Calculations of neutron penetration through graphite medium with Monte Carlo code MCNP
Experiments for fast neutron penetration through graphite are analysed with the continuous energy Monte Carlo code MCNP. Reaction rates and energy spectra obtained with the MCNP are compared with measured values and calculated ones with McBEND code. And validity of penetration calculation with the MCNP is comfirmed. In addition, it is revealed that the MCNP code using Weight-Window method is well applicable to calculations of neutron penetration through graphite up to 70 cm in depth. (author)
The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids
A comparison between the Monte Carlo radiation transport codes MCNP and MCBEND
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)
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
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
MCNP: a general Monte Carlo code for neutron and photon transport
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
MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2
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
MCNP, a general Monte Carlo code for neutron and photon transport: a summary
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
The present of shielding analysis with nuclear data for continuous energy Monte Carlo code MCNP
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)
Installation of Monte Carlo neutron and photon transport code system MCNP4
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)
Sohrabpour, M. [Gamma Irradiation Center, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Shahriari, M. [Physics Department, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Zarifian, V.; Moghadam, K.K. [Nuclear Research Center, Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of)
1999-04-01
A borehole experiment using prompt gamma neutron activation analysis has been performed in a large sample box having a volume of 1 m{sup 3}. Brine solutions having a salt concentration in the range of 0-10 wt% of sodium chloride has been used. Chlorine prompt gamma spectral response as a function of the salt concentrations have been obtained. A simulation of the above experiments has also been carried out using the MCNP4A Monte Carlo code. Comparison of the experimental spectral response versus the simulated MCNP4A data has produced excellent agreement. In view of the good benchmark data it is proposed that due to the inherent problems associated with the ordinary calibration procedures for the borehole logging tools, one could employ a combined calibration/simulation scheme to circumvent these difficulties and achieve more effective results.
Parallel processing of Monte Carlo code MCNP for particle transport problem
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)
Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics
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)
MCNP: a general Monte Carlo code for neutron and photon transport
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
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.)
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
Performance of the improved version of Monte Carlo Code A{sup 3}MCNP for cask shielding design
Hasegawa, T. [Mitsubishi Heavy Industries, Yokohama (Japan); Ueki, K. [Tokai Univ., Kanagawa (Japan); Sato, O. [Mitsubishi Research Inst., Tokyo (Japan); Sjoden, G.E. [Dept. of Nuclear and Radiological Engineering, Univ. of Florida, Gainesville, FL (United States); Miyake, Y.; Ohmura, M.; Haghighat, A.
2004-07-01
A{sup 3}MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A{sup 3}MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A{sup 3}MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A{sup 3}MCNP (referred to as A{sup 3}MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A{sup 3}MCNPV for cask neutron and gamma-ray shielding problem.
MCNP: a general Monte Carlo code for neutron and photon transport
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
MCNP trademark Monte Carlo: A precis of MCNP
MCNP trademark is a general purpose three-dimensional time-dependent neutron, photon, and electron transport code. It is highly portable and user-oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 years and an ongoing commitment to excellence
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
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
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)
Monte Carlo Simulation Of Absorbed Dose From LINAC On VOXEL Phantom By Using MCNP5 Code
In this work, we use MCNP5 code for simulating dose distribution calculation from LINAC on phantom CT. CT images obtained from cancer treatment cases at Cho Ray hospital. In order to transform CT images into data of MCNP5 input file we also build a program CODIM by using MATLAB programming software. The results show that there is a difference of 5% in comparison to DSS program - a semi-empirical simulation program which is being used for treatment planning in Cho Ray hospital. (author)
Uncertainty analysis in the simulation of an HPGe detector using the Monte Carlo Code MCNP5
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)
In order to improve the accuracy and calculating speed of shielding analyses, MCNP 4, a Monte Carlo neutron and photon transport code system, has been parallelized and measured of its efficiency in the highly parallel distributed memory type computer, AP1000. The code has been analyzed statically and dynamically, then the suitable algorithm for parallelization has been determined for the shielding analysis functions of MCNP 4. This includes a strategy where a new history is assigned to the idling processor element dynamically during the execution. Furthermore, to avoid the congestion of communicative processing, the batch concept, processing multi-histories by a unit, has been introduced. By analyzing a sample cask problem with 2,000,000 histories by the AP1000 with 512 processor elements, the 82 % of parallelization efficiency is achieved, and the calculational speed has been estimated to be around 50 times as fast as that of FACOM M-780. (author)
Electron absorbed dose comparison between MCNP5 and Penelope Monte Carlo code for microdosimetry
The objective of the present work was to compare electron absorbed dose results between two widespread used codes in international scientific community: MCNP5 and Penelope-2003. Individual water spheres with masses between 10-9 g up to 10-3 g immersed in an infinite water medium (density of 1g/cm3) and monoenergetic electron sources with energy from 0.002 MeV to 0.1 MeV have been considered. The absorbed dose in the spheres was evaluated by both codes and the relative differences have been quantified. The results shown that Penelope gives, in general, higher results that, in some cases saturate or reach a maximum point and then rapidly drops. Particularly, for the 40 keV electron source we have done additional tests in three different scenarios: more points in the region of lower masses to a better definition of the curve behavior; MCNP used 200 substeps and Penelope was set to a full detail history methodology, and almost same parameters of case B but with the density of exterior medium increased to 10 g/cm3. The three cases show the influence of the backscattering that contribute with an important fraction of absorbed dose, finally we can infer a range of reliability to use the codes in this kind of simulations: both codes can calculate close results for up to 10-4 g.Even though MCNP5 uses the condensed history method, if simulation parameters are chosen carefully it can reproduce results very close to those obtained using detailed history mode. In some cases, the use of higher number of electron substeps causes significant differences in the result. (author)
Monte Carlo simulation has been used by many researchers to calculate organ and effective dose of patients arising from conventional X-ray examinations. In this study the radiation transport code, MCNP4C, has been used to perform Monte Carlo simulations to estimate radiation dose delivered to different organs in conventional X-ray examinations. Materials and Methods: In this work we have made use of ORNL mathematical phantoms with few modifications which have been made. The source has been defined as a point source, emitting photons into a solid angle. The X-ray beam was shaped by a collimator to produce a rectangular field at the midline of the phantom. Results: to validate the simulation executed in this study normalized organs doses to unit ESD for hermaphrodite phantom were computed. Our results were compared with corresponding values presented by NRPB. In general organs doses obtained by application of MCNP-4C (present study) and corresponding values presented in NRPB were in good agreement. For further evaluation of our phantom, the values acquired for organ and effective doses by MCNP-4C and ODS-60 were compared. Conclusion: the technique we have developed is capable of estimating organ and effective doses with a better accuracy than dose values obtained by employment of NRPB and ODS-60 technique
A good model on experimental data (criticality, control rod worth, and fuel element worth distributions) is encouraged to provide from the Musashi-TRIGA Mark 2 reactor. In the previous paper, as the keff values for different fuel loading patterns had been provided ranging from the minimum core to the full one, the data would be candidate for an ICSBEP evaluation. Evaluation of the control rod worth and fuel element worth distributions presented in this paper could be an excellent benchmark data applicable for validation of calculation technique used in the field of modern research reactor. As a result of simulation on the TRIGA-2 benchmark experiment, which was performed by three-dimensional continuous-energy Monte Carlo code (MCNP4A), it was found that the MCNP calculated values of control rod worth were consisted to the experimental data for both rod-drop and period methods. And for the fuel and the graphite element worth distributions, the MCNP calculated values agreed well with the measured ones though consideration of real control rod positions was needed for calculating fuel element reactivity positioned in inner ring. (G.K.)
The criticality analysis of the TRIGA-II benchmark experiment at the Musashi Institute of Technology Research Reactor (MuITR, 100kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). To minimize errors due to an inexact geometry model, all fresh fuels and control rods as well as vicinity of the core were precisely modeled. Effective multiplication factors (keff) in the initial core critical experiment and in the excess reactivity adjustment for the several fuel-loading patterns as well as the fuel element reactivity worth distributions were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated keff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements (fuels or graphite elements were added only to the outer-ring), but the discrepancy increased to 1.8%Δk/k for the some fuel-loading patterns (graphite elements were inserted into the inner-ring). The comparison result of the fuel element worth distribution showed above tendency. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicates that the Monte Carlo model is enough to simulate criticality of the TRIGA-II reactor. (author)
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.
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)
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)
Parreno Z, F.; Paucar J, R.; Picon C, C. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, San Borja, Lima 41 (Peru)
1998-12-31
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)
MCNP-POLIMI v1.0, Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities
1 - Description of program or function: MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. Based on the Los Alamos National Laboratory code MCNP4C (formerly distributed as CCC-700), MCNP-PoliMi was developed to simulate time-analysis quantities. In particular, the code includes the correlation between neutron interaction and the corresponding photon production. Conversely to the technique adopted by standard MCNP, MCNP PoliMi samples secondary photons according to the neutron collision type. A post-processing code, i.e. the Matlab script 'postmain', is included and can be tailored to model specific detector characteristics. These features make MCNP-PoliMi a versatile tool to simulate particle interactions and detection processes. 2 - Methods: MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. The MCNP-PoliMi code was developed to simulate each neutron-nucleus interaction as closely as possible. In particular, neutron interaction and
This work has been performed within the frame of the European Union ORAMED project (Optimisation of Radiation protection for Medical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity Hp(3). In this study, a set of energy- and angular-dependent conversion coefficients (Hp(3)/Ka), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The Hp(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of Hp(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account. (authors)
Kim, Kyung-O; Roh, Gyuhong; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-10-15
As a result, there was a difference within about 300-400 pcm between keff values at each enrichment due to the difference of codes and nuclear data used in the evaluations. The FTC was changed to be less negative with the increase of uranium enrichment, and it followed the form of asymptotic curve. However, it is necessary to perform additional study for investigating what factor causes the differences more than two standard deviation (2σ) among the FTCs at partial enrichment region. The interaction probability of incident neutron with nuclear fuel is depended on the relative velocity between the neutron and the target nuclei. The Fuel Temperature Coefficient (FTC) is defined as the change of Doppler effect with respect to the change in fuel temperature without any other change such as moderator temperature, moderator density, etc. In this study, the FTCs for UO{sub 2} fuel were evaluated by using MCNP6.1 and KENO6 codes based on a Monte Carlo method. In addition, the latest neutron cross-sections (ENDF/B-VI and VII) were applied to analyze the effect of these data on the evaluation of FTC, and nuclear data used in MCNP calculations were generated from the makxsf code. An evaluation of the Doppler effect and FTC for UO{sub 2} fuel widely used in PWR was conducted using MCNP6.1 and KENO6 codes. The ENDF/B-VI and VII were also applied to analyze what effect these data has on those evaluations. All cross-sections needed for MCNP calculation were produced using makxsf code. The calculation models used in the evaluations were based on the typical PWR UO{sub 2} lattice.
Parallel MCNP Monte Carlo transport calculations with MPI
The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
Morgan C. White
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of 6Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high 6Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x1013 n cm-2 per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2
As part of developing its nuclear infrastructure base, the National Science and Technology Center Nuclear (CNSTN) examines the technical feasibility of setting up a new installation of subcritical assembly. Our study focuses on determining the neutron parameters of a nuclear zero power reactor based on Monte Carlo simulation MCNP. The objective of the simulation is to model the installation, determine the effective multiplication factor, and spatial distribution of neutron flux.
MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials
1 - Description of program or function: MCNP4B-GN is used to treat (gamma,n) production and transport in medical accelerator heads, to study the undesired neutron dose to patients, employing a single code for both the electromagnetic and the neutron transport. 2 - Methods: The code simulates the production of giant dipole resonance (GDR) photoneutrons in thick layers of high-Z elements. Neutrons are generated through evaporation of the compound nucleus or through direct channel; the photoneutron origin coordinates, evaluated as the electromagnetic shower develops are taken into account as well as the energy spectrum of the generated photoneutron. Photoneutron production routines have been inserted into MCNP4B, thus allowing a handling of complicated geometries with a single input definition, a fundamental requirement for this kind of application. The modifications to the standard MCNP4B were made as a 'patch' (i.e. a series of instructions on how to modify the basic code) which is distributed together with a preprocessor. The preprocessor reads the patch, reads MCNP4B and writes MCNP4B-GN. A new subroutine 'GAMMN' has been written, analogous to the subroutine 'ACEGAM' for (n,gamma). GAMMN is called from subroutine 'COLIDP' at a photon collision with one of the 6 elements in question when the energy of the photon is such that there is a non-zero probability of producing a neutron. In analogy with the (n,gamma) capability, the production or not of a neutron is not correlated with the subsequent history of the photon and in particular the choice of the type of photon collision at that spatial point. 3 - Restrictions on the complexity of the problem: The upper energy limit, imposed by theoretical considerations, is 30 MeV for photons and 20 MeV for photoneutrons. The physical model breaks down at higher photon energies, whilst the neutron transport performed with MCNP cannot be simulated for neutron energies greater than 20 MeV (the neutron energy regime in MCNP is
The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared
MCNP4, a parallel Monte Carlo implementation on a workstation network
The Monte Carlo code MCNP4 has been implemented on a workstation network to allow parallel computing of Monte Carlo transport processes. This has been achieved by making use of the communication tool PVM (Parallel Virtual Machine) and introducing some changes in the MCNP4 code. The PVM daemons and user libraries have been installed on different workstations to allow working on the same platform. Essential features of PVM and the structure of the parallelized MCNP4 version are discussed in this paper. Experiences are described and problems are explained and solved with the extended version of MCNP. The efficiency of the parallelized MCNP4 is assessed for two realistic sample problems from the field of fusion neutronics. Compared with the fastest workstation in the network, a speed-up factor near five has been obtained by using a network of ten workstations, different in architecture and performance. (orig.)
Monte Carlo N-Particle Transport Code (MCNP) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle is internationally recognized as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant was used to enhance the capabilities of the MCNP Visual Editor to allow it to read in both 2D and 3D Computer Aided Design (CAD) files, allowing the user to electronically generate a valid MCNP input geometry
MCNP-REN a Monte Carlo tool for neutron detector design
Abhold, M E
2002-01-01
The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel w...
