Burnup calculation methodology in the serpent 2 Monte Carlo code
International Nuclear Information System (INIS)
This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)
Development of 3d reactor burnup code based on Monte Carlo method and exponential Euler method
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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)
Burnup calculation capability in the PSG2 / Serpent Monte Carlo reactor physics code
International Nuclear Information System (INIS)
The PSG continuous-energy Monte Carlo reactor physics code has been developed at VTT Technical Research Centre of Finland since 2004. The code is mainly intended for group constant generation for coupled reactor simulator calculations and other tasks traditionally handled using deterministic lattices physics codes. The name was recently changed from acronym PSG to 'Serpent', and the capabilities have been extended by implementing built-in burnup calculation routines that enable the code to be used for fuel cycle studies and the modelling of irradiated fuels. This paper presents the methodology used for burnup calculation. Serpent has two fundamentally different options for solving the Bateman depletion equations: 1) the Transmutation Trajectory Analysis method (TTA), based on the analytical solution of linearized depletion chains and 2) the Chebyshev Rational Approximation Method (CRAM), an advanced matrix exponential solution developed at VTT. The first validation results are compared to deterministic CASMO-4E calculations. It is also shown that the overall running time in Monte Carlo burnup calculation can be significantly reduced using specialized calculation techniques, and that the continuous-energy Monte Carlo method is becoming a viable alternative to deterministic assembly burnup codes. (authors)
Burnup calculations of TR-2 Research Reactor with Monteburns Monte Carlo Code
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Full text: In this study, some neutronic calculations of first and second core cycles of 5 MW pool type TR-2 Research Reactor have been performed using Multi-Step Monte Carlo Burnup Code System MONTEBURNS and the results were compared with the values of experiments and other codes. Time dependent keff distribution and burnup ratios belong to first and second core cycles of TR-2 Research Reactor were compared and quite good consistence in the results were observed. After modeling the first and second core cycles of TR-2 with MCNP5 Monte Carlo code, MCNP5 used in MONTEBURNS code has been parallelized in 8 HP ProLiant BL680C G5 systems with 4 quad-core Intel Xeon E7330 CPU, utilizing the MPI parallel protocol and simulations were performed on the 128 cores Linux parallel computing machine system. The computation time was reduced by parallelization of MONTEBURNS which uses MCNP in many steps. (authors)
Progress on burnup calculation methods coupling Monte Carlo and depletion codes
Energy Technology Data Exchange (ETDEWEB)
Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar
2005-07-01
Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)
ALEPH 1.1.2: A Monte Carlo burn-up code
International Nuclear Information System (INIS)
In the last 40 years, Monte Carlo particle transport has been applied to a multitude of problems such as shielding and medical applications, to various types of nuclear reactors, . . . The success of the Monte Carlo method is mainly based on its broad application area, on its ability to handle nuclear data not only in its most basic but also most complex form (namely continuous energy cross sections, complex interaction laws, detailed energy-angle correlations, multi-particle physics, . . . ), on its capability of modeling geometries from simple 1D to complex 3D, . . . There is also a current trend in Monte Carlo applications toward high detail 3D calculations (for instance voxel-based medical applications), something for which deterministic codes are neither suited nor performant as to computational time and precision. Apart from all these fields where Monte Carlo particle transport has been applied successfully, there is at least one area where Monte Carlo has had limited success, namely burn-up and activation calculations where the time parameter is added to the problem. The concept of Monte Carlo burn-up consists of coupling a Monte Carlo code to a burn-up module to improve the accuracy of depletion and activation calculations. For every time step the Monte Carlo code will provide reaction rates to the burn-up module which will return new material compositions to the Monte Carlo code. So if static Monte Carlo particle transport is slow, then Monte Carlo particle transport with burn-up will be even slower as calculations have to be performed for every time step in the problem. The computational issues to perform accurate Monte Carlo calculations are however continuously reduced due to improvements made in the basic Monte Carlo algorithms, due to the development of variance reduction techniques and due to developments in computer architecture (more powerful processors, the so-called brute force approach through parallel processors and networked systems
Development of burnup calculation function in reactor Monte Carlo code RMC
International Nuclear Information System (INIS)
This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (authors)
Accuracy assessment of a new Monte Carlo based burnup computer code
International Nuclear Information System (INIS)
Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.
International Nuclear Information System (INIS)
In the design of the incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI), the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional burnup and fuel depletion analysis codes have been found to be inadequate for these calculations, largely because of the material and geometry modeled and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. More sophisticated codes such as the transport lattice type code WIMS often lack some materials, such as ZrH. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the fuel element is needed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations. This effort required the modification of MCNP itself to perform the additional task of accomplishing burnup calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up
International Nuclear Information System (INIS)
There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)
Development and validation of Monte-Carlo burnup calculation code MCNTRANS
International Nuclear Information System (INIS)
A new nuclear fuel burnup calculation code MCNTRANS based on MCNP was introduced in this paper. The neutronics calculation parameter was extracted from the MCNP5 reaction rate tally result, while a graph theory algorithm was implemented to track the burnup chain and the analytic solution of the Bateman equation was given. At the same time, the detailed physical process was considered to improve the accuracy and serviceability of this code, and prediction-correction method was used to allow a large burnup step. The OECD/NEA and JAERI pin cell benchmark problems were used to validate the code MCNTRANS while a reference result was given by other code. It can be concluded that the calculation results of MCNTRANS are generally consistent with the experimental result and that of the other burnup codes, and part of the actinides and fission products calculation result show better accuracy. (authors)
Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly
International Nuclear Information System (INIS)
The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.
Using SERPENT Monte Carlo and Burnup code to model Traveling Wave Reactors - TWR
International Nuclear Information System (INIS)
This paper is mainly devoted to the proof-of-principle implementation of the SERPENT code for the simulation of traveling wave reactors. Traveling wave reactors are both fast reactors and nuclear burning wave reactors in which the breeding and burning of nuclear fuel appear almost simultaneously. SERPENT is a neutron transport code whose last official update package is SERPENT 1.1.19 and whose SERPENT 2 version is currently in progress. The investigation of SERPENT 1.1.19 and of SERPENT 2 codes for multiprocessor tasks with long burnup steps was performed. It appears that SERPENT 2 has eliminated parallelization problems efficiently. Methods to remove the influence of the ignition zone were considered, and neutron transport simulations with various fragmentations of the burnup zone were performed. (authors)
Burnup simulations of different fuel grades using the MCNPX Monte Carlo code
Directory of Open Access Journals (Sweden)
Asah-Opoku Fiifi
2014-01-01
Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.
Energy Technology Data Exchange (ETDEWEB)
Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, Service d' Etude des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA/LTSD, F91191 Gif-sur-Yvette cedex (France)
2013-07-01
For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)
SFR whole core burnup calculations with TRIPOLI-4 Monte Carlo code
International Nuclear Information System (INIS)
Under the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD/NEA, an international collaboration benchmark was recently established on the neutronic analysis of four Sodium-cooled Fast Reactor (SFR) concepts of the Generation- IV nuclear energy systems. As the whole core Monte Carlo depletion calculation is one of the essential challenges of current reactor physics studies, the continuous-energy TRIPOLI-4 Monte Carlo transport code was firstly used in this study to perform whole core 3D neutronic calculations for these four SFR cores. Two medium size (1000 MWt) and two large size (3600 MWt) SFR of GEN-IV systems were analyzed. The medium size SFR concepts are from the Advanced Burner Reactors (ABR). The large size SFR concepts are from the self-breeding reactors. The TRIPOLI-4 depletion calculations were made with MOX and metallic U-Pu-Zr fuels for the ABR cores and with MOX and Carbide (U,Pu)C fuels for the self-breeding cores. The whole core reactor physics parameters calculations were performed for the BOEC and EOEC (Beginning and End of Equilibrium Cycle) conditions. This paper summarizes the TRIPOLI-4 calculation results of Keff, βeff, sodium void worth, Doppler constant, control rod worth, and core power distributions for the BOEC and EOEC conditions. The one-cycle depletion calculation results of the core inventory of U and TRU (Pu, Am, Cm, and Np) are also analyzed, after 328.5 days depletion irradiation for the ABR cores, 410 days for the large MOX core, and 500 days for the large carbide core. (author)
International Nuclear Information System (INIS)
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)
Energy Technology Data Exchange (ETDEWEB)
Borio Di Tigliole, A.; Bruni, J.; Panza, F. [Dept. of Nuclear and Theoretical Physics, Univ. of Pavia, 27100 Pavia (Italy); Italian National Inst. of Nuclear Physics INFN, Section of Pavia, Via A. Bassi, 6, 27100 Pavia (Italy); Alloni, D.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Prata, M.; Salvini, A. [Italian National Inst. of Nuclear Physics INFN, Section of Pavia, Via A. Bassi, 6, 27100 Pavia (Italy); Applied Nuclear Energy Laboratory LENA, Univ. of Pavia, Via Aselli, 41, 27100 Pavia (Italy); Chiesa, D.; Clemenza, M.; Pattavina, L.; Previtali, E.; Sisti, M. [Physics Dept. G. Occhialini, Univ. of Milano Bicocca, 20126 Milano (Italy); Italian National Inst. of Nuclear Physics INFN, Section of Milano Bicocca, P.zza della Scienza, 3, 20126 Milano (Italy); Cammi, A. [Italian National Inst. of Nuclear Physics INFN, Section of Milano Bicocca, P.zza della Scienza, 3, 20126 Milano (Italy); Dept. of Energy Enrico Fermi Centre for Nuclear Studies CeSNEF, Polytechnic Univ. of Milan, Via U. Bassi, 34/3, 20100 Milano (Italy)
2012-07-01
Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the Univ. of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation. (authors)
International Nuclear Information System (INIS)
Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the Univ. of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation. (authors)
International Nuclear Information System (INIS)
The deep burn fuel cycle for the incineration of military plutonium in the GT-MHR is studied using the Monte-Carlo burnup code. The irradiation is DF is so rich in fissile isotopes that the TF cannot guarantee a negative reactive feedback, and the presence of erbium as burnable poison is absolutely necessary for the reactivity safety reasons. At beginning of life (BOL) the fuel composed of DF, consisting of fresh military plutonium, after an irradiation period of three years the fuel is reprocessed into post driver fuel (PDF). The mass flow of the GT-MHR fuelled by military plutonium at the equilibrium of the fuel composition shows that 66% of 239Pu is burned in three years and 92% in six years. (authors)
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When the residence time of nuclear fuel rods exceeds a given threshold value, several properties of the pellet material suffer changes and hence the posterior behaviour of the rod is significantly altered. Structural modifications start at the pellet periphery, which is usually referred to as rim zone. It is presently believed that these changes are a consequence of the localized absorption of epithermal neutrons by 238U, which effective cross section presents resonant peaks. Due to the chain of nuclear reactions that take place, several Pu isotopes are born especially at the rim. In particular, the fissile character of 239Pu and 241Pu is the cause of the increased number of fission events that occur in the pellet periphery. For this reason, the power generation rate and the burnup adopt a non uniform distribution in the pellet, reaching at the rim values two or three times higher than the average [1]. The rim zone starts to form for a burnup threshold value of about 50-60 MWd/kgHM and its width increases as the irradiation progresses. The microstructure of this zone is characterized by the presence of small grains, with a typical size of 200 nm, and large pores, of some μm. Even though the rim zone is very thin, it has a significant effect on the mechanical integrity of the pellet, particularly when it makes contact with the cladding, and on the temperature distribution in the whole pellet, because of its low thermal conductivity [1,2]. The numerical codes designed to simulate fuel behaviour under irradiation must include the phenomena associated to high burnup if they aim at extending the prediction range, and this is the purpose with our DIONISIO code. But a detailed analysis of the phenomena that take place in this region demands the use of neutronic codes that solve the Boltzmann transport equations [3] in a number of energy intervals (groups), including adequate considerations in the region of the resonant absorption peaks of 238U. These cell codes predict
Development and verification of Monte Carlo burnup calculation system
International Nuclear Information System (INIS)
Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)
Burnup calculation code system COMRAD96
International Nuclear Information System (INIS)
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)
Validation of IRBURN calculation code system through burnup benchmark analysis
International Nuclear Information System (INIS)
Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.
Integrated burnup calculation code system SWAT
International Nuclear Information System (INIS)
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)
Systemization of burnup sensitivity analysis code. 2
International Nuclear Information System (INIS)
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For
Systemization of burnup sensitivity analysis code
International Nuclear Information System (INIS)
To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this
Burnup calculations using serpent code in accelerator driven thorium reactors
International Nuclear Information System (INIS)
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
Burnup calculations using serpent code in accelerator driven thorium reactors
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.
2013-07-15
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
High burnup models in computer code fair
International Nuclear Information System (INIS)
An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs
BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies
International Nuclear Information System (INIS)
BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inventory as a function of time and power density. The code allows for multiple fueled regions to be analyzed. The companion code, RELOAD, can be used to shuffle fueled regions or reload regions with fresh fuel. BURNCAL can be used to study the reactivity effects and isotopic inventory as a function of time for a nuclear reactor system. Neutron transmutation, fission, and radioactive decay are included in the modeling of the production and removal terms for each isotope of interest. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. Fueled and un-fueled regions, such as cladding and moderator, can be analyzed simultaneously. The nuclides analyzed are limited only by the neutron cross section availability in the MCNP cross-section library. BURNCAL is unique in comparison to other burnup codes in that it does not use the calculated neutron flux as input to other computer codes to generate the nuclide mixture for the next time step. Instead, BURNCAL directly uses the neutron absorption tally/reaction information generated by MCNP for each nuclide of interest to determine the nuclide inventory for that region. This allows for the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations
Energy Technology Data Exchange (ETDEWEB)
Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)
2008-04-15
Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.
Revised SWAT. The integrated burnup calculation code system
International Nuclear Information System (INIS)
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
Revised SWAT. The integrated burnup calculation code system
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)
2000-07-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises
International Nuclear Information System (INIS)
The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.
Development and validation of burnup function in reactor Monte Carlo RMC
International Nuclear Information System (INIS)
This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including middle-of-step approximation and predictor-corrector method, are adopted by RMC to assure accuracy under large step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably save computational time with negligible accuracy loss. According to validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (author)
International Nuclear Information System (INIS)
Graphical abstract: - Highlights: • Continuous Energy Monte-Carlo burnup code. • Instabilities of depletion calculation in loosely coupled system. • Advanced step model for burnup calculations. • Xenon profile oscillation in thermal reactor. • Parametrical study of instabilities. - Abstract: In this paper we use the Continuous Energy Monte-Carlo tool to expose the problem of burnup instabilities occurring in 1D and 2D systems based on PWR geometry. The intensity of power profile oscillations is studied as a function of geometry properties and time step length. We compare two step models for depletion procedure: classic staircase step model and stochastic implicit Euler method, that belongs to the family of predictor–corrector schemes. What is more, we consider the usage of better neutron source intensity value than beginning-of-step approximation. Required methodology was implemented into MCB5 simulation code. The practical conclusions about depletion calculations were formulated and the efficiency of advanced step model was confirmed
Miniature neutron source reactor burnup calculations using IRBURN code system
International Nuclear Information System (INIS)
Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.
SERPENT Monte Carlo reactor physics code
International Nuclear Information System (INIS)
SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)
Systemization of burnup sensitivity analysis code (2) (Contract research)
International Nuclear Information System (INIS)
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion
Methods of RECORD, an LWR fuel assembly burnup code
International Nuclear Information System (INIS)
The RECORD computer code is a detailed rector physics code for performing efficient LWR fuel assembly calculations, taking into account most of the features found in BWR and PWR fuel designs. The code calculates neutron spectrum, reaction rates and reactivity as a function of fuel burnup, and it generates the few-group data required for use in full scale core simulation and fuel management calculations. The report describes the methods of the RECORD computer code and the basis for fundamental models selected, and gives a review of code qualifications against measured data. (Auth. /RF)
Evolution of the ELESTRES code for application to extended burnups
International Nuclear Information System (INIS)
The computer code ELESTRES is frequently used at Atomic Energy of Canada Limited to assess the integrity of CANDU fuel under normal operating conditions. The code also provides initial conditions for evaluating fuel behaviour during high-temperature transients. This paper describes recent improvements in the code in the areas of pellet expansion and of fission gas release. Both of these are very important considerations in ensuring fuel integrity at extended burnups. Firstly, in calculations of pellet expansion, the code now accounts for the effect of thermal stresses on the volume of gas bubbles at the boundaries of UO2 grains. This has a major influence on the expansion of the pellet during power-ramps. Secondly, comparisons with data showed that the previous fission gas package significantly underpredicted the fission gas release at high burnups. This package has now been improved via modifications to the following modules: distance between neighbouring bubbles on grain boundaries; diffusivity; and thermal conductivity. The predictions of the revised version of the code show reasonable agreement with measurements of ridge strains and of fission gas release. An illustrative example demonstrates that the code can be used to identify a fuel design that would: reduce the sheath stresses at circumferential ridges by a factor of 2-10; and keep the gas pressure at very high burnups to below the coolant pressure
International Nuclear Information System (INIS)
External linking scripts between Monte Carlo transport codes and burnup codes, and complete integration of burnup capability into Monte Carlo transport codes, have been or are currently being developed. Monte Carlo linked burnup methodologies may serve as an excellent benchmark for new deterministic burnup codes used for advanced systems; however, there are some instances where deterministic methodologies break down (i.e., heavily angularly biased systems containing exotic materials without proper group structure) and Monte Carlo burn up may serve as an actual design tool. Therefore, researchers are also developing these capabilities in order to examine complex, three-dimensional exotic material systems that do not contain benchmark data. Providing a reference scheme implies being able to associate statistical errors to any neutronic value of interest like k(eff), reaction rates, fluxes, etc. Usually in Monte Carlo, standard deviations are associated with a particular value by performing different independent and identical simulations (also referred to as 'cycles', 'batches', or 'replicas'), but this is only valid if the calculation itself is not biased. And, as will be shown in this paper, there is a bias in the methodology that consists of coupling transport and depletion codes because Bateman equations are not linear functions of the fluxes or of the reaction rates (those quantities being always measured with an uncertainty). Therefore, we have to quantify and correct this bias. This will be achieved by deriving an unbiased minimum variance estimator of a matrix exponential function of a normal mean. The result is then used to propose a reference scheme to solve Boltzmann/Bateman coupled equations, thanks to Monte Carlo transport codes. Numerical tests will be performed with an ad hoc Monte Carlo code on a very simple depletion case and will be compared to the theoretical results obtained with the reference scheme. Finally, the statistical error propagation
MCOR - Monte Carlo depletion code for reference LWR calculations
International Nuclear Information System (INIS)
Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations
MCOR - Monte Carlo depletion code for reference LWR calculations
Energy Technology Data Exchange (ETDEWEB)
Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)
2011-04-15
Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
Energy Technology Data Exchange (ETDEWEB)
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)
2015-09-15
A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.
OTTER 3 - A single channel, axial burnup code
International Nuclear Information System (INIS)
OTTER 3 is a single channel, axial burnup code, written in Fortran for the KDF 9 computer, and suitable for studying fuel management schemes of the continuous charge/discharge type. A general fuel shuffling scheme is allowed, and both unidirectional and bidirectional fuel feed can be studied. A 2-group neutron diffusion code is incorporated, the flux equations being solved by the forward elimination - backward substitution technique for the inner problem and a source iteration technique accelerated by Chebyshev extrapolation for the outer problem. (author)
Coupled Monte Carlo and burnup calculations for research and power reactors
International Nuclear Information System (INIS)
Reactor physics calculations require the solution of transport equation under consideration of complex energy structure of cross sections and spatial distribution of materials. Since the material composition changes during operation due to burn-up this changes have to be regarded in detail. An accurate method for reactor physics calculations is the continuous energy Monte Carlo method e.g. represented by the code MCNP(4C). The performance and quality of this code is demonstrated by own and world-wide applications. With the Monte Carlo method coupled with a program for the solution of the equations for nuclide build-up and decay not only the composition of fresh fuel but also the composition of irradiated fuel can be taken into account without external calculation of nuclide composition of irradiated materials. The coupled method represents at time the most accurate solution of criticality problems if the calculated parameters are determined with sufficient low statistical uncertainty. By this method important safety related parameters and reactivity conditions can be analysed for research reactors, standard LWR design as well as for innovative reactor or fuel designs or systems for actinide burning. (author)
Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment
International Nuclear Information System (INIS)
Highlights: • MCNP and ORIGEN are coupled to perform nuclides depletion and decay calculation. • Coupled system MCORE uses “modified predictor corrector” approach. • MCORE can use different depletion schemes and simulate fuel shuffling. • MCORE is assessed by a “VVER-1000 LEU Assembly Computational Benchmark”. • MCORE is also assessed by a fast reactor benchmark problem. - Abstract: An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA). The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k∞ within 800 pcm at the end of the burn-up for LEU assembly (40 MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for 243Am. In general, MCORE results show a good agreement with the benchmark values
A multi-platform linking code for fuel burnup and radiotoxicity analysis
International Nuclear Information System (INIS)
A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory. - Highlights: • Multi-platform execution and automatic results analysis are implemented. • Improvements allow simulations in a simpler and faster way. • The results show good agreement between GB and Monteburns codes
International Nuclear Information System (INIS)
The Monte Carlo code MONK is a general program written to provide a high degree of flexibility to the user. MONK is distinguished by its detailed representation of nuclear data in point form i.e., the cross-section is tabulated at specific energies instead of the more usual group representation. The nuclear data are unadjusted in the point form but recently the code has been modified to accept adjusted group data as used in fast and thermal reactor applications. The various geometrical handling capabilities and importance sampling techniques are described. In addition to the nuclear data aspects, the following features are also described; geometrical handling routines, tracking cycles, neutron source and output facilities. 12 references. (U.S.)
Burnup calculations using the ORIGEN code in the CONKEMO computing system
International Nuclear Information System (INIS)
This article describes the CONKEMO computing system for kinetic multigroup calculations of nuclear reactors and their physical characteristics during burnup. The ORIGEN burnup calculation code has been added to the system. The results of an international benchmark calculation are also presented. (author)
Dufek, Jan; Anglart, Henryk
2013-01-01
Numerically stable Monte Carlo burnup calculations of nuclear fuel cycles are now possible with the previously derived Stochastic Implicit Euler method based coupling scheme. In this paper, we show that this scheme can be easily extended to include the thermal-hydraulic feedback during the Monte Carlo burnup simulations, while preserving its unconditional stability property. At each time step, the implicit solution (for the end-of-step neutron flux, fuel nuclide densities and thermal-hydrauli...
International Nuclear Information System (INIS)
The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)
Development of high-burnup fuel analysis code EXBURN-I
International Nuclear Information System (INIS)
A computer code EXBURN-I has been developed which analyses LWR fuel behavior in high-burnup region in normal operation and transient conditions. In the high-burnup region, fuel behavior is affected considerably by such burnup-dependent factors as FP gas release, waterside corrosion of cladding, and pellet property change. To analyze these phenomena, in the present version, the base code FEMAXI-IV has been improved and incorporated such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding waterside corrosion. The present report describes the whole structure of the code, adopted models, and material properties, followed by input manual and sample input/output. Verification and further improvement of the code performance by experimental data will be done in the next stage. (author)
Energy Technology Data Exchange (ETDEWEB)
Holly R. Trellue
1998-12-01
Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.
Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor
International Nuclear Information System (INIS)
Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanium (HPGe) gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the 137Cs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% (1σ). The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors. (authors)
Monte Carlo solver for UWB1 nuclear fuel depletion code
International Nuclear Information System (INIS)
Highlights: • A new Monte Carlo solver was developed in order to speed-up depletion calculations. • For LWR model, UWB1 Monte Carlo solver is on average 10 times faster than MCNP6. • The UWB1 code will allow faster calculation analysis of BA parameters in fuel design. - Abstract: Recent nuclear reactor burnable absorber research tries to introduce new materials in the nuclear fuel. As a part of this effort, a fast computational tool is being developed for the advanced nuclear fuel. The first version of the newly developed UWB1 fast nuclear fuel depletion code significantly reduced calculation time by omitting the solution step for the Boltzmann transport equation. However, estimation of neutron multiplication factor during depletion was not sufficiently calculated. Therefore, at least one transport calculation for fuel depletion is necessary. This paper presents a new Monte Carlo solver that is implemented into the UWB1 code. The UWB1 Monte Carlo solver calculates neutron multiplication factor and neutron flux in the fuel for collapsed cross sections. Accuracy of the solver is supported by using current nuclear data stored in the ENDF/B-VII.1 library. Speed of the solver is the product of development focusing on minimization of CPU utilization at the expense of RAM demands. The UWB1 Monte Carlo solver is approximately 14 times faster than the MCNP6 reference code when one transport equation solution within fuel depletion is compared. Another speed-up can be achieved by employing advanced depletion scheme in the coupled transport and burnup equations. The resulting faster code will be used in optimization studies for ideal burnable absorber material selection where many various materials and concentrations will be evaluated
International Nuclear Information System (INIS)
The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that keff and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)
Monte Carlo uncertainty propagation approaches in ADS burn-up calculations
International Nuclear Information System (INIS)
Highlights: ► Two Monte Carlo uncertainty propagation approaches are compared. ► How to make both approaches equivalent is presented and applied. ► ADS burn-up calculation is selected as the application of approaches. ► The cross-section uncertainties of 239Pu and 241Pu are propagated. ► Cross-correlations appear as a source of differences between approaches. - Abstract: In activation calculations, there are several approaches to quantify uncertainties: deterministic by means of sensitivity analysis, and stochastic by means of Monte Carlo. Here, two different Monte Carlo approaches for nuclear data uncertainty are presented: the first one is the Total Monte Carlo (TMC). The second one is by means of a Monte Carlo sampling of the covariance information included in the nuclear data libraries to propagate these uncertainties throughout the activation calculations. This last approach is what we named Covariance Uncertainty Propagation, CUP. This work presents both approaches and their differences. Also, they are compared by means of an activation calculation, where the cross-section uncertainties of 239Pu and 241Pu are propagated in an ADS activation calculation
Revised Burnup Code System SWAT: Description and Validation Using Postirradiation Examination Data
International Nuclear Information System (INIS)
The burnup code system Step-Wise Burnup Analysis Code System (SWAT) is revised for use in a burnup credit analysis. An important feature of the revised SWAT is that its functions are achieved by calling validated neutronics codes without any changes to the original codes. This feature is realized with a system function of the operating system, which allows the revised SWAT to be independent of the development status of each code.A package of the revised SWAT contains the latest libraries based on JENDL-3.2 and the second version of the JNDC FP library. These libraries allow us to analyze burnup problems, such as an analysis of postirradiation examination (PIE), using the latest evaluated data of not only cross sections but also fission yield and decay constants.Another function of the revised SWAT is a library generator for the ORIGEN2 code, which is one of the most reliable burnup codes. ORIGEN2 users can obtain almost the same results with the revised SWAT using the library prepared by this function.The validation of the revised SWAT is conducted by calculation of the Organization for Economic Cooperation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark Phase I-B and analyses of PIE data for spent fuel from Takahama Unit 3. The analysis of PIE data shows that the revised SWAT can predict the isotopic composition of main uranium and plutonium with a deviation of 5% from experimental results taken from UO2 fuels of 17 x 17 fuel assemblies. Many results of fission products including samarium are within a deviation of 10%. This means that the revised SWAT has high reliability to predict the isotopic composition for pressurized water reactor spent fuel
Development and validation of a fast reactor core burnup code - FARCOB
Energy Technology Data Exchange (ETDEWEB)
Mohanakrishnan, P. [Indira Gandhi Centre for the Atomic Research, CDO, Reactor Physics Division, Kalpakkam, TN 603 102 (India)], E-mail: mohana@igcar.gov.in
2008-02-15
A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X-Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.
Development of a New Monte Carlo reactor physics code
International Nuclear Information System (INIS)
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. An interesting near-future application for the Monte Carlo method is the generation of input parameters for deterministic reactor simulator codes. These codes are used in coupled LWR full-core analyses and typically based on few-group nodal diffusion methods. The input data consists of homogenised few-group constants, presently generated using deterministic lattice transport codes. The task is becoming increasingly challenging, along with the development in nuclear technology. Calculations involving high-burnup fuels, advanced MOX technology and next-generation reactor systems are likely to cause problems in the future, if code development cannot keep up with the applications. A potential solution is the use of Monte Carlo based lattice transport codes, which brings all the advantages of the calculation method. So far there has been only a handful of studies on group constant generation using the Monte Carlo method, although the interest has clearly increased during the past few years. The homogenisation of reaction cross sections is simple and straightforward, and it can be carried out using any Monte Carlo code. Some of the parameters, however, require the use of special techniques that are usually not available in general-purpose codes. The main problem is the calculation of neutron diffusion coefficients, which
Siemens PWR burnup credit criticality analysis methodology: Depletion code and verification methods
International Nuclear Information System (INIS)
Application of burnup credit requires knowledge of the reactivity state of the irradiated fuel for which burnup credit is taken. The isotopic inventory of the irradiated fuel has to be calculated, therefore, by means of depletion codes. Siemens performs depletion calculations for PWR fuel burnup credit applications with the aid of the code package SAV. This code package is based on the first principles approach, i.e., avoids cycle or reactor specific fitting or adjustment parameters. This approach requires a general and comprehensive qualification of SAV by comparing experimental with calculational results. In the paper on hand the attention is focused mainly on the evaluation of chemical assay data received from different experimental programmes. (author)
International Nuclear Information System (INIS)
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
Development of Monte Carlo-based pebble bed reactor fuel management code
International Nuclear Information System (INIS)
Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower keff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher keff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed
New high burnup fuel models for NRC`s licensing audit code, FRAPCON
Energy Technology Data Exchange (ETDEWEB)
Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)
1996-03-01
Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.
SPQR: a Monte Carlo reactor kinetics code
International Nuclear Information System (INIS)
The SPQR Monte Carlo code has been developed to analyze fast reactor core accident problems where conventional methods are considered inadequate. The code is based on the adiabatic approximation of the quasi-static method. This initial version contains no automatic material motion or feedback. An existing Monte Carlo code is used to calculate the shape functions and the integral quantities needed in the kinetics module. Several sample problems have been devised and analyzed. Due to the large statistical uncertainty associated with the calculation of reactivity in accident simulations, the results, especially at later times, differ greatly from deterministic methods. It was also found that in large uncoupled systems, the Monte Carlo method has difficulty in handling asymmetric perturbations
The Serpent Monte Carlo Code: Status, Development and Applications in 2013
Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni
2014-06-01
The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.
DELIGHT-6: one dimensional lattice burn-up code for high temperature gas-cooled reactors
International Nuclear Information System (INIS)
The code, DELIGHT-6, performs multi-group neutron spectrum calculation and provides few-group constans for succeeding core calculations. The main objective of the code is to serve as the lattice burn-up code for the core of a very high temperature gas-cooled reactor. The fuel rods of the reactor contain many coated fuel particles resulting double heterogeneous arrangement. The main calculational schema of DELIGHT-6 code is as follows; (1) Energy range for fast neutrons covers from 10 MeV to 2.38 eV and is divided into 61 fine groups. The thermal neutrons covers the rest of the energy range from 2.38 eV to 0 eV. Thermal spectrum is calculated by P1 or P0 approximation with 50 fine groups. (2) To treat resonance absorption, IR method is employed. (3) Zero and one dimensional models are available for the fuel lattice geometry and used for criticality and burn-up calculations. Collision probability method is adopted for the calculation of one dimensional model. (4) Shielding factor of burnable poison is calculated by collision probability method. (5) Other functions of the code are; 1. Spatial shielding factor calculation of 240Pu, 2. Calculation of neutron streaming effect caused by a gap or a hole in the fuel lattice, 3. Calculation of neutron flux distribution in the fuel lattice by diffusion theory, 4. Calculation of Xe and Sm absorption cross sections with burn-up. (6) Cross section library in both fast and thermal energy range is compiled from ENDF/B-4 except burn-up data of Xm, Sm and pseudo FPs which are supplied by ENDF/B-3. (7) The code provides the macroscopic group constants of fuel lattice with burn-up in CITATION input format. (jin)
Coded aperture optimization using Monte Carlo simulations
International Nuclear Information System (INIS)
Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.
Point reactivity burnup code DELIGHT-4 for high temperature, gas-cooled reactor cells
International Nuclear Information System (INIS)
The code DELIGHT-4 has been developed for analizing burnup characteristics of the graphite moderated reactor cells and producing the few-group constants. Calculation models for the code are as follows: (1) The number of neutron energy groups is 61 for fast neutrons (10 MeV -- 2.38 eV) and 50 for thermal neutrons (2.38 eV -- 0 eV). (2) The doubly space-heterogeneous effect of fuel (dispersion of coated fuel particles in fuel compacts and regular array of fuel rods in graphite blocks) is considered in the calculation of resonance absorption. (3) The double heterogenity of burnable poison (dispersion of absorber grains in rods) can be considered. (4) The chemical binding effect of graphite is introduced in the scattering of thermal neutrons. (5) The calculations of criticality and burnup are by a few-energy-group models (up to 10 groups for both fast and thermal neutrons), and nuclide chains of thorium-uranium and uranium-plutonium are used for burnup calculation. (6) Neutron streaming effect through holes and gaps in cells can be considered in criticality calculation. (7) The flux distribution in cells can be calculated. The cell-averaged few group constants can be produced in card form for 1-D transport approximation code SLALOM, 2-D S sub( n) code TWOTRAN, 1-D diffusion code BRIQUET, 2-D diffusion code ZADOC-3 and 3-D diffusion code CITATION-DEGA. (author)
Successful vectorization - reactor physics Monte Carlo code
International Nuclear Information System (INIS)
Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)
Recent developments of the TRANSURANUS code with emphasis on high burnup phenomena
International Nuclear Information System (INIS)
TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors, which is developed at the Institute for Transuranium Elements. The code is in use in several European organisations, both in research and industry. In the paper the recent developments are summarised: the burnup degradation of the fuel thermal conductivity as well as the effects of gadolinium on the radial power distribution and thermal conductivity. Fission gas release from the High Burnup Structure is discussed. Finally, a new numerical method is outlined that is able to treat the highly non-linear mechanical equations in transients (RIAs and LOCAs). (author)
International Nuclear Information System (INIS)
The burnup analysis program based on three dimensional discrete ordinates (SN) neutron/photon transport method has been developed by the FDS team, China, to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The program uses output parameters generated by three-dimensional SN trans- port code to determine the isotopic inventory and anisotropic flux distribution as a function of time. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. The IAEA benchmark test problem has been correctly calculated and analyzed to validate the system. (authors)
International Nuclear Information System (INIS)
The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX.
Parallel processing Monte Carlo radiation transport codes
International Nuclear Information System (INIS)
Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine
SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2
International Nuclear Information System (INIS)
1 - Description of program or function: SWAT evaluates isotopic composition of spent nuclear fuel, especially for burnup credit issues by driving codes SRAC95 and ORIGEN2.1 or ORIGEN2. SWAT is an automated driver code system. At the initial development phase, it was constructed by combining source programs of SRAC and ORIGEN2. To overcome the problem associated with code updates, SWAT chose to use system function of UNIX operating system to execute SRAC95 and ORIGEN2. So that, SWAT is independent of development and modification of SRAC95 and ORIGEN2.1. In SWAT, ORIGEN2(82) or ORIGEN2.1 is used for burnup calculations using the matrix exponential method. An updated decay library is included in the distribution. SWAT uses SRAC95 for neutron spectrum and effective cross section calculation in 107 groups, using the collision probability method for given geometry and isotopic composition. One or two dimensional cell geometries are supported in SRAC95. NEA-1698/02: The main purpose of new package is to run SWAT on several machines not supported in previous package (IA64 under Linux, Windows with cygwin and Sun,...) and several commercial FORTRAN compiler (Intel, PGI, Fujitsu). 2 - Methods: In calculating the problem-dependent cross section in SWAT, the total burnup history is divided into 'burnup steps'. Power, boric acid concentration, temperature of each region, and void ratio of coolant are given as history data. For each burnup step, the neutron spectrum and effective cross section are evaluated by SRAC95 using the information given in previous burnup calculation and cell geometry information. The user can select geometry options for the collision probability method in SRAC95. 3 - Restrictions on the complexity of the problem: Resonance absorption calculation with ultra-fine group cross section can not be directly applicable for 2D geometry
Monte Carlo simulation code modernization
CERN. Geneva
2015-01-01
The continual development of sophisticated transport simulation algorithms allows increasingly accurate description of the effect of the passage of particles through matter. This modelling capability finds applications in a large spectrum of fields from medicine to astrophysics, and of course HEP. These new capabilities however come at the cost of a greater computational intensity of the new models, which has the effect of increasing the demands of computing resources. This is particularly true for HEP, where the demand for more simulation are driven by the need of both more accuracy and more precision, i.e. better models and more events. Usually HEP has relied on the "Moore's law" evolution, but since almost ten years the increase in clock speed has withered and computing capacity comes in the form of hardware architectures of many-core or accelerated processors. To harness these opportunities we need to adapt our code to concurrent programming models taking advantages of both SIMD and SIMT architectures. Th...
Modification of the code SCTEMP and RIA transient analysis at high burnup
International Nuclear Information System (INIS)
The code SCTEMP has been modified for RIA analysis of high burnup fuel. New models were introduced for this purpose, including rim type radial power profile, thermal conductivity degradation, and heat transfer coefficient covering different regimes. Thus, the thermal response during a RIA transient at high burnup can be simulated. An analytical exercise was made with parameter variations of pulse time amplitude, radial power profile, thermal conductivity degradation, energy deposition and heat transfer boundary conditions. In order to keep this exercise within a realistic frame, the fuel used as reference is a high burnup fuel which was tested at Halden and for which the fuel thermal characteristics are known based on in-pile fuel temperature measurements. The calculation results are analysed and discussed in terms of their significance for fuel behaviour under the transient and for design of possible RIA experiments. Pulse time amplitude effects and energy deposition effects of high burnup fuel were evaluated quantitatively. The rim of high burnup fuel is assumed to play a very important role for RIA transients. (author)
Energy Technology Data Exchange (ETDEWEB)
Horhoianu, G.; Moscalu, D.R.; Popescu, I.A
1998-06-01
Extending burnup is a practical way to improve the economics of water-reactor operation, via enhanced fuel utilisation and reduced spent fuel volume. A dedicated fuel behaviour modeling computer code (entitled ROFEM-1B) has been developed in order to analyse high burnup fuel performance. The code was benchmarked on an experimental data base which include a significant number of irradiation experiments performed in TRIGA-INR Pitesti research reactor. Five fuel rod behaviour during irradiation up to 50 MWd kg{sup -1}UO{sub 2}{sup -1} burnup have been analysed by the code in the framework of the first phase of the international FUMEX code exercise co-ordinated by IAEA Vienna. The input experimental data package has been prepared by IFE-OECD Halden, Norway laboratory. In the second phase of the FUMEX exercise the participants have analysed eight simplified theoretical cases. The paper presents and discuss the results obtained with ROFEM-1B in this exercise and the comparison between code predictions and experimental data.
