Parallel processing of Monte Carlo code MCNP for particle transport problem

Energy Technology Data Exchange (ETDEWEB)

Higuchi, Kenji; Kawasaki, Takuji

1996-06-01

It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)

RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes

International Nuclear Information System (INIS)

Parisi, C.; D'Auria, F.

2007-01-01

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)

MCNP: a general Monte Carlo code for neutron and photon transport

International Nuclear Information System (INIS)

1979-11-01

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

Coupling the MCNP Monte Carlo code and the FISPACT activation code with automatic visualization of the results of simulations

International Nuclear Information System (INIS)

Bourauel, Peter; Nabbi, Rahim; Biel, Wolfgang; Forrest, Robin

2009-01-01

The MCNP 3D Monte Carlo computer code is used not only for criticality calculations of nuclear systems but also to simulate transports of radiation and particles. The findings so obtained about neutron flux distribution and the associated spectra allow information about materials activation, nuclear heating, and radiation damage to be obtained by means of activation codes such as FISPACT. The stochastic character of particle and radiation transport processes normally links findings to the materials cells making up the geometry model of MCNP. Where high spatial resolution is required for the activation calculations with FISPACT, fine segmentation of the MCNP geometry becomes compulsory, which implies considerable expense for the modeling process. For this reason, an alternative simulation technique has been developed in an effort to automate and optimize data transfer between MCNP and FISPACT. (orig.)

Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

1987-02-01

Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

Verification of the shift Monte Carlo code with the C5G7 reactor benchmark

International Nuclear Information System (INIS)

Sly, N. C.; Mervin, B. T.; Mosher, S. W.; Evans, T. M.; Wagner, J. C.; Maldonado, G. I.

2012-01-01

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)

Parallel MCNP Monte Carlo transport calculations with MPI

International Nuclear Information System (INIS)

Wagner, J.C.; Haghighat, A.

1996-01-01

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

MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. 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. 3 - Restrictions on the complexity of the problem: None noted

A parallelization study of the general purpose Monte Carlo code MCNP4 on a distributed memory highly parallel computer

International Nuclear Information System (INIS)

Yamazaki, Takao; Fujisaki, Masahide; Okuda, Motoi; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

1993-01-01

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.)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S; Lewis, B.J.; Bonin, H.W.

2013-01-01

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 k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff 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)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

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 k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} 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)

Electron absorbed dose comparison between MCNP5 and Penelope Monte Carlo code for microdosimetry

International Nuclear Information System (INIS)

Cintra, Felipe B. de; Yoriyaz, Helio

2009-01-01

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/cm 3 ) 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/cm 3 . 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)

Performance of the improved version of Monte Carlo Code A3MCNP for cask shielding design

International Nuclear Information System (INIS)

Hasegawa, T.; Ueki, K.; Sato, O.; Sjoden, G.E.; Miyake, Y.; Ohmura, M.; Haghighat, A.

2004-01-01

A 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 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 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 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for cask neutron and gamma-ray shielding problem

MCNP-REN a Monte Carlo tool for neutron detector design

CERN Document Server

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...

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

International Nuclear Information System (INIS)

Kang, H-S; Jang, M-S; Kim, S-R; Park, J-M; Kim, K-N

2015-01-01

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

Energy Technology Data Exchange (ETDEWEB)

Kang, H-S; Jang, M-S; Kim, S-R [NESS, Daejeon (Korea, Republic of); Park, J-M; Kim, K-N [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis.

Tally and geometry definition influence on the computing time in radiotherapy treatment planning with MCNP Monte Carlo code.

Science.gov (United States)

Juste, B; Miro, R; Gallardo, S; Santos, A; Verdu, G

2006-01-01

The present work has simulated the photon and electron transport in a Theratron 780 (MDS Nordion) (60)Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle), version 5. In order to become computationally more efficient in view of taking part in the practical field of radiotherapy treatment planning, this work is focused mainly on the analysis of dose results and on the required computing time of different tallies applied in the model to speed up calculations.

MCNP5 development, verification, and performance

International Nuclear Information System (INIS)

Forrest B, Brown

2003-01-01

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)

MCNP5 development, verification, and performance

Energy Technology Data Exchange (ETDEWEB)

Forrest B, Brown [Los Alamos National Laboratory (United States)

2003-07-01

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)

Wielandt acceleration for MCNP5 Monte Carlo eigenvalue calculations

International Nuclear Information System (INIS)

Brown, F.

2007-01-01

Monte Carlo criticality calculations use the power iteration method to determine the eigenvalue (k eff ) and eigenfunction (fission source distribution) of the fundamental mode. A recently proposed method for accelerating convergence of the Monte Carlo power iteration using Wielandt's method has been implemented in a test version of MCNP5. The method is shown to provide dramatic improvements in convergence rates and to greatly reduce the possibility of false convergence assessment. The method is effective and efficient, improving the Monte Carlo figure-of-merit for many problems. In addition, the method should eliminate most of the underprediction bias in confidence intervals for Monte Carlo criticality calculations. (authors)

Obtaining of primary rays of spectrum X codes Penelope and MCNP5; Obtencion del espectro primario de Rayos X con los codigos Penelope y MCNP5

Energy Technology Data Exchange (ETDEWEB)

Pozuelo, F.; Querol, A.; Gallardo, S.; Rodenas, J.; Verdu, G.

2012-07-01

In this case, used codes PENELOPE MCNP5, based on the Monte Carlo method for x-ray spectrum taking into account the characteristics of the x-ray tube. In order to achieve a greater fit of simulated by the theoretical spectrum. It carried out a sensitivity analysis of the parameters available in both codes. The obtaining of the simulated spectrum could lead to an improvement in quality control of the x-ray tube to incorporate it as a method complementary to techniques.

Evaluation of the methodology for dose calculation in microdosimetry with electrons sources using the MCNP5 Code

International Nuclear Information System (INIS)

Cintra, Felipe Belonsi de

2010-01-01

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)

Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system.

Science.gov (United States)

Natto, S A; Lewis, D G; Ryde, S J

1998-01-01

The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

Performance of the improved version of Monte Carlo code A 3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Omura, M.; Miyake, Y.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G. E.

2005-01-01

A 3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which 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 3MCNP uses the three-dimensional (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 A 3MCNP 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 3MCNP 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 3MCNP (referred to as A 3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem. (authors)

Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code

Science.gov (United States)

Peri, Eyal; Orion, Itzhak

2017-09-01

High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.

Convergence testing for MCNP5 Monte Carlo eigenvalue calculations

International Nuclear Information System (INIS)

Brown, F.; Nease, B.; Cheatham, J.

2007-01-01

Determining convergence of Monte Carlo criticality problems is complicated by the statistical noise inherent in the random, walks of the neutrons in each generation. The latest version of MCNP5 incorporates an important new tool for assessing convergence: the Shannon entropy of the fission source distribution, H src . Shannon entropy is a well-known concept from information theory and provides a single number for each iteration to help characterize convergence trends for the fission source distribution. MCNP5 computes H src for each iteration, and these values may be plotted to examine convergence trends. Convergence testing should include both k eff and H src , since the fission distribution will converge more slowly than k eff , especially when the dominance ratio is close to 1.0. (authors)

SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

International Nuclear Information System (INIS)

Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi

2009-05-01

Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)

Parallelization of MCNP Monte Carlo neutron and photon transport code in parallel virtual machine and message passing interface

International Nuclear Information System (INIS)

Deng Li; Xie Zhongsheng

1999-01-01

The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)

Development of a consistent Monte Carlo-deterministic transport methodology based on the method of characteristics and MCNP5

International Nuclear Information System (INIS)

Karriem, Z.; Ivanov, K.; Zamonsky, O.

2011-01-01

This paper presents work that has been performed to develop an integrated Monte Carlo- Deterministic transport methodology in which the two methods make use of exactly the same general geometry and multigroup nuclear data. The envisioned application of this methodology is in reactor lattice physics methods development and shielding calculations. The methodology will be based on the Method of Long Characteristics (MOC) and the Monte Carlo N-Particle Transport code MCNP5. Important initial developments pertaining to ray tracing and the development of an MOC flux solver for the proposed methodology are described. Results showing the viability of the methodology are presented for two 2-D general geometry transport problems. The essential developments presented is the use of MCNP as geometry construction and ray tracing tool for the MOC, verification of the ray tracing indexing scheme that was developed to represent the MCNP geometry in the MOC and the verification of the prototype 2-D MOC flux solver. (author)

Using lattice tools and unfolding methods for hpge detector efficiency simulation with the Monte Carlo code MCNP5

International Nuclear Information System (INIS)

Querol, A.; Gallardo, S.; Ródenas, J.; Verdú, G.

2015-01-01

In environmental radioactivity measurements, High Purity Germanium (HPGe) detectors are commonly used due to their excellent resolution. Efficiency calibration of detectors is essential to determine activity of radionuclides. The Monte Carlo method has been proved to be a powerful tool to complement efficiency calculations. In aged detectors, efficiency is partially deteriorated due to the dead layer increasing and consequently, the active volume decreasing. The characterization of the radiation transport in the dead layer is essential for a realistic HPGe simulation. In this work, the MCNP5 code is used to calculate the detector efficiency. The F4MESH tally is used to determine the photon and electron fluence in the dead layer and the active volume. The energy deposited in the Ge has been analyzed using the ⁎F8 tally. The F8 tally is used to obtain spectra and to calculate the detector efficiency. When the photon fluence and the energy deposition in the crystal are known, some unfolding methods can be used to estimate the activity of a given source. In this way, the efficiency is obtained and serves to verify the value obtained by other methods. - Highlights: • The MCNP5 code is used to estimate the dead layer thickness of an HPGe detector. • The F4MESH tally is applied to verify where interactions occur into the Ge crystal. • PHD and the energy deposited are obtained with F8 and ⁎F8 tallies, respectively. • An average dead layer between 70 and 80 µm is obtained for the HPGe studied. • The efficiency is calculated applying the TSVD method to the response matrix.

MCNP-REN: a Monte Carlo tool for neutron detector design

International Nuclear Information System (INIS)

Abhold, M.E.; Baker, M.C.

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 were taken with the Underwater Coincidence Counter, and measurements of highly enriched uranium reactor fuel were taken with the active neutron interrogation Research Reactor Fuel Counter and compared to calculation. Simulations completed for other detector design applications are described. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions

SWAT4.0 - The integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

International Nuclear Information System (INIS)

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki

2015-03-01

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)

Acceleration of the MCNP branch of the OCTOPUS depletion code system

Energy Technology Data Exchange (ETDEWEB)

Pijlgroms, B.J.; Hogenbirk, A.; Oppe, J. [Section Nuclear and Reactor Physics, ECN Nuclear Research, Petten (Netherlands)

1998-09-01

OCTOPUS depletion calculations using the 3D Monte Carlo spectrum code MCNP (Monte Carlo Code for Neutron and Photon Transport) require much computing time. In a former implementation, the time required by OCTOPUS to perform multi-zone calculations, increased roughly proportional to the number of burnable zones. By using a different method the situation has improved considerably. In the new implementation described here, the dependence of the computing time on the number of zones has been moved from the MCNP code to a faster postprocessing code. By this, the overall computing time will reduce substantially. 11 refs.

Acceleration of the MCNP branch of the OCTOPUS depletion code system

International Nuclear Information System (INIS)

Pijlgroms, B.J.; Hogenbirk, A.; Oppe, J.

1998-09-01

OCTOPUS depletion calculations using the 3D Monte Carlo spectrum code MCNP (Monte Carlo Code for Neutron and Photon Transport) require much computing time. In a former implementation, the time required by OCTOPUS to perform multi-zone calculations, increased roughly proportional to the number of burnable zones. By using a different method the situation has improved considerably. In the new implementation described here, the dependence of the computing time on the number of zones has been moved from the MCNP code to a faster postprocessing code. By this, the overall computing time will reduce substantially. 11 refs

MCNP: Photon benchmark problems

International Nuclear Information System (INIS)

Whalen, D.J.; Hollowell, D.E.; Hendricks, J.S.

1991-09-01

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

Potential of the MCNP computer code

International Nuclear Information System (INIS)

Kyncl, J.

1995-01-01

The MCNP code is designed for numerical solution of neutron, photon, and electron transport problems by the Monte Carlo method. The code is based on the linear transport theory of behavior of the differential flux of the particles. The code directly uses data from the cross section point data library for input. Experience is outlined, gained in the application of the code to the calculation of the effective parameters of fuel assemblies and of the entire reactor core, to the determination of the effective parameters of the elementary fuel cell, and to the numerical solution of neutron diffusion and/or transport problems of the fuel assembly. The agreement between the calculated and observed data gives evidence that the MCNP code can be used with advantage for calculations involving WWER type fuel assemblies. (J.B.). 4 figs., 6 refs

Monte Carlo codes and Monte Carlo simulator program

International Nuclear Information System (INIS)

Higuchi, Kenji; Asai, Kiyoshi; Suganuma, Masayuki.

1990-03-01

Four typical Monte Carlo codes KENO-IV, MORSE, MCNP and VIM have been vectorized on VP-100 at Computing Center, JAERI. The problems in vector processing of Monte Carlo codes on vector processors have become clear through the work. As the result, it is recognized that these are difficulties to obtain good performance in vector processing of Monte Carlo codes. A Monte Carlo computing machine, which processes the Monte Carlo codes with high performances is being developed at our Computing Center since 1987. The concept of Monte Carlo computing machine and its performance have been investigated and estimated by using a software simulator. In this report the problems in vectorization of Monte Carlo codes, Monte Carlo pipelines proposed to mitigate these difficulties and the results of the performance estimation of the Monte Carlo computing machine by the simulator are described. (author)

Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics

International Nuclear Information System (INIS)

Parreno Z, F.; Paucar J, R.; Picon C, C.

1998-01-01

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)

Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60

Directory of Open Access Journals (Sweden)

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.

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

Science.gov (United States)

Blanchet, David; Pénéliau, Yannick; Eschbach, Romain; Fontaine, Bruno; Cantone, Bruno; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Letellier, Laurent; Proust, Maxime; Mota, Fernando; Palermo, Iole; Rios, Luis; Guern, Frédéric Le; Kocan, Martin; Reichle, Roger

2017-09-01

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.

Monte Carlo importance sampling for the MCNP trademark general source

International Nuclear Information System (INIS)

Lichtenstein, H.

1996-01-01

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

MCNP: a general Monte Carlo code for neutron and photon transport

International Nuclear Information System (INIS)

1978-07-01

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

Modification to the Monte Carlo N-Particle (MCNP) Visual Editor (MCNPVised) to Read in Computer Aided Design (CAD) Files

International Nuclear Information System (INIS)

Randolph Schwarz; Leland L. Carter; Alysia Schwarz

2005-01-01

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

Features of MCNP6

International Nuclear Information System (INIS)

Goorley, T.; James, M.; Booth, T.; Brown, F.; Bull, J.; Cox, L.J.; Durkee, J.; Elson, J.; Fensin, M.; Forster, R.A.; Hendricks, J.; Hughes, H.G.; Johns, R.; Kiedrowski, B.; Martz, R.; Mashnik, S.; McKinney, G.; Pelowitz, D.; Prael, R.; Sweezy, J.

2016-01-01

Highlights: • MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. • MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. • These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and Nonproliferation Division, Radiation Transport Modeling Team (NEN-5) respectively, have combined their code development efforts to produce the next evolution of MCNP. • While maintenance and major bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. • In fact, the initial release of MCNP6 contains numerous new features not previously found in either code. • These new features are summarized in this document. • Packaged with MCNP6 is also the new production release of the ENDF/B-VII.1 nuclear data files usable by MCNP. • The high quality of the overall merged code, usefulness of these new features, along with the desire in the user community to start using the merged code, have led us to make the first MCNP6 production release: MCNP6 version 1. • High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, our automated nightly software debugger tests, the underlying high quality nuclear and atomic databases, and significant testing by many beta testers. - Abstract: MCNP6 can be described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and

SU-E-T-212: Comparison of TG-43 Dosimetric Parameters of Low and High Energy Brachytherapy Sources Obtained by MCNP Code Versions of 4C, X and 5

Energy Technology Data Exchange (ETDEWEB)

Zehtabian, M; Zaker, N; Sina, S [Shiraz University, Shiraz, Fars (Iran, Islamic Republic of); Meigooni, A Soleimani [Comprehensive Cancer Center of Nevada, Las Vegas, Nevada (United States)

2015-06-15

Purpose: Different versions of MCNP code are widely used for dosimetry purposes. The purpose of this study is to compare different versions of the MCNP codes in dosimetric evaluation of different brachytherapy sources. Methods: The TG-43 parameters such as dose rate constant, radial dose function, and anisotropy function of different brachytherapy sources, i.e. Pd-103, I-125, Ir-192, and Cs-137 were calculated in water phantom. The results obtained by three versions of Monte Carlo codes (MCNP4C, MCNPX, MCNP5) were compared for low and high energy brachytherapy sources. Then the cross section library of MCNP4C code was changed to ENDF/B-VI release 8 which is used in MCNP5 and MCNPX codes. Finally, the TG-43 parameters obtained using the MCNP4C-revised code, were compared with other codes. Results: The results of these investigations indicate that for high energy sources, the differences in TG-43 parameters between the codes are less than 1% for Ir-192 and less than 0.5% for Cs-137. However for low energy sources like I-125 and Pd-103, large discrepancies are observed in the g(r) values obtained by MCNP4C and the two other codes. The differences between g(r) values calculated using MCNP4C and MCNP5 at the distance of 6cm were found to be about 17% and 28% for I-125 and Pd-103 respectively. The results obtained with MCNP4C-revised and MCNPX were similar. However, the maximum difference between the results obtained with the MCNP5 and MCNP4C-revised codes was 2% at 6cm. Conclusion: The results indicate that using MCNP4C code for dosimetry of low energy brachytherapy sources can cause large errors in the results. Therefore it is recommended not to use this code for low energy sources, unless its cross section library is changed. Since the results obtained with MCNP4C-revised and MCNPX were similar, it is concluded that the difference between MCNP4C and MCNPX is their cross section libraries.

Suitability study of MCNP Monte Carlo program for use in medical physics

International Nuclear Information System (INIS)

Jeraj, R.

1998-01-01

MCNP is widely used Monte Carlo program in reactor and nuclear physics. However, an option of simulating electrons was added into the code a few years ago. With this extension MCNP became a code, potentially applicable for applications in medical physics. In 1997, a new version of the code, named MCNP4B was released, which contains several improvements in electron transport modeling. To test suitability of the code, several important issues were considered and examined. Default sampling in MCNP electron transport was found to be inappropriate, because it gives wrong depth dose curves for electron energies of interest in radiotherapy (Me V range). The problem can be solved if ITS-style energy sampling is used instead. One of the most difficult problems in electron transport is simulation of electron backscattering, which MCNP predicts well for all, low and high Z materials. One of the potential drawbacks, if somebody wanted to use MCNP for dosimetry on real patient geometries is that MCNP lattice calculation (e.g. when calculating dose distributions) becomes very slow for large number of scoring voxels. However, if just one scoring voxel is used, the number of geometry voxels only slightly affects the speed. In the study it was found that MCNP could be reliability used for many applications in medical physics. However, the established limitations should be taken into account when MCNP is used for a particular application.(author)

Monte Carlo determination of the conversion coefficients Hp(3)/Ka in a right cylinder phantom with 'PENELOPE' code. Comparison with 'MCNP' simulations.

Science.gov (United States)

Daures, J; Gouriou, J; Bordy, J M

2011-03-01

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 H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(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 H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

The new MCNP6 depletion capability

International Nuclear Information System (INIS)

Fensin, M. L.; James, M. R.; Hendricks, J. S.; Goorley, J. T.

2012-01-01

The first MCNP based in-line 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. (authors)

The New MCNP6 Depletion Capability

International Nuclear Information System (INIS)

Fensin, Michael Lorne; James, Michael R.; Hendricks, John S.; Goorley, John T.

2012-01-01

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.

General introduction to MCNP

International Nuclear Information System (INIS)

Naito, Yoshitaka

2001-01-01

To assist succeeding reports which will be presented in this research meeting, following items on the computer code MCNP developed in USA are presented: (1) history of development of MCNP, (2) meaning of the development, (3) progress of study on Monte Carlo codes in the nuclear code committee and (4) expectation to Monte Carlo codes. (author)

UNR. A code for processing unresolved resonance data for MCNP

International Nuclear Information System (INIS)

Hogenbirk, A.

1994-09-01

In neutron transport problems the correct treatment of self-shielding is important for those nuclei present in large concentrations. Monte Carlo calculations using continuous-energy cross section data, such as calculations with the code MCNP, offer the advantage that neutron transport is calculated in a very accurate way. Self-shielding in the resolved resonance region is taken into account exactly in MCNP. However, self-shielding in the unresolved resonance region can not be taken into account by MCNP, although the effect of it may be important in many applications. In this report a description is given of the computer code UNR. With this code problem-dependent cross section libraries can be produced for MCNP. In these libraries self-shielded cross section data in the unresolved resonance range are given, which are produced by NJOY-module UNRESR. It is noted, that the treatment for resonance self-shielding presented in this report is approximate. However, the current version of MCNP does not allow the use of probability tables, which would be a general solution. (orig.)