Nuclear densimeter of soil simulated in MCNP-4C code
The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)
Distributed processor Monte Carlo. MCNP results on a 16-node IBM cluster
MCNP, a general 3-D Monte Carlo neutron, photon, and electron transport code developed by LANL (Los Alamos National Laboratory, X-6 Group) is presented. With the widespread use of high-performance workstations, an increased interest for multi-processing MCNP on distributed memory systems has emerged. Such systems, connected by high-speed communication networks, provide the capability for codes like MCNP to achieve order-of-magnitude higher performance over current common memory systems. The software communication package chosen for use within MCNP is Parallel Virtual Machine (PVM), available as public domain software. This package supports a variety of communication networks and computer systems. The PVM version of MCNP on a 16 processor IBM RS/6000 cluster produced speedups that approach the number of processors. (author)
Benchmark study of TRIPOLI-4 through experiment and MCNP codes
Michel, M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Coulon, R. [Canberra France, F-78182 Saint Quentin en Yvelines (France); Normand, S. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Huot, N.; Petit, O. [CEA, DEN DANS, SERMA, F-91191 Gif-sur-Yvette (France)
2011-07-01
Reliability on simulation results is essential in nuclear physics. Although MCNP5 and MCNPX are the world widely used 3D Monte Carlo radiation transport codes, alternative Monte Carlo simulation tools exist to simulate neutral and charged particles' interactions with matter. Therefore, benchmark are required in order to validate these simulation codes. For instance, TRIPOLI-4.7, developed at the French Alternative Energies and Atomic Energy Commission for neutron and photon transport, now also provides the user with a full feature electron-photon electromagnetic shower. Whereas the reliability of TRIPOLI-4.7 for neutron and photon transport has been validated yet, the new development regarding electron-photon matter interaction needs additional validation benchmarks. We will thus demonstrate how accurately TRIPOLI-4's 'deposited spectrum' tally can simulate gamma spectrometry problems, compared to MCNP's 'F8' tally. The experimental setup is based on an HPGe detector measuring the decay spectrum of an {sup 152}Eu source. These results are then compared with those given by MCNPX 2.6d and TRIPOLI-4 codes. This paper deals with both the experimental aspect and simulation. We will demonstrate that TRIPOLI-4 is a potential alternative to both MCNPX and MCNP5 for gamma-electron interaction simulation. (authors)
Benchmark study of TRIPOLI-4 through experiment and MCNP codes
Reliability on simulation results is essential in nuclear physics. Although MCNP5 and MCNPX are the world widely used 3D Monte Carlo radiation transport codes, alternative Monte Carlo simulation tools exist to simulate neutral and charged particles' interactions with matter. Therefore, benchmark are required in order to validate these simulation codes. For instance, TRIPOLI-4.7, developed at the French Alternative Energies and Atomic Energy Commission for neutron and photon transport, now also provides the user with a full feature electron-photon electromagnetic shower. Whereas the reliability of TRIPOLI-4.7 for neutron and photon transport has been validated yet, the new development regarding electron-photon matter interaction needs additional validation benchmarks. We will thus demonstrate how accurately TRIPOLI-4's 'deposited spectrum' tally can simulate gamma spectrometry problems, compared to MCNP's 'F8' tally. The experimental setup is based on an HPGe detector measuring the decay spectrum of an 152Eu source. These results are then compared with those given by MCNPX 2.6d and TRIPOLI-4 codes. This paper deals with both the experimental aspect and simulation. We will demonstrate that TRIPOLI-4 is a potential alternative to both MCNPX and MCNP5 for gamma-electron interaction simulation. (authors)
BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies
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
Investigation of the applicability of MCNP code to complicated geometries
Applicability of MCNP code, which is a general purpose Monte Carlo code for particle transport problems, to complicated geometries, has been investigated as a study in Human Acts Simulation Program (HASP), in which basic studies for intelligent robot for patrol and inspection of nuclear facilities are being performed. In HASP, basic software systems simulating the behavior of intelligent robot of human shape working in Japan Research Reactor No.3 are being developed. The aim of Dose Evaluation system in HASP is to establish the methodology to evaluate irradiation damage of the LSI/VLSI circuits embedded within a robot body and to give design criteria of intelligent robot. Monte Carlo method is used to solve particle transport problem in a complicated geometry such as robot body. Preliminary evaluation to establish the methodology has been conducted using continuous energy Monte Carlo code, MCNP with the anthropomorphic phantom. The phantom has the same degree of geometric complexity as robot body and is widely used for the calculation of the effective dose equivalent for radiological protection. It allowed us to verify the validity of the methodology by comparison of calculation results with the data in ICRP Pub. 51. In this report, the method used in the calculation of effective dose equivalent, visualization system supporting visualization of input data for complicated geometry and the results in the evaluation of validity of the method by the comparison of the calculated results with the data in the ICRP publication are described. (author)
Monte Carlo parameter studies and uncertainty analyses with MCNP5
A software tool called mcnppstudy has been developed to automate the setup, execution, and collection of results from a series of MCNP5 Monte Carlo calculations. This tool provides a convenient means of performing parameter studies, total uncertainty analyses, parallel job execution on clusters, stochastic geometry modeling, and other types of calculations where a series of MCNP5 jobs must be performed with varying problem input specifications. (authors)
Fission Matrix Capability for MCNP Monte Carlo
Carney, Sean E. [Los Alamos National Laboratory; Brown, Forrest B. [Los Alamos National Laboratory; Kiedrowski, Brian C. [Los Alamos National Laboratory; Martin, William R. [Los Alamos National Laboratory
2012-09-05
In a Monte Carlo criticality calculation, before the tallying of quantities can begin, a converged fission source (the fundamental eigenvector of the fission kernel) is required. Tallies of interest may include powers, absorption rates, leakage rates, or the multiplication factor (the fundamental eigenvalue of the fission kernel, k{sub eff}). Just as in the power iteration method of linear algebra, if the dominance ratio (the ratio of the first and zeroth eigenvalues) is high, many iterations of neutron history simulations are required to isolate the fundamental mode of the problem. Optically large systems have large dominance ratios, and systems containing poor neutron communication between regions are also slow to converge. The fission matrix method, implemented into MCNP[1], addresses these problems. When Monte Carlo random walk from a source is executed, the fission kernel is stochastically applied to the source. Random numbers are used for: distances to collision, reaction types, scattering physics, fission reactions, etc. This method is used because the fission kernel is a complex, 7-dimensional operator that is not explicitly known. Deterministic methods use approximations/discretization in energy, space, and direction to the kernel. Consequently, they are faster. Monte Carlo directly simulates the physics, which necessitates the use of random sampling. Because of this statistical noise, common convergence acceleration methods used in deterministic methods do not work. In the fission matrix method, we are using the random walk information not only to build the next-iteration fission source, but also a spatially-averaged fission kernel. Just like in deterministic methods, this involves approximation and discretization. The approximation is the tallying of the spatially-discretized fission kernel with an incorrect fission source. We address this by making the spatial mesh fine enough that this error is negligible. As a consequence of discretization we get a
Heart simulation with surface equations for using on MCNP code
Rezaei-Ochbelagh, D.; Salman-Nezhad, S.; Asadi, A.; Rahimi, A.
2011-12-01
External photon beam radiotherapy is carried out in a way to achieve an "as low as possible" a dose in healthy tissues surrounding the target. One of these surroundings can be heart as a vital organ of body. As it is impossible to directly determine the absorbed dose by heart, using phantoms is one way to acquire information around it. The other way is Monte Carlo method. In this work we have presented a simulation of heart geometry by introducing of different surfaces in MCNP code. We used 14 surface equations in order to determine human heart modeling. Those surfaces are borders of heart walls and contents.
Heart simulation with surface equations for using on MCNP code
External photon beam radiotherapy is carried out in a way to achieve an 'as low as possible' a dose in healthy tissues surrounding the target. One of these surroundings can be heart as a vital organ of body. As it is impossible to directly determine the absorbed dose by heart, using phantoms is one way to acquire information around it. The other way is Monte Carlo method. In this work we have presented a simulation of heart geometry by introducing of different surfaces in MCNP code. We used 14 surface equations in order to determine human heart modeling. Those surfaces are borders of heart walls and contents.
Salome, Jean A.D.; Cardoso, Fabiano; Faria, Rochkhudson B.; Pereira, Claubia, E-mail: jadsalome@yahoo.com.br, E-mail: fabinuclear@yahoo.com.br, E-mail: rockdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2015-07-01
The IPEN/MB-01 reactor, located in the city of Sao Paulo - Brazil, reached its first criticality on the year of 1988. The reactor is characterized by a low output power of 100 W only, even because its purpose is to produce knowledge about nuclear power plants on a smaller geometric scale without the requirement of an extremely complex cooling system. The use of devices such as this it is very interesting because it achieves the demands of nuclear engineering about the neutronic parameters needed in the design of large nuclear plants through relatively simple and inexpensive methods. In this paper, the computational mathematical code MCNP5 is used to perform the calculation of the thermal neutron flux in the core of the IPEN/MB-01 reactor. To do this is used an experiment from the LEU-COMP-THERM-077 benchmark that represents the standard rectangular configuration of the IPEN/MB-01 reactor. The thermal neutron flux is calculated at some axial planes of different heights and, after that, axial profiles of the thermal neutron flux are done and compared to experimental results issued previously. The experimental values used as reference refer to a cylindrical configuration of the core of the reactor. Finally, the pertinence and relevance of the results are checked. With this work is expected to produce more knowledge about the dynamics of neutron flux in the core of the IPEN/MB-01 reactor. (author)
Parallel processing Monte Carlo radiation transport codes
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
MCNP variance reduction overview
The MCNP code is rich in variance reduction features. Standard variance reduction methods found in most Monte Carlo codes are available as well as a number of methods unique to MCNP. We discuss the variance reduction features presently in MCNP as well as new ones under study for possible inclusion in future versions of the code
Monte Carlo importance sampling for the MCNP{trademark} general source
Lichtenstein, H.
1996-01-09
Research was performed to develop an importance sampling procedure for a radiation source. The procedure was developed for the MCNP radiation transport code, but the approach itself is general and can be adapted to other Monte Carlo codes. The procedure, as adapted to MCNP, relies entirely on existing MCNP capabilities. It has been tested for very complex descriptions of a general source, in the context of the design of spent-reactor-fuel storage casks. Dramatic improvements in calculation efficiency have been observed in some test cases. In addition, the procedure has been found to provide an acceleration to acceptable convergence, as well as the benefit of quickly identifying user specified variance-reduction in the transport that effects unstable convergence.
Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code
C. A. M. Silva; J. A. D. Salomé; B. T. Guerra; Pereira, C; Costa, A. L.; Veloso, M. A. F.; M. A. B. C. Menezes; Dalle, H. M.
2014-01-01
In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport) code. The sensitivity analyses included sma...
A comparison is made of results obtained from neutron transmissions analysis of RPV performed by MCNP with ENDF/B-VI and JENDL-3.1 iron data. At first, a one-dimensional discrete ordinates transport calculation using VITAMIN-C fine-group library based on ENDF/B-IV was performed for a cylindrical model of a PWR to generate the source spectrum at the front of the RPV. And then, the transmission of neutrons through RPV was calculated by MCNP with the moderated fission spectrum incident on the vessel face. For these ENDF/B-IV, -VI and JENDL-3.1 iron data were processed into continuous energy point data form by NJOY91.91. The fast neutron fluxes and dosimeter reaction rates through RPV using each iron data were intercompared
Monte Carlo simulation of the Greek research reactor neutron irradiation positions using MCNP
Prediction of neutron flux at the irradiation devices of a research reactor facility is essential for the design and evaluation of experiments involving material irradiations. A computational model of the Greek Research Reactor (GRR-1) was developed using the Monte Carlo code MCNP with continuous energy neutron cross-section data evaluations from ENDF/B-VI library. The model included detailed geometrical representation of the fuel and control assemblies, beryllium reflectors, irradiation devices and the graphite pile. The MCNP model was applied to predict neutron flux at the in-pool irradiation positions and the graphite pile. The MCNP estimated neutron fluxes were compared with measurements using activation foils and a good agreement between calculated and experimental results was observed. (author)
Regulation and optimization of the design of Qimingxing experimental facility utilizing MCNP code
Qimingxing Experimental Facility is designed to validate principle of Accelerator Driven Systems (ADS). Monte Carlo Code (MCNP) is used to carry out Keff calculations. The thickness of thermal zone and the pitch of fuel rods in thermal zone are optimized. Calculation results show that the optimized design can meet the Keff requirement of the facility. (authors)
Calculate Some Characteristic Parameters Of VVER-1000's Fuel Assembly By MCNP4C2 Code
This report presents the descriptions of parameters characteristics of the LEU and MOX Fuel Assemblies of VVER-1000 reactor, and calculation results such as infinite neutron multiplication factor kinf, two groups energies constants, neutron flux distribution by using Monte Carlo code MCNP. (author)
In this study, azimuthal particle redistribution (APR), and azimuthal particle rotational splitting (APRS) methods are implemented in MCNPX2.4 source code. First of all, the efficiency of these methods was compared to two tallying methods. The APRS is more efficient than the APR method in track length estimator tallies. However in the energy deposition tally, both methods have nearly the same efficiency. Latent variance reduction factors were obtained for 6, 10 and 18 MV photons as well. The APRS relative efficiency contours were obtained. These obtained contours reveal that by increasing the photon energies, the contours depth and the surrounding areas were further increased. The relative efficiency contours indicated that the variance reduction factor is position and energy dependent. The out of field voxels relative efficiency contours showed that latent variance reduction methods increased the Monte Carlo (MC) simulation efficiency in the out of field voxels. The APR and APRS average variance reduction factors had differences less than 0.6% for splitting number of 1000. -- Highlights: ► The efficiency of APR and APRS methods was compared to two tallying methods. ► The APRS is more efficient than the APR method in track length estimator tallies. ► In the energy deposition tally, both methods have nearly the same efficiency. ► Variance reduction factors of these methods are position and energy dependent.
