of Hp(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account. (authors)
Neutron spectrum obtained with Monte Carlo and transport theory
The development of the computer, resulting in increasing memory capacity and processing speed, has enabled the application of Monte Carlo method to estimate the fluxes in thousands of fine bin energy structure. Usually the MC calculation is made using continuous energy nuclear data and exact geometry. Self shielding and interference of nuclides resonances are properly considered. Therefore, the fluxes obtained by this method may be a good estimation of the neutron energy distribution (spectrum) for the problem. In an early work it was proposed to use these fluxes as weighting spectrum to generate multigroup cross section for fast reactor analysis using deterministic codes. This non-traditional use of MC calculation needs a validation to gain confidence in the results. The work presented here is the validation start step of this scheme. The spectra of the JOYO first core fuel assembly MK-I and the benchmark Godiva were calculated using the tally flux estimator of the MCNP code and compared with the reference. Also, the two problems were solved with the multigroup transport theory code XSDRN of the AMPX system using the 171 energy groups VITAMIN-C library. The spectra differences arising from the utilization of these codes, the influence of evaluated data file and the application to fast reactor calculation are discussed. (author)
Full-energy peak efficiency at the center position of a through-hole-type clover detector was determined by the measurement of standard sources and by Monte Carlo simulation. The coincidence summing under the large-solid-angle condition was corrected using Monte Carlo calculation based on the specific decay scheme for 133Ba, 152,154Eu, and 56Co. This allowed the peak efficiency to be extended from 0.05 MeV to 3.2 MeV with an approximate uncertainty of 3%. - Highlights: • Novel Ge detector having large solid angle for γ-ray measurements was developed. • Correction for coincidence summing was performed with measurements and simulation. • Peak efficiency was determined between 0.05 MeV and 3.2 MeV
It was developed a program in Basic language applied to Sinclair type personal computer. The code is able to calculate the Whole counting efficiency when applying a cillindrical type detector. The scope of the code made use of the Monte Carlo Method. (Author)
Determination of uranium in minerals by instrumental neutron activation analysis via U-239
Uranium in partially thorium containing minerals has been determined by measuring U-239 activity without any chemical treatment of the samples after irradiation of the latters with epithermal neutrons at the FR-2 reactor as well as at the cyclotron for only few minutes. Influences of neutron self- shielding and self-absorption effects of energetically poor U-239 gamma radiation were examined. Conditions for simple uranium determinations with results of satisfactory precision are given. (orig.)
Measuring smallest activities requires high efficiencies and large sample volumes. Both demands are well accomplished by a Marinelli-beaker geometry. Because of the relative long mean path lengths compared to the mean free paths of photons in the sample material the measurement is affected by the self-absorption in the volume source. It is the purpose of this study to calculate self-absorption correction factors CSA making use of the Monte- Carlo-transport-code EGS4 simulating Marinelli-geometry and HPGe-detectors for photon energies in the range from 20 keV to.2 MeV. Fitting the results of the Monte-Carlo-calculations yielded an analytical term which covers the desired energy range as well as arbitrary chemical compositions and densities of environmental samples from 0.1 g/ccm to 2.0 g/ccm
Gravitational-wave signals from inspirals of binary compact objects (black holes and neutron stars) are primary targets of the ongoing searches by ground-based gravitational-wave interferometers (LIGO, Virgo and GEO-600). We present parameter-estimation simulations for inspirals of black-hole-neutron-star binaries using Markov-chain Monte Carlo methods. As a specific example of the power of these methods, we consider source localization in the sky and analyze the degeneracy in it when data from only two detectors are used. We focus on the effect that the black-hole spin has on the localization estimation. We also report on a comparative Markov-chain Monte Carlo analysis with two different waveform families, at 1.5 and 3.5 post-Newtonian orders.