Implementation of a Monte Carlo based inverse planning model for clinical IMRT with MCNP code

International Nuclear Information System (INIS)

He, Tongming Tony

2003-01-01

Inaccurate dose calculations and limitations of optimization algorithms in inverse planning introduce systematic and convergence errors to treatment plans. This work was to implement a Monte Carlo based inverse planning model for clinical IMRT aiming to minimize the aforementioned errors. The strategy was to precalculate the dose matrices of beamlets in a Monte Carlo based method followed by the optimization of beamlet intensities. The MCNP 4B (Monte Carlo N-Particle version 4B) code was modified to implement selective particle transport and dose tallying in voxels and efficient estimation of statistical uncertainties. The resulting performance gain was over eleven thousand times. Due to concurrent calculation of multiple beamlets of individual ports, hundreds of beamlets in an IMRT plan could be calculated within a practical length of time. A finite-sized point source model provided a simple and accurate modeling of treatment beams. The dose matrix calculations were validated through measurements in phantoms. Agreements were better than 1.5% or 0.2 cm. The beamlet intensities were optimized using a parallel platform based optimization algorithm that was capable of escape from local minima and preventing premature convergence. The Monte Carlo based inverse planning model was applied to clinical cases. The feasibility and capability of Monte Carlo based inverse planning for clinical IMRT was demonstrated. Systematic errors in treatment plans of a commercial inverse planning system were assessed in comparison with the Monte Carlo based calculations. Discrepancies in tumor doses and critical structure doses were up to 12% and 17%, respectively. The clinical importance of Monte Carlo based inverse planning for IMRT was demonstrated

Shielding properties of 80TeO2–5TiO2–(15−x) WO3–xAnOm glasses using WinXCom and MCNP5 code

International Nuclear Information System (INIS)

Dong, M.G.; El-Mallawany, R.; Sayyed, M.I.; Tekin, H.O.

2017-01-01

Gamma ray shielding properties of 80TeO 2 –5TiO 2 –(15−x) WO 3 –xA n O m glasses, where A n O m is Nb 2 O 5 = 0.01, 5, Nd 2 O 3 = 3, 5 and Er 2 O 3 = 5 mol% have been achieved. Shielding parameters; mass attenuation coefficients, half value layers, and macroscopic effective removal cross section for fast neutrons have been computed by using WinXCom program and MCNP5 Monte Carlo code. In addition, by using Geometric Progression method (G-P), exposure buildup factor values were also calculated. Variations of shielding parameters are discussed for the effect of REO addition into the glasses and photon energy. - Highlights: • The shielding properties of 80TeO 2 –5TiO 2 –(15−x) WO 3 –xA n O m glasses were evaluated. • WinXCom program and MCNP simulation codes were used in the calculations. • Good agreement was noticed between the WinXCom and MCNP5 code results.

Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics; Introduccion a la simulacion con el codigo de Monte Carlo MCNP y sus aplicaciones en Fisica Medica

Energy Technology Data Exchange (ETDEWEB)

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)

MCOR - Monte Carlo depletion code for reference LWR calculations

Energy Technology Data Exchange (ETDEWEB)

Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

2011-04-15

Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

MCOR - Monte Carlo depletion code for reference LWR calculations

International Nuclear Information System (INIS)

Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan

2011-01-01

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

Voxel2MCNP: a framework for modeling, simulation and evaluation of radiation transport scenarios for Monte Carlo codes

International Nuclear Information System (INIS)

Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian

2013-01-01

The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX’s MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application. (paper)

Advanced local dose rate calculations with the Monte Carlo code MCNP for plutonium nitrate storage containers

International Nuclear Information System (INIS)

Quade, U.

1994-01-01

Neutron- und Gamma dose rate calculations were performed for the storage containers filled with plutonium nitrate of the MOX fabrication facility of Siemens. For the particle transport calculations the Monte Carlo Code MCNP 4.2 was used. The calculated results were compared with experimental dose rate measurements. It can be stated that the choice of the code system was appropriate since all aspects of the many facettes of the problem were well reproduced in the calculations. The position dependency as well as the influence of the shieldings, the reflections and the mutual influences of the sources were well described by the calculations for the gamma and for the neutron dose rates. However, good agreement with the experimental results on the gamma dose rates could only be reached when the lead shielding of the detector was integrated into the geometry modelling of the calculations. For some few cases of thick shieldings and soft gamma ray sources the statistics of the calculational results were not sufficient. In such cases more elaborate variance reduction methods must be applied in future calculations. Thus the MCNP code in connection with NGSRC has been proven as an effective tool for the solution of this type of problems. (orig./HP) [de

Treating voxel geometries in radiation protection dosimetry with a patched version of the Monte Carlo codes MCNP and MCNPX.

Science.gov (United States)

Burn, K W; Daffara, C; Gualdrini, G; Pierantoni, M; Ferrari, P

2007-01-01

The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

Energy Technology Data Exchange (ETDEWEB)

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)

MCNP(trademark) Version 5

International Nuclear Information System (INIS)

Cox, Lawrence J.; Barrett, Richard F.; Booth, Thomas Edward; Briesmeister, Judith F.; Brown, Forrest B.; Bull, Jeffrey S.; Giesler, Gregg Carl; Goorley, John T.; Mosteller, Russell D.; Forster, R. Arthur; Post, Susan E.; Prael, Richard E.; Selcow, Elizabeth Carol; Sood, Avneet

2002-01-01

The Monte Carlo transport workhorse, MCNP, is undergoing a massive renovation at Los Alamos National Laboratory (LANL) in support of the Eolus Project of the Advanced Simulation and Computing (ASCI) Program. MCNP Version 5 (V5) (expected to be released to RSICC in Spring, 2002) will consist of a major restructuring from FORTRAN-77 (with extensions) to ANSI-standard FORTRAN-90 with support for all of the features available in the present release (MCNP-4C2/4C3). To most users, the look-and-feel of MCNP will not change much except for the improvements (improved graphics, easier installation, better online documentation). For example, even with the major format change, full support for incremental patching will still be provided. In addition to the language and style updates, MCNP V5 will have various new user features. These include improved photon physics, neutral particle radiography, enhancements and additions to variance reduction methods, new source options, and improved parallelism support (PVM, MPI, OpenMP).

The development of depletion program coupled with Monte Carlo computer code

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan

2015-01-01

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 MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)

Image enhancement using MCNP5 code and MATLAB in neutron radiography.

Science.gov (United States)

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. Copyright © 2014 Elsevier Ltd. All rights reserved.

MCNP variance reduction overview

International Nuclear Information System (INIS)

Hendricks, J.S.; Booth, T.E.

1985-01-01

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

Comparison of MCNP5 and experimental results on neutron shielding effects for materials

Energy Technology Data Exchange (ETDEWEB)

Torres, D. A. (Daniel A.); Mosteller, R. D. (Russell D.); Sweezy, J. E. (Jeremy E.)

2004-01-01

The MCNP Radiation-Shielding Validation Suite was created to assess the impact on dose rates and attenuation factors of future improvements in the MCNP Monte Carlo code or its nuclear data libraries. However, it does not currently contain any deep-penetration cases. For this reason, a set of deep-penetration benchmarks has been investigated for possible inclusion in the Suite. Overall, the MCNP5 results match the measured values quite well. Furthermore, with the exception of Resin-F, there is no systematic trend in the ratio of calculated to measured results.

MCNP-X Monte Carlo Code Application for Mass Attenuation Coefficients of Concrete at Different Energies by Modeling 3 × 3 Inch NaI(Tl Detector and Comparison with XCOM and Monte Carlo Data

Directory of Open Access Journals (Sweden)

Huseyin Ozan Tekin

2016-01-01

Full Text Available Gamma-ray measurements in various research fields require efficient detectors. One of these research fields is mass attenuation coefficients of different materials. Apart from experimental studies, the Monte Carlo (MC method has become one of the most popular tools in detector studies. An NaI(Tl detector has been modeled, and, for a validation study of the modeled NaI(Tl detector, the absolute efficiency of 3 × 3 inch cylindrical NaI(Tl detector has been calculated by using the general purpose Monte Carlo code MCNP-X (version 2.4.0 and compared with previous studies in literature in the range of 661–2620 keV. In the present work, the applicability of MCNP-X Monte Carlo code for mass attenuation of concrete sample material as building material at photon energies 59.5 keV, 80 keV, 356 keV, 661.6 keV, 1173.2 keV, and 1332.5 keV has been tested by using validated NaI(Tl detector. The mass attenuation coefficients of concrete sample have been calculated. The calculated results agreed well with experimental and some other theoretical results. The results specify that this process can be followed to determine the data on the attenuation of gamma-rays with other required energies in other materials or in new complex materials. It can be concluded that data from Monte Carlo is a strong tool not only for efficiency studies but also for mass attenuation coefficients calculations.

First results of saturation curve measurements of heat-resistant steel using GEANT4 and MCNP5 codes

International Nuclear Information System (INIS)

Hoang, Duc-Tam; Tran, Thien-Thanh; Le, Bao-Tran; Vo, Hoang-Nguyen; Chau, Van-Tao; Tran, Kim-Tuyet; Huynh, Dinh-Chuong

2015-01-01

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. (author)

Criticality Calculations with MCNP6 - Practical Lectures

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)

2016-11-29

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

Criticality Calculations with MCNP6 - Practical Lectures

International Nuclear Information System (INIS)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

2016-01-01

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

Development of a Fully-Automated Monte Carlo Burnup Code Monteburns

International Nuclear Information System (INIS)

Poston, D.I.; Trellue, H.R.

1999-01-01

Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the use of continuous energy cross sections and (2) the ability to model detailed, complex, three-dimensional (3-D) geometries. These advantages allow more accurate burnup results to be obtained, provided that the user possesses the required computing power (which is required for discrete ordinate methods as well). Several linkage codes have been written that combine a Monte Carlo N-particle transport code (such as MCNP TM ) with a radioactive decay and burnup code. This paper describes one such code that was written at Los Alamos National Laboratory: monteburns. Monteburns links MCNP with the isotope generation and depletion code ORIGEN2. The basis for the development of monteburns was the need for a fully automated code that could perform accurate burnup (and other) calculations for any 3-D system (accelerator-driven or a full reactor core). Before the initial development of monteburns, a list of desired attributes was made and is given below. o The code should be fully automated (that is, after the input is set up, no further user interaction is required). . The code should allow for the irradiation of several materials concurrently (each material is evaluated collectively in MCNP and burned separately in 0RIGEN2). o The code should allow the transfer of materials (shuffling) between regions in MCNP. . The code should allow any materials to be added or removed before, during, or after each step in an automated fashion. . The code should not require the user to provide input for 0RIGEN2 and should have minimal MCNP input file requirements (other than a working MCNP deck). . The code should be relatively easy to use

Comparison study of photon attenuation characteristics of Lead-Boron Polyethylene by MCNP code, XCOM and experimental data

Science.gov (United States)

Zhang, Lei; Jia, Mingchun; Gong, Junjun; Xia, Wenming

2017-08-01

The linear attenuation coefficient, mass attenuation coefficient and mean free path of various Lead-Boron Polyethylene (PbBPE) samples which can be used as the photon shielding materials in marine reactor have been simulated using the Monte Carlo N-Particle (MCNP)-5 code. The MCNP simulation results are in good agreement with the XCOM values and the reported experimental data for source Cesium-137 and Cobalt-60. Thus, this method based on MCNP can be used to simulate the photon attenuation characteristics of various types of PbBPE materials.

Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis

Directory of Open Access Journals (Sweden)

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.

GB - a preliminary linking code between MCNP4C and Origen2.1 - DEN/UFMG version

International Nuclear Information System (INIS)

Campolina, Daniel; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Cavatoni, Andre

2009-01-01

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)

Nuclear reactor multi-physics simulations with coupled MCNP5 and STAR-CCM+

International Nuclear Information System (INIS)

Cardoni, Jeffrey Neil; Rizwan-uddin

2011-01-01

The MCNP5 Monte Carlo particle transport code has been coupled to the computational fluid dynamics code, STAR-CCM+, to provide a high fidelity multi-physics simulation tool for pressurized water nuclear reactors. The codes are executed separately and coupled externally through a Perl script. The Perl script automates the exchange of temperature, density, and volumetric heating information between the codes using ASCII text data files. Fortran90 and Java utility programs assist job automation with data post-processing and file management. The MCNP5 utility code, MAKXSF, pre-generates temperature dependent cross section libraries for the thermal feedback calculations. The MCNP5–STAR-CCM+ coupled simulation tool, dubbed MULTINUKE, was applied to a steady state, PWR cell model to demonstrate its usage and capabilities. The demonstration calculation showed reasonable results that agree with PWR values typically reported in literature. Temperature and fission reaction rate distributions were realistic and intuitive. Reactivity coefficients were also deemed reasonable in comparison to historically reported data. The demonstration problem consisted of 9,984 CFD cells and 7,489 neutronic cells. MCNP5 tallied fission energy deposition over 3,328 UO_2 cells. The coupled solution converged within eight hours and in three MULTINUKE iterations. The simulation was carried out on a 64 bit, quad core, Intel 2.8 GHz microprocessor with 1 GB RAM. The simulations on a quad core machine indicated that a massively parallelized implementation of MULTINUKE can be used to assess larger multi-million cell models. (author)

Validation of the MCNP-DSP Monte Carlo code for calculating source-driven noise parameters of subcritical systems

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.

1995-01-01

This paper describes calculations performed to validate the modified version of the MCNP code, the MCNP-DSP, used for: the neutron and photon spectra of the spontaneous fission of californium 252; the representation of the detection processes for scattering detectors; the timing of the detection process; and the calculation of the frequency analysis parameters for the MCNP-DSP code

MatMCNP: A Code for Producing Material Cards for MCNP

Energy Technology Data Exchange (ETDEWEB)

DePriest, Kendall Russell [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saavedra, Karen C. [American Structurepoint, Inc., Indianapolis, IN (United States)

2014-09-01

A code for generating MCNP material cards (MatMCNP) has been written and verified for naturally occurring, stable isotopes. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.

MCNP capabilities for nuclear well logging calculations

International Nuclear Information System (INIS)

Forster, R.A.; Little, R.C.; Briesmeister, J.F.; Hendricks, J.S.

1990-01-01

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo neutron photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data

MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

International Nuclear Information System (INIS)

Finfrock, S.H.

2009-01-01

The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL(reg s ign) processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 (le) H/Fissile (le) 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k eff ) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k eff is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately ± 0.001. For the cases where the reported benchmark k eff was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k eff is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k eff limit for calculations of the intermediate enriched uranium type systems.

Parallelization of MCNP 4, a Monte Carlo neutron and photon transport code system, in highly parallel distributed memory type computer

International Nuclear Information System (INIS)

Masukawa, Fumihiro; Takano, Makoto; Naito, Yoshitaka; Yamazaki, Takao; Fujisaki, Masahide; Suzuki, Koichiro; Okuda, Motoi.

1993-11-01

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)

Improved response function calculations for scintillation detectors using an extended version of the MCNP code

CERN Document Server

Schweda, K

2002-01-01

The analysis of (e,e'n) experiments at the Darmstadt superconducting electron linear accelerator S-DALINAC required the calculation of neutron response functions for the NE213 liquid scintillation detectors used. In an open geometry, these response functions can be obtained using the Monte Carlo codes NRESP7 and NEFF7. However, for more complex geometries, an extended version of the Monte Carlo code MCNP exists. This extended version of the MCNP code was improved upon by adding individual light-output functions for charged particles. In addition, more than one volume can be defined as a scintillator, thus allowing the simultaneous calculation of the response for multiple detector setups. With the implementation of sup 1 sup 2 C(n,n'3 alpha) reactions, all relevant reactions for neutron energies E sub n <20 MeV are now taken into consideration. The results of these calculations were compared to experimental data using monoenergetic neutrons in an open geometry and a sup 2 sup 5 sup 2 Cf neutron source in th...

Computation of a voxelized anthropomorphic phantom from Computer Tomography slices and 3D dose distribution calculation utilizing the MCNP5 Code

International Nuclear Information System (INIS)

Abella, V.; Miro, R.; Juste, B.; Verdu, G.

2008-01-01

Full text: The purpose of this work is to obtain the voxelization of a series of tomography slices in order to provide a voxelized human phantom throughout a MatLab algorithm, and the consequent simulation of the irradiation of such phantom with the photon beam generated in a Theratron 780 (MDS Nordion) 60 Co radiotherapy unit, using the Monte Carlo transport code MCNP (Monte Carlo N-Particle), version 5. The project provides as results dose mapping calculations inside the voxelized anthropomorphic phantom. Prior works have validated the cobalt therapy model utilizing a simple heterogeneous water cube-shaped phantom. The reference phantom model utilized in this work is the Zubal phantom, which consists of a group of pre-segmented CT slices of a human body. The CT slices are to be input into the Matlab program which computes the voxelization by means of two-dimensional pixel and material identification on each slice, and three-dimensional interpolation, in order to depict the phantom geometry via small cubic cells. Each slice is divided in squares with the size of the desired voxelization, and then the program searches for the pixel intensity with a predefined material at each square, making a subsequent three-dimensional interpolation. At the end of this process, the program produces a voxelized phantom in which each voxel defines the mixture of the different materials that compose it. In the case of the Zubal phantom, the voxels result in pure organ materials due to the fact that the phantom is presegmented. The output of this code follows the MCNP input deck format and is integrated in a full input model including the 60 Co radiotherapy unit. Dose rates are calculated using the MCNP5 tool FMESH, superimposed mesh tally. This feature allows to tally particles on an independent mesh over the problem geometry, and to obtain the length estimation of the particle flux, in units of particles/cm 2 (tally F4). Furthermore, the particle flux is transformed into dose by

Monte Carlo simulation on nuclear energy study. Annual report of Nuclear Code Evaluation Committee

International Nuclear Information System (INIS)

Sakurai, Kiyoshi; Yamamoto, Toshihiro

1999-03-01

In this report, research results discussed in 1998 fiscal year at Nuclear Code Evaluation Special Committee of Nuclear Code Committee were summarised. Present status of Monte Carlo calculation in high energy region investigated / discussed at Monte Carlo simulation working-group and automatic compilation system for MCNP cross sections developed at MCNP high temperature library compilation working-group were described. The 6 papers are indexed individually. (J.P.N.)

Uncertainty analysis in the simulation of an HPGe detector using the Monte Carlo Code MCNP5

International Nuclear Information System (INIS)

Gallardo, Sergio; Pozuelo, Fausto; Querol, Andrea; Verdu, Gumersindo; Rodenas, Jose; Ortiz, J.; Pereira, Claubia

2013-01-01

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)

Uncertainty analysis in the simulation of an HPGe detector using the Monte Carlo Code MCNP5

Energy Technology Data Exchange (ETDEWEB)

Gallardo, Sergio; Pozuelo, Fausto; Querol, Andrea; Verdu, Gumersindo; Rodenas, Jose, E-mail: sergalbe@upv.es [Universitat Politecnica de Valencia, Valencia, (Spain). Instituto de Seguridad Industrial, Radiofisica y Medioambiental (ISIRYM); Ortiz, J. [Universitat Politecnica de Valencia, Valencia, (Spain). Servicio de Radiaciones. Lab. de Radiactividad Ambiental; Pereira, Claubia [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2013-07-01

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)

Comparison of TG-43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes.

Science.gov (United States)

Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, Ali S

2016-03-08

Monte Carlo simulations are widely used for calculation of the dosimetric parameters 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 simulating 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.

Monte Carlo calculations of thermal neutron capture in gadolinium: a comparison of GEANT4 and MCNP with measurements.

Science.gov (United States)

Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans

2006-02-01

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

MCNP Version 6.2 Release Notes

Energy Technology Data Exchange (ETDEWEB)

Werner, Christopher John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Solomon, C. J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKinney, Gregg Walter [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dixon, David A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martz, Roger Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hughes, Henry G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cox, Lawrence James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zukaitis, Anthony J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Armstrong, J. C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Forster, Robert Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-02-05

Monte Carlo N-Particle or MCNP^® is a general-purpose Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP Version 6.2 follows the MCNP6.1.1 beta version and has been released in order to provide the radiation transport community with the latest feature developments and bug fixes for MCNP. Since the last release of MCNP major work has been conducted to improve the code base, add features, and provide tools to facilitate ease of use of MCNP version 6.2 as well as the analysis of results. These release notes serve as a general guide for the new/improved physics, source, data, tallies, unstructured mesh, code enhancements and tools. For more detailed information on each of the topics, please refer to the appropriate references or the user manual which can be found at http://mcnp.lanl.gov. This release of MCNP version 6.2 contains 39 new features in addition to 172 bug fixes and code enhancements. There are still some 33 known issues the user should familiarize themselves with (see Appendix).

Parallel processing Monte Carlo radiation transport codes

International Nuclear Information System (INIS)

McKinney, G.W.

1994-01-01

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

Igo - A Monte Carlo Code For Radiotherapy Planning

International Nuclear Information System (INIS)

Goldstein, M.; Regev, D.

1999-01-01

The goal of radiation therapy is to deliver a lethal dose to the tumor, while minimizing the dose to normal tissues and vital organs. To carry out this task, it is critical to calculate correctly the 3-D dose delivered. Monte Carlo transport methods (especially the Adjoint Monte Carlo have the potential to provide more accurate predictions of the 3-D dose the currently used methods. IG0 is a Monte Carlo code derived from the general Monte Carlo Program - MCNP, tailored specifically for calculating the effects of radiation therapy. This paper describes the IG0 transport code, the PIG0 interface and some preliminary results

Investigation of the applicability of MCNP code to complicated geometries

International Nuclear Information System (INIS)

Higuchi, Kenji; Yamaguchi, Yukichi

1994-03-01

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)

A detailed investigation of interactions within the shielding to HPGe detector response using MCNP code

Energy Technology Data Exchange (ETDEWEB)

Thanh, Tran Thien; Tao, Chau Van; Loan, Truong Thi Hong; Nhon, Mai Van; Chuong, Huynh Dinh; Au, Bui Hai [Vietnam National Univ., Ho Chi Minh City (Viet Nam). Dept. of Nuclear Physics

2012-12-15

The accuracy of the coincidence-summing corrections in gamma spectrometry depends on the total efficiency calibration that is hardly obtained over the whole energy as the required experimental conditions are not easily attained. Monte Carlo simulations using MCNP5 code was performed in order to estimate the affect of the shielding to total efficiency. The effect of HPGe response are also shown. (orig.)