Problem and solution of tally segment card in MCNP code
Wrong results may be given when FS card (tally segment card) was used for tally with other tally cards in Monte Carlo code MCNP. According to the comparison of segment tally results which were obtained by FS card of three different models of the same geometry, the tally results of fuel regions were found to be wrong in fill pattern. The reason is that the fuel cells were described by Universe card and FILL card, and the filled cells were always considered at Universe card definition place. A proposed solution was that the segment tally for filled cells was done at Universe card definition place. Radial flux distribution of one example was calculated in this way. The results show that the fault of segment tally with FS card in fill pattern could be solved by this method. (authors)
Modeling the PUSPATI TRIGA Reactor using MCNP code
The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core and shielding with literally no physical approximation. (author)
Qualification of MCNP4A code with a lithium-6 neutron detector's
This paper is a qualification of MCNP code (Monte Carlo N-Particle transport code system). In this study MCNP code has been used for the simulation of a neutron detector. The purpose of this work is benchmarking of the code by comparing its results with the physical measurements made by Wang, Jensen and Czirr of neutron detection in a Lithium-6 enriched glass scintillator with two different moderator materials and two different neutron sources. The purpose of the referenced paper was to test to which extent MCNP could predict within a 1% accuracy the relative efficiency of a neutron detector under measurable conditions. For this purpose, the efficiency of ''6 Li-glass moderating neutron detector using Lucite as moderator relative to the efficiency of the detector using mineral oil as moderator has been calculated using the Monte Carlo transport code MCNP-4A, and the results compared with the measurements. The experiments were performed using two different fast neutron sources, a ''252 Cf fission neutrons sources and the 12 MeV neutrons produced by a Van de Graff accelerator. (Author) 2 refs
Computers cluster for parallel processing with the MCNP4B code: Linux + PVM + MCNP4B
Procedures and results obtained in the implantation process of a microcomputers cluster (Linux and PVM) to reduce processing time in MCNP simulations are described. The necessities and specificity concerned to the hardware, the Linux operational system, the PVM software package and an adequate compiled version for MCNP4B are presented. The system was tested to verify the gain in the processing time in function of the number of computers in the cluster using the MCNP input developed to simulate the research reactor TRIGA IPR - R1 of the CDTN/CNEN. Advantages and disadvantages of MCNP code generated with several compilers are discussed, as well as the performance of the executable code generated by each compiler is compared. The results show the viability of the system. It is possible to reduce significantly the simulations processing time of the MCNP code by utilizing PVM computers cluster. It was found that the cost/performance of such a system is often better than the sequential machine (workstations or even mainframes). (author)
C.C. Arwui
2011-05-01
Full Text Available Design objectives for brachytherapy treatment facilities require sufficient shielding to reduce primary and scatter radiation to design limit in order to limit exposure to patients, staff and the general public. The primary aim of this study is to verify whether shielding of the brachytherapy unit at the Korle Bu teaching Hospital in Ghana provides adequate protection in order to assess any radiological health and safety impact and also test the suitability of other available sources. The study evaluates the effectiveness of the biological shielding design of a Cs-137 brachytherapy unit at the Korle-Bu Teaching Hospital in Ghana using MCNP5. The facility was modeled based on the design specifications for LDR Cs-137, MDR Cs-137, HDR Co-60 and HDR Ir-192 treatment modalities. The estimated dose rate ranged from (0.01-0.15 μSv/h and (0.37-3.05 μSv/h for the existing initial and decayed activities of LDR Cs-137 for the public and controlled areas respectively, (0.03-0.57 μSv/h and (1.53-8.06 μSv/h for MDR Cs-137, (7.47-59.46 μSv/h and (144.87-178.74 μSv/h for HDR Co- 60, (0.13-6.95 μSv/h and (19.47-242.98 μSv/h for HDR Ir-192 for the public and controlled areas respectively. The results were verified by dose rates measurement for the current LDR setup at the Brachytherapy unit and agreed quiet well. It was also compared with the reference values of 0.5 μSv/h for public areas and 7.5 μSv/h for controlled areas respectively. It can be concluded that the shielding is adequate for the existing source.
The MCNPX Monte Carlo Radiation Transport Code
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
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.
Hammam Oktajianto
2015-01-01
Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW.
MCNP Perturbation Capability for Monte Carlo Criticality Calculations
The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of keff in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward
MCNP Perturbation Capability for Monte Carlo Criticality Calculations
Hendricks, J.S.; Carter, L.L.; McKinney, G.W.
1999-09-20
The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k{sub eff} in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward.
Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60
Kim Kyung-O
2016-01-01
Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.
Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises
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%.
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...
Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis
O. Mazzantini; F. D'Auria; M. Pecchia; Parisi, C
2011-01-01
The geometrical complexity and the peculiarities of Atucha-2 PHWR require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Core models of Atucha-2 PHWR were developed using both MCNP5 and KENO-VI codes. The developed models were applied for calculating reactor criticality states at beginning of life, reactor cell constants, and control rods volumes. The last two applications were relevant for performing successive three dimensional neutron kinetic ana...
Physics and Algorithm Enhancements for a Validated MCNP/X Monte Carlo Simulation Tool, Phase VII
McKinney, Gregg W [Los Alamos National Laboratory
2012-07-17
Currently the US lacks an end-to-end (i.e., source-to-detector) radiation transport simulation code with predictive capability for the broad range of DHS nuclear material detection applications. For example, gaps in the physics, along with inadequate analysis algorithms, make it difficult for Monte Carlo simulations to provide a comprehensive evaluation, design, and optimization of proposed interrogation systems. With the development and implementation of several key physics and algorithm enhancements, along with needed improvements in evaluated data and benchmark measurements, the MCNP/X Monte Carlo codes will provide designers, operators, and systems analysts with a validated tool for developing state-of-the-art active and passive detection systems. This project is currently in its seventh year (Phase VII). This presentation will review thirty enhancements that have been implemented in MCNPX over the last 3 years and were included in the 2011 release of version 2.7.0. These improvements include 12 physics enhancements, 4 source enhancements, 8 tally enhancements, and 6 other enhancements. Examples and results will be provided for each of these features. The presentation will also discuss the eight enhancements that will be migrated into MCNP6 over the upcoming year.
Atucha-2 PHWR Monte Carlo MCNP5 and KENO-VI models development and application
The geometrical complexity and the peculiarities of Atucha-2 PHWR require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Core models of Atucha-2 PHWR were developed using both MCNP5 and KENOVI codes. The developed models were applied for calculating reactor criticality states at beginning of life, reactor cell constants and control rods volumes. The last two applications were relevant for performing successive three dimensional neutron kinetic analyses since it was necessary to correctly evaluate the effect of each oblique control rod in each cell discretizing the reactor. These corrective factors were then applied to the cell cross sections calculated by the two dimensional deterministic lattice physics code HELIOS. (authors)
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors keff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors keff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors keff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)
GB - a preliminary linking code between MCNP4C and Origen2.1 - DEN/UFMG version
Nowadays it is possible to perform burnup simulation in a detailed 3D geometry and a continuous energy description by the Monte Carlo method. This paper describes an initial project to create and verify a connection code to link Origen2.1 (Oak Ridge National Laboratory) and MCNP4C (Los Alamos National Laboratory). Essentially the code includes point depletion capability to the MCNP code. The incorporation of point depletion capability is explicit and can be summarized by three steps: 1-Monte Carlo determines reaction rates, 2-the reaction rates are used to determine microscopic cross sections for depletion equations, 3-solution of depletion equations (given by Origen2.1) determines number densities for next MCNP step. To evaluate the initial version of the program, we focused on comparing the results with one of the major Monte Carlo burnup codes: MCNPX version 2.6.0. The input files for all codes share the same MCNP geometry, nuclear data library and core thermal power. While simulating 75 time steps at 800 kw of a Heat Pipe Power System model, we have found that the codes generate very similar results. The neutron flux and criticality value of the core agree, especially in the begin of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB (author)
Neutron fluence at the pressure vessel of a pressurized water reactor determined by the MCNP code
Pressure vessel fluence and reaction rates for dosimetry foils in the cavity surrounding the pressure vessel of a pressurized water reactor were determined with a Monte Carlo calculation using the MCNP code. Source neutrons were sampled from a position probability distribution derived from the utility-provided normalized assembly segment power output. The MCNP model was based on one-eighth core symmetry. Source segment spatial biasing, energy cutoff, spatial importance functions, and weight windows were employed as variance reduction techniques. Computed reaction rates were compared with measured ones and in one case to discrete ordinates transport code calculations. Computed reaction rates matched the measured ones within ±10% for 21 of 33 cases and within ±15% for 26 of 33 cases. Neutron flux and fluence >0.1111 and 1 MeV at the pressure vessel location were computed to 17 n/cm2
RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes
The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)
Parallelization of MCNP4 code by using simple FORTRAN algorithms
Simple FORTRAN algorithms, that rely only on open, close, read and write statements, together with disk files and some UNIX commands have been applied to parallelization of MCNP4. The code, named MCNPNFS, maintains almost all capabilities of MCNP4 in solving shielding problems. It is able to perform parallel computing on a set of any UNIX workstations connected by a network, regardless of the heterogeneity in hardware system, provided that all processors produce a binary file in the same format. Further, it is confirmed that MCNPNFS can be executed also on Monte-4 vector-parallel computer. MCNPNFS has been tested intensively by executing 5 photon-neutron benchmark problems, a spent fuel cask problem and 17 sample problems included in the original code package of MCNP4. Three different workstations, connected by a network, have been used to execute MCNPNFS in parallel. By measuring CPU time, the parallel efficiency is determined to be 58% to 99% and 86% in average. On Monte-4, MCNPNFS has been executed using 4 processors concurrently and has achieved the parallel efficiency of 79% in average. (author)
Noelle, P
2006-12-15
In vivo lung counting, one of the preferred methods for monitoring people exposed to the risk of actinide inhalation, is nevertheless limited by the use of physical calibration phantoms which, for technical reasons, can only provide a rough representation of human tissue. A new approach to in vivo measurements has been developed to take advantage of advances in medical imaging and computing; this consists of numerical phantoms based on tomographic images (CT) or magnetic resonance images (R.M.I.) combined with Monte Carlo computing techniques. Under laboratory implementation of this innovative method using specific software called O.E.D.I.P.E., the main thrust of this thesis was to provide answers to the following question: what do numerical phantoms and new techniques like O.E.D.I.P.E. contribute to the improvement in calibration of low-energy in vivo counting systems? After a few developments of the O.E.D.I.P.E. interface, the numerical method was validated for systems composed of four germanium detectors, the most widespread configuration in radio bioassay laboratories (a good match was found, with less than 10% variation). This study represents the first step towards a person-specific numerical calibration of counting systems, which will improve assessment of the activity retained. A second stage focusing on an exhaustive evaluation of uncertainties encountered in in vivo lung counting was possible thanks to the approach offered by the previously-validated O.E.D.I.P.E. software. It was shown that the uncertainties suggested by experiments in a previous study were underestimated, notably morphological differences between the physical phantom and the measured person. Some improvements in the measurement procedure were then proposed, particularly new bio-metric equations specific to French measurement configurations that allow a more sensible choice of the calibration phantom, directly assessing the thickness of the torso plate to be added to the Livermore phantom
VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment
Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)
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.
Generation of Bondarenko F factors with MCNP5 for use in scale transport codes
Hart, S.; Maldonado, G. I. [Univ. of Tennessee, 315 Pasqua Engineering Building, Knoxville, TN 37996 (United States)
2012-07-01
Generally there are three methods of cross-section processing available when using the Scale computer code. They are NITAWL, BONAMI, and CENTRM, with CENTRM being the most common and accurate, but computationally expensive. In order to improve the accuracy of BONAMI (which uses the Bondarenko Method), new Bondarenko/F Factors were to be generated that will smooth out the current F Factors that are being generated using CENTRM. The case discussed here involves using a dedicated Monte Carlo code (MCNP5) to calculate the transport solution from which shielded cross-sections can be produced directly for the transport geometry/mesh. A simple program was created to parse and collect the tallied cross-sections from the MCNP output, which was fed into a modified CLAROL input (a module that replaces or adds data in an AMPX master library), which used the cross-sections obtained from MCNP to calculate new F-factors and update the SCALE library. This approach allows the use of other methods to generate the shielded cross-sections and for easy comparison to existing results. Initial proof-of-principle calculations were carried out for an various cases using various transport solvers, such as NEWT, KENO, and XSDRN, with BONAMI in SCALE. Poor results were obtained using cross-sections generated using infinite homogeneous cases, but good results were obtained by using pin cells in an infinite lattice. (authors)
Generation of Bondarenko F factors with MCNP5 for use in scale transport codes
Generally there are three methods of cross-section processing available when using the Scale computer code. They are NITAWL, BONAMI, and CENTRM, with CENTRM being the most common and accurate, but computationally expensive. In order to improve the accuracy of BONAMI (which uses the Bondarenko Method), new Bondarenko/F Factors were to be generated that will smooth out the current F Factors that are being generated using CENTRM. The case discussed here involves using a dedicated Monte Carlo code (MCNP5) to calculate the transport solution from which shielded cross-sections can be produced directly for the transport geometry/mesh. A simple program was created to parse and collect the tallied cross-sections from the MCNP output, which was fed into a modified CLAROL input (a module that replaces or adds data in an AMPX master library), which used the cross-sections obtained from MCNP to calculate new F-factors and update the SCALE library. This approach allows the use of other methods to generate the shielded cross-sections and for easy comparison to existing results. Initial proof-of-principle calculations were carried out for an various cases using various transport solvers, such as NEWT, KENO, and XSDRN, with BONAMI in SCALE. Poor results were obtained using cross-sections generated using infinite homogeneous cases, but good results were obtained by using pin cells in an infinite lattice. (authors)
Optimum Water Level for Spent Fuel Pool using MCNP Code
TRIGA reactor (RTP) has been operated for more than 30 years. Some of the part of the reactor become degraded by the time. Sooner or later, all these part either will be changed with a new part and proceed with upgrading plan or the reactor itself will be decommissioned. By that time, spent fuel pool (SFP) need to be ready to keep all the fuel from the core. The conceptual design of the SFP has been established. This paper will determine optimum water level to avoid any radiation hazard expose to the workers during managing the fuel later. This determination will use MCNP computer code. (author)
Monte Carlo Simulation of Electron Beams for Radiotherapy - EGS4, MCNP4b and GEANT3 Intercomparison
Trindade, A; Alves, C M; Chaves, A; Lopes, C; Oliveira, C; Peralta, L
2000-01-01
In medical radiation physics, an increasing number of Monte Carlo codes are being used, which requires intercomparison between them to evaluated the accuracy of the simulated results against benchmark experiments. The Monte Carlo code EGS4, commonly used to simulate electron beams from medical linear accelerators, was compared with GEANT3 and MCNP4b. Intercomparison of electron energy spectra, angular and spatial distribution were carried out for the Siemens KD2 linear accelerator, at beam energies of 10 and 15 MeV for a field size of 10x10 cm2. Indirect validation was performed against electron depth doses curves and beam profiles measured in a MP3-PTW water phantom using a Markus planar chamber. Monte Carlo isodose lines were reconstructed and compared to those from commercial treatment planning systems (TPS's) and with experimental data.
Calibration curves of a PGNAA system for cement raw material analysis using the MCNP code
Oliveira, Carlos; Salgado, Jose
1998-12-01
In large samples, the {gamma}-ray count rate of a prompt gamma neutron activation analysis system is a multi-variable function of the elemental dry composition, density, water content and thickness of the material. The experimental calibration curves require tremendous laboratory work, using a great number of standards with well-known compositions. Although a Monte Carlo simulation study does not avoid the experimental calibration work, it reduces the number of experimental calibration standards. This paper is part of a feasibility study for a PGNAA system for on-line continuous characterisation of cement raw material conveyed on a belt (Oliveira, C., Salgado, J. and Carvalho, F. G. (1997) Optimisation of PGNAA instrument design for cement raw materials using the MCNP code. J. Radioanal. Nucl. Chem. 216(2), 191-198; Oliveira, C., Salgado, J., Goncalves, I. F., Carvalho, F. G. and Leitao, F. (1997a) A Monte Carlo study of the influence of geometry arrangements and structural materials on a PGNAA system performance for cement raw materials analysis. Appl. Radiat. Isot. (accepted); Oliveira, C., Salgado, J. and Leitao, F. (1997b) Density and water content corrections in the gamma count rate of a PGNAA system for cement raw material analysis using the MCNP code. Appl. Radiat. Isot. (accepted).]. It reports on the influence of the density, mass water content and thickness on the calibration curves of the PGNAA system. The MCNP-4A code, running in a Pentium-PC and in a DEC workstation, was used to simulate the PGNAA configuration system.