Development of a fuel rod thermal-mechanical analysis code for high burnup fuel
International Nuclear Information System (INIS)
The thermal-mechanical analysis code for high burnup BWR fuel rod has been developed by NFI. The irradiation data accumulated up to the assembly burnup of 55 GWd/t in commercial BWRs were adopted for the modeling. In the code, pellet thermal conductivity degradation with burnup progress was considered. Effects of the soluble FPs, irradiation defects and porosity increase due to RIM effect were taken into the model. In addition to the pellet thermal conductivity degradation, the pellet swelling due to the RIM porosity was studied. The modeling for the high burnup effects was also carried out for (U, Gd)O2 and MOX fuel. The thermal conductivities of all pellet types, UO2, (U, Gd)O2 and (U, Pu)O2 pellets, are expressed by the same form of equation with individual coefficient γ in the code. The pellet center temperature was calculated using this modeling code, and compared with measured values for the code verification. The pellet center temperature calculated using the thermal conductivity degradation model agreed well with the measured values within ±150 deg. C. The influence of rim porosity on pellet center temperature is small, and the temperature increase in only 30 deg. C at 75 GWd/t and 200 W/cm. The pellet center temperature of MOX fuel was also calculated, and it was found that the pellet center temperature of MOX fuel with 10wt% PuO2 is about 60 deg. C higher than UO2 fuel at 75 GWd/t and 200 W/cm. (author)
Morse Monte Carlo Radiation Transport Code System
Energy Technology Data Exchange (ETDEWEB)
Emmett, M.B.
1975-02-01
The report contains sections containing descriptions of the MORSE and PICTURE codes, input descriptions, sample problems, deviations of the physical equations and explanations of the various error messages. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. The PICTURE code provide aid in preparing correct input data for the combinatorial geometry package CG. It provides a printed view of arbitrary two-dimensional slices through the geometry. By inspecting these pictures one may determine if the geometry specified by the input cards is indeed the desired geometry. 23 refs. (WRF)
Monte Carlo calculations of the REBUS critical experiment for validation of burnup credit
International Nuclear Information System (INIS)
The application of burnup credit (BUC) to criticality safety analysis for Spent Nuclear Fuel (SNF) configurations requires the implementation of both estimation of the SNF composition with the aid of depletion calculation tools and estimation of the SNF reactivity with the aid of criticality calculation tools. Amongst the several experimental programs dedicated to the validation of both calculation tools, REBUS is distinguished by a combination of chemical analysis and critical experiment. In addition to detailed assays of irradiated fuel, the reactivity worth of the fuel rods under investigation is measured both before and after irradiation. Since a whole bundle of fuel rods is used in the experiment, the change in reactivity is significant enough to be observable by Monte Carlo calculations. Thus, the calculation tools which see the most widespread use in SNF critical safety applications can be validated directly. Apart from the effective neutron multiplication factor keff, REBUS also provides measurements of the flux and fission rate distributions. While the program comprises investigation of commercial UO2 fuel rods and mixed oxide (MOX) fuel from a research reactor, the presentation will focus on the commercial UO2 fuel with an overview of the experimental setup and first results from the analysis. (author)
First steps towards a validation of the new burnup and depletion code TNT
International Nuclear Information System (INIS)
In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)
The MCNPX Monte Carlo Radiation Transport Code
International Nuclear Information System (INIS)
MCNPX (Monte Carlo N-Particle eXtended) is a general-purpose Monte Carlo radiation transport code with three-dimensional geometry and continuous-energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi-processing computer platforms. MCNPX is based on MCNP4c and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low-energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics, particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development
THE MCNPX MONTE CARLO RADIATION TRANSPORT CODE
Energy Technology Data Exchange (ETDEWEB)
WATERS, LAURIE S. [Los Alamos National Laboratory; MCKINNEY, GREGG W. [Los Alamos National Laboratory; DURKEE, JOE W. [Los Alamos National Laboratory; FENSIN, MICHAEL L. [Los Alamos National Laboratory; JAMES, MICHAEL R. [Los Alamos National Laboratory; JOHNS, RUSSELL C. [Los Alamos National Laboratory; PELOWITZ, DENISE B. [Los Alamos National Laboratory
2007-01-10
MCNPX (Monte Carlo N-Particle eXtended) is a general-purpose Monte Carlo radiation transport code with three-dimensional geometry and continuous-energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi-processing computer platforms. MCNPX is based on MCNP4B, and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low-energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics; particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development.
International Nuclear Information System (INIS)
An iterative method for burn-up determination in the non-destructive analysis of irradiated fuel elements using the ORIGEN computer code is presented. On the bases of data obtained from ORIGEN code the calibration coefficient for the neutron flux is determined as a function of one fission product activity while the burn-up is determined as a function of the calibration coefficient for a given irradiation history. These functions are used for determining the burn-up of nuclear fuel elements measured by gamma-scanning. The method is tested for fuel elements irradiated in a TRIGA reactor facility. (Author)
DELIGHT-6(revised): one dimensional lattice burnup code for high temperature gas cooled reactors
International Nuclear Information System (INIS)
The code, DELIGHT-6, performs the multi-group neutron spectrum calculation and provides the few-group constants for burnup calculations of a high temperature gas-cooled reactor core, whose fuel elements containing many coated fuel particles are arranged in double heterogeneity. The main revisions in the DELIGHT-6 (Revised) are as follows; (1)The option of a sphere fuel cell calculation is added for the core design of pebble bed type high temperature gas-cooled reactor. (2)The yield and decay constants of fission products for burnup calculation is revised. (3)The following auxiliary functions are added; (i) Automatic calculation of averaged atom number density in the fuel region, (ii) Estimation of local neutron flux distribution (disadvantage factor), (iii) Preparation of the data for the fine mesh core calculation. (author)
MODRIB - a zero dimensional code for criticality and burn-up of LWR's
International Nuclear Information System (INIS)
The computer program MODRIB is a zero-dimensional code for calculating criticality and burn-up of light water reactors (LWR's). It is a version of an Italian code RIBOT-2 with an updated cross-section data library. The nuclear constants of MODRIB-code are calculated with a two group scheme (fast and thermal), where the fast group is an average of three fast groups. The code requires as input data essential extensive reactor parameters such as fuel rod radius, clad thickness, fuel enrichment, lattice pitch, water density and temperature etc. A summary of the physical model description and the input-output procedures are given in this report. Selected results of two sample problems are also given for the purpose of checking the validity and reliability of the code. The first is BWR and the second is PWR. The calculation time for a criticality problem with burn-up is about 8 seconds for the first time step and about 3 seconds for each subsequent time step on the ICL-1906 computer facility. The requirements on the memory size is less than 32 K-word. (author)
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Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))
1999-12-15
Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)
ELESTRES 2.1 computer code for high burnup CANDU fuel performance analysis
International Nuclear Information System (INIS)
The ELESTRES (ELEment Simulation and sTRESses) computer code models the thermal, mechanical and micro structural behaviours of CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains in fuel element design analysis and assessments. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. ELESTRES 2.1 was developed for high burnup fuel application, based on an industry standard tool version of the code, through the implementation or modification to code models such as fission gas release, fuel pellet densification, flux depression (radial power distribution in the fuel pellet), fuel pellet thermal conductivity, fuel sheath creep, fuel sheath yield strength, fuel sheath oxidation, two dimensional heat transfer between the fuel pellet and the fuel sheath; and an automatic finite element meshing capability to handle various fuel pellet shapes. The ELESTRES 2.1 code design and development was planned, implemented, verified, validated, and documented in accordance with the AECL software quality assurance program, which meets the requirements of the Canadian Standards Association standard for software quality assurance CSA N286.7-99. This paper presents an overview of the ELESTRES 2.1 code with descriptions of the code's theoretical background, solution methodologies, application range, input data, and interface with other analytical tools. Code verification and validation results, which are also discussed in the paper, have confirmed that ELESTRES 2.1 is capable of modelling important fuel phenomena and the code can be used in the design assessment and the verification of high burnup fuels. (author)
Criticality benchmarking of ANET Monte Carlo code
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In this work the new Monte Carlo code ANET is tested on criticality calculations. ANET is developed based on the high energy physics code GEANT of CERN and aims at progressively satisfying several requirements regarding both simulations of GEN II/III reactors, as well as of innovative nuclear reactor designs such as the Accelerator Driven Systems (ADSs). Here ANET is applied on three different nuclear configurations, including a subcritical assembly, a Material Testing Reactor and the conceptual configuration of an ADS. In the first case, calculation of the effective multiplication factor (keff) are performed for the Training Nuclear Reactor of the Aristotle University of Thessaloniki, while in the second case keff is computed for the fresh fueled core of the Portuguese research reactor (RPJ) just after its conversion to Low Enriched Uranium, considering the control rods at the position that renders the reactor critical. In both cases ANET computations are compared with corresponding results obtained by three different well established codes, including both deterministic (XSDRNPM/CITATION) and Monte Carlo (TRIPOLI, MCNP). In the RPI case, keff computations are also compared with observations during the reactor core commissioning since the control rods are considered at criticality position. The above verification studies show ANET to produce reasonable results since they are satisfactorily compared with other models as well as with observations. For the third case (ADS), preliminary ANET computations of keff for various intensities of the proton beam are presented, showing also a reasonable code performance concerning both the order of magnitude and the relative variation of the computed parameter. (author)
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In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor keff (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System
International Nuclear Information System (INIS)
1 - Description of program or function: The LOLA-SYSTEM is a part of the JEN-UPM code package for PWR fuel management, scope or design calculations. It is a code package for core burnup calculations using nodal theory based on a FLARE type code. The LOLA-SYSTEM includes four modules: the first one (MELON-3) generates the constants of the K-inf and M2 correlations to be input into SIMULA-3. It needs the K-inf and M2 fuel assembly values at different conditions of moderator temperature, Boron concentration, burnup, etc., which are provided by MARIA fuel assembly calculations. The main module (SIMULA-3) is the core burnup calculation code in three dimensions and one group of energy. It normally uses a geometrical representation of one node per fuel assembly or per quarter of fuel assembly. It has included a thermal hydraulic feedback on flow and voids and criticality searches on boron concentration and control rods insertion. The CONCON code makes the calculation of the albedo, transport factors, K-inf and M2 correction factors to be input into SIMULA-3. The calculation is made in the XY transversal plane. The CONAXI code is similar to CONCON, but in the axial direction. 2 - Method of solution: MELON-3 makes a mean squares fit of K-inf and M2 values at different conditions in order to determine the constants of the feedback correlations. SIMULA-3 uses a modified one-group nodal theory, with a new transport kernel that provides the same node interface leakages as a fine mesh diffusion calculation. CONCON and CONAXI determine the transport and correction factors, as well as the albedo, to be input into SIMULA-3. They are determined by a method of leakages equivalent to the detailed diffusion calculation of CARMEN or VENTURE; these factors also include the heterogeneity effects inside the node. 3 - Restrictions on the complexity of the problem: Number of axial nodes less than or equal 34. Number of material types less than or equal 30. Number of fuel assembly types less
High burnup fuel simulation during a reactivity initiated transient using the FRAPTRAN code
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CIP01 test simulated a RIA (Reactivity Initiated Accident) transient at the experimental CABRI reactor under sodium cooling conditions on November 29th 2001. The father rod was fabricated by ENUSA and irradiated during five cycles in Vandellos 2 NPP and it reached a maximum burnup at EOL (End of Life) of 75 GWd/tU. The test rodlet was refabricated in Studsvik from the span 5 of the father rod. The cladding material was ZIRLO with a mean oxide layer of 75 μm. This note describes the modelling of the CIP01 test carried out in the CABRI reactor under RIA conditions by the use of the transient FRAPTRAN code (1.2 version). The CIP01 test consisted in a power pulse of 32.4 ms with a deposited energy of 98.2 cal/g at PPN (Peak Power Node). The code results will be presented and compared to the experimental data available. In addition, the FRAPTRAN and the SCANAIR code estimations will be also contrasted. This work is framed within the CSN-CIEMAT agreement on 'Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup'. (Author)
Startup of “CANDLE” burnup in a Gas-cooled Fast Reactor using Monte Carlo method
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Highlights: ► In equilibrium state of a CANDLE core, the burning region contains fission products and actinides. ► These isotopes are not available. The solution is startup of a reactor using easily available materials. ► At the end of core life the fuel for the equilibrium core is produced. ► In this work the startup of a CANDLE-GFR has been evaluated using Monte Carlo technique. ► The results show that the equilibrium state could be achieved after some minor transients. -- Abstract: During the past decade, the CANDLE burnup strategy has been proposed as an innovative fuel cycle and reactor design for complete utilization of uranium resources. In this strategy the shapes of neutron flux, nuclide densities and power density distribution remain constant but the burning region moves in axial direction. The feasibility of this strategy has been demonstrated widely by using the diffusion technique in conjunction with nuclide transmutation equations. On the other hand since the Monte Carlo method provides the exact solution to the neutron transport, the Monte Carlo technique is becoming more widely used in routine burnup calculations. The main objective of this work is startup of CANDLE burnup in a Gas cooled Fast Reactor using a Monte Carlo burnup scheme. In this case only natural or depleted uranium is required for fresh fuel region. However, the construction of the first CANDLE core is faced with a big problem. In equilibrium state the burning region contains a spectrum of fission products as well as higher actinides. These isotopes are not easily available for constructing the initial CANDLE core. The solution is startup of a special reactor using the enriched uranium in starter zone. At the end of core life the fuel for the next core is produced with the composition close to the equilibrium state. An originally MCNP–ORIGEN linkage program named MOBC has been used for criticality and isotopic evaluation of the core. The results of analysis showed that
MORSE Monte Carlo radiation transport code system
International Nuclear Information System (INIS)
This report is an addendum to the MORSE report, ORNL-4972, originally published in 1975. This addendum contains descriptions of several modifications to the MORSE Monte Carlo Code, replacement pages containing corrections, Part II of the report which was previously unpublished, and a new Table of Contents. The modifications include a Klein Nishina estimator for gamma rays. Use of such an estimator required changing the cross section routines to process pair production and Compton scattering cross sections directly from ENDF tapes and writing a new version of subroutine RELCOL. Another modification is the use of free form input for the SAMBO analysis data. This required changing subroutines SCORIN and adding new subroutine RFRE. References are updated, and errors in the original report have been corrected
Burnup span sensitivity analysis of different burnup coupling schemes
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Highlights: ► The objective of this work is the burnup span sensitivity analysis of different coupling schemes. ► Three kinds of schemes have been implemented in a new MCNP–ORIGEN linkage program. ► Two kinds of schemes are based predictor–corrector technique and the third is based on Euler explicit method. ► The analysis showed that the predictor–corrector approach better accounts for nonlinear behavior of burnup. ► It is sufficiently good to use the Euler method at small spans but for large spans use of second order scheme is mandatory. - Abstract: The analysis of core composition changes is complicated by the fact that the time and spatial variations in isotopic composition depend on the neutron flux distribution and vice versa. Fortunately, changes in core composition occur relatively slowly and hence the burnup analysis can be performed by dividing the burnup period into some burnup spans and assuming that the averaged flux and cross sections are constant during each burn up span. The burnup span sensitivity analysis attempts to find how much the burnup spans could be increased without any significant change in results. This goal has been achieved by developing a new MCNP–ORIGEN linkage program named MOBC (MCNP–ORIGEN Burnup Calculation). Three kinds of coupling scheme have been implemented in MOBC. Two of these are based on second order predictor–corrector technique and enable us to choose larger time steps, whilst the third one is based on Euler explicit first order method and is faster than the other two. The validity of the developed program has been evaluated by the code vs. code comparison technique. Two different types of codes are employed. The first one is based on deterministic two dimensional transport method, like CASMO-4 and HELIOS codes, and the second one is based on Monte Carlo method, like MCODE code. Only one coupling technique is employed in each of these state of the art codes, while the MOBC excels in its ability to
Development of a code and models for high burnup fuel performance analysis
International Nuclear Information System (INIS)
First the high burnup LWR fuel behavior is discussed and necessary models for the analysis are reviewed. These aspects of behavior are the changes of power history due to the higher enrichment, the temperature feedback due to fission gas release and resultant degradation of gap conductance, axial fission gas transport in fuel free volume, fuel conductivity degradation due to fission product solution and modification of fuel micro-structure. Models developed for these phenomena, modifications in the code, and the benchmark results mainly based on Risoe fission gas project is presented. Finally the rim effect which is observe only around the fuel periphery will be discussed focusing into the fuel conductivity degradation and swelling due to the porosity development. (author). 18 refs, 13 figs, 3 tabs
International Nuclear Information System (INIS)
The predictions of the ENIGMA code have been compared with data from high burn-up fuel experiments from the Halden and RISO reactors. The experiments modelled were IFA-504 and IFA-558 from Halden and the test II-5 from the RISO power burnup test series. The code has well modelled the fuel thermal performance and has provided a good measure of iodine release from pre-interlinked fuel. After interlinkage the iodine predictions remain a good fit for one experiment, but there is significant overprediction for a second experiment (IFA-558). Stable fission gas release is also well modelled and the predictions are within the expected uncertainly band throughout the burn-up range. This report presents code predictions for stable fission gas release to 40GWd/tU, iodine release measurements to 50GWd/tU and thermal performance (fuel centre temperature) to 55GWd/tU. Fuel ratings of up to 38kW/m were modelled at the high burn-up levels. The code is shown to accurately or conservatively predict all these parameters. (author). 1 ref., 6 figs
Criticality qualification of a new Monte Carlo code for reactor core analysis
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In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.
On the use of SERPENT Monte Carlo code to generate few group diffusion constants
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Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)
Criticality qualification of a new Monte Carlo code for reactor core analysis
Energy Technology Data Exchange (ETDEWEB)
Catsaros, N. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Gaveau, B. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Jaekel, M. [Laboratoire de Physique Theorique, Ecole Normale Superieure, 24 rue Lhomond, 75231 Paris (France); Maillard, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); CNRS-IDRIS, Bt 506, BP167, 91403 Orsay (France); CNRS-IN2P3, 3 rue Michel Ange, 75794 Paris (France); Maurel, G. [Faculte de Medecine, Universite Paris VI, 27 rue de Chaligny, 75012 Paris (France); MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Savva, P., E-mail: savvapan@ipta.demokritos.g [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece); Silva, J. [MAPS, Universite Paris VI, 4 Place Jussieu, 75005 Paris (France); Varvayanni, M.; Zisis, Th. [Institute of Nuclear Technology - Radiation Protection, NCSR ' DEMOKRITOS' , P.O. Box 60228, 15310 Aghia Paraskevi (Greece)
2009-11-15
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.
International Nuclear Information System (INIS)
The high burnup database for CANDU fuel includes several cases from both power station and experimental reactor irradiations, with achieved burnups of up to 800 MW.h/kgU. The power history for each of these cases is different, encompassing low steady-state, declining, and power-ramps. This variety offers a good opportunity to check the models of fuel behaviour, and to identify areas for improvement. The main parameters for comparing calculated versus measured data are the fission gas release and the sheath hoop strain. Good agreement of calculated values of these two parameters with experimental data indicates that the global behaviour of the fuel element is adequately simulated by our codes. The ELESIM computer code was used as the simulation tool. The models for fission gas release, swelling and for fuel pellet expansion were thoroughly analysed. Changes were proposed for both models. The fuel pellet expansion model was modified to account for gaseous swelling, which becomes very important at high burnups. As well, the mathematics of the fission gas release model was upgraded for the diffusional release of fission gas atoms to the grain boundaries. A revised version of the ELESIM computer code was used to simulate the cases from the high burnup database. Satisfactory agreement was found for most cases. The discrepancies are discussed in view of alternative mechanisms that can operate and be enhanced at high burnup. These include stoichiometry changes with burnup that affects fission gas release, and also outer pellet rim fission gas release by a grain boundary diffusion process. The main conclusion of this study is that the revised version of the ELESIM code is able to simulate with reasonable accuracy high burnup as well as low burnup CANDU fuel. This includes irradiations of steady-state, declining, or ramped fuel power histories with a prolonged hold at high power. However, future improvements to ELESIM are needed to model fuel power histories with short dwell
A PWR PCI failure criterion to burnups of 60 GW·d/t using the ENIGMA code
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A fuel performance modelling code (ENIGMA) has been used to analyse the empirical PCI failure criterion in terms of a clad failure stress as a function of burnup and fast neutron dose. The Studsvik database has been analysed. Results indicate a rising and then saturating failure stress with burnup and fast neutron dose. Using the PCI failure limits, equivalent to 95/95 confidence limits, an ENIGMA stress-based methodology is used to derive PWR PCI failure limits up to 60 GW·d/t U using a conservative assumption that the failure stress does not increase at high burnup and neutron dose. In addition experimental ramp data on gadolinia-doped fuel rods do not indicate any increased susceptibility to PCI failure implying that the UO2 criterion can be used for gadolinia doped fuel. (author)
International Nuclear Information System (INIS)
By coupling a three-dimensional Discrete Ordinate (SN) code with an Exponential Euler Method inventory code, a three-dimensional code system was developed for accurate and rapid fuel burnup and material activation analyses. Resonance self-shielding effect was corrected dynamically in each step and each fuel zone. The update of compositions and nuclide densities in transport calculation was treated by setting nuclide density limitation, keeping important nuclides and using pseudo fission products (FPeff). The validation was described, which illustrates the robustness and reliability of this code system. (authors)
Ramamoorthy, Karthikeyan
The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant
Fast code for Monte Carlo simulations
International Nuclear Information System (INIS)
A computer code to generate the dynamic evolution of the Ising model on a square lattice, following the Metropolis algorithm is presented. The computer time consumption is reduced by a factor of 8 when one compares our code with traditional multiple spin codes. The memory allocation size is also reduced by a factor of 4. The code is easily generalizable for other lattices and models. (author)
MORET: Version 4.B. A multigroup Monte Carlo criticality code
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MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)
MOx benchmark calculations by deterministic and Monte Carlo codes
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Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.
A New Monte Carlo Neutron Transport Code at UNIST
International Nuclear Information System (INIS)
Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results
Benchmarking Monte Carlo codes for criticality safety using subcritical measurements
International Nuclear Information System (INIS)
Monte Carlo codes that are used for criticality safety evaluations are typically validated using critical experiments in which the neutron multiplication factor is unity. However, the conditions for most fissile material operations do not coincide to those of the critical experiments. This paper demonstrates that Monte Carlo methods and nuclear data can be validated using subcritical measurements whose conditions may coincide more closely to actual configurations of fissile material. (orig.)
Design of shielding of LILW containers by Monte Carlo codes
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Accurate prediction of dose rates from containers with radioactive waste is becoming more important regarding more rigorous regulative in this area. The usual approach to the problem consists in combining numerical and measuring methods. In this paper a Monte Carlo calculations were used for calculating doses from a standard 200 liter drum which contains the intermediate level radioactive waste. Two different Monte Carlo codes were applied and compared, for the same combination of parameters. (author)
Verification of the spectral history correction method with fully coupled Monte Carlo code BGCore
International Nuclear Information System (INIS)
Recently, a new method for accounting for burnup history effects on few-group cross sections was developed and implemented in the reactor dynamic code DYN3D. The method relies on the tracking of the local Pu-239 density which serves as an indicator of burnup spectral history. The validity of the method was demonstrated in PWR and VVER applications. However, the spectrum variation in BWR core is more pronounced due to the stronger coolant density change. Therefore, the purpose of the current work is to further investigate the applicability of the method to BWR analysis. The proposed methodology was verified against recently developed BGCore system, which couples Monte Carlo neutron transport with depletion and thermal hydraulic solvers and thus capable of providing a reference solution for 3D simulations. The results dearly show that neglecting the spectral history effects leads to a very large deviation (e.g. 2000 pcm in reactivity) from fee reference solution. However, a very good agreement between DYN3D and BGCore is observed (on the order of 200 pcm in reactivity), when the. Pu-correction method is applied. (author)
Current status of the PSG Monte Carlo neutron transport code
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PSG is a new Monte Carlo neutron transport code, developed at the Technical Research Centre of Finland (VTT). The code is mainly intended for fuel assembly-level reactor physics calculations, such as group constant generation for deterministic reactor simulator codes. This paper presents the current status of the project and the essential capabilities of the code. Although the main application of PSG is in lattice calculations, the geometry is not restricted in two dimensions. This paper presents the validation of PSG against the experimental results of the three-dimensional MOX fuelled VENUS-2 reactor dosimetry benchmark. (authors)
Vectorization techniques for neutron transport Monte Carlo codes
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Four Monte Carlo codes, KENO IV, MORSE-DD, MCNP and VIM, have been vectorized already at JAERI Computing Center aiming at an increase in clculation performance, and speed-up ratios of vectorized codes to the original ones were found to be low values between 1.3 and 1.5. In this report the vectorization processes for these four codes are reviewed comprehensively, and methods of analysis for vectorization, modification of control structures of codes and debugging techniques are discussed. The reason for low speed-up ratios is also discussed. (author)
A Burnup Analysis of PBMR-400MWth Reactor Core
International Nuclear Information System (INIS)
The purpose of this study is to analyze the burnup characteristics of 400MWth PBMR using Monte Carlo method. In the world, the deterministic method is widely used to model such that system but it still has a disadvantage which is not flexible in simulating the burnup cycle. Although this method applies some techniques to increase the accuracy of calculation results but it is necessary to model this system by a suitable computer code that can verify and validate the results of the deterministic method. A method which uses a Monte Carlo technique for simulating the burnup cycle was performed. A reactor physics computer code uses in this method is MONTEBURN 2.0 which accurately and efficiently computes the neutronic and material properties of the fuel cycle. MONTEBURN is a fully automated tool that links the MCNP Monte Carlo transport code with a radioactive decay and burnup code ORIGEN. In this model, the calculations are based on a detailed core modeling using MCNP. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and fuel kernels in the pebble. For the burnup model, a start-up core was studied with considering the movement of pebbles. By shifting down one layer at each discrete time step and inserting fresh fuel from the top, this cyclic calculation is continued until equilibrium burnup cycle is achieved. In this study, the time dependence of multiplication factor keff, the spatial dependence of flux profile, power distribution, burnup, and inventory of isotopes in the start up process are analyzed. The results will provide the basis data of the burnup process and be also utilized as the verified data to validate a compute code for PBMR core analysis which will be developed in near future
A Monte Carlo code for ion beam therapy
Anaïs Schaeffer
2012-01-01
Initially developed for applications in detector and accelerator physics, the modern Fluka Monte Carlo code is now used in many different areas of nuclear science. Over the last 25 years, the code has evolved to include new features, such as ion beam simulations. Given the growing use of these beams in cancer treatment, Fluka simulations are being used to design treatment plans in several hadron-therapy centres in Europe. Fluka calculates the dose distribution for a patient treated at CNAO with proton beams. The colour-bar displays the normalized dose values. Fluka is a Monte Carlo code that very accurately simulates electromagnetic and nuclear interactions in matter. In the 1990s, in collaboration with NASA, the code was developed to predict potential radiation hazards received by space crews during possible future trips to Mars. Over the years, it has become the standard tool to investigate beam-machine interactions, radiation damage and radioprotection issues in the CERN accelerator com...
Development of a New Monte Carlo reactor physics code
Leppänen, Jaakko
2007-01-01
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to so...
The Monte Carlo code TRAMO - Capabilities and instructions for application
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The report is intended for readers familiar with the fundamentals of the Monte Carlo method. Those readers might be interested in learning about successful generalisations as well as new ideas for curbing the statistical errors involved. Another intention however is to explain the significant basic features of the multigroup Monte Carlo code TRAMO, including the required input, so that readers will be able to performing the required adjustments to the specific calculation technique and develop their own tools for performing their specific calculations. An indispensable code needed for such TRAMO applications is the TRAWEI Monte Carlo code which calculates he required weightings for applications of the variance reducing Weight Window Method; other codes required are those for generating the neutron cross-section data and the group data. The TRAMO code calculates, with given source distribution of neutrons in multigroup approximation, multigroup flux data, integrated group flux data, and dose values for given partial volumes and surfaces. There are further code versions for calculation of neutron and gamma fluxes, or criticality data, but these are not considered in the report. (orig./CB)
MORSE Monte Carlo radiation transport code system
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For a number of years the MORSE user community has requested additional help in setting up problems using various options. The sample problems distributed with MORSE did not fully demonstrate the capability of the code. At Oak Ridge National Laboratory the code originators had a complete set of sample problems, but funds for documenting and distributing them were never available. Recently the number of requests for listings of input data and results for running some particular option the user was trying to implement has increased to the point where it is not feasible to handle them on an individual basis. Consequently it was decided to package a set of sample problems which illustrates more adequately how to run MORSE. This write-up may be added to Part III of the MORSE report. These sample problems include a combined neutron-gamma case, a neutron only case, a gamma only case, an adjoint case, a fission case, a time-dependent fission case, the collision density case, an XCHEKR run and a PICTUR run
Taylor series development in the Monte Carlo code Tripoli-4
Mazzolo, Alain; Zoia, Andrea; Martin, Brunella
2014-06-01
Perturbation methods for one or several variables based on the Taylor series development up to the second order is presented for the collision estimator in the framework of the Monte Carlo code Tripoli-4. Comparisons with the correlated sampling method implemented in Tripoli-4 demonstrate the need of including the cross derivatives in the development.
Energy Technology Data Exchange (ETDEWEB)
Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik
2014-06-15
In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
International Nuclear Information System (INIS)
The effect of fuel burn-up on the radioactivation behavior of cladding hull materials was investigated using the ORIGEN-S code for various materials of Zircaloy-4, Zirlo, HANA-4, and HANA-6 and for various fuel burn-ups of 30, 45, 60, and 75 GWD/MTU. The Zircaloy-4 material is the only one that does not contain Nb as an alloy constituent, and it was revealed that 125Sb, 125mTe, and 55Fe are the major sources of radioactivity. On the other hand, 93mNb was identified as the most radioactive nuclide for the other materials although minor radioactive nuclides varied owing to their different initial constituents. The radioactivity of 94Nb was of particular focus owing to its acceptance limit against a Korean intermediate-/low-level waste repository. The radioactivation calculation results revealed that only Zircaloy-4 is acceptable for the Korean repository, while the other materials required at least 4,900 of Nb decontamination factor owing to the high radioactivity of 94Nb regardless of the fuel burn-up. A discussion was also made on the feasibility of Zr recovery methods (chlorination and electrorefining) for selective recovery of Zr so that it can be disposed of in the Korean repository. (author)
Acceleration of a Monte Carlo radiation transport code
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Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics
Study on random number generator in Monte Carlo code
International Nuclear Information System (INIS)
The Monte Carlo code uses a sequence of pseudo-random numbers with a random number generator (RNG) to simulate particle histories. A pseudo-random number has its own period depending on its generation method and the period is desired to be long enough not to exceed the period during one Monte Carlo calculation to ensure the correctness especially for a standard deviation of results. The linear congruential generator (LCG) is widely used as Monte Carlo RNG and the period of LCG is not so long by considering the increasing rate of simulation histories in a Monte Carlo calculation according to the remarkable enhancement of computer performance. Recently, many kinds of RNG have been developed and some of their features are better than those of LCG. In this study, we investigate the appropriate RNG in a Monte Carlo code as an alternative to LCG especially for the case of enormous histories. It is found that xorshift has desirable features compared with LCG, and xorshift has a larger period, a comparable speed to generate random numbers, a better randomness, and good applicability to parallel calculation. (author)
A semianalytic Monte Carlo code for modelling LIDAR measurements
Palazzi, Elisa; Kostadinov, Ivan; Petritoli, Andrea; Ravegnani, Fabrizio; Bortoli, Daniele; Masieri, Samuele; Premuda, Margherita; Giovanelli, Giorgio
2007-10-01
LIDAR (LIght Detection and Ranging) is an optical active remote sensing technology with many applications in atmospheric physics. Modelling of LIDAR measurements appears useful approach for evaluating the effects of various environmental variables and scenarios as well as of different measurement geometries and instrumental characteristics. In this regard a Monte Carlo simulation model can provide a reliable answer to these important requirements. A semianalytic Monte Carlo code for modelling LIDAR measurements has been developed at ISAC-CNR. The backscattered laser signal detected by the LIDAR system is calculated in the code taking into account the contributions due to the main atmospheric molecular constituents and aerosol particles through processes of single and multiple scattering. The contributions by molecular absorption, ground and clouds reflection are evaluated too. The code can perform simulations of both monostatic and bistatic LIDAR systems. To enhance the efficiency of the Monte Carlo simulation, analytical estimates and expected value calculations are performed. Artificial devices (such as forced collision, local forced collision, splitting and russian roulette) are moreover foreseen by the code, which can enable the user to drastically reduce the variance of the calculation.
ABRAC: A microcomputer-based Fortran code for multi-cyle burnup
International Nuclear Information System (INIS)
Pressurized-water reactors have reactor physics and fuel management characteristics which are very amenable to simplified analysis. Given models which account for the dominant features of core and fuel performance, it is possible to rapidly perform relatively accurate scoping studies of many years of reactor operation in just a few hours on a modern (386-class) microcomputer. Models are described for burnup-dependent cross-section generation, for burnup of fuel under irradiation, and for computation of radial power distributions in hexagonal geometry assuming hexagonal fuel assemblies. Comparisons with more elaborate methods are given in order to validate the models, and to quantify the accuracy of the results. 16 refs., 5 figs., 5 tabs
International Nuclear Information System (INIS)
PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO2, UO2-Gd2O3, inhomogeneous MOX, and UO2-ThO2. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of 92U233-239, 93Np237-239, 94Pu238-243, 95Am241-244 (including isomers), and 96Cm242-245. Poisoning fission products are represented by 54Xe131,133,135, 48Cd113, 62Sm149,151,152, 64Gd154-160, 63Eu153,155, 36Kr83,85, 42Mo95, 43Tc99, 45Rh103, 47Ag109, 53I127,129,131, 55Cs133, 57La139, 59Pr141, 60Nd143-150, 61Pm147. Fission gases and volatiles included in the code are 36Kr83-86, 54Xe129-136, 52Te125-130, 53I127-131, 55Cs133-137, and 56Ba135-140. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)
TRIPOLI-3: a neutron/photon Monte Carlo transport code
International Nuclear Information System (INIS)
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.)
SPAMCART: a code for smoothed particle Monte Carlo radiative transfer
Lomax, O
2016-01-01
We present a code for generating synthetic SEDs and intensity maps from Smoothed Particle Hydrodynamics simulation snapshots. The code is based on the Lucy (1999) Monte Carlo Radiative Transfer method, i.e. it follows discrete luminosity packets, emitted from external and/or embedded sources, as they propagate through a density field, and then uses their trajectories to compute the radiative equilibrium temperature of the ambient dust. The density is not mapped onto a grid, and therefore the calculation is performed at exactly the same resolution as the hydrodynamics. We present two example calculations using this method. First, we demonstrate that the code strictly adheres to Kirchhoff's law of radiation. Second, we present synthetic intensity maps and spectra of an embedded protostellar multiple system. The algorithm uses data structures that are already constructed for other purposes in modern particle codes. It is therefore relatively simple to implement.
ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®
Damian, F.; Brun, E.
2014-06-01
ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.
International Nuclear Information System (INIS)
Highlights: • Coupling of Monte Carlo code Serpent and thermal–hydraulics code RELAP5. • A convergence criterion is developed based on the statistical uncertainty of power. • Correlation between MC statistical uncertainty and coupled error is quantified. • Both UO2 and MOX single assembly models are used in the coupled simulation. • Validation of coupling results with a multi-group transport code DeCART. - Abstract: Coupled multi-physics approach plays an important role in improving computational accuracy. Compared with deterministic neutronics codes, Monte Carlo codes have the advantage of a higher resolution level. In the present paper, a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, Serpent, is coupled with a thermal–hydraulics safety analysis code, RELAP5. The coupled Serpent/RELAP5 code capability is demonstrated by the improved axial power distribution of UO2 and MOX single assembly models, based on the OECD-NEA/NRC PWR MOX/UO2 Core Transient Benchmark. Comparisons of calculation results using the coupled code with those from the deterministic methods, specifically heterogeneous multi-group transport code DeCART, show that the coupling produces more precise results. A new convergence criterion for the coupled simulation is developed based on the statistical uncertainty in power distribution in the Monte Carlo code, rather than ad-hoc criteria used in previous research. The new convergence criterion is shown to be more rigorous, equally convenient to use but requiring a few more coupling steps to converge. Finally, the influence of Monte Carlo statistical uncertainty on the coupled error of power and thermal–hydraulics parameters is quantified. The results are presented such that they can be used to find the statistical uncertainty to use in Monte Carlo in order to achieve a desired precision in coupled simulation
Taking burnup credit for interim storage and transportation system for BWR fuels
International Nuclear Information System (INIS)
In the back-end issues of nuclear fuel cycle, selection of reprocessing or one-through is a big issue. For both of the cases, a reasonable interim storage and transportation system is required. This study proposes an advanced practical monitoring and evaluation system. The system features the followings: (l) Storage racks and transportation casks taking credit for burnup. (2) A burnup estimation system using a compact monitor with Cd- Te detectors and fission chambers. (3) A neutron emission-rate evaluation methodology, especially important for high burnup MOX fuels. (4) A nuclear materials management system for safeguards. Current storage system and transport casks are designed on the basis of a fresh fuel assumption. The assumption is too conservative. Taking burnup credit gives a reasonable design while keeping conservatism. In order to establish a reasonable burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of some modules such as TGBLA, ORIGEN, CITATION, MCNP and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. The code takes operational history such as, power density, void fraction into account. This code is applied to the back-end issues for a more accurate design of a storage and a transportation system. The ORIGEN code is well-known one-point isotope depletion code. In the calculation system, the code calculates isotope compositions using libraries generated from the TGBLA code. The CITATION code, the MCNP code, and the KENO code are three dimensional diffusion code, continuous energy Monte Carlo code, discrete energy Monte Carlo code, respectively. Those codes calculate k- effective of the storage and transportation systems using isotope compositions generated from the ORIGEN code. The CITATION code and the KENO code are usually used for practical designs. The MCNP code is used for reference
Verification of Monte Carlo transport codes by activation experiments
International Nuclear Information System (INIS)
With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is focused on obtaining the experimental data on activation of the targets by heavy-ion beams. Several experiments were performed at GSI Helmholtzzentrum fuer Schwerionenforschung. The interaction of nitrogen, argon and uranium beams with aluminum targets, as well as interaction of nitrogen and argon beams with copper targets was studied. After the irradiation of the targets by different ion beams from the SIS18 synchrotron at GSI, the γ-spectroscopy analysis was done: the γ-spectra of the residual activity were measured, the radioactive nuclides were identified, their amount and depth distribution were detected. The obtained experimental results were compared with the results of the Monte Carlo simulations using FLUKA, MARS and SHIELD. The discrepancies and agreements between experiment and simulations are pointed out. The origin of discrepancies is discussed. Obtained results allow for a better verification of the Monte Carlo transport codes, and also provide information for their further development. The necessity of the activation studies for accelerator applications is discussed. The limits of applicability of the heavy-ion beam-loss criteria were studied using the FLUKA code. FLUKA-simulations were done to determine the most preferable from the radiation protection point of view materials for use in accelerator components.
Burnup and plutonium distribution of WWER-440 fuel pin at extended burnup
International Nuclear Information System (INIS)
The formation of rim region in LWR UO2 based nuclear fuel at high burnup is a common observation. This region has very high porosity due to excessive gas release. Such a region is also characterized by a significantly high plutonium concentration and high local burnup compared to the internal fuel region. Spatial distribution of these parameters has been incorporated with fuel behavior and performance analysis codes by using mostly empirical relations. Variation of these parameters depends on the neutron flux as well as neutron energy spectrum. Detailed neutronics analysis is necessary for the accurate prediction of these parameters. This study is performed by MCNP4B Monte Carlo code for the calculation of local neutron flux, ORIGEN2 for burnup and depletion calculations, and MONTEBURNS for coupling these codes. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell. Fuel pin is divided into a number of radial segments. A relatively small mesh size is used at the region near the surface to reveal the rim effect. The variation of plutonium and local burnup are obtained for high burnup. Results are compared with existing experimental observations for WWER-440 fuel and other theoretical predictions
Nanodosimetric verification in proton therapy: Monte Carlo Codes Comparison
International Nuclear Information System (INIS)
Full text: Nanodosimetry strives to develop a novel dosimetry concept suitable for advanced modalities of cancer radiotherapy, such as proton therapy. This project aims to evaluate the plausibility of the physical models implemented in the Geant4 Very Low Energy (Geant4-DNA) extensions by comparing nanodosimetric quantities calculated with Geant4-DNA and the PTB Monte Carlo track structure code. Nanodosimetric track structure parameters were calculated for cylindrical targets representing DNA and nucleosome segments and converted into the probability of producing a DSB using the model proposed by Garty et al. [1]. Monoenergetic protons and electrons of energies typical for 6-electron spectra were considered as primary particles. Good agreement was found between the two codes for electrons of energies above 200 eV. Below this energy Geant4-DNA produced slightly higher numbers of ionisations in the sensitive volumes and higher probabilities for DSB formation. For protons, Geant4-DNA also gave higher numbers of ionisations and DSB probabilities, particularly in the low energy range, while a satisfactory agreement was found for energies higher than I MeV. Comparing two codes can be useful as any observed divergence in results between the two codes provides valuable information as to where further consideration of the underlying physical models used in each code may be required. Consistently it was seen that the largest difference between the codes was in the low energy ranges for each particle type. (author)
Proton therapy Monte Carlo SRNA-VOX code
Directory of Open Access Journals (Sweden)
Ilić Radovan D.