Comparison of TG‐43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes

Science.gov (United States)

Zaker, Neda; Sina, Sedigheh; Koontz, Craig; Meigooni1, Ali S.

2016-01-01

Monte Carlo simulations are widely used for calculation of the dosimetric parameters 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 simulating 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. PACS number(s): 87.56.bg PMID:27074460

Modeling of the YALINA booster facility by the Monte Carlo code MONK

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Kondev, F.; Kiyavitskaya, H.; Serafimovich, I.; Bournos, V.; Fokov, Y.; Routkovskaya, C.

2007-01-01

The YALINA-Booster facility has been modeled according to the benchmark specifications defined for the IAEA activity without any geometrical homogenization using the Monte Carlo codes MONK and MCNP/MCNPX/MCB. The MONK model perfectly matches the MCNP one. The computational analyses have been extended through the MCB code, which is an extension of the MCNP code with burnup capability because of its additional feature for analyzing source driven multiplying assemblies. The main neutronics arameters of the YALINA-Booster facility were calculated using these computer codes with different nuclear data libraries based on ENDF/B-VI-0, -6, JEF-2.2, and JEF-3.1.

The OpenMC Monte Carlo particle transport code

International Nuclear Information System (INIS)

Romano, Paul K.; Forget, Benoit

2013-01-01

Highlights: ► An open source Monte Carlo particle transport code, OpenMC, has been developed. ► Solid geometry and continuous-energy physics allow high-fidelity simulations. ► Development has focused on high performance and modern I/O techniques. ► OpenMC is capable of scaling up to hundreds of thousands of processors. ► Results on a variety of benchmark problems agree with MCNP5. -- Abstract: A new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high performance scalable algorithms as well as modern software design practices. The present work describes the methods used in the OpenMC code and demonstrates the performance and accuracy of the code on a variety of problems.

Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4

International Nuclear Information System (INIS)

Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

2004-01-01

The expanding clinical use of low-energy photon emitting 125 I and 103 Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst ±5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately ±2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV

Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4.

Science.gov (United States)

Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

2004-02-07

The expanding clinical use of low-energy photon emitting 125I and 103Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst +/- 5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately +/- 2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV.

Monte Carlo dose calculations in homogeneous media and at interfaces: a comparison between GEPTS, EGSnrc, MCNP, and measurements.

Science.gov (United States)

Chibani, Omar; Li, X Allen

2002-05-01

Three Monte Carlo photon/electron transport codes (GEPTS, EGSnrc, and MCNP) are bench-marked against dose measurements in homogeneous (both low- and high-Z) media as well as at interfaces. A brief overview on physical models used by each code for photon and electron (positron) transport is given. Absolute calorimetric dose measurements for 0.5 and 1 MeV electron beams incident on homogeneous and multilayer media are compared with the predictions of the three codes. Comparison with dose measurements in two-layer media exposed to a 60Co gamma source is also performed. In addition, comparisons between the codes (including the EGS4 code) are done for (a) 0.05 to 10 MeV electron beams and positron point sources in lead, (b) high-energy photons (10 and 20 MeV) irradiating a multilayer phantom (water/steel/air), and (c) simulation of a 90Sr/90Y brachytherapy source. A good agreement is observed between the calorimetric electron dose measurements and predictions of GEPTS and EGSnrc in both homogeneous and multilayer media. MCNP outputs are found to be dependent on the energy-indexing method (Default/ITS style). This dependence is significant in homogeneous media as well as at interfaces. MCNP(ITS) fits more closely the experimental data than MCNP(DEF), except for the case of Be. At low energy (0.05 and 0.1 MeV), MCNP(ITS) dose distributions in lead show higher maximums in comparison with GEPTS and EGSnrc. EGS4 produces too penetrating electron-dose distributions in high-Z media, especially at low energy (MCNP results depend significantly on the electron energy-indexing method.

Performance Study of Monte Carlo Codes on Xeon Phi Coprocessors — Testing MCNP 6.1 and Profiling ARCHER Geometry Module on the FS7ONNi Problem

Science.gov (United States)

Liu, Tianyu; Wolfe, Noah; Lin, Hui; Zieb, Kris; Ji, Wei; Caracappa, Peter; Carothers, Christopher; Xu, X. George

2017-09-01

This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.

Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

International Nuclear Information System (INIS)

Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

1989-01-01

Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

Monte Carlo Codes Invited Session

International Nuclear Information System (INIS)

Trama, J.C.; Malvagi, F.; Brown, F.

2013-01-01

This document lists 22 Monte Carlo codes used in radiation transport applications throughout the world. For each code the names of the organization and country and/or place are given. We have the following computer codes. 1) ARCHER, USA, RPI; 2) COG11, USA, LLNL; 3) DIANE, France, CEA/DAM Bruyeres; 4) FLUKA, Italy and CERN, INFN and CERN; 5) GEANT4, International GEANT4 collaboration; 6) KENO and MONACO (SCALE), USA, ORNL; 7) MC21, USA, KAPL and Bettis; 8) MCATK, USA, LANL; 9) MCCARD, South Korea, Seoul National University; 10) MCNP6, USA, LANL; 11) MCU, Russia, Kurchatov Institute; 12) MONK and MCBEND, United Kingdom, AMEC; 13) MORET5, France, IRSN Fontenay-aux-Roses; 14) MVP2, Japan, JAEA; 15) OPENMC, USA, MIT; 16) PENELOPE, Spain, Barcelona University; 17) PHITS, Japan, JAEA; 18) PRIZMA, Russia, VNIITF; 19) RMC, China, Tsinghua University; 20) SERPENT, Finland, VTT; 21) SUPERMONTECARLO, China, CAS INEST FDS Team Hefei; and 22) TRIPOLI-4, France, CEA Saclay

Analysis of parallel computing performance of the code MCNP

International Nuclear Information System (INIS)

Wang Lei; Wang Kan; Yu Ganglin

2006-01-01

Parallel computing can reduce the running time of the code MCNP effectively. With the MPI message transmitting software, MCNP5 can achieve its parallel computing on PC cluster with Windows operating system. Parallel computing performance of MCNP is influenced by factors such as the type, the complexity level and the parameter configuration of the computing problem. This paper analyzes the parallel computing performance of MCNP regarding with these factors and gives measures to improve the MCNP parallel computing performance. (authors)

A graphical user interface for diagnostic radiology dosimetry using Monte Carlo (MCNP) simulation

International Nuclear Information System (INIS)

Collins, P.J.; Gorbatkov, D.; Schultz, F.W.

2000-01-01

Monte Carlo methods (for example, MCNP, EGGS4) are the 'gold standard' for both external and internal dosimetry in humans. These powerful simulation tools are, however, general-purpose codes and consequently do not provide a simple user interface for specific dosimetry tasks. We have developed a graphical user interface, for external radiation dosimetry (diagnostic radiology) using MCNP and an anthropomorphic mathematical phantom (Adam/Eva), which enables convenient modification and processing of the MCNP input and output files. The input form displays a colour coded, 3D representation of the phantom with a superimposed 'beam' for the required x-ray projection. The phantom can be rotated through 360 degrees and a transverse section at the level of the mid-point of the beam is also displayed. Text fields enable entry of input data (beam dimensions, source position, kVp, total filtration, focus-to-skin distance). A pull-down menu enables the user to select from 22 standard radiographic views. A standard projection can be modified, or new projection data entered if required. The input program modifies the MCNP input file and initiates processing. An output form displays the organ doses, normalised to unit skin entrance dose (with backscatter) (SED). The user can also enter the SED (calculated or measured) for a particular machine, to obtain the effective dose. To validate the program, the results for a PA Chest study (80 kVp, 2.5 mm Al total filtration) were compared with NRPB data (Jones and Wall, 1985). In conclusion, a convenient and reliable graphical user interface has been developed for MCNP, which enables dosimetry calculation for a full range of diagnostic radiological studies. (author)

Radiation field characterization of a BNCT research facility using Monte Carlo method - code MCNP-4B

International Nuclear Information System (INIS)

Hernandez, Antonio Carlos

2002-01-01

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 fluencyN T = 1,35x10 8 n/cm , a fast neutron dose of 5,86x10 -10 Gy/N T and a gamma ray dose of 8,30x10 -14 Gy/N T . (author)

Radiation field characterization of a BNCT research facility using Monte Carlo Method - Code MCNP-4B

International Nuclear Information System (INIS)

Hernandes, Antonio Carlos

2002-01-01

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 Ν Τ = 1,35x10 8 n/cm 2 , a fast neutron dose of 5,86x -1 0 Gy/Ν Τ and a gamma ray dose of 8,30x -14 Gy/Ν Τ . (author)

Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry

International Nuclear Information System (INIS)

Sohrabpour, M.; Hassanzadeh, M.; Shahriari, M.; Sharifzadeh, M.

2002-01-01

The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators

Nuclear densimeter of soil simulated in MCNP-4C code

International Nuclear Information System (INIS)

Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T.; Silva, Clemente J.G.C.

2009-01-01

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)

MCNP and OMEGA criticality calculations

International Nuclear Information System (INIS)

Seifert, E.

1998-04-01

The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

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

Monte Carlo MCNP-4B-based absorbed dose distribution estimates for patient-specific dosimetry.

Science.gov (United States)

Yoriyaz, H; Stabin, M G; dos Santos, A

2001-04-01

This study was intended to verify the capability of the Monte Carlo MCNP-4B code to evaluate spatial dose distribution based on information gathered from CT or SPECT. A new three-dimensional (3D) dose calculation approach for internal emitter use in radioimmunotherapy (RIT) was developed using the Monte Carlo MCNP-4B code as the photon and electron transport engine. It was shown that the MCNP-4B computer code can be used with voxel-based anatomic and physiologic data to provide 3D dose distributions. This study showed that the MCNP-4B code can be used to develop a treatment planning system that will provide such information in a time manner, if dose reporting is suitably optimized. If each organ is divided into small regions where the average energy deposition is calculated with a typical volume of 0.4 cm(3), regional dose distributions can be provided with reasonable central processing unit times (on the order of 12-24 h on a 200-MHz personal computer or modest workstation). Further efforts to provide semiautomated region identification (segmentation) and improvement of marrow dose calculations are needed to supply a complete system for RIT. It is envisioned that all such efforts will continue to develop and that internal dose calculations may soon be brought to a similar level of accuracy, detail, and robustness as is commonly expected in external dose treatment planning. For this study we developed a code with a user-friendly interface that works on several nuclear medicine imaging platforms and provides timely patient-specific dose information to the physician and medical physicist. Future therapy with internal emitters should use a 3D dose calculation approach, which represents a significant advance over dose information provided by the standard geometric phantoms used for more than 20 y (which permit reporting of only average organ doses for certain standardized individuals)

Recent MCNP developments

International Nuclear Information System (INIS)

Hendricks, J.S.; Briesmeister, J.F.

1991-01-01

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

Verification of Compton scattering spectrum of a 662 keV photon beam scattered on a cylindrical steel target using MCNP5 code

International Nuclear Information System (INIS)

Thanh, Tran Thien; Nguyen, Vo Hoang; Chuong, Huynh Dinh; Tran, Le Bao; Tam, Hoang Duc; Binh, Nguyen Thi; Tao, Chau Van

2015-01-01

This article focuses on the possible application of a "1"3"7Cs low-radioactive source (5 mCi) 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 27 mm using gamma energy of 662 keV ("1"3"7Cs) at a scattering angle of 120°. This study aims at measuring the diameter of solid cylindrical objects by gamma-scattering technique. - Highlights: • This study aims a possible application a "1"3"7Cs low-radioactive source (5 mCi) and a NaI(Tl) detector for measuring the saturation thickness of solid cylindrical steel targets by gamma-scattering technique. • Monte Carlo N-particle (MCNP5) code is performed to verify on the detector response function of Compton scattering spectrum. • The results show a good agreement in response function of the experimental and simulation scattering spectra. • The saturation depth of a steel cylinder is determined by experiment and simulation at about 27 mm using gamma energy of 662 keV ("1"3"7Cs) at a scattering angle of 120°.

Analysis of the variation of the attenuation curve in function of the radiation field size for k Vp X-ray beams using the MCNP-5C code

Energy Technology Data Exchange (ETDEWEB)

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)

Use of the MCNP Monte Carlo code for characterization of a pencil-type ionization chamber; Uso do código de Monte Carlo MCNP para caracterização de uma câmara de ionização tipo lápis

Energy Technology Data Exchange (ETDEWEB)

Mendonça, Dalila Souza Costa; Santos, William S.; Perini, Ana Paula, E-mail: anapaula.perini@ufu.br [Universidade Federal de Uberlândia (INFIS/UFU), MG (Brazil). Instituto de Física; Neves, Lucio Pereira; Caldas, Linda V. E. [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Belinato, Walmir [Instituto Federal de Educação, Ciência e Tecnologia da Bahia (IFBA), Vitória da Conquista, BA (Brazil)

2017-07-01

Ionization chambers are widely used in diagnostic radiology dosimetry. In this work, a special pencil-type ionization chamber, with different dimensions, configuration and materials in relation to commercial ones, was studied computationally. For this, the MCNP-4C Monte Carlo code and different radiation spectra were used to determine the influence of its components on its response. It was possible to observe that the highest influence was for the PVC wall. (author)

Potential MCNP enhancements for NCT

International Nuclear Information System (INIS)

Estes, G.P.; Taylor, W.M.

1992-01-01

MCNP a Monte Carlo radiation transport code, is currently widely used in the medical community for a variety of purposes including treatment planning, diagnostics, beam design, tomographic studies, and radiation protection. This is particularly true in the Neutron Capture Therapy (NCT) community. The current widespread medical use of MCNP after its general public distribution in about 1980 attests to the code's general versatility and usefulness, particularly since its development to date has not been influenced by medical applications. This paper discusses enhancements to MCNP that could be implemented at Los Alamos for the benefit of the NCT community. These enhancements generally fall into two categories, namely those that have already been developed to some extent but are not yet publicly available, and those that seem both needed based on our current understanding of NCT goals, and achievable based on our working knowledge of the MCNP code. MCNP is a general, coupled neutron/photon/electron Monte Carlo code developed and maintained by the Radiation Transport Group at Los Alamos. It has been used extensively for radiation shielding studies, reactor analysis, detector design, physics experiment interpretation, oil and gas well logging, radiation protection studies, accelerator design, etc. over the years. MCNP is a three-dimensional geometry, continuous energy physics code capable of modeling complex geometries, specifying material regions such as organs by the intersections of analytical surfaces

MCNP-DSP users manual

International Nuclear Information System (INIS)

Valentine, T.E.

1997-01-01

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 252 Cf-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

Shielding properties of 80TeO2-5TiO2-(15-x) WO3-xAnOm glasses using WinXCom and MCNP5 code

Science.gov (United States)

Dong, M. G.; El-Mallawany, R.; Sayyed, M. I.; Tekin, H. O.

2017-12-01

Gamma ray shielding properties of 80TeO2-5TiO2-(15-x) WO3-xAnOm glasses, where AnOm is Nb2O5 = 0.01, 5, Nd2O3 = 3, 5 and Er2O3 = 5 mol% have been achieved. Shielding parameters; mass attenuation coefficients, half value layers, and macroscopic effective removal cross section for fast neutrons have been computed by using WinXCom program and MCNP5 Monte Carlo code. In addition, by using Geometric Progression method (G-P), exposure buildup factor values were also calculated. Variations of shielding parameters are discussed for the effect of REO addition into the glasses and photon energy.

Development of MCNP interface code in HFETR

International Nuclear Information System (INIS)

Qiu Liqing; Fu Rong; Deng Caiyu

2007-01-01

In order to describe the HFETR core with MCNP method, the interface code MCNPIP for HFETR and MCNP code is developed. This paper introduces the core DXSY and flowchart of MCNPIP code, and the handling of compositions of fuel elements and requirements on hardware and software. Finally, MCNPIP code is validated against the practical application. (authors)

Implementation of multileaf collimator in a LINAC MCNP5 simulation coupled with the radiation treatment planing system PLUNC

International Nuclear Information System (INIS)

Abella, Vicente; Miro, Rafael; Juste, Belen; Verdu, Gumersindo

2010-01-01

Multileaf collimators are used on linear accelerators to provide conformal shaping of radiotherapy treatment beams, being an important tool for radiation therapy dose delivery. In this work, a multileaf collimator has been designed and implemented in the MCNP model of an Elekta Precise Linear Accelerator and introduced in PLUNC, a set of software tools for radiotherapy treatment planning (RTP) which was coupled in previous works with MCNP5 (Monte Carlo N-Particle transport code), with the purpose of comparing its effect on deterministic and Monte Carlo dose calculations. A 3D Shepp-Logan phantom was utilized as the patient model for validation purposes. Once the multileaf collimator model is implemented in the PLUNC LINAC model, a series of Matlab interfaces extract phantom and beam information created with PLUNC during the treatment plan and write it in MCNP5 input deck format. After the Monte Carlo simulation is performed, results are input back again in PLUNC in order to continue with the plan evaluation. The comparison is made via mapping of dose distribution inside the phantom with different field sizes, utilizing the MCNP5 tool EMESH, superimposed mesh tally, which allows registering the results over the problem geometry. This work follows a valid methodology for multileaf LINAC MC calculations during radiation treatment plans. (author)

About the application of MCNP4 code in nuclear reactor core design calculations

International Nuclear Information System (INIS)

Svarny, J.

2000-01-01

This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

Image enhancement using MCNP5 code and MATLAB in neutron radiography

International Nuclear Information System (INIS)

Tharwat, Montaser; Mohamed, Nader; Mongy, T.

2014-01-01

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. - 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

Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code

International Nuclear Information System (INIS)

Ferreira, Vanessa M.; Oliveira, Renato C.M.; Barros, Graiciany P.; Oliveira, Arno H.; Veloso, M. Auxiliadora F.

2011-01-01

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)

Physical models, cross sections, and numerical approximations used in MCNP and GEANT4 Monte Carlo codes for photon and electron absorbed fraction calculation.

Science.gov (United States)

Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir

2009-11-01

Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for BHTR Analysis

Energy Technology Data Exchange (ETDEWEB)

William R. Martin; John C. Lee

2009-12-30

Monte Carlo capability has been combined with a production LWR lattice physics code to allow analysis of high temperature gas reactor configurations, accounting for the double heterogeneity due to the TRISO fuel. The Monte Carlo code MCNP5 has been used in conjunction with CPM3, which was the testbench lattice physics code for this project. MCNP5 is used to perform two calculations for the geometry of interest, one with homogenized fuel compacts and the other with heterogeneous fuel compacts, where the TRISO fuel kernels are resolved by MCNP5.

An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for VHTR Analysis

International Nuclear Information System (INIS)

Martin, William R.; Lee, John C.

2009-01-01

Monte Carlo capability has been combined with a production LWR lattice physics code to allow analysis of high temperature gas reactor configurations, accounting for the double heterogeneity due to the TRISO fuel. The Monte Carlo code MCNP5 has been used in conjunction with CPM3, which was the testbench lattice physics code for this project. MCNP5 is used to perform two calculations for the geometry of interest, one with homogenized fuel compacts and the other with heterogeneous fuel compacts, where the TRISO fuel kernels are resolved by MCNP5.

Detailed resonance absorption calculations with the Monte Carlo code MCNP and collision probability version of the slowing down code ROLAIDS

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Janssen, A.J.

1994-01-01

Very accurate Mote Carlo calculations with Monte Carlo Code have been performed to serve as reference for benchmark calculations on resonance absorption by U 238 in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (authors). 9 refs., 5 figs., 2 tabs

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part I: boron neutron capture therapy models.

Science.gov (United States)

Culbertson, C N; Wangerin, K; Ghandourah, E; Jevremovic, T

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.

Simulation and study on the γ response spectrum of BGO detector by the application of monte carlo code MOCA

International Nuclear Information System (INIS)

Jia Wenbao; Chen Xiaowen; Xu Aiguo; Li Anmin

2010-01-01

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 137 Cs, 60 Co 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)

MCNP neutron benchmarks

International Nuclear Information System (INIS)

Hendricks, J.S.; Whalen, D.J.; Cardon, D.A.; Uhle, J.L.

1991-01-01

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

MCNP application for the 21 century

International Nuclear Information System (INIS)

McKinney, G.W.

2000-01-01

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

Implementation of a tree algorithm in MCNP code for nuclear well logging applications

Energy Technology Data Exchange (ETDEWEB)

Li Fusheng, E-mail: fusheng.li@bakerhughes.com [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States); Han Xiaogang [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States)

2012-07-15

The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. - Highlights: Black-Right-Pointing-Pointer Tree structure programming is suitable for Monte-Carlo based particle tracking. Black-Right-Pointing-Pointer Enhanced pulse height tally is developed for oilwell logging tool simulation. Black-Right-Pointing-Pointer Neutron interaction tally and gamma ray index tally for geochemical logging.