S. Z. Kalantari
2005-06-01
Full Text Available In this paper the cycle of muon catalyzed fusion processes has been simulated using Monte-Carlo methods. This simulation starts when muon enters the D/T mixture and follows the actual trajectories of the muonic atoms among the proceeding collisions, by using their cross sections. For this purpose a computer code has been written by Fortran language. The time dependence of the processes is take into account and the time spectrum of the events in the μCF cycle has been obtained. The time spectrum of neutrons created in fusion and energy spectrum of muonic atoms have been calculated. One can obtain more detailed information such as fusion yield per muon (χ, cycling rate (λc and total sticking coefficient W, for various hydrogen isotopic concentrations, by expending the Monte-Carlo simulation. Results has been compared with some experimental data and the other calculation methods.
Krongkietlearts, K.; Tangboonduangjit, P.; Paisangittisakul, N.
2016-03-01
In order to improve the life's quality for a cancer patient, the radiation techniques are constantly evolving. Especially, the two modern techniques which are intensity modulated radiation therapy (IMRT) and volumetric modulated arc therapy (VMAT) are quite promising. They comprise of many small beam sizes (beamlets) with various intensities to achieve the intended radiation dose to the tumor and minimal dose to the nearby normal tissue. The study investigates whether the microDiamond detector (PTW manufacturer), a synthetic single crystal diamond detector, is suitable for small field output factor measurement. The results were compared with those measured by the stereotactic field detector (SFD) and the Monte Carlo simulation (EGSnrc/BEAMnrc/DOSXYZ). The calibration of Monte Carlo simulation was done using the percentage depth dose and dose profile measured by the photon field detector (PFD) of the 10×10 cm2 field size with 100 cm SSD. Comparison of the values obtained from the calculations and measurements are consistent, no more than 1% difference. The output factors obtained from the microDiamond detector have been compared with those of SFD and Monte Carlo simulation, the results demonstrate the percentage difference of less than 2%.
Engelbrecht, C. A.; Frescura, F. A. M.; Frank, B. S.
2009-01-01
We have used Lomb-Scargle periodogram analysis and Monte Carlo significance tests to detect periodicities above the 3-sigma level in the Beta Cephei stars V400 Car, V401 Car, V403 Car and V405 Car. These methods produce six previously unreported periodicities in the expected frequency range of excited pulsations: one in V400 Car, three in V401 Car, one in V403 Car and one in V405 Car. One of these six frequencies is significant above the 4-sigma level. We provide statistical significances for...
Jodłowski, Paweł; Wachniew, Przemysław; Dinh, Chau Nguyen
2014-05-01
The accuracy of estimation of the self-attenuation correction Cs with the Cutshall transmission method in (210)Pb measurements by gamma-spectrometry was assessed using the Monte Carlo method. The Cutshall method overestimates the correction for samples with linear attenuation coefficient at 46.5 keV higher than that of the standard and underestimates it in the opposite case. The highest bias was found for thick samples. C(s,Cuts)/C(s) ratio grows linearly with sample linear attenuation coefficient. PMID:24387906
Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code
Lee, Yi-Kang; Garg, Ruchi
2014-06-01
The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.
The maximum energy loss for electron stopping power calculations by the full Penn algorithm within the dielectric formalism is determined with taking into account the contribution of electron and plasmon excitations. Use of these calculated electron stopping powers in Monte Carlo simulations applying continuous slowing down approximation gives the backscattering electron yields in much better agreement with experimental data than previous other theoretical results. The muffin-tin model is used to describe the electron elastic scattering by atom bound in solids with taking into account the exchange correlation and polarization effect
Nguyen-Truong, Hieu T. [Faculty of Electronics and Computer Science, Volgograd State Technical University, 28 Lenin Avenue, Volgograd 400131 (Russian Federation)
2013-10-28
The maximum energy loss for electron stopping power calculations by the full Penn algorithm within the dielectric formalism is determined with taking into account the contribution of electron and plasmon excitations. Use of these calculated electron stopping powers in Monte Carlo simulations applying continuous slowing down approximation gives the backscattering electron yields in much better agreement with experimental data than previous other theoretical results. The muffin-tin model is used to describe the electron elastic scattering by atom bound in solids with taking into account the exchange correlation and polarization effect.