2012-01-01
Full Text Available The most powerful feature of the Monte Carlo method is the possibility of simulating all individual particle interactions in three dimensions and performing numerical experiments with a preset error. These facts were the motivation behind the development of a general-purpose Monte Carlo SRNA program for proton transport simulation in technical systems described by standard geometrical forms (plane, sphere, cone, cylinder, cube. Some of the possible applications of the SRNA program are: (a a general code for proton transport modeling, (b design of accelerator-driven systems, (c simulation of proton scattering and degrading shapes and composition, (d research on proton detectors; and (e radiation protection at accelerator installations. This wide range of possible applications of the program demands the development of various versions of SRNA-VOX codes for proton transport modeling in voxelized geometries and has, finally, resulted in the ISTAR package for the calculation of deposited energy distribution in patients on the basis of CT data in radiotherapy. All of the said codes are capable of using 3-D proton sources with an arbitrary energy spectrum in an interval of 100 keV to 250 MeV.
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This paper describes results of testing of the TRANSURANUS burn-up model (TUBRNP routine) for Gd-doped WWER-1000 fuel pin based on results of HELIOS code. The testing covers the analysis of different types of nuclear fuel rods from a neutronic point of view that one can encounter in the VVER-1000 reactor core. The HELIOS computations simulate the assembly geometry, and combine 4 different 235U enrichment configurations with 4 different Gd2O3-concentrations. For each of these combinations the radial distribution of the concentrations of 155Gd and 157Gd compute in one Gd-doped rod. Based on these results the recommendations on using cross section of Gd in TRANSURANUS TUBRNP model were proposed. (author)
Computed radiography simulation using the Monte Carlo code MCNPX
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Simulating x-ray images has been of great interest in recent years as it makes possible an analysis of how x-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data. (author)
Adjoint Monte Carlo techniques and codes for organ dose calculations
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Adjoint Monte Carlo simulations can be effectively used for the estimation of doses in small targets when the sources are extended in large volumes or surfaces. The main features of two computer codes for calculating doses at free points or in organs of an anthropomorphic phantom are described. In the first program (REBEL-3) natural gamma-emitting sources are contained in the walls of a dwelling room; in the second one (POKER-CAMP) the user can specify arbitrary gamma sources with different spatial distributions in the environment: in (or on the surface of) the ground and in the air. 3 figures
International Nuclear Information System (INIS)
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
International Nuclear Information System (INIS)
Highlights: → The specifications required for the analyses of the destructive assay data taken from irradiated fuel in Ohi-1 and Ohi-2 PWRs were documented in this paper. → These data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. → These destructive assay data are suitable for the benchmarking of the burnup calculation code systems. - Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.
Energy Technology Data Exchange (ETDEWEB)
Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-08-01
PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)
Vectorization of continuous energy Monte Carlo code VIM
International Nuclear Information System (INIS)
VIM is a continuous energy Monte Carlo code for criticality calculation. The random walk control system which uses combinatorial geometry system has been vectorized on FACOM VP-100. Vectorization has been done by the event bank method which controls simultaneous multiple particle's random walks, since behavior of neutron is independent. In vectorization of VIM code, we have two problems. One is a large overhead introduced by program modifications for vectorization. Another is a lowering of vector processing efficiency, since the vector length decreases with time according to the absorption and leakage of neutron and cut off of neutron for variance reduction. The average vector length during the random walks has been kept long by utilizing cross section library of single energy band and by reducing the number of the event banks. The performance ratio of vectorized version to the original one is 1.39 for the simple geometry and 1.13 for the complex geometry. (author)
Parallel computing by Monte Carlo codes MVP/GMVP
International Nuclear Information System (INIS)
General-purpose Monte Carlo codes MVP/GMVP are well-vectorized and thus enable us to perform high-speed Monte Carlo calculations. In order to achieve more speedups, we parallelized the codes on the different types of parallel computing platforms or by using a standard parallelization library MPI. The platforms used for benchmark calculations are a distributed-memory vector-parallel computer Fujitsu VPP500, a distributed-memory massively parallel computer Intel paragon and a distributed-memory scalar-parallel computer Hitachi SR2201, IBM SP2. As mentioned generally, linear speedup could be obtained for large-scale problems but parallelization efficiency decreased as the batch size per a processing element(PE) was smaller. It was also found that the statistical uncertainty for assembly powers was less than 0.1% by the PWR full-core calculation with more than 10 million histories and it took about 1.5 hours by massively parallel computing. (author)
Parallelization of a Monte Carlo particle transport simulation code
Hadjidoukas, P.; Bousis, C.; Emfietzoglou, D.
2010-05-01
We have developed a high performance version of the Monte Carlo particle transport simulation code MC4. The original application code, developed in Visual Basic for Applications (VBA) for Microsoft Excel, was first rewritten in the C programming language for improving code portability. Several pseudo-random number generators have been also integrated and studied. The new MC4 version was then parallelized for shared and distributed-memory multiprocessor systems using the Message Passing Interface. Two parallel pseudo-random number generator libraries (SPRNG and DCMT) have been seamlessly integrated. The performance speedup of parallel MC4 has been studied on a variety of parallel computing architectures including an Intel Xeon server with 4 dual-core processors, a Sun cluster consisting of 16 nodes of 2 dual-core AMD Opteron processors and a 200 dual-processor HP cluster. For large problem size, which is limited only by the physical memory of the multiprocessor server, the speedup results are almost linear on all systems. We have validated the parallel implementation against the serial VBA and C implementations using the same random number generator. Our experimental results on the transport and energy loss of electrons in a water medium show that the serial and parallel codes are equivalent in accuracy. The present improvements allow for studying of higher particle energies with the use of more accurate physical models, and improve statistics as more particles tracks can be simulated in low response time.
A comparison between the Monte Carlo radiation transport codes MCNP and MCBEND
Energy Technology Data Exchange (ETDEWEB)
Sawamura, Hidenori; Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)
2001-01-01
In Japan, almost of all radiation analysts are using the MCNP code and MVP code on there studies. But these codes have not had automatic variance reduction. MCBEND code made by UKAEA have automatic variance reduction. And, MCBEND code is user friendly more than other Monte Carlo Radiation Transport Codes. Our company was first introduced MCBEND code in Japan. Therefore, we compared with MCBEND code and MCNP code about functions and production capacity. (author)
Burnup instabilities in the full-core HTR model simulation
International Nuclear Information System (INIS)
Highlights: • We performed full-core burnup calculation coupled with Monte Carlo code. • Depletion instabilities have been detected for HTR system at high burnup. • We assess the stability of time step models in application to core calculation. • Discussion of the modeling factors related to burnup core simulation is presented. - Abstract: The phenomenon of numerical instabilities present in the Monte Carlo burnup calculations has been shown and explained by many authors using models of LWR, often simplified. Some theoretical considerations about origins of oscillations are very general, however it may be difficult to apply it easily to other models as a prediction of stability. Physics of HTR core differs significantly from the properties of light water system and the reliable extrapolation of the current numerical results is not possible. Moreover, most of the works concerning HTR burnup calculations put no emphasis on the spatial stability of the simulation and apply very long time steps. The awareness in this field of research seems to be not sufficient. In this paper, we focus on the demonstration of depletion instabilities in the simulations of HTR core dedicated for deep burnup of plutonium and minor actinides. We apply various methodology of time step implemented in advanced Continuous Energy Monte Carlo burnup code MCB version 5. Stability analysis is very rare for the full core calculations and the awareness of the oscillation’s problem is obligatory for the reliable modeling of a fuel cycle. In the summary of this work we systematize and discuss factors related to the stability of depletion and review available solutions
On the use of the Serpent Monte Carlo code for few-group cross section generation
International Nuclear Information System (INIS)
Research highlights: → B1 methodology was used for generation of leakage-corrected few-group cross sections in the Serpent Monte-Carlo code. → Few-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. → 3D analysis of a PWR core was performed by a nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. → An excellent agreement in the results of 3D core calculations obtained with Helios and Serpent generated cross-section libraries was observed. - Abstract: Serpent is a recently developed 3D continuous-energy Monte Carlo (MC) reactor physics burnup calculation code. Serpent is specifically designed for lattice physics applications including generation of homogenized few-group constants for full-core core simulators. Currently in Serpent, the few-group constants are obtained from the infinite-lattice calculations with zero neutron current at the outer boundary. In this study, in order to account for the non-physical infinite-lattice approximation, B1 methodology, routinely used by deterministic lattice transport codes, was considered for generation of leakage-corrected few-group cross sections in the Serpent code. A preliminary assessment of the applicability of the B1 methodology for generation of few-group constants in the Serpent code was carried out according to the following steps. Initially, the two-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. Then, a 3D analysis of a Pressurized Water Reactor (PWR) core was performed by the nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. At this stage thermal-hydraulic (T-H) feedback was neglected. The DYN3D results were compared with those obtained from the 3D full core Serpent MC calculations. Finally, the full core DYN3D calculations were repeated taking into account T-H feedback and
Monte Carlo simulation of medical linear accelerator using primo code
International Nuclear Information System (INIS)
The use of monte Carlo simulation has become very important in the medical field and especially in calculation in radiotherapy. Various Monte Carlo codes were developed simulating interactions of particles and photons with matter. One of these codes is PRIMO that performs simulation of radiation transport from the primary electron source of a linac to estimate the absorbed dose in a water phantom or computerized tomography (CT). PRIMO is based on Penelope Monte Carlo code. Measurements of 6 MV photon beam PDD and profile were done for Elekta precise linear accelerator at Radiation and Isotopes Center Khartoum using computerized Blue water phantom and CC13 Ionization Chamber. accept Software was used to control the phantom to measure and verify dose distribution. Elektalinac from the list of available linacs in PRIMO was tuned to model Elekta precise linear accelerator. Beam parameter of 6.0 MeV initial electron energy, 0.20 MeV FWHM, and 0.20 cm focal spot FWHM were used, and an error of 4% between calculated and measured curves was found. The buildup region Z max was 1.40 cm and homogenous profile in cross line and in line were acquired. A number of studies were done to verily the model usability one of them is the effect of the number of histories on accuracy of the simulation and the resulted profile for the same beam parameters. The effect was noticeable and inaccuracies in the profile were reduced by increasing the number of histories. Another study was the effect of Side-step errors on the calculated dose which was compared with the measured dose for the same setting.It was in range of 2% for 5 cm shift, but it was higher in the calculated dose because of the small difference between the tuned model and measured dose curves. Future developments include simulating asymmetrical fields, calculating the dose distribution in computerized tomographic (CT) volume, studying the effect of beam modifiers on beam profile for both electron and photon beams.(Author)
KAMCCO, a reactor physics Monte Carlo neutron transport code
International Nuclear Information System (INIS)
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.)
Verification of Monte Carlo transport codes FLUKA, Mars and Shield
International Nuclear Information System (INIS)
The present study is a continuation of the project 'Verification of Monte Carlo Transport Codes' which is running at GSI as a part of activation studies of FAIR relevant materials. It includes two parts: verification of stopping modules of FLUKA, MARS and SHIELD-A (with ATIMA stopping module) and verification of their isotope production modules. The first part is based on the measurements of energy deposition function of uranium ions in copper and stainless steel. The irradiation was done at 500 MeV/u and 950 MeV/u, the experiment was held at GSI from September 2004 until May 2005. The second part is based on gamma-activation studies of an aluminium target irradiated with an argon beam of 500 MeV/u in August 2009. Experimental depth profiling of the residual activity of the target is compared with the simulations. (authors)
Monte Carlo Code System Development for Liquid Metal Reactor
Energy Technology Data Exchange (ETDEWEB)
Kim, Chang Hyo; Shim, Hyung Jin; Han, Beom Seok; Park, Ho Jin; Park, Dong Gyu [Seoul National University, Seoul (Korea, Republic of)
2007-03-15
We have implemented the composition cell class and the use cell to MCCARD for hierarchy input processing. For the inputs of KALlMER-600 core consisted of 336 assemblies, we require the geometric data of 91,056 pin cells. Using hierarchy input processing, it was observed that the system geometries are correctly handled with the geometric data of total 611 cells; 2 cells for fuel rods, 2 cells for guide holes, 271 translation cells for rods, and 336 translation cells for assemblies. We have developed monte carlo decay-chain models based on decay chain model of REBUS code for liquid metal reactor analysis. Using developed decay-chain models, the depletion analysis calculations have performed for the homogeneous and heterogeneous model of KALlMER-600. The k-effective for the depletion analysis agrees well with that of REBUS code. and the developed decay chain models shows more efficient performance for time and memories, as compared with the existing decay chain model The chi-square criterion has been developed to diagnose the temperature convergence for the MC TjH feedback calculations. From the application results to the KALlMER pin and fuel assembly problem, it is observed that the new criterion works well Wc have applied the high efficiency variance reduction technique by splitting Russian roulette to estimate the PPPF of the KALIMER core at BOC. The PPPF of KALlMER core at BOC is 1.235({+-}0.008). The developed technique shows four time faster calculation, as compared with the existin2 calculation Subject Keywords Monte Carlo
Detailed Burnup Calculations for Testing Nuclear Data
Leszczynski, F.
2005-05-01
A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross
Application of two-dimensional burnup computer codes to the operation of nuclear power plants
International Nuclear Information System (INIS)
The needs for three-dimensional computer calculations of the power density distribution in WWER type reactors are outlined. In most cases, however, two-dimensional calculations provide sufficiently exact results and result in a decrease in computer costs. The application, performance and computer codes of two-dimensional calculations are dealt with. (author)
A burnup credit calculation methodology for PWR spent fuel transportation
International Nuclear Information System (INIS)
A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)
A Monte Carlo track structure code for low energy protons
Endo, S; Nikjoo, H; Uehara, S; Hoshi, M; Ishikawa, M; Shizuma, K
2002-01-01
A code is described for simulation of protons (100 eV to 10 MeV) track structure in water vapor. The code simulates molecular interaction by interaction for the transport of primary ions and secondary electrons in the form of ionizations and excitations. When a low velocity ion collides with the atoms or molecules of a target, the ion may also capture or lose electrons. The probabilities for these processes are described by the quantity cross-section. Although proton track simulation at energies above Bragg peak (>0.3 MeV) has been achieved to a high degree of precision, simulations at energies near or below the Bragg peak have only been attempted recently because of the lack of relevant cross-section data. As the hydrogen atom has a different ionization cross-section from that of a proton, charge exchange processes need to be considered in order to calculate stopping power for low energy protons. In this paper, we have used state-of-the-art Monte Carlo track simulation techniques, in conjunction with the pub...
Monte Carlo simulation in UWB1 depletion code
International Nuclear Information System (INIS)
UWB1 depletion code is being developed as a fast computational tool for the study of burnable absorbers in the University of West Bohemia in Pilsen, Czech Republic. In order to achieve higher precision, the newly developed code was extended by adding a Monte Carlo solver. Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers in nuclear fuel. Burnable absorbers (BA) allow the compensation of the initial reactivity excess of nuclear fuel and result in an increase of fuel cycles lengths with higher enriched fuels. The paper describes the depletion calculations of VVER nuclear fuel doped with rare earth oxides as burnable absorber based on performed depletion calculations, rare earth oxides are divided into two equally numerous groups, suitable burnable absorbers and poisoning absorbers. According to residual poisoning and BA reactivity worth, rare earth oxides marked as suitable burnable absorbers are Nd, Sm, Eu, Gd, Dy, Ho and Er, while poisoning absorbers include Sc, La, Lu, Y, Ce, Pr and Tb. The presentation slides have been added to the article
The Monte Carlo code MCSHAPE: Main features and recent developments
International Nuclear Information System (INIS)
MCSHAPE is a general purpose Monte Carlo code developed at the University of Bologna to simulate the diffusion of X- and gamma-ray photons with the special feature of describing the full evolution of the photon polarization state along the interactions with the target. The prevailing photon–matter interactions in the energy range 1–1000 keV, Compton and Rayleigh scattering and photoelectric effect, are considered. All the parameters that characterize the photon transport can be suitably defined: (i) the source intensity, (ii) its full polarization state as a function of energy, (iii) the number of collisions, and (iv) the energy interval and resolution of the simulation. It is possible to visualize the results for selected groups of interactions. MCSHAPE simulates the propagation in heterogeneous media of polarized photons (from synchrotron sources) or of partially polarized sources (from X-ray tubes). In this paper, the main features of MCSHAPE are illustrated with some examples and a comparison with experimental data. - Highlights: • MCSHAPE is an MC code for the simulation of the diffusion of photons in the matter. • It includes the proper description of the evolution of the photon polarization state. • The polarization state is described by means of the Stokes vector, I, Q, U, V. • MCSHAPE includes the computation of the detector influence in the measured spectrum. • MCSHAPE features are illustrated with examples and comparison with experiments
Inclusion of historical dependences of fuel burn-up into MOBY-DICK Code
International Nuclear Information System (INIS)
The paper briefly describes inclusion of historical dependences of cross sections into MOBY-DICK code. Changes in program and its library are specified, especially from the point of view of programs user. Preliminary testing on assembly level is described and also testing on core level for 'transient' loading pattern calculations is demonstrated on examples of 18th and 19th cycles of the Dukovany NPP Unit III. Some features are addressed in the end of the paper (Authors)
Incorporation of the variation in conductivity with burnup in the stability of code predictive LAPUR
International Nuclear Information System (INIS)
In the field of nuclear safety, the analysis of the stability of boiling water reactors is one of the biggest challenges for researchers. LAPUR code that allows to obtain the parameters of stability of the plant (Decay rate and frequency), being one of the programs used by IBERDROLA can be used for these calculations. With the collaboration of the research group TIN of the Polytechnic University of Valencia, a model of loss of conductivity of uranium has joined with the burned LAPUR. This update allows you to play the phenomenon in a more realistic way. This improvement has been validated and verified contrasting results with reference values.
International Nuclear Information System (INIS)
Highlights: • The inventory of the radioactive nuclides was calculated using the GETERA code. • The consumptions of 235U for the HEU, MEU and LEU were: 2360, 2334 and 2320 g. • The amounts of 239Pu produced were: 67.07, 157.86 g for the MEU and LEU. • The core radioactivity for the MEU and LEU were: 8.84 × 1016 and 9.31 × 1016 Bq. - Abstract: Efforts have been made recently to study the possibility of core conversion of the 10 MW MTR type research reactor from the HEU to LEU fuels due to the proliferation issue. Since the inventory of the reactor core is a required parameter to study the atmospheric dispersion calculation for a postulated accidental airborne radionuclide release from the reactor, the inventory of the radioactive nuclides accumulated in the UAlx–Al fuels: HEU (93% 235U), MEU (45% 235U) and LEU (20% 235U) after 200 days of the reactor operating time was calculated using the GETERA code. The result showed, after 200 days of the reactor operation time (35% burnup), that the total consumptions of 235U and 238U for the HEU, MEU and LEU fuels were: 2360, 2334 and 2320 g for the 235U and 13, 105 and 238 g for the 238U, respectively. The amounts of 239Pu produced in the core were: 67.07, 157.86 g for the MEU and LEU fuels, respectively, compared with 7.95 g. The total core radioactivity after 200 days for the MEU and LEU cores were: 8.84 × 1016 and 9.31 × 1016 Bq, respectively, compared with 8.63 × 1016 Bq for the HEU core
Gholamzadeh Zohreh; Hossein Feghhi Seyed Amir; Soltani Leila; Rezazadeh Marzieh; Tenreiro Claudio; Joharifard Mahdi
2014-01-01
Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. N...
Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms
H A Nedaie; Mosleh-Shirazi, M. A.; Allahverdi, M.
2013-01-01
Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous...
Parallel implementation of the Monte Carlo transport code EGS4 on the hypercube
International Nuclear Information System (INIS)
Monte Carlo transport codes are commonly used in the study of particle interactions. The CALOR89 code system is a combination of several Monte Carlo transport and analysis programs. In order to produce good results, a typical Monte Carlo run will have to produce many particle histories. On a single processor computer, the transport calculation can take a huge amount of time. However, if the transport of particles were divided among several processors in a multiprocessor machine, the time can be drastically reduced
Directory of Open Access Journals (Sweden)
Gholamzadeh Zohreh
2014-12-01
Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view
Method of tallying adjoint fluence and calculating kinetics parameters in Monte Carlo codes
International Nuclear Information System (INIS)
A method of using iterated fission probability to estimate the adjoint fluence during particles simulation, and using it as the weighting function to calculate kinetics parameters βeff and A in Monte Carlo codes, was introduced in this paper. Implements of this method in continuous energy Monte Carlo code MCNP and multi-group Monte Carlo code MCMG are both elaborated. Verification results show that, with regardless additional computing cost, using this method, the adjoint fluence accounted by MCMG matches well with the result computed by ANISN, and the kinetics parameters calculated by MCNP agree very well with benchmarks. This method is proved to be reliable, and the function of calculating kinetics parameters in Monte Carlo codes is carried out effectively, which could be the basement for Monte Carlo codes' utility in the analysis of nuclear reactors' transient behavior. (authors)
Analysis of the burnup of the control rods with the COREMASTER-Presto code
International Nuclear Information System (INIS)
An evaluation of the capacity of the COREMASTER-Presto code, to evaluate generically the burnt of the control bars in the Laguna Verde reactors plant (CLV) is made. It was found that the code only reports burnt values of the control rods in MWD/TM, in spite of having with a second order polynomial model, for the conversion to remainder of the Boron-10 (B-10). It was observed that said model is adequate only for burnt smaller to 45,000 MWD/TM. To evaluate the burnt of the control rods it was reproduced the balance cycle of 18 months for the CLV, executing Cm-Presto during 13 consecutive cycles. First without rod burnt, taking this as the base case. Later on, cases with 1, 2 and up to 13 cycles with rod burnt were generated. When comparing results it was observed that the control rods pattern it loses reactivity lineally with the burnt one. By each 10 G Wd/T of burnt of the nucleus it is decreased the reactivity of the pattern rods ∼ 1 pcm in hot condition and of ∼ 20 pcm in cold condition. When burning three cycles those rods more burnt reached the 13,900 MWD/TM, equivalent to 36% of B-10 reduction, near value to 34% proposed by aging in the one lost study of B-10. It was observed that Cm-Presto it doesn't burn the superior node of the control rods when these are completely extracted. A one big lost of B-10, of the order of 50%, it represents only a decrease of 11% of the reactivity value of the rod. One can affirm that even when it is strongly decreased the content of B-10, the rod is continue considering as a black absorber, that is to say, thermal neutron that enters in the neutron rod that is absorbed. (Author)
Dependence of control rod worth on fuel burnup
Energy Technology Data Exchange (ETDEWEB)
Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)
2011-02-15
Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.
International Nuclear Information System (INIS)
Highlights: • Pu-239 based spectral history method was tested on 3D BWR single assembly case. • Burnup of a BWR fuel assembly was performed with the nodal code DYN3D. • Reference solution was obtained by coupled Monte-Carlo thermal-hydraulic code BGCore. • The proposed method accurately reproduces moderator density history effect for BWR test case. - Abstract: This research focuses on the verification of a recently developed methodology accounting for spectral history effects in 3D full core nodal simulations. The traditional deterministic core simulation procedure includes two stages: (1) generation of homogenized macroscopic cross section sets and (2) application of these sets to obtain a full 3D core solution with nodal codes. The standard approach adopts the branch methodology in which the branches represent all expected combinations of operational conditions as a function of burnup (main branch). The main branch is produced for constant, usually averaged, operating conditions (e.g. coolant density). As a result, the spectral history effects that associated with coolant density variation are not taken into account properly. Number of methods to solve this problem (such as micro-depletion and spectral indexes) were developed and implemented in modern nodal codes. Recently, we proposed a new and robust method to account for history effects. The methodology was implemented in DYN3D and involves modification of the few-group cross section sets. The method utilizes the local Pu-239 concentration as an indicator of spectral history. The method was verified for PWR and VVER applications. However, the spectrum variation in BWR core is more pronounced due to the stronger coolant density change. The purpose of the current work is investigating the applicability of the method to BWR analysis. The proposed methodology was verified against recently developed BGCore system, which couples Monte Carlo neutron transport with depletion and thermal-hydraulic solvers and
Burnup dependent core neutronic analysis for PBMR
International Nuclear Information System (INIS)
The strategy for core neutronics modeling is based on SCALE4.4 code KENOV.a module that uses Monte Carlo calculational methods. The calculations are based on detailed unit cell and detailed core modeling. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and the fuel kernels in the pebble. The core is then modeled by placing these pebbles randomly throughout the core, yet not loosing track of any one of them. For the burnup model, a cyclic manner is adopted by coupling the KENOV.a and ORIGEN-S modules. Shifting down one slice at each discrete time step, and inserting fresh fuel from the top, this cyclic calculation model continues until equilibrium burnup cycle is achieved. (author)
International Nuclear Information System (INIS)
Highlights: • Development of a neural network model to predict the requiered plutonium content. • The accuracy of this model is very good (0.37% of error). • Development of a neural network model to predict evolution of average cross sections. • Predictions allow calculating fuel depletion quickly and with a very good accuracy. • This approach has been applied to the PWR MOX case in a dynamic fuel cycle code. - Abstract: Dynamic fuel cycle simulation tools calculate nuclei inventories and mass flows evolution in an entire fuel cycle, from the mine to the final disposal. Usually, the fuel depletion in reactor is handled by a fuel loading model and a mean cross section predictor. In the case of a PWR–MOX, a fuel loading model provides from a plutonium stock the plutonium fraction in the fresh fuel needed to reach a specific burnup. A mean cross section predictor aims to assess isotopic cross sections required for building Bateman equations for any fresh fuel composition with a sufficient accuracy and a reasonable computing time. This paper presents a methodology based on neural networks for building a fuel loading model and a cross section predictor for a PWR reactor loaded with MOX fuel. The mean error of the plutonium content prediction from the fuel loading model is 0.37%. Furthermore, the mean cross section predictor allows completion of the fuel depletion calculation in less than one minute with excellent accuracy. A maximum deviation of 3% on main nuclei is obtained at the end of cycle between inventories calculated from neural networks and from the reference coupled neutron transport/fuel depletion calculation
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Results of the RTOP-CA code calculations for experiments in the research MIR reactor are presented. The MIR-reactor tests were made to study the activity release from defective WWER fuel at high burnup (∼60 MWd/kgU). The RTOP-CA calculations are compared to experimental data on radial distributions of burnup as well as radial profiles of Pu and Xe concentrations in fuel pellets. The RTOP-CA predictions are also compared to the data on activity release (radionuclides of I, Cs, Xe and Kr) from the test fuel rod with an artificial defect in cladding. Additional calculations were performed for WWER-1000 fuel of an advanced design. In these calculation series the effect of design innovations on activity release from defective fuel rods was estimated. It is demonstrated that in case of a failure the new generation of WWER fuel shows lower levels of activity release into primary coolant. (authors)
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A fuel performance code FEMAXI-6 has been developed for the analysis of LWR fuel rod behaviors in normal operation and transient (not accident) conditions. The code uses FEM for mechanical analysis, and has incorporated thermal and mechanical models of phenomena anticipated in high burn-up fuel rods, such as fuel thermal conductivity degradation and pellet-clad bonding. In the present study, PCMI induced by swelling in a high burn-up BWR type fuel rod has been analyzed by the FEMAXI-6 code. During a power ramp for the high burn-up fuel, instantaneous pellet swelling can significantly exceed the level that is predicted by a 'steady-rate' swelling model, causing a large circumferential strain in cladding. This phenomenon has been simulated by a new swelling model to take into account the fission gas bubble growth, and we found that the new model can give satisfactory predictions on cladding diametral expansion in comparison with measurements in test rod. The bubble growth model assumes an equilibrium between bubble size and external pressure on the bubble, and simultaneous solution is obtained with both bubble size determination and diffusion equation of fission gas atoms. In addition, a pellet-clad bonding model which assumes firm mechanical coupling between pellet outer surface and cladding inner surface predicts an elevated tensile stress in the axial direction of cladding during ramp, indicating the generation of bi-axial stress state in the cladding. These analyses by the FEMAXI-6 code enable us to predict the magnitude of mechanical loading on cladding during transient and also serve for failure investigation. Clearly, prediction by code calculation depends on the creep and stress-strain properties of highly irradiated cladding. (author)
Fast reactor 3D core and burnup analysis using VESTA
Energy Technology Data Exchange (ETDEWEB)
Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)
2012-07-01
Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)
Verification of the Monte Carlo code RMC with a whole PWR MOX/UO2 core benchmark
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Several types of V and V work are being carried out for the Reactor Monte Carlo code RMC, including the heterogeneous whole core configurations. In this paper, a whole PWR MOX/UO2 core benchmark which contains both UO2 and MOX assemblies with different enrichments and various burn-up points is chosen to verify RMC's criticality calculation capability, and the results of RMC and other codes are discussed and compared, such as eigenvalues, assembly power distributions, pin power distributions and so on. The discrepancies in eigenvalues and power distributions are satisfactory, which proves the accuracy of RMC's criticality calculation. Also, the influences of different cross-section libraries are discussed upon the results of RMC. Besides these results, the detailed comparisons between RMC and MCNP with the same ENDF/B-VII.0 cross-section library are carried out in this paper, including the comparisons of control rod worths calculated by both RMC and MCNP. According to the results, RMC and MCNP agree quite well in eigenvalues, power distributions and other results. The discrepancies of eigenvalues and control rod worth are fairly small and the relative differences of assembly and pin power distributions are acceptable. All these results contribute to the conclusion that the criticality calculation performance of RMC is accurate and excellent. (author)
Fuel burnup analysis for the Moroccan TRIGA research reactor
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Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of
Supercell burnup model for the physics design of BWR fuel assemblies
International Nuclear Information System (INIS)
A code called SUPERB has been developed for the BWR fuel assembly burnup analyses using supercell model. Each of the characteristic heterogeneities of a BWR fuel assembly like water gap, poisoned pins, control blade etc., is treated by invoking appropriate supercell concept. The burnup model of SUPERB is so devised as to strike a balance between accuracy and speed. This is achieved by building isotopic densities in each fuel pin separately while the depletion equations are solved only in a few groups of pins or burnup zones and the multigroup neutron spectra are differentiated in fewer group of pincell types. Multiple fuel ring burnup is considered only for Gd isotopes. A special empirical formula allows the microscopic cross section of Gd isotopes to be varied even during burnup integration. The supercell model has been tested against Monte Carlo results for the fresh cold clean Tarapur fuel assembly with two Gd fuel pins. The burnup model of SUPERB has been validated against one of the most sophisticated codes LWR-WIMS for a benchmark problem involving all the complexities of a BWR fuel assembly. The agreement of SUPERB results with both Monte Carlo and LWR-WIMS results is found to be excellent. (auth.)
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Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)
Energy Technology Data Exchange (ETDEWEB)
Boyarinov, V. F.; Davidenko, V. D.; Polismakov, A. A.; Tsibulsky, V. F. [Russian Research Center Kurchatov Inst., Nuclear Reactor Inst., 123182, Moscow (Russian Federation)
2006-07-01
Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)
On the inner workings of Monte Carlo codes
Dubbeldam, D.; Torres Knoop, A.; Walton, K.S.
2013-01-01
We review state-of-the-art Monte Carlo (MC) techniques for computing fluid coexistence properties (Gibbs simulations) and adsorption simulations in nanoporous materials such as zeolites and metal-organic frameworks. Conventional MC is discussed and compared to advanced techniques such as reactive MC, configurational-bias Monte Carlo and continuous fractional MC. The latter technique overcomes the problem of low insertion probabilities in open systems. Other modern methods are (hyper-)parallel...
Proton therapy Monte Carlo SRNA-VOX code
Ilić Radovan D.
2012-01-01
The most powerful feature of the Monte Carlo method is the possibility of simulating all individual particle interactions in three dimensions and performing numerical experiments with a preset error. These facts were the motivation behind the development of a general-purpose Monte Carlo SRNA program for proton transport simulation in technical systems described by standard geometrical forms (plane, sphere, cone, cylinder, cube). Some of the possible applications of the SRNA program are:...
The influence of pitch, burnup and absorber rods on the spent fuel pool criticality
International Nuclear Information System (INIS)
It has been shown that supercriticality might occur for some postulated accidents for the TRIGA spent fuel pool at ''Josef Stefan'' Institute in Ljubljana, Slovenia. However, in the previous studies, the effect of burnup was not accounted for. In this work the dependence of criticality on fuel burnup, the pitch among the elements and the number of uniformly mixed absorber rods for a square arrangement is presented. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. WIMS-D code was used to model the isotopic composition of the fuel for 5, 10, 20 and 30 % burnup without cooling time. The results show, that out of the three studied effects: pitch from contact (3.75cm) up to rack design pitch (8cm), number of absorbers from 0 to 8 and burnup up to 30 %, the pitch has the greatest influence on the multiplication factor keff. In the interval in which the pitch was changed, keff decreased for up to 0.45. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g. for contact of fuel elements with 8 absorber rods among them, keff values are smaller for almost 0.20 than for arrangement without absorber rods regardless of the burnup. The effect of burnup is the smallest since in no case keff decreases for more than 0.10, even for high burnups of 30 %. (author)
Recent developments of JAEA’s Monte Carlo code MVP for reactor physics applications
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Highlights: • This paper describes the recent development status of the Monte Carlo code MVP. • The basic features and capabilities of MVP are briefly described. • New capabilities useful for reactor analysis are also described. - Abstract: This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described
International Nuclear Information System (INIS)
In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)
Development of a Monte-Carlo Radiative Transfer Code for the Juno/JIRAM Limb Measurements
Sindoni, G.; Adriani, A.; Mayorov, B.; Aoki, S.; Grassi, D.; Moriconi, M.; Oliva, F.
2013-09-01
The Juno/JIRAM instrument will acquire limb spectra of the Jupiter atmosphere in the infrared spectral range. The analysis of these spectra requires a radiative transfer code that takes into account the multiple scattering by particles in a spherical-shell atmosphere. Therefore, we are developing a code based on the Monte-Carlo approach to simulate the JIRAM observations. The validation of the code was performed by comparison with DISORT-based codes.
Monte Carlo capabilities of the SCALE code system
International Nuclear Information System (INIS)
Highlights: • Foundational Monte Carlo capabilities of SCALE are described. • Improvements in continuous-energy treatments are detailed. • New methods for problem-dependent temperature corrections are described. • New methods for sensitivity analysis and depletion are described. • Nuclear data, users interfaces, and quality assurance activities are summarized. - Abstract: SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a “plug-and-play” framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE’s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2
Application of Monte Carlo code EGS4 to calculate gamma exposure buildup factors
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Exposure buildup factors up to 40 mean free paths ranging from 0.015 MeV to 15 MeV photon energy were calculated by using the Monte Carlo simulation code EGS4 for ordinary concrete. The calculation involves PHOTX cross section library, a point isotropic source, infinite uniform medium model and a particle splitting method and considers the Bremsstrahlung, fluorescent effect, correlative (Rayleigh) scatter. The results were compared with the relevant data. Results show that the data of the buildup factors calculated by the Monte Carlo code EGS4 was reliable. The Monte Carlo method can be used widely to calculate gamma-ray exposure buildup factors. (authors)
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FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes' integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6)
The analog linear interpolation approach for Monte Carlo simulation of PGNAA: The CEARPGA code
Zhang, Wenchao; Gardner, Robin P.
2004-01-01
The analog linear interpolation approach (ALI) has been developed and implemented to eliminate the big weight problem in the Monte Carlo simulation code CEARPGA. The CEARPGA code was previously developed to generate elemental library spectra for using the Monte Carlo - library least-squares (MCLLS) approach in prompt gamma-ray neutron activation analysis (PGNAA). In addition, some other improvements to this code have been introduced, including (1) adopting the latest photon cross-section data, (2) using an improved detector response function, (3) adding the neutron activation backgrounds, (4) generating the individual natural background libraries, (5) adding the tracking of annihilation photons from pair production interactions outside of the detector and (6) adopting a general geometry package. The simulated result from the new CEARPGA code is compared with those calculated from the previous CEARPGA code and the MCNP code and experimental data. The new CEARPGA code is found to give the best result.
On the inner workings of Monte Carlo codes
D. Dubbeldam; A. Torres Knoop; K.S. Walton
2013-01-01
We review state-of-the-art Monte Carlo (MC) techniques for computing fluid coexistence properties (Gibbs simulations) and adsorption simulations in nanoporous materials such as zeolites and metal-organic frameworks. Conventional MC is discussed and compared to advanced techniques such as reactive MC
Data libraries as a collaborative tool across Monte Carlo codes
Augelli, Mauro; Han, Mincheol; Hauf, Steffen; Kim, Chan-Hyeung; Kuster, Markus; Pia, Maria Grazia; Quintieri, Lina; Saracco, Paolo; Seo, Hee; Sudhakar, Manju; Eidenspointner, Georg; Zoglauer, Andreas
2010-01-01
The role of data libraries in Monte Carlo simulation is discussed. A number of data libraries currently in preparation are reviewed; their data are critically examined with respect to the state-of-the-art in the respective fields. Extensive tests with respect to experimental data have been performed for the validation of their content.
Monte Carlo Capabilities of the SCALE Code System
Rearden, B. T.; Petrie, L. M.; Peplow, D. E.; Bekar, K. B.; Wiarda, D.; Celik, C.; Perfetti, C. M.; Ibrahim, A. M.; Hart, S. W. D.; Dunn, M. E.
2014-06-01
SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a "plug-and-play" framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2, to be released in 2014, will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
Aurora T: a Monte Carlo code for transportation of neutral atoms in a toroidal plasma
International Nuclear Information System (INIS)
This paper contains a short description of Aurora code. This code have been developed at Princeton with Monte Carlo method for calculating neutral gas in cylindrical plasma. In this work subroutines such one can take in account toroidal geometry are developed
MCNP, a general Monte Carlo code for neutron and photon transport: a summary
International Nuclear Information System (INIS)
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces
MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis
International Nuclear Information System (INIS)
The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2, and the Weapons-Grade Mixed Oxide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
International Nuclear Information System (INIS)
This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs
Energy Technology Data Exchange (ETDEWEB)
Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)
2016-06-15
Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.