Calibration curves of a PGNAA system for cement raw material analysis using the MCNP code
In large samples, the γ-ray count rate of a prompt gamma neutron activation analysis system is a multi-variable function of the elemental dry composition, density, water content and thickness of the material. The experimental calibration curves require tremendous laboratory work, using a great number of standards with well-known compositions. Although a Monte Carlo simulation study does not avoid the experimental calibration work, it reduces the number of experimental calibration standards. This paper is part of a feasibility study for a PGNAA system for on-line continuous characterisation of cement raw material conveyed on a belt (Oliveira, C., Salgado, J. and Carvalho, F. G. (1997) Optimisation of PGNAA instrument design for cement raw materials using the MCNP code. J. Radioanal. Nucl. Chem. 216(2), 191-198; Oliveira, C., Salgado, J., Goncalves, I. F., Carvalho, F. G. and Leitao, F. (1997a) A Monte Carlo study of the influence of geometry arrangements and structural materials on a PGNAA system performance for cement raw materials analysis. Appl. Radiat. Isot. (accepted); Oliveira, C., Salgado, J. and Leitao, F. (1997b) Density and water content corrections in the gamma count rate of a PGNAA system for cement raw material analysis using the MCNP code. Appl. Radiat. Isot. (accepted).]. It reports on the influence of the density, mass water content and thickness on the calibration curves of the PGNAA system. The MCNP-4A code, running in a Pentium-PC and in a DEC workstation, was used to simulate the PGNAA configuration system
The use of the MCNP code for the quantitative analysis of elements in geological formations
The Monte Carlo modelling calculations using the MCNP code have been performed, which support the spectrometric neutron-gamma (SNGL) borehole logging. The SNGL enables the lithology identification through the quantitative analysis of the elements in geological formations and thus can be very useful for the oil and gas industry as well as for prospecting of the potential host rocks for radioactive waste disposal. In the SNGL experiment, gamma-rays induced by the neutron interactions with the nuclei of the rock elements are detected using the gamma-ray probe of complex mechanical and electronic construction. The probe has to be calibrated for a wide range of the elemental concentrations, to assure the proper quantitative analysis. The Polish Calibration Station in Zielona Gora is equipped with a limited number of calibration standards. An extension of the experimental calibration and the evaluation of the effect of the so-called side effects (for example the borehole and formation salinity variation) on the accuracy of the SNGL method can be done by the use of the MCNP code. The preliminary MCNP results showing the effect of the borehole and formation fluids salinity variations on the accuracy of silicon (Si), calcium (Ca) and iron (Fe) content determination are presented in the paper. The main effort has been focused on a modelling of the complex SNGL probe situated in a fluid filled borehole, surrounded by a geological formation. Track length estimate of the photon flux from the (n,gamma) interactions as a function of gamma-rays energy was used. Calculations were run on the PC computer with AMD Athlon 1.33 GHz processor. Neutron and photon cross-sections libraries were taken from the MCNP4c package and based mainly on the ENDF/B-6, ENDF/B-5 and MCPLIB02 data. The results of simulated experiment are in conformity with results of the real experiment performed with the use of the main lithology models (sandstones, limestones and dolomite). (authors)
Vectorization techniques for neutron transport Monte Carlo codes
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)
Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code
C. A. M. Silva
2014-01-01
Full Text Available In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport code. The sensitivity analyses included small differences of the core and the rods dimensions and different levels of model detailing. Four models were simulated and neutronic parameters such as effective multiplication factor (keff, reactivity (ρ, and thermal and total neutron flux in central thimble in some different conditions of the reactor operation were analysed. The simulated models presented good agreement between them, as well as in comparison with available experimental data. In this way, the sensitivity analyses demonstrated that simulations of the TRIGA IPR-R1 reactor can be performed using any one of the four investigated MCNP models to obtain the referenced neutronic parameters.
Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; S Meigooni, Ali
2016-01-01
Monte Carlo simulations are widely used for calculation of the dosimetric param-eters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulat-ing different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes. PMID:27074460
MCNP: Photon benchmark problems
The recent widespread, markedly increased use of radiation transport codes has produced greater user and institutional demand for assurance that such codes give correct results. Responding to these pressing requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on six different photon problem families. MCNP was used to simulate these six sets numerically. Results for each were compared to the set's analytical or experimental data. MCNP successfully predicted the analytical or experimental results of all six families within the statistical uncertainty inherent in the Monte Carlo method. From this we conclude that MCNP can accurately model a broad spectrum of photon transport problems. 8 refs., 30 figs., 5 tabs
Method of tallying adjoint fluence and calculating kinetics parameters in Monte Carlo codes
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)
Study of photon attenuation coefficient in brine using MCNP code
Barbosa, Caroline M.; Salgado, Cesar M.; Brandao, Luis E.B., E-mail: carolmattosb@yahoo.com.br, E-mail: otero@ien.gov.br, E-mail: brandao@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2015-07-01
In petroleum industry, multiphase flows are common and the relative salt content of the water component depends on the location of oil extraction. The salt present in the water component causes incrustations in the pipeline and may interfere in the flow measurement. This paper presents an elaborate model using MCNP code to simulate a narrow beam gamma ray source, a brine sample and a NaI(Tl) detector, with beam energies ranging from 59,54 keV to 662 keV. Through this model, we can relate the photon attenuation coefficient to the salinity of water. This model can be experimentally reproduced, and used to measure the salinity in situ without affecting the medium. (author)
MCOR - Monte Carlo depletion code for reference LWR calculations
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
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
Highlights: ► We used Monte Carlo method and MCNP-4C code for simulation. ► We simulated linear attenuation coefficients, Mfp and transmission factor for three barites and lead. ► We compared MCNP results and experimental data for various concretes in different energies. ► MCNP results showed a good agreement with experimental data. - Abstract: In this study shielding properties of various barites concretes and lead in three high gamma energies 0.662, 1.173, and 1.332 MeV were investigated using the MCNP-4C code and compared with predictions from the XCOM code and experimental data. In the three selected energies, the simulated and available data values were compared and results showed a good agreement. The results of the three methods show that lead, and pure (100%) barite have higher linear attenuation coefficients and lower transmission factors and mean free path values relative to 50% barite and 0% barite concretes
Criticality benchmarking of ANET Monte Carlo code
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)
MCMG: a 3-D multigroup P3 Monte Carlo code and its benchmarks
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)
This study made a comparison between some of the major transport codes that employ the Monte Carlo stochastic approach in dosimetric calculations in nuclear medicine. We analyzed in detail the various physical and numerical models used by MCNP5 code in relation with codes like EGS and Penelope. The identification of its potential and limitations for solving microdosimetry problems were highlighted. The condensed history methodology used by MCNP resulted in lower values for energy deposition calculation. This showed a known feature of the condensed stories: its underestimates both the number of collisions along the trajectory of the electron and the number of secondary particles created. The use of transport codes like MCNP and Penelope for micrometer scales received special attention in this work. Class I and class II codes were studied and their main resources were exploited in order to transport electrons, which have particular importance in dosimetry. It is expected that the evaluation of available methodologies mentioned here contribute to a better understanding of the behavior of these codes, especially for this class of problems, common in microdosimetry. (author)
In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)
Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis
M. Pecchia
2011-01-01
Full Text Available The geometrical complexity and the peculiarities of Atucha-2 PHWR require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Core models of Atucha-2 PHWR were developed using both MCNP5 and KENO-VI codes. The developed models were applied for calculating reactor criticality states at beginning of life, reactor cell constants, and control rods volumes. The last two applications were relevant for performing successive three dimensional neutron kinetic analyses since it was necessary to correctly evaluate the effect of each oblique control rod in each cell discretizing the reactor. These corrective factors were then applied to the cell cross sections calculated by the two-dimensional deterministic lattice physics code HELIOS. These results were implemented in the RELAP-3D model to perform safety analyses for the licensing process.
Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis
The geometrical complexity and the peculiarities of Atucha-2 PHWR require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Core models of Atucha-2 PHWR were developed using both MCNP5 and KENO-VI codes. The developed models were applied for calculating reactor criticality states at beginning of life, reactor cell constants, and control rods volumes. The last two applications were relevant for performing successive three dimensional neutron kinetic analyses since it was necessary to correctly evaluate the effect of each oblique control rod in each cell discretizing the reactor. These corrective factors were then applied to the cell cross sections calculated by the two-dimensional deterministic lattice physics code Helios. These results were implemented in the RELAP-3D model to perform safety analyses for the licensing process.
Research on parallel computing of MCNP code based on MPI
This paper introduces the method that develops a parallel computing platform with ordinary PCs and the mpi.nt.1.2.5 software based on MPI (Message Interface Passing) standard specification on Windows operating system. The parallel computing of MCNP on this platform is realized, and the parallel computing performance of MCNP is analyzed in this paper. (authors)
Image enhancement using MCNP5 code and MATLAB in neutron radiography.
Tharwat, Montaser; Mohamed, Nader; Mongy, T
2014-07-01
This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. PMID:24583508
Calculation of age-dependent effective doses for external exposure using the MCNP code
Hung, Tran Van [Research and Development Center for Radiation Technology, ThuDuc, HoChiMinh City (VT)
2013-07-15
Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)
Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm
Burnup calculations of TR-2 Research Reactor with Monteburns Monte Carlo Code
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)
Full-core pin-power calculations using Monte Carlo codes
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)
Simulations of X-ray spectrum and HVL for mammographic equipment using MCNP5 code
Souza, Rafael Toledo F. de; Alvarez, Matheus; Velo, Alexandre F.; Oliveira, Marcela de; Miranda, Jose Ricardo A. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Inst. de Biociencias de Botucatu. Dept. de Fisica e Biofisica; Pina, Diana R. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Dept. de Doencas Tropicais e Diagnostico por Imagem
2012-07-01
Full text: The main goal of mammography is early detection of breast cancer. Thus, the mammograph should be designed so that the X-ray photons are emitted within an appropriate energy range, to distinguish the normal breast tissue and cancerous tissue. The distribution of the photons amount of X-ray beam, with their respective energies, is called the spectrum. From the spectrum it is possible to estimate the quality of the X-ray beam from the Half Value Layer (HVL). Objectives: This study aims to simulate the Senographe 600T mammography unit, manufactured by General Electric (GE), using the MCNP5 Monte Carlo code, to obtain its spectrum and HVL, and compare the HVL of the simulated model with experimental data. Method: the mammography unit was simulated using a simplified model which a beam of 2x10{sup 8} electrons focuses on a Mo target angled 12 degrees, within a capsule filled with vacuum. The incident electrons were converted into photons. The capsule has a beryllium window, allowing the passage of the X-ray beam. The beam is detected by an air cylinder with 1 cm thickness placed 60 cm from the target. On the path of X-ray beam, is inserted a 0.03 mm Mo filter located 1.6 cm after the beryllium window. The space between the capsule and the detector cylinder was filled with air. The quality of X-ray beam was verified from the HVL using the MCNP5 code and the experimental method for the voltage range typically used in clinical routine (26-31 kVp). Results and discussion: the X-ray spectrum of the mammography device is satisfactorily simulated by MCNP5, showing the characteristic radiation peaks of molybdenum at 17.479 keV and 19.602 keV, the filtered spectrum generated by Bremsstrahlung, and reducing the total number of photons with the decrease in applied tension (kVp). The HVL obtained by MCNP5 and experimental measurements show a maximum difference of 5.31% (for 31 kVp). The result of both methods are within acceptable limits established by national
Verification of the shift Monte Carlo code with the C5G7 reactor benchmark
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)
MCNP is a widely used and actively developed Monte Carlo radiation transport code. Many important features have recently been added and more are under development. Benchmark studies not only indicate that MCNP is accurate but also that modern computer codes can give answers basically as accurate as the physics data that goes in them. Even deep penetration problems can be correct to within a factor of two after 10 to 25 mean free paths of penetration. And finally, Monte Carlo calculations, once thought to be too expensive to run routinely, can now be run effectively on desktop computers which compete with the supercomputers of yesteryear. 21 refs., 3 tabs
Monte Carlo solver for UWB1 nuclear fuel depletion code
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
The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from the 252Cf-source-driven frequency analysis measurements. This code can be used to validate calculational methods and cross section data sets from subcritical experiments. This code provides a more general model for interpretation and planning of experiments for nuclear criticality safety, nuclear safeguards, and nuclear weapons identification and replaces the use of point kinetics models for interpreting the measurements. The use of MCNP-DSP extends the usefulness of this measurement method to systems with much lower neutron multiplication factors
An accurate knowledge of the output energy spectra of an x-ray tube is essential in many areas of radiological studies. It forms the basis of almost all image quality simulations and enable system designers to predict patient dose more accurately. Many radiological physics problems that can be solved by Monte Carlo simulation methods require an x-ray spectra as input data. Computer simulation of x-ray spectra is one of the most important tools for investigation of patient dose and image quality in diagnostic radiology systems. In this work the general purpose Monte Carlo N-particle radiation transport computer code (MCNP-4C) was used for the simulation of x-ray spectra in diagnostic radiology, Electron's path in the target was followed until it's energy was reduced to 10 keV. A user friendly interface named 'Diagnostic X-ray Spectra by Monte Carlo Simulation (DXRaySMCS)' was developed to facilitate the application of MCNP-4C code for diagnostic radiology spectrum prediction. The program provides a user friendly interface for modifying the MCNP input file, launching the MCNP program to simulate electron and photon transport and processing the MCNP output file to yield a summary of the results (Relative Photon Number per Energy Bin). In this article the development and characteristics of DXRaySMCS are outlined. As part of the validation process, out put spectra for 46 diagnostic radiology system settings produced by DXRaySMCS were compared with the corresponding IPEM78. Generally, there is a good agreement between the two sets of spectra. No statistically significant differences have been observed between IPEM78 reported spectra and the simulated spectra generated in this study. (author)
A voxel-based mouse for internal dose calculations using Monte Carlo simulations (MCNP)
Murine models are useful for targeted radiotherapy pre-clinical experiments. These models can help to assess the potential interest of new radiopharmaceuticals. In this study, we developed a voxel-based mouse for dosimetric estimates. A female nude mouse (30 g) was frozen and cut into slices. High-resolution digital photographs were taken directly on the frozen block after each section. Images were segmented manually. Monoenergetic photon or electron sources were simulated using the MCNP4c2 Monte Carlo code for each source organ, in order to give tables of S-factors (in Gy Bq sup - sup 1 s sup - sup 1) for all target organs. Results obtained from monoenergetic particles were then used to generate S-factors for several radionuclides of potential interest in targeted radiotherapy. Thirteen source and 25 target regions were considered in this study. For each source region, 16 photon and 16 electron energies were simulated. Absorbed fractions, specific absorbed fractions and S-factors were calculated for 16 radionuclides of interest for targeted radiotherapy. The results obtained generally agree well with data published previously. For electron energies ranging from 0.1 to 2.5 MeV, the self-absorbed fraction varies from 0.98 to 0.376 for the liver, and from 0.89 to 0.04 for the thyroid. Electrons cannot be considered as 'non-penetrating' radiation for energies above 0.5 MeV for mouse organs. This observation can be generalized to radionuclides: for example, the beta self-absorbed fraction for the thyroid was 0.616 for I-131; absorbed fractions for Y-90 for left kidney-to-left kidney and for left kidney-to-spleen were 0.486 and 0.058, respectively. Our voxel-based mouse allowed us to generate a dosimetric database for use in preclinical targeted radiotherapy experiments. (author)
Shielding analysis of the microtron MT-25 bunker using the MCNP-4C code and NCRP report 51
A cyclic electron accelerator Microtron MT-25 will be installed in Havana (Cuba)). Electrons, neutrons and gamma radiation up to 25 MeV can be produced in the MT-25. A detailed shielding analysis for the bunker is carried out using two ways: the NCRP-51 Report and the Monte Carlo Method (MCNP-4C Code). The walls and ceiling thicknesses are estimated with dose constraints of 0.5 and 20 mSv y-1, respectively, and an area occupancy factor of 1/16. Both results are compared and a preliminary bunker design is shown. (authors)
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.)
Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems
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)
Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation
Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)
2012-07-01
Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)
Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation
Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3DC/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)
Valentine, T.E.
2001-01-19
The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from subcritical measurements. The code can be used to simulate a variety of subcritical measurements including source-driven noise analysis, Rossi-{alpha}, pulsed source, passive frequency analysis, multiplicity, and Feynman variance measurements. This code can be used to validate Monte Carlo methods and cross section data sets with subcritical measurements and replaces the use of point kinetics models for interpreting subcritical measurements.
TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation
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.
Validation the Monte Carlo code RMC with C5G7 benchmark
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
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.
Monte Carlo photon benchmark problems
Photon benchmark calculations have been performed to validate the MCNP Monte Carlo computer code. These are compared to both the COG Monte Carlo computer code and either experimental or analytic results. The calculated solutions indicate that the Monte Carlo method, and MCNP and COG in particular, can accurately model a wide range of physical problems. 8 refs., 5 figs
Simulation of absorbed dose in human blood with MCNP 4C code
Biological dosimetry, based on the analysis of solid stained dicentric chromosomes, has been used since the mid 1960s. The intervening years have seen great improvements bringing the technique to a point where dicentric analysis has become a routine component of the radiological protection programs of many countries. Experience of its application in thousands of cases of actual or suspected overexposures has proved the worth of the method. The aberrations scored in the lymphocytes are interpreted in terms of absorbed dose by reference to a dose response calibration curve. This curve will have been produced by exposure of blood in vitro to doses of the appropriate quality of radiation. The doses given to the specimens should be traceable via a physical instrument such as an ionization chamber, to a primary or secondary standard. An alternative to obtain the information about absorbed dose in a specific blood volume is through the Monte Carlo method. The use of such technique is worldwide when physical measurements are inconvenient or impossible, and particularly useful for the solution of complex problems that cannot be modeled by codes that use deterministic methods. It is applied to particle systems as neutrons and electrons, as well as photons or still in mixed systems. Due to difficulties that involve the use of neutrons, this technique has shown extreme importance for preliminary research and experimental arrangements with neutron sources. In this study, the main objective was to simulate the dose absorbed by a blood sample in an experimental arrangement through the irradiation with sources of 241AmBe. It was used the code Monte Carlo N-Particle version 4C (MCNP 4C) whose data had been processed parallel in a computational structure in a cluster. This method allowed estimating the absorbed dose in a specific blood volume, making possible the experimental setup arrangement. (author)
The New MCNP6 Depletion Capability
Fensin, Michael Lorne [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory
2012-06-19
The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.
This study deals with the neutronic analysis of the 2MW TRIGA MARK II Moroccan research reactor. The reactor was commissioned at Centre des Etudes Nucleaires de la Maamora (CENM) and it went critical on May 2, 2007. The 3-D continuous energy Monte Carlo code MCNP5 was used to develop a full model of the TRIGA reactor, using the maximum details allowed by the constructor General Atomics of USA. Continuous energy cross section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross section libraries were generated by using the NJOY99 system updated to its more recent patch file 'up259'. The consistency and accuracy of both Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. (author)
McKinney, Gregg W [Los Alamos National Laboratory; Armstrong, Hirotatsu [Los Alamos National Laboratory; James, Michael R [Los Alamos National Laboratory; Clem, John [University of Delaware, BRI; Goldhagen, Paul [DHS, National Urban Security Technology Laboratory
2012-06-19
MCNP is a Monte Carlo radiation transport code that has been under development for over half a century. Over the last decade, the development team of a high-energy offshoot of MCNP, called MCNPX, has implemented several physics and algorithm improvements important for modeling galactic cosmic-ray (GCR) interactions with matter. In this presentation, we discuss the latest of these improvements, a new Cosmic-Source option, that has been implemented in MCNP6.
In vivo lung counting, one of the preferred methods for monitoring people exposed to the risk of actinide inhalation, is nevertheless limited by the use of physical calibration phantoms which, for technical reasons, can only provide a rough representation of human tissue. A new approach to in vivo measurements has been developed to take advantage of advances in medical imaging and computing; this consists of numerical phantoms based on tomographic images (CT) or magnetic resonance images (R.M.I.) combined with Monte Carlo computing techniques. Under laboratory implementation of this innovative method using specific software called O.E.D.I.P.E., the main thrust of this thesis was to provide answers to the following question: what do numerical phantoms and new techniques like O.E.D.I.P.E. contribute to the improvement in calibration of low-energy in vivo counting systems? After a few developments of the O.E.D.I.P.E. interface, the numerical method was validated for systems composed of four germanium detectors, the most widespread configuration in radio bioassay laboratories (a good match was found, with less than 10% variation). This study represents the first step towards a person-specific numerical calibration of counting systems, which will improve assessment of the activity retained. A second stage focusing on an exhaustive evaluation of uncertainties encountered in in vivo lung counting was possible thanks to the approach offered by the previously-validated O.E.D.I.P.E. software. It was shown that the uncertainties suggested by experiments in a previous study were underestimated, notably morphological differences between the physical phantom and the measured person. Some improvements in the measurement procedure were then proposed, particularly new bio-metric equations specific to French measurement configurations that allow a more sensible choice of the calibration phantom, directly assessing the thickness of the torso plate to be added to the Livermore phantom
The MCNP perturbation feature assessment for nuclear logging Monte Carlo calculations
In the oil industry a log is a record of the measurements taken by the instrument, called a sonde, in an attempt to determine a geology through which a drill is passing. Nuclear sondes probe the formation with very energetic neutrons and gamma rays. The sonde is typically about half the diameter of the borehole and contains a source, detectors and shielding. In this work the sensitivity of detector response to the borehole fluid, detector helium gas and formation matrix density changes has been tested with well logging benchmark, containing a neutron source and two 3He neutron detectors. The MCNP4C code standard perturbation feature, using the first- and second-term Taylor series expansion of perturbed quantity, has been applied for differential sampling. The volume averaged neutron flux and (n,p) reaction rates were the tallying quantities. Calculation results show that the MCNP perturbation feature can be fairly successfully implemented as initial step of well logging sensitivity studies. However, one should always keep in mind the limitations resulting from a two-term Taylor series expansion, currently used in MCNP. (author)
Use of MCNP perturbation feature for nuclear well logging Monte Carlo sensitivity calculations
In the oil industry a log is a record of the measurements taken by the instrument, called a sonde, in an attempt to determine a geology through which a drill is passing. Nuclear sondes probe the formation with very energetic neutrons and gamma rays. The sonde is typically about half the diameter of the borehole and contains a source, detectors and shielding. In this work the sensitivity of detector response to the borehole fluid, detector helium gas and formation matrix density changes has been tested with well logging benchmark, containing a neutron source and two 3He neutron detectors. The MCNP4C code standard perturbation feature, using the first- and second-term Taylor series expansion of perturbed quantity, has been applied for differential sampling. The volume averaged neutron flux and (n,p) reaction rates were the tallying quantities. Calculation results show that the MCNP perturbation feature can be fairly successfully implemented as initial step of well logging sensitivity studies. However, one should always keep in mind the limitations resulting from a two-term Taylor series expansion, currently used in MCNP. (author)
Calibration of a foot borne spectrometry system using the MCNP 4C code
The increased interest for the cycling of radioactive Caesium in natural ecosystems has gained need for rapid and reliable methods to investigate the deposition density in natural soils. One commonly used method, soil sampling, is a good method that correctly used gives information of both the horizontal and vertical distribution of the desired nuclide. The main disadvantage is that the method is time consuming regarding sampling, preparation and measurements. An alternative method is the use of semiconductors or scintillation detectors in the field i.e. in cars, airplanes, or helicopters. Theses methods are rapid and integrate over large areas which gives a more reliable mean value provided that the operator has some basic knowledge about the depth distribution of the radio nuclides and bulk density in the soil. To be effective the systems are often connected to a GPS to give the exact coordinate for each measurement. In a situation where the area of interest is too large to cover by soil samples and measurements by airplane not will give a spatial resolution good enough, one feasible method is to use a foot borne gamma spectrometry system. The advantage of a foot borne system is that the operator can cover a quite large area within a few hours and that the method can detect small anomalies in the deposition field which may be difficult to discover with soil samples. This abstract describes the calibration of a foot borne gamma-spectrometry system carried in a back-pack and consisting of a NaI-detector, a GPS and a system for logging activity and position. The detector system and surroundings has been modeled in the Monte Carlo code MCNP 4C (Figure 1). The Monte Carlo method gives the possibility to study the influence of complex geometries that are difficult to create for a practical calibration using real activity. The results of the MCNP calibration model, has been compared to foot borne gamma-spectrometry field measurements in a Cs-137 deposition area. A
New developments enhancing MCNP for criticality safety
Since the early 80's MCNP has had three estimates of keff: collision, absorption, and track length. MCNP has also had collision and absorption estimators of removal lifetime. These are calculated for every cycle and are averaged over the cycles as simple averages and covariance weighted averages. Correlation coefficients between estimators are also calculated. These criticality estimators are all in addition to the extensive summary information and tally edits used in shielding and other problems. A number of significant new developments have been made to enhance the MCNP Monte Carlo radiation transport code for criticality safety applications. These are available in the newly released MCNP4A version of the code
Simulation of the BNCT of Brain Tumors Using MCNP Code: Beam Designing and Dose Evaluation
BNCT is an effective method to destroy brain tumoral cells while sparing the healthy tissues. The recommended flux for epithermal neutrons is 109 n/cm2s, which has the most effectiveness on deep-seated tumors. In this paper, it is indicated that using D-T neutron source and optimizing of Beam Shaping Assembly leads to treating brain tumors in a reasonable time where all International Atomic Energy Agency recommended criteria are met. The proposed Beam Shaping Assembly based on a D-T neutron generator consists of a neutron multiplier system, moderators, reflector, and collimator. The simulated Snyder head phantom is used to evaluate dose profiles in tissues due to the irradiation of designed beam. Monte Carlo Code, MCNP-4C, was used in order to perform these calculations. The neutron beam associated with the designed and optimized Beam Shaping Assembly has an adequate epithermal flux at the beam port and neutron and gamma contaminations are removed as much as possible. Moreover, it was showed that increasing J/Φ, as a measure of beam directionality, leads to improvement of beam performance and survival of healthy tissues surrounding the tumor. According to the simulation results, the proposed system based on D-T neutron source, which is suitable for in-hospital installations, satisfies all in-air parameters. Moreover, depth-dose curves investigate proper performance of designed beam in tissues. The results are comparable with the performances of other facilities.
Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code
Ferreira, Vanessa M.; Oliveira, Renato C.M., E-mail: vanessamachado@ufmg.br [Curso Superior de Tecnologia em Radiologia. Faculdade de Medicina da Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil); Barros, Graiciany P.; Oliveira, Arno H.; Veloso, M. Auxiliadora F. [Departamento de Engenharia Nuclear. Escola de Engenharia. Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil)
2011-07-01
One of basic principles for radiological protection is the optimization of techniques for obtain radiographic images, in way that the dose in the patient is kept as low as reasonably achievable (ALARA). Intensifying screens are used in medical radiology, which reduce considerably the dose rates in the production of radiographic images, maintaining the quality of these, while in dental radiology, there is no a intensifying screen available for intraoral examinations. From this technological requirement, this paper evaluates a computational modeling of an intensifying screen for use in intraoral radiography. For this, it was used the Monte Carlo code MCNP5 that allows the radiography simulation through the transport of electrons and photons in the different materials present in this examination. The goal of an intensifying screen is the conversion of X-ray photons to photons in the visible spectrum, knowing that radiographic films are more sensitive to light photons than to X-ray photons. So the screen should be composed of an efficient material for converting x-rays photons in light photons, therefore was made simulations using different materials, thicknesses and positions possible for placing screen in radiographic film in order to find the way more technically feasible. (author)
Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code
One of basic principles for radiological protection is the optimization of techniques for obtain radiographic images, in way that the dose in the patient is kept as low as reasonably achievable (ALARA). Intensifying screens are used in medical radiology, which reduce considerably the dose rates in the production of radiographic images, maintaining the quality of these, while in dental radiology, there is no a intensifying screen available for intraoral examinations. From this technological requirement, this paper evaluates a computational modeling of an intensifying screen for use in intraoral radiography. For this, it was used the Monte Carlo code MCNP5 that allows the radiography simulation through the transport of electrons and photons in the different materials present in this examination. The goal of an intensifying screen is the conversion of X-ray photons to photons in the visible spectrum, knowing that radiographic films are more sensitive to light photons than to X-ray photons. So the screen should be composed of an efficient material for converting x-rays photons in light photons, therefore was made simulations using different materials, thicknesses and positions possible for placing screen in radiographic film in order to find the way more technically feasible. (author)
Image enhancement using MCNP5 code and MATLAB in neutron radiography
This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. - Highlights: • This work is applicable for static based film neutron radiography and digital neutron imaging. • MATLAB is a useful tool for imaging enhancement in radiographic film. • Advanced imaging processing is available in the ETRR-2 for imaging processing and data extraction. • The digital imaging system is suitable for complex shapes and sizes, while MATLAB technique is suitable for simple shapes and sizes. • Quantitative measurements are available
Progress on burnup calculation methods coupling Monte Carlo and depletion codes
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)
First Results of Saturation Curve Measurements of Heat-Resistant Steel Using GEANT4 and MCNP5 Codes
Hoang, Duc-Tam; Tran, Thien-Thanh; Le, Bao-Tran; Tran, Kim-Tuyet; Huynh, Dinh-Chuong; Vo, Hoang-Nguyen; Chau, Van-Tao
A gamma backscattering technique is applied to calculate the saturation curve and the effective mass attenuation coefficient of material. A NaI(Tl) detector collimated by collimator of large diameter is modeled by Monte Carlo technique using both MCNP5 and GEANT4 codes. The result shows a good agreement in response function of the scattering spectra for the two codes. Based on such spectra, the saturation curve of heat-resistant steel is determined. The results represent a strong confirmation that it is appropriate to use the detector collimator of large diameter to obtain the scattering spectra and this work is also the basis of experimental set-up for determining the thickness of material.