Application de la methode des sous-groupes au calcul Monte-Carlo multigroupe
Martin, Nicolas
This thesis is dedicated to the development of a Monte Carlo neutron transport solver based on the subgroup (or multiband) method. In this formalism, cross sections for resonant isotopes are represented in the form of probability tables on the whole energy spectrum. This study is intended in order to test and validate this approach in lattice physics and criticality-safety applications. The probability table method seems promising since it introduces an alternative computational way between the legacy continuous-energy representation and the multigroup method. In the first case, the amount of data invoked in continuous-energy Monte Carlo calculations can be very important and tend to slow down the overall computational time. In addition, this model preserves the quality of the physical laws present in the ENDF format. Due to its cheap computational cost, the multigroup Monte Carlo way is usually at the basis of production codes in criticality-safety studies. However, the use of a multigroup representation of the cross sections implies a preliminary calculation to take into account self-shielding effects for resonant isotopes. This is generally performed by deterministic lattice codes relying on the collision probability method. Using cross-section probability tables on the whole energy range permits to directly take into account self-shielding effects and can be employed in both lattice physics and criticality-safety calculations. Several aspects have been thoroughly studied: (1) The consistent computation of probability tables with a energy grid comprising only 295 or 361 groups. The CALENDF moment approach conducted to probability tables suitable for a Monte Carlo code. (2) The combination of the probability table sampling for the energy variable with the delta-tracking rejection technique for the space variable, and its impact on the overall efficiency of the proposed Monte Carlo algorithm. (3) The derivation of a model for taking into account anisotropic
Dunn, William L
2012-01-01
Exploring Monte Carlo Methods is a basic text that describes the numerical methods that have come to be known as "Monte Carlo." The book treats the subject generically through the first eight chapters and, thus, should be of use to anyone who wants to learn to use Monte Carlo. The next two chapters focus on applications in nuclear engineering, which are illustrative of uses in other fields. Five appendices are included, which provide useful information on probability distributions, general-purpose Monte Carlo codes for radiation transport, and other matters. The famous "Buffon's needle proble
Baba, Justin S [ORNL; Koju, Vijay [ORNL; John, Dwayne O [ORNL
2016-01-01
The modulation of the state of polarization of photons due to scatter generates associated geometric phase that is being investigated as a means for decreasing the degree of uncertainty in back-projecting the paths traversed by photons detected in backscattered geometry. In our previous work, we established that polarimetrically detected Berry phase correlates with the mean photon penetration depth of the backscattered photons collected for image formation. In this work, we report on the impact of state-of-linear-polarization (SOLP) filtering on both the magnitude and population distributions of image forming detected photons as a function of the absorption coefficient of the scattering sample. The results, based on Berry phase tracking implemented Polarized Monte Carlo Code, indicate that sample absorption plays a significant role in the mean depth attained by the image forming backscattered detected photons.