Using deterministic codes to accelerate continuous energy Monte-Carlo standards calculations
International Nuclear Information System (INIS)
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
Calculations of neutron penetration through graphite medium with Monte Carlo code MCNP
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Experiments for fast neutron penetration through graphite are analysed with the continuous energy Monte Carlo code MCNP. Reaction rates and energy spectra obtained with the MCNP are compared with measured values and calculated ones with McBEND code. And validity of penetration calculation with the MCNP is comfirmed. In addition, it is revealed that the MCNP code using Weight-Window method is well applicable to calculations of neutron penetration through graphite up to 70 cm in depth. (author)
International Nuclear Information System (INIS)
This paper is focused on the description and validation of the new dynamic model aimed at the processes taking place on grain boundaries in polycrystalline Light Water Reactor fuel based on the use of UO2, which has been recently developed for the START-3 code. The analysis embraces such processes as formation of fine surface clusters and larger intergranular pores, equi-axial grain growth, direct release and percolation of fission gas to the open surfaces. This model is also closely linked to intragranular behaviour of fission gas and essentially overlaps modelling of High Burnup Structure, as well as high temperature- and power transient-assisted processes. The model embodies some of the state-of-the-art approaches to numerical description of the processes taking place on grain boundaries, incorporating considerations of diffusion theory with respect to analysis of the dynamics of grain boundary pore growth/shrinkage caused by self-diffusion of the fuel material around them. Besides, it widely uses the elements of probability theory thereby accounting for stochastic nature of the analyzed phenomena. The several examples of model validation, illustrating credibility of pertinent results as applied to a wide enough range of application, including high-temperature out-of-pile annealing, High Burnup Structure Effects and transient behaviour of Light Water Reactor fuel, are also presented. As the validation shows, the developed model can be accepted as an important element of overall dynamic modelling with a view to justification of reliability of high burnup Light Water Reactor fuel, and safety analysis, as well
Extension of the TRANSURANUS burn-up model
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The validation range of the model in the TRANSURANUS fuel performance code for calculating the radial power density and burn-up in UO2 fuel has been extended from 64 MWd/kgHM up to 102 MWd/kgHM, thereby improving also its precision. In addition, the first verification of calculations with post-irradiation examination data is reported for LWR-MOX fuel with a rod average burn-up up to 45 MWd/kgHM. The extension covers the inclusion of new isotopes in order to account for the production of 238Pu. The corresponding one-group cross-sections used in the equations rely on results obtained with ALEPH, a new Monte Carlo burn-up code. The experimental verification is based on electron probe microanalysis (EPMA) and on secondary ion mass spectrometry (SIMS) as well as radiochemical data of fuel irradiated in commercial power plants. The deviations are quantified in terms of frequency distributions of the relative errors. The relative errors on the burn-up distributions in both fuel types remain below 12%, corresponding to the experimental scatter
MCMG: a 3-D multigroup P3 Monte Carlo code and its benchmarks
International Nuclear Information System (INIS)
In this paper a 3-D Monte Carlo multigroup neutron transport code MCMG has been developed from a coupled neutron and photon transport Monte Carlo code MCNP. The continuous-energy cross section library of the MCNP code is replaced by the multigroup cross section data generated by the transport lattice code, such as the WIMS code. It maintains the strong abilities of MCNP for geometry treatment, counting, variance reduction techniques and plotting. The multigroup neutron scattering cross sections adopt the Pn (n ≤ 3) approximation. The test results are in good agreement with the results of other methods and experiments. The number of energy groups can be varied from few groups to multigroup, and either macroscopic or microscopic cross section can be used. (author)
A vectorized Monte Carlo code for modeling photon transport in SPECT
International Nuclear Information System (INIS)
A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT
Burnup analysis of the VVER-1000 reactor using thorium-based fuel
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science
2014-12-15
This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.
Burnup analysis of the VVER-1000 reactor using thorium-based fuel
International Nuclear Information System (INIS)
This paper aims to investigate 232Th/233U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. 232Th/235U/238U oxide mixture was considered as fuel in the core, when the mass fraction of 232Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of 238U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the 232Th, 233U, 238U, 237Np, 239Pu, 241Am and 244Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.
Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes
International Nuclear Information System (INIS)
Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)
International Nuclear Information System (INIS)
Tokyo Metropolitan University of Health Sciences has done The Information Education using EGS4 Monte Carlo code since the 1998 fiscal year. Two items under practical training item were done. 1. The interaction between photon of 0.1 ∼ 10 MeV (Mega Electron Volt: MeV) and Aluminum (Al), Iron (Fe) and Lead (Pb). 2. The simulation of gamma ray energy measurement of the radiation detector. As the result, the student was possible the understanding of the radiation physics for the easiness at Practical training of EGS4 Monte Carlo code. (author)
The Monte Carlo code MCBEND - where it is and where it's going
International Nuclear Information System (INIS)
The Monte Carlo method forms a corner stone to the calculational procedures established in the UK for shielding design and assessment. The emphasis of the work in the shielding area is centred on the Monte Carlo code MCBEND. The work programme in support of the code is broadly directed towards utilisation of new hardware, the development of improved modelling algorithms, the development of new acceleration methods for specific applications and enhancements to user image. This paper summarises the current status of MCBEND and reviews developments carried out over the past two years and planned for the future. (author)
Parallelization of MCATNP MONTE CARLO particle transport code by using MPI
International Nuclear Information System (INIS)
A Monte Carlo code for simulating Atmospheric Transport of Neutrons and Photons (MCATNP) is used to simulate the ionization effects caused by high altitude nuclear detonation (HAND) and it was parallelized in MPI by adopting the leap random number producer and modifying the original serial code. The parallel results and serial results are identical. The speedup increases almost linearly with the number of processors used. The parallel efficiency is up to to 97% while 16 processors are used, and 94% while 32 are used. The experimental results show that parallelization can obviously reduce the calculation time of Monte Carlo simulation of HAND ionization effects. (authors)
Generalized Albedo option on the Morse Monte Carlo code
International Nuclear Information System (INIS)
The advisability of using the albedo procedure for solving deep penetration shielding problems which have ducts and other penetrations is investigated. It is generally accepted that the use of albedo data can dramatically improve the computational efficiency of certain Monte Carlo calculations - however the accuracy of these results may be unacceptable because of lost information during the albedo event and serious errors in the available differential albedo data. This study has been done to evaluate and appropriately modify the MORSE/BREESE package, to develop new methods for generating the required albedo data, and to extend the adjoint capability to the albedo modified calculations. The major modifications include the tracking of special particles inside albedo media, an option to displace the point-of-emergence during an albedo event, and an option to read, process, and use spatially-dependent albedo data for both forward and adjoint calculations. (author)
The three-dimensional Monte-Carlo code TRIPOLI-02
International Nuclear Information System (INIS)
TRIPOLI-2 solves the transport equation for neutrons or gamma rays in tridimensional geometrical configurations. TRIPOLI uses the Monte Carlo method. This method allows to treat exactly the geometrical configurations, the energy losses and the scattering laws. TRIPOLI 2 allows to treat the following problems: gamma transport problems, neutrons transport problems with fixed source (the problems can be time dependent or not), critical problems without fixed source and research of multiplication factor due to fissions, subcritical problems with fixed source and with multiplication by fission. These problems can be separate in two types. First type: shielding problems essentially with deep penetration and streaming through voids. Biasing technics are used to reduce the computing time. Second type: core problems for cell calculations or for small core calculations. In this case, it is necessary to have a fine representation of the cross sections. The thermalization is also treated exactly
Importance function by collision probabilities for Monte Carlo code Tripoli
International Nuclear Information System (INIS)
We present a completely automatic biasing technique where the parameters of the biased simulation are deduced from the solution of the adjoint transport equation calculated by collision probabilities. In this study we shall estimate the importance function through collision probabilities method and we shall evaluate its possibilities thanks to a Monte Carlo calculation. We have run simulations with this new biasing method for one-group transport problems with isotropic shocks (one dimension geometry and X-Y geometry) and for multigroup problems with anisotropic shocks (one dimension geometry). For the anisotropic problems we solve the adjoint equation with anisotropic collision probabilities. The results show that for the one-group and homogeneous geometry transport problems the method is quite optimal without Splitting and Russian Roulette technique but for the multigroup and heterogeneous X-Y geometry ones the figures of merit are higher if we add Splitting and Russian Roulette technique
Longitudinal development of extensive air showers: hybrid code SENECA and full Monte Carlo
Ortiz, J A; De Souza, V; Ortiz, Jeferson A.; Tanco, Gustavo Medina
2004-01-01
New experiments, exploring the ultra-high energy tail of the cosmic ray spectrum with unprecedented detail, are exerting a severe pressure on extensive air hower modeling. Detailed fast codes are in need in order to extract and understand the richness of information now available. Some hybrid simulation codes have been proposed recently to this effect (e.g., the combination of the traditional Monte Carlo scheme and system of cascade equations or pre-simulated air showers). In this context, we explore the potential of SENECA, an efficient hybrid tridimensional simulation code, as a valid practical alternative to full Monte Carlo simulations of extensive air showers generated by ultra-high energy cosmic rays. We extensively compare hybrid method with the traditional, but time consuming, full Monte Carlo code CORSIKA which is the de facto standard in the field. The hybrid scheme of the SENECA code is based on the simulation of each particle with the traditional Monte Carlo method at two steps of the shower devel...
Depletion of a BWR lattice using the racer continuous energy Monte Carlo code
International Nuclear Information System (INIS)
In the past several years there has been a renewed interest in the accuracy of a new generation of lattice physics codes. Most of the time these codes are benchmarked against Monte Carlo codes only at beginning of cycle. In this paper a highly heterogeneous BWR lattice depletion benchmark problem is presented. Results of a 40% void depletion using the RACER continuous energy Monte Carlo code are also presented. Complete problem specifications are given so that comparisons with lattice physics codes or other Monte Carlo codes is possible. The RACER calculations were performed with the ENDF/B-V cross section set. Each flux calculation utilized 2.7 million histories resulting in 95% confidence intervals of ∼1 milli-k on the eigenvalue and ∼1% uncertainties on pin-wise power fractions. Timing statistics for the calculation using the vectorized RACER code averaged ∼ 24,000 neutrons/minute on a single processor of a CRAY-C90 computer
MKENO-DAR: a direct angular representation Monte Carlo code for criticality safety analysis
International Nuclear Information System (INIS)
Improving the Monte Carlo code MULTI-KENO, the MKENO-DAR (Direct Angular Representation) code has been developed for criticality safety analysis in detail. A function was added to MULTI-KENO for representing anisotropic scattering strictly. With this function, the scattering angle of neutron is determined not by the average scattering angle μ-bar of the Pl Legendre polynomial but by the random work operation using probability distribution function produced with the higher order Legendre polynomials. This code is avilable for the FACOM-M380 computer. This report is a computer code manual for MKENO-DAR. (author)
Full-core pin-power calculations using Monte Carlo codes
International Nuclear Information System (INIS)
Pin wise calculations of core power distribution have been performed for a criticality mock up installation that models a WWER-1000 reactor. Two Monte Carlo codes have been applied for solving of this problem: the MCNP4B code and the KENO-VI code from the SCALE 4.4 system. The codes use different kinds of neutron cross section data: pointwise continuous-energy ENDF/B-VI data and multigroup ENDF/B-V data. Comparisons of calculated results show that the MCNP4B and KENO-VI results are in good agreement. (authors)
International Nuclear Information System (INIS)
The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)
Parallel processing of Monte Carlo code MCNP for particle transport problem
Energy Technology Data Exchange (ETDEWEB)
Higuchi, Kenji; Kawasaki, Takuji
1996-06-01
It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)
MCAM 5: an advanced interface program for multiple Monte Carlo Codes
International Nuclear Information System (INIS)
The Automatic Modeling Program for Neutronics and Radiation Transport Simulation (MCAM) developed in China, is an advanced interface program between CAD (Computer Aided Design) systems and Monte Carlo (MC) codes. It can significantly reduce the manpower and enhance reliability for constructing MC codes input of complex systems. The latest version MCAM 4.8 was a mature and efficient version which was benchmarked with ITER benchmark model and has been used by hundreds of institutes in more than 40 countries all over the world. It can deal with MCNP and TRIPOLI models. The main function of MCAM is to convert geometries in CAD systems to geometries in MC codes input files. The MCAM version 5.2 is going to be released with added capabilities to support SuperMC, Geant4 and FLUKA Monte Carlo codes
International Nuclear Information System (INIS)
The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor
Modelling photon transport in non-uniform media for SPECT with a vectorized Monte Carlo code.
Smith, M F
1993-10-01
A vectorized Monte Carlo code has been developed for modelling photon transport in non-uniform media for single-photon-emission computed tomography (SPECT). The code is designed to compute photon detection kernels, which are used to build system matrices for simulating SPECT projection data acquisition and for use in matrix-based image reconstruction. Non-uniform attenuating and scattering regions are constructed from simple three-dimensional geometric shapes, in which the density and mass attenuation coefficients are individually specified. On a Stellar GS1000 computer, Monte Carlo simulations are performed between 1.6 and 2.0 times faster when the vector processor is utilized than when computations are performed in scalar mode. Projection data acquired with a clinical SPECT gamma camera for a line source in a non-uniform thorax phantom are well modelled by Monte Carlo simulations. The vectorized Monte Carlo code was used to stimulate a 99Tcm SPECT myocardial perfusion study, and compensations for non-uniform attenuation and the detection of scattered photons improve activity estimation. The speed increase due to vectorization makes Monte Carlo simulation more attractive as a tool for modelling photon transport in non-uniform media for SPECT. PMID:8248288
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
International Nuclear Information System (INIS)
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided
Energy Technology Data Exchange (ETDEWEB)
Perfetti, Christopher M [ORNL; Martin, William R [University of Michigan; Rearden, Bradley T [ORNL; Williams, Mark L [ORNL
2012-01-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the SHIFT Monte Carlo code within the Scale code package. The methods were used for several simple test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods.
QCDMPI - pure QCD Monte Carlo simulation code with MPI
International Nuclear Information System (INIS)
QCDMPI is a pure QCD simulation code with MPI calls. QCDMPI is very portable because; - you can simulate any-dimensional QCD, - on any-dimensional partitioning, - on any number of processors, - with rather small working area. Also by this program, you can get two performances, - calculation (link update time) - communication (MB/sec). In this paper, outline of QCDMPI is reported. Comparison of the performances on several parallel machines; AP1000, AP1000+, AP3000, Cenju-3, Paragon, SR2201 and Workstation Cluster, is also reported. (orig.)
Verification of the shift Monte Carlo code with the C5G7 reactor benchmark
Energy Technology Data Exchange (ETDEWEB)
Sly, N. C.; Mervin, B. T. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States); Mosher, S. W.; Evans, T. M.; Wagner, J. C. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)
2012-07-01
Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)
International Nuclear Information System (INIS)
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
The present of shielding analysis with nuclear data for continuous energy Monte Carlo code MCNP
International Nuclear Information System (INIS)
Following three problems are analyzed by continuous energy Monte Carlo code MCNP with JENDL-3.2, 3.3, and ENDF/B-VI. 1. Shielding analysis of WINFRITH-Aspins iron deep penetration experiment. 2. Shielding analysis of TN-12A spent fuel transport cask experiment. 3. Shielding analysis of modular shielding house keeping spent fuel transportable casks. (author)
A comparative study of MONTEBURNS and MCNPX 2.6.0 codes in ADS simulations
International Nuclear Information System (INIS)
The possible use of the MONTEBURNS and MCNPX 2.6.0 codes in Accelerator-driven systems (ADSs) simulations for fuel evolution description is discussed. ADSs are investigated for fuel breeding and long-lived fission product transmutation so simulations of fuel evolution have a great relevance. The burnup/depletion capability is present in both studied codes. MONTEBURNS code links Monte Carlo N-Particle Transport Code (MCNP) to the radioactive decay burnup code ORIGEN2, whereas MCNPX depletion/ burnup capability is a linked process involving steady-state flux calculations by MCNPX and nuclide depletion calculations by CINDER90. A lead-cooled accelerator-driven system fueled with thorium was simulated and the results obtained using MONTEBURNS code and the results from MCNPX 2.6.0 code were compared. The system criticality and the variation of the actinide inventory during the burnup were evaluated and the results indicate a similar behavior between the results of each code. (author)
Subroutines to Simulate Fission Neutrons for Monte Carlo Transport Codes
Lestone, J P
2014-01-01
Fortran subroutines have been written to simulate the production of fission neutrons from the spontaneous fission of 252Cf and 240Pu, and from the thermal neutron induced fission of 239Pu and 235U. The names of these four subroutines are getnv252, getnv240, getnv239, and getnv235, respectively. These subroutines reproduce measured first, second, and third moments of the neutron multiplicity distributions, measured neutron-fission correlation data for the spontaneous fission of 252Cf, and measured neutron-neutron correlation data for both the spontaneous fission of 252Cf and the thermal neutron induced fission of 235U. The codes presented here can be used to study the possible uses of neutron-neutron correlations in the area of transparency measurements and the uses of neutron-neutron correlations in coincidence neutron imaging.
Homma, Yuto; Moriwaki, Hiroyuki; Ohki, Shigeo; Ikeda, Kazumi
2014-06-01
This paper deals with verification of three dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at beginning of cycle of an initial core and at beginning and end of cycle of equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multi-plication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity.
Reactivity effects of nonuniform axial burnup distributions on spent fuel
International Nuclear Information System (INIS)
When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup
Reactivity effects of nonuniform axial burnup distributions on spent fuel
Energy Technology Data Exchange (ETDEWEB)
Leary, R.W. II; Parish, T.A. [Texas A & M Univ., College Station, TX (United States)
1995-12-01
When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup.
International Nuclear Information System (INIS)
Computational Monte Carlo (MC) codes have been used for simulation of nuclear installations mainly for internal monitoring of workers, the well known as Whole Body Counters (WBC). The main goal of this project was the modeling and simulation of the counting efficiency (CE) of a WBC system using three different MC codes: MCNPX, EGSnrc and VMC in-vivo. The simulations were performed for three different groups of analysts. The results shown differences between the three codes, as well as in the results obtained by the same code and modeled by different analysts. Moreover, all the results were also compared to the experimental results obtained in laboratory for meaning of validation and final comparison. In conclusion, it was possible to detect the influence on the results when the system is modeled by different analysts using the same MC code and in which MC code the results were best suited, when comparing to the experimental data result. (author)
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported
Applications of FLUKA Monte Carlo code for nuclear and accelerator physics
Battistoni, Giuseppe; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fasso, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M; Morone, Christina; Muraro, Silvia; Parodi, Katerina; Patera, Vincenzo; Pelliccioni, Maurizio; Pinsky, Lawrence; Ranft, Johannes; Roesler, Stefan; Rollet, Sofia; Sala, Paola R; Santana, Mario; Sarchiapone, Lucia; Sioli, Maximiliano; Smirnov, George; Sommerer, Florian; Theis, Christian; Trovati, Stefania; Villari, R; Vincke, Heinz; Vincke, Helmut; Vlachoudis, Vasilis; Vollaire, Joachim; Zapp, Neil
2011-01-01
FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such top...
Simulating fast transients with fuel behavior feedback using the Serpent 2 Monte Carlo code
International Nuclear Information System (INIS)
Simulating transients with reactivity feedback effects using Monte Carlo neutron transport codes can be used for validating deterministic transient codes or estimating for example the total deposited energy in a fuel rod following a known reactivity insertion in the system. Recent increases in computational power as well as developments in calculation methodology makes it possible to obtain a coupled solution for several aspects of the multi-physics problem in a single calculation. This paper describes the different methods implemented in Serpent 2 Monte Carlo code that enable it to model fast transients with fuel behavior feedback. The capability is demonstrated in a prompt critical pin-cell case, where the transient is shut down by the negative reactivity from rising fuel temperature. (author)
Dose conversion coefficients for ICRP110 voxel phantom in the Geant4 Monte Carlo code
Martins, M. C.; Cordeiro, T. P. V.; Silva, A. X.; Souza-Santos, D.; Queiroz-Filho, P. P.; Hunt, J. G.
2014-02-01
The reference adult male voxel phantom recommended by International Commission on Radiological Protection no. 110 was implemented in the Geant4 Monte Carlo code. Geant4 was used to calculate Dose Conversion Coefficients (DCCs) expressed as dose deposited in organs per air kerma for photons, electrons and neutrons in the Annals of the ICRP. In this work the AP and PA irradiation geometries of the ICRP male phantom were simulated for the purpose of benchmarking the Geant4 code. Monoenergetic photons were simulated between 15 keV and 10 MeV and the results were compared with ICRP 110, the VMC Monte Carlo code and the literature data available, presenting a good agreement.
ERSN-OpenMC, a Java-based GUI for OpenMC Monte Carlo code
Directory of Open Access Journals (Sweden)
Jaafar EL Bakkali
2016-07-01
Full Text Available OpenMC is a new Monte Carlo transport particle simulation code focused on solving two types of neutronic problems mainly the k-eigenvalue criticality fission source problems and external fixed fission source problems. OpenMC does not have any Graphical User Interface and the creation of one is provided by our java-based application named ERSN-OpenMC. The main feature of this application is to provide to the users an easy-to-use and flexible graphical interface to build better and faster simulations, with less effort and great reliability. Additionally, this graphical tool was developed with several features, as the ability to automate the building process of OpenMC code and related libraries as well as the users are given the freedom to customize their installation of this Monte Carlo code. A full description of the ERSN-OpenMC application is presented in this paper.
A new Monte Carlo code for absorption simulation of laser-skin tissue interaction
Institute of Scientific and Technical Information of China (English)
Afshan Shirkavand; Saeed Sarkar; Marjaneh Hejazi; Leila Ataie-Fashtami; Mohammad Reza Alinaghizadeh
2007-01-01
In laser clinical applications, the process of photon absorption and thermal energy diffusion in the target tissue and its surrounding tissue during laser irradiation are crucial. Such information allows the selection of proper operating parameters such as laser power, and exposure time for optimal therapeutic. The Monte Carlo method is a useful tool for studying laser-tissue interaction and simulation of energy absorption in tissue during laser irradiation. We use the principles of this technique and write a new code with MATLAB 6.5, and then validate it against Monte Carlo multi layer (MCML) code. The new code is proved to be with good accuracy. It can be used to calculate the total power bsorbed in the region of interest. This can be combined for heat modelling with other computerized programs.
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
Energy Technology Data Exchange (ETDEWEB)
Iandola, F N; O' Brien, M J; Procassini, R J
2010-11-29
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code
International Nuclear Information System (INIS)
Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.
The use of an inbuilt importance generator for acceleration of the Monte Carlo code MCBEND
International Nuclear Information System (INIS)
Monte Carlo is currently the most accurate method for the analysis of neutron and gamma-ray transport. However its application, especially to deep penetration studies, is costly in terms of the man-days to set up the calculation and in terms of computer usage. The MAGIC module, developed at the Winfrith Technology Centre, addresses both these problems. It employs an automated procedure based upon the established technique of splitting/roulette with an importance function derived from the solution of the adjoint diffusion equation. Examples are given of the application of the module with Monte Carlo code MCBEND
Burnup study of 18 months and 16/20 months cycle AP1000 cores using CASMO4E and SIMULATE-3 codes
International Nuclear Information System (INIS)
AP1000 reactor is an advanced pressurized water reactor equipped with passive safety systems. AP1000 reactor core is designed for 18 month cycle length and can also be used for 16/20 month alternate cycle lengths to meet energy requirements during high demand periods. The purpose of this study is to analyze the feasibility of AP1000 core for both 18 and 16/20 alternate cycle lengths by using CASMO4E and SIMULATE-3 code package. For this purpose, burnup analysis of both the schemes is carried out from initial core loading through optimized transition cores to equilibrium core. The study is performed by modeling three dimensional full core in SIMULATE-3 with each fuel assembly divided into 40 axial and 4 radial quadrant nodes. Once and twice burned fuel reloading from one cycle to the next and removal of burnable poison rods from the core after first cycle options are used in these codes. The results of this study indicate that both the cycle schemes can be utilized by varying the core loading pattern. Moreover, reactivity coefficients, total power peaking factors and enthalpy rise factors are calculated which indicate that the AP1000 core provide adequate safety margins in both the cycle schemes. (author)
The development of depletion program coupled with Monte Carlo computer code
International Nuclear Information System (INIS)
The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNPREBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)
A new assembly-level Monte Carlo neutron transport code for reactor physics calculations
International Nuclear Information System (INIS)
This paper presents a new assembly-level Monte Carlo neutron transport code, specifically intended for diffusion code group-constant generation and other reactor physics calculations. The code is being developed at the Technical Research Centre of Finland (VTT), under the working title 'Probabilistic Scattering Game', or PSG. The PSG code uses a method known as Woodcock tracking to simulate neutron histories. The advantages of the method include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. The main drawback is the inability to calculate reaction rates in optically thin volumes. This narrows the field of application to calculations involving parameters integrated over large volumes. The main features of the PSG code and the Woodcock tracking method are introduced. The code is applied in three example cases, involving infinite lattices of two-dimensional LWR fuel assemblies. Comparison calculations are carried out using MCNP4C and CASMO-4E. The results reveal that the code performs quite well in the calculation cases of this study, especially when compared to MCNP. The PSG code is still under extensive development and there are both flaws in the simulation of the interaction physics and programming errors in the source code. The results presented here, however, seem very encouraging, especially considering the early development stage of the code. (author)
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes
Pinsky, L; Ferrari, A; Sala, P; Carminati, F; Brun, R
2001-01-01
This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be usef...
Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Directory of Open Access Journals (Sweden)
Ilić Radovan D.
2002-01-01
Full Text Available This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice.
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
International Nuclear Information System (INIS)
This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice. (author)
Calculation of effective delayed neutron fraction with modified library of Monte Carlo code
International Nuclear Information System (INIS)
Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data
Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics
International Nuclear Information System (INIS)
The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)
DgSMC-B code: A robust and autonomous direct simulation Monte Carlo code for arbitrary geometries
Kargaran, H.; Minuchehr, A.; Zolfaghari, A.
2016-07-01
In this paper, we describe the structure of a new Direct Simulation Monte Carlo (DSMC) code that takes advantage of combinatorial geometry (CG) to simulate any rarefied gas flows Medias. The developed code, called DgSMC-B, has been written in FORTRAN90 language with capability of parallel processing using OpenMP framework. The DgSMC-B is capable of handling 3-dimensional (3D) geometries, which is created with first-and second-order surfaces. It performs independent particle tracking for the complex geometry without the intervention of mesh. In addition, it resolves the computational domain boundary and volume computing in border grids using hexahedral mesh. The developed code is robust and self-governing code, which does not use any separate code such as mesh generators. The results of six test cases have been presented to indicate its ability to deal with wide range of benchmark problems with sophisticated geometries such as airfoil NACA 0012. The DgSMC-B code demonstrates its performance and accuracy in a variety of problems. The results are found to be in good agreement with references and experimental data.
The Physical Models and Statistical Procedures Used in the RACER Monte Carlo Code
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This report describes the MCV (Monte Carlo - Vectorized)Monte Carlo neutron transport code [Brown, 1982, 1983; Brown and Mendelson, 1984a]. MCV is a module in the RACER system of codes that is used for Monte Carlo reactor physics analysis. The MCV module contains all of the neutron transport and statistical analysis functions of the system, while other modules perform various input-related functions such as geometry description, material assignment, output edit specification, etc. MCV is very closely related to the 05R neutron Monte Carlo code [Irving et al., 1965] developed at Oak Ridge National Laboratory. 05R evolved into the 05RR module of the STEMB system, which was the forerunner of the RACER system. Much of the overall logic and physics treatment of 05RR has been retained and, indeed, the original verification of MCV was achieved through comparison with STEMB results. MCV has been designed to be very computationally efficient [Brown, 1981, Brown and Martin, 1984b; Brown, 1986]. It was originally programmed to make use of vector-computing architectures such as those of the CDC Cyber- 205 and Cray X-MP. MCV was the first full-scale production Monte Carlo code to effectively utilize vector-processing capabilities. Subsequently, MCV was modified to utilize both distributed-memory [Sutton and Brown, 1994] and shared memory parallelism. The code has been compiled and run on platforms ranging from 32-bit UNIX workstations to clusters of 64-bit vector-parallel supercomputers. The computational efficiency of the code allows the analyst to perform calculations using many more neutron histories than is practical with most other Monte Carlo codes, thereby yielding results with smaller statistical uncertainties. MCV also utilizes variance reduction techniques such as survival biasing, splitting, and rouletting to permit additional reduction in uncertainties. While a general-purpose neutron Monte Carlo code, MCV is optimized for reactor physics calculations. It has the
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
International Nuclear Information System (INIS)
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)
Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code
Energy Technology Data Exchange (ETDEWEB)
Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan
2007-05-15
This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.
TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation
Energy Technology Data Exchange (ETDEWEB)
Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Cayla, Pierre-Yves; Fausser, Clement [MILLENNIUM, 16 Av du Québec Silic 628, F-91945 Villebon sur Yvette (France); Damian, Frederic; Lee, Yi-Kang; Puma, Antonella Li; Trama, Jean-Christophe [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France)
2014-10-15
3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4{sup ®} is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4{sup ®}, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4{sup ®} A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4{sup ®} is shown; discrepancies are mainly included in the statistical error.
Françoise Benz
2006-01-01
2005-2006 ACADEMIC TRAINING PROGRAMME LECTURE SERIES 27, 28, 29 June 11:00-12:00 - TH Conference Room, bldg. 4 The use of Monte Carlo radiation transport codes in radiation physics and dosimetry F. Salvat Gavalda,Univ. de Barcelona, A. FERRARI, CERN-AB, M. SILARI, CERN-SC Lecture 1. Transport and interaction of electromagnetic radiation F. Salvat Gavalda,Univ. de Barcelona Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interaction models and multiple-scattering theories will be analyzed. Benchmark comparisons of simu...
Shielding evaluation for e-Linac - Inter-comparison of Monte Carlo codes and analytical calculations
International Nuclear Information System (INIS)
Estimation of optimum shielding thickness is an important aspect in radiation protection as well as in assessment of cost effectiveness of any upcoming accelerator facility. Analytical calculations for shielding estimates are fast and being frequently used even though they are very approximate. Estimates by Monte Carlo codes, on the other hand is accurate, provided used in a judicious manner, but they are very time consuming and require high end computational hardware. The purpose of this work is to compare the results from various available Monte Carlo codes, such as FLUKA and EGSmc. The estimated output was also compared with the analytical techniques. For the work, an e-Linac facility of 50 MeV electron beam was used and calculations were carried out with 1 mA beam current. (author)
International Nuclear Information System (INIS)
COG is a major multiparticle simulation code in the LLNL Monte Carlo radiation transport toolkit. It was designed to solve deep-penetration radiation shielding problems in arbitrarily complex 3D geometries, involving coupled transport of photons, neutrons, and electrons. COG was written to provide as much accuracy as the underlying cross-sections will allow, and has a number of variance-reduction features to speed computations. Recently COG has been applied to the simulation of high- resolution radiographs of complex objects and the evaluation of contraband detection schemes. In this paper we will give a brief description of the capabilities of the COG transport code and show several examples of neutron and gamma-ray imaging simulations. Keywords: Monte Carlo, radiation transport, simulated radiography, nonintrusive inspection, neutron imaging
Vectorization and multitasking with a Monte-Carlo code for neutron transport problems
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This paper summarizes two improvements of a Monte Carlo code by resorting to vectorization and multitasking techniques. After a short presentation of the physical problem to solve and a description of the main difficulties to produce an efficient coding, this paper introduces the vectorization principles employed and briefly describes how the vectorized algorithm works. Next, measured performances on CRAY 1S, CYBER 205 and CRAY X-MP are compared. The second part of this paper is devoted to multitasking technique. Starting from the standard multitasking tools available with FORTRAN on CRAY X-MP/4, a multitasked algorithm and its measured speed-ups are presented. In conclusion we prove that vector and parallel computers are a great opportunity for such Monte Carlo algorithms
Penelope - A code system for Monte Carlo simulation of electron and photon transport
International Nuclear Information System (INIS)
The computer code system PENELOPE (version 2001) performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials for a wide energy range, from a few hundred eV to about 1 GeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A geometry package called PENGEOM permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the PENELOPE code system, but also to provide the user with the necessary information to understand the details of the Monte-Carlo algorithm. (authors)
PEREGRINE: An all-particle Monte Carlo code for radiation therapy
International Nuclear Information System (INIS)
The goal of radiation therapy is to deliver a lethal dose to the tumor while minimizing the dose to normal tissues. To carry out this task, it is critical to calculate correctly the distribution of dose delivered. Monte Carlo transport methods have the potential to provide more accurate prediction of dose distributions than currently-used methods. PEREGRINE is a new Monte Carlo transport code developed at Lawrence Livermore National Laboratory for the specific purpose of modeling the effects of radiation therapy. PEREGRINE transports neutrons, photons, electrons, positrons, and heavy charged-particles, including protons, deuterons, tritons, helium-3, and alpha particles. This paper describes the PEREGRINE transport code and some preliminary results for clinically relevant materials and radiation sources
New burnup calculation of TRIGA IPR-R1 reactor
International Nuclear Information System (INIS)
The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)
New burnup calculation of TRIGA IPR-R1 reactor
Energy Technology Data Exchange (ETDEWEB)
Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2015-07-01
The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)
Platt, M. E.; Lewis, E. E.; Boehm, F.
1991-01-01
A Monte Carlo Fortran computer program was developed that uses two variance reduction techniques for computing system reliability applicable to solving very large highly reliable fault-tolerant systems. The program is consistent with the hybrid automated reliability predictor (HARP) code which employs behavioral decomposition and complex fault-error handling models. This new capability is called MC-HARP which efficiently solves reliability models with non-constant failures rates (Weibull). Common mode failure modeling is also a specialty.
International Nuclear Information System (INIS)
A model of a gamma sterilizer was built using the ITS/ACCEPT Monte Carlo code and verified through dosimetry. Individual dosimetry measurements in homogeneous material were pooled to represent larger bodies that could be simulated in a reasonable time. With the assumptions and simplifications described, dose predictions were within 2-5% of dosimetry. The model was used to simulate product movement through the sterilizer and to predict information useful for process optimization and facility design
Validation of GEANT4 Monte Carlo Simulation Code for 6 MV Varian Linac Photon Beam
International Nuclear Information System (INIS)
The head of a clinical linear accelerator based on the manufacturer detailed information is simulated by using GEANT4. Percentage Depth Dose (PDD) and flatness symmetry (lateral dose profiles) in water phantom were evaluated. Comparisons between experimental and simulated data were carried out for two field sizes; 5 × 5, and 10 ×10 cm2. The obtained results indicated that GEANT4 code is a promising and validated Monte Carlo program for using in radiotherapy applications
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies
International Nuclear Information System (INIS)
TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
Energy Technology Data Exchange (ETDEWEB)
Perfetti, C.; Martin, W. [Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States); Rearden, B.; Williams, M. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Div., Bldg. 5700, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)
2012-07-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)
Exact modeling of the torus geometry with Monte Carlo transport code
International Nuclear Information System (INIS)
It is valuable to model torus geometry exactry for the neutronics design of fusion reactor in order to assess neutronics characteristics such as tritium breeding ratio, heat generation rate, etc, near the plasma. Monte Carlo code MORSE-GG which plays important role in the radiation streaming calculation of fusion reactors had been able to deal with the geometry composed of second order surfaces. The MORSE-GG program is modified to be able to deal with torus geometry which has fourth order surface by solving biquadratic equations, hoping that MORSE-GG code becomes more effective for the neutronics calculation of the Tokamak fusion reactor. (author)
Efficient data management techniques implemented in the Karlsruhe Monte Carlo code KAMCCO
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The Karlsruhe Monte Carlo Code KAMCCO is a forward neutron transport code with an eigenfunction and a fixed source option, including time-dependence. A continuous energy model is combined with a detailed representation of neutron cross sections, based on linear interpolation, Breit-Wigner resonances and probability tables. All input is processed into densely packed, dynamically addressed parameter fields and networks of pointers (addresses). Estimation routines are decoupled from random walk and analyze a storage region with sample records. This technique leads to fast execution with moderate storage requirements and without any I/O-operations except in the input and output stages. 7 references. (U.S.)
Installation of Monte Carlo neutron and photon transport code system MCNP4
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The continuous energy Monte Carlo code MCNP-4 including its graphic functions has been installed on the Sun-4 sparc-2 work station with minor corrections. In order to validate the installed MCNP-4 code, 25 sample problems have been executed on the work station and these results have been compared with the original ones. And, the most of the graphic functions have been demonstrated by using 3 sample problems. Further, additional 14 nuclides have been included to the continuous cross section library edited from JENDL-3. (author)
International Nuclear Information System (INIS)
A code to simulate almost any electron--photon transport problem conceivable is described. The report begins with a lengthy historical introduction and a description of the shower generation process. Then the detailed physics of the shower processes and the methods used to simulate them are presented. Ideas of sampling theory, transport techniques, particle interactions in general, and programing details are discussed. Next, EGS calculations and various experiments and other Monte Carlo results are compared. The remainder of the report consists of user manuals for EGS, PEGS, and TESTSR codes; options, input specifications, and typical output are included. 38 figures, 12 tables
Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry
International Nuclear Information System (INIS)
Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman and Richtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the
ASCOT: redesigned Monte Carlo code for simulations of minority species in tokamak plasmas
Hirvijoki, Eero; Koskela, Tuomas; Kurki-Suonio, Taina; Miettunen, Juho; Sipilä, Seppo; Snicker, Antti; Äkäslompolo, Simppa
2013-01-01
A comprehensive description of methods for Monte Carlo studies of fast ions and impurity species in tokamak plasmas is presented. The described methods include Hamiltonian orbit-following in particle and guiding center phase space, test particle or guiding center solution of the kinetic equation applying stochastic differential equations in the presence of Coulomb collisions, Neoclassical tearing modes and Alfv\\'en eigenmodes as electromagnetic perturbations relevant for fast ions, together with plasma flow and atomic reactions relevant for impurity studies. Applying the methods, a complete reimplementation of a well-established minority species code is carried out as a response both to the increase in computing power during the last twenty years and to the weakly structured growth of the previous code which has made implementation of additional models impractical. Also, a thorough benchmark between the previous code and the reimplementation is accomplished, showing good agreement between the codes.
Uncertainties associated with the use of the KENO Monte Carlo criticality codes
International Nuclear Information System (INIS)
The KENO multi-group Monte Carlo criticality codes have earned the reputation of being efficient, user friendly tools especially suited for the analysis of situations commonly encountered in the storage and transportation of fissile materials. Throughout their twenty years of service, a continuing effort has been made to maintain and improve these codes to meet the needs of the nuclear criticality safety community. Foremost among these needs is the knowledge of how to utilize the results safely and effectively. Therefore it is important that code users be aware of uncertainties that may affect their results. These uncertainties originate from approximations in the problem data, methods used to process cross sections, and assumptions, limitations and approximations within the criticality computer code itself. 6 refs., 8 figs., 1 tab
Specific Monte Carlo code development for nuclear well-logging tool responses
International Nuclear Information System (INIS)
McPNL is a specific Monte Carlo computer code that has been developed to simulate a pulsed neutron oil well logging tool and uses implicit capture, Russian roulette and statistical estimation techniques as primary variance reduction methods. The code has been validated by benchmarking against six sets of laboratory test pit data on water, limestone and quartz formations with widely varying sets of borehole and formation conditions. McDNL is a specific Monte Carlo computer code that has been developed to simulate a dual-spaced neutron porosity tool. The low counting yield in the far detector of the tool requires the use of biasing schemes to obtain adequate efficiency. Exponential transform and directional biasing techniques have been applied with remarkable success for this problem, along with source biasing, implicit capture, Russian roulette and statistical estimation techniques. The code has been benchmarked against five sets of laboratory test pit data and found to be valid. Correlated sampling can be optionally used in the code to accurately predict the relative change in the detector response due to small perturbations in the formation porosity. (author)
An analytical solution to a simplified EDXRF model for Monte Carlo code verification
International Nuclear Information System (INIS)
The objective of this study is to obtain an analytical solution to the scalar photon transport equation that can be used to obtain benchmark results for the verification of energy dispersive X-Ray fluorescence (EDXRF) Monte Carlo simulation codes. The multi-collided flux method (multiple scattering method) is implemented to obtain analytical expressions for the space-, energy-, and angle-dependent scalar photon flux for a one dimensional EDXRF model problem. In order to obtain benchmark results, higher-order multiple scattering terms are included in the multi-collided flux method. The details of the analytical solution and of the proposed EDXRF model problem are presented. Analytical expressions obtained are then used to calculate the energy-dependent current. The analytically-calculated energy-dependent current is compared with Monte Carlo code results. The findings of this study show that analytical solutions to the scalar photon transport equation with the proposed model problem can be used as a verification tool in EDXRF Monte Carlo code development.