Simulation of density curve for slim borehole using the Monte Carlo code MCNPX
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
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)
MCNP applications for the 21st century
The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions. (author)
MCNP application for the 21 century
The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions
This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using 'TORT-coupled MCNP' method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency
Development of a New Monte Carlo reactor physics code
have no continuous-energy counterparts in the Monte Carlo calculation. This study is focused on the development of an entirely new Monte Carlo neutron transport code, specifically intended for reactor physics calculations at the fuel assembly level. The PSG code is developed at VTT Technical Research Centre of Finland and one of the main applications is the generation of homogenised group constants for deterministic reactor simulator codes. The theoretical background on general transport theory, nodal diffusion calculation and the Monte Carlo method are discussed. The basic methodology used in the PSG code is introduced and previous studies related to the topic are briefly reviewed. PSG is validated by comparison to reference results produced by MCNP4C and CASMO-4E in infinite two-dimensional LWR lattice calculations. Group constants generated by PSG are used in ARES reactor simulator calculations and the results compared to reference calculations using CASMO-4E data.
MCNPX{trademark} -- The LAHET{trademark}/MCNP{trademark} code merger
Hughes, H.G.; Adams, K.J.; Chadwick, M.B. [and others
1997-08-01
The MCNP code is written and maintained by Group X-TM at Los Alamos National Laboratory. In response to the demands of the accelerator community, the authors have undertaken a major effort to expand the capabilities of MCNP to increase the set of transportable particles; to make use of newly evaluated high-energy nuclear data tables for neutrons, protons, and potentially other particles; and to incorporate physics models for use where tabular data are unavailable. A preliminary version of the expanded code, called MCNPX, has now been issued for testing. The new code includes all existing LAHET physics modules, and has the ability to utilize the 150-MeV data libraries that have recently been released by LANL Group T-2.
The dose distribution calculation is one of major steps in cancer radiotherapy. This paper applies Monte Carlo code, MCNP5, in simulation 15 MV photon beams from linear accelerator of General Hospital of Kien Giang in a case treatment of lung cancer. The settings for beam direction, field size and isocenter position used in MCNP5 must be the same as in treatment plan at hospital to ensure the results from MCNP5 are accurate. We also built a program CODIM by using MATLAB® programming software. This program is used to construct digital voxel phantoms from lung CT images obtained from cancer treatment cases at Kien Giang hospital and then simulate the delivered dose of linac in these phantoms by using MCNP5 simulation code. The results show that there is a difference of 5% in comparison to Prowess Panther program - a semi-empirical simulation program which is being used for treatment planning in Kien Giang hospital. (author)
MCNP and Monte Carlo method was used to calculate dose rate in the air-space of irradiation room at Hanoi Irradiation Center. Experiment measurements were also carried out to investigate the real distribution of dose field in air of the irradiator as well as the distribution of absorbed dose in sample product containers. The results show that there is a deviation between calculated data given by MCNP and measurements. The data of MCNP give a symmetric distribution of dose field against the axes going through the center of the source rack meanwhile the experiment data show that dose rate get higher values in the lower part of the space. Going to lower position to the floor dose rate getting higher value. This phenomenon was also occurred for the measurements of absorbed dose in sample product container. (author)
El-Khayatt, A. M.; Ali, A. M.; Singh, Vishwanath P.
2014-01-01
The mass attenuation coefficients, μ/ρ, total interaction cross-section, σt, and mean free path (MFP) of some Heavy Metal Oxides (HMO) glasses, with potential applications as gamma ray shielding materials, have been investigated using the MCNP-4C code. Appreciable variations are noted for all parameters by changing the photon energy and the chemical composition of HMO glasses. The numerical simulations parameters are compared with experimental data wherever possible. Comparisons are also made with predictions from the XCOM program in the energy region from 1 keV to 100 MeV. Good agreement noticed indicates that the chosen Monte Carlo method may be employed to make additional calculations on the photon attenuation characteristics of different glass systems, a capability particularly useful in cases where no analogous experimental data exist.
Ivanov, A., E-mail: aleksandar.ivanov@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Herman-vom-Helmholtz-Platz-1, 76344 Eggenstein-Leopoldshafen (Germany); Sanchez, V., E-mail: victor.sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Herman-vom-Helmholtz-Platz-1, 76344 Eggenstein-Leopoldshafen (Germany); Stieglitz, R., E-mail: robert.stieglitz@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Herman-vom-Helmholtz-Platz-1, 76344 Eggenstein-Leopoldshafen (Germany); Ivanov, K., E-mail: kni1@psu.edu [The Pennsylvania State University, Department of Mechanical and Nuclear Engineering 206 Reber, University Park, PA 16802 (United States)
2013-09-15
Highlights: • Coupled scheme between MCNP and in-house developed sub-channel code. • We discuss the temperature dependence of nuclear data. • We test the validity of the pseudo material mixing for thermal scattering data. • The correlation between statistical uncertainty and convergence is investigated. • The coupled scheme was applied to PWR, BWR and High Conversion PWR problems. -- Abstract: In order to increase the accuracy and the degree of spatial and energy resolution of core design studies, coupled 3D neutronic (multi-group deterministic and continuous energy Monte-Carlo) and 3D thermal-hydraulic (CFD and sub-channel) codes are being developed worldwide. At KIT, both deterministic and Monte-Carlo codes were coupled with sub-channel codes and applied to predict the safety-related design parameters such as critical power ratio, maximal cladding, fuel temperature and DNB. These coupling approaches were revised and improved based on the experience gained. One particular example is replacing COBRA-TF with SUBCHANFLOW, an in-house developed sub-channel code, in the COBRA-TF/MCNP coupling, accompanied with new way of radial mapping between the neutronic and thermal-hydraulic domains. The new coupled system MCNP5/SUBCHANFLOW makes it possible to investigate a variety of fuel assembly types. Key issues in such a coupled system are the implementation of the thermal-hydraulic/neutronic feedback mechanisms, the precision of the Monte-Carlo solutions, and the supervision of convergence during the iterative solution process. Another key issue considered is the optimal application of parallel computing. Using multi-processor computer architectures, it is possible to reduce the Monte-Carlo running time and obtain converged results within reasonable time limit. In particular, it is shown that by exploiting the capabilities of multi-processor calculation, large fuel assemblies on a pin-by-pin basis with a resolution at sub-channel level can be analyzed. One of the
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.
Depleted Reactor Analysis With MCNP-4B
Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool
The filter effects in X-ray tube XRF simulated by using the MCNP5 code
In X-ray fluorescence (XRF) analysis, the use of filter reduces the background interference effectively, and improves the analysis sensitivity. In this paper, the MCNP5 code is used to simulate the filter effects in X-ray tube XRF. The XRF spectra by 140 keV electron beams, with filters of different materials and thicknesses, are compared. Their attenuation rates, i.e. background reduction from the original XRF spectra, are analyzed. (authors)
Determination of dosimetric parameters for 125I seed source using MCNP5 and EGSnrc MC codes
Background: Seed source has become a popular treatment option in the management of various tumors, particularly in the prostate. Purpose: The aim is to develop accurate and reliable dosimetric parameters that could be used to measure the dose delivered to organs at risk. Methods: Dosimetric parameters (dose rate constant, radial dose function and anisotropy function) of model 6711 125I seed source were calculated with MCNP5 and EGSnrc MC codes following AAPM TG43U1 recommendations. Results: The two results were compared with the relative data recommend by AAPM TG43U1, and the data were as follows: dose rate constant with MCNP5 was in agreement with 0.62%, while that with EGSnrc was 2.07%; radial dose function with MCNP5 was within 0.15%-5.12%, while that with EGSnrc was within 0%-2.48%. Conclusion: The results of two MC codes are in accordance with the recommendations. But that with EGSnrc MC code is better. (authors)
Cadenas Mendicoa, A. M.
2016-08-01
Due to the lack of graphical representation capability of same nuclear codes like MCNP of GOTHIC, widely used in the industry, the following article describes the development of an interface to use a graphical representation open source (Paraview) with the outputs generated by the nuclear codes. Moreover, this article aims at describing the advantage of this type of visualization programs for the modeling and decision making in the calculation. (Author)
SPQR: a Monte Carlo reactor kinetics code
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
Application of ENDF nuclear data for testing a Monte-Carlo neutron and photon transport code
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)
Calculation of the Fast Flux Test Facility fuel pin tests with the WIMS-E and MCNP codes
The Fuel Assembly Area (FAA) at the Fast Flux Test Facility site on the Hanford Site at Richland, Washington currently is being prepared to fabricate mixed oxide fuel (U, Pu) for the FFTF. Calculational tools are required to perform criticality safety analyses for various process locations and to establish safe limits for fissile material handling at the FAA. These codes require validation against experimental data appropriate for the compositions that will be handled. Critical array experiments performed by Bierman provide such data for mixed oxide fuel in the range Pu/(U+Pu) = 22 wt %, and with Pu-240 contents equal to 12 wt %. Both the Monte Carlo Neutron Photon (MCNP) and the Winfrith Improved Multigroup Scheme (WIMS-E) computer codes were used to calculate the neutron multiplication factor for explicit models of the various critical arrays. The W-CACTUS modules within the WIMS-E code system was used to calculate k∞ for the explicit array configuration, as well as few-group cross sections that were then used in a three-dimensional diffusion theory code for the calculation of keff for the finite array. 10 refs., 15 figs., 7 tabs
Coded aperture optimization using Monte Carlo simulations
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.
Successful vectorization - reactor physics Monte Carlo code
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.)
Fernandes, Marco A.R., E-mail: marco@cetea.com.b, E-mail: marfernandes@fmb.unesp.b [Universidade Estadual Paulista Julio de Mesquita Filho (FMB/UNESP), Botucatu, SP (Brazil). Fac. de Medicina; Ribeiro, Victor A.B. [Universidade Estadual Paulista Julio de Mesquita Filho (IBB/UNESP), Botucatu, SP (Brazil). Inst. de Biociencias; Viana, Rodrigo S.S.; Coelho, Talita S. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
The paper illustrates the use of the Monte Carlo method, MCNP-5C code, to analyze the attenuation curve behavior of the 50 kVp radiation beam from superficial radiotherapy equipment as Dermopan2 model. The simulations seek to verify the MCNP-5C code performance to study the variation of the attenuation curve - percentage depth dose (PDD) curve - in function of the radiation field dimension used at radiotherapy of skin tumors with 50 kVp X-ray beams. The PDD curve was calculated for six different radiation field sizes with circular geometry of 1.0, 2.0, 3.0, 4.0, 5.0 and 6.0 cm in diameter. The radiation source was modeled considering a tungsten target with inclination 30 deg, focal point of 6.5 mm in diameter and energy beam of 50 kVp; the X-ray spectrum was calculated with the MCNP-5C code adopting total filtration (beryllium window of 1 mm and aluminum additional filter of 1 mm). The PDD showed decreasing behavior with the attenuation depth similar what is presented on the literature. There was not significant variation at the PDD values for the radiation field between 1.0 and 4.0 cm in diameter. The differences increased for fields of 5.0 and 6.0 cm and at attenuation depth higher than 1.0 cm. When it is compared the PDD values for fields of 3.0 and 6.0 cm in diameter, it verifies the greater difference (12.6 %) at depth of 5.7 cm, proving the scattered radiation effect. The MCNP-5C code showed as an appropriate procedure to analyze the attenuation curves of the superficial radiotherapy beams. (author)
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...
Bot3p consists of a set of standard Fortran 77 language programs that gives the users of the deterministic transport codes Dort and Tort 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. Bot3p produces at the same time the geometrical and material distribution data for the deterministic transport codes Twodant and Threedant and, only in three-dimensional (3D) Cartesian geometry, for the Monte Carlo Transport Code MCNP. 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. Through the use of Bot3p, radiation transport problems with complex 3D 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. (author)
The present work had as objective to analyze the distribution profile of a therapeutic dose of radiation produced by radioactive sources used in radiotherapy procedures in superficial lesions on the skin. The experimental measurements for analysis of dosimetric radiation sources were compared with calculations obtained from the computer system based on the Monte Carlo Method. The results obtained by the computations calculations using the code MCNP-4C showed a good agreement with the experimental measurements. A comparison of different treatment modalities allows an indication of more appropriate procedures for each clinical case. (author)
Verification of the MCNP (TM) Perturbation Correction Feature for Cross-Section Dependent Tallies
A. K. Hess; G. W. McKinney; J. S. Hendricks; L. L. Carter
1998-10-01
The Monte Carlo N-Particle Transport Code MCNP version 4B perturbation capability has been extended to cross-section dependent tallies and to the track-length estimate of Iqff in criticality problems. We present the complete theory of the MCNP perturbation capability including the correction to MCNP4B which enables cross-section dependent perturbation tallies. We also present the MCNP interface as an upgrade to the MCNP4B manual. Finally, we present test results demonstrating the validity of the perturbation capability in MCNP, particularly cross-section dependent problems.
MCAM 5: an advanced interface program for multiple Monte Carlo Codes
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
More than 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The benchmark calculations reported here are part of an ongoing multiyear, multiperson effort to benchmark version 4 of the MCNP code. The MCNP is a Monte Carlo three-dimensional general-purpose, continuous-energy neutron, photon, and electron transport code. It is used around the world for many applications including aerospace, oil-well logging, physics experiments, criticality safety, reactor analysis, medical imaging, defense applications, accelerator design, radiation hardening, radiation shielding, health physics, fusion research, and education. The first phase of the benchmark project consisted of analytic and photon problems. The second phase consists of the ENDF/B-V neutron problems reported in this paper and in more detail in the comprehensive report. A cooperative program being carried out a General Electric, San Jose, consists of light water reactor benchmark problems. A subsequent phase focusing on electron problems is planned
Mosteller, R. D. (Russell D.)
2002-01-01
Two validation suites, one for criticality and another for radiation shielding, have been defined and tested for the MCNP Monte Carlo code. All of the cases in the validation suites are based on experiments so that calculated and measured results can be compared in a meaningful way. The cases in the validation suites are described, and results from those cases are discussed. For several years, the distribution package for the MCNP Monte Carlo code1 has included an installation test suite to verify that MCNP has been installed correctly. However, the cases in that suite have been constructed primarily to test options within the code and to execute quickly. Consequently, they do not produce well-converged answers, and many of them are physically unrealistic. To remedy these deficiencies, sets of validation suites are being defined and tested for specific types of applications. All of the cases in the validation suites are based on benchmark experiments. Consequently, the results from the measurements are reliable and quantifiable, and calculated results can be compared with them in a meaningful way. Currently, validation suites exist for criticality and radiation-shielding applications.
The development of depletion program coupled with Monte Carlo computer code
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
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)
Neutronic conceptual design of the ETRR-2 cold-neutron source using the MCNP code
Khalil, M. Y.; Shaat, M. K.; Abdelfattah, A. Y.