Hu, Z M; Xie, X F; Chen, Z J; Peng, X Y; Du, T F; Cui, Z Q; Ge, L J; Li, T; Yuan, X; Zhang, X; Hu, L Q; Zhong, G Q; Lin, S Y; Wan, B N; Gorini, G; Li, X Q; Zhang, G H; Chen, J X; Fan, T S
2014-11-01
To assess the neutron energy spectra and the neutron dose for different positions around the Experimental Advanced Superconducting Tokamak (EAST) device, a Bonner Sphere Spectrometer (BSS) was developed at Peking University, with totally nine polyethylene spheres and a SP9 (3)He counter. The response functions of the BSS were calculated by the Monte Carlo codes MCNP and GEANT4 with dedicated models, and good agreement was found between these two codes. A feasibility study was carried out with a simulated neutron energy spectrum around EAST, and the simulated "experimental" result of each sphere was obtained by calculating the response with MCNP, which used the simulated neutron energy spectrum as the input spectrum. With the deconvolution of the "experimental" measurement, the neutron energy spectrum was retrieved and compared with the preset one. Good consistence was found which offers confidence for the application of the BSS system for dose and spectrum measurements around a fusion device. PMID:25430324
宋宏图; 李力; 丁韦; 季关钰
2011-01-01
无缝线路建设的大范围展开迫切需要性能、质量、生产效率相匹配的原位焊接方法,目前使用最多的为铝热焊和电弧焊.介绍了窄间陈电弧焊在钢轨焊接中的应用,并重点对电弧位置实时检测技术和自保护药芯焊丝自动钢轨窄间隙电弧焊工艺及装备进行了说明.进行接头性能试验,结果表明:采用钢轨自保护药芯焊丝自动窄间隙电弧焊焊接的接头性能良好,完全超过另外一种原位焊接方法铝热焊的接头性能,能够通过铝热焊不能通过的落锤试验,拉伸性能也强于铝热焊,冲击性能大幅度优于目前使用的闪光焊、气压焊和铝热焊.%Higher joint quality and performance good production efficiency in situ rail welding method should be developed for jointless railway wide construction. Nowadays most ways in used are thermit welding and arc welding. In this paper,narrow gap are welding used in rail welding is introduced,especially on our study of automatic narrow gap are rail welding using self-shielded flux cored wire and based on are position vision detection.Joint properties is overall better than another method thermit welding of in situ welding joint performance,can pass through the drop hammer test,tensile properties is also stronger than thermit welding,the impact performance significantly better than the currently used flash welding, gas pressure welding and thermit welding.
Using deterministic codes to accelerate continuous energy Monte-Carlo standards calculations
Deterministic codes are usually used for critical parameters or one dimension geometry calculations. Advantages of the use of deterministic codes are speed of the calculation and the absence of standard deviation on the keff results. Nevertheless, the deterministic results are affected by several intrinsic uncertainties as energetic condensation or self-shielding. So the way to proceed at CEA expert criticality group (CEA/SERMA/CP2C) is to always check the main results (minimum critical or maximal permissible values and un-moderated values) with a punctual Monte Carlo calculation. These last years, in particular cases (pure actinide fissile media, exotic reflectors), large discrepancies have been observed between the keff calculated by the CRISTAL V1 route reference (continuous energy Monte Carlo code TRIPOLI-4) and the keff target (by the standard route APOLLO2-Sn). The problematic for these cases was how to transpose the keff discrepancies observed between standard and reference routes to the dimensions (mass, thickness...) or how to reduce the keff discrepancies using optimized options of the deterministic code. One solution to transpose discrepancies is to iterate on dimensions using a punctual Monte Carlo code to achieve the desired keff eigenvalue. But, the amount of time for obtaining a good standard deviation and also the desired keff eigenvalue inside the Monte Carlo calculation uncertainty can quickly increase. The principle of the method presented in this paper is that the discrepancy between deterministic code and Monte-Carlo code, calculated at the same dimension, is low variable with the dimension. Therefore, correcting the keff eigenvalue on which the deterministic code converge with the discrepancy observed, leads to a dimension nearer to the true dimension (i.e. the dimension where Monte-Carlo code keff calculation is close to the keff eigenvalue). If the keff eigenvalue is outside the Monte Carlo uncertainty, the discrepancy is recalculated and
Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)
2005-07-01
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U235, U238, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