International Nuclear Information System (INIS)
This paper presents an unstructured mesh based multi-physics interface implemented in the Serpent 2 Monte Carlo code, for the purpose of coupling the neutronics solution to component-scale thermal hydraulics calculations, such as computational fluid dynamics (CFD). The work continues the development of a multi-physics coupling scheme, which relies on the separation of state-point information from the geometry input, and the capability to handle temperature and density distributions by a rejection sampling algorithm. The new interface type is demonstrated by a simplified molten-salt reactor test case, using a thermal hydraulics solution provided by the CFD solver in OpenFOAM. (author)
ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence
The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety
International Nuclear Information System (INIS)
It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool keff is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor keff. In the interval in which the pitch was changed, keff decreased for up to ∼0.4 for standard and ∼0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, keff values are smaller for ∼0.2 (∼0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is ∼0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the keff of spent-fuel pool and can be neglected in spent-fuel pool design
Implementation of a Monte Carlo based inverse planning model for clinical IMRT with MCNP code
He, Tongming Tony
In IMRT inverse planning, inaccurate dose calculations and limitations in optimization algorithms introduce both systematic and convergence errors to treatment plans. The goal of this work is to practically implement a Monte Carlo based inverse planning model for clinical IMRT. The intention is to minimize both types of error in inverse planning and obtain treatment plans with better clinical accuracy than non-Monte Carlo based systems. The strategy is to calculate the dose matrices of small beamlets by using a Monte Carlo based method. Optimization of beamlet intensities is followed based on the calculated dose data using an optimization algorithm that is capable of escape from local minima and prevents possible pre-mature convergence. The MCNP 4B Monte Carlo code is improved to perform fast particle transport and dose tallying in lattice cells by adopting a selective transport and tallying algorithm. Efficient dose matrix calculation for small beamlets is made possible by adopting a scheme that allows concurrent calculation of multiple beamlets of single port. A finite-sized point source (FSPS) beam model is introduced for easy and accurate beam modeling. A DVH based objective function and a parallel platform based algorithm are developed for the optimization of intensities. The calculation accuracy of improved MCNP code and FSPS beam model is validated by dose measurements in phantoms. Agreements better than 1.5% or 0.2 cm have been achieved. Applications of the implemented model to clinical cases of brain, head/neck, lung, spine, pancreas and prostate have demonstrated the feasibility and capability of Monte Carlo based inverse planning for clinical IMRT. Dose distributions of selected treatment plans from a commercial non-Monte Carlo based system are evaluated in comparison with Monte Carlo based calculations. Systematic errors of up to 12% in tumor doses and up to 17% in critical structure doses have been observed. The clinical importance of Monte Carlo based
Energy Technology Data Exchange (ETDEWEB)
Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)
2005-07-01
We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.
International Nuclear Information System (INIS)
We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am242m is Am241(n,γ-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am242m concentration.The uncertainties in the inventory of fission products reach up to 30%
Validation the Monte Carlo code RMC with C5G7 benchmark
International Nuclear Information System (INIS)
Highlights: • The RMC code was verified based on the benchmark of C5G7. • Calculation speed of RMC is better than MCNP, especially in the flux tallies. • Eigenvalues calculated by RMC were within 2σ of the benchmark in all cases. • The pin by pin flux tallies of RMC are consistent with MCNP well. - Abstract: RMC (Reactor Monte Carlo code) is a new 3D Monte Carlo neutron transport code being developed by Department of Engineering Physics in Tsinghua University. The current version of RMC is a β version. In this paper, based on 2D and 3D benchmark of C5G7, the criticality calculation capacity of RMC was verified. Comparisons were made between the benchmark eigenvalues and those outputs by the RMC code. The RMC-generated eigenvalues were within two standard deviations of the benchmark and MCNP values in all cases. Additionally, the flux was compared pin by pin between MCNP and RMC. The flux tallies generated by RMC were found to be in well agreement with those from MCNP
Validation of the Monteburns code for criticality calculation of TRIGA reactors
Energy Technology Data Exchange (ETDEWEB)
Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)
2002-07-01
Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)
Li, Junli; Li, Chunyan; Qiu, Rui; Yan, Congchong; Xie, Wenzhang; Wu, Zhen; Zeng, Zhi; Tung, Chuanjong
2015-09-01
The method of Monte Carlo simulation is a powerful tool to investigate the details of radiation biological damage at the molecular level. In this paper, a Monte Carlo code called NASIC (Nanodosimetry Monte Carlo Simulation Code) was developed. It includes physical module, pre-chemical module, chemical module, geometric module and DNA damage module. The physical module can simulate physical tracks of low-energy electrons in the liquid water event-by-event. More than one set of inelastic cross sections were calculated by applying the dielectric function method of Emfietzoglou's optical-data treatments, with different optical data sets and dispersion models. In the pre-chemical module, the ionised and excited water molecules undergo dissociation processes. In the chemical module, the produced radiolytic chemical species diffuse and react. In the geometric module, an atomic model of 46 chromatin fibres in a spherical nucleus of human lymphocyte was established. In the DNA damage module, the direct damages induced by the energy depositions of the electrons and the indirect damages induced by the radiolytic chemical species were calculated. The parameters should be adjusted to make the simulation results be agreed with the experimental results. In this paper, the influence study of the inelastic cross sections and vibrational excitation reaction on the parameters and the DNA strand break yields were studied. Further work of NASIC is underway. PMID:25883312
International Nuclear Information System (INIS)
The method of Monte Carlo simulation is a powerful tool to investigate the details of radiation biological damage at the molecular level. In this paper, a Monte Carlo code called NASIC (Nanodosimetry Monte Carlo Simulation Code) was developed. It includes physical module, pre-chemical module, chemical module, geometric module and DNA damage module. The physical module can simulate physical tracks of low-energy electrons in the liquid water event-by-event. More than one set of inelastic cross sections were calculated by applying the dielectric function method of Emfietzoglou's optical-data treatments, with different optical data sets and dispersion models. In the pre-chemical module, the ionised and excited water molecules undergo dissociation processes. In the chemical module, the produced radiolytic chemical species diffuse and react. In the geometric module, an atomic model of 46 chromatin fibres in a spherical nucleus of human lymphocyte was established. In the DNA damage module, the direct damages induced by the energy depositions of the electrons and the indirect damages induced by the radiolytic chemical species were calculated. The parameters should be adjusted to make the simulation results be agreed with the experimental results. In this paper, the influence study of the inelastic cross sections and vibrational excitation reaction on the parameters and the DNA strand break yields were studied. Further work of NASIC is underway (authors)
Overview of TRIPOLI-4 version 7, Continuous-energy Monte Carlo Transport Code
International Nuclear Information System (INIS)
The TRIPOLI-4 code is used essentially for four major classes of applications: shielding studies, criticality studies, core physics studies, and instrumentation studies. In this updated overview of the Monte Carlo transport code TRIPOLI-4, we list and describe its current main features, including recent developments or extended capacities like effective beta estimation, photo-nuclear reactions or extended mesh tallies. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. TRIPOLI-4 has support for execution in parallel mode. Special features and applications are also presented concerning: 'particles storage', resuming a stopped TRIPOLI-4 run, collision bands, Green's functions, source convergence in criticality mode, and mesh tally
Application of ENDF nuclear data for testing a Monte-Carlo neutron and photon transport code
International Nuclear Information System (INIS)
A Monte-Carlo photon and neutron transport code was developed at OAEP. The code was written in C and C++ languages in an object-oriented programming style. Constructive solid geometry (CSG), rather than combinatorial, was used such that making its input file more readable and recognizable. As the first stage of code validation, data from some ENDF files, in the MCNP's specific format, were used and compared with experimental data. The neutron (from a 300 mCi Am/Be source) attenuation by water was chosen to compare the results. The agreement of the quantity 1/Σ among the calculation from SIPHON and MCNP, and the experiment - which are 10.39 cm, 9.71 cm and 10.25 cm respectively - was satisfactorily well within the experimental uncertainties. These results also agree with the 10.8 cm result of N.M., Mirza, et al. (author)
Parallel Grand Canonical Monte Carlo (ParaGrandMC) Simulation Code
Yamakov, Vesselin I.
2016-01-01
This report provides an overview of the Parallel Grand Canonical Monte Carlo (ParaGrandMC) simulation code. This is a highly scalable parallel FORTRAN code for simulating the thermodynamic evolution of metal alloy systems at the atomic level, and predicting the thermodynamic state, phase diagram, chemical composition and mechanical properties. The code is designed to simulate multi-component alloy systems, predict solid-state phase transformations such as austenite-martensite transformations, precipitate formation, recrystallization, capillary effects at interfaces, surface absorption, etc., which can aid the design of novel metallic alloys. While the software is mainly tailored for modeling metal alloys, it can also be used for other types of solid-state systems, and to some degree for liquid or gaseous systems, including multiphase systems forming solid-liquid-gas interfaces.
Rabie, M.; Franck, C. M.
2016-06-01
We present a freely available MATLAB code for the simulation of electron transport in arbitrary gas mixtures in the presence of uniform electric fields. For steady-state electron transport, the program provides the transport coefficients, reaction rates and the electron energy distribution function. The program uses established Monte Carlo techniques and is compatible with the electron scattering cross section files from the open-access Plasma Data Exchange Project LXCat. The code is written in object-oriented design, allowing the tracing and visualization of the spatiotemporal evolution of electron swarms and the temporal development of the mean energy and the electron number due to attachment and/or ionization processes. We benchmark our code with well-known model gases as well as the real gases argon, N2, O2, CF4, SF6 and mixtures of N2 and O2.
International Nuclear Information System (INIS)
Criticality studies in nuclear fuel cycle are based on Monte Carlo method. These codes use multigroup cross sections which can verify by experimental configurations or by use of reference codes such Tripoli 2. In this Tripoli 2 code nuclear data are errors attached and asked for experimental studies with critical experiences. This is one of the aim of this thesis. To calculate the keff of interacted fissile units we have used the multigroup Monte Carlo code Moret with convergence problems. A new estimator of reactions rates permit to better approximate the neutrons exchange between units and a new importance function has been tested. 2 annexes
Analysing the statistics of group constants generated by Serpent 2 Monte Carlo code
International Nuclear Information System (INIS)
An important topic in Monte Carlo neutron transport calculations is to verify that the statistics of the calculated estimates are correct. Undersampling, non-converged fission source distribution and inter-cycle correlations may result in inaccurate results. In this paper, we study the effect of the number of neutron histories on the distributions of homogenized group constants and assembly discontinuity factors generated using Serpent 2 Monte Carlo code. We apply two normality tests and a so-called “drift-in-mean” test to the batch-wise distributions of selected parameters generated for two assembly types taken from the MIT BEAVRS benchmark. The results imply that in the tested cases the batch-wise estimates of the studied group constants can be regarded as normally distributed. We also show that undersampling is an issue with the calculated assembly discontinuity factors when the number of neutron histories is small. (author)
International Nuclear Information System (INIS)
This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000® problems. These benchmark and scaling studies show promising results
International Nuclear Information System (INIS)
The current basis for conversion coefficients for calibrating individual photon dosimeters in terms of dose equivalents is found in the series of papers by Grosswent. In his calculation the collision kerma inside the phantom is determined by calculation of the energy fluence at the point of interest and the use of the mass energy absorption coefficient. This approximates the local absorbed dose. Other Monte Carlo methods can be sued to provide calculations of the conversion coefficients. Rogers has calculated fluence-to-dose equivalent conversion factors with the Electron-Gamma Shower Version 3, EGS3, Monte Carlo program and produced results similar to Grosswent's calculations. This paper will report on calculations using the Integrated TIGER Series Version 3, ITS3, code to calculate the conversion coefficients in ICRU Tissue and in PMMA. A complete description of the input parameters to the program is given and comparison to previous results is included
Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.
2016-03-01
This work discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package authored at Oak Ridge National Laboratory. Shift has been developed to scale well from laptops to small computing clusters to advanced supercomputers and includes features such as support for multiple geometry and physics engines, hybrid capabilities for variance reduction methods such as the Consistent Adjoint-Driven Importance Sampling methodology, advanced parallel decompositions, and tally methods optimized for scalability on supercomputing architectures. The scaling studies presented in this paper demonstrate good weak and strong scaling behavior for the implemented algorithms. Shift has also been validated and verified against various reactor physics benchmarks, including the Consortium for Advanced Simulation of Light Water Reactors' Virtual Environment for Reactor Analysis criticality test suite and several Westinghouse AP1000® problems presented in this paper. These benchmark results compare well to those from other contemporary Monte Carlo codes such as MCNP5 and KENO.
An object-oriented implementation of a parallel Monte Carlo code for radiation transport
Santos, Pedro Duarte; Lani, Andrea
2016-05-01
This paper describes the main features of a state-of-the-art Monte Carlo solver for radiation transport which has been implemented within COOLFluiD, a world-class open source object-oriented platform for scientific simulations. The Monte Carlo code makes use of efficient ray tracing algorithms (for 2D, axisymmetric and 3D arbitrary unstructured meshes) which are described in detail. The solver accuracy is first verified in testcases for which analytical solutions are available, then validated for a space re-entry flight experiment (i.e. FIRE II) for which comparisons against both experiments and reference numerical solutions are provided. Through the flexible design of the physical models, ray tracing and parallelization strategy (fully reusing the mesh decomposition inherited by the fluid simulator), the implementation was made efficient and reusable.
MCT: a Monte Carlo code for time-dependent neutron thermalization problems
International Nuclear Information System (INIS)
In the Monte Carlo simulation of pulse source experiments, the neutron energy spectrum, spatial distribution and total density may be required for a long time after the pulse. If the assemblies are very small, as often occurs in the cases of interest, sophisticated Monte Carlo techniques must be applied which force neutrons to remain in the system during the time interval investigated. In the MCT code a splitting technique has been applied to neutrons exceeding assigned target times, and we have found that this technique compares very favorably with more usual ones, such as the expected leakage probability, giving large gains in computational time and variance. As an example, satisfactory asymptotic thermal spectra with a neutron attenuation of 10-5 were quickly obtained. (U.S.)
Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries
Ilic, R D; Stankovic, S J
2002-01-01
This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...
ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.
Energy Technology Data Exchange (ETDEWEB)
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2008-04-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.
International Nuclear Information System (INIS)
Application of Monte Carlo method to build spectra library is useful to reduce experiment workload in Prompt Gamma Neutron Activation Analysis (PGNAA). The new Monte Carlo Code MOCA was used to simulate the response spectra of BGO detector for gamma rays from 137Cs, 60Co and neutron induced gamma rays from S and Ti. The results were compared with general code MCNP, show that the agreement of MOCA between simulation and experiment is better than MCNP. This research indicates that building spectra library by Monte Carlo method is feasible. (authors)
Calibration and simulation of a HPGe well detector using Monte Carlo computer code
International Nuclear Information System (INIS)
Monte Carlo methods are often used in simulating physical and mathematical systems. This computer code is a class of computational algorithms that rely on repeated random sampling to compute their results. Because of their reliance on repeated computation of random or pseudo-random numbers, these methods are most suited to calculation by a computer and tend to be used when it is unfeasible or impossible to compute an exact result with a deterministic algorithm. The Monte Carlo method is used to determine a detector's response curves which are difficult to obtain experimentally. It deals with random numbers for the simulation of the decay conditions and angle of incidence at a given energy value, studying, thus, the random behavior of the radiation, providing response and efficiency curves. The MCNP5 computer code provides means to simulate gamma ray detectors and has been used for this work for the 50keV - 2000 keV energy range. The HPGe well detector was simulated with the MCNP5 computer code and compared with experimental data. The dimensions of both dead layer and the transition layer were determined, and the response curve for a particular geometry was then obtained and compared with the experimental results, in order to verify the detector's simulation. Both results were in very good agreement. (author)
Analysis of the tritium breeding ratio benchmark experiments using the Monte Carlo code TRIPOLI-4
International Nuclear Information System (INIS)
Tritium breeding is an essential element of fusion nuclear technology. A tritium breeding ratio greater than unity is necessary for self-sufficient fueling. To simulate the 14 MeV neutron transport in tritium breeding systems from the D-T fusion reaction, the 3D realistic modeling with Monte Carlo code and the point-wise nuclear data are recommended. Continuous-energy TRIPOLI-4 Monte Carlo transport code has been widely used on the radiation shielding, criticality safety, and fission reactor physics. For supporting the ITER TBM (test blanket module) neutronics study with TRIPOLI-4 code, this paper presents the TRIPOLI-4 simulation of TBR (tritium breeding ratio) for six OKTAVIAN spherical assemblies of Osaka University: Li, Li-C, Pb-Li, Pb-Li-C, Be-Li, and Be-Li-C. It also investigates the impact of nuclear data libraries on TBR calculations from ENDF/B-VI.4, ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and FENDL-2.1. In general, TRIPOLI-4 produced satisfactory C/E values. Only beryllium of JEFF-3.1 library introduces higher uncertainties.
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.
Dorval, Eric
2016-01-01
Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient model...
Current status of safety analysis code MARS and uncertainty quantification by Monte-Carlo method
International Nuclear Information System (INIS)
MARS (Multi-dimensional Analysis of Reactor Safety) code has been developed since 1997 for a realistic multi-dimensional thermal-hydraulic system analysis of light water reactor transients. The backbones of MARS are the RELAP5/MOD3.2.1.2 and COBRA-TF codes of USNRC. These two codes were consolidated into a single code by integrating the hydrodynamic solution schemes. New multidimensional TH model has been developed and extended to enable integrated coupled TH analysis through code coupling technique, DLL. The motivation for uncertainty quantification of MARS is considered twofold, 1) to provide “best estimate plus uncertainty” analysis for licensing of commercial power reactor with realistic margins, and 2) to provide support to design and/or validation related analysis for research and production reactors. An assessment of the current LBLOCA uncertainty analysis methodology has been done using data from an integral thermal-hydraulic experiment LOFT L2-5. Monte Carlo calculation has been performed and compared with the tolerance level determined by Wilks formula. The calculation has been done within reasonable CPU time on PC cluster system. Monte-Carlo exercise shows that the 95% upper limit value can be obtained well with 95% confidence level by Wilks formula, although we have to endure 5% risk of PCT under-prediction. The result also shows the statistical fluctuation of limit value using Wilks 1st order is as large as PCT uncertainty itself. The main conclusion is that it is desirable to increase the order of Wilks formula to be higher than the second order to get the reliable safety margin of current design feature. (author)
International Nuclear Information System (INIS)
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
Energy Technology Data Exchange (ETDEWEB)
Cullen, D E
1998-11-22
TART98 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART98 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART98 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART98 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART98 and its data files.
HERMES: a Monte Carlo Code for the Propagation of Ultra-High Energy Nuclei
De Domenico, Manlio; Lyberis, Haris; Settimo, Mariangela
2013-01-01
Although the recent experimental efforts to improve the observation of Ultra-High Energy Cosmic Rays (UHECRs) above $10^{18}$ eV, the origin and the composition of such particles is still unknown. In this work, we present the novel Monte Carlo code (HERMES) simulating the propagation of UHE nuclei, in the energy range between $10^{16}$ and $10^{22}$ eV, accounting for propagation in the intervening extragalactic and Galactic magnetic fields and nuclear interactions with relic photons of the e...
Sampling-Based Nuclear Data Uncertainty Quantification for Continuous Energy Monte Carlo Codes
Zhu, Ting
2015-01-01
The goal of the present PhD research is to establish a methodology of nuclear data uncertainty quantification (NDUQ) for MCNPX, the continuous-energy Monte-Carlo (M-C) code. The high fidelity (continuous-energy treatment and flexible geometry modelling) of MCNPX makes it the choice of routine criticality safety calculations at PSI/LRS, but also raises challenges for NDUQ by conventional sensitivity/uncertainty (S/U) methods. The methodology developed during this PhD research is fundamentally ...
Simulation of clinical X-ray tube using the Monte Carlo Method - PENELOPE code
International Nuclear Information System (INIS)
Breast cancer is the most common type of cancer among women. The main strategy to increase the long-term survival of patients with this disease is the early detection of the tumor, and mammography is the most appropriate method for this purpose. Despite the reduction of cancer deaths, there is a big concern about the damage caused by the ionizing radiation to the breast tissue. To evaluate these measures it was modeled a mammography equipment, and obtained the depth spectra using the Monte Carlo method - PENELOPE code. The average energies of the spectra in depth and the half value layer of the mammography output spectrum. (author)
3-D Monte Carlo neutron-photon transport code JMCT and its algorithms
International Nuclear Information System (INIS)
JMCT Monte Carlo neutron and photon transport code has been developed which is based on the JCOGIN toolbox. JCOGIN includes the geometry operation, tally, the domain decomposition and the parallel computation about particle (MPI) and spatial domain (OpenMP) etc. The viewdata of CAD is equipped in JMCT preprocessor. The full-core pin-mode, which is from Chinese Qinshan-II nuclear power station, is design and simulated by JMCT. The detail pin-power distribution and keff results are shown in this paper. (author)
International Nuclear Information System (INIS)
Two methods of calculating criticality are available in the 3D generalised geometry Monte Carlo particle transport code SPARTAN (Bending and Heffer, 1975). The first is a matrix technique in which the multiplication constant and source distribution of the system under study are calculated from estimates of fission probabilities and the second a method in which the multiplication constant is inferred from estimates of changes in neutron population over a number of neutron generations. Modifications are described which have been made to the way in which these methods are used in SPARTAN in order to improve the efficiency of criticality calculations. (author)
International Nuclear Information System (INIS)
Aiming at validation of depletion codes against WWER-440 spent fuel data some calculations of isotope composition of WWER-440 spent fuel assembly have been carried out by the NESSEL-NUKO code system. The initial data and data for the comparisons were taken from the ISTS burn up credit project data, recently published in the ISTC report 'Radiochemical Assays of Irradiated WWER-440 Fuel for Use in Spent Fuel Burnup Credit Activities. The specific work scope included the destructive assay (DA) of spent fuel assembly rod segments with an - -38.5 MWd/KgU burn up from a single WWER-440 fuel assembly from the Novovorenezh reactor in Russia (Authors)
MULTI-KENO: a Monte Carlo code for criticality safety analysis
International Nuclear Information System (INIS)
Modifying the Monte Carlo code KENO-IV, the MULTI-KENO code was developed for criticality safety analysis. The following functions were added to the code; (1) to divide a system into many sub-systems named super boxes where the size of box types in each super box can be selected independently, (2) to output graphical view of a system for examining geometrical input data, (3) to solve fixed source problems, (4) to permit intersection of core boundaries and inner geometries, (5) to output ANISN type neutron balance table. With the above function (1), many cases which had to be applied a general geometry option of KENO-IV, became to be treated as box type geometry. In such a case, input data became simpler and required computer time became shorter than those of KENO-IV. This code is now available for the FACOM-M200 computer and the CDC 6600 computer. This report is a computer code manual for MULTI-KENO. (author)
The use of Monte Carlo radiation transport codes in radiation physics and dosimetry
CERN. Geneva; Ferrari, Alfredo; Silari, Marco
2006-01-01
Transport and interaction of electromagnetic radiation Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. In these codes, photon transport is simulated by using the detailed scheme, i.e., interaction by interaction. Detailed simulation is easy to implement, and the reliability of the results is only limited by the accuracy of the adopted cross sections. Simulations of electron and positron transport are more difficult, because these particles undergo a large number of interactions in the course of their slowing down. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interacti...
Simulation of density curve for slim borehole using the Monte Carlo code MCNPX
International Nuclear Information System (INIS)
Borehole logging for formation density has been an important geophysical measurement in oil industry. For calibration of the Gamma Ray nuclear logging tool, numerous rock models of different lithology and densities are necessary. However, the full success of this calibration process is determined by a reliable benchmark, where the complete and precise chemical composition of the standards is necessary. Simulations using the Monte Carlo MCNP have been widely employed in well logging application once it serves as a low-cost substitute for experimental test pits, as well as a means for obtaining data that are difficult to obtain experimentally. Considering this, the purpose of this work is to use the code MCNP to obtain density curves for slim boreholes using Gamma Ray logging tools. For this, a Slim Density Gamma Probe, named TRISONDR, and a 100 mCi Cs-137 gamma source has been modeled with the new version of MCNP code MCNPX. (author)
Simulation of density curve for slim borehole using the Monte Carlo code MCNPX
Energy Technology Data Exchange (ETDEWEB)
Souza, Edmilson Monteiro de; Silva, Ademir Xavier da; Lopes, Ricardo Tadeu, E-mail: emonteiro@nuclear.ufrj.b, E-mail: ademir@nuclear.ufrj.b, E-mail: ricardo@lin.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Correa, Samanda Cristine Arruda, E-mail: scorrea@nuclear.ufrj.b [Centro Universitario Estadual da Zona Oeste (CCMAT/UEZO), Rio de Janeiro, RJ (Brazil); Lima, Inaya C.B., E-mail: inaya@lin.ufrj.b [Universidade Estadual do Rio de Janeiro (IPRJ/UERJ) Nova Friburgo, Rio de Janeiro, RJ (Brazil). Instituto Politecnico do Rio de Janeiro; Rocha, Paula L.F., E-mail: ferrucio@acd.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ) RJ (Brazil). Dept. de Geologia
2010-07-01
Borehole logging for formation density has been an important geophysical measurement in oil industry. For calibration of the Gamma Ray nuclear logging tool, numerous rock models of different lithology and densities are necessary. However, the full success of this calibration process is determined by a reliable benchmark, where the complete and precise chemical composition of the standards is necessary. Simulations using the Monte Carlo MCNP have been widely employed in well logging application once it serves as a low-cost substitute for experimental test pits, as well as a means for obtaining data that are difficult to obtain experimentally. Considering this, the purpose of this work is to use the code MCNP to obtain density curves for slim boreholes using Gamma Ray logging tools. For this, a Slim Density Gamma Probe, named TRISOND{sup R}, and a 100 mCi Cs-137 gamma source has been modeled with the new version of MCNP code MCNPX. (author)
An EGS4 Monte Carlo user code for radiation therapy planning
International Nuclear Information System (INIS)
An EGS4 Monte Carlo user code (the UCRTP code) with voxel geometry has been developed as a prototype of the dose calculation engine for radiation therapy planning. A series of dose calculations for photon beam irradiation to a simplified heterogenous voxel phantom of a lung cancer patient has shown that significant build-up in lung tumor and build-down in surrounding normal lung tissue region exist due to the heterogeneity of the media and small field size. Most of the heterogeneity correction algorithms employed by the current commercial treatment planning systems are not satisfactory enough to account for the build-up/down. Since the commercial systems may significantly underestimate the dose in normal lung tissues, sufficient verification and quality assurance of the radiation therapy planning is needed especially in the lung cancer treatment. (author)
Shielding properties of iron at high energy proton accelerators studied by a Monte Carlo code
International Nuclear Information System (INIS)
Shielding properties of a lateral iron shield and of iron and concrete shields at angles between 5deg and 30deg are studied by means of the Monte Carlo program FLUNEV (DESY-D3 version of the FLUKA code extended for emission and transport of low energy neutrons). The following quantities were calculated for a high energy proton beam hitting an extended iron target: total and partial dose equivalents, attenuation coefficients, neutron spectra, star densities (compared also with the CASIM code) and quality factors. The dependence of the dose equivalent on the energy of primary protons, the effect of a concrete layer behind a lateral iron shielding and the total number of neutrons produced in the target were also estimated. (orig.)
Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP
International Nuclear Information System (INIS)
Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)
Speedup of MCACE, a Monte Carlo code for evaluation of shielding safety, by parallel computer, 1
International Nuclear Information System (INIS)
In order to improve the accuracy of shielding analysis, we have modified MCACE, a Monte Carlo code for shielding analysis, to be able to execute on a parallel computer. The suitable algorithms for efficient paralleling has been investigated by static and dynamic analyses of the code. This includes a strategy where new units of batches are assigned to the idling cells dynamically during the execution. The efficiency of paralleling has been measured by using a simulator of a parallel computer. It is found that the load factor of all cells reached nearly 100%, and consequently, it can be said that the most effective paralleling has been achieved. The simulator has estimated the effect of paralleling as the speedup of 7.13 times when a sample problem of 8 batches, 400 particles per one batch, is loaded on parallel computer equipped with 8 cells. (author)
A 3DHZETRN Code in a Spherical Uniform Sphere with Monte Carlo Verification
Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.
2014-01-01
The computationally efficient HZETRN code has been used in recent trade studies for lunar and Martian exploration and is currently being used in the engineering development of the next generation of space vehicles, habitats, and extra vehicular activity equipment. A new version (3DHZETRN) capable of transporting High charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation is under development. In the present report, new algorithms for light ion and neutron propagation with well-defined convergence criteria in 3D objects is developed and tested against Monte Carlo simulations to verify the solution methodology. The code will be available through the software system, OLTARIS, for shield design and validation and provides a basis for personal computer software capable of space shield analysis and optimization.
Implementation of mathematical phantom of hand and forearm in GEANT4 Monte Carlo code
International Nuclear Information System (INIS)
In this work, the implementation of a hand and forearm Geant4 phantom code, for further evaluation of occupational exposure of ends of the radionuclides decay manipulated during procedures involving the use of injection syringe. The simulation model offered by Geant4 includes a full set of features, with the reconstruction of trajectories, geometries and physical models. For this work, the values calculated in the simulation are compared with the measurements rates by thermoluminescent dosimeters (TLDs) in physical phantom REMAB®. From the analysis of the data obtained through simulation and experimentation, of the 14 points studied, there was a discrepancy of only 8.2% of kerma values found, and these figures are considered compatible. The geometric phantom implemented in Geant4 Monte Carlo code was validated and can be used later for the evaluation of doses at ends
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed
Energy Technology Data Exchange (ETDEWEB)
Hart, S. W. D. [University of Tennessee, Knoxville (UTK); Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK); Celik, Cihangir [ORNL; Leal, Luiz C [ORNL
2014-01-01
For many Monte Carlo codes cross sections are generally only created at a set of predetermined temperatures. This causes an increase in error as one moves further and further away from these temperatures in the Monte Carlo model. This paper discusses recent progress in the Scale Monte Carlo module KENO to create problem dependent, Doppler broadened, cross sections. Currently only broadening the 1D cross sections and probability tables is addressed. The approach uses a finite difference method to calculate the temperature dependent cross-sections for the 1D data, and a simple linear-logarithmic interpolation in the square root of temperature for the probability tables. Work is also ongoing to address broadening theS (alpha , beta) tables. With the current approach the temperature dependent cross sections are Doppler broadened before transport starts, and, for all but a few isotopes, the impact on cross section loading is negligible. Results can be compared with those obtained by using multigroup libraries, as KENO currently does interpolation on the multigroup cross sections to determine temperature dependent cross-sections. Current results compare favorably with these expected results.
Nelson, Adam
Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system
Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes.
Pinsky, L S; Wilson, T L; Ferrari, A; Sala, P; Carminati, F; Brun, R
2001-01-01
This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be useful in the design and analysis of experiments such as ACCESS (Advanced Cosmic-ray Composition Experiment for Space Station), which is an Office of Space Science payload currently under evaluation for deployment on the International Space Station (ISS). FLUKA will be significantly improved and tailored for use in simulating space radiation in four ways. First, the additional physics not presently within the code that is necessary to simulate the problems of interest, namely the heavy ion inelastic processes, will be incorporated. Second, the internal geometry package will be replaced with one that will substantially increase the calculation speed as well as simplify the data input task. Third, default incident flux packages that include all of the different space radiation sources of interest will be included. Finally, the user interface and internal data structure will be melded together with ROOT, the object-oriented data analysis infrastructure system. Beyond
International Nuclear Information System (INIS)
The general purpose Monte Carlo code MCNP4 has been implemented on the Fujitsu AP1000 distributed memory highly parallel computer. Parallelization techniques developed and studied are reported. A shielding analysis function of the MCNP4 code is parallelized in this study. A technique to map a history to each processor dynamically and to map control process to a certain processor was applied. The efficiency of parallelized code is up to 80% for a typical practical problem with 512 processors. These results demonstrate the advantages of a highly parallel computer to the conventional computers in the field of shielding analysis by Monte Carlo method. (orig.)
Coded aperture coherent scatter imaging for breast cancer detection: a Monte Carlo evaluation
Lakshmanan, Manu N.; Morris, Robert E.; Greenberg, Joel A.; Samei, Ehsan; Kapadia, Anuj J.
2016-03-01
It is known that conventional x-ray imaging provides a maximum contrast between cancerous and healthy fibroglandular breast tissues of 3% based on their linear x-ray attenuation coefficients at 17.5 keV, whereas coherent scatter signal provides a maximum contrast of 19% based on their differential coherent scatter cross sections. Therefore in order to exploit this potential contrast, we seek to evaluate the performance of a coded- aperture coherent scatter imaging system for breast cancer detection and investigate its accuracy using Monte Carlo simulations. In the simulations we modeled our experimental system, which consists of a raster-scanned pencil beam of x-rays, a bismuth-tin coded aperture mask comprised of a repeating slit pattern with 2-mm periodicity, and a linear-array of 128 detector pixels with 6.5-keV energy resolution. The breast tissue that was scanned comprised a 3-cm sample taken from a patient-based XCAT breast phantom containing a tomosynthesis- based realistic simulated lesion. The differential coherent scatter cross section was reconstructed at each pixel in the image using an iterative reconstruction algorithm. Each pixel in the reconstructed image was then classified as being either air or the type of breast tissue with which its normalized reconstructed differential coherent scatter cross section had the highest correlation coefficient. Comparison of the final tissue classification results with the ground truth image showed that the coded aperture imaging technique has a cancerous pixel detection sensitivity (correct identification of cancerous pixels), specificity (correctly ruling out healthy pixels as not being cancer) and accuracy of 92.4%, 91.9% and 92.0%, respectively. Our Monte Carlo evaluation of our experimental coded aperture coherent scatter imaging system shows that it is able to exploit the greater contrast available from coherently scattered x-rays to increase the accuracy of detecting cancerous regions within the breast.
Implementation of the probability table method in a continuous-energy Monte Carlo code system
Energy Technology Data Exchange (ETDEWEB)
Sutton, T.M.; Brown, F.B. [Lockheed Martin Corp., Schenectady, NY (United States)
1998-10-01
RACER is a particle-transport Monte Carlo code that utilizes a continuous-energy treatment for neutrons and neutron cross section data. Until recently, neutron cross sections in the unresolved resonance range (URR) have been treated in RACER using smooth, dilute-average representations. This paper describes how RACER has been modified to use probability tables to treat cross sections in the URR, and the computer codes that have been developed to compute the tables from the unresolved resonance parameters contained in ENDF/B data files. A companion paper presents results of Monte Carlo calculations that demonstrate the effect of the use of probability tables versus the use of dilute-average cross sections for the URR. The next section provides a brief review of the probability table method as implemented in the RACER system. The production of the probability tables for use by RACER takes place in two steps. The first step is the generation of probability tables from the nuclear parameters contained in the ENDF/B data files. This step, and the code written to perform it, are described in Section 3. The tables produced are at energy points determined by the ENDF/B parameters and/or accuracy considerations. The tables actually used in the RACER calculations are obtained in the second step from those produced in the first. These tables are generated at energy points specific to the RACER calculation. Section 4 describes this step and the code written to implement it, as well as modifications made to RACER to enable it to use the tables. Finally, some results and conclusions are presented in Section 5.
Calculation of the CB1 burnup credit benchmark reaction rates with MCNP4B
International Nuclear Information System (INIS)
The first calculational VVER-440 burnup credit benchmark CB1 in 1996. VTT Energy participated in the calculation of the CB1 benchmark with three different codes: CASMO-4, KENO-VI and MCNP4B. However, the reaction rates and the fission ν were calculated only with CASMO-4. Now, the neutron absorption and production reaction rates and the fission ν values have been calculated at VTT Energy with the MCNP4B Monte Carlo code using the ENDF60 neutron data library. (author)
Monte Carlo simulation of MOSFET dosimeter for electron backscatter using the GEANT4 code.
Chow, James C L; Leung, Michael K K
2008-06-01
The aim of this study is to investigate the influence of the body of the metal-oxide-semiconductor field effect transistor (MOSFET) dosimeter in measuring the electron backscatter from lead. The electron backscatter factor (EBF), which is defined as the ratio of dose at the tissue-lead interface to the dose at the same point without the presence of backscatter, was calculated by the Monte Carlo simulation using the GEANT4 code. Electron beams with energies of 4, 6, 9, and 12 MeV were used in the simulation. It was found that in the presence of the MOSFET body, the EBFs were underestimated by about 2%-0.9% for electron beam energies of 4-12 MeV, respectively. The trend of the decrease of EBF with an increase of electron energy can be explained by the small MOSFET dosimeter, mainly made of epoxy and silicon, not only attenuated the electron fluence of the electron beam from upstream, but also the electron backscatter generated by the lead underneath the dosimeter. However, this variation of the EBF underestimation is within the same order of the statistical uncertainties as the Monte Carlo simulations, which ranged from 1.3% to 0.8% for the electron energies of 4-12 MeV, due to the small dosimetric volume. Such small EBF deviation is therefore insignificant when the uncertainty of the Monte Carlo simulation is taken into account. Corresponding measurements were carried out and uncertainties compared to Monte Carlo results were within +/- 2%. Spectra of energy deposited by the backscattered electrons in dosimetric volumes with and without the lead and MOSFET were determined by Monte Carlo simulations. It was found that in both cases, when the MOSFET body is either present or absent in the simulation, deviations of electron energy spectra with and without the lead decrease with an increase of the electron beam energy. Moreover, the softer spectrum of the backscattered electron when lead is present can result in a reduction of the MOSFET response due to stronger
ACCEPT: three-dimensional electron/photon Monte Carlo transport code using combinatorial geometry
International Nuclear Information System (INIS)
The ACCEPT code provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through three-dimensional multimaterial geometries described by the combinational method. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. ACCEPT combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. The ACCEPT code is currently running on the CDC-7600 (66000) where the bulk of the cross-section data and the statistical variables are stored in Large Core Memory
International Nuclear Information System (INIS)
The numerical simulation of the dynamics of fast ions coming from neutral beam injection (NBI) heating is an important task in fusion devices, since these particles are used as sources to heat and fuel the plasma and their uncontrolled losses can damage the walls of the reactor. This paper shows a new application that simulates these dynamics on the grid: FastDEP. FastDEP plugs together two Monte Carlo codes used in fusion science, namely FAFNER2 and ISDEP, and add new functionalities. Physically, FAFNER2 provides the fast ion initial state in the device while ISDEP calculates their evolution in time; as a result, the fast ion distribution function in TJ-II stellerator has been estimated, but the code can be used on any other device. In this paper a comparison between the physics of the two NBI injectors in TJ-II is presented, together with the differences between fast ion confinement and the driven momentum in the two cases. The simulations have been obtained using Montera, a framework developed for achieving grid efficient executions of Monte Carlo applications. (paper)
International Nuclear Information System (INIS)
This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in our neutron transport problems, we briefly describe the work we have done in order to get a vector code. Vectorization principles will be presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples will be presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example we propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion we prove that Monte Carlo algorithms are very well suited to future vector and parallel computers. (orig.)