2005-04-01
A conceptual neutronic design of the cold-neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was done using the MCNP code. Parametric analysis to chose the type and geometry of the moderator, and the required CNS dimensions to maximize the cold neutron production was performed. The moderator cell has a spherical annulus structure containing liquid hydrogen. The cold neutron gain and cold neutron brightness are calculated together with the nuclear heat load of the CNS. Analysis of the estimated performance of the CNS has been done regarding the effect of void fraction in the moderator cell together with the ortho: para ratio.
Neutronic conceptual design of the ETRR-2 cold-neutron source using the MCNP code
A conceptual neutronic design of the cold-neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was done using the MCNP code. Parametric analysis to chose the type and geometry of the moderator, and the required CNS dimensions to maximize the cold neutron production was performed. The moderator cell has a spherical annulus structure containing liquid hydrogen. The cold neutron gain and cold neutron brightness are calculated together with the nuclear heat load of the CNS. Analysis of the estimated performance of the CNS has been done regarding the effect of void fraction in the moderator cell together with the ortho: para ratio
Shielding analysis of high level waste water storage facilities using MCNP code
The neutron and gamma-ray transport analysis for the facility as a reprocessing facility with large buildings having thick shielding was made. Radiation shielding analysis consists of a deep transmission calculation for the concrete wall and a skyshine calculation for the space out of the buildings. An efficient analysis with a short running time and high accuracy needs a variance reduction technique suitable for all the calculation regions and structures. In this report, the shielding analysis using MCNP and a discrete ordinate transport code is explained and the idea and procedure of decision of variance reduction parameter is completed. (J.P.N.)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms
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.
Morse Monte Carlo Radiation Transport Code System
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)
A group of neutronics calculations in the MNSR using the MCNP-4C code
The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl4-Al) to 19.75 % LEU fuel (UO2). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ιp l and βeff). (authors)
Leal, B. [Centro Regional de Estudios Nucleares, A.P. 579C, 98068 Zacatecas (Mexico)
2004-07-01
In the majority of the laboratories, the calibration in efficiency of the detector is carried out by means of the standard sources measurement of gamma photons that have a determined activity, or for matrices that contain a variety of radionuclides that can embrace the energy range of interest. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the energy range of 80 keV to 1400 keV varying the density of the matrix, by means of the application of the Monte Carlo code MCNP-4A. The adjustment obtained shows an acceptance grade in the range of 100 to 600 keV, with a smaller percentage discrepancy to 5%. (Author)
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.
SERPENT Monte Carlo reactor physics code
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)
Accuracy assessment of a new Monte Carlo based burnup computer code
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.
A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. Each carrier can be loaded with 4 cardboard boxes (0.4x0.4x 0.4 m3 ). Each box was divided in eight equal cubes. Absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated inside the eight cubes. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to reach the desired density were used. The contributions from source, irradiator structures, sample material and other origins (ceiling, floor and walls) for the total photon spectra were also calculated. The dose distribution in the irradiator depends on the material and its density. These results show that the MCNP is an important tool to perform a dose mapping of the irradiator for each material to be irradiated. The economic benefits of the knowledge of the dose mapping for each material are important because allow to save time utilisation of UTR, dosimeters and man power. The previous knowledge of the dose mapping permits to establish an appropriate irradiation planning, which results in good dose uniformity in the material and reducing previous experimental work. (author)
Highlights: • The MCNP4C was used to calculate the gamma ray dose rate spatial distribution in for the SGIF. • Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter was conducted as well. • Good agreements were noticed between the calculated and measured results. • The maximum relative differences were less than 7%, 4% and 4% in the x, y and z directions respectively. - Abstract: A three dimensional model for the Syrian gamma irradiation facility (SGIF) is developed in this paper to calculate the gamma ray dose rate spatial distribution in the irradiation room at the 60Co source board using the MCNP-4C code. Measurement of the gamma ray dose rate spatial distribution using the Chlorobenzene dosimeter is conducted as well to compare the calculated and measured results. Good agreements are noticed between the calculated and measured results with maximum relative differences less than 7%, 4% and 4% in the x, y and z directions respectively. This agreement indicates that the established model is an accurate representation of the SGIF and can be used in the future to make the calculation design for a new irradiation facility
Criticality calculations with MCNP trademark: A primer
With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand
Importance sampling techniques and treatment of electron transport in MCNP 4A
The continuous energy Monte Carlo code MCNP was developed by the Radiation Transport Group at Los Alamos National Laboratory and the MCNP 4A version is available, now. The MCNP 4A is able to do the coupled neutron-secondary gamma-ray-electron-bremsstrahlung calculation. The calculated results, such as energy spectra, tally fluctuation chart, and geometrical input data can be displayed by using a work station. The document of the MCNP 4A code has no description on the subroutines, except few ones of 'SOURCE', 'TALLYX'. However, when we want to improve the MCNP Monte Carlo sampling techniques to get more accuracy or efficiency results for some problems, some subroutines are required or needed to revised. Three subroutines have been revised and built in the MCNP 4A code. (author)
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
Comparison of a laboratory spectrum of Eu-152 with results of simulation using the MCNP code
Rodenas, J. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain); Gallardo, S. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: sergalbe@iqn.upv.es; Ortiz, J. [Laboratorio de Radiactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)
2007-09-21
Detectors used for gamma spectrometry must be calibrated for each geometry considered in environmental radioactivity laboratories. This calibration is performed using a standard solution containing gamma emitter sources. Nevertheless, the efficiency curves obtained are periodically checked using a source such as {sup 152}Eu emitting many gamma rays that cover a wide energy range (20-1500 keV). {sup 152}Eu presents a problem because it has a lot of peaks affected by True Coincidence Summing (TCS). Two experimental measures have been performed placing the source (a Marinelli beaker) at 0 and 10 cm from the detector. Both spectra are simulated by the MCNP 4C code, where the TCS is not reproduced. Therefore, the comparison between experimental and simulated peak net areas permits one to choose the most convenient peaks to check the efficiency curves of the detector.
Installation and validation of MCNP-4A
MCNP-4A is a multi-purpose Monte Carlo program suitable for the modelling of neutron, photon, and electron transport problems. It is a particularly useful technique when studying systems containing irregular shapes. MCNP has been developed over the last 25 years by Los Alamos, and is distributed internationally via RSIC at Oak Ridge. This document describes the installation of MCNP-4A (henceforth referred to as MCNP) on the Silicon Graphics workstation (bluey.ansto.gov.au). A limited number of benchmarks pertaining to fast and thermal systems were performed to check the installation and validate the code. The results are compared to deterministic calculations performed using the AUS neutronics code system developed at ANSTO. (author)
Analysis Of Criticality Experiments Of Bandung Triga 2000 Reactor By Using MCNP-4B Code
During the first core loading of Bandung TRIGA 2000 reactor, two kinds of criticality experiment have been conducted, i.e, sub critical core loading and critical core loading experiments. The purpose of the experiments is to maximize the utilization of the reactor as well as to provide benchmark data for neutronic computer codes. In the sub critical core loading experiment, the core is loaded up to 42 fuel elements ring D, 13 fuel elements in ring, D, 6 fuel elements and 3 graphite dummies in ring E, 2 fuel elements in ring B, 2 fuel elements in ring B, 1 fuel element in ring B. In the other case, during the critical loading experiment, the core is loaded following the loading pattern planned by General Atomics, i.e: 20 fuel elements in ring B, C and D plus 5 control rods in ring D, 11 fuel elements in ring D, 6 fuel elements and 3 graphite dummies in ring E, and then the core is loaded with additional fuel elements, step by step, until the core reached its first criticality, i.e., 55 fuel elements. Prior to conduct of criticality experiments MCNP-4B code is used to plan the fuel loading pattern of the sub critical loading experiment, i.e. to assure that the core is still in sub critical state with 42 fuel elements in the core. In the calculation is assumed that the mass of U-235 in each fuel element depends on the documented burnup data, the mass of U-238 is assumed to be the same as the one in fresh fuels. Furthermore, all fission patricides as well as poisonous materials in each fuel element are ignored. The experiment results showed that the calculations of MCNP-4B also predicted that TRIGA 2000 reactor with the above assumptions, is appropriate for predicting for predicting the neutronic characteristics of Bandung TRIGA 2000 reactor
Performance of scientific computing platforms running MCNP4B
A performance study has been made for the MCNP4B Monte Carlo radiation transport code on a wide variety of scientific computing platforms ranging from personal computers to Cray mainframes. We present the timing study results using MCNP4B and its new test set and libraries. This timing study is unlike other timing studies because of its widespread reproducibility, its direct comparability to the predecessor study in 1993, and its focus upon a nuclear engineering code
Calibration and simulation of a HPGe well detector using Monte Carlo computer code
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)
Lecture Notes on Criticality Safety Validation Using MCNP & Whisper
Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,@@) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection - Ck's, weights; extreme value theory - bias, bias uncertainty; MOS for nuclear data uncertainty - GLLS) and usage are discussed.
Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code
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)
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
MCNP Code in Assessment of Variations of Effective Dose with Torso Adipose Tissue Thickness
The effective dose is the unite used in the field of radiation protection. It is a well defined doubly weighted uantity involving both physical and biological variables. Several factors may induce variation in the effective dose in different individuals of similar exposure data. One of these factors is the variation of adipose tissue thickness in different exposed individuals. This study essentially concenrs the assessment of the possible variation in the effective dose due to variation in the thickness of adipose tissue. The study was done using MCNP4b code to perform mathematical model of the human body depending on that given to the reference man developed by International Commission of Radiological Protection (ICRP), and calculate the effective dose with different thicknessess of adipose tissues. The study includes a comprehensive appraisal of the Monte Cario simulation, the Medical Internal Radiation Dose (MIRD) model for the human body, and the various mathematical considerations involved in the radiation dose calculations for the various pertinent parts of the human body. The radiation energies considered were 80 KeV, 300 KeV and I MeV, applying two exposure positions; anteroposterior (AP), postero-anterior (PA) with different adipose tissue thickness. This study is a theoretical approach based on detailed mathematical calculations of great precision that deals with all considerations involved in the mechanisms of radiation energy absorption in biological system depending on the variation in the densities of the particular in biological system depending on the variation in the densities of the particular tissues. The results obtained indicate that maximum decrease in effective dose occures with the lowest energy at 5cm adipose tissues thickeness for both AP and PA exposure positions. The results obtained were compared to similar work previsouly done using MCNP4 b showing very good agreement
MORSE Monte Carlo radiation transport code system
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
Analysis of the TRIGA Mark-II benchmark IEU-COMP-THERM-003 with Monte Carlo code MVP
The benchmark experiments of the TRIGA Mark-II reactor in the ICSBEP handbook have been analyzed with the Monte Carlo code MVP using the cross section libraries based on JENDL-3.3, JENDL-3.2 and ENDF/B-VI.8. The MCNP calculations have been also performed with the ENDF/B-VI.6 library for comparison between the MVP and MCNP results. For both cores labeled 132 and 133, which have different core configurations, the ratio of the calculated to the experimental results (C/E) for keff obtained by the MVP code is 0.999 for JENDL-3.3, 1.003 for JENDL-3.2, and 0.998 for ENDF/B-VI.8. For the MCNP code, the C/E values are 0.998 for both Core 132 and 133. All the calculated results agree with the reference values within the experimental uncertainties. The results obtained by MVP with ENDF/B-VI.8 and MCNP with ENDF/B-VI.6 differ only by 0.02% for Core 132, and by 0.01% for Core 133. (author)
MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis
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
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
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.
One of the direct analysis methods of the activity of 238U nuclide in an HPGe detector based gamma spectrometer is to measure the 63.3 keV gamma ray emitted from 234Th nuclide. The advantage of this method is to overcome the difficulty of sample sealing in the analysis of 226Ra nuclide. However for the samples of low activity as environmental ones, they must have large masses and volumes. Then another difficulty appears relating with the significant self-absorption effect of this low energy gamma ray in the sample. This effect depends on the sample geometry, density and chemical composition. The present work suggests a method to correct the self-absorption effect by applying the Monte Carlo MCNP5 code. The sample containers used in this study have cylindrical and Marinelli forms with the sample masses of hundreds grams and sample volumes of hundreds cm3. (author)
Thanh, Tran Thien; Nguyen, Vo Hoang; Chuong, Huynh Dinh; Tran, Le Bao; Tam, Hoang Duc; Binh, Nguyen Thi; Tao, Chau Van
2015-11-01
This article focuses on the possible application of a (137)Cs low-radioactive source (5mCi) and a NaI(Tl) detector for measuring the saturation thickness of solid cylindrical steel targets. In order to increase the reliability of the obtained experimental results and to verify the detector response function of Compton scattering spectrum, simulation using Monte Carlo N-particle (MCNP5) code is performed. The obtained results are in good agreement with the response functions of the simulation scattering and experimental scattering spectra. On the basis of such spectra, the saturation depth of a steel cylinder is determined by experiment and simulation at about 27mm using gamma energy of 662keV ((137)Cs) at a scattering angle of 120°. This study aims at measuring the diameter of solid cylindrical objects by gamma-scattering technique. PMID:26363240
MCNP5 development, verification, and performance
MCNP is a well-known and widely used Monte Carlo code for neutron, photon, and electron transport simulations. During the past 18 months, MCNP was completely reworked to provide MCNP5, a modernized version with many new features, including plotting enhancements, photon Doppler broadening, radiography image tallies, enhancements to source definitions, improved variance reduction, improved random number generator, tallies on a superimposed mesh, and edits of criticality safety parameters. Significant improvements in software engineering and adherence to standards have been made. Over 100 verification problems have been used to ensure that MCNP5 produces the same results as before and that all capabilities have been preserved. Testing on large parallel systems shows excellent parallel scaling. (author)
Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes
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
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.