Development and validation of a model TRIGA Mark III reactor with code MCNP5

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

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 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)

Design of tallying function for general purpose Monte Carlo particle transport code JMCT

International Nuclear Information System (INIS)

Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin

2013-01-01

A new postponed accumulation algorithm was proposed. Based on JCOGIN (J combinatorial geometry Monte Carlo transport infrastructure) framework and the postponed accumulation algorithm, the tallying function of the general purpose Monte Carlo neutron-photon transport code JMCT was improved markedly. JMCT gets a higher tallying efficiency than MCNP 4C by 28% for simple geometry model, and JMCT is faster than MCNP 4C by two orders of magnitude for complicated repeated structure model. The available ability of tallying function for JMCT makes firm foundation for reactor analysis and multi-step burnup calculation. (authors)

DXRaySMCS: a user-friendly interface developed for prediction of diagnostic radiology X-ray spectra produced by Monte Carlo (MCNP-4C) simulation.

Science.gov (United States)

Bahreyni Toossi, M T; Moradi, H; Zare, H

2008-01-01

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. The electron's path in the target was followed until its 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: (i) modifying the MCNP input file, (ii) launching the MCNP program to simulate electron and photon transport and (iii) 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, output 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.

Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

International Nuclear Information System (INIS)

White, Morgan C.

2000-01-01

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 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

Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

Energy Technology Data Exchange (ETDEWEB)

White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)

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

Importance sampling techniques and treatment of electron transport in MCNP 4A

International Nuclear Information System (INIS)

Ueki, K.

1994-01-01

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)

Evaluation of a 50-MV photon therapy beam from a racetrack microtron using MCNP4B Monte Carlo code

International Nuclear Information System (INIS)

Gudowska, I.; Svensson, R.

2001-01-01

High energy photon therapy beam from the 50 MV racetrack microtron has been evaluated using the Monte Carlo code MCNP4B. The spatial and energy distribution of photons, radial and depth dose distributions in the phantom are calculated for the stationary and scanned photon beams from different targets. The calculated dose distributions are compared to the experimental data using a silicon diode detector. Measured and calculated depth-dose distributions are in fairly good agreement, within 2-3% for the positions in the range 2-30 cm in the phantom, whereas the larger discrepancies up to 10% are observed in the dose build-up region. For the stationary beams the differences in the calculated and measured radial dose distributions are about 2-10%. (orig.)

Criticality safety analysis of spent fuel storage for NPP Mochovce using MCNP5

International Nuclear Information System (INIS)

Farkas, G.; Hascik, J.; Lueley, J.; Vrban, B.; Petriska, M.; Slugen, V.; Urban, P.

2011-01-01

The paper presents results of nuclear criticality safety analysis of spent fuel storage for the first and second unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of k eff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal k eff values. The requirement of current safety regulations to ensure 5% subcriticality was met except one especially conservative case. (Authors)

Comparisons between MCNP, EGS4 and experiment for clinical electron beams.

Science.gov (United States)

Jeraj, R; Keall, P J; Ostwald, P M

1999-03-01

Understanding the limitations of Monte Carlo codes is essential in order to avoid systematic errors in simulations, and to suggest further improvement of the codes. MCNP and EGS4, Monte Carlo codes commonly used in medical physics, were compared and evaluated against electron depth dose data and experimental backscatter results obtained using clinical radiotherapy beams. Different physical models and algorithms used in the codes give significantly different depth dose curves and electron backscattering factors. The default version of MCNP calculates electron depth dose curves which are too penetrating. The MCNP results agree better with experiment if the ITS-style energy-indexing algorithm is used. EGS4 underpredicts electron backscattering for high-Z materials. The results slightly improve if optimal PRESTA-I parameters are used. MCNP simulates backscattering well even for high-Z materials. To conclude the comparison, a timing study was performed. EGS4 is generally faster than MCNP and use of a large number of scoring voxels dramatically slows down the MCNP calculation. However, use of a large number of geometry voxels in MCNP only slightly affects the speed of the calculation.

Modelling of the RA-1 reactor using a Monte Carlo code

International Nuclear Information System (INIS)

Quinteiro, Guillermo F.; Calabrese, Carlos R.

2000-01-01

It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

Physics and Algorithm Enhancements for a Validated MCNP/X Monte Carlo Simulation Tool, Phase VII

International Nuclear Information System (INIS)

McKinney, Gregg W.

2012-01-01

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.

Semi-Analytical Benchmarks for MCNP6

Energy Technology Data Exchange (ETDEWEB)

Grechanuk, Pavel Aleksandrovi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-11-07

Code verification is an extremely important process that involves proving or disproving the validity of code algorithms by comparing them against analytical results of the underlying physics or mathematical theory on which the code is based. Monte Carlo codes such as MCNP6 must undergo verification and testing upon every release to ensure that the codes are properly simulating nature. Specifically, MCNP6 has multiple sets of problems with known analytic solutions that are used for code verification. Monte Carlo codes primarily specify either current boundary sources or a volumetric fixed source, either of which can be very complicated functions of space, energy, direction and time. Thus, most of the challenges with modeling analytic benchmark problems in Monte Carlo codes come from identifying the correct source definition to properly simulate the correct boundary conditions. The problems included in this suite all deal with mono-energetic neutron transport without energy loss, in a homogeneous material. The variables that differ between the problems are source type (isotropic/beam), medium dimensionality (infinite/semi-infinite), etc.

Practical Application of Monte Carlo Code in RTP

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Julia Abdul Karim; Muhammad Rawi Mohamed Zin; Na'im Syauqi Hamzah; Mark Dennis Anak Usang; Abi Muttaqin Jalal Bayar; Muhammad Khairul Ariff Mustafa

2015-01-01

Monte Carlo neutron transport codes are widely used in various reactor physics applications in RTP and other related nuclear and radiation research in Nuklear Malaysia. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. This paper presents several calculation activities, its achievements and challenges in using MCNP code for neutronics analysis, nuclide inventory and source term calculation, shielding and dose evaluation. (author)

A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J. Eduard, E-mail: J.E.Hoogenboom@tudelft.nl [Delft University of Technology (Netherlands); Ivanov, Aleksandar; Sanchez, Victor, E-mail: Aleksandar.Ivanov@kit.edu, E-mail: Victor.Sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Diop, Cheikh, E-mail: Cheikh.Diop@cea.fr [CEA/DEN/DANS/DM2S/SERMA, Commissariat a l' Energie Atomique, Gif-sur-Yvette (France)

2011-07-01

A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

International Nuclear Information System (INIS)

Hoogenboom, J. Eduard; Ivanov, Aleksandar; Sanchez, Victor; Diop, Cheikh

2011-01-01

A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

Criticality calculations with MCNP trademark: A primer

International Nuclear Information System (INIS)

Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

1994-01-01

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

Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code

International Nuclear Information System (INIS)

Rafhayudi Jamro; Redzuwan Yahaya; Abdul Aziz Mohamed; Eid Abdel-Munem; Megat Harun Al-Rashid; Julia Abdul Karim; Ikki Kurniawan; Hafizal Yazid; Azraf Azman; Shukri Mohd

2008-01-01

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 (k e ff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of k e ff 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 k e ff from MCNP5 were obtained when the control fuel rods were fully inserted, the k e ff 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 k e ff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)

Validation suite for MCNP

International Nuclear Information System (INIS)

Mosteller, 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.

Comparisons between MCNP, EGS4 and experiment for clinical electron beams

International Nuclear Information System (INIS)

Jeraj, R.; Keall, P.J.; Ostwald, P.M.

1999-01-01

Understanding the limitations of Monte Carlo codes is essential in order to avoid systematic errors in simulations, and to suggest further improvement of the codes. MCNP and EGS4, Monte Carlo codes commonly used in medical physics, were compared and evaluated against electron depth dose data and experimental backscatter results obtained using clinical radiotherapy beams. Different physical models and algorithms used in the codes give significantly different depth dose curves and electron backscattering factors. The default version of MCNP calculates electron depth dose curves which are too penetrating. The MCNP results agree better with experiment if the ITS-style energy-indexing algorithm is used. EGS4 underpredicts electron backscattering for high- Z materials. The results slightly improve if optimal PRESTA-I parameters are used. MCNP simulates backscattering well even for high- Z materials. To conclude the comparison, a timing study was performed. EGS4 is generally faster than MCNP and use of a large number of scoring voxels dramatically slows down the MCNP calculation. However, use of a large number of geometry voxels in MCNP only slightly affects the speed of the calculation. (author)

Calibration curves of a PGNAA system for cement raw material analysis using the MCNP code

International Nuclear Information System (INIS)

Oliveira, Carlos; Salgado, Jose

1998-01-01

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

Using MCNP code for neutron and photon skyshine analysis

Energy Technology Data Exchange (ETDEWEB)

Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Netecha, M.E. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2000-03-01

The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector's dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field. (author)

Lecture note on neutron and photon transport calculation with MCNP

International Nuclear Information System (INIS)

Sakurai, Kiyoshi

2003-01-01

This paper is a lecture note on the continuous energy Monte Carlo method. The contents are as follows; history of the Monte Carlo study, continuous energy Monte Carlo codes, libraries, evaluation method for calculation results, integral emergent particle density equation, pseudorandom number, random walk, variance reduction techniques, MCNP weight window method, MCNP weight window generator, exponential transform, estimators, criticality problem and research subjects. This paper is a textbook for beginners on the Monte Carlo calculation. (author)

MCNP4C2, Coupled Neutron, Electron Gamma 3-D Time-Dependent Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

2002-01-01

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. MCNP4C2 is an interim release of MCNP4C with distribution restricted to the Criticality Safety community and attendees of the LANL MCNP workshops. The major new features of MCNP4C2 include: - Photonuclear physics; - Interactive plotting; - Plot superimposed weight window mesh; - Implement remaining macro-body surfaces; - Upgrade macro-bodies to surface sources and other capabilities; - Revised summary tables; - Weight window improvements. See the MCNP home page more information http://www-xdiv.lanl.gov/XCI/PROJECTS/MCNP with a link to the MCNP Forum. See the Electronic Notebook at http://www-rsicc.ornl.gov/rsic.html for information on user experiences with MCNP. 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. Pointwise continuous-energy cross section data are used, although multigroup data may also be used. Fixed-source adjoint calculations may be made with the multigroup data option. 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

Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

Science.gov (United States)

El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

2008-09-01

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

Development and validation of ALEPH Monte Carlo burn-up code

International Nuclear Information System (INIS)

Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.

2011-01-01

The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)

Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

International Nuclear Information System (INIS)

Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.

2007-01-01

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 [fr

Characteristics of Multihole Collimator Gamma Camera Simulation Modeled Using MCNP5

International Nuclear Information System (INIS)

Saripan, M. I.; Mashohor, S.; Adnan, W. A. Wan; Marhaban, M. H.; Hashim, S.

2008-01-01

This paper describes the characteristics of the multihole collimator gamma camera that is simulated using the combination of the Monte Carlo N-Particles Code (MCNP) version 5 and in-house software. The model is constructed based on the GCA-7100A Toshiba Gamma Camera at the Royal Surrey County Hospital, Guildford, Surrey, UK. The characteristics are analyzed based on the spatial resolution of the images detected by the Sodium Iodide (NaI) detector. The result is recorded in a list-mode file referred to as a PTRAC file within MCNP5. All pertinent nuclear reaction mechanisms, such as Compton and Rayleigh scattering and photoelectric absorption are undertaken by MCNP5 for all materials encountered by each photon. The experiments were conducted on Tl-201, Co-57, Tc-99 m and Cr-51 radio nuclides. The comparison of full width half maximum value of each datasets obtained from experimental work, simulation and literature are also reported in this paper. The relationship of the simulated data is in agreement with the experimental results and data obtained in the literature. A careful inspection at each of the data points of the spatial resolution of Tc-99 m shows a slight discrepancy between these sets. However, the difference is very insignificant, i.e. less than 3 mm only, which corresponds to a size of less than 1 pixel only (of the segmented detector)

Gamma streaming experiments for validation of Monte Carlo code

International Nuclear Information System (INIS)

Thilagam, L.; Mohapatra, D.K.; Subbaiah, K.V.; Iliyas Lone, M.; Balasubramaniyan, V.

2012-01-01

In-homogeneities in shield structures lead to considerable amount of leakage radiation (streaming) increasing the radiation levels in accessible areas. Development works on experimental as well as computational methods for quantifying this streaming radiation are still continuing. Monte Carlo based radiation transport code, MCNP is usually a tool for modeling and analyzing such problems involving complex geometries. In order to validate this computational method for streaming analysis, it is necessary to carry out some experimental measurements simulating these inhomogeneities like ducts and voids present in the bulk shields for typical cases. The data thus generated will be analysed by simulating the experimental set up employing MCNP code and optimized input parameters for the code in finding solutions for similar radiation streaming problems will be formulated. Comparison of experimental data obtained from radiation streaming experiments through ducts will give a set of thumb rules and analytical fits for total radiation dose rates within and outside the duct. The present study highlights the validation of MCNP code through the gamma streaming experiments carried out with the ducts of various shapes and dimensions. Over all, the present study throws light on suitability of MCNP code for the analysis of gamma radiation streaming problems for all duct configurations considered. In the present study, only dose rate comparisons have been made. Studies on spectral comparison of streaming radiation are in process. Also, it is planned to repeat the experiments with various shield materials. Since the penetrations and ducts through bulk shields are unavoidable in an operating nuclear facility the results on this kind of radiation streaming simulations and experiments will be very useful in the shield structure optimization without compromising the radiation safety

Modelling of the RA-1 reactor using a Monte Carlo code; Modelado del reactor RA-1 utilizando un codigo Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Quinteiro, Guillermo F; Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Reactores y Centrales Nucleares

2000-07-01

It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

Modeling the PUSPATI TRIGA Reactor using MCNP code

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Mark Dennis Usang; Naim Syauqi Hamzah; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

2012-01-01

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)

Monte Carlo Simulation of Electron Beams for Radiotherapy - EGS4, MCNP4b and GEANT3 Intercomparison

CERN Document Server

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.

Calculation of age-dependent effective doses for external exposure using the MCNP code

International Nuclear Information System (INIS)

Hung, Tran Van

2013-01-01

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.)

Calculation of age-dependent effective doses for external exposure using the MCNP code

Energy Technology Data Exchange (ETDEWEB)

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.)

Calculation of power density with MCNP in TRIGA reactor

International Nuclear Information System (INIS)

Snoj, L.; Ravnik, M.

2006-01-01

Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

Study of bremsstrahlung photons in bulk target using MCNP code

Directory of Open Access Journals (Sweden)

S. Sangaroon

2017-11-01

Full Text Available The aim of this research was to study the feasibility of bremsstrahlung photon production in target bombarded by 1 GeV electrons. The calculations were performed by the Monte Carlo code MCNP. Six target materials with densities between 2 and 20 g/cm3 were studied. The bremsstrahlung photon flux is high for the target density above 8 g/cm3. Copper is the best target for 1 GeV electron beam due to high bremsstrahlung photon production, low scattering and low transmission electron flux. The copper target was altered to have different thicknesses between 0.01 and 2.5 cm. The results showed that the bremsstrahlung photon flux significantly increased when the target thickness increased from 0.01 to 1.5 cm. The angular distribution of the bremsstrahlung photons with angles between 0 and 120 degrees was determined for copper target. The maximum angle of the photon scattering was about 20 degree.

Monte Carlo dose calculations of beta-emitting sources for intravascular brachytherapy: a comparison between EGS4, EGSnrc, and MCNP.

Science.gov (United States)

Wang, R; Li, X A

2001-02-01

The dose parameters for the beta-particle emitting 90Sr/90Y source for intravascular brachytherapy (IVBT) have been calculated by different investigators. At a distant distance from the source, noticeable differences are seen in these parameters calculated using different Monte Carlo codes. The purpose of this work is to quantify as well as to understand these differences. We have compared a series of calculations using an EGS4, an EGSnrc, and the MCNP Monte Carlo codes. Data calculated and compared include the depth dose curve for a broad parallel beam of electrons, and radial dose distributions for point electron sources (monoenergetic or polyenergetic) and for a real 90Sr/90Y source. For the 90Sr/90Y source, the doses at the reference position (2 mm radial distance) calculated by the three code agree within 2%. However, the differences between the dose calculated by the three codes can be over 20% in the radial distance range interested in IVBT. The difference increases with radial distance from source, and reaches 30% at the tail of dose curve. These differences may be partially attributed to the different multiple scattering theories and Monte Carlo models for electron transport adopted in these three codes. Doses calculated by the EGSnrc code are more accurate than those by the EGS4. The two calculations agree within 5% for radial distance <6 mm.

Installation and validation of MCNP-4A

International Nuclear Information System (INIS)

Marks, N.A.

1997-01-01

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)

New developments enhancing MCNP for criticality safety

International Nuclear Information System (INIS)

Hendricks, J.S.; McKinney, G.W.; Forster, R.A.

1993-01-01

Since the early 80's MCNP has had three estimates of k eff : 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

Application of SN and Monte Carlo codes to the SHEBA critical assemblies

International Nuclear Information System (INIS)

O'Dell, R.D.

1993-01-01

The Solution High-Energy Burst Assembly (SHEBA) at Los Alamos is a low-enriched (4.95 wt. %) aqueous uranyl fluoride solution critical assembly. There are two SHEBA configurations, both consisting of right circular cylinders with a central control rod. The first configuration, hereafter called the old SHEBA, had a fuel solution diameter of 54.6 cm and a measured critical solution height of 36.5 cm. An improved modification, hereafter called the new SHEBA, has a fuel solution diameter of 48.9 cm but since it is not yet operational, the critical solution height has not yet been measured. In this presentation the application of the discrete-ordinates (S N ) code TWODANT using Hansen-Roach cross sections and the MCNP Monte Carlo code using continuous-energy cross sections for calculating the critical solution heights for both the old and new SHEBA assemblies is described. The code's predictions are compared and it is shown that a single calculation with a standard computer code may yield misleading results, especially when using a Monte Carlo code

Application of dose evaluation of the MCNP code for interim spent fuel cask storage facility

International Nuclear Information System (INIS)

Kosako, Toshiso; Iimoto, Takeshi; Ishikawa, Satoshi; Tsuboi, Takafumi; Teramura, Masahiro; Okamura, Tomomi; Narumiya, Yoshiyuki

2007-01-01

The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results. (author)

A DRAGON-MCNP comparison of void reactivity calculations

Energy Technology Data Exchange (ETDEWEB)

Marleau, G [Ecole Polytechnique, Montreal, PQ (Canada). Inst. de Genie Nucleaire; Milgram, M S [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs.

A DRAGON-MCNP comparison of void reactivity calculations

International Nuclear Information System (INIS)

Marleau, G.

1995-01-01

The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs

Application of Monte-Carlo Code to dose distribution calculation in a case of lung cancer by the emitted photon beams from linear accelerator

International Nuclear Information System (INIS)

Le Thanh Xuan; Nguyen Thi Cam Thu; Tran Van Nghia; Truong Thi Hong Loan; Vo Thanh Nhon

2015-01-01

The dose distribution calculation is one of the major steps in radiotherapy. In this paper the Monte Carlo code MCNP5 has been applied for simulation 15 MV photon beams emitted from linear accelerator in a case of lung cancer of the General Hospital of Kien Giang. The settings for beam directions, field sizes and isocenter position used in MCNP5 must be the same as those in treatment plan at the hospital to ensure the results from MCNP5 are accurate. We also built a program CODIM by using MATLAB® programming software. This program was used to construct patient model from lung CT images obtained from cancer treatment cases at the General Hospital of Kien Giang and then MCNP5 code was used to simulate the delivered dose in the patient. The results from MCNP5 show that there is a difference of 5% in comparison with Prowess Panther program - a semi-empirical simulation program which is being used for treatment planning in the General Hospital of Kien Giang. The success of the work will help the planners to verify the patient dose distribution calculated from the treatment planning program being used at the hospital. (author)

Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

Li Fei; Ge Liangquan; Zhang Qingxian

2010-01-01

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)

MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System

International Nuclear Information System (INIS)

2007-01-01

A - Description of program or function: MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC, nor from the NEA Databank. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included. The following changes have been made: 1) An increase in the number of removal group information that must be provided for each material in each step in the feed input file. 2) The capability to use CINDER90 instead of ORIGEN2.1 as the depletion/decay part of the code. 3) ORIGEN2.2 can also be used instead of ORIGEN2.1 in Monteburns. 4) The correction of including the capture cross sections to metastable as well as ground states if applicable for an isotope (i.e. Am-241 and Am-243 in particular). 5) The ability to use a MCNP input file that has a title card starting with 'm' (this was a bug in the first version of Monteburns). 6) A decrease in run time for cases involving decay-only steps (power of 0.0). Monteburns does not run MCNP to calculate cross sections for a step unless it is an irradiation step. 7) The ability to change the cross section libraries used each step. If different cross section libraries are desired for multiple steps. 8

Low Enrichment Uranium (LEU)-fueled SLOWPOKE-2 nuclear reactor simulation with the Monte-Carlo based MCNP 4A code

International Nuclear Information System (INIS)

Pierre, J.R.M.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.

Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.

2016-09-01

The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

Evaluation of dose equivalent to the people accompanying patients in diagnostic radiology using MCNP4C Monte Carlo code

International Nuclear Information System (INIS)

Mehdizadeh, S.; Faghihi, R.; Sina, S.; Zehtabian, M.