OpenMC: A state-of-the-art Monte Carlo code for research and development
International Nuclear Information System (INIS)
Highlights: • OpenMC is an open source Monte Carlo particle transport code. • Solid geometry and continuous-energy physics allow high-fidelity simulations. • Development has focused on high performance and modern I/O techniques. • OpenMC is capable of scaling up to hundreds of thousands of processors. • Other features include plotting, CMFD acceleration, and variance reduction. - Abstract: This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes
Point KENO V.a: A continuous-energy Monte Carlo code for transport applications
International Nuclear Information System (INIS)
KENO V.a is a multigroup Monte Carlo code that solves the Boltzmann transport equation and is used extensively in the criticality safety community to calculate the effective multiplication factor of systems with fissionable material. In this work, a continuous-energy or pointwise version of KENO V.a has been developed by first designing a new continuous-energy cross-section format and then by developing the appropriate Monte Carlo transport procedures to sample the new cross-section format. In order to generate pointwise cross sections for a test library, a series of cross-section processing modules were developed and used to process 50 ENDF/B-VI Release 7 nuclides for the test library. Once the cross-section processing procedures were in place, a continuous-energy version of KENO V.a was developed and tested by calculating 46 test cases that include critical and calculational benchmark problems. The Point KENO-calculated results for the test problems are in agreement with calculated results obtained with the multigroup version of KENO V.a and MCNP4C. Based on the calculated results with the prototypic cross-section library, a continuous-energy version of the KENO V.a code has been successfully developed and demonstrated for modeling systems with fissionable material. (authors)
Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms
Directory of Open Access Journals (Sweden)
H A Nedaie
2013-01-01
Full Text Available Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous phantom and around inhomogeneities. Different types of phantoms ranging in complexity were used; namely, a homogeneous water phantom and phantoms made of polymethyl methacrylate slabs containing different-sized, low- and high-density inserts of heterogeneous materials. Electron beams with 8 and 15 MeV nominal energy generated by an Elekta Synergy linear accelerator were investigated. Measurements were performed for a 10 cm × 10 cm applicator at a source-to-surface distance of 100 cm. Individual parts of the beam-defining system were introduced into the simulation one at a time in order to show their effect on depth doses. In contrast to the first scattering foil, the secondary scattering foil, X and Y jaws and applicator provide up to 5% of the dose. A 2%/2 mm agreement between MCNP and measurements was found in the homogenous phantom, and in the presence of heterogeneities in the range of 1-3%, being generally within 2% of the measurements for both energies in a "complex" phantom. A full-component simulation is necessary in order to obtain a realistic model of the beam. The MCNP4C results agree well with the measured electron dose distributions.
International Nuclear Information System (INIS)
The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)
Spent fuel pool storage calculations using the ISOCRIT burnup credit tool
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Highlights: ► Depletion isotopics are needed for burnup credit in spent fuel pool analyses. ► We developed ISOCRIT to generate the isotopics using conservative depletion assumptions. ► ISOCRIT works in an automated fashion passing data between lattice physics and 3D Monte Carlo codes. ► Analyses to assess the impact of different depletion parameters on the reactivity of the spent fuel in pool conditions. - Abstract: In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse’s state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.
Monali-Rev.1: a Monte Carlo code for analysing fuel assemblies of nuclear reactors
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MONALI-Rev.1 is a multigroup Monte Carlo program developed on ND computers for analysing fuel assemblies of nuclear reactors. This version of the code is flexibly dimensioned so that the allowed size of a problem is limited only by the total data storage required. The code can read multigroup data for various nuclides directly from WIMS multigroup (69/27) cross section sets. Most of the input data, with the exception of cross sections, if needed, are read in free format. The treatment of anisotropy (up to P1 at present) may be in selective mixtures. The input to the geometry module has been simplified. The code has flexibility in the definition of regions. The results calculated by the code include Keff, multigroup leakages and absorptions, group- and region-dependent fluxes. The multigroup leakages are calculated for each outer-most surface. Statistical confidence limits are also assigned to the results. In the end frequency distributions are found for the multiplication factor and optionally a normality test is also performed on the multiplication factors. (author). 8 refs., 4 figs., 2 tabs., 2 appendices
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RodrIguez, M L [Centro Medico Paitilla. Calle 53 y ave Balboa, Paitilla (Panama)], E-mail: milrocas@gmail.com
2008-09-07
In this work we present PENLINAC, a code package developed to facilitate the use of the Monte Carlo code PENELOPE for the simulation of therapeutic beams, including high-energy electrons, photons and {sup 60}Co beams. The code simplifies the creation of the treatment machine geometry, allowing the modeling of their components from elementary geometric bodies and their further conversion to the quadric functions-based structure handled by PENELOPE. The code is implemented in various subroutines that allow the user to handle several models of radiation sources and phase spaces. The phase spaces are not part of the geometry and can store many variables of the particle in a relatively small data space. The set of subroutines does not alter the PENELOPE algorithms; thus, the main program implemented by the user can maintain its kind-of-particle-independent structure. A support program can handle and analyze the phase spaces to generate, among others, last interaction maps and probability distributions that can be used as sources in simulation. Results from simulations of a Clinac linear accelerator head are presented in order to demonstrate the package capabilities. Dose distributions calculated in a water phantom for a variety of beams of this accelerator showed good agreement with measurements.
Load balancing in highly parallel processing of Monte Carlo code for particle transport
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In parallel processing of Monte Carlo(MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)
Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor
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A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data
Activity ratio measurement and burnup analysis for high burnup PWR fuels
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Applying burnup credit to spent fuel transportation and storage system is beneficial. To take burnup credit to criticality safety design for a spent fuel transportation cask and storage rack, the burnup of target fuel assembly based on core management data must be confirmed by experimental methods. Activity ratio method, in which measured the ratio of the activity of a nuclide to that of another, is one of the ways to confirm burnup history. However, there is no public data of gamma-ray spectrum from high burnup fuels and validation of depletion calculation codes is not sufficient in the evaluation of the burnup or nuclide inventories. In this study, applicability evaluation of activity ratio method was carried out for high burnup fuel samples taken from PWR lead use assembly. In the gamma-ray measurement experiments, energy spectrum was taken in the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA), and 134Cs/137Cs and 154Eu/137Cs activity ratio were obtained. With the MVP-BURN code, the activity ratios were calculated by depletion calculation tracing the operation history. As a result, 134Cs/137Cs and 154Eu/137Cs activity ratios for UO2 fuel samples show good agreements and the activity ratio method has good applicability to high burnup fuels. 154Eu/134Cs activity ratio for Gd2O3+UO2 fuels also shows good agreements between calculation results and experimental results as well as the activity ratio for UO2 fuels. It also becomes clear that it is necessary to pay attention to not only burnup but also axial burnup distribution history when confirming the burnup of UO2+Gd2O3 fuel with 134Cs/137Cs activity ratios. (author)
Analysis of high burnup fuel safety issues
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Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S
2000-12-01
Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.
Analysis of high burnup fuel safety issues
International Nuclear Information System (INIS)
Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development
MCNP: a general Monte Carlo code for neutron and photon transport
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The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
SU-E-T-578: MCEBRT, A Monte Carlo Code for External Beam Treatment Plan Verifications
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Chibani, O; Ma, C [Fox Chase Cancer Center, Philadelphia, PA (United States); Eldib, A [Fox Chase Cancer Center, Philadelphia, PA (United States); Al-Azhar University, Cairo (Egypt)
2014-06-01
Purpose: Present a new Monte Carlo code (MCEBRT) for patient-specific dose calculations in external beam radiotherapy. The code MLC model is benchmarked and real patient plans are re-calculated using MCEBRT and compared with commercial TPS. Methods: MCEBRT is based on the GEPTS system (Med. Phys. 29 (2002) 835–846). Phase space data generated for Varian linac photon beams (6 – 15 MV) are used as source term. MCEBRT uses a realistic MLC model (tongue and groove, rounded ends). Patient CT and DICOM RT files are used to generate a 3D patient phantom and simulate the treatment configuration (gantry, collimator and couch angles; jaw positions; MLC sequences; MUs). MCEBRT dose distributions and DVHs are compared with those from TPS in absolute way (Gy). Results: Calculations based on the developed MLC model closely matches transmission measurements (pin-point ionization chamber at selected positions and film for lateral dose profile). See Fig.1. Dose calculations for two clinical cases (whole brain irradiation with opposed beams and lung case with eight fields) are carried out and outcomes are compared with the Eclipse AAA algorithm. Good agreement is observed for the brain case (Figs 2-3) except at the surface where MCEBRT dose can be higher by 20%. This is due to better modeling of electron contamination by MCEBRT. For the lung case an overall good agreement (91% gamma index passing rate with 3%/3mm DTA criterion) is observed (Fig.4) but dose in lung can be over-estimated by up to 10% by AAA (Fig.5). CTV and PTV DVHs from TPS and MCEBRT are nevertheless close (Fig.6). Conclusion: A new Monte Carlo code is developed for plan verification. Contrary to phantombased QA measurements, MCEBRT simulate the exact patient geometry and tissue composition. MCEBRT can be used as extra verification layer for plans where surface dose and tissue heterogeneity are an issue.
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There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)
International Nuclear Information System (INIS)
There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.
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As The Monte Carlo (MC) particle transport analysis for a complex system such as research reactor, accelerator, and fusion facility may require accurate modeling of the complicated geometry. Its manual modeling by using the text interface of a MC code to define the geometrical objects is tedious, lengthy and error-prone. This problem can be overcome by taking advantage of modeling capability of the computer aided design (CAD) system. There have been two kinds of approaches to develop MC code systems utilizing the CAD data: the external format conversion and the CAD kernel imbedded MC simulation. The first approach includes several interfacing programs such as McCAD, MCAM, GEOMIT etc. which were developed to automatically convert the CAD data into the MCNP geometry input data. This approach makes the most of the existing MC codes without any modifications, but implies latent data inconsistency due to the difference of the geometry modeling system. In the second approach, a MC code utilizes the CAD data for the direct particle tracking or the conversion to an internal data structure of the constructive solid geometry (CSG) and/or boundary representation (B-rep) modeling with help of a CAD kernel. MCNP-BRL and OiNC have demonstrated their capabilities of the CAD-based MC simulations. Recently we have developed a CAD-based geometry processing module for the MC particle simulation by using the OpenCASCADE (OCC) library. In the developed module, CAD data can be used for the particle tracking through primitive CAD surfaces (hereafter the CAD-based tracking) or the internal conversion to the CSG data structure. In this paper, the performances of the text-based model, the CAD-based tracking, and the internal CSG conversion are compared by using an in-house MC code, McSIM, equipped with the developed CAD-based geometry processing module
Verification of Monte Carlo transport codes: FLUKA, MARS and SHIELD-A
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Monte Carlo transport codes like FLUKA, MARS and SHIELD are widely used for the estimation of radiation hazards in accelerator facilities. Accurate simulations are especially important with increasing energies and intensities of the machines. As the physical models implied in the codes are being constantly further developed, the verification is needed to make sure that the simulations give reasonable results. We report on the verification of electronic stopping modules and the verification of nuclide production modules of the codes. The verification of electronic stopping modules is based on the results of irradiation of stainless steel, copper and aluminum by 500 MeV/u and 950 MeV/u uranium ions. The stopping ranges achieved experimentally are compared with the simulated ones. The verification of isotope production modules is done via comparing the experimental depth profiles of residual activity (aluminum targets were irradiated by 500 MeV/u and 950 MeV/u uranium ions) with the results of simulations. Correspondences and discrepancies between the experiment and the simulations are discussed.
Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes
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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)
Verification of Monte Carlo transport codes: FLUKA, MARS and SHIELD-A
Energy Technology Data Exchange (ETDEWEB)
Chetvertkova, Vera [IAP, J. W. Goethe-University, Frankfurt am Main (Germany); GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Mustafin, Edil; Strasik, Ivan [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Ratzinger, Ulrich [IAP, J. W. Goethe-University, Frankfurt am Main (Germany); Latysheva, Ludmila; Sobolevskiy, Nikolai [Institute for Nuclear Research RAS, Moscow (Russian Federation)
2011-07-01
Monte Carlo transport codes like FLUKA, MARS and SHIELD are widely used for the estimation of radiation hazards in accelerator facilities. Accurate simulations are especially important with increasing energies and intensities of the machines. As the physical models implied in the codes are being constantly further developed, the verification is needed to make sure that the simulations give reasonable results. We report on the verification of electronic stopping modules and the verification of nuclide production modules of the codes. The verification of electronic stopping modules is based on the results of irradiation of stainless steel, copper and aluminum by 500 MeV/u and 950 MeV/u uranium ions. The stopping ranges achieved experimentally are compared with the simulated ones. The verification of isotope production modules is done via comparing the experimental depth profiles of residual activity (aluminum targets were irradiated by 500 MeV/u and 950 MeV/u uranium ions) with the results of simulations. Correspondences and discrepancies between the experiment and the simulations are discussed.
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The conversion coefficients, H'(d,α)/φ, for monoenergetic positrons and positron-emitting radionuclides were calculated by using the user code UCICRPM of the Monte Carlo code EGS5 to estimate the radiation dose for medical staff involved in positron emission tomography examinations. From these coefficients, the dose equivalent rates per unit activity at 0.07 and 10 mm depths in a soft tissue for a straight-line source of 2-deoxy-2-[18F]fluoro-d-glucose (18F-FDG) were calculated by using the developed user code UCF18DOSE. The dose equivalent rates per unit activity at 0.07 and 10 mm depths were measured by using a personal dosemeter (DOSE 3) under the same conditions as those considered in the calculation. The calculated dose equivalent rates per unit activity at 0.07 and 10 mm depths were 0.116 and 0.0352 pSv min-1 Bq-1, respectively, at 20 cm from the 18F-FDG injection tube. (authors)
Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation
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Hugot, F.X.; Lee, Y.K.; Malvagi, F. [CEA - DEN/DANS/DM2S/SERMA/LTSD, Saclay (France)
2008-07-01
TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)
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A reliable Monte Carlo based investigation of ion chambers in medical physics problems depends on the accuracy of the charged particle transport and implementations of the condensed history technique. Improper handling of media interfaces can lead to anomalous results or 'interface artefacts'. This work presents a rigorous investigation of the electron transport algorithm in the general purpose Monte Carlo (MC) code FLUKA (2008.3b.1). A 'Fano test' was implemented in order to benchmark the accuracy of the algorithm. Furthermore, the calculation of wall perturbation factors pwall of a Roos type chamber irradiated by electrons were performed and compared with values based on the EGSnrc MC code
Gas bremsstrahlung studies for medium energy electron storage rings using FLUKA Monte Carlo code
Sahani, Prasanta Kumar; Haridas, G.; Sinha, Anil K.; Hannurkar, P. R.
2016-02-01
Gas bremsstrahlung is generated due to the interaction of the stored electron beam with residual gas molecules of the vacuum chamber in a storage ring. As the opening angle of the bremsstrahlung is very small, the scoring area used in Monte Carlo simulation plays a dominant role in evaluating the absorbed dose. In the present work gas bremsstrahlung angular distribution and absorbed dose for the energies ranging from 1 to 5 GeV electron storage rings are studied using the Monte Carlo code, FLUKA. From the study, an empirical formula for gas bremsstrahlung dose estimation was deduced. The results were compared with the data obtained from reported experimental values. The results obtained from simulations are found to be in very good agreement with the reported experimental data. The results obtained are applied in estimating the gas bremsstrahlung dose for 2.5 GeV synchrotron radiation source, Indus-2 at Raja Ramanna Centre for Advanced Technology, India. The paper discusses the details of the simulation and the results obtained.
Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code
International Nuclear Information System (INIS)
A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (keff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of keff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of keff from MCNP5 were obtained when the control fuel rods were fully inserted, the keff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of keff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)
Energy Technology Data Exchange (ETDEWEB)
Matthew Ellis; Derek Gaston; Benoit Forget; Kord Smith
2011-07-01
In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes. An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.
Characterization of 60Co dose distribution using BEAMnrc Monte Carlo code
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In this study BEAMnrc based on EGSnrc as Monte Carlo code has been used for modeling and simulating 60Co machine in radioisotope centre of Khartoum (RICK), Two fields size ( 5 cm x 5 cm and 35 cm x 35 cm), were been studied, to define the characterization of 60Co machine and to investigate the effect of increasing the surface to skin distance (SSD) on the 60Co machine properties, e.g.; beam profile and percentage depth dose (Pdd). For the narrow field size there is a small change observed in the curves representing beam profile and the percentage depth dose when increasing the distance by 5 cm, for the wide fi ld size there relatively clear different in curves. The study results been compared with other previous studies and clear consistence observed. (Author)
OpenMC: a state-of-the-Art Monte Carlo code for research and development
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This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes. (authors)
Application of a Monte Carlo Penelope code at diverse dosimetric problems in radiotherapy
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In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: 192 Ir, 125 I, 106 Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)
The FLUKA code for application of Monte Carlo methods to promote high precision ion beam therapy
Parodi, K; Cerutti, F; Ferrari, A; Mairani, A; Paganetti, H; Sommerer, F
2010-01-01
Monte Carlo (MC) methods are increasingly being utilized to support several aspects of commissioning and clinical operation of ion beam therapy facilities. In this contribution two emerging areas of MC applications are outlined. The value of MC modeling to promote accurate treatment planning is addressed via examples of application of the FLUKA code to proton and carbon ion therapy at the Heidelberg Ion Beam Therapy Center in Heidelberg, Germany, and at the Proton Therapy Center of Massachusetts General Hospital (MGH) Boston, USA. These include generation of basic data for input into the treatment planning system (TPS) and validation of the TPS analytical pencil-beam dose computations. Moreover, we review the implementation of PET/CT (Positron-Emission-Tomography / Computed- Tomography) imaging for in-vivo verification of proton therapy at MGH. Here, MC is used to calculate irradiation-induced positron-emitter production in tissue for comparison with the +-activity measurement in order to infer indirect infor...
MCPT: A Monte Carlo code for simulation of photon transport in tomographic scanners
International Nuclear Information System (INIS)
MCPT is a special-purpose Monte Carlo code designed to simulate photon transport in tomographic scanners. Variance reduction schemes and sampling games present in MCPT were selected to characterize features common to most tomographic scanners. Combined splitting and biasing (CSB) games are used to systematically sample important detection pathways. An efficient splitting game is used to tally particle energy deposition in detection zones. The pulse height distribution of each detector can be found by convolving the calculated energy deposition distribution with the detector's resolution function. A general geometric modelling package, HERMETOR, is used to describe the geometry of the tomographic scanners and provide MCPT information needed for particle tracking. MCPT's modelling capabilites are described and preliminary experimental validation is presented. (orig.)
International Nuclear Information System (INIS)
A computer code package (PTSIM) for particle transport Monte Carlo simulation was developed using object oriented techniques of design and programming. A flexible system for simulation of coupled photon, electron transport, facilitating development of efficient simulation applications, was obtained. For photons: Compton and photo-electric effects, pair production and Rayleigh interactions are simulated, while for electrons, a class II condensed history scheme was considered, in which catastrophic interactions (Moeller electron-electron interaction, bremsstrahlung, etc.) are treated in detail and all other interactions with reduced individual effect on electron history are grouped together using continuous slowing down approximation and energy straggling theories. Electron angular straggling is simulated using Moliere theory or a mixed model in which scatters at large angles are treated as distinct events. Comparisons with experimentally benchmarks for electron transmission and bremsstrahlung emissions energy and angular spectra, and for dose calculations are presented
Running the EGS4 Monte Carlo code with Fortran 90 on a pentium computer
International Nuclear Information System (INIS)
The possibility to run the EGS4 Monte Carlo code radiation transport system for medical radiation modelling on a microcomputer is discussed. This has been done using a Fortran 77 compiler with a 32-bit memory addressing system running under a memory extender operating system. In addition a virtual memory manager such as QEMM386 was required. It has successfully run on a SUN Sparcstation2. In 1995 faster Pentium-based microcomputers became available as did the Windows 95 operating system which can handle 32-bit programs, multitasking and provides its own virtual memory management. The paper describe how with simple modification to the batch files it was possible to run EGS4 on a Pentium under Fortran 90 and Windows 95. This combination of software and hardware is cheaper and faster than running it on a SUN Sparcstation2. 8 refs., 1 tab
Zhao, L; Cluggish, B; Kim, J S; Pardo, R; Vondrasek, R
2010-02-01
A Monte Carlo charge breeding code (MCBC) is being developed by FAR-TECH, Inc. to model the capture and charge breeding of 1+ ion beam in an electron cyclotron resonance ion source (ECRIS) device. The ECRIS plasma is simulated using the generalized ECRIS model which has two choices of boundary settings, free boundary condition and Bohm condition. The charge state distribution of the extracted beam ions is calculated by solving the steady state ion continuity equations where the profiles of the captured ions are used as source terms. MCBC simulations of the charge breeding of Rb+ showed good agreement with recent charge breeding experiments at Argonne National Laboratory (ANL). MCBC correctly predicted the peak of highly charged ion state outputs under free boundary condition and similar charge state distribution width but a lower peak charge state under the Bohm condition. The comparisons between the simulation results and ANL experimental measurements are presented and discussed. PMID:20192325
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms
TRIPOLI-4®, CEA, EDF and AREVA Reference Monte Carlo Code
2014-06-01
This paper presents an overview of TRIPOLI-4®, the fourth generation of the 3D continuous-energy Monte Carlo code developed by the Service d'Etudes des Réacteurs et de Mathématiques Appliquées (SERMA) at CEA Saclay. The paper surveys the generic features: programming language, parallel operation, tracked particles, nuclear data, geometry, simulation modes, standard variance reduction techniques, sources, tracking and collision algorithms, tallies, sensitivity studies. Moreover, specific and recent features are also detailed: Doppler broadening of the elastic scattering kernel, neutron and photon material irradiation, advanced variance reduction techniques, Green's functions, cycle correlation correction, nuclear data management and depletion capabilities. The productivity tools (T4G, SALOME TRIPOLI, T4RootTools), the Verification & Validation process and the distribution and licensing policy are finally presented.
The impact of time dependant spectra on fusion blanket burn-up
International Nuclear Information System (INIS)
Highlights: ► We modelled tritium production and nuclide burn-up within a spherical, solid-breeder blanket, with a 1 GW DT fusion source. ► The effect of updating reaction rates regularly is not significant for parent nuclides. ► Updating reaction rates regularly can change the daughter nuclide inventories by several hundred percent. ► Hydrogen and helium production within steels are not significantly effected by reaction rate update. ► A time step duration of 2 weeks or less is required for tritium breeding calculations. -- Abstract: Knowledge of nuclide burn-up within tritium breeding blankets has a crucial part to play in the safety, reliability and efficiency of fusion reactors. The modelling of burn-up requires a series of neutron transport calculations which can compute the reaction rate either directly, via Monte-Carlo estimators, or by implementing the multi-group method. These reaction rates can then be directly substituted into the burn-up equations, which can calculate nuclide number densities after a specified period of burn-up. The material burn-up will change the neutron spectra and the rate of nuclear reactions. Hence, a new neutron transport calculation needs to be performed after burn-up and the sequence is repeated for several time-steps. Radiation transport calculations are computationally expensive, therefore the minimisation of reaction rate calculations via Monte-Carlo simulations is desirable. Thus, time-intervals between Monte-Carlo simulations should be as large as possible. This paper addresses the effect of neutron spectra on the burn-up of parent and daughter nuclides found in EUROFER steel and the tritium self-sufficiency time. Using a spherical reactor geometry with lithium–lead tritium breeding material, a neutron spectrum is computed at time = 0 and time = 2 years after a detailed depletion calculation using 1 day time intervals. These two spectra are then used to calculate reaction rates for every isotope listed within
New strategies of sensitivity analysis capabilities in continuous-energy Monte Carlo code RMC
International Nuclear Information System (INIS)
Highlights: • Data decomposition techniques are proposed for memory reduction. • New strategies are put forward and implemented in RMC code to improve efficiency and accuracy for sensitivity calculations. • A capability to compute region-specific sensitivity coefficients is developed in RMC code. - Abstract: The iterated fission probability (IFP) method has been demonstrated to be an accurate alternative for estimating the adjoint-weighted parameters in continuous-energy Monte Carlo forward calculations. However, the memory requirements of this method are huge especially when a large number of sensitivity coefficients are desired. Therefore, data decomposition techniques are proposed in this work. Two parallel strategies based on the neutron production rate (NPR) estimator and the fission neutron population (FNP) estimator for adjoint fluxes, as well as a more efficient algorithm which has multiple overlapping blocks (MOB) in a cycle, are investigated and implemented in the continuous-energy Reactor Monte Carlo code RMC for sensitivity analysis. Furthermore, a region-specific sensitivity analysis capability is developed in RMC. These new strategies, algorithms and capabilities are verified against analytic solutions of a multi-group infinite-medium problem and against results from other software packages including MCNP6, TSUANAMI-1D and multi-group TSUNAMI-3D. While the results generated by the NPR and FNP strategies agree within 0.1% of the analytic sensitivity coefficients, the MOB strategy surprisingly produces sensitivity coefficients exactly equal to the analytic ones. Meanwhile, the results generated by the three strategies in RMC are in agreement with those produced by other codes within a few percent. Moreover, the MOB strategy performs the most efficient sensitivity coefficient calculations (offering as much as an order of magnitude gain in FoMs over MCNP6), followed by the NPR and FNP strategies, and then MCNP6. The results also reveal that these
Improvements in the Monte Carlo code for simulating 4πβ(PC)–γ coincidence system measurements
Energy Technology Data Exchange (ETDEWEB)
Dias, M.S., E-mail: msdias@ipen.br [Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil); Takeda, M.N. [Universidade Santo Amaro, UNISA Rua Prof. Enéas da Siqueira Neto 340, 04829-300 São Paulo, SP (Brazil); Toledo, F.; Brancaccio, F.; Tongu, M.L.O.; Koskinas, M.F. [Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil)
2013-01-11
A Monte Carlo simulation code known as ESQUEMA has been developed by the Nuclear Metrology Laboratory (Laboratório de Metrologia Nuclear—LMN) in the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares—IPEN) to be used as a benchmark for radionuclide standardization. The early version of this code simulated only β−γ and ec−γ emitters with reasonably high electron and X-ray energies. To extend the code to include other radionuclides and enable the code to be applied to software coincidence counting systems, several improvements have been made and are presented in this work. -- Highlights: ► Improvements to the Monte Carlo code ESQUEMA are described. ► The experimental extrapolation curve was compared to Monte Carlo simulation. ► Eu-152 was standardized by 4π(PC)β-γ coincidence system and compared to Monte Carlo simulation. ► 4π proportional counter gamma-ray efficiency was calculated by MCNPX and compared with experiment. ► X-ray and positron decay emitters were included in the simulation.
Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual
International Nuclear Information System (INIS)
The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)
Use of the FLUKA Monte Carlo code for Hadron Therapy Application
International Nuclear Information System (INIS)
Monte Carlo (M C) codes are increasingly spreading in the hadron therapy community due to their detailed description of radiation transport and interaction with matter. M C methods are being utilized at several institutions for a wide range of activities spanning from beam characterization to quality assurance and dosimetric/radiobiological studies. The suitability of a M C code for application to hadron therapy demands accurate and reliable physical models for the description of the transport and the interaction of all components of the expected radiation field (ions, hadrons, electrons, positrons and photons). This becomes extremely important for correctly performing not only physical but also biologically based dose calculations especially in cases where ions heavier than protons are involved. In addition, accurate prediction of emerging secondary radiation is of utmost importance in emerging areas of research aiming to in vivo treatment verification. This contribution will address the specific case of the general-purpose particle and interaction code FLUKA. Validations and applications at several experimental sites as well as proton/ion therapy facilities with active beam delivery systems will be presented: Generation of synchrotron accelerator library of proton/carbon ion beam energies and foci (i.e., lateral widths at the iso centre of the treatment unit). Physical database generation: laterally integrated depth-dose profiles, lateral-dose distributions at different depths, secondary fragments yields and fragment energy spectra at different depths. Forward M C re-calculations of physical/RBE-weighted dose distributions of proton and carbon ion treatment plans. M C-based treatment planning in proton therapy. The satisfactorily agreement of FLUKA against several dosimetric/nuclear yields data indicates that the code already represents a valuable choice for supporting a large variety of applications in proton and ion beam therapy
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
International Nuclear Information System (INIS)
The basis of the NDP is the irradiation of a sample with a thermal or cold neutron beam and the subsequent release of charged particles due to neutron-induced exoergic charged particle reactions. Neutrons interact with the nuclei of elements and release mono-energetic charged particles, e.g. alpha particles or protons, and recoil atoms. Depth profile of the analyzed element can be obtained by making a linear transformation of the measured energy spectrum by using the stopping power of the sample material. A few micrometer of the material can be analyzed nondestructively, and on the order of 10nm depth resolution can be obtained depending on the material type with NDP method. In the NDP method, the one first steps of the analytical process is a channel-energy calibration. This calibration is normally made with the experimental measurement of NIST Standard Reference Material sample (SRM-93a). In this study, some Monte Carlo (MC) codes were tried to calculate the Si detector response function when this detector accounted the energy charges particles emitting from an analytical sample. In addition, these MC codes were also tried to calculate the depth distributions of some light elements (10B, 3He, 6Li, etc.) in SRM-93a and SRM-2137 samples. These calculated profiles were compared with the experimental profiles and SIMS profiles. In this study, some popular MC neutron transport codes are tried and tested to calculate the detector response function in the NDP method. The simulations were modeled based on the real CN-NDP system which is a part of Cold Neutron Activation Station (CONAS) at HANARO (KAERI). The MC simulations are very successful at predicting the alpha peaks in the measured energy spectrum. The net area difference between the measured and predicted alpha peaks are less than 1%. A possible explanation might be bad cross section data set usage in the MC codes for the transport of low energetic lithium atoms inside the silicon substrate
REBUS: A burnup credit experimental programme
International Nuclear Information System (INIS)
An international programme called REBUS (REactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK-CEN and Belgonucleaire. At present it is sponsored by USNRC, EdF from France and VGB, representing German nuclear utilities. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark would qualify the codes to perform calculations of the burn-up credit. The benchmark exercise will investigate the following fuel types with associated burn-up. 1. Reference absorber test bundle, 2. Fresh commercial PWR UO2 fuel, 3. Irradiated commercial PWR UO2 fuel (50 GWd/tM), 4. Fresh PWR UO2 fuel, 5. Irradiated PWR UO2 fuel (30 GWd/tM). Reactivity effects will be measured in the critical facility VENUS. The accumulated burn-up of all rods will be measured non-destructively by gamma-spectrometry. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-18 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). The experimental implementation of the programme will start in 2000. (author)
International Nuclear Information System (INIS)
Serpent is a continuous-energy Monte Carlo (MC) reactor physics code recently developed at VTT Technical Research Centre of Finland. Serpent can be used for 2D fuel lattice calculations as well as for 3D full core simulations. Due to its built-in decay and burnup routine Serpent can perform depletion and decay analysis to provide time-dependent isotopic compositions and spent fuel characteristics including radioactivity and decay heat. Serpent uses matrix exponential method to solve the Bateman decay and depletion equations while the solution of the matrix exponential relies on the Chebyshev Rational Approximation Method (CRAM) Serpent runs significantly faster than other MC codes due to the two main reasons: 1) the use of the Woodcock delta-tracking in a combination with a typical surface-to-surface ray-tracing in a geometry routine, and 2) the use of the unionized energy grid for all point-wise reaction cross sections. The later, however, considerably increases the memory requirements and can be a bottleneck in simulations with a large number of involved nuclides. Serpent is especially designed to generate homogenized constants for deterministic 3D core analysis. For any region of interest the code automatically calculates homogenized few-group cross sections, group-to-group scattering matrices, diffusion coefficients, assembly discontinuity factors, kinetics parameters, etc. More details can be found in Serpent User's Manual. Recently some new calculation methods related to the production of homogenized few-group constants were implemented in the Serpent code including homogenization in leakage-corrected criticality spectrum, group constant generation in reflectors and other non-fissile regions, and improved treatment of neutron-multiplying scattering reactions. The capability to generate homogenized few-group constants can be considered as one of the most attractive features of Serpent. Being a MC code, Serpent is capable of handling complex geometries without
A user`s manual for MASH 1.0: A Monte Carlo Adjoint Shielding Code System
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O. [ed.
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the ``dose importance`` of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user`s manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
A user's manual for MASH 1. 0: A Monte Carlo Adjoint Shielding Code System
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O. (ed.)
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the dose importance'' of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
International Nuclear Information System (INIS)
The ITS (Integrated Tiger Series) Monte Carlo code package developed at Sandia National Laboratories and distributed as CCC-467/ITS by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) consists of eight codes - the standard codes, TIGER, CYLTRAN, ACCEPT; the P-codes, TIGERP, CYLTRANP, ACCEPTP; and the M-codes ACCEPTM, CYLTRANM. The codes have been adapted to run on the IBM 3081, VAX 11/780, CDC-7600, and Cray 1 with the use of the update emulator UPEML. This manual should serve as a guide to a user running the codes on IBM computers having 370 architecture. The cases listed were tested on the IBM 3033, under the MVS operating system using the VS Fortran Level 1.3.1 compiler
A combined 1D/3D fuel burnup analysis of generation IV light water reactor IRIS
International Nuclear Information System (INIS)
A combined 1D/3D methodology for the fuel burnup analysis of generation IV light water reactors with thin boron coating that covers the fuel rods is described in this paper. This methodology is founded on three approximations. The first approximation assumes that the problem of fuel depletion in the entire 3D core can be resolved into two independent problems. One is a 3D Monte Carlo evolution of power distribution in large volumes (nodes) with the KENO-V.a code, and the other is a transport method evolution of burnup dependent fuel composition in 1D Wigner-Seitz cell for each node independently. With the second approximation, the time-dependent fuel composition in the node (e.g., in the fuel assembly) is calculated by using a 1D fuel depletion analysis with the SAS2H control module from the SCALE-4.4a code system. The third approximation involves smearing the boron coating with the clad (by volume homogenization). The proposed SAS2H/KENO-V.a methodology is verified for the case of 2D x-y model of IRIS 15x15 fuel assembly (with a reflective boundary condition) by using two well benchmarked code systems. The first one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. It has been found that the proposed SAS2H/KENO-V.a methodology gives a satisfactory accuracy for keff and nuclide composition. Finally, this methodology was applied for 3D burnup analysis of IRIS-1000 benchmark≠44 core. Detailed keff and power density evolution with burnup are reported. (author)
Theory analysis and simple calculation of travelling wave burnup scheme
International Nuclear Information System (INIS)
Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)
International Nuclear Information System (INIS)
One of the major problems in burnup studies is the reasonably fast and accurate calculation of the space-and-energy dependent neutron flux and reaction rates for realistic power reactor fuel geometries and compositions, and its optimal integration in the global reactor calculations. The scope of the present research was to develop improved methods trying to satisfy the above requirements. In the epithermal region, simple and efficient approximation is proposed which allows the analytical solution for the space dependence of the spherical harmonics flux moments, and hence the derivation of the recurrence relations between he flux moments at successive lethargy pivotal points. A new matrix formalism to invert the coefficient matrix of band structure resulted in a reduce computer time and memory demands. The research on epithermal region is finalized in computing programme SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, and the related integral quantities as reaction rates and resonance integrals. For partial verification of the above methods a Monte Carlo procedure was developed. Using point-wise representation of variables, a flexible and fast convergent integral transport method SEPT i developed. Expanding the neutron source and flux in finite series of arbitrary polynomials, the space-and-energy dependent integral transport equation is transformed into a general linear algebraic form, which is solved numerically. A simple and efficient procedure for deriving multipoint equations and constructing matrix is proposed and examined, and no unwanted oscillations were noticed. The energy point method was combined with the spherical harmonics method as well. A multi zone few-group program SPECTAR for global reactor calculations was developed. For testing, the flux distribution, neutron leakage and effective multiplication factor for the PWR reactor of the power station San Onofre were calculated. In order to verify
Brogan, John
Understanding the dosimetry for high-energy, heavy ions (HZE), especially within living systems, is complex and requires the use of both experimental and computational methods. Tissue-equivalent proportional counters (TEPCs) have been used experimentally to measure energy deposition in volumes similar in dimension to a mammalian cell. As these experiments begin to include a wider range of ions and energies, considerations to cost, time, and radiation protection are necessary and may limit the extent of these studies. Multiple Monte Carlo computational codes have been created to remediate this problem and serve as a mode of verification for pervious experimental methods. One such code, Relativistic-Ion Tracks (RITRACKS), is currently being developed at the NASA Johnson Space center. RITRACKS was designed to describe patterns of ionizations responsible for DNA damage on the molecular scale (nanometers). This study extends RITRACKS version 3.07 into the microdosimetric scale (microns), and compares computational results to previous experimental TEPC data. Energy deposition measurements for 1000 MeV nucleon-1 Fe ions in a 1 micron spherical target were compared. Different settings within RITRACKS were tested to verify their effects on dose to a target and the resulting energy deposition frequency distribution. The results were then compared to the TEPC data.
Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code
Energy Technology Data Exchange (ETDEWEB)
Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)
2015-07-01
Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)
International Nuclear Information System (INIS)
Recently, sensitivity and uncertainty (S/U) techniques have been used to determine the area of applicability (AOA) of critical experiments used for code and data validation. These techniques require the computation of energy-dependent sensitivity coefficients for multiple reaction types for every nuclide in each system included in the validation. The sensitivity coefficients, as used for this application, predict the relative change in the system multiplication factor due to a relative change in a given cross-section data component or material number density. Thus, a sensitivity coefficient, S, for some macroscopic cross section, Σ, is expressed as S = Σ/k ∂k/∂Σ, where k is the effective neutron multiplication factor for the system. The sensitivity coefficient for the density of a material is equivalent to that of the total macroscopic cross section. Two distinct techniques have been employed in Monte Carlo radiation transport codes for the computation of sensitivity coefficients. The first, and most commonly employed, is the differential sampling technique. The second is the adjoint-based perturbation theory approach. This paper briefly describes each technique and presents the results of a simple test case, pointing out discrepancies in the computed results and proposing a remedy to these discrepancies
PINSPEC. A Monte Carlo code for pin cell spectral calculations for educational applications
International Nuclear Information System (INIS)
Students in many reactor physics courses are exposed to canonical reactor physics concepts through theoretical problems simplified to allow for tractable analytical solutions. Such problems typically require tedious mathematical derivation which is often not the most effective approach to teaching basic reactor physics concepts. A new complementary methodology to introduce these concepts is made possible with PINSPEC, a pin cell Monte Carlo code for educational use. PINSPEC enables students to simulate pin cell models for various reactor types with a simple-to-use Python interface. PINSPEC uses point-wise cross section data and includes a module for Single-Level Breit-Wigner cross-section generation and Doppler broadening. The PINSPEC code supports a variety of tallies which students may use to compute resonance integrals, multi-group cross sections, and more for various materials and pin configurations. PINSPEC is undergoing review for open source release in the near future such that it will be a free and accessible tool for instructors developing reactor physics curricula with an applied and interactive approach to learning. (author)
Development of an unstructured mesh based geometry model in the Serpent 2 Monte Carlo code
International Nuclear Information System (INIS)
This paper presents a new unstructured mesh based geometry type, developed in the Serpent 2 Monte Carlo code as a by-product of another study related to multi-physics applications and coupling to CFD codes. The new geometry type is intended for the modeling of complicated and irregular objects, which are not easily constructed using the conventional CSG based approach. The capability is put to test by modeling the 'Stanford Critical Bunny' – a variation of a well-known 3D test case for methods used in the world of computer graphics. The results show that the geometry routine in Serpent 2 can handle the unstructured mesh, and that the use of delta-tracking results in a considerable reduction in the overall calculation time as the geometry is refined. The methodology is still very much under development, with the final goal of implementing a geometry routine capable of reading standardized geometry formats used by 3D design and imaging tools in industry and medical physics. (author)
BREESE-II: auxiliary routines for implementing the albedo option in the MORSE Monte Carlo code
International Nuclear Information System (INIS)
The routines in the BREESE package implement the albedo option in the MORSE Monte Carlo Code by providing (1) replacements for the default routines ALBIN and ALBDO in the MORSE Code, (2) an estimating routine ALBDOE compatible with the SAMBO package in MORSE, and (3) a separate program that writes a tape of albedo data in the proper format for ALBIN. These extensions of the package initially reported in 1974 were performed jointly by ORNL, Bechtel Power Corporation, and Science Applications, Inc. The first version of BREESE had a fixed number of outgoing polar angles and the number of outgoing azimuthal angles was a function of the value of the outgoing polar angle only. An examination of differential albedo data led to this modified version which allows the number of outgoing polar angles to be dependent upon the value of the incoming polar angle and the number of outgoing azimuthal angles to be a function of the value of both incoming and outgoing polar angles
Development of a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport
Jia, Xun; Sempau, Josep; Choi, Dongju; Majumdar, Amitava; Jiang, Steve B
2009-01-01
Monte Carlo simulation is the most accurate method for absorbed dose calculations in radiotherapy. Its efficiency still requires improvement for routine clinical applications, especially for online adaptive radiotherapy. In this paper, we report our recent development on a GPU-based Monte Carlo dose calculation code for coupled electron-photon transport. We have implemented the Dose Planning Method (DPM) Monte Carlo dose calculation package (Sempau et al, Phys. Med. Biol., 45(2000)2263-2291) on GPU architecture under CUDA platform. The implementation has been tested with respect to the original sequential DPM code on CPU in two cases. Our results demonstrate the adequate accuracy of the GPU implementation for both electron and photon beams in radiotherapy energy range. A speed up factor of 4.5 and 5.5 times have been observed for electron and photon testing cases, respectively, using an NVIDIA Tesla C1060 GPU card against a 2.27GHz Intel Xeon CPU processor .