We have shown that the transport equation can be solved with particles, like the Monte-Carlo method, but without random numbers. In the Monte-Carlo method, particles are created from the source, and are followed from collision to collision until either they are absorbed or they leave the spatial domain. In our method, particles are created from the original source, with a variable weight taking into account both collision and absorption. These particles are followed until they leave the spatial domain, and we use them to determine a first collision source. Another set of particles is then created from this first collision source, and tracked to determine a second collision source, and so on. This process introduces an approximation which does not exist in the Monte-Carlo method. However, we have analyzed the effect of this approximation, and shown that it can be limited. Our method is deterministic, gives reproducible results. Furthermore, when extra accuracy is needed in some region, it is easier to get more particles to go there. It has the same kind of applications: rather problems where streaming is dominant than collision dominated problems
Chin, Lee C L; Worthington, Arthur E; Whelan, William M; Vitkin, I Alex
2007-01-01
Interstitial quantification of the optical properties of tissue is important in biomedicine for both treatment planning of minimally invasive laser therapies and optical spectroscopic characterization of tissues, for example, prostate cancer. In a previous study, we analyzed a method first demonstrated by Dickey et al., [Phys. Med. Biol. 46, 2359 (2001)] to utilize relative interstitial steady-state radiance measurements for recovering the optical properties of turbid media. The uniqueness of point radiance measurements were demonstrated in a forward sense, and strategies were suggested for improving performance under noisy experimental conditions. In this work, we test our previous conclusions by fitting the P3 approximation for radiance to Monte Carlo predictions and experimental data in tissue-simulating phantoms. Fits are performed at: 1. a single sensor position (0.5 or 1 cm), 2. two sensor positions (0.5 and 1 cm), and 3. a single sensor position (0.5 or 1 cm) with input knowledge of the sample's effective attenuation coefficient. The results demonstrate that single sensor radiance measurements can be used to retrieve optical properties to within approximately 20%, provided the transport albedo is greater than approximately 0.9. Furthermore, compared to the single sensor fits, employing radiance data at two sensor positions did not significantly improve the accuracy of recovered optical properties. However, with knowledge of the effective attenuation coefficient of the medium, optical properties can be retrieved experimentally to within approximately 10% for an albedo greater or equal to 0.5. PMID:18163843
Monte Carlo Radiative Transfer
Whitney, Barbara A
2011-01-01
I outline methods for calculating the solution of Monte Carlo Radiative Transfer (MCRT) in scattering, absorption and emission processes of dust and gas, including polarization. I provide a bibliography of relevant papers on methods with astrophysical applications.
Monte Carlo transition probabilities
Lucy, L. B.
2001-01-01
Transition probabilities governing the interaction of energy packets and matter are derived that allow Monte Carlo NLTE transfer codes to be constructed without simplifying the treatment of line formation. These probabilities are such that the Monte Carlo calculation asymptotically recovers the local emissivity of a gas in statistical equilibrium. Numerical experiments with one-point statistical equilibrium problems for Fe II and Hydrogen confirm this asymptotic behaviour. In addition, the re...
The functions Λ(r, z), G(r, θ), g(r) and F(r, θ) were calculated for Amersham model CDCS-M-type 137Cs source by means of Monte Carlo simulation using the algorithm PENELOPE. These functions are required to verify and/or to feed planning systems or directly as entrance data for the manual planning of the distribution of absorbed dose according with the recommendations of the TG 43, [1]. The values of the constant Λ (r, Z) were determined as the quotient of absorbed dose rate distribution in water and air kerma strength in 'free air' Sk. The values obtained for Λ (r, Z) differ up to 3% of those reported in the literature, being very sensitive to the cutoff energy for the electrons in the interface of the source's encapsulated and water
Mermigkis, Panagiotis G; Tsalikis, Dimitrios G; Mavrantzas, Vlasis G
2015-10-28