2007-01-01

Complete text of publication follows. Objective: X rays used in diagnostic radiology contribute a major share to population doses from man-made sources of radiation. In some branches of radiology, it is necessary that another person stay in the imaging room and immobilize the patient to carry out radiological operation. ICRP 70 recommends that this should be done by parents or accompanying nursing or ancillary personnel and not in any case by radiation workers. Methods: Dose measurements were made previously using standard methods employing LiF TLD-100 dosimeters. A TLD card was installed on the main trunk of the body of the accompanying people where the maximum dose was probable. In this research the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) is used to calculate the equivalent dose to the people accompanying patients exposed to radiation scattered from the patient (Without protective clothing). To do the simulations, all components of the geometry are placed within an air-filled box. Two homogeneous water phantoms are used to simulate the patient and the accompanying person. The accompanying person leans against the table at one side of the patient. Finally in case of source specification, only the focus of the X-ray tube is modelled, i.e. as a standard MCNP point source emitting a cone of photons. Photon stopping material is used as a collimator model to reduce the circular cross section of the cone to a rectangle. The X-ray spectra to be used in the MCNP simulations are generated with spectrum generator software, taking the X-ray voltage and all filtration applied in the clinic as input parameters. These calculations are done for different patient sizes and for different radiological operations. Results: In case of TL dosimetry, for a group of 100 examinations, the dose equivalents ranged from 0.01 μsv to 0.13 msv with the average of 0.05 msv. The results are seen to be in close agreement with Monte Carlo simulations

DXRaySMCS. First user friendly interface developed for prediction of diagnostic radiology X-ray spectra produced by Monte Carlo (MCNP-4C) simulation in Iran

International Nuclear Information System (INIS)

Bahreyni Toossi, M.T.; Zare, H.; Moradi Faradanbe, H.

2008-01-01

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)

Neutron-induced photon production in MCNP

International Nuclear Information System (INIS)

Little, R.C.; Seamon, R.E.

1983-01-01

An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems

Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2014-01-01

The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

2014-07-01

The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

CREOLE experiment study on the reactivity temperature coefficient with sensitivity and uncertainty analysis using the MCNP5 code and different neutron cross section evaluations

International Nuclear Information System (INIS)

Boulaich, Y.; El Bardouni, T.; Erradi, L.; Chakir, E.; Boukhal, H.; Nacir, B.; El Younoussi, C.; El Bakkari, B.; Merroun, O.; Zoubair, M.

2011-01-01

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 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 2 and UO 2 -PuO 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 discrepancy is

Monte Carlo simulation using MCNP4B for an optimal shielding design of a 252 Cf source

International Nuclear Information System (INIS)

Silva, Ademir X. da; Crispim, Verginia R.

2001-01-01

This study aim to investigate an optimum shielding design against neutrons and gamma-rays from a source of 252 Cf, using Monte Carlo simulation. The shielding materials studied were: borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP, version 4B, was used to design shielding for 252 Cf based neutron irradiator systems. By normalizing the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independents of the intensity of actual 252 Cf source. The results shown what the total dose equivalent rates were reduced significantly by the shielding system optimization. (author)

Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X and NJOY2016

Directory of Open Access Journals (Sweden)

Kabach Ouadie

2017-12-01

Full Text Available To validate the new Evaluated Nuclear Data File (ENDF/B-VIII.0β4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016. The results obtained with the ENDF/B-VIII.0β4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X code. All the MCNP(X calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-11

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,β) 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 – C_k's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usage are discussed.

MCB. A continuous energy Monte Carlo burnup simulation code

International Nuclear Information System (INIS)

Cetnar, J.; Wallenius, J.; Gudowski, W.

1999-01-01

A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (author)

Modification to the Monte N-Particle (MCNP) Visual Editor (MCNPVised) to read in Computer Aided Design (CAD) files

International Nuclear Information System (INIS)

Schwarz, Randy A.; Carter, Leeland L.

2004-01-01

Monte Carlo N-Particle Transport Code (MCNP) (Reference 1) 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 (References 2 to 11) is recognized internationally as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant enhanced the capabilities of the MCNP Visual Editor to allow it to read in a 2D Computer Aided Design (CAD) file, allowing the user to modify and view the 2D CAD file and then electronically generate a valid MCNP input geometry with a user specified axial extent

Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes

International Nuclear Information System (INIS)

Trgina, M.

1993-12-01

The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs

Whole core burnup calculations using `MCNP`

Energy Technology Data Exchange (ETDEWEB)

Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

1996-12-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

Whole core burnup calculations using 'MCNP'

International Nuclear Information System (INIS)

Haran, O.; Shaham, Y.

1996-01-01

Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

Calculation of the effective dose from natural radioactivity in soil using MCNP code.

Science.gov (United States)

Krstic, D; Nikezic, D

2010-01-01

Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this work. Calculations have been done for the most common natural radionuclides in soil (238)U, (232)Th series and (40)K. A ORNL human phantoms and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs. The effective dose was calculated according to ICRP 74 recommendations. Conversion factors of effective dose per air kerma were determined. Results obtained here were compared with other authors. Copyright 2009 Elsevier Ltd. All rights reserved.

Simulation of density curve for slim borehole using the Monte Carlo code MCNPX

International Nuclear Information System (INIS)

Souza, Edmilson Monteiro de; Silva, Ademir Xavier da; Lopes, Ricardo Tadeu; Lima, Inaya C.B.; Rocha, Paula L.F.

2010-01-01

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 R , and a 100 mCi Cs-137 gamma source has been modeled with the new version of MCNP code MCNPX. (author)

Simulation of density curve for slim borehole using the Monte Carlo code MCNPX

Energy Technology Data Exchange (ETDEWEB)

Souza, Edmilson Monteiro de; Silva, Ademir Xavier da; Lopes, Ricardo Tadeu, E-mail: emonteiro@nuclear.ufrj.b, E-mail: ademir@nuclear.ufrj.b, E-mail: ricardo@lin.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Correa, Samanda Cristine Arruda, E-mail: scorrea@nuclear.ufrj.b [Centro Universitario Estadual da Zona Oeste (CCMAT/UEZO), Rio de Janeiro, RJ (Brazil); Lima, Inaya C.B., E-mail: inaya@lin.ufrj.b [Universidade Estadual do Rio de Janeiro (IPRJ/UERJ) Nova Friburgo, Rio de Janeiro, RJ (Brazil). Instituto Politecnico do Rio de Janeiro; Rocha, Paula L.F., E-mail: ferrucio@acd.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ) RJ (Brazil). Dept. de Geologia

2010-07-01

Borehole logging for formation density has been an important geophysical measurement in oil industry. For calibration of the Gamma Ray nuclear logging tool, numerous rock models of different lithology and densities are necessary. However, the full success of this calibration process is determined by a reliable benchmark, where the complete and precise chemical composition of the standards is necessary. Simulations using the Monte Carlo MCNP have been widely employed in well logging application once it serves as a low-cost substitute for experimental test pits, as well as a means for obtaining data that are difficult to obtain experimentally. Considering this, the purpose of this work is to use the code MCNP to obtain density curves for slim boreholes using Gamma Ray logging tools. For this, a Slim Density Gamma Probe, named TRISOND{sup R}, and a 100 mCi Cs-137 gamma source has been modeled with the new version of MCNP code MCNPX. (author)

Analysis of the TRIGA Mark-II benchmark IEU-COMP-THERM-003 with Monte Carlo code MVP

International Nuclear Information System (INIS)

Mahmood, Mohammad Sayem; Nagaya, Yasunobu; Mori, Takamasa

2004-03-01

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 k eff 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)

Parallelization of MCNP4 code by using simple FORTRAN algorithms

International Nuclear Information System (INIS)

Yazid, P.I.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka.

1993-12-01

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)

Application of MCNP in the criticality calculation for reactors

International Nuclear Information System (INIS)

Zhong Zhaopeng; Shi Gong; Hu Yongming

2003-01-01

The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

Benchmark of physics design of a proposed 30 MW Multi Purpose Research Reactor using a Monte Carlo code MCNP

International Nuclear Information System (INIS)

Singh, Tej; Kumar, Jainendra; Sharma, Archana; Singh, Kanchhi; Raina, V.K.; Srinivasan, P.

2009-01-01

At present Dhruva and Cirus reactors provide majority of research reactor based experimental/irradiation facilities to cater to various needs of the vast pool of researchers in the field of sciences research and development work for nuclear power plants and production of radioisotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 30 MWt Multi Purpose Research Reactor is proposed to be constructed. This paper describes some of the physics design features of this reactor using MCNP code to validate the deterministic methods. The criticality calculations for 100 material testing reactor (JHR) of France and 610 MW SAVANNAH thermal reactor were performed using MCNP computer codes to boost the confidence level in designing the physics design of reactor core. (author)

Absorbed fractions in a voxel-based phantom calculated with the MCNP-4B code.

Science.gov (United States)

Yoriyaz, H; dos Santos, A; Stabin, M G; Cabezas, R

2000-07-01

A new approach for calculating internal dose estimates was developed through the use of a more realistic computational model of the human body. The present technique shows the capability to build a patient-specific phantom with tomography data (a voxel-based phantom) for the simulation of radiation transport and energy deposition using Monte Carlo methods such as in the MCNP-4B code. MCNP-4B absorbed fractions for photons in the mathematical phantom of Snyder et al. agreed well with reference values. Results obtained through radiation transport simulation in the voxel-based phantom, in general, agreed well with reference values. Considerable discrepancies, however, were found in some cases due to two major causes: differences in the organ masses between the phantoms and the occurrence of organ overlap in the voxel-based phantom, which is not considered in the mathematical phantom.

Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project

International Nuclear Information System (INIS)

Campolina, Daniel de Almeida Magalhaes

2009-01-01

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)

Monte Carlo simulation of x-ray spectra in diagnostic radiology and mammography using MCNP4C

Energy Technology Data Exchange (ETDEWEB)

Ay, M R [Department of Physics and Nuclear Sciences, AmirKabir University of Technology, Tehran (Iran, Islamic Republic of); Shahriari, M [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Sarkar, S [Department of Medical Physics, Tehran University of Medical Science, Tehran (Iran, Islamic Republic of); Adib, M [TPP Co., GE Medical Systems, Iran Authorized Distributor, Tehran (Iran, Islamic Republic of); Zaidi, H [Division of Nuclear Medicine, Geneva University Hospital, 1211 Geneva (Switzerland)

2004-11-07

The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combinations to investigate the effect of tube voltage, target material and filter thickness on x-ray spectra in the diagnostic radiology and mammography energy ranges. The simulated x-ray spectra were compared with experimental measurements and spectra calculated by IPEM report number 78. In addition, the anode heel effect and off-axis x-ray spectra were assessed for different anode angles and target materials and the results were compared with EGS4-based Monte Carlo simulations and measured data. Quantitative evaluation of the differences between our Monte Carlo simulated and comparison spectra was performed using student's t-test statistical analysis. Generally, there is a good agreement between the simulated x-ray and comparison spectra, although there are systematic differences between the simulated and reference spectra especially in the K-characteristic x-rays intensity. Nevertheless, no statistically significant differences have been observed between IPEM spectra and the simulated spectra. It has been shown that the difference between MCNP simulated spectra and IPEM spectra in the low energy range is the result of the overestimation of characteristic photons following the normalization procedure. The transmission curves produced by MCNP4C have good agreement with the IPEM report especially for tube voltages of 50 kV and 80 kV. The systematic discrepancy for higher tube voltages is the result of systematic differences between the corresponding spectra.

Monte Carlo simulation of x-ray spectra in diagnostic radiology and mammography using MCNP4C

Science.gov (United States)

Ay, M. R.; Shahriari, M.; Sarkar, S.; Adib, M.; Zaidi, H.

2004-11-01

The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combinations to investigate the effect of tube voltage, target material and filter thickness on x-ray spectra in the diagnostic radiology and mammography energy ranges. The simulated x-ray spectra were compared with experimental measurements and spectra calculated by IPEM report number 78. In addition, the anode heel effect and off-axis x-ray spectra were assessed for different anode angles and target materials and the results were compared with EGS4-based Monte Carlo simulations and measured data. Quantitative evaluation of the differences between our Monte Carlo simulated and comparison spectra was performed using student's t-test statistical analysis. Generally, there is a good agreement between the simulated x-ray and comparison spectra, although there are systematic differences between the simulated and reference spectra especially in the K-characteristic x-rays intensity. Nevertheless, no statistically significant differences have been observed between IPEM spectra and the simulated spectra. It has been shown that the difference between MCNP simulated spectra and IPEM spectra in the low energy range is the result of the overestimation of characteristic photons following the normalization procedure. The transmission curves produced by MCNP4C have good agreement with the IPEM report especially for tube voltages of 50 kV and 80 kV. The systematic discrepancy for higher tube voltages is the result of systematic differences between the corresponding spectra.

Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages

International Nuclear Information System (INIS)

Matijevic, M.; Jecmenica, R.; Pevec, D.; Trontl, K.

2012-01-01

The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals of the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working enviroment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 code system, with the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analysed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be

Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes

CERN Document Server

Mainardi, E; Donahue, R J

2002-01-01

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 ...

Accuracy assessment of a new Monte Carlo based burnup computer code

International Nuclear Information System (INIS)

El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.

2012-01-01

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.

C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

Energy Technology Data Exchange (ETDEWEB)

Yesilyurt, Gokhan [ORNL; Clarno, Kevin T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL; Fox, Patricia B [ORNL

2011-01-01

The C5 benchmark problem proposed by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency was modeled to examine the capabilities of Denovo, a three-dimensional (3-D) parallel discrete ordinates (S{sub N}) radiation transport code, for problems with no spatial homogenization. Denovo uses state-of-the-art numerical methods to obtain accurate solutions to the Boltzmann transport equation. Problems were run in parallel on Jaguar, a high-performance supercomputer located at Oak Ridge National Laboratory. Both the two-dimensional (2-D) and 3-D configurations were analyzed, and the results were compared with the reference MCNP Monte Carlo calculations. For an additional comparison, SCALE/KENO-V.a Monte Carlo solutions were also included. In addition, a sensitivity analysis was performed for the optimal angular quadrature and mesh resolution for both the 2-D and 3-D infinite lattices of UO{sub 2} fuel pin cells. Denovo was verified with the C5 problem. The effective multiplication factors, pin powers, and assembly powers were found to be in good agreement with the reference MCNP and SCALE/KENO-V.a Monte Carlo calculations.

MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

Energy Technology Data Exchange (ETDEWEB)

Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-05-22

MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 using CEM03.03 and LAQGSM03.03 describes well fragmentation reactions induced on light and medium target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below, and can serve as a reliable simulation tool for different applications, like cosmic-ray-induced single event upsets (SEU’s), radiation protection, and cancer therapy with proton and ion beams, to name just a few. Future improvements of the predicting capabilities of MCNP6 for such reactions are possible, and are discussed in this work.

Radiation shielding calculation using MCNP

International Nuclear Information System (INIS)

Masukawa, Fumihiro

2001-01-01

To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

Calculation of the effective dose from natural radioactivity sources in soil using MCNP code

International Nuclear Information System (INIS)

Krstic, D.; Nikezic, D.

2008-01-01

Full text: Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this report. Calculations have been done for the most common natural radionuclides in soil as 238 U, 232 Th series and 40 K. A ORNL age-dependent phantom and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs of phantom.The effective dose was calculated according to ICRP74 recommendations. Conversion coefficients of effective dose per air kerma were determined. Results obtained here were compared with other authors

Radiation calculations using LAHET/MCNP/CINDER90

International Nuclear Information System (INIS)

Waters, L.

1994-01-01

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

Calculation of the X-Ray Spectrum of a Mammography System with Various Voltages and Different Anode-Filter Combinations Using MCNP Code

Directory of Open Access Journals (Sweden)

Lida Gholamkar

2016-09-01

Full Text Available Introduction One of the best methods in the diagnosis and control of breast cancer is mammography. The importance of mammography is directly related to its value in the detection of breast cancer in the early stages, which leads to a more effective treatment. The purpose of this article was to calculate the X-ray spectrum in a mammography system with Monte Carlo codes, including MCNPX and MCNP5. Materials and Methods The device, simulated using the MCNP code, was Planmed Nuance digital mammography device (Planmed Oy, Finland, equipped with an amorphous selenium detector. Different anode/filter materials, such as molybdenum-rhodium (Mo-Rh, molybdenum-molybdenum (Mo-Mo, tungsten-tin (W-Sn, tungsten-silver (W-Ag, tungsten-palladium (W-Pd, tungsten-aluminum (W-Al, tungsten-molybdenum (W-Mo, molybdenum-aluminum (Mo-Al, tungsten-rhodium (W-Rh, rhodium-aluminum (Rh-Al, and rhodium-rhodium (Rh-Rh, were simulated in this study. The voltage range of the X-ray tube was between 24 and 34 kV with a 2 kV interval. Results The charts of changing photon flux versus energy were plotted for different types of anode-filter combinations. The comparison with the findings reported by others indicated acceptable consistency. Also, the X-ray spectra, obtained from MCNP5 and MCNPX codes for W-Ag and W-Rh combinations, were compared. We compared the present results with the reported data of MCNP4C and IPEM report No. 78 for Mo-Mo, Mo-Rh, and W-Al combinations. Conclusion The MCNPX calculation outcomes showed acceptable results in a low-energy X-ray beam range (10-35 keV. The obtained simulated spectra for different anode/filter combinations were in good conformity with the finding of previous research.

Preliminary evaluation of pin power distribution for fuel assemblies of SMART by MCNP

International Nuclear Information System (INIS)

Kim, Kyo Youn

1998-08-01

Monte Carlo transport code MCNP can describe an object sophisticately by use of three-dimensional modelling and can adopt a continuous energy cross-section library. Therefore MCNP has been widely utilized in the field of radiation physics to estimate fluxes and dose rates for nuclear facilities and to review results from conventional methods such a as discrete ordinates method and point kernel method. The Monte Carlo method has recently been introduced to estimated the neutron multiplication factor and pin power distribution in the fuel assembly of a reactor core. The operating thermal power of SMART core is 330 MWt and there are 57 fuel assemblies in the core. In this study it was assumed that the core has 4 types of fuel assemblies. In this study, MCNP4a was used to perform to estimate criticality and normalized pin power distribution in a fuel assembly of SMART core. The results from MCNP4a calculations are able to be used review those from nuclear design/analysis code. It is very complicated to pick up interested data from MCNP output list and to normalize pin power distribution in a fuel assembly because MCNP is not only a nuclear design/analysis code. In this study a program FAPIN was developed to generated a generate a normalized pin power distribution from the MCNP output list. (author). 11 refs

Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

International Nuclear Information System (INIS)

Leppaenen, J.

2007-01-01

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)

The use of the MCNP code for the quantitative analysis of elements in geological formations

Energy Technology Data Exchange (ETDEWEB)

Cywicka-Jakiel, T.; Woynicka, U. [The Henryk Niewodniczanski Institute of Nuclear Physics, Krakow (Poland); Zorski, T. [University of Mining and Metallurgy, Faculty of Geology, Geophysics and Environmental Protection, Krakow (Poland)

2003-07-01

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)

The use of the MCNP code for the quantitative analysis of elements in geological formations

International Nuclear Information System (INIS)

Cywicka-Jakiel, T.; Woynicka, U.; Zorski, T.

2003-01-01

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)

Generating and verification of ACE-multigroup library for MCNP

International Nuclear Information System (INIS)

Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai

2012-01-01

The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)

Utilization of MCNP code in the research and design for China advanced research reactor

International Nuclear Information System (INIS)

Shen Feng

2006-01-01

MCNP, which is the internationalized neutronics code, is used for nuclear research and design in China Advanced Research Reactor (CARR). MCNP is an important neutronics code in the research and design for CARR since many calculation tasks could be undertaken by it. Many nuclear parameters on reactor core, the design and optimization research for many reactor utilizations, much verification for other nuclear calculation code and so on are conducted with help of MCNP. (author)

Evaluation of the WIMS (KAERI) - VENTURE code system for peak power prediction of KMRR core using MCNP

International Nuclear Information System (INIS)

Park, W.S.; Lee, K.M.; Lee, C.S.; Lee, J.T.; Oh, S.K.

1992-01-01

In this work, the validity and quantitative uncertainty of WIMS (KAERI) - VENTURE code system for the design and analysis of KMRR core was tried to be inferred using a well known benchmark code, MCNP. WIMS (KAERI) showed an excellent agreement with MCNP code. For three different control rod positions at a simulated core which has a quarter symmetry, total peaking factors and three sub-factors (radial, axial, and local) obtained from VENTURE were compared with those of MCNP. The comparison proved the validity of VENTURE and showed better agreement in the order of radial, axial, and local factors. The uncertainty of WIMS (KAERI) - VENTURE system was inferred using the 2σ band of total peaking obtained by MCNP. The uncertainty of WIMS (KAERI) - VENTURE system were found to be 18.5 % for the operating condition. (author)

NaI(Tl) detectors modeling in MCNP-X and Gate/Geant4 codes

Energy Technology Data Exchange (ETDEWEB)

Affonso, Renato Raoni Werneck; Silva, Ademir Xavier da, E-mail: raoniwa@yahoo.com.br, E-mail: ademir@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, Cesar Marques, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

NaI (Tl) detectors are widely used in gamma-ray densitometry, but their modeling in Monte Carlo codes, such as MCNP-X and Gate/Geant4, needs a lot of work and does not yield comparable results with experimental arrangements, possibly due to non-simulated physical phenomena, such as light transport within the scintillator. Therefore, it is necessary a methodology that positively impacts the results of the simulations while maintaining the real dimensions of the detectors and other objects to allow validating a modeling that matches up with the experimental arrangement. Thus, the objective of this paper is to present the studies conducted with the MCNPX and Gate/Geant4 codes, in which the comparisons of their results were satisfactory, showing that both can be used for the same purposes. (author)

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

International Nuclear Information System (INIS)

Tzika, F.; Stamatelatos, I.E.

2004-01-01

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

MCNPX{trademark} -- The LAHET{trademark}/MCNP{trademark} code merger

Energy Technology Data Exchange (ETDEWEB)

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.

Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

International Nuclear Information System (INIS)

Nagao, Y.; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H.

2000-01-01

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 6 Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high 6 Li 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 1-6x10 13 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

Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

Energy Technology Data Exchange (ETDEWEB)

Nagao, Y. E-mail: nagao@jmtr.oarai.jaeri.go.jp; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H

2000-11-01

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 {sup 6}Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high {sup 6}Li 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 1-6x10{sup 13} n cm{sup -2} per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2.

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium; Comparacion y validacion de los resultados del codigo AZNHEX v.1.0 con el codigo MCNP simulando el nucleo de un reactor rapido refrigerado con sodio

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Esquivel E, J., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

Feasibility of a Monte Carlo-deterministic hybrid method for fast reactor analysis

Energy Technology Data Exchange (ETDEWEB)

Heo, W.; Kim, W.; Kim, Y. [Korea Advanced Institute of Science and Technology - KAIST, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of); Yun, S. [Korea Atomic Energy Research Institute - KAERI, 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2013-07-01

A Monte Carlo and deterministic hybrid method is investigated for the analysis of fast reactors in this paper. Effective multi-group cross sections data are generated using a collision estimator in the MCNP5. A high order Legendre scattering cross section data generation module was added into the MCNP5 code. Both cross section data generated from MCNP5 and TRANSX/TWODANT using the homogeneous core model were compared, and were applied to DIF3D code for fast reactor core analysis of a 300 MWe SFR TRU burner core. For this analysis, 9 groups macroscopic-wise data was used. In this paper, a hybrid calculation MCNP5/DIF3D was used to analyze the core model. The cross section data was generated using MCNP5. The k{sub eff} and core power distribution were calculated using the 54 triangle FDM code DIF3D. A whole core calculation of the heterogeneous core model using the MCNP5 was selected as a reference. In terms of the k{sub eff}, 9-group MCNP5/DIF3D has a discrepancy of -154 pcm from the reference solution, 9-group TRANSX/TWODANT/DIF3D analysis gives -1070 pcm discrepancy. (authors)

Problem and solution of tally segment card in MCNP code

International Nuclear Information System (INIS)

Xie Jiachun; Zhao Shouzhi; Sun Zheng; Jia Baoshan

2010-01-01

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 dose-rate on/over the surface of cylindrical radio-models using Monte Carlo methods

International Nuclear Information System (INIS)

Xiao Xuefu; Ma Guoxue; Wen Fuping; Wang Zhongqi; Wang Chaohui; Zhang Jiyun; Huang Qingbo; Zhang Jiaqiu; Wang Xinxing; Wang Jun

2004-01-01

Objective: To determine the dose-rates on/over the surface of 10 cylindrical radio-models, which belong to the Metrology Station of Radio-Geological Survey of CNNC. Methods: The dose-rates on/over the surface of 10 cylindrical radio-models were modeled using the famous Monte Carlo code-MCNP. The dose-rates on/over the surface of 10 cylindrical radio-models were measured by a high gas pressurized ionization chamber dose-rate meter, respectively. The values of dose-rate modeled using MCNP code were compared with those obtained by authors in the present experimental measurement, and with those obtained by other workers previously. Some factors causing the discrepancy between the data obtained by authors using MCNP code and the data obtained using other methods are discussed in this paper. Results: The data of dose-rates on/over the surface of 10 cylindrical radio-models, obtained using MCNP code, were in good agreement with those obtained by other workers using the theoretical method. They were within the discrepancy of ±5% in general, and the maximum discrepancy was less than 10%. Conclusions: As if each factor needed for the Monte Carlo code is correct, the dose-rates on/over the surface of cylindrical radio-models modeled using the Monte Carlo code are correct with an uncertainty of 3%

MCNP analysis of the nine-cell LWR gadolinium benchmark

International Nuclear Information System (INIS)

Arkuszewski, J.J.

1988-01-01

The Monte Carlo results for a 9-cell fragment of the light water reactor square lattice with a central gadolinium-loaded pin are presented. The calculations are performed with the code MCNP-3A and the ENDF-B/5 library and compared with the results obtained from the BOXER code system and the JEF-1 library. The objective of this exercise is to study the feasibility of BOXER for the analysis of a Gd-loaded LWR lattice in the broader framework of GAP International Benchmark Analysis. A comparison of results indicates that, apart from unavoidable discrepancies originating from different data evaluations, the BOXER code overestimates the multiplication factor by 1.4 % and underestimates the power release in a Gd cell by 4.66 %. It is hoped that further similar studies with use of the JEF-1 library for both BOXER and MCNP will help to isolate and explain these discrepancies in a cleaner way. (author) 4 refs., 9 figs., 10 tabs

Modeling and Simulation Monte Carlo by the MCNP code for determining neutron parameters of the nuclear reactor-subcritical assembly in CNSTN

International Nuclear Information System (INIS)

Romdhani, Ibtissem

2014-01-01

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.

Development of automatic cross section compilation system for MCNP

International Nuclear Information System (INIS)

Maekawa, Fujio; Sakurai, Kiyoshi

1999-01-01

A development of a code system to automatically convert cross-sections for MCNP is in progress. The NJOY code is, in general, used to convert the data compiled in the ENDF format (Evaluated Nuclear Data Files by BNL) into the cross-section libraries required by various reactor physics codes. While the cross-section library: FSXLIB-J3R2 was already converted from the JENDL-3.2 version of Japanese Evaluated Nuclear Data Library for a continuous energy Monte Carlo code MCNP, the library keeps only the cross-sections at room temperature (300 K). According to the users requirements which want to have cross-sections at higher temperature, say 600 K or 900 K, a code system named 'autonj' is under development to provide a set of cross-section library of arbitrary temperature for the MCNP code. This system can accept any of data formats adopted JENDL that may not be treated by NJOY code. The input preparation that is repeatedly required at every nuclide on NJOY execution is greatly reduced by permitting the conversion process of as many nuclides as the user wants in one execution. A few MCNP runs were achieved for verification purpose by using two libraries FSXLIB-J3R2 and the output of autonj'. The almost identical MCNP results within the statistical errors show the 'autonj' output library is correct. In FY 1998, the system will be completed, and in FY 1999, the user's manual will be published. (K. Tsuchihashi)

Development of a coupling code for PWR reactor cavity radiation streaming calculation

International Nuclear Information System (INIS)

Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

2012-01-01

PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

Monte Carlo Techniques for Nuclear Systems - Theory Lectures

International Nuclear Information System (INIS)

Brown, Forrest B.; Univ. of New Mexico, Albuquerque, NM

2016-01-01

These are lecture notes for a Monte Carlo class given at the University of New Mexico. The following topics are covered: course information; nuclear eng. review & MC; random numbers and sampling; computational geometry; collision physics; tallies and statistics; eigenvalue calculations I; eigenvalue calculations II; eigenvalue calculations III; variance reduction; parallel Monte Carlo; parameter studies; fission matrix and higher eigenmodes; doppler broadening; Monte Carlo depletion; HTGR modeling; coupled MC and T/H calculations; fission energy deposition. Solving particle transport problems with the Monte Carlo method is simple - just simulate the particle behavior. The devil is in the details, however. These lectures provide a balanced approach to the theory and practice of Monte Carlo simulation codes. The first lectures provide an overview of Monte Carlo simulation methods, covering the transport equation, random sampling, computational geometry, collision physics, and statistics. The next lectures focus on the state-of-the-art in Monte Carlo criticality simulations, covering the theory of eigenvalue calculations, convergence analysis, dominance ratio calculations, bias in Keff and tallies, bias in uncertainties, a case study of a realistic calculation, and Wielandt acceleration techniques. The remaining lectures cover advanced topics, including HTGR modeling and stochastic geometry, temperature dependence, fission energy deposition, depletion calculations, parallel calculations, and parameter studies. This portion of the class focuses on using MCNP to perform criticality calculations for reactor physics and criticality safety applications. It is an intermediate level class, intended for those with at least some familiarity with MCNP. Class examples provide hands-on experience at running the code, plotting both geometry and results, and understanding the code output. The class includes lectures & hands-on computer use for a variety of Monte Carlo calculations

Monte Carlo Techniques for Nuclear Systems - Theory Lectures

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications Group; Univ. of New Mexico, Albuquerque, NM (United States). Nuclear Engineering Dept.

2016-11-29

These are lecture notes for a Monte Carlo class given at the University of New Mexico. The following topics are covered: course information; nuclear eng. review & MC; random numbers and sampling; computational geometry; collision physics; tallies and statistics; eigenvalue calculations I; eigenvalue calculations II; eigenvalue calculations III; variance reduction; parallel Monte Carlo; parameter studies; fission matrix and higher eigenmodes; doppler broadening; Monte Carlo depletion; HTGR modeling; coupled MC and T/H calculations; fission energy deposition. Solving particle transport problems with the Monte Carlo method is simple - just simulate the particle behavior. The devil is in the details, however. These lectures provide a balanced approach to the theory and practice of Monte Carlo simulation codes. The first lectures provide an overview of Monte Carlo simulation methods, covering the transport equation, random sampling, computational geometry, collision physics, and statistics. The next lectures focus on the state-of-the-art in Monte Carlo criticality simulations, covering the theory of eigenvalue calculations, convergence analysis, dominance ratio calculations, bias in Keff and tallies, bias in uncertainties, a case study of a realistic calculation, and Wielandt acceleration techniques. The remaining lectures cover advanced topics, including HTGR modeling and stochastic geometry, temperature dependence, fission energy deposition, depletion calculations, parallel calculations, and parameter studies. This portion of the class focuses on using MCNP to perform criticality calculations for reactor physics and criticality safety applications. It is an intermediate level class, intended for those with at least some familiarity with MCNP. Class examples provide hands-on experience at running the code, plotting both geometry and results, and understanding the code output. The class includes lectures & hands-on computer use for a variety of Monte Carlo calculations

Verification of the Monte Carlo differential operator technique for MCNP trademark

International Nuclear Information System (INIS)

McKinney, G.W.; Iverson, J.L.

1996-02-01

The differential operator perturbation technique has been incorporated into the Monte Carlo N-Particle transport code MCNP and will become a standard feature of future releases. This feature includes first and second order terms of the Taylor series expansion for response perturbations related to cross-section data (i.e., density, composition, etc.). Perturbation and sensitivity analyses can benefit from this technique in that predicted changes in one or more tally responses may be obtained for multiple perturbations in a single run. The user interface is intuitive, yet flexible enough to allow for changes in a specific microscopic cross section over a specified energy range. With this technique, a precise estimate of a small change in response is easily obtained, even when the standard deviation of the unperturbed tally is greater than the change. Furthermore, results presented in this report demonstrate that first and second order terms can offer acceptable accuracy, to within a few percent, for up to 20-30% changes in a response

Development of a coupling scheme between MCNP5 and subchanflow for the PIN- and fuel Assembly-Wise simulation of LWR and innovative reactors

International Nuclear Information System (INIS)

Ivanov, A.; Sanchez, V.; Imke, U.

2011-01-01

In order to increase the accuracy and the degree of spatial resolution of core design studies, coupled 3D neutronic (deterministic and Monte Carlo) and 3D thermal hydraulics (CFD and subchannel) codes are being developed worldwide. At KIT both deterministic and Monte Carlo codes were coupled with subchannel codes and applied to predict the safety-related design parameters such as pin power, maximal cladding and fuel temperature, DNB. These coupling approaches were revised and improved based on the experience gained. One particular example is replacing COBRA-TF with SUBCHANFLOW, in-house development subchannel 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 variety of fuel assembly types (BWR, PWR or SCFR). Key issues in such a coupled system are the way in which thermal-hydraulic/neutronic feedbacks, accuracy of the Monte Carlo solutions and observation of convergence during the iterative solution are handled. Another key issue that might be 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, it is possible to investigate large fuel assemblies in a pin-by-pin manner with a resolution at pin and subchannel level. One of the most important issues addressed in the current work is the temperature effects on nuclear data. For the particular studies pseudo material approach was used, which produces interpolated results for Doppler broadened cross sections from NJOY pre-generated nuclear data. (author)

SU-E-T-521: Investigation of the Uncertainties Involved in Secondary Neutron/gamma Production in Geant4/MCNP6 Monte Carlo Codes for Proton Therapy Application

International Nuclear Information System (INIS)

Mirzakhanian, L; Enger, S; Giusti, V

2015-01-01

Purpose: A major concern in proton therapy is the production of secondary neutrons causing secondary cancers, especially in young adults and children. Most utilized Monte Carlo codes in proton therapy are Geant4 and MCNP. However, the default versions of Geant4 and MCNP6 do not have suitable cross sections or physical models to properly handle secondary particle production in proton energy ranges used for therapy. In this study, default versions of Geant4 and MCNP6 were modified to better handle production of secondaries by adding the TENDL-2012 cross-section library. Methods: In-water proton depth-dose was measured at the “The Svedberg Laboratory” in Uppsala (Sweden). The proton beam was mono-energetic with mean energy of 178.25±0.2 MeV. The measurement set-up was simulated by Geant4 version 10.00 (default and modified version) and MCNP6. Proton depth-dose, primary and secondary particle fluence and neutron equivalent dose were calculated. In case of Geant4, the secondary particle fluence was filtered by all the physics processes to identify the main process responsible for the difference between the default and modified version. Results: The proton depth-dose curves and primary proton fluence show a good agreement between both Geant4 versions and MCNP6. With respect to the modified version, default Geant4 underestimates the production of secondary neutrons while overestimates that of gammas. The “ProtonInElastic” process was identified as the main responsible process for the difference between the two versions. MCNP6 shows higher neutron production and lower gamma production than both Geant4 versions. Conclusion: Despite the good agreement on the proton depth dose curve and primary proton fluence, there is a significant discrepancy on secondary neutron production between MCNP6 and both versions of Geant4. Further studies are thus in order to find the possible cause of this discrepancy or more accurate cross-sections/models to handle the nuclear

Utilization of the MCNP-3A code for criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.; Moreira, J.M.L.

1996-01-01

In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

MCNP capabilities at the dawn of the 21st century: Neutron-gamma applications

International Nuclear Information System (INIS)

Selcow, E.C.; McKinney, G.W.

2000-01-01

The Los Alamos National Laboratory Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron-gamma radiation transport applications. These include nuclear criticality safety, radiation shielding, nuclear safeguards, nuclear well-logging, fission and fusion reactor design, accelerator target design, detector design and analysis, health physics, medical radiation therapy and imaging, radiography, decontamination and decommissioning, and waste storage and disposal. The latest version of the code, MCNP4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000.This paper described the new features and capabilities of the code, and discusses the specific applicability to neutron-gamma problems. We will also discuss the future directions for MCNP code development, including rewriting the code in Fortran 90

Status of electron transport in MCNP trademark

International Nuclear Information System (INIS)

Hughes, H.G.

1997-01-01

The latest version of MCNP, the Los Alamos Monte Carlo transport code, has now been officially released. MCNP4B has been sent to the Radiation Safety Information Computational Center (RSICC), in Oak Ridge, Tennessee, which is responsible for the further distribution of the code within the US. International distribution of MCNP is done by the Nuclear Energy Agency (ECD/NEA), in Paris, France. Readers with access to the World-Wide-Web should consult the MCNP distribution site http://www-xdiv.lanl.gov/XTM/mcnp/about.html for specific information about contacting RSICC and OECD/NEA. A variety of new features are available in MCNP4B. Among these are differential operator perturbations, cross-section plotting capabilities, enhanced diagnostics for transport in repeated structures and lattices, improved efficiency in distributed-memory multiprocessing, corrected particle lifetime and lifespan estimators, and expanded software quality assurance procedures and testing, including testing of the multigroup Boltzmann-Fokker-Planck capability. New and improved cross section sets in the form of ENDF/B-VI evaluations have also been recently released and can be used in MCNP4B. Perhaps most significant for the interests of this special session, the electron transport algorithm has been improved, especially in the collisional energy-loss straggling and the angular-deflection treatments. In this paper, the author concentrates on a fairly complete documentation of the current status of the electron transport methods in MCNP

Use of McCad for the conversion of ITER CAD data to MCNP geometry

International Nuclear Information System (INIS)

Tsige-Tamirat, H.; Fischer, U.; Serikov, A.; Stickel, S.

2008-01-01

The program McCad provides a CAD interface for the Monte Carlo transport code MCNP. It is able to convert CAD data into MCNP input geometry description and provides GUI components for modeling, visualization, and data exchange. It performs sequences of tests on CAD data to check its validity and neutronics appropriateness including completion of the final MCNP model by void geometries. McCad has been used to convert a 40 deg. ITER torus sector CAD model to a suitable MCNP geometry model. Results of MCNP calculations performed to validate the converted geometry are presented

Conversion of Input Data between KENO and MCNP File Formats for Computer Criticality Assessments

International Nuclear Information System (INIS)

Schwarz, Randolph A.; Carter, Leland L.; Schwarz Alysia L.

2006-01-01

KENO is a Monte Carlo criticality code that is maintained by Oak Ridge National Laboratory (ORNL). KENO is included in the SCALE (Standardized Computer Analysis for Licensing Evaluation) package. KENO is often used because it was specifically designed for criticality calculations. Because KENO has convenient geometry input, including the treatment of lattice arrays of materials, it is frequently used for production calculations. Monte Carlo N-Particle (MCNP) is a Monte Carlo transport code maintained by Los Alamos National Laboratory (LANL). MCNP has a powerful 3D geometry package and an extensive cross section database. It is a general-purpose code and may be used for calculations involving shielding or medical facilities, for example, but can also be used for criticality calculations. MCNP is becoming increasingly more popular for performing production criticality calculations. Both codes have their own specific advantages. After a criticality calculation has been performed with one of the codes, it is often desirable (or may be a safety requirement) to repeat the calculation with the other code to compare the important parameters using a different geometry treatment and cross section database. This manual conversion of input files between the two codes is labor intensive. The industry needs the capability of converting geometry models between MCNP and KENO without a large investment in manpower. The proposed conversion package will aid the user in converting between the codes. It is not intended to be used as a ''black box''. The resulting input file will need to be carefully inspected by criticality safety personnel to verify the intent of the calculation is preserved in the conversion. The purpose of this package is to help the criticality specialist in the conversion process by converting the geometry, materials, and pertinent data cards

Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

International Nuclear Information System (INIS)

Shiba, T.; Fallot, M.

2015-01-01

To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

A new effective Monte Carlo Midway coupling method in MCNP applied to a well logging problem

Energy Technology Data Exchange (ETDEWEB)

Serov, I.V.; John, T.M.; Hoogenboom, J.E

1998-12-01

The background of the Midway forward-adjoint coupling method including the black absorber technique for efficient Monte Carlo determination of radiation detector responses is described. The method is implemented in the general purpose MCNP Monte Carlo code. The utilization of the method is fairly straightforward and does not require any substantial extra expertise. The method was applied to a standard neutron well logging porosity tool problem. The results exhibit reliability and high efficiency of the Midway method. For the studied problem the efficiency gain is considerably higher than for a normal forward calculation, which is already strongly optimized by weight-windows. No additional effort is required to adjust the Midway model if the position of the detector or the porosity of the formation is changed. Additionally, the Midway method can be used with other variance reduction techniques if extra gain in efficiency is desired.

Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

International Nuclear Information System (INIS)

Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.

2015-01-01

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

Estimation and interpretation of keff confidence intervals in MCNP

International Nuclear Information System (INIS)

Urbatsch, T.J.

1995-01-01

The Monte Carlo code MCNP has three different, but correlated, estimators for calculating k eff in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k eff estimator, is shown to be the best k eff estimator available in MCNP for estimating k eff confidence intervals. Theoretically, the Gauss-Markov theorem provides a solid foundation for MCNP's three-combined estimator. Analytically, a statistical study, where the estimates are drawn using a known covariance matrix, shows that the three-combined estimator is superior to the estimator with the smallest variance. Empirically, MCNP examples for several physical systems demonstrate the three-combined estimator's superiority over each of the three individual estimators and its correct coverage rates. Additionally, the importance of MCNP's statistical checks is demonstrated

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part II: gadolinium neutron capture therapy models and therapeutic effects.

Science.gov (United States)

Wangerin, K; Culbertson, C N; Jevremovic, T

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam.

Simulation of reactor noise analysis measurement for light-water critical assembly TCA using MCNP-DSP

International Nuclear Information System (INIS)

Yamamoto, Toshihiro; Sakurai, Kiyoshi; Tonoike, Kotaro; Miyoshi, Yoshinori

2001-01-01

Reactor noise analysis methods using Monte Carlo technique have been proposed and developed in the field of nuclear criticality safety. The Monte Carlo simulation for noise analysis can be made by simulating physical phenomena in the course of neutron transport in a nuclear fuel as practically as possible. MCNP-DSP was developed by T. Valentine of ORNL for this purpose and it is a modified version of MCNP-4A. The authors applied this code to frequency analysis measurements performed in light-water critical assembly TCA. Prompt neutron generation times for critical and subcritical cores were measured by doing the frequency analysis of detector signals. The Monte Carlo simulations for these experiments were carried out using MCNP-DSP, and prompt neutron generation times were calculated. (author)

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

International Nuclear Information System (INIS)

Talamo, Alberto; Ji, Wei; Cetnar, Jerzy; Gudowski, Waclaw

2006-01-01

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides

Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)]. E-mail: alby@anl.gov; Ji, Wei [University of Michigan, Bonisteel Boulevard 2355, Ann Arbor, MI 48109-2104 (United States); Cetnar, Jerzy [AGH-University of Science and Technology, Al. Mickiewicza 30 Cracow (Poland); Gudowski, Waclaw [Royal Institute of Technology (KTH), Roslagstullsbacken 21, Stockholm S-10691 (Sweden)

2006-09-15

Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.