Energy Technology Data Exchange (ETDEWEB)
Radaev, A. I., E-mail: radaev-aleksandr@mail.ru; Schurovskaya, M. V., E-mail: mvhchurovskaya@mephi.ru [National Research Nuclear University MEPhI (Russian Federation)
2015-12-15
The choice of the spatial nodalization for the calculation of the power density and burnup distribution in a research reactor core with fuel assemblies of the IRT-3M and VVR-KN type using the program based on the Monte Carlo code is described. The influence of the spatial nodalization on the results of calculating basic neutronic characteristics and calculation time is investigated.
A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System
Energy Technology Data Exchange (ETDEWEB)
C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler
1998-10-01
The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
Energy Technology Data Exchange (ETDEWEB)
Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)
2014-08-15
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables
A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom
International Nuclear Information System (INIS)
As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)
Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code
International Nuclear Information System (INIS)
The Monte Carlo N-Particle transport code (MCNP) was adopted to calculate the skyshine dose from the turbine building of a BWR plant for obtaining precise estimations at the site boundary. In MCNP calculation, the equipment and piping arranged on the operating floor of the turbine building were considered and modeled in detail. The inner and outer walls of the turbine building, the shielding materials around the high-pressure turbine, and the piping connected from the moisture separator to the low-pressure turbine were all considered. A three-step study was conducted to estimate the applicability of MCNP code. The first step is confirming the propriety of calculation models. The atmospheric relief diaphragms, which are installed on top of the low-pressure turbine exhaust hood, are not considered in the calculation model. There was little difference between the skyshine dose distributions that were considered when using and not using the atmospheric relief diaphragms. The calculated dose rates agreed well with the measurements taken around the turbine. The second step is estimating the dose rates on the outer roof surface of the turbine building. This calculation was made to confirm the dose distribution of gamma-rays on the turbine roof before being scattered into the air. The calculated dose rates agreed well with the measured data. The third step is making a final confirmation by comparing the calculations and measurements of skyshine dose rates around the turbine building. The source terms of the main steam system are based on the measured activity data of N-16 and C-15. As a conclusion, we were able to calculate reasonable skyshine dose rates by using MCNP code. (authors)
Verification of SMART Neutronics Design Methodology by the MCNAP Monte Carlo Code
International Nuclear Information System (INIS)
SMART is a small advanced integral pressurized water reactor (PWR) of 330 MW(thermal) designed for both electricity generation and seawater desalinization. The CASMO-3/MASTER nuclear analysis system, a design-basis of Korean PWR plants, has been employed for the SMART core nuclear design and analysis because the fuel assembly (FA) characteristics and reactor operating conditions in temperature and pressure are similar to those of PWR plants. However, the SMART FAs are highly poisoned with more than 20 Al2O3-B4C plus additional Gd2O3/UO2 BPRs each FA. The reactor is operated with control rods inserted. Therefore, the flux and power distribution may become more distorted than those of commercial PWR plants. In addition, SMART should produce power from room temperature to hot-power operating condition because it employs nuclear heating from room temperature. This demands reliable predictions of core criticality, shutdown margin, control rod worth, power distributions, and reactivity coefficients at both room temperature and hot operating condition, yet no such data are available to verify the CASMO-3/MASTER (hereafter MASTER) code system. In the absence of experimental verification data for the SMART neutronics design, the Monte Carlo depletion analysis program MCNAP is adopted as near-term alternatives for qualifying MASTER neutronics design calculations. The MCNAP is a personal computer-based continuous energy Monte Carlo neutronics analysis program written in C++ language. We established its qualification by presenting its prediction accuracy on measurements of Venus critical facilities and core neutronics analysis of a PWR plant in operation, and depletion characteristics of integral burnable absorber FAs of the current PWR. Here, we present a comparison of MASTER and MCNAP neutronics design calculations for SMART and establish the qualification of the MASTER system
Performance of code 'FAIR' in IAEA CRP on FUMEX
International Nuclear Information System (INIS)
A modern fuel performance analysis code FAIR has been developed for analysing high burnup fuel pins of water/heavy water cooled reactors. The code employs finite element method for modelling thermo mechanical behaviour of fuel pins and mechanistic models for modelling various physical and chemical phenomena affecting the behaviour of nuclear reactor fuel pins. High burnup affects such as pellet thermal conductivity degradation, enhanced fission gas release and radial flux redistribution are incorporated in the code FAIR. The code FAIR is capable of performing statistical analysis of fuel pins using Monte Carlo technique. The code is implemented on BARC parallel processing system ANUPAM. The code has recently participated in an International Atomic Energy Agency (IAEA) coordinated research program (CRP) on fuel modelling at extended burnups (FUMEX). Nineteen agencies from different countries participated in this exercise. In this CRP, spread over a period of three years, a number of high burnup fuel pins irradiated at Halden reactor are analysed. The first phase of the CRP is a blind code comparison exercise, where the computed results are compared with experimental results. The second phase consists of modifications to the code based on the experimental results of first phase and statistical analysis of fuel pins. The performance of the code FAIR in this CRP has been very good. The present report highlights the main features of code FAIR and its performance in the IAEA CRP on FUMEX. 14 refs., 5 tabs., ills
Full modelling of the MOSAIC animal PET system based on the GATE Monte Carlo simulation code
International Nuclear Information System (INIS)
Positron emission tomography (PET) systems dedicated to animal imaging are now widely used for biological studies. The scanner performance strongly depends on the design and the characteristics of the system. Many parameters must be optimized like the dimensions and type of crystals, geometry and field-of-view (FOV), sampling, electronics, lightguide, shielding, etc. Monte Carlo modelling is a powerful tool to study the effect of each of these parameters on the basis of realistic simulated data. Performance assessment in terms of spatial resolution, count rates, scatter fraction and sensitivity is an important prerequisite before the model can be used instead of real data for a reliable description of the system response function or for optimization of reconstruction algorithms. The aim of this study is to model the performance of the Philips Mosaic(TM) animal PET system using a comprehensive PET simulation code in order to understand and describe the origin of important factors that influence image quality. We use GATE, a Monte Carlo simulation toolkit for a realistic description of the ring PET model, the detectors, shielding, cap, electronic processing and dead times. We incorporate new features to adjust signal processing to the Anger logic underlying the Mosaic(TM) system. Special attention was paid to dead time and energy spectra descriptions. Sorting of simulated events in a list mode format similar to the system outputs was developed to compare experimental and simulated sensitivity and scatter fractions for different energy thresholds using various models of phantoms describing rat and mouse geometries. Count rates were compared for both cylindrical homogeneous phantoms. Simulated spatial resolution was fitted to experimental data for 18F point sources at different locations within the FOV with an analytical blurring function for electronic processing effects. Simulated and measured sensitivities differed by less than 3%, while scatter fractions agreed
International Nuclear Information System (INIS)
Highlights: ► New coupled Monte Carlo code system for reference results at operating conditions. ► Automated methodology to create and use temperature-dependent cross section libraries. ► Multi-level coupling scheme between MCNP5 and COBRA-TF with different options. ► Acceleration strategy for coupled Monte Carlo calculations including hybrid approach. ► Sensitivity studies on thermal-scattering models and different sub-channel approaches. -- Abstract: High accuracy code systems are necessary to model core environments with considerable geometry complexity and great material heterogeneity. These features are typical of current and innovative nuclear reactor core designs. Advanced methodologies and state-of-the art coupled code systems must be put into practice in order to model with high accuracy these challenging core designs. The presented research comprises the development and implementation of the thermal–hydraulic feedback to the Monte Carlo method and of speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal–hydraulic models. The development and verification of such reference high-fidelity coupled multi-physics scheme is performed at the Pennsylvania State University (PSU) in cooperation with AREVA, AREVA NP GmbH in Erlangen, Germany, on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This paper presents the latest studies and ameliorations developed to this coupled hybrid system, which includes a new methodology for generation and interpolation of Temperature-Dependent Thermal Scattering Cross Section Libraries for MCNP5, a comparison between sub-channel approaches, and acceleration schemes.
Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia
Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto
2015-01-01
A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the n...
Assessment of MIRD data for internal dosimetry using the GATE Monte Carlo code.
Parach, Ali Asghar; Rajabi, Hossein; Askari, Mohammad Ali
2011-08-01
GATE/GEANT is a Monte Carlo code dedicated to nuclear medicine that allows calculation of the dose to organs of voxel phantoms. On the other hand, MIRD is a well-developed system for estimation of the dose to human organs. In this study, results obtained from GATE/GEANT using Snyder phantom are compared to published MIRD data. For this, the mathematical Snyder phantom was discretized and converted to a digital phantom of 100 × 200 × 360 voxels. The activity was considered uniformly distributed within kidneys, liver, lungs, pancreas, spleen, and adrenals. The GATE/GEANT Monte Carlo code was used to calculate the dose to the organs of the phantom from mono-energetic photons of 10, 15, 20, 30, 50, 100, 200, 500, and 1000 keV. The dose was converted into specific absorbed fraction (SAF) and the results were compared to the corresponding published MIRD data. On average, there was a good correlation (r (2)>0.99) between the two series of data. However, the GATE/GEANT data were on average -0.16 ± 6.22% lower than the corresponding MIRD data for self-absorption. Self-absorption in the lungs was considerably higher in the MIRD compared to the GATE/GEANT data, for photon energies of 10-20 keV. As for cross-irradiation to other organs, the GATE/GEANT data were on average +1.5 ± 8.1% higher than the MIRD data, for photon energies of 50-1000 keV. For photon energies of 10-30 keV, the relative difference was +7.5 ± 67%. It turned out that the agreement between the GATE/GEANT and the MIRD data depended upon absolute SAF values and photon energy. For 10-30 keV photons, where the absolute SAF values were small, the uncertainty was high and the effect of cross-section prominent, and there was no agreement between the GATE/GEANT results and the MIRD data. However, for photons of 50-1,000 keV, the bias was negligible and the agreement was acceptable. PMID:21573984
COG10, Multiparticle Monte Carlo Code System for Shielding and Criticality Use
International Nuclear Information System (INIS)
1 - Description of program or function: COG is a modern, full-featured Monte Carlo radiation transport code which provides accurate answers to complex shielding, criticality, and activation problems. COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computing Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://www-phys.llnl.gov/N_Div/COG/. Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, path-length stretching, point detectors, scattered direction biasing, and forced collisions. Criticality - For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system. Activation - COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source. Coupled Problems - COG can solve coupled problems involving neutrons, photons, and electrons. 2 - Methods:COG uses Monte Carlo methods to solve the Boltzmann transport equation for particles traveling through arbitrary 3-dimensional geometries. Neutrons, photons
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After a description of the context of radiological accidents (definition, history, context, exposure types, associated clinic symptoms of irradiation and contamination, medical treatment, return on experience) and a presentation of dose assessment in the case of external exposure (clinic, biological and physical dosimetry), this research thesis describes the principles of numerical reconstruction of a radiological accident, presents some computation codes (Monte Carlo code, MCNPX code) and the SESAME tool, and reports an application to an actual case (an accident which occurred in Equator in April 2009). The next part reports the developments performed to modify the posture of voxelized phantoms and the experimental and numerical validations. The last part reports a feasibility study for the reconstruction of radiological accidents occurring in external radiotherapy. This work is based on a Monte Carlo simulation of a linear accelerator, with the aim of identifying the most relevant parameters to be implemented in SESAME in the case of external radiotherapy
International Nuclear Information System (INIS)
Highlights: • We present a new Monte Carlo method to perform sensitivity/perturbation calculations. • Sensitivity of keff, reaction rates, point kinetics parameters to nuclear data. • Fully continuous implicitly constrained Monte Carlo sensitivities to scattering distributions. • Implementation of the method in the continuous energy Monte Carlo code SERPENT. • Verification against ERANOS and TSUNAMI generalized perturbation theory results. - Abstract: In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the effects of nuclear data perturbation on several response functions: the effective multiplication factor, reaction rate ratios and bilinear ratios (e.g., effective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
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Tuan, Hoang Sy Minh; Sun, Gwang Min; Park, Byung Gun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The basis of the NDP is the irradiation of a sample with a thermal or cold neutron beam and the subsequent release of charged particles due to neutron-induced exoergic charged particle reactions. Neutrons interact with the nuclei of elements and release mono-energetic charged particles, e.g. alpha particles or protons, and recoil atoms. Depth profile of the analyzed element can be obtained by making a linear transformation of the measured energy spectrum by using the stopping power of the sample material. A few micrometer of the material can be analyzed nondestructively, and on the order of 10nm depth resolution can be obtained depending on the material type with NDP method. In the NDP method, the one first steps of the analytical process is a channel-energy calibration. This calibration is normally made with the experimental measurement of NIST Standard Reference Material sample (SRM-93a). In this study, some Monte Carlo (MC) codes were tried to calculate the Si detector response function when this detector accounted the energy charges particles emitting from an analytical sample. In addition, these MC codes were also tried to calculate the depth distributions of some light elements ({sup 10}B, {sup 3}He, {sup 6}Li, etc.) in SRM-93a and SRM-2137 samples. These calculated profiles were compared with the experimental profiles and SIMS profiles. In this study, some popular MC neutron transport codes are tried and tested to calculate the detector response function in the NDP method. The simulations were modeled based on the real CN-NDP system which is a part of Cold Neutron Activation Station (CONAS) at HANARO (KAERI). The MC simulations are very successful at predicting the alpha peaks in the measured energy spectrum. The net area difference between the measured and predicted alpha peaks are less than 1%. A possible explanation might be bad cross section data set usage in the MC codes for the transport of low energetic lithium atoms inside the silicon substrate.
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Highlights: • A successful validation of the burn-up simulation system EVOLCODE is presented here. • A Sensitivity/Uncertainty model was applied for uncertainty propagation/assessment. • Cross sections are for most cases the main contributors to inventory uncertainties. • The improved model helps to explain some simulation-experiment discrepancies. • Some hints for the improvement of basic data libraries are provided. - Abstract: A validation of the burn-up simulation system EVOLCODE 2.0 is presented here, involving the experimental measurement of U and Pu isotopes and some fission fragments production ratios after a burn-up of around 30 GWd/tU in a Pressurized Light Water Reactor (PWR). This work provides an in-depth analysis of the validation results, including the possible sources of the uncertainties. An uncertainty analysis based on the sensitivity methodology has been also performed, providing the uncertainties in the isotopic content propagated from the cross sections uncertainties. An improvement of the classical Sensitivity/Uncertainty (S/U) model has been developed to take into account the implicit dependence of the neutron flux normalization, that is, the effect of the constant power of the reactor. The improved S/U methodology, neglected in this kind of studies, has proven to be an important contribution to the explanation of some simulation-experiment discrepancies for which, in general, the cross section uncertainties are, for the most relevant actinides, an important contributor to the simulation uncertainties, of the same order of magnitude and sometimes even larger than the experimental uncertainties and the experiment-simulation differences. Additionally, some hints for the improvement of the JEFF3.1.1 fission yield library and for the correction of some errata in the experimental data are presented
Icarus: A 2-D Direct Simulation Monte Carlo (DSMC) Code for Multi-Processor Computers
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Icarus is a 2D Direct Simulation Monte Carlo (DSMC) code which has been optimized for the parallel computing environment. The code is based on the DSMC method of Bird[11.1] and models from free-molecular to continuum flowfields in either cartesian (x, y) or axisymmetric (z, r) coordinates. Computational particles, representing a given number of molecules or atoms, are tracked as they have collisions with other particles or surfaces. Multiple species, internal energy modes (rotation and vibration), chemistry, and ion transport are modeled. A new trace species methodology for collisions and chemistry is used to obtain statistics for small species concentrations. Gas phase chemistry is modeled using steric factors derived from Arrhenius reaction rates or in a manner similar to continuum modeling. Surface chemistry is modeled with surface reaction probabilities; an optional site density, energy dependent, coverage model is included. Electrons are modeled by either a local charge neutrality assumption or as discrete simulational particles. Ion chemistry is modeled with electron impact chemistry rates and charge exchange reactions. Coulomb collision cross-sections are used instead of Variable Hard Sphere values for ion-ion interactions. The electro-static fields can either be: externally input, a Langmuir-Tonks model or from a Green's Function (Boundary Element) based Poison Solver. Icarus has been used for subsonic to hypersonic, chemically reacting, and plasma flows. The Icarus software package includes the grid generation, parallel processor decomposition, post-processing, and restart software. The commercial graphics package, Tecplot, is used for graphics display. All of the software packages are written in standard Fortran
Tyagi, Neelam; Bose, Abhijit; Chetty, Indrin J
2004-09-01
We have parallelized the Dose Planning Method (DPM), a Monte Carlo code optimized for radiotherapy class problems, on distributed-memory processor architectures using the Message Passing Interface (MPI). Parallelization has been investigated on a variety of parallel computing architectures at the University of Michigan-Center for Advanced Computing, with respect to efficiency and speedup as a function of the number of processors. We have integrated the parallel pseudo random number generator from the Scalable Parallel Pseudo-Random Number Generator (SPRNG) library to run with the parallel DPM. The Intel cluster consisting of 800 MHz Intel Pentium III processor shows an almost linear speedup up to 32 processors for simulating 1 x 10(8) or more particles. The speedup results are nearly linear on an Athlon cluster (up to 24 processors based on availability) which consists of 1.8 GHz+ Advanced Micro Devices (AMD) Athlon processors on increasing the problem size up to 8 x 10(8) histories. For a smaller number of histories (1 x 10(8)) the reduction of efficiency with the Athlon cluster (down to 83.9% with 24 processors) occurs because the processing time required to simulate 1 x 10(8) histories is less than the time associated with interprocessor communication. A similar trend was seen with the Opteron Cluster (consisting of 1400 MHz, 64-bit AMD Opteron processors) on increasing the problem size. Because of the 64-bit architecture Opteron processors are capable of storing and processing instructions at a faster rate and hence are faster as compared to the 32-bit Athlon processors. We have validated our implementation with an in-phantom dose calculation study using a parallel pencil monoenergetic electron beam of 20 MeV energy. The phantom consists of layers of water, lung, bone, aluminum, and titanium. The agreement in the central axis depth dose curves and profiles at different depths shows that the serial and parallel codes are equivalent in accuracy. PMID:15487756
International Nuclear Information System (INIS)
We have parallelized the Dose Planning Method (DPM), a Monte Carlo code optimized for radiotherapy class problems, on distributed-memory processor architectures using the Message Passing Interface (MPI). Parallelization has been investigated on a variety of parallel computing architectures at the University of Michigan-Center for Advanced Computing, with respect to efficiency and speedup as a function of the number of processors. We have integrated the parallel pseudo random number generator from the Scalable Parallel Pseudo-Random Number Generator (SPRNG) library to run with the parallel DPM. The Intel cluster consisting of 800 MHz Intel Pentium III processor shows an almost linear speedup up to 32 processors for simulating 1x108 or more particles. The speedup results are nearly linear on an Athlon cluster (up to 24 processors based on availability) which consists of 1.8 GHz+ Advanced Micro Devices (AMD) Athlon processors on increasing the problem size up to 8x108 histories. For a smaller number of histories (1x108) the reduction of efficiency with the Athlon cluster (down to 83.9% with 24 processors) occurs because the processing time required to simulate 1x108 histories is less than the time associated with interprocessor communication. A similar trend was seen with the Opteron Cluster (consisting of 1400 MHz, 64-bit AMD Opteron processors) on increasing the problem size. Because of the 64-bit architecture Opteron processors are capable of storing and processing instructions at a faster rate and hence are faster as compared to the 32-bit Athlon processors. We have validated our implementation with an in-phantom dose calculation study using a parallel pencil monoenergetic electron beam of 20 MeV energy. The phantom consists of layers of water, lung, bone, aluminum, and titanium. The agreement in the central axis depth dose curves and profiles at different depths shows that the serial and parallel codes are equivalent in accuracy
Energy Technology Data Exchange (ETDEWEB)
Blazy-Aubignac, L
2007-09-15
The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
Modeling Monte Carlo of multileaf collimators using the code GEANT4
Energy Technology Data Exchange (ETDEWEB)
Oliveira, Alex C.H.; Lima, Fernando R.A., E-mail: oliveira.ach@yahoo.com, E-mail: falima@cnen.gov.br [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Lima, Luciano S.; Vieira, Jose W., E-mail: lusoulima@yahoo.com.br [Instituto Federal de Educacao, Ciencia e Tecnologia de Pernambuco (IFPE), Recife, PE (Brazil)
2014-07-01
Radiotherapy uses various techniques and equipment for local treatment of cancer. The equipment most often used in radiotherapy to the patient irradiation is linear accelerator (Linac). Among the many algorithms developed for evaluation of dose distributions in radiotherapy planning, the algorithms based on Monte Carlo (MC) methods have proven to be very promising in terms of accuracy by providing more realistic results. The MC simulations for applications in radiotherapy are divided into two parts. In the first, the simulation of the production of the radiation beam by the Linac is performed and then the phase space is generated. The phase space contains information such as energy, position, direction, etc. of millions of particles (photons, electrons, positrons). In the second part the simulation of the transport of particles (sampled phase space) in certain configurations of irradiation field is performed to assess the dose distribution in the patient (or phantom). Accurate modeling of the Linac head is of particular interest in the calculation of dose distributions for intensity modulated radiation therapy (IMRT), where complex intensity distributions are delivered using a multileaf collimator (MLC). The objective of this work is to describe a methodology for modeling MC of MLCs using code Geant4. To exemplify this methodology, the Varian Millennium 120-leaf MLC was modeled, whose physical description is available in BEAMnrc Users Manual (20 11). The dosimetric characteristics (i.e., penumbra, leakage, and tongue-and-groove effect) of this MLC were evaluated. The results agreed with data published in the literature concerning the same MLC. (author)
HERMES: a Monte Carlo Code for the Propagation of Ultra-High Energy Nuclei
De Domenico, Manlio; Settimo, Mariangela
2013-01-01
Although the recent experimental efforts to improve the observation of Ultra-High Energy Cosmic Rays (UHECRs) above $10^{18}$ eV, the origin and the composition of such particles is still unknown. In this work, we present the novel Monte Carlo code (HERMES) simulating the propagation of UHE nuclei, in the energy range between $10^{16}$ and $10^{22}$ eV, accounting for propagation in the intervening extragalactic and Galactic magnetic fields and nuclear interactions with relic photons of the extragalactic background radiation. In order to show the potential applications of HERMES for astroparticle studies, we estimate the expected flux of UHE nuclei in different astrophysical scenarios, the GZK horizons and we show the expected arrival direction distributions in the presence of turbulent extragalactic magnetic fields. A stable version of HERMES will be released in the next future for public use together with libraries of already propagated nuclei to allow the community to perform mass composition and energy sp...
Analysis of the KANT experiment on beryllium using TRIPOLI-4 Monte Carlo code
International Nuclear Information System (INIS)
Beryllium is an important material in fusion technology for multiplying neutrons in blankets. However, beryllium nuclear data are differently presented in modern nuclear data evaluations. Recent investigations with the TRIPOLI-4 Monte Carlo simulation of the tritium breeding ratio (TBR) demonstrated that beryllium reaction data are the main source of the calculation uncertainties between ENDF/B-VII.0 and JEFF-3.1. To clarify the calculation uncertainties from data libraries on beryllium, in this study TRIPOLI-4 calculations of the Karlsruhe Neutron Transmission (KANT) experiment have been performed by using ENDF/B-VII.0 and new JEFF-3.1.1 data libraries. The KANT Experiment on beryllium has been used to validate neutron transport codes and nuclear data libraries. An elaborated KANT experiment benchmark has been compiled and published in the NEA/SINBAD database and it has been used as reference in the present work. The neutron multiplication in bulk beryllium assemblies was considered with a central D-T neutron source. Neutron leakage spectra through the 5, 10, and 17 cm thick spherical beryllium shells were calculated and five-group partial leakage multiplications were reported and discussed. In general, improved C/E ratios on neutron leakage multiplications have been obtained. Both ENDF/B-VII.0 and JEFF-3.1.1 beryllium data libraries of TRIPOLI-4 are acceptable now for fusion neutronics calculations.
Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code
Nunomiya, T; Nakamura, T; Nakao, N
2002-01-01
A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were trans...
International Nuclear Information System (INIS)
The most dental imaging is performed by means a imaging system consisting of a film/screen combination. Fluorescent intensifying screens for X-ray films are used in order to reduce the radiation dose. They produce visible light which increases the efficiency of the film. In addition, the primary radiation can be scattered elastically (Rayleigh scattering) and inelastically (Compton scattering) which will degrade the image resolution. Scattered radiation produced in Gd2O2S:Tb intensifying screens was simulated by using a Monte Carlo radiation transport code - the EGS4. The magnitude of scattered radiation striking the film is typically quantified using the scatter to primary radiation and the scatter fraction. The angular distribution of the intensity of the scattered radiation (sum of both the scattering effects) was simulated, showing that the ratio of secondary-to-primary radiation incident on the X-ray film is about 5.67% and 3.28 % and the scatter function is about 5.27% and 3.18% for the front and back screen, respectively, over the range from 0 to π rad. (author)
Modelling of a general purpose irradiation chamber using a Monte Carlo particle transport code
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Full-text: The aim of this research is to stimulate the effectiveness use of a general purpose irradiation chamber to contain pure neutron particles obtained from a research reactor. The secondary neutron and gamma particles dose discharge from the chamber layers will be used as a platform to estimate the safe dimension of the chamber. The chamber, made up of layers of lead (Pb), shielding, polyethylene (PE), moderator and commercial grade aluminium (Al) cladding is proposed for the use of interacting samples with pure neutron particles in a nuclear reactor environment. The estimation was accomplished through simulation based on general Monte Carlo N-Particle transport code using Los Alamos MCNPX software. Simulations were performed on the model of the chamber subjected to high neutron flux radiation and its gamma radiation product. The model of neutron particle used is based on the neutron source found in PUSPATI TRIGA MARK II research reactor which holds a maximum flux value of 1 x 1012 neutron/ cm2s. The expected outcomes of this research are zero gamma dose in the core of the chamber and neutron dose rate of less than 10 μSv/ day discharge from the chamber system. (author)
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The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor; 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions. (author)
Radiation field characterization of a BNCT research facility using Monte Carlo method - code MCNP-4B
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Boron Neutron Capture Therapy - BNCT - is a selective cancer treatment and arises as an alternative therapy to treat cancer when usual techniques - surgery, chemotherapy or radiotherapy - show no satisfactory results. The main proposal of this work is to project a facility to BNCT studies. This facility relies on the use of an Am Be neutron source and on a set of moderators, filters and shielding which will provide the best neutron/gamma beam characteristic for these Becton studies, i.e., high intensity thermal and/or epithermal neutron fluxes and with the minimum feasible gamma rays and fast neutrons contaminants. A computational model of the experiment was used to obtain the radiation field in the sample irradiation position. The calculations have been performed with the MCNP 4B Monte Carlo Code and the results obtained can be regarded as satisfactory, i.e., a thermal neutron fluencyNT = 1,35x108 n/cm , a fast neutron dose of 5,86x10-10 Gy/NT and a gamma ray dose of 8,30x10-14 Gy/NT. (author)
Radiation field characterization of a BNCT research facility using Monte Carlo Method - Code MCNP-4B
International Nuclear Information System (INIS)
Boron Neutron Capture Therapy - BNCT- is a selective cancer treatment and arises as an alternative therapy to treat cancer when usual techniques - surgery, chemotherapy or radiotherapy - show no satisfactory results. The main proposal of this work is to project a facility to BNCT studies. This facility relies on the use of an AmBe neutron source and on a set of moderators, filters and shielding which will provide the best neutron/gamma beam characteristic for these BNCT studies, i.e., high intensity thermal and/or epithermal neutron fluxes and with the minimum feasible gamma rays and fast neutrons contaminants. A computational model of the experiment was used to obtain the radiation field in the sample irradiation position. The calculations have been performed with the MCNP 4B Monte Carlo Code and the results obtained can be regarded as satisfactory, i.e., a thermal neutron fluency ΝΤ = 1,35x108 n/cm2, a fast neutron dose of 5,86x-10 Gy/ΝΤ and a gamma ray dose of 8,30x-14 Gy/ΝΤ. (author)
MOCRA: a Monte Carlo code for the simulation of radiative transfer in the atmosphere.
Premuda, Margherita; Palazzi, Elisa; Ravegnani, Fabrizio; Bortoli, Daniele; Masieri, Samuele; Giovanelli, Giorgio
2012-03-26
This paper describes the radiative transfer model (RTM) MOCRA (MOnte Carlo Radiance Analysis), developed in the frame of DOAS (Differential Optical Absorption Spectroscopy) to correctly interpret remote sensing measurements of trace gas amounts in the atmosphere through the calculation of the Air Mass Factor. Besides the DOAS-related quantities, the MOCRA code yields: 1- the atmospheric transmittance in the vertical and sun directions, 2- the direct and global irradiance, 3- the single- and multiple- scattered radiance for a detector with assigned position, line of sight and field of view. Sample calculations of the main radiometric quantities calculated with MOCRA are presented and compared with the output of another RTM (MODTRAN4). A further comparison is presented between the NO2 slant column densities (SCDs) measured with DOAS at Evora (Portugal) and the ones simulated with MOCRA. Both comparisons (MOCRA-MODTRAN4 and MOCRA-observations) gave more than satisfactory results, and overall make MOCRA a versatile tool for atmospheric radiative transfer simulations and interpretation of remote sensing measurements. PMID:22453470
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The System of Computerized Analysis for Licensing at Atomic industry (SCALA) is a Russian analogue of the well-known SCALE system. For criticality evaluations the ABBN-93 system is used with TWODANT and with joined American KENO and Russian MMK Monte-Carlo code MMKKENO. Using the same cross sections and input models, all these codes give results that coincide within the statistical uncertainties (for Monte-Carlo codes). Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Another task of the work was to test the burnup capability of SCALA system in complex geometry in compare with other codes. Benchmark models of VVER type reactor assemblies with UO2 and MOX fuel including the cases with burnable gadolinium absorbers were calculated. KENO-VI and MMK codes were used for power distribution calculations, ORIGEN code was used for the isotopic kinetics calculations. (authors)
COGEMA/TRANSNUCLEAIRE's experience with burnup credit
International Nuclear Information System (INIS)
Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)
International Nuclear Information System (INIS)
The crucial problem for radiation shielding design at heavy ion accelerator facilities with beam energies of several GeV/n is the source term problem. Experimental data on double differential neutron yields from thick targets irradiated with high-energy uranium nuclei are lacking. At present there are not many Monte Carlo multipurpose codes that can work with primary high-energy uranium nuclei. These codes use different physical models for simulating nucleus-nucleus reactions. Therefore, verification of the codes with available experimental data is very important for selection of the most reliable code for practical tasks. This paper presents comparisons of the FLUKA, GEANT4 and SHIELD code simulations with experimental data on neutron production at 1 GeV/n 238U beam interaction with a thick Fe target
International Nuclear Information System (INIS)
The gas tag burnup changes investigated were limited to the three tags (Kr-78/Kr-80, Xe-126/Xe-129 and Kr-82/Kr-80) currently accepted as being the most desirable. Control rod tag burnup was significantly greater than fuel rod tag burnup. This occurs because control rods stay in the reactor longer and occupy positions of greater low-energy flux. Thus, minimum tag spacings were set by the control rods as 1.079 for Kr-78/Kr-80, 1.189 for Xe-126/Xe-129 and 1.134 for Kr-82/Kr-80
International Nuclear Information System (INIS)
An advanced model GRSWEL-A for fission gas behavior and micro-structural evolutions in Light Water Reactor (LWR) fuels was developed for and embedded in the START-3 fuel performance code. This paper represents the physical basis and verification of the model with emphasis on analysis of High Burn-up Structure (HBS), which is generally ascribed to a so-called rim-layer of high burn-up fuel pellets. Specifically, the issues of microscopic polygonization, loss of matrix fission gas, growth of fuel porosity and fission gas release are highlighted. The effects of HBS on total fission gas release, temperature distribution in the pellet, pellet swelling and permanent strain of the cladding are considered in the appropriate section of the paper by means of comparative and sensitivity analysis with the use of both modeling and available experimental data. In all the cases, an accounting for the present effects is found to be an important integral part of thorough analysis of LWR fuel behavior. Aside from the description of current capabilities of modeling, some priority directions of further improvement are outlined. (author)
Energy Technology Data Exchange (ETDEWEB)
Carrazana Gonzalez, J.; Cornejo Diaz, N. [Centre for Radiological Protection and Hygiene, P.O. Box 6195, Habana (Cuba); Jurado Vargas, M., E-mail: mjv@unex.es [Departamento de Fisica, Universidad de Extremadura, 06071 Badajoz (Spain)
2012-05-15
We studied the applicability of the Monte Carlo code DETEFF for the efficiency calibration of detectors for in situ gamma-ray spectrometry determinations of ground deposition activity levels. For this purpose, the code DETEFF was applied to a study case, and the calculated {sup 137}Cs activity deposition levels at four sites were compared with published values obtained both by soil sampling and by in situ measurements. The {sup 137}Cs ground deposition levels obtained with DETEFF were found to be equivalent to the results of the study case within the uncertainties involved. The code DETEFF could thus be used for the efficiency calibration of in situ gamma-ray spectrometry for the determination of ground deposition activity using the uniform slab model. It has the advantage of requiring far less simulation time than general Monte Carlo codes adapted for efficiency computation, which is essential for in situ gamma-ray spectrometry where the measurement configuration yields low detection efficiency. - Highlights: Black-Right-Pointing-Pointer Application of the code DETEFF to in situ gamma-ray spectrometry. Black-Right-Pointing-Pointer {sup 137}Cs ground deposition levels evaluated assuming a uniform slab model. Black-Right-Pointing-Pointer Code DETEFF allows a rapid efficiency calibration.
International Nuclear Information System (INIS)
A new specific purpose Monte Carlo code called McENL for modeling the time response of epithermal neutron lifetime tools is described. The code was developed so that the Monte Carlo neophyte can easily use it. A minimum amount of input preparation is required and specified fixed values of the parameters used to control the code operation can be used. The weight windows technique, employing splitting and Russian Roulette, is used with an automated importance function based on the solution of an adjoint diffusion model to improve the code efficiency. Complete composition and density correlated sampling is also included in the code and can be used to study the effect on tool response of small variations in the formation, borehole, or logging tool composition and density. An illustration of the latter application is given here for the density of a thermal neutron filter. McENL was benchmarked against test-pit data for the Mobil pulsed neutron porosity (PNP) tool and found to be very accurate. Results of the experimental validation and details of code performance are presented
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The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
International Nuclear Information System (INIS)
High accuracy code systems are necessary to model core environments with considerable geometry complexity and great material heterogeneity. These features are typical of current and innovative nuclear reactor core designs. Advanced methodologies and state-of-the art coupled code systems must be put into practice in order to model with high accuracy these challenging core designs. The presented research comprises the development and implementation of the thermal-hydraulic feedback to the Monte Carlo method and of speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal-hydraulic models. The development and verification of such reference high-fidelity coupled multi-physics scheme is performed at the Pennsylvania State University (PSU) in cooperation with AREVA, AREVA NP GmbH in Erlangen, Germany, on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This paper presents the latest studies and ameliorations developed to this coupled hybrid system, which includes a new methodology for generation and interpolation of Temperature-Dependent Thermal Scattering Cross Section Libraries for MCNP5, a comparison between sub-channel approaches, and acceleration schemes. (authors)
Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit
International Nuclear Information System (INIS)
The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States (U.S.) Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized water reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% Δk. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs, they
Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit
Energy Technology Data Exchange (ETDEWEB)
Wagner, J.C.
2001-09-28
The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs
Automated generation of burnup chain for reactor analysis applications
International Nuclear Information System (INIS)
This paper presents the development of an automated generation of a new burnup chain for reactor analysis applications. The JENDL FP Decay Data File 2011 and Fission Yields Data File 2011 were used as the data sources. The nuclides in the new chain are determined by restrictions of the half-life and cumulative yield of fission products or from a given list. Then, decay modes, branching ratios and fission yields are recalculated taking into account intermediate reactions. The new burnup chain is output according to the format for the SRAC code system. Verification was performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Further development and applications are being planned with the burnup chain code. (author)
Investigation of burnup credit implementation for BWR fuel
International Nuclear Information System (INIS)
Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)
Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes
International Nuclear Information System (INIS)
The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior
Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes
Energy Technology Data Exchange (ETDEWEB)
Mainardi, Enrico; Donahue, Richard J.; Blakely, Eleanor A.
2002-09-11
The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior.
International Nuclear Information System (INIS)
The analysis of void reactivity effect is prominent interest for Sodium-cooled Fast Reactor (SFR) safety. Indeed, in case of sodium leakage of the primary circuit, void reactivity represents the main passive negative feedback to ensure reactivity control. The core can be designed to maximize neutron leakage and lower the average neutron multiplication factor in the event of sodium disappearing from within assemblies. Thus, the nuclear chain reaction is stopped. The most promising solution is to place a sodium region above the fuel in order for neutrons to be reflected when the region is filled and escape when the region is empty. In terms of simulation, this configuration is a challenge for usual calculation schemes: 1. Deterministic codes are typically limited in their ability to homogenize a sub-critical medium as the sodium plenum. 2. Monte Carlo codes are typically not able to split the total reactivity effect on different components, which prevents to achieve straightforward uncertainty analysis. Furthermore, since experimental values can sometimes be small, Monte Carlo codes may not converge within a reasonable computation time. A new feature recently available in the Monte Carlo TRIPOLI-4® based on the Exact Perturbation Theory allows very small reactivity perturbations to be computed accurately as well as reactivity effect to be estimated on distinct isotopes cross-sections. In the first part of this paper, this new feature of the code is described and then applied in the second part to a core configuration composed of several layers of fuel and fertile zones below a sodium plenum. Reactivity and its contributions from specific reactions and energy groups are calculated and compared with the results of the deterministic code ERANOS. The aim of this work is twofold: (1) Achieve a numerical validation of the new TRIPOLI-4® features and (2) Identify where deterministic codes might be less accurate and why – even when using them at full capacity (S16
High Burnup Fuel Performance and Safety Research
Energy Technology Data Exchange (ETDEWEB)
Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)
2007-03-15
The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.