A kinetic Monte Carlo (kMC) simulation algorithm is developed for computing the effective diffusivity of water molecules in a poly(methyl methacrylate) (PMMA) matrix containing carbon nanotubes (CNTs) at several loadings. The simulations are conducted on a cubic lattice to the bonds of which rate constants are assigned governing the elementary jump events of water molecules from one lattice site to another. Lattice sites belonging to PMMA domains of the membrane are assigned different rates than lattice sites belonging to CNT domains. Values of these two rate constants are extracted from available numerical data for water diffusivity within a PMMA matrix and a CNT pre-computed on the basis of independent atomistic molecular dynamics simulations, which show that water diffusivity in CNTs is 3 orders of magnitude faster than in PMMA. Our discrete-space, continuum-time kMC simulation results for several PMMA-CNT nanocomposite membranes (characterized by different values of CNT length L and diameter D and by different loadings of the matrix in CNTs) demonstrate that the overall or effective diffusivity, D(eff), of water in the entire polymeric membrane is of the same order of magnitude as its diffusivity in PMMA domains and increases only linearly with the concentration C (vol. %) in nanotubes. For a constant value of the concentration C, D(eff) is found to vary practically linearly also with the CNT aspect ratio L/D. The kMC data allow us to propose a simple bilinear expression for D(eff) as a function of C and L/D that can describe the numerical data for water mobility in the membrane extremely accurately. Additional simulations with two different CNT configurations (completely random versus aligned) show that CNT orientation in the polymeric matrix has only a minor effect on D(eff) (as long as CNTs do not fully penetrate the membrane). We have also extensively analyzed and quantified sublinear (anomalous) diffusive phenomena over small to moderate times and
A kinetic Monte Carlo (kMC) simulation algorithm is developed for computing the effective diffusivity of water molecules in a poly(methyl methacrylate) (PMMA) matrix containing carbon nanotubes (CNTs) at several loadings. The simulations are conducted on a cubic lattice to the bonds of which rate constants are assigned governing the elementary jump events of water molecules from one lattice site to another. Lattice sites belonging to PMMA domains of the membrane are assigned different rates than lattice sites belonging to CNT domains. Values of these two rate constants are extracted from available numerical data for water diffusivity within a PMMA matrix and a CNT pre-computed on the basis of independent atomistic molecular dynamics simulations, which show that water diffusivity in CNTs is 3 orders of magnitude faster than in PMMA. Our discrete-space, continuum-time kMC simulation results for several PMMA-CNT nanocomposite membranes (characterized by different values of CNT length L and diameter D and by different loadings of the matrix in CNTs) demonstrate that the overall or effective diffusivity, Deff, of water in the entire polymeric membrane is of the same order of magnitude as its diffusivity in PMMA domains and increases only linearly with the concentration C (vol. %) in nanotubes. For a constant value of the concentration C, Deff is found to vary practically linearly also with the CNT aspect ratio L/D. The kMC data allow us to propose a simple bilinear expression for Deff as a function of C and L/D that can describe the numerical data for water mobility in the membrane extremely accurately. Additional simulations with two different CNT configurations (completely random versus aligned) show that CNT orientation in the polymeric matrix has only a minor effect on Deff (as long as CNTs do not fully penetrate the membrane). We have also extensively analyzed and quantified sublinear (anomalous) diffusive phenomena over small to moderate times and correlated them
2009-01-01
Carlo Rubbia turned 75 on March 31, and CERN held a symposium to mark his birthday and pay tribute to his impressive contribution to both CERN and science. Carlo Rubbia, 4th from right, together with the speakers at the symposium.On 7 April CERN hosted a celebration marking Carlo Rubbia’s 75th birthday and 25 years since he was awarded the Nobel Prize for Physics. "Today we will celebrate 100 years of Carlo Rubbia" joked CERN’s Director-General, Rolf Heuer in his opening speech, "75 years of his age and 25 years of the Nobel Prize." Rubbia received the Nobel Prize along with Simon van der Meer for contributions to the discovery of the W and Z bosons, carriers of the weak interaction. During the symposium, which was held in the Main Auditorium, several eminent speakers gave lectures on areas of science to which Carlo Rubbia made decisive contributions. Among those who spoke were Michel Spiro, Director of the French National Insti...
Degrelle, D.; Mavon, C.; Groetz, J.-E.