Validation and verification of MCNP6 against intermediate and high-energy experimental data and results by other codes

International Nuclear Information System (INIS)

Mashnik, Stepan G.

2011-01-01

MCNP6, the latest and most advanced LANL transport code representing a recent merger of MCNP5 and MCNPX, has been Validated and Verified (V and V) against a variety of intermediate and high-energy experimental data and against results by different versions of MCNPX and other codes. In the present work, we V and V MCNP6 using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.02 and LAQGSM03.03. We found that MCNP6 describes reasonably well various reactions induced by particles and nuclei at incident energies from 18 MeV to about 1 TeV per nucleon measured on thin and thick targets and agrees very well with similar results obtained with MCNPX and calculations by CEM03.02, LAQGSM03.01 (03.03), INCL4 + ABLA, and Bertini INC + Dresner evaporation, EPAX, ABRABLA, HIPSE, and AMD, used as stand alone codes. Most of several computational bugs and more serious physics problems observed in MCNP6/X during our V and V have been fixed; we continue our work to solve all the known problems before MCNP6 is distributed to the public. (author)

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

Science.gov (United States)

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. Copyright © 2014 Elsevier Ltd. All rights reserved.

Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

International Nuclear Information System (INIS)

Aredes, Vitor Ottoni; Bitelli, Ulysses d'Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza

2015-01-01

This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10 8 ± 5.25% n/cm 2 s. (author)

Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

Energy Technology Data Exchange (ETDEWEB)

Aredes, Vitor Ottoni; Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)

Utilization of new 150-MeV neutron and proton evaluations in MCNP

International Nuclear Information System (INIS)

Little, R.C.; Frankle, S.C.; Hughes, H.G. III; Prael, R.E.

1997-01-01

MCNP trademark and LAHET trademark are two of the codes included in the LARAMIE (Los Alamos Radiation Modeling Interactive Environment) code system. Both MCNP and LAHET are three-dimensional continuous-energy Monte Carlo radiation transport codes. The capabilities of MCNP and LAHET are currently being merged into one code for the Accelerator Production of Tritium (APT) program at Los Alamos National Laboratory. Concurrently, a significant effort is underway to improve the accuracy of the physics in the merged code. In particular, full nuclear-data evaluations (in ENDF6 format) for many materials of importance to APT are being produced for incident neutrons and protons up to an energy of 150-MeV. After processing, cross-section tables based on these new evaluations will be available for use fin the merged code. In order to utilize these new cross-section tables, significant enhancements are required for the merged code. Neutron cross-section tables for MCNP currently specify emission data for neutrons and photons only; the new evaluations also include complete neutron-induced data for protons, deuterons, tritons, and alphas. In addition, no provision in either MCNP or LAHET currently exists for the use of incident charged-particle tables other than for electrons. To accommodate the new neutron-induced data, it was first necessary to expand the format definition of an MCNP neutron cross-section table. The authors have prepared a 150-MeV neutron cross-section library in this expanded format for 15 nuclides. Modifications to MCNP have been implemented so that this expanded neutron library can be utilized

MCNP evaluation of top node control rod depletion below the core in KKL

International Nuclear Information System (INIS)

Beran, Tâm; Seltborg, Per; Lindahl, Sten-Örjan; Bieli, Roger; Ledergerber, Guido

2014-01-01

In previous studies, there has been identified a significant discrepancy in the BWR control rod top node depletion between the two core simulator nodal codes POLCA7 and PRESTO-2, which indicates that there is a large general uncertainty in nodal codes in calculating the top node depletion of fully withdrawn control rods. In this study, the stochastic Monte Carlo code MCNP has been used to calculate the top node control rod depletion for benchmarking the nodal codes. By using the TIP signal obtained from an extended TIP campaign below the core performed in the KKL reactor, the MCNP model has been verified by comparing the axial profile between the TIP data and the gamma flux calculated by MCNP. The MCNP results have also been compared with calculations from POLCA7, which was found to yield slightly higher depletion rates than MCNP. It was also found that the 10 B depletion in the top node is very sensitive to the exact axial location of the control rod top when it is fully withdrawn. By using the MCNP results, the neutron flux model below the core in the nodal codes can be improved by implementing an exponential function for the neutron flux. (author)

Experiment on neutron transmission through depleted uranium layers and analysis with DOT 3.5 and MCNP

International Nuclear Information System (INIS)

Oka, Y.; Kodama, T.; Akiyama, M.; Hashikura, H.; Kondo, S.

1987-01-01

The reaction rates in the multi-layers containing depleted uranium were measured by activation foils and micro-fission chambers. The analysis of the experiment was carried out by using the multi-group transport calculation code, DOT 3.5 and the continuous energy Monte Carlo code, MCNP. The multi-group calculation overpredicted the low energy reaction rates in the DU layers, while the continuous energy calculation agreed well. The multi-group and continuous energy calculation was compared for the one-dimensional transmission of iron spheres. The results revealed overprediction of the multi-group calculation near the fast neutron source. The averaging of the resonance shapes in generating the multi-group cross sections made minima of the resonance valleys higher than that of the pointwise cross section. This increased the scattering of the neutrons inside and caused the overprediction of the multi-group calculation

MCNP load balancing and fault tolerance with PVM

International Nuclear Information System (INIS)

McKinney, G.W.

1995-01-01

Version 4A of the Monte Carlo neutron, photon, and electron transport code MCNP, developed by LANL (Los Alamos National Laboratory), supports distributed-memory multiprocessing through the software package PVM (Parallel Virtual Machine, version 3.1.4). Using PVM for interprocessor communication, MCNP can simultaneously execute a single problem on a cluster of UNIX-based workstations. This capability provided system efficiencies that exceeded 80% on dedicated workstation clusters, however, on heterogeneous or multiuser systems, the performance was limited by the slowest processor (i.e., equal work was assigned to each processor). The next public release of MCNP will provide multiprocessing enhancements that include load balancing and fault tolerance which are shown to dramatically increase multiuser system efficiency and reliability

Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP

Energy Technology Data Exchange (ETDEWEB)

Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Aguilera, Pablo [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, Depto. de Física, Facultad de Ciencias, Las Palmeras 3425, Ñuñoa, Santiago (Chile); Arellano, H. F. [Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile)

2016-07-07

The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.

Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Kiyoshi; Yamamoto, Toshihiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; eds.

1998-03-01

`MCNP Use Experience` Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year`s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile `Guideline of Monte Carlo Calculation` which will be a standard in the future. The appendices of this report include this `Guideline`, the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)

Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques

International Nuclear Information System (INIS)

Sakurai, Kiyoshi; Yamamoto, Toshihiro

1998-03-01

''MCNP Use Experience'' Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year''s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile ''Guideline of Monte Carlo Calculation'' which will be a standard in the future. The appendices of this report include this ''Guideline'', the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)

Reactor physics simulations with coupled Monte Carlo calculation and computational fluid dynamics

International Nuclear Information System (INIS)

Seker, V.; Thomas, J.W.; Downar, T.J.

2007-01-01

A computational code system based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program 'McSTAR'. McSTAR is written in FORTRAN90 programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STARCD for every region. Three different methods were investigated and implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. Preliminary testing of the code was performed using a 3x3 PWR pin-cell problem. The preliminary results are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code DeCART. Good agreement in the k eff and the power profile was observed. Increased computational capabilities and improvements in computational methods have accelerated interest in high fidelity modeling of nuclear reactor cores during the last several years. High-fidelity has been achieved by utilizing full core neutron transport solutions for the neutronics calculation and computational fluid dynamics solutions for the thermal-hydraulics calculation. Previous researchers have reported the coupling of 3D deterministic neutron transport method to CFD and their application to practical reactor analysis problems. One of the principal motivations of the work here was to utilize Monte Carlo methods to validate the coupled deterministic neutron transport

First validation of the new continuous energy version of the MORET5 Monte Carlo code

International Nuclear Information System (INIS)

Miss, Joachim; Bernard, Franck; Forestier, Benoit; Haeck, Wim; Richet, Yann; Jacquet, Olivier

2008-01-01

The 5.A.1 version is the next release of the MORET Monte Carlo code dedicated to criticality and reactor calculations. This new version combines all the capabilities that are already available in the multigroup version with many new and enhanced features. The main capabilities of the previous version are the powerful association of a deterministic and Monte Carlo approach (like for instance APOLLO-MORET), the modular geometry, five source sampling techniques and two simulation strategies. The major advance in MORET5 is the ability to perform calculations either a multigroup or a continuous energy simulation. Thanks to these new developments, we now have better control over the whole process of criticality calculations, from reading the basic nuclear data to the Monte Carlo simulation itself. Moreover, this new capability enables us to better validate the deterministic-Monte Carlo multigroup calculations by performing continuous energy calculations with the same code, using the same geometry and tracking algorithms. The aim of this paper is to describe the main options available in this new release, and to present the first results. Comparisons of the MORET5 continuous-energy results with experimental measurements and against another continuous-energy Monte Carlo code are provided in terms of validation and time performance. Finally, an analysis of the interest of using a unified energy grid for continuous energy Monte Carlo calculations is presented. (authors)

CREPT-MCNP code for efficiency calibration of HPGe detectors with the representative point method.

Science.gov (United States)

Saegusa, Jun

2008-01-01

The representative point method for the efficiency calibration of volume samples has been previously proposed. For smoothly implementing the method, a calculation code named CREPT-MCNP has been developed. The code estimates the position of a representative point which is intrinsic to each shape of volume sample. The self-absorption correction factors are also given to make correction on the efficiencies measured at the representative point with a standard point source. Features of the CREPT-MCNP code are presented.

Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

CERN Document Server

2002-01-01

This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

Estimation of skyshine dose from turbine building of BWR plant using Monte Carlo code

Energy Technology Data Exchange (ETDEWEB)

Yuji, Nemoto; Toshihisa, Tsukiyama; Shigeki, Nemezawa [Hitachi. Ltd., Saiwai-cho, Hitachi (Japan); Tadashi, Yamasaki; Hidetsugu, Okada [Chubu Electric Power Company, Inc., Odaka-cho, Midori-ku Nagoya (Japan)

2007-07-01

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)

Modeling of LVRF critical experiments in ZED-2 using WIMS9A/PANTHER and MCNP5

International Nuclear Information System (INIS)

Sissaoui, M.T.; Carlson, P.A.; Lebenhaft, J.R.

2009-01-01

The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) underpredicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 overpredicted k eff and underpredicted the CVR bias relative to MCNP5 by 100-200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately

On the effect of updated MCNP photon cross section data on the simulated response of the HPA TLD.

Science.gov (United States)

Eakins, Jonathan

2009-02-01

The relative response of the new Health Protection Agency thermoluminescence dosimeter (TLD) has been calculated for Narrow Series X-ray distribution and (137)Cs photon sources using the Monte Carlo code MCNP5, and the results compared with those obtained during its design stage using the predecessor code, MCNP4c2. The results agreed at intermediate energies (approximately 0.1 MeV to (137)Cs), but differed at low energies (<0.1 MeV) by up to approximately 10%. This disparity has been ascribed to differences in the default photon interaction data used by the two codes, and derives ultimately from the effect on absorbed dose of the recent updates to the photoelectric cross sections. The sources of these data have been reviewed.

TET_2MCNP: A conversion program to implement tetrahearal-mesh models in MCNP

International Nuclear Information System (INIS)

Han, Min Cheol; Yeom, Yeon Soo; Nguyen, Thng Tat; Choi, Chan Soo; Lee, Hyun Su; Kim, Chan Hyeong

2016-01-01

Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET_2MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. TET_2MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET_2MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET_2MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. In the present study, we have developed a computer program, TET_2MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code

New strategies of sensitivity analysis capabilities in continuous-energy Monte Carlo code RMC

International Nuclear Information System (INIS)

Qiu, Yishu; Liang, Jingang; Wang, Kan; Yu, Jiankai

2015-01-01

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

Impact of MCNP Unresolved Resonance Probability-Table Treatment on Uranium and Plutonium Benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.; Little, R.C.

1999-01-01

A probability-table treatment recently has been incorporated into an intermediate version of the MCNP Monte Carlo code named MCNP4XS. This paper presents MCNP4XS results for a variety of uranium and plutonium criticality benchmarks, calculated with and without the probability-table treatment. It is shown that the probability-table treatment can produce small but significant reactivity changes for plutonium and 233 U systems with intermediate spectra. More importantly, it can produce substantial reactivity increases for systems with large amounts of 238 U and intermediate spectra

Performance of scientific computing platforms with MCNP4B

International Nuclear Information System (INIS)

McLaughlin, H.E.; Hendricks, J.S.

1998-01-01

Several computing platforms were evaluated with the MCNP4B Monte Carlo radiation transport code. The DEC AlphaStation 500/500 was the fastest to run MCNP4B. Compared to the HP 9000-735, the fastest platform 4 yr ago, the AlphaStation is 335% faster, the HP C180 is 133% faster, the SGI Origin 2000 is 82% faster, the Cray T94/4128 is 1% faster, the IBM RS/6000-590 is 93% as fast, the DEC 3000/600 is 81% as fast, the Sun Sparc20 is 57% as fast, the Cray YMP 8/8128 is 57% as fast, the sun Sparc5 is 33% as fast, and the Sun Sparc2 is 13% as fast. All results presented are reproducible and allow for comparison to computer platforms not included in this study. Timing studies are seen to be very problem dependent. The performance gains resulting from advances in software were also investigated. Various compilers and operating systems were seen to have a modest impact on performance, whereas hardware improvements have resulted in a factor of 4 improvement. MCNP4B also ran approximately as fast as MCNP4A

Monte Carlo simulation applied to radiosurgery narrow beams using MCNP-4C

International Nuclear Information System (INIS)

Chaves, A.; Lopes, M.C.; Oliveira, C.

2001-01-01

Dose measurements for the narrow photon beams used in radiosurgery are complicated by the lack of electron equilibrium which is a requirement namely for ionometric methods. To overcome this difficulty the use of different dosimetric supports is strongly recommended in order to appreciate the influence of each type of detector. Monte Carlo simulation is another kind of tool to assess the details of the energy deposition phenomena in such narrow photon beams. In this study output factors and depth dose calculated by the Monte Carlo MCNP-4C code are presented and compared with experimental data measured with a diode, a Markus chamber, a 0.125 cc thimble chamber and a Pinpoint chamber. Simulated energy spectra for narrow beams are also presented in order to compare them with the reference 10 cm x 10 cm beam field size and thus discuss the different contributions of the absorbed energy in water, in each case. A detailed analysis on the photon energy spectra showed a slight decrease on the photon mean energy that can be explained by the increased scattering inside the additional collimators. Calculated and measured depth doses curves are in good agreement for most of the collimators. For the two smallest collimators some differences have been pointed and explained according to the characteristics of the detectors (author)

Recent developments in the Los Alamos radiation transport code system

International Nuclear Information System (INIS)

Forster, R.A.; Parsons, K.

1997-01-01

A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results

Applications of the Los Alamos High Energy Transport code

International Nuclear Information System (INIS)

Waters, L.; Gavron, A.; Prael, R.E.

1992-01-01

Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter

MORSE Monte Carlo code

International Nuclear Information System (INIS)

Cramer, S.N.

1984-01-01

The MORSE code is a large general-use multigroup Monte Carlo code system. Although no claims can be made regarding its superiority in either theoretical details or Monte Carlo techniques, MORSE has been, since its inception at ORNL in the late 1960s, the most widely used Monte Carlo radiation transport code. The principal reason for this popularity is that MORSE is relatively easy to use, independent of any installation or distribution center, and it can be easily customized to fit almost any specific need. Features of the MORSE code are described

Calculation of gamma ray dose buildup factors in water for isotropic point, plane mono directional and line sources using MCNP code

International Nuclear Information System (INIS)

Atak, H.; Celikten, O. S.; Tombakoglu, M.

2009-01-01

Gamma ray dose buildup factors in water for isotropic point, plane mono directional and infinite/finite line sources were calculated using the MCNP code. The buildup factors are determined for gamma ray energies of 1, 2, 3 and 4 Mev and for shield thicknesses of 1, 2, 4 and 7 mean free paths. The calculated buildup factors were then fitted in the Taylor and Berger forms. For the line sources a buildup factor table was also constructed using the Sievert function and the constants in Taylor form derived in this study to compare with the Monte Carlo results. All buildup factors were compared with the tabulated data given in literature. In order to reduce the statistical errors on buildup factors, 'forced collision' option was used in the MCNP calculations.

TET{sub 2}MCNP: A conversion program to implement tetrahearal-mesh models in MCNP

Energy Technology Data Exchange (ETDEWEB)

Han, Min Cheol; Yeom, Yeon Soo; Nguyen, Thng Tat; Choi, Chan Soo; Lee, Hyun Su; Kim, Chan Hyeong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

2016-12-15

Tetrahedral-mesh geometries can be used in the MCNP code, but the MCNP code accepts only the geometry in the Abaqus input file format; hence, the existing tetrahedral-mesh models first need to be converted to the Abacus input file format to be used in the MCNP code. In the present study, we developed a simple but useful computer program, TET{sub 2}MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. TET{sub 2}MCNP is written in C++ and contains two components: one for converting a TetGen output file to the Abacus input file and the other for the reverse conversion process. The TET{sub 2}MCP program also produces an MCNP input file. Further, the program provides some MCNP-specific functions: the maximum number of elements (i.e., tetrahedrons) per part can be limited, and the material density of each element can be transferred to the MCNP input file. To test the developed program, two tetrahedral-mesh models were generated using TetGen and converted to the Abaqus input file format using TET{sub 2}MCNP. Subsequently, the converted files were used in the MCNP code to calculate the object- and organ-averaged absorbed dose in the sphere and phantom, respectively. The results show that the converted models provide, within statistical uncertainties, identical dose values to those obtained using the PHITS code, which uses the original tetrahedral-mesh models produced by the TetGen program. The results show that the developed program can successfully convert TetGen tetrahedral-mesh models to Abacus input files. In the present study, we have developed a computer program, TET{sub 2}MCNP, which can be used to convert TetGen-generated tetrahedral-mesh models to the Abaqus input file format for use in the MCNP code. We believe this program will be used by many MCNP users for implementing complex tetrahedral-mesh models, including computational human phantoms, in the MCNP code.

Modeling of LVRF Critical Experiments in ZED-2 Using WIMS9A/PANTHER and MCNP5

International Nuclear Information System (INIS)

Sissaoui, M.T.; Lebenhaft, J.R; Carlson, P.A.

2008-01-01

The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) under-predicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 over-predicted k eff and under-predicted the CVR bias relative to MCNP5 by 100 pcm to 200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately. (authors)

Performance of MPI parallel processing implemented by MCNP5/ MCNPX for criticality benchmark problems

International Nuclear Information System (INIS)

Mark Dennis Usang; Mohd Hairie Rabir; Mohd Amin Sharifuldin Salleh; Mohamad Puad Abu

2012-01-01

MPI parallelism are implemented on a SUN Workstation for running MCNPX and on the High Performance Computing Facility (HPC) for running MCNP5. 23 input less obtained from MCNP Criticality Validation Suite are utilized for the purpose of evaluating the amount of speed up achievable by using the parallel capabilities of MPI. More importantly, we will study the economics of using more processors and the type of problem where the performance gain are obvious. This is important to enable better practices of resource sharing especially for the HPC facilities processing time. Future endeavours in this direction might even reveal clues for best MCNP5/ MCNPX coding practices for optimum performance of MPI parallelisms. (author)

Specialized Monte Carlo codes versus general-purpose Monte Carlo codes

International Nuclear Information System (INIS)

Moskvin, Vadim; DesRosiers, Colleen; Papiez, Lech; Lu, Xiaoyi

2002-01-01

The possibilities of Monte Carlo modeling for dose calculations and optimization treatment are quite limited in radiation oncology applications. The main reason is that the Monte Carlo technique for dose calculations is time consuming while treatment planning may require hundreds of possible cases of dose simulations to be evaluated for dose optimization. The second reason is that general-purpose codes widely used in practice, require an experienced user to customize them for calculations. This paper discusses the concept of Monte Carlo code design that can avoid the main problems that are preventing wide spread use of this simulation technique in medical physics. (authors)

3D dose distribution calculation in a voxelized human phantom by means of Monte Carlo method

International Nuclear Information System (INIS)

Abella, V.; Miro, R.; Juste, B.; Verdu, G.

2010-01-01

The aim of this work is to provide the reconstruction of a real human voxelized phantom by means of a MatLab program and the simulation of the irradiation of such phantom with the photon beam generated in a Theratron 780 (MDS Nordion) 60 Co radiotherapy unit, by using the Monte Carlo transport code MCNP (Monte Carlo N-Particle), version 5. The project results in 3D dose mapping calculations inside the voxelized antropomorphic head phantom. The program provides the voxelization by first processing the CT slices; the process follows a two-dimensional pixel and material identification algorithm on each slice and three-dimensional interpolation in order to describe the phantom geometry via small cubic cells, resulting in an MCNP input deck format output. Dose rates are calculated by using the MCNP5 tool FMESH, superimposed mesh tally, which gives the track length estimation of the particle flux in units of particles/cm 2 . Furthermore, the particle flux is converted into dose by using the conversion coefficients extracted from the NIST Physical Reference Data. The voxelization using a three-dimensional interpolation technique in combination with the use of the FMESH tool of the MCNP Monte Carlo code offers an optimal simulation which results in 3D dose mapping calculations inside anthropomorphic phantoms. This tool is very useful in radiation treatment assessments, in which voxelized phantoms are widely utilized.

Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

International Nuclear Information System (INIS)

Donders, R.E.; Douglas, S.R.

2002-01-01

Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)

MCNP Perturbation Capability for Monte Carlo Criticality Calculations

International Nuclear Information System (INIS)

Hendricks, J.S.; Carter, L.L.; McKinney, G.W.

1999-01-01

The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k 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

Organ dose conversion coefficients based on a voxel mouse model and MCNP code for external photon irradiation.

Science.gov (United States)

Zhang, Xiaomin; Xie, Xiangdong; Cheng, Jie; Ning, Jing; Yuan, Yong; Pan, Jie; Yang, Guoshan

2012-01-01

A set of conversion coefficients from kerma free-in-air to the organ absorbed dose for external photon beams from 10 keV to 10 MeV are presented based on a newly developed voxel mouse model, for the purpose of radiation effect evaluation. The voxel mouse model was developed from colour images of successive cryosections of a normal nude male mouse, in which 14 organs or tissues were segmented manually and filled with different colours, while each colour was tagged by a specific ID number for implementation of mouse model in Monte Carlo N-particle code (MCNP). Monte Carlo simulation with MCNP was carried out to obtain organ dose conversion coefficients for 22 external monoenergetic photon beams between 10 keV and 10 MeV under five different irradiation geometries conditions (left lateral, right lateral, dorsal-ventral, ventral-dorsal, and isotropic). Organ dose conversion coefficients were presented in tables and compared with the published data based on a rat model to investigate the effect of body size and weight on the organ dose. The calculated and comparison results show that the organ dose conversion coefficients varying the photon energy exhibits similar trend for most organs except for the bone and skin, and the organ dose is sensitive to body size and weight at a photon energy approximately <0.1 MeV.

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Monte Carlo radiation transport: A revolution in science

International Nuclear Information System (INIS)

Hendricks, J.

1993-01-01

When Enrico Fermi, Stan Ulam, Nicholas Metropolis, John von Neuman, and Robert Richtmyer invented the Monte Carlo method fifty years ago, little could they imagine the far-flung consequences, the international applications, and the revolution in science epitomized by their abstract mathematical method. The Monte Carlo method is used in a wide variety of fields to solve exact computational models approximately by statistical sampling. It is an alternative to traditional physics modeling methods which solve approximate computational models exactly by deterministic methods. Modern computers and improved methods, such as variance reduction, have enhanced the method to the point of enabling a true predictive capability in areas such as radiation or particle transport. This predictive capability has contributed to a radical change in the way science is done: design and understanding come from computations built upon experiments rather than being limited to experiments, and the computer codes doing the computations have become the repository for physics knowledge. The MCNP Monte Carlo computer code effort at Los Alamos is an example of this revolution. Physicians unfamiliar with physics details can design cancer treatments using physics buried in the MCNP computer code. Hazardous environments and hypothetical accidents can be explored. Many other fields, from underground oil well exploration to aerospace, from physics research to energy production, from safety to bulk materials processing, benefit from MCNP, the Monte Carlo method, and the revolution in science

Generating Importance Map for Geometry Splitting using Discrete Ordinates Code in Deep Shielding Problem

International Nuclear Information System (INIS)

Kim, Jong Woon; Lee, Young Ouk

2016-01-01

When we use MCNP code for a deep shielding problem, we prefer to use variance reduction technique such as geometry splitting, or weight window, or source biasing to have relative error within reliable confidence interval. To generate importance map for geometry splitting in MCNP calculation, we should know the track entering number and previous importance on each cells since a new importance is calculated based on these information. If a problem is deep shielding problem such that we have zero tracks entering on a cell, we cannot generate new importance map. In this case, discrete ordinates code can provide information to generate importance map easily. In this paper, we use AETIUS code as a discrete ordinates code. Importance map for MCNP is generated based on a zone average flux of AETIUS calculation. The discretization of space, angle, and energy is not necessary for MCNP calculation. This is the big merit of MCNP code compared to the deterministic code. However, deterministic code (i.e., AETIUS) can provide a rough estimate of the flux throughout a problem relatively quickly. This can help MCNP by providing variance reduction parameters. Recently, ADVANTG code is released. This is an automated tool for generating variance reduction parameters for fixed-source continuous-energy Monte Carlo simulations with MCNP5 v1.60

Particle Track Visualization using the MCNP Visual Editor

International Nuclear Information System (INIS)

Schwarz, Randolph A.; Carter, Lee; Brown, Wendi A.

2001-01-01

The Monte Carlo N-Particle (MCNP) visual editor1,2,3 is used throughout the world for displaying and creating complex MCNP geometries. The visual editor combines the Los Alamos MCNP Fortran code with a C front end to provide a visual interface. A big advantage of this approach is that the particle transport routines for MCNP are available to the visual front end. The latest release of the visual editor by Pacific Northwest National Laboratory enables the user to plot transport data points on top of a two-dimensional geometry plot. The user can plot source points, collisions points, surface crossings, and tally contributions. This capability can be used to show where particle collisions are occurring, verify the effectiveness of the particle biasing, or show which collisions contribute to a tally. For a KCODE (criticality source) calculation, the visual editor can be used to plot the source points for specific cycles

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

Science.gov (United States)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

Science.gov (United States)

Li, Fusheng; Han, Xiaogang

2012-07-01

The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. Copyright © 2011 Elsevier Ltd. All rights reserved.

Accelerating Pseudo-Random Number Generator for MCNP on GPU

Science.gov (United States)

Gong, Chunye; Liu, Jie; Chi, Lihua; Hu, Qingfeng; Deng, Li; Gong, Zhenghu

2010-09-01

Pseudo-random number generators (PRNG) are intensively used in many stochastic algorithms in particle simulations, artificial neural networks and other scientific computation. The PRNG in Monte Carlo N-Particle Transport Code (MCNP) requires long period, high quality, flexible jump and fast enough. In this paper, we implement such a PRNG for MCNP on NVIDIA's GTX200 Graphics Processor Units (GPU) using CUDA programming model. Results shows that 3.80 to 8.10 times speedup are achieved compared with 4 to 6 cores CPUs and more than 679.18 million double precision random numbers can be generated per second on GPU.

An Electron/Photon/Relaxation Data Library for MCNP6

Energy Technology Data Exchange (ETDEWEB)

Hughes, III, H. Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-08-07

The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.

Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

Science.gov (United States)

Faghihi, F.; Mehdizadeh, S.; Hadad, K.

Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

Practice of calculation of neutron-physical characteristics of reactors and radiating shielding in structure SNPS with program complex MCNP

International Nuclear Information System (INIS)

Krotov, A.D.; Son'ko, A.V.

2009-01-01

Calculation of neutron-physical properties and radiation protection of space power reactor was made by means of the MCNP code allowing simulation of neutron, γ- and electron transport by the Monte Carlo method in the systems with combined geometry. Universality of the MCNP code has been demonstrated both for the calculation of reactor-converter so for the optimization of radiation protection that allows to reserve a new level of complex simulation of SNPS [ru

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

2016-11-15

Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

MCNP6 fragmentation of light nuclei at intermediate energies

Energy Technology Data Exchange (ETDEWEB)

Mashnik, Stepan G., E-mail: mashnik@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Kerby, Leslie M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of Idaho, Moscow, ID 83844 (United States)

2014-11-11

Fragmentation reactions induced on light target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the latest Los Alamos Monte Carlo transport code MCNP6 and with its cascade-exciton model (CEM) and Los Alamos version of the quark-gluon string model (LAQGSM) event generators, version 03.03, used as stand-alone codes. Such reactions are involved in different applications, like cosmic-ray-induced single event upsets (SEU's), radiation protection, and cancer therapy with proton and ion beams, among others; therefore, it is important that MCNP6 simulates them as well as possible. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. Both CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to {sup 4}He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

Benchmark of neutron production cross sections with Monte Carlo codes

Science.gov (United States)

Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun

2018-02-01

Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first

Adjoint acceleration of Monte Carlo simulations using TORT/MCNP coupling approach: A case study on the shielding improvement for the cyclotron room of the Buddhist Tzu Chi General Hospital

International Nuclear Information System (INIS)

Sheu, R. J.; Sheu, R. D.; Jiang, S. H.; Kao, C. H.

2005-01-01

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted. (authors)

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

Science.gov (United States)

Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

2017-09-01

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

International Nuclear Information System (INIS)

Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

2003-01-01

One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

MCNP modelling of a combined neutron/gamma counter

CERN Document Server

Bourva, L C A; Ottmar, H; Weaver, D R

1999-01-01

A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of device, part of a suite of non-destructive assay instruments utilised for the enforcement of the Euratom nuclear safeguards within the European Union, is to be used for high accuracy measurements of the plutonium content of small samples of nuclear materials. The multi-purpose Monte Carlo N-particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO sub 2 and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study to both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, epsilon, to, alpha, the ratio of (alpha...

An improved algorithm to convert CAD model to MCNP geometry model based on STEP file

International Nuclear Information System (INIS)

Zhou, Qingguo; Yang, Jiaming; Wu, Jiong; Tian, Yanshan; Wang, Junqiong; Jiang, Hai; Li, Kuan-Ching

2015-01-01

Highlights: • Fully exploits common features of cells, making the processing efficient. • Accurately provide the cell position. • Flexible to add new parameters in the structure. • Application of novel structure in INP file processing, conveniently evaluate cell location. - Abstract: MCNP (Monte Carlo N-Particle Transport Code) is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. Its input file, the INP file, has the characteristics of complicated form and is error-prone when describing geometric models. Due to this, a conversion algorithm that can solve the problem by converting general geometric model to MCNP model during MCNP aided modeling is highly needed. In this paper, we revised and incorporated a number of improvements over our previous work (Yang et al., 2013), which was proposed and targeted after STEP file and INP file were analyzed. Results of experiments show that the revised algorithm is more applicable and efficient than previous work, with the optimized extraction of geometry and topology information of the STEP file, as well as the production efficiency of output INP file. This proposed research is promising, and serves as valuable reference for the majority of researchers involved with MCNP-related researches

MCNP simulation of a Theratron 780 radiotherapy unit.

Science.gov (United States)

Miró, R; Soler, J; Gallardo, S; Campayo, J M; Díez, S; Verdú, G

2005-01-01

A Theratron 780 (MDS Nordion) 60Co radiotherapy unit has been simulated with the Monte Carlo code MCNP. The unit has been realistically modelled: the cylindrical source capsule and its housing, the rectangular collimator system, both the primary and secondary jaws and the air gaps between the components. Different collimator openings, ranging from 5 x 5 cm2 to 20 x 20 cm2 (narrow and broad beams) at a source-surface distance equal to 80 cm have been used during the study. In the present work, we have calculated spectra as a function of field size. A study of the variation of the electron contamination of the 60Co beam has also been performed.

Determination of the detection efficiency of a HPGe detector by means of the MCNP 4A simulation code

International Nuclear Information System (INIS)

Leal, B.

2004-01-01

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)

A new MCNP trademark test set

International Nuclear Information System (INIS)

Brockhoff, R.C.; Hendricks, J.S.

1994-09-01

The MCNP test set is used to test the MCNP code after installation on various computer platforms. For MCNP4 and MCNP4A this test set included 25 test problems designed to test as many features of the MCNP code as possible. A new and better test set has been devised to increase coverage of the code from 85% to 97% with 28 problems. The new test set is as fast as and shorter than the MCNP4A test set. The authors describe the methodology for devising the new test set, the features that were not covered in the MCNP4A test set, and the changes in the MCNP4A test set that have been made for MCNP4B and its developmental versions. Finally, new bugs uncovered by the new test set and a compilation of all known MCNP4A bugs are presented

Dose distribution and dosimetry parameters calculation of MED3633 Palladium-103 source in water phantom using MCNP

International Nuclear Information System (INIS)

Mowlavi, A. A.; Binesh, A.; Moslehitabar, H.

2006-01-01

Palladium-103 ( 103 Pd) is a brachytherapy source for cancer treatment. The Monte Carlo codes are usually applied for dose distribution and effect of shieldings. Monte Carlo calculation of dose distribution in water phantom due to a MED3633 103 Pd source is presented in this work. Materials and Methods: The dose distribution around the 10 3Pd Model MED3633 located in the center of 30*30*30 m 3 water phantom cube was calculated using MCNP code by the Monte Carlo method. The percentage depth dose variation along the different axis parallel and perpendicular to the source was also calculated. Then, the isodose curves for 100%, 75%, 50% and 25% percentage depth dose and dosimetry parameters of TG-43 protocol were determined. Results: The results show that the Monte Carlo Method could calculate dose deposition in high gradient region, near the source, accurately. The isodose curves and dosimetric characteristics obtained for MED3633 103 Pd source are in good agreement with published results. Conclusion: The isodose curves of the MED3633 103 Pd source have been derived form dose calculation by MCNP code. The calculated dosimetry parameters for the source agree quite well with their Monte Carlo calculated and experimental measurement values

Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

International Nuclear Information System (INIS)

Thiagu Supramaniam

2007-01-01

The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

Visualization of geometry and tally data using MCNP and Justine

International Nuclear Information System (INIS)

Cox, L.J.; Favorite, J.A.

1999-01-01

The Monte Carlo N-Particle (MCNP) transport code is a general-purpose code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport, including the capability to calculate eigenvalues for neutron-multiplying systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori. Justine is the graphical user interface and problem setup tool for the Los Alamos Radiation Modeling Interactive Environment (LARAMIE). Its purpose is to serve as a convenient and very general interface for setting up physics calculations and linking together the disparate radiation transport codes under a single front-end. Currently, the LARAMIE system includes MCNP and the deterministic transport code suit DANTSYS (ONEDANT, TWODANT, and THREEDANT, for one-, two-, and three-dimensional geometries, respectively). Justine is currently available through the Radiation Safety Information Computational Center to members of the criticality safety community for evaluation and use. The authors will demonstrate the capabilities of both codes for visualization of geometries and results from a variety of criticality problems

Development of burnup methods and capabilities in Monte Carlo code RMC

International Nuclear Information System (INIS)

She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

2013-01-01

Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Uranium Metal, Oxide, and Solution Systems on the High Performance Computing Platform Moonlight

International Nuclear Information System (INIS)

Chapman, Bryan Scott; MacQuigg, Michael Robert; Wysong, Andrew Russell

2016-01-01

In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k eff .

Analysis of Topaz-II reactor performance using MCNP and TFEHX

International Nuclear Information System (INIS)

Lee, H.H.; Klein, A.C.

1993-01-01

Data reported by Russian scientist and engineers for the TOPAZ-II Space Nuclear Power is compared with analytical results calculated using the Monte Carlo Neutron and Photon (MCNP) and TFEHX computer codes. The results of these comparisons show good agreement with the TOPAZ-II neutronics, thermionic and thermal hydraulics performance. A detailed description of the TOPAZ-II reactor and of the TFE should enhance the performance of the both codes in modeling the reactor and TFE performances

Status of Monte Carlo at Los Alamos

International Nuclear Information System (INIS)

Thompson, W.L.; Cashwell, E.D.

1980-01-01

At Los Alamos the early work of Fermi, von Neumann, and Ulam has been developed and supplemented by many followers, notably Cashwell and Everett, and the main product today is the continuous-energy, general-purpose, generalized-geometry, time-dependent, coupled neutron-photon transport code called MCNP. The Los Alamos Monte Carlo research and development effort is concentrated in Group X-6. MCNP treats an arbitrary three-dimensional configuration of arbitrary materials in geometric cells bounded by first- and second-degree surfaces and some fourth-degree surfaces (elliptical tori). Monte Carlo has evolved into perhaps the main method for radiation transport calculations at Los Alamos. MCNP is used in every technical division at the Laboratory by over 130 users about 600 times a month accounting for nearly 200 hours of CDC-7600 time

Load balancing in highly parallel processing of Monte Carlo code for particle transport

International Nuclear Information System (INIS)

Higuchi, Kenji; Takemiya, Hiroshi; Kawasaki, Takuji

1998-01-01

In parallel processing of Monte Carlo (MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)

MCNP modelling of scintillation-detector gamma-ray spectra from natural radionuclides

NARCIS (Netherlands)

Hendriks, Peter; Maucec, M; de Meijer, RJ

gamma-ray spectra of natural radionuclides are simulated for a BGO detector in a borehole geometry using the Monte Carlo code MCNP. All gamma-ray emissions of the decay of K-40 and the series of Th-232 and U-238 are used to describe the source. A procedure is proposed which excludes the

Monte Carlo electron/photon transport

International Nuclear Information System (INIS)

Mack, J.M.; Morel, J.E.; Hughes, H.G.

1985-01-01

A review of nonplasma coupled electron/photon transport using Monte Carlo method is presented. Remarks are mainly restricted to linerarized formalisms at electron energies from 1 keV to 1000 MeV. Applications involving pulse-height estimation, transport in external magnetic fields, and optical Cerenkov production are discussed to underscore the importance of this branch of computational physics. Advances in electron multigroup cross-section generation is reported, and its impact on future code development assessed. Progress toward the transformation of MCNP into a generalized neutral/charged-particle Monte Carlo code is described. 48 refs

Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

Directory of Open Access Journals (Sweden)

Giang Phan

2017-01-01

Full Text Available Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.

Modeling of a planning system in radiotherapy and Nuclear Medicine using the MCNP6 code

International Nuclear Information System (INIS)

Massicano, Felipe

2015-01-01

Cancer therapy has many branches and one of them is the use of radiation sources as treatment leading method. Radiotherapy and nuclear medicine are examples of these treatment types. For using the ionization radiation as main tool for the therapy, there is the need of crafting many treatment simulation in order to maximum the tumoral tissue dose without surpass the dose limit in health tissue surrounding. Treatment planning systems (TPS) are systems which have the purpose of simulating these therapy types. Nuclear medicine and radiotherapy have many distinct features linked to the therapy mode and consequently they have different TPS destined for each. The radiotherapy TPS is more developed than the nuclear medicine TPS and by that reason the development of a TPS that was similar to the radiotherapy TPS, but enough generic for include other therapy types, it will contribute with significant advances in nuclear medicine and in others therapy types with radiation. Based on this, the goal of work was to model a TPS that utilizes the Monte Carlo N-Particle Transport code (MCNP6) in order to simulate radiotherapy therapy, nuclear medicine therapy and with potential for simulating other therapy types too. The result of this work was the creation of a Framework in Java language, object oriented, named IBMC which will assist in the development of new TPS with MCNP6 code. The IBMC allowed to develop rapidly and easily TPS for radiotherapy and nuclear medicine and the results were validated with systems already consolidated. The IBMC showed high potential for developing TPS by new therapy types. (author)

Application of the NJOY code for unresolved resonance treatment in the MCNP utility code

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J. . E-mail addresses of corresponding authors: mmilos@vin.bg.ac.yu , vujic@nuc.berkeley.edu ,; Milosevic, M.; Vujic, J.)

2005-01-01

There are numerous uncertainties in the prediction of neutronic characteristics of reactor cores, particularly in the case of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) benchmark core and the new procedures and cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and it is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include: MOCUP - a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the ENDF/B-VI evaluations; and KWO2 - a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 group library. Calculations made for the ENHS benchmark have shown that the differences between the results obtained using different code systems and cross section libraries are significant and should be taken into account in assessing the quality of nuclear data libraries. (author)

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

International Nuclear Information System (INIS)

Dawahra, S.; Khattab, K.

2011-01-01

Highlights: → The MCNP4C code was used to calculate the power distribution in 3-D geometry in the MNSR reactor. → 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 kW. - Abstract: 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 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. 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.

Monte Carlo calculation of dose rate conversion factors for external exposure to photon emitters in soil

CERN Document Server

Clouvas, A; Antonopoulos-Domis, M; Silva, J

2000-01-01

The dose rate conversion factors D/sub CF/ (absorbed dose rate in air per unit activity per unit of soil mass, nGy h/sup -1/ per Bq kg/sup -1/) are calculated 1 m above ground for photon emitters of natural radionuclides uniformly distributed in the soil. Three Monte Carlo codes are used: 1) The MCNP code of Los Alamos; 2) The GEANT code of CERN; and 3) a Monte Carlo code developed in the Nuclear Technology Laboratory of the Aristotle University of Thessaloniki. The accuracy of the Monte Carlo results is tested by the comparison of the unscattered flux obtained by the three Monte Carlo codes with an independent straightforward calculation. All codes and particularly the MCNP calculate accurately the absorbed dose rate in air due to the unscattered radiation. For the total radiation (unscattered plus scattered) the D/sub CF/ values calculated from the three codes are in very good agreement between them. The comparison between these results and the results deduced previously by other authors indicates a good ag...

Vectorization, parallelization and porting of nuclear codes (porting). Progress report fiscal 1998

International Nuclear Information System (INIS)

Nemoto, Toshiyuki; Kawai, Wataru; Ishizuki, Shigeru; Kawasaki, Nobuo; Kume, Etsuo; Adachi, Masaaki; Ogasawara, Shinobu

2000-03-01

Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system, the AP3000 system and the Paragon system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 12 codes in fiscal 1998. These results are reported in 3 parts, i.e., the vectorization and parallelization on vector processors part, the parallelization on scalar processors part and the porting part. In this report, we describe the porting. In this porting part, the porting of Monte Carlo N-Particle Transport code MCNP4B2 and Reactor Safety Analysis code RELAP5 on the AP3000 are described. In the vectorization and parallelization on vector processors part, the vectorization of General Tokamak Circuit Simulation Program code GTCSP, the vectorization and parallelization of Molecular Dynamics Ntv Simulation code MSP2, Eddy Current Analysis code EDDYCAL, Thermal Analysis Code for Test of Passive Cooling System by HENDEL T2