DeCART code verifications by numerical benchmark calculations of HTTR
International Nuclear Information System (INIS)
DeCART code verifications have been performed through the numerical benchmark calculations of HTTR. The reference calculations have been carried out using the Monte Carlo McCARD code in which a double heterogeneity model was used. Verification results show that the DeCART code gives less negative MTC and RTC than the McCARD code does and thus the DeCART code underestimates the multiplication factors at states with high moderator and reflector temperatures. However, the DeCART code predicts more negative FTC than McCARD code does. In the depletion calculation for the HTTR single cell and single block, the error of the DeCART code increases with burnup. While the DeCART code error in a 2-dimensional core depletion calculation decreases with burnup up to around 500 FPD. (author)
International Nuclear Information System (INIS)
The major aim of this work is a sensitivity analysis related to the influence of the different nuclear data libraries on the k-infinity values and on the void coefficient estimations performed for various CANDU fuel projects, and on the simulations related to the replacement of the original stainless steel adjuster rods by cobalt assemblies in the CANDU reactor core. The computations are performed using the Monte Carlo transport codes MCNP5 and MONTEBURNS 1.0 for the actual, detailed geometry and material composition of the fuel bundles and reactivity devices. Some comparisons with deterministic and probabilistic codes involving the WIMS library are also presented
Its version 3.0. The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
The ITS system is described, which is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures. (author)
International Nuclear Information System (INIS)
Consideration is given of a technique and algorithms of constructing neutron trajectories in the Monte-Carlo method taking into account the data on adjoint transport equation solution. When simulating the transport part of transfer kernel the use is made of piecewise-linear approximation of free path length density along the particle motion direction. The approach has been implemented in programs within the framework of the BRAND code system. The importance is calculated in the multigroup P1-approximation within the framework of the DD-30 code system. The efficiency of the developed computation technique is demonstrated by means of solution of two model problems. 4 refs.; 2 tabs
Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code
International Nuclear Information System (INIS)
A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements
Antiproton annihilation physics in the Monte Carlo particle transport code SHIELD-HIT12A
International Nuclear Information System (INIS)
The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An experimental depth dose curve obtained by the AD-4/ACE collaboration was compared with an earlier version of SHIELD-HIT, but since then inelastic annihilation cross sections for antiprotons have been updated and a more detailed geometric model of the AD-4/ACE experiment was applied. Furthermore, the Fermi–Teller Z-law, which is implemented by default in SHIELD-HIT12A has been shown not to be a good approximation for the capture probability of negative projectiles by nuclei. We investigate other theories which have been developed, and give a better agreement with experimental findings. The consequence of these updates is tested by comparing simulated data with the antiproton depth dose curve in water. It is found that the implementation of these new capture probabilities results in an overestimation of the depth dose curve in the Bragg peak. This can be mitigated by scaling the antiproton collision cross sections, which restores the agreement, but some small deviations still remain. Best agreement is achieved by using the most recent antiproton collision cross sections and the Fermi–Teller Z-law, even if experimental data conclude that the Z-law is inadequately describing annihilation on compounds. We conclude that more experimental cross section data are needed in the lower energy range in order to resolve this contradiction, ideally combined with more rigorous models for annihilation on compounds
Uncertainty analysis in the simulation of an HPGe detector using the Monte Carlo Code MCNP5
International Nuclear Information System (INIS)
A gamma spectrometer including an HPGe detector is commonly used for environmental radioactivity measurements. Many works have been focused on the simulation of the HPGe detector using Monte Carlo codes such as MCNP5. However, the simulation of this kind of detectors presents important difficulties due to the lack of information from manufacturers and due to loss of intrinsic properties in aging detectors. Some parameters such as the active volume or the Ge dead layer thickness are many times unknown and are estimated during simulations. In this work, a detailed model of an HPGe detector and a petri dish containing a certified gamma source has been done. The certified gamma source contains nuclides to cover the energy range between 50 and 1800 keV. As a result of the simulation, the Pulse Height Distribution (PHD) is obtained and the efficiency curve can be calculated from net peak areas and taking into account the certified activity of the source. In order to avoid errors due to the net area calculation, the simulated PHD is treated using the GammaVision software. On the other hand, it is proposed to use the Noether-Wilks formula to do an uncertainty analysis of model with the main goal of determining the efficiency curve of this detector and its associated uncertainty. The uncertainty analysis has been focused on dead layer thickness at different positions of the crystal. Results confirm the important role of the dead layer thickness in the low energy range of the efficiency curve. In the high energy range (from 300 to 1800 keV) the main contribution to the absolute uncertainty is due to variations in the active volume. (author)
Accurate simulation of ionization chamber response with the Monte Carlo code PENELOPE
Energy Technology Data Exchange (ETDEWEB)
Sempau, Josep [Technical University of Catalonia (Spain)
2010-07-01
Full text. Ionization chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, during the last 20 years, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects are extremely important when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artifact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics will be discussed in the context of the transport model implemented in the Penelope code. The degree of violation of the Fano theorem for a simple, planar geometry, will be used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It will be shown that, with a suitable choice of transport parameters, Penelope can simulate IC response with an accuracy of the order of 0.1%. (author)
Modeling of realistic pebble bed reactor geometries using the Serpent Monte Carlo code
International Nuclear Information System (INIS)
Highlights: • The explicit stochastic geometry model in Serpent is documented. • A pebble bed criticality benchmark was calculated demonstrating the geometry model. • Stochastic pebble configurations were obtained from discrete element simulations. • Results deviate from experiments but are in line with example calculations. - Abstract: This paper documents the models available in Serpent for high temperature reactor (HTR) calculations. It is supplemented by a calculation example of ASTRA critical pebble bed experiments. In the pebble bed reactor modeling, different methods have been used to model the double heterogeneity problem occurring in pebble bed reactor calculations. A solution was sought to avoid unphysical simplifications in the pebble bed modeling and the stochastic geometry modeling features available in the Monte Carlo code Serpent were applied for exact placement of pebbles and fuel particles. Randomly packed pebble beds were produced in discrete element method (DEM) simulations and fuel particles were positioned randomly inside the pebbles. Pebbles and particles are located using a Cartesian search mesh, which provides necessary computational efficiency. Serpent uses Woodcock delta-tracking which provides efficient neutron tracking in the complicated geometries. This detailed pebble bed modeling approach was tested by calculating the ASTRA criticality benchmark experiment done at the Kurchatov Institute in 2004. The calculation results are in line with the sample calculations provided with the benchmark documentation. The material library selected for the calculations has a major effect on the results. The difference in graphite absorption cross section is considered the cause of this result. The model added in Serpent is very efficient with a calculation time slightly higher than with a regular lattice approximation. It is demonstrated that Serpent can be used for pebble bed reactor calculations with minimal geometric approximations as it
Accurate simulation of ionisation chamber response with the Monte Carlo code PENELOPE
International Nuclear Information System (INIS)
Ionisation chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, for various decades, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects can be sizeable when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artefact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics are discussed in the context of the transport model implemented in the PENELOPE code. The degree of violation of the Fano theorem for a simple, planar geometry, is used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It is shown that, with a suitable choice of transport parameters, PENELOPE simulates IC response with an accuracy of the order of 0.1%.
Accurate simulation of ionization chamber response with the Monte Carlo code PENELOPE
International Nuclear Information System (INIS)
Full text. Ionization chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, during the last 20 years, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects are extremely important when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artifact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics will be discussed in the context of the transport model implemented in the Penelope code. The degree of violation of the Fano theorem for a simple, planar geometry, will be used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It will be shown that, with a suitable choice of transport parameters, Penelope can simulate IC response with an accuracy of the order of 0.1%. (author)
Deep-penetration calculation for the ISIS target station shielding using the MARS Monte Carlo code
International Nuclear Information System (INIS)
A calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation with good statistics, the following three techniques were used in this study. First, the geometry of the bulk shield was three-dimensionally divided into several layers of about 50-cm thickness, and a step-by-step calculation was carried out to multiply the number of penetrated particles at the boundaries between the layers. Second, the source particles in the layers were divided into two parts to maintain the statistical balance on the spatial-flux distribution. Third, only high-energy particles above 20 MeV were transported up to approximately 1 m before the region for benchmark calculation. Finally, the energy spectra of neutrons behind the very thick shield were calculated down to the thermal energy with good statistics, and typically agree well within a factor of two with the experimental data over a broad energy range. The 12C(n,2n)11C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem. In this report, the calculation conditions, geometry and the variance reduction techniques used in the deep-penetration calculation with the MARS14 code are clarified, and several subroutines of MARS14 which were used in our calculation are also given in the appendix. The numerical data of the calculated neutron energy spectra, reaction rates, dose rates and their C/E (Calculation/Experiment) values are also summarized. The
Directory of Open Access Journals (Sweden)
Ilić Radovan D.
2004-01-01
Full Text Available This paper describes the application of SRNA Monte Carlo package for proton transport simulations in complex geometry and different material composition. SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The compound nuclei decay was simulated by our own and the Russian MSDM models using ICRU 63 data. The developed package consists of two codes SRNA-2KG, which simulates proton transport in the combinatorial geometry and SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield’s data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of proton beam characterization by Multi-Layer Faraday Cup, spatial distribution of positron emitters obtained by SRNA-2KG code, and intercomparison of computational codes in radiation dosimetry, indicate the immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in SRNA pack age, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumor.
International Nuclear Information System (INIS)
This paper describes the application of SRNA Monte Carlo package for proton transport simulations in complex geometry and different material composition. SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The compound nuclei decay was simulated by our own and the Russian MSDM models using ICRU 63 data. The developed package consists of two codes: SRNA-2KG, which simulates proton transport in the combinatorial geometry and SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of proton beam characterization by Multi-Layer Faraday Cup, spatial distribution of positron emitters obtained by SRNA-2KG code, and intercomparison of computational codes in radiation dosimetry, indicate the immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumor. (author)
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
Energy Technology Data Exchange (ETDEWEB)
Ilic, Radovan D [Laboratory of Physics (010), Vinca Institute of Nuclear Sciences, PO Box 522, 11001 Belgrade (Serbia and Montenegro); Spasic-Jokic, Vesna [Laboratory of Physics (010), Vinca Institute of Nuclear Sciences, PO Box 522, 11001 Belgrade (Serbia and Montenegro); Belicev, Petar [Laboratory of Physics (010), Vinca Institute of Nuclear Sciences, PO Box 522, 11001 Belgrade (Serbia and Montenegro); Dragovic, Milos [Center for Nuclear Medicine MEDICA NUCLEARE, Bulevar Despota Stefana 69, 11000 Belgrade (Serbia and Montenegro)
2005-03-07
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour.
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
Ilic, Radovan D.; Spasic-Jokic, Vesna; Belicev, Petar; Dragovic, Milos
2005-03-01
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour.
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
International Nuclear Information System (INIS)
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour
Fission gas release modelling at high burnup
International Nuclear Information System (INIS)
A large quantity of experimental data on fission gas release is now available in the public domain. It covers a wide variety of fuel types and burnups of up to more than 70 GWd/tU. This data, together with gas release measurements from British Energy's AGRs, has been used to build a comprehensive validation database for the fuel performance code ENIGMA. Validation of ENIGMA version 5.11 against this database has identified a requirement for model development to improve predictions at high burnup. A modified gas release model has been produced and tested. (author)
International Nuclear Information System (INIS)
Dosimetric studies are necessary for all patients treated with targeted radiotherapy. In order to attain the precision required, we have developed Oedipe, a dosimetric tool based on the MCNPX Monte Carlo code. The anatomy of each patient is considered in the form of a voxel-based geometry created using computed tomography (CT) images or magnetic resonance imaging (MRI). Oedipe enables dosimetry studies to be carried out at the voxel scale. Validation of the results obtained by comparison with existing methods is complex because there are multiple sources of variation: calculation methods (different Monte Carlo codes, point kernel), patient representations (model or specific) and geometry definitions (mathematical or voxel-based). In this paper, we validate Oedipe by taking each of these parameters into account independently. Monte Carlo methodology requires long calculation times, particularly in the case of voxel-based geometries, and this is one of the limits of personalized dosimetric methods. However, our results show that the use of voxel-based geometry as opposed to a mathematically defined geometry decreases the calculation time two-fold, due to an optimization of the MCNPX2.5e code. It is therefore possible to envisage the use of Oedipe for personalized dosimetry in the clinical context of targeted radiotherapy
Energy Technology Data Exchange (ETDEWEB)
Somasundaram, E.; Palmer, T. S. [Department of Nuclear Engineering and Radiation Health Physics, Oregon State University, 116 Radiation Center, Corvallis, OR 97332-5902 (United States)
2013-07-01
In this paper, the work that has been done to implement variance reduction techniques in a three dimensional, multi group Monte Carlo code - Tortilla, that works within the frame work of the commercial deterministic code - Attila, is presented. This project is aimed to develop an integrated Hybrid code that seamlessly takes advantage of the deterministic and Monte Carlo methods for deep shielding radiation detection problems. Tortilla takes advantage of Attila's features for generating the geometric mesh, cross section library and source definitions. Tortilla can also read importance functions (like adjoint scalar flux) generated from deterministic calculations performed in Attila and use them to employ variance reduction schemes in the Monte Carlo simulation. The variance reduction techniques that are implemented in Tortilla are based on the CADIS (Consistent Adjoint Driven Importance Sampling) method and the LIFT (Local Importance Function Transform) method. These methods make use of the results from an adjoint deterministic calculation to bias the particle transport using techniques like source biasing, survival biasing, transport biasing and weight windows. The results obtained so far and the challenges faced in implementing the variance reduction techniques are reported here. (authors)
International Nuclear Information System (INIS)
In this paper, the work that has been done to implement variance reduction techniques in a three dimensional, multi group Monte Carlo code - Tortilla, that works within the frame work of the commercial deterministic code - Attila, is presented. This project is aimed to develop an integrated Hybrid code that seamlessly takes advantage of the deterministic and Monte Carlo methods for deep shielding radiation detection problems. Tortilla takes advantage of Attila's features for generating the geometric mesh, cross section library and source definitions. Tortilla can also read importance functions (like adjoint scalar flux) generated from deterministic calculations performed in Attila and use them to employ variance reduction schemes in the Monte Carlo simulation. The variance reduction techniques that are implemented in Tortilla are based on the CADIS (Consistent Adjoint Driven Importance Sampling) method and the LIFT (Local Importance Function Transform) method. These methods make use of the results from an adjoint deterministic calculation to bias the particle transport using techniques like source biasing, survival biasing, transport biasing and weight windows. The results obtained so far and the challenges faced in implementing the variance reduction techniques are reported here. (authors)
Energy Technology Data Exchange (ETDEWEB)
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2004-06-01
ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2)multigroup codes with adjoint transport capabilities, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.
ITS version 5.0 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2)multigroup codes with adjoint transport capabilities, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability
COOL: A code for Dynamic Monte Carlo Simulation of molecular dynamics
Barletta, Paolo
2012-02-01
Cool is a program to simulate evaporative and sympathetic cooling for a mixture of two gases co-trapped in an harmonic potential. The collisions involved are assumed to be exclusively elastic, and losses are due to evaporation from the trap. Each particle is followed individually in its trajectory, consequently properties such as spatial densities or energy distributions can be readily evaluated. The code can be used sequentially, by employing one output as input for another run. The code can be easily generalised to describe more complicated processes, such as the inclusion of inelastic collisions, or the possible presence of more than two species in the trap. New version program summaryProgram title: COOL Catalogue identifier: AEHJ_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHJ_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1 097 733 No. of bytes in distributed program, including test data, etc.: 18 425 722 Distribution format: tar.gz Programming language: C++ Computer: Desktop Operating system: Linux RAM: 500 Mbytes Classification: 16.7, 23 Catalogue identifier of previous version: AEHJ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 182 (2011) 388 Does the new version supersede the previous version?: Yes Nature of problem: Simulation of the sympathetic process occurring for two molecular gases co-trapped in a deep optical trap. Solution method: The Direct Simulation Monte Carlo method exploits the decoupling, over a short time period, of the inter-particle interaction from the trapping potential. The particle dynamics is thus exclusively driven by the external optical field. The rare inter-particle collisions are considered with an acceptance/rejection mechanism, that is, by comparing a random number to the collisional probability
A Monte Carlo study of the effect of coded-aperture material and thickness on neutron imaging
International Nuclear Information System (INIS)
In this paper, a coded-aperture design for a scintillator-based neutron imaging system has been simulated using a series of Monte Carlo simulations. Using Monte Carlo simulations, work to optimise a system making use of the EJ-426 neutron scintillator detector has been conducted. This type of scintillator has a low sensitivity to gamma rays and is therefore particularly useful for neutron detection in a mixed radiation environment. Simulations have been conducted using varying coded-aperture materials and different coded-aperture thicknesses. From this, neutron images have been produced, compared qualitatively and quantitatively for each case to find the best material for the MURA (modified uniformly redundant array) pattern. The neutron images generated also allow observations on how differing thicknesses of coded-aperture impact the system. A system in which a neutron sensitive scintillator has been used in conjunction with a MURA coded aperture to detect and locate a neutron emitting point source centralised in the system has been simulated. A comparison between the results of the different coded-aperture thicknesses is discussed, via the calculation of system error between the reconstructed source location and the actual location. As the system is small scale with a relatively large step along the axis (0.5 cm), it is justifiable to say that the smaller error values provide satisfactory results, which correlate with only a few centimetres difference in the reconstructed source location to actual source location. A general trend of increasing error can be deduced both as the thickness of the coded-aperture material decreases and the capture cross section of the different materials reduces. (authors)
Thyroid cell irradiation by radioiodines: a new Monte Carlo electron track-structure code
Directory of Open Access Journals (Sweden)
Christophe Champion
2007-09-01
Full Text Available The most significant impact of the Chernobyl accident is the increased incidence of thyroid cancer among children who were exposed to short-lived radioiodines and 131-iodine. In order to accurately estimate the radiation dose provided by these radioiodines, it is necessary to know where iodine is incorporated. To do that, the distribution at the cellular level of newly organified iodine in the immature rat thyroid was performed using secondary ion mass microscopy (NanoSIMS50. Actual dosimetric models take only into account the averaged energy and range of beta particles of the radio-elements and may, therefore, imperfectly describe the real distribution of dose deposit at the microscopic level around the point sources. Our approach is radically different since based on a track-structure Monte Carlo code allowing following-up of electrons down to low energies (~ 10eV what permits a nanometric description of the irradiation physics. The numerical simulations were then performed by modelling the complete disintegrations of the short-lived iodine isotopes as well as of 131I in new born rat thyroids in order to take into account accurate histological and biological data for the thyroid gland.O impacto mais significante do acidente de Chernobyl é o crescimento da incidência de câncer de tireóide em crianças que foram expostas a radioiodos de vida curta e ao Iodo-131. Na estimativa precisa da dose de radiação fornecida por esses radioiodos, é necessário conhecer onde o iodo está incorporado. Para obtermos esse resultado, a distribuição em nível celular de iodo recentemente organificado na tireóde de ratos imaturos foi realizada usando microscopia de massa iônica secundária (NanoSIMS50. Modelos dosimétricos atuais consideram apenas a energia média das partículas beta dos radioelementos e pode, imperfeitamente descrever a distribuição real de dose ao nível microscópico em torno dos pontos pesquisados. Nossa abordagem
A SAS2H/KENO-V.a methodology for a combined 1D/3D full core fuel burnup analysis
International Nuclear Information System (INIS)
An efficient methodology for 3D fuel burnup analysis of LWR reactors is described in this paper. This methodology is founded on coupling Monte Carlo method for 3D calculation of node power distribution, and transport method for depletion calculation in ID Wigner-Seitz equivalent cell for each node independently. The proposed fuel burnup modeling, based on application of SCALE-4.4a control modules SAS2H and KENO-V.a is verified for the case of 2D x-y model of IRIS 15 x 15 fuel assembly (with reflective boundary condition) by using two well benchmarked code systems. The one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. The proposed SAS2H/KENO-V.a methodology was applied for 3D burnup analysis of IRIS-1000 benchmark no.44 core. Detailed keff and power density evolution with burnup are reported. (author)
De Geyter, Gert; Fritz, Jacopo; Camps, Peter
2012-01-01
We present FitSKIRT, a method to efficiently fit radiative transfer models to UV/optical images of dusty galaxies. These images have the advantage that they have better spatial resolution compared to FIR/submm data. FitSKIRT uses the GAlib genetic algorithm library to optimize the output of the SKIRT Monte Carlo radiative transfer code. Genetic algorithms prove to be a valuable tool in handling the multi- dimensional search space as well as the noise induced by the random nature of the Monte Carlo radiative transfer code. FitSKIRT is tested on artificial images of a simulated edge-on spiral galaxy, where we gradually increase the number of fitted parameters. We find that we can recover all model parameters, even if all 11 model parameters are left unconstrained. Finally, we apply the FitSKIRT code to a V-band image of the edge-on spiral galaxy NGC4013. This galaxy has been modeled previously by other authors using different combinations of radiative transfer codes and optimization methods. Given the different...
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The RTP is a light-water moderated and pool-type TRIGA MARK II reactor with power capacity of 1MWt. It was built in 1979 and attained the first criticality on 28 June 1982. The RTP was designed mainly for neutron activation analysis, small angle neutron scattering, neutron radiography, radioisotope production, education and training purposes. It uses standard TRIGA fuel developed by General Atomic in which the zirconium hydride moderator is homogeneously combined with enriched uranium. It has a cylindrical core with which possibility of locating 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector is made of high purity graphite. Because of its relatively small power, it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been carried out several times. Until now, there were 12 configurations of the core, the most recent change being in July 2006. This paper will describe the RTP core calculation using the Monte Carlo MVP code system. VP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation in order to have an accurate and fast Monte Carlo simulation of neutron and photon transport problems. The MVP Monte Carlo code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique. When compared to the conventional scalar method, this code could achieve higher computation speed by up to a factor of 10 on the vector super-computer. The RTP core has been modelled using cylinder geometry along the z-coordinate geometry with the MVP code system while its material cross section data is calculated beforehand. The JENDL3.3 data library was used in the whole calculation. The objectives of the calculation are to calculate the multiplication factor values (keff), fission density and flux distribution from the tally data. The calculation also
Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
International Nuclear Information System (INIS)
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ''end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified
Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
Energy Technology Data Exchange (ETDEWEB)
Wagner, J.C.; DeHart, M.D.
2000-03-01
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.
Analysis of void coefficient in fast spectrum BWR core with Monte Carlo code 'MVP'
International Nuclear Information System (INIS)
An innovative large BWR core concept has been proposed for aiming at fuel breeding as well as negative void reactivity coefficient. The core consists of two types of MOX fuel assemblies. One is a triangular tight lattice bundle 1.6 m in active core height and the other is the same bundle 0.8 m. The ratio of flow area to fuel area of the bundle is set at about 0.5 in order to increase breeding ratio. A neutron-streaming channel that consists of a cavity-can containing helium gas and a flow gap between the cavity-can and the channel box is located above each short bundle. It will decrease void reactivity coefficient by enhancing neutron leakage from the core when the void fraction is increased in the flow gap. A core composed of tight lattice bundles provides a much harder neutron spectrum than that of conventional BWRs but a slightly softer one than that of typical FBRs. The cavity-can and the flow gap will cause a steep gradient of neutron flux. The neutronics for such a complicated core structure could not be properly analyzed by conventional analysis methods. In particular, the analysis of void reactivity coefficient requires a sophisticated method because it deals with a small change in core composition. In the analysis of the void reactivity coefficient, we adopted a three-dimensional Monte Carlo code 'MVP', which has been developed by JAERI and has many advantages such as an easy input form for lattice structures, a short run time and a continuous neutron energy method. The continuous neutron energy method is important for the analysis of this core because fission reactions occur mainly in the resonance energy region, where the evaluation of accurate cross sections is difficult with conventional methods. The library used is JENDL-3.2. The multi-layer structure of lattices is also essential for the analysis because its hard spectrum and relatively long neutron mean free path require a modeling for the full core with a lot of bundles. The analysis indicates that
Energy Technology Data Exchange (ETDEWEB)
Procassini, R.J. [Lawrence Livermore National lab., CA (United States)
1997-12-31
The fine-scale, multi-space resolution that is envisioned for accurate simulations of complex weapons systems in three spatial dimensions implies flop-rate and memory-storage requirements that will only be obtained in the near future through the use of parallel computational techniques. Since the Monte Carlo transport models in these simulations usually stress both of these computational resources, they are prime candidates for parallelization. The MONACO Monte Carlo transport package, which is currently under development at LLNL, will utilize two types of parallelism within the context of a multi-physics design code: decomposition of the spatial domain across processors (spatial parallelism) and distribution of particles in a given spatial subdomain across additional processors (particle parallelism). This implementation of the package will utilize explicit data communication between domains (message passing). Such a parallel implementation of a Monte Carlo transport model will result in non-deterministic communication patterns. The communication of particles between subdomains during a Monte Carlo time step may require a significant level of effort to achieve a high parallel efficiency.
Mairani, A; Valente, M; Battistoni, G; Botta, F; Pedroli, G; Ferrari, A; Cremonesi, M; Di Dia, A; Ferrari, M; Fasso, A
2011-01-01
Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy ((89)Sr, (90)Y, (131)I, (153)Sm, (177)Lu, (186)Re, and (188)Re). Point isotropic...
International Nuclear Information System (INIS)
High energy photon (over 10 MeV) and neutron beams adopted in radiobiology and radiotherapy always produce mixed neutron/gamma-ray fields. The Mg(Ar) ionization chambers are commonly applied to determine the gamma-ray dose because of its neutron insensitive characteristic. Nowadays, many perturbation corrections for accurate dose estimation and lots of treatment planning systems are based on Monte Carlo technique. The Monte Carlo codes EGSnrc, FLUKA, GEANT4, MCNP5, and MCNPX were used to evaluate energy dependent response functions of the Exradin M2 Mg(Ar) ionization chamber to a parallel photon beam with mono-energies from 20 keV to 20 MeV. For the sake of validation, measurements were carefully performed in well-defined (a) primary M-100 X-ray calibration field, (b) primary 60Co calibration beam, (c) 6-MV, and (d) 10-MV therapeutic beams in hospital. At energy region below 100 keV, MCNP5 and MCNPX both had lower responses than other codes. For energies above 1 MeV, the MCNP ITS-mode greatly resembled other three codes and the differences were within 5%. Comparing to the measured currents, MCNP5 and MCNPX using ITS-mode had perfect agreement with the 60Co, and 10-MV beams. But at X-ray energy region, the derivations reached 17%. This work shows us a better insight into the performance of different Monte Carlo codes in photon-electron transport calculation. Regarding the application of the mixed field dosimetry like BNCT, MCNP with ITS-mode is recognized as the most suitable tool by this work.
International Nuclear Information System (INIS)
The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analytical activities in the country are described, as well as international collaborations in this field. Perspectives for future research and development of new applications are finally resumed. (author)
Energy Technology Data Exchange (ETDEWEB)
Sanchez, R.A.; Fernandez V, J.M.; Salvat, F. [Servicio de Oncologia Radioterapica. Hospital Clinico de Barcelona. Villarroel 170 08036 Barcelona (Spain)
1998-12-31
In the present communication it is presented the results of the simulation utilizing the Penelope code (Penetration and Energy loss of Positrons and Electrons) in several applications of radiotherapy which can be the radioactive sources simulation: {sup 192} Ir, {sup 125} I, {sup 106} Ru or the electron beams simulation of a linear accelerator Siemens KDS. The simulations presented in this communication have been on computers of type Pentium PC of 100 throughout 300 MHz, and the times of execution were from some hours until several days depending of the complexity of the problem. It is concluded that Penelope is a very useful tool for the Monte Carlo calculations due to its great ability and its relative handling facilities. (Author)
Verification of a Multi-group Cross Section Library for Burnup Calculation
Energy Technology Data Exchange (ETDEWEB)
Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of); Joo, Hang Yu [Seoul National Univ., Seoul (Korea, Republic of)
2013-05-15
Despite satisfying the estimation of the neutronic parameters without depletion to some extent, it still requires detailed investigation of the behavior of a fuel with strong neutron absorber over its operating life time by nTRACER, the direct whole core calculation code with the conventional semi Predictor-Corrector method. This study is mainly focused on the verification of the newly generated multi-group library for burnup calculation by nTRACER through the analysis of its performance of depletion calculation of UO{sub 2} fuel with strong neutron absorbers such as Gadolinium. Firstly, the depletion calculation results of nTRACER are presented by comparing the evolution of k-inf and the inventories of commonly found important isotopes as a function of burnup in the cases of gadolinia(GAD)-bearing fuel pin and fuel assembly (FA) with those of MCNPX-version.2.6.0. The newly generated multi-group library for burnup calculation by nTRACER was verified through GAD-bearing fuel after the new approach of resonance treatment had been employed. Though very good agreement in the overall effect reflected on the multiplication factor of FA at BOC, the evolution of k-inf along fuel irradiation history was systematically well underestimated by nTRACER when compared to Monte Carlo results.
Burnup studies of the subcritical fusion-driven in-zinerator
International Nuclear Information System (INIS)
A fusion-driven subcritical core, 'In-Zinerator', has been proposed for nuclear waste transmutation [1]. In this concept, a powerful Z-pinch neutron source will produce pulses of 14 MeV neutrons that multiply in a surrounding subcritical core consisting of spent fuel from the LWR fuel cycle or from deep burn high temperature reactors. The proposed design has pulse frequency 0.1 Hz and a thermal power of 3 GWth. The Z-pinch fusion experiment is located at Sandia Laboratories, USA, and can today fire once a day. However, investigations have been made how to increase the frequency to several fires per minute. Each fire yields 300 MJ corresponding to 1020 neutrons per pulse. The source chamber will in the In-Zinerator concept be surrounded by spent fuel to reach an effective multiplication factor, keff, of 0.97. The core will be cooled by liquid lead. In this paper, the burnup of different fuel compositions in the In-Zinerator will be studied as function of initial keff. The Monte Carlo based continuous energy burnup code MCB [2][3]will be used. References: [1] B.B. Cipiti, Fusion Transmutation of Waste and the Role of the In-Zinerator in the Nuclear Fuel Cycle, Sandia Report SAND2006-3522, Sandia National Laboratories, USA, 2006. [2] J. Cetnar, J Wallenius and W Gudowski, MCB: A continuous energy Monte-Carlo burnup simulation code, Actinide and fission product partitioning and transmutation, Proc. of the Fifth Int. Information Exchange Meeting, Mol, Belgium, 25-27 November 1998, 523, OECD/NEA, 1998. [3] http://www.nea.fr/abs/html/nea-1643.html
Comparison of Space Radiation Calculations from Deterministic and Monte Carlo Transport Codes
Adams, J. H.; Lin, Z. W.; Nasser, A. F.; Randeniya, S.; Tripathi, r. K.; Watts, J. W.; Yepes, P.
2010-01-01
The presentation outline includes motivation, radiation transport codes being considered, space radiation cases being considered, results for slab geometry, results from spherical geometry, and summary. ///////// main physics in radiation transport codes hzetrn uprop fluka geant4, slab geometry, spe, gcr,
Results of the isotopic concentrations of VVER calculational burnup credit benchmark No. 2(CB2)
International Nuclear Information System (INIS)
Results of the nuclide concentrations are presented of VVER Burnup Credit Benchmark No. 2(CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is summarized. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2' code and other codes were used for the calculation of the spent fuel concentration. (author)
Burn-up measurements coupling gamma spectrometry and neutron measurement
International Nuclear Information System (INIS)
The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)
International Nuclear Information System (INIS)
BOT3P consists of a set of standard Fortran 77 language programs that gives the users of the deterministic transport codes DORT, TORT, TWODANT, THREEDANT, PARTISN and the sensitivity code SUSD3D some useful diagnostic tools to prepare and check the geometry of their input data files for both Cartesian and cylindrical geometries, including graphical display modules. Users can produce the geometrical and material distribution data for all the cited codes for both two-dimensional and three-dimensional applications and, only in 3-dimensional Cartesian geometry, for the Monte Carlo Transport Code MCNP, starting from the same BOT3P input. Moreover, BOT3P stores the fine mesh arrays and the material zone map in a binary file, the content of which can be easily interfaced to any deterministic and Monte Carlo transport code. This makes it possible to compare directly for the same geometry the effects stemming from the use of different data libraries and solution approaches on transport analysis results. BOT3P Version 5.0 lets users optionally and with the desired precision compute the area/volume error of material zones with respect to the theoretical values, if any, because of the stair-cased representation of the geometry, and automatically update material densities on the whole zone domains to conserve masses. A local (per mesh) density correction approach is also available. BOT3P is designed to run on Linux/UNIX platforms and is publicly available from the Organization for Economic Cooperation and Development (OECD/NEA)/Nuclear Energy Agency Data Bank. Through the use of BOT3P, radiation transport problems with complex 3-dimensional geometrical structures can be modelled easily, as a relatively small amount of engineer-time is required and refinement is achieved by changing few parameters. This tool is useful for solving very large challenging problems, as successfully demonstrated not only in some complex neutron shielding and criticality benchmarks but also in a power
Memory bottlenecks and memory contention in multi-core Monte Carlo transport codes
International Nuclear Information System (INIS)
Highlights: • The performance of nuclear reactor Monte Carlo transport applications is examined. • A “proxy-application” (XSBench) is presented representing the key kernel. • In-depth performance analyses reveal the algorithm is bottlenecked by bandwidth. • Strategies are discussed to improve scalability on next generation HPC systems. - Abstract: We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm – the calculation of macroscopic cross sections – in an effort to expose bottlenecks within multi-core, shared memory architectures
Proceedings of the first symposium on Monte Carlo simulation
International Nuclear Information System (INIS)
The first symposium on Monte Carlo simulation was held at Mitsubishi Research Institute, Otemachi, Tokyo, on 10th and 11st of September, 1998. This symposium was organized by Nuclear Code Research Committee at Japan Atomic Energy Research Institute. In the sessions, were presented orally 21 papers on code development, parallel calculation, reactor physics, burn-up, criticality, shielding safety, dose evaluation, nuclear fusion reactor, thermonuclear fusion plasma, nuclear transmutation, electromagnetic cascade, fuel cycle facility. Those presented papers are compiled in this proceedings. The 21 of the presented papers are indexed individually. (J.P.N.)
Proceedings of the first symposium on Monte Carlo simulation
Energy Technology Data Exchange (ETDEWEB)
NONE
2001-01-01
The first symposium on Monte Carlo simulation was held at Mitsubishi Research Institute, Otemachi, Tokyo, on 10th and 11st of September, 1998. This symposium was organized by Nuclear Code Research Committee at Japan Atomic Energy Research Institute. In the sessions, were presented orally 21 papers on code development, parallel calculation, reactor physics, burn-up, criticality, shielding safety, dose evaluation, nuclear fusion reactor, thermonuclear fusion plasma, nuclear transmutation, electromagnetic cascade, fuel cycle facility. Those presented papers are compiled in this proceedings. The 21 of the presented papers are indexed individually. (J.P.N.)
Effect of Self-Shielding on Burn-Up Calculation of ETRR-2 Reactor
International Nuclear Information System (INIS)
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
Analysis of burnup credit on spent fuel storage
International Nuclear Information System (INIS)
Chemical analyses were carried out on high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins. Measured data of the composition of nuclides from 234U to 242Pu were used for evaluation of ORIGEN-2/82 code. Criticality calculations were executed for the casks which were being designed to store 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for (1) axial and horizontal profiles of burnup, and void history (BWR), (2) operational histories such as control rod insertion history, BPR insertion history and others, and (3) calculational accuracy of ORIGEN-2/82 code on the composition of nuclides. Present evaluation shows that introduction of burnup credit has a substantial merit in criticality safety analysis of the cask, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for present reactivity bias evaluation and showed a possibility of simplifying the reactivity bias evaluation in burnup credit. Finally, adapting procedures of burnup credit such as the burnup meter were evaluated. (author)
User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code
International Nuclear Information System (INIS)
This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate keff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)
International Nuclear Information System (INIS)
The Monte Carlo method was used to build a new code for the simulation of particle transport. Several calculations were done after that for verification, where different sources were used, the source term was obtained using the ORIGEN-S code. Water and lead shield were used with spherical geometry, and the tally results were obtained on the external surface of the shield, afterward the results were compared with the results of MCNPX for verification of the new code. The variance reduction techniques of splitting and Russian Roulette were implemented in the code to be more efficient, by reducing the amount of custom programming required, by artificially increasing the particles being tallied with decreasing the weight. The code shows lower results than the results of MCNPX, this can be interpreted by the effect of the secondary gamma radiation that can be produced by the electron, which is ejected by the primary radiation. In the future a more study will be made on the effect of the electron production and transport, either by a real transport of the electron or by simply using an approximation such the thick target bremsstahlung(TTB) option which is used in MCNPX
Energy Technology Data Exchange (ETDEWEB)
Alnajjar, Alaaddin [Univ. of Science and Technology, Daejeon (Korea, Republic of); Park, Chang Je; Lee, Byunchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-10-15
The Monte Carlo method was used to build a new code for the simulation of particle transport. Several calculations were done after that for verification, where different sources were used, the source term was obtained using the ORIGEN-S code. Water and lead shield were used with spherical geometry, and the tally results were obtained on the external surface of the shield, afterward the results were compared with the results of MCNPX for verification of the new code. The variance reduction techniques of splitting and Russian Roulette were implemented in the code to be more efficient, by reducing the amount of custom programming required, by artificially increasing the particles being tallied with decreasing the weight. The code shows lower results than the results of MCNPX, this can be interpreted by the effect of the secondary gamma radiation that can be produced by the electron, which is ejected by the primary radiation. In the future a more study will be made on the effect of the electron production and transport, either by a real transport of the electron or by simply using an approximation such the thick target bremsstahlung(TTB) option which is used in MCNPX.
Monte Carlo simulation code for photon collection in S(T)EM scintillation detectors
Czech Academy of Sciences Publication Activity Database
Schauer, Petr; Autrata, Rudolf
Ljubljana : Jožef Stefan Institute, 2005, s. 199-200. ISBN 961-6303-69-4. [Multinational Congress on Microscopy /7./. Portorož (SI), 26.06.2005-30.06.2005] R&D Projects: GA ČR(CZ) GA102/04/2144 Keywords : collection of photons * scintillation detector * Monte Carlo simulation Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering
Fuel burnup characteristics for the NRU research reactor
International Nuclear Information System (INIS)
The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U3Si, consisting of particles of U3Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)
Fuel burnup characteristics for the NRU research reactor
Energy Technology Data Exchange (ETDEWEB)
Leung, T.C., E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)
2013-07-01
The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U{sub 3}Si, consisting of particles of U{sub 3}Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)
Calculation study of TNPS spent fuel pool using burnup credit
International Nuclear Information System (INIS)
Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks, the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied. The MONK9A code was used to analyze keff, of different enrichment fuels at different burnups. A reference loading curve was proposed in accordance with the system keff's changing with the burnup of different initially enriched nuclear fuels. The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC. (authors)
Chiesa,
2014-01-01
In recent years, many computer codes, based on Monte Carlo methods or deterministic calculations, have been developed to separately analyze different aspects regarding nuclear reactors. Nuclear reactors are very complex systems, which require an integrated analysis of all the variables which are intrinsically correlated: neutron fluxes, reaction rates, neutron moderation and absorption, thermal and power distributions, heat generation and transfer, criticality coefficients, fuel burnup, e...