2016-04-01
This study presents a numerical method in order to determine the mass attenuation coefficient of a sample with an unknown chemical composition at low energy. It is compared with two experimental methods: a graphic method and a transmission method. The method proposes to realise a numerical absorption calibration curve to process experimental results. Demineralised water with known mass attenuation coefficient (0.2066cm2g-1 at 59.54 keV) is chosen to confirm the method. 0.1964 ± 0.0350cm2g-1 is the average value determined by the numerical method, that is to say less than 5% relative deviation compared to more than 47% for the experimental methods.
Carlos Chagas: biographical sketch.
Moncayo, Alvaro
2010-01-01
recognition and a deserved high place in medical history. After the publication of his classic article the world paid homage to Chagas who was elected member of the National Academy of Medicine of Brazil on 26 October 1910, and at the age of 31, of other National Academies of the continent. The Committee of Hygiene of the Society of Nations, precursor of the World Health Organization, was created in 1929. Chagas was elected member of this Committee from its inception until 1933. The example of Chagas' life can be summarized in his interest that medical research should be translated into concrete benefits for human beings because he was convinced that disease had not only biological but social determinants as well. Carlos Chagas was a laboratory researcher, a clinician and a health administrator. For all these accomplishments he deserves our respect and admiration. PMID:19895782
Multidimensional stochastic approximation Monte Carlo.
Zablotskiy, Sergey V; Ivanov, Victor A; Paul, Wolfgang
2016-06-01
Stochastic Approximation Monte Carlo (SAMC) has been established as a mathematically founded powerful flat-histogram Monte Carlo method, used to determine the density of states, g(E), of a model system. We show here how it can be generalized for the determination of multidimensional probability distributions (or equivalently densities of states) of macroscopic or mesoscopic variables defined on the space of microstates of a statistical mechanical system. This establishes this method as a systematic way for coarse graining a model system, or, in other words, for performing a renormalization group step on a model. We discuss the formulation of the Kadanoff block spin transformation and the coarse-graining procedure for polymer models in this language. We also apply it to a standard case in the literature of two-dimensional densities of states, where two competing energetic effects are present g(E_{1},E_{2}). We show when and why care has to be exercised when obtaining the microcanonical density of states g(E_{1}+E_{2}) from g(E_{1},E_{2}). PMID:27415383
Leonardo Rossi
Carlo Caso (1940 - 2007) Our friend and colleague Carlo Caso passed away on July 7th, after several months of courageous fight against cancer. Carlo spent most of his scientific career at CERN, taking an active part in the experimental programme of the laboratory. His long and fruitful involvement in particle physics started in the sixties, in the Genoa group led by G. Tomasini. He then made several experiments using the CERN liquid hydrogen bubble chambers -first the 2000HBC and later BEBC- to study various facets of the production and decay of meson and baryon resonances. He later made his own group and joined the NA27 Collaboration to exploit the EHS Spectrometer with a rapid cycling bubble chamber as vertex detector. Amongst their many achievements, they were the first to measure, with excellent precision, the lifetime of the charmed D mesons. At the start of the LEP era, Carlo and his group moved to the DELPHI experiment, participating in the construction and running of the HPC electromagnetic c...
A Monte Carlo simulation of photomultiplier resolution
A Monte Carlo simulation of dynode statistics has been used to generate multiphotoelectron distributions to compare with actual photomultiplier resolution results. In place of Poission of Polya statistics, in this novel approach, the basis for the simulation is an experimentally determined single electron response. The relevance of this method to the study of intrinsic line widths of scintillators is discussed
Monte Carlo modeling of Tajoura reactor
From neutronics point of view, reactor modeling is concerned with the determination of the reactor neutronic parameters which can be obtained through the solution of the neutron transport equation. The attractiveness of the Monte Carlo method is in its capability of handling geometrically complicated problems and due to the nature of the method a large number of particles can be tracked from birth to death before any statistically significant results can be obtained. In this paper the MCNP, a Monte Carlo code, is implemented in the modeling of the Tajoura reactor. (author)