The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 k W. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 k W. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively. (author)
Flow regime identification methodology with MCNP-X code and artificial neural network
This paper presents flow regimes identification methodology in multiphase system in annular, stratified and homogeneous oil-water-gas regimes. The principle is based on recognition of the pulse height distributions (PHD) from gamma-ray with supervised artificial neural network (ANN) systems. The detection geometry simulation comprises of two NaI(Tl) detectors and a dual-energy gamma-ray source. The measurement of scattered radiation enables the dual modality densitometry (DMD) measurement principle to be explored. Its basic principle is to combine the measurement of scattered and transmitted radiation in order to acquire information about the different flow regimes. The PHDs obtained by the detectors were used as input to ANN. The data sets required for training and testing the ANN were generated by the MCNP-X code from static and ideal theoretical models of multiphase systems. The ANN correctly identified the three different flow regimes for all data set evaluated. The results presented show that PHDs examined by ANN may be applied in the successfully flow regime identification. (author)
K0-PGNAA of pollutants in aqueous samples using MCNP code
Prompt γ-neutron activation analysis (PGNAA) using the k0 method by employing the 1951.1 keV γ-line of the 35Cl(n, γ)36Cl thermal neutron reaction as monostandard comparator was described. The method has been applied and evaluated using the anti-Compton prompt γ-ray neutron activation analysis facility using 252Cf neutron source with a neutron flux of 6.16·106 n· cm-2· s-1. A well-type HPGe detector as the main detector surrounded by NaI(Tl) guard detector has been arranged to investigate the performance of the Compton suppression spectrometer using the simplified slow circuit. The properties of neutron flux were determined by MCNP code calculations. In order to determine the efficiency curve of an HPGe detector, the prompt γ-rays from chlorine were used and an exponential curve was fitted. AC-PGNAA method has been used for the determination of high neutron absorbing elements like Cd, Sm and Gd as well as 20 light and heavy elements (Na, Mg, Al, Si, P, K, Ca, Ti, V, Mn, Sc, Fe, Co, Zn, La, Rb, Cs, As and Th) in standard reference materials (IAEA, Soil-7) and ten sediment samples collected from El-Manzala lake in northern part of Egypt. The reference material IAEA, Soil-7 was analyzed for data validation and good agreement between the experimental values and the certified values have been obtained
Simulation of detection of total content of N, P in water using PGNAA by MCNP code
The total content of N, P is an important index of water quality detection. The content of a special element in a water sample could be determined by prompt γ rays neutron activation analysis (PGNAA) quickly. The process, γ rays were emitted while the water sample was irradiated by neutron beam, was simulated by a model set up MCNP code and a pulse neutron generator as neutron source. The total content of N, P of class Ⅳ-Ⅴ water demanded by the surface water environment quality standard were used as basis. So that detection limit of N, P using PGNAA could be gained. If the total content of N, P in the water sample were small, the detection precision could be improved by increasing the neutron flux or concentrating the water sample. For contaminated water, the total content of N, P can be obtained quickly by PGNAA so that related departments could take measures to deal with polluted water in time when emergency of water pollution takes place. (authors)
Performance of the MTR core with MOX fuel using the MCNP4C2 code.
Shaaban, Ismail; Albarhoum, Mohamad
2016-08-01
The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. PMID:27213809
MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution
Cox, Lawrence J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2014-06-12
This report documents the preparations for and testing of the production release of MCNP6™1.1 beta through RSICC at ORNL. It addresses tests on supported operating systems (Linux, MacOSX, Windows) with the supported compilers (Intel, Portland Group and gfortran). Verification and Validation test results are documented elsewhere. This report does not address in detail the overall packaging of the distribution. Specifically, it does not address the nuclear and atomic data collection, the other included software packages (MCNP5, MCNPX and MCNP6) and the collection of reference documents.
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
The objective of the study is to compare the thermal neutron fluxes at specimen positions of neutron radiography facility calculated by MCNP4C code with the measurement. A model for calculation was developed using details of the reactor core configuration no. 14 and neutron radiography facility installed at the existing research reactor, TRR-1/M1 reactor. Assuming all fresh fuel elements and all control rod out condition, the thermal neutron fluxes at various specimen positions were calculated using MCNP4C code. The calculation are verified by the measurement using foil activation method. Generally, the calculated neutron fluxes are overestimated by 16-20% which is reasonably good agreement and acceptable for the complex system. The discrepancy is expected to the assumption of using fresh fuel elements, all control rod out condition, and also lacks of information in develop a more accurate model for calculation. This study shows the possibility of using the MCNP4C code to verify the thermal neutron fluxes at specimen position and shielding design of the new neutron radiography facility at the new Thai research reactor
Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code
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)
The research performed in common these last 3 years by the French Atomic Commission CEA, COGEMA and Eurisys Mesures had for main subject the realization of a complete tool of modelization for the largest range of realistic cases, the Pascalys modelization software. The main purpose of the modelization was to calculate the global measurement efficiency, which delivers the most accurate relationship between the photons emitted by the nuclear source in volume, punctual or deposited form and the germanium hyper pure detector, which detects and analyzes the received photons. It has been stated since long time that experimental global measurement efficiency becomes more and more difficult to address especially for complex scene as we can find in decommissioning and dismantling or in case of high activities for which the use of high activity reference sources become difficult to use for both health physics point of view and regulations. The choice of a calculation code is fundamental if accurate modelization is searched. MCNP represents the reference code but its use is long time calculation consuming and then not practicable in line on the field. Direct line-of-sight point kernel code as the French Atomic Commission 3-D analysis Mercure code can represent the practicable compromise between the most accurate MCNP reference code and the realistic performances needed in modelization. The comparison between the results of Pascalys-Mercure and MCNP code taking in account the last improvements of Mercure in the low energy range where the most important errors can occur, is presented in this paper, Mercure code being supported in line by the recent Pascalys 3-D modelization scene software. The incidence of the intrinsic efficiency of the Germanium detector is also approached for the total efficiency of measurement. (authors)
Luneville, L.; Chiron, M. [CEA Saclay, Dept. de Mecanique et de Technologie, 91 - Gif sur Yvette (France); Toubon, H. [Cogema BCR, 78 - Saint Quentin Yveline (France); Dogny, S. [Cogema 30 - Marcoule (France); Huver, M.; Berger, L. [Eurisys Mesures, 78 - Montigny (France)
2001-07-01
The research performed in common these last 3 years by the French Atomic Commission CEA, COGEMA and Eurisys Mesures had for main subject the realization of a complete tool of modelization for the largest range of realistic cases, the Pascalys modelization software. The main purpose of the modelization was to calculate the global measurement efficiency, which delivers the most accurate relationship between the photons emitted by the nuclear source in volume, punctual or deposited form and the germanium hyper pure detector, which detects and analyzes the received photons. It has been stated since long time that experimental global measurement efficiency becomes more and more difficult to address especially for complex scene as we can find in decommissioning and dismantling or in case of high activities for which the use of high activity reference sources become difficult to use for both health physics point of view and regulations. The choice of a calculation code is fundamental if accurate modelization is searched. MCNP represents the reference code but its use is long time calculation consuming and then not practicable in line on the field. Direct line-of-sight point kernel code as the French Atomic Commission 3-D analysis Mercure code can represent the practicable compromise between the most accurate MCNP reference code and the realistic performances needed in modelization. The comparison between the results of Pascalys-Mercure and MCNP code taking in account the last improvements of Mercure in the low energy range where the most important errors can occur, is presented in this paper, Mercure code being supported in line by the recent Pascalys 3-D modelization scene software. The incidence of the intrinsic efficiency of the Germanium detector is also approached for the total efficiency of measurement. (authors)
The MCNP code development program is a relatively large and rapidly changing project in the small and highly-specialized field of radiation transport, specifically radiation protection and shielding. A number of major new MCNP initiatives are described in the subsequent papers in this session. The focus of this paper is the important new developments not described elsewhere and a number of recent developments that have been available since MCNP4A but have gone unnoticed. In particular, we report for the first time a new MCNP quality assurance initiative providing 97% test coverage, a new MCNP feature enabling plotting of nuclear data, and the other new features developed so far for MCNP4B. Finally, an attempt is made to articulate how all these fit together into the overall MCNP development program
Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.
Henry, R; Tiselj, I; Snoj, L
2015-03-01
New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. PMID:25576735
National Laboratory, UCRL-ID-126455, Rev. 1, November, 1997] and MCNP4B [MCNP - A General Monte Carlo N-Particle Transport Code, Version 4B, La-12625-m, March 1997, Los Alamos National Laboratory] for two different configurations of the system is discussed, separating the n and γ contributions, in the light of the physical interpretation of the results in terms of first flight and of scattered neutron fluxes, of primary γ and of secondary γ generated by inelastically scattered or radiatively captured neutrons. The final conclusions indicate some guidelines and suggest possible improvements for the future neutronic shielding design for a HIF facility
Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2015-09-15
The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)
E language based on MCNP modeling software for autonomous
MCNP (Monte Carlo N-Particle Code) is based on the Monte Carlo method for computing neutron, photon and other particles as the object of the movement simulation computer program. Because of its powerful computing simulation, flexible and universal features in many fields has been widely used, but due to a software professional in the operating area has been greatly restricted, so that in later development has been greatly hindered. E-language was used in order to develop the autonomy of MCNP modeling software, used to address users not familiar with MCNP and can not create object model, get rid of dull red tape 'notebook' type of program type and built a new MCNP modeling system. (authors)
New strategies of sensitivity analysis capabilities in continuous-energy Monte Carlo code RMC
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
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
Comparison of CAP88 and MCNP for Overhead Gamma-emitting Plumes
Mcnaughton, Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gillis, Jessica Mcdonnel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McClory, Aysha Reede [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Whicker, Jeffrey Jay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fuehne, David Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-01-08
The purpose of this paper is to use the Monte Carlo N-Particle Code (MCNP) to investigate the dose from gamma-emitting radionuclides such as Carbon-11 when a plume passes overhead. MCNP results are compared with results from the EPA program, CAP88. In some cases, typically near the source during stable conditions, the CAP88 results are less than the MCNP results. However, in the case of a receptor 800 m from a source at the Los Alamos Neutron Science Center (LANSCE), the CAP88 result is greater than the MCNP result.
Criticality calculations with MCNP{sup TM}: A primer
Mendius, P.W. [ed.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.
1994-08-01
The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.
Fast code for Monte Carlo simulations
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)
Criticality calculations with MCNP{trademark}: A primer
Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A. [New Mexico Univ., Albuquerque, NM (United States)
1994-06-06
With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.
MCNP Progress & Performance Improvements
Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-04-14
Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.
Use experience of MCNP in nuclear energy study
'MCNP Use Experience' Working Group was organized in the Special Committee on Nuclear Code Evaluation. Each calculation sample for nuclear energy study using the continuous energy Monte Carlo code MCNP has been reported in this working group. This report consists of (1) promotion of calculation science and its subjects, (2) compilation of point-wise cross sections for MCNP, (3) neutron and gamma-ray transport calculation of fusion reactor system, (4) concept design of nuclear transmutation system using accelerator, (5) JMTR core calculation, (6) prompt neutron time attenuation constant calculation of TCA core, (7) criticality safety analysis, (8) neutron and gamma-ray transport calculation for exposure evaluation, (9) neutron and gamma-ray transport calculation of shielding system. The MCNP input list is also shown in each calculation sample. The adaptation of the best variance reduction technique is a important subject. (author)
MORET: Version 4.B. A multigroup Monte Carlo criticality code
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)
Boulaich, Y., E-mail: boulaich@cnesten.org.ma [Radiations and Nuclear Systems Laboratory, University Abdelmalek Essaadi, Faculty of Sciences of Tetouan (Morocco); CEN-Maamora, CNESTEN, Rabat (Morocco); El Bardouni, T. [Radiations and Nuclear Systems Laboratory, University Abdelmalek Essaadi, Faculty of Sciences of Tetouan (Morocco); Erradi, L. [University Mohammed V of Rabat (Morocco); Chakir, E. [LRM/EPTN, Department of Physics, Faculty of Sciences, Kenitra (Morocco); Boukhal, H. [Radiations and Nuclear Systems Laboratory, University Abdelmalek Essaadi, Faculty of Sciences of Tetouan (Morocco); Nacir, B. [CEN-Maamora, CNESTEN, Rabat (Morocco); El Younoussi, C.; El Bakkari, B. [Radiations and Nuclear Systems Laboratory, University Abdelmalek Essaadi, Faculty of Sciences of Tetouan (Morocco); CEN-Maamora, CNESTEN, Rabat (Morocco); Merroun, O.; Zoubair, M. [Radiations and Nuclear Systems Laboratory, University Abdelmalek Essaadi, Faculty of Sciences of Tetouan (Morocco)
2011-08-15
Highlights: > In the present work, we have analyzed the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. > Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values. > In order to specify the source of the relatively large discrepancy in the case of ENDF-BVII nuclear data evaluation, the k{sub eff} discrepancy between ENDF-BVII and JENDL3.3 was decomposed by using sensitivity and uncertainty analysis technique. - Abstract: In the present work, we analyze the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA/Cadarache, was mainly dedicated to the RTC studies for both UO{sub 2} and UO{sub 2}-PuO{sub 2} PWR type lattices covering the whole temperature range from 20 deg. C to 300 deg. C. We have developed an accurate 3D model of the EOLE reactor by using the MCNP5 Monte Carlo code which guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII, JEFF3.1 and JENDL3.3 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values; the
On-The-Fly neutron Doppler broadening in MCNP
Multi-physics calculations may involve coupling continuous-energy Monte Carlo neutronics codes to CFD codes that provide many thousands or even millions of region temperatures. The traditional Monte Carlo approach - using pre-calculated Doppler broadened nuclear cross-sections - is not feasible for these large multi-physics problems. Instead, an On-the-Fly (OTF) Doppler broadening methodology is required, whereby neutron cross-sections are broadened during the Monte Carlo transport. To this end, we have developed a methodology for MCNP to provide OTF broadening based on cell temperatures during neutron tracking. The method enables the use of many thousands or more temperatures in MCNP Monte Carlo calculations for multi-physics applications, significantly advancing the state-of-the-art by permitting the solution of problems that were not previously possible with continuous-energy Monte Carlo codes. A production library with an extended set of isotopes has been developed for use with MCNP6. Calculations of test problems with MCNP6 and the new library demonstrate the accuracy and effectiveness of the OTF approach. (authors)
Verification of the Monte Carlo code RMC with a whole PWR MOX/UO2 core benchmark
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)
MOx benchmark calculations by deterministic and Monte Carlo codes
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
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
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
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.
Radiation calculations using LAHET/MCNP/CINDER90
The LAHET Monte Carlo code system has recently been expanded to include high energy hadronic interactions via the FLUKA code, while retaining the original Los Alamos versions of HETC and ISABEL at lower energies. Electrons and photons are transported with EGS4 or ITS, while the MCNP coupled neutron/photon Monte Carlo code provides analysis of neutrons with kinetic energies less than 20 MeV. An interface with the CINDER activation code is now in common use. Various other changes have been made to facilitate analysis of high energy accelerator radiation environments and experimental physics apparatus, such as those found at SSC and RHIC. Current code developments and applications are reviewed
Varian 2100C/D Clinac 18 MV photon phase space file characterization and modeling by using MCNP Code
Ezzati, Ahad Ollah
2015-07-01
Multiple points and a spatial mesh based surface source model (MPSMBSS) was generated for 18MV Varian 2100 C/D Clinac phase space file (PSF) and implemented in MCNP code. The generated source model (SM) was benchmarked against PSF and measurements. PDDs and profiles were calculated using the SM and original PSF for different field sizes from 5 × 5 to 20 × 20 cm2. Agreement was within 2% of the maximum dose at 100cm SSD for beam profiles at the depths of 4cm and 15cm with respect to the original PSF. Differences between measured and calculated points were less than 2% of the maximum dose or 2mm distance to agreement (DTA) at 100 cm SSD. Thus it can be concluded that the modified MCNP code can be used for radiotherapy calculations including multiple source model (MSM) and using the source biasing capability of MPSMBSS can increase the simulation speed up to 3600 for field sizes smaller than 5 × 5 cm2.
Point KENO V.a: A continuous-energy Monte Carlo code for transport applications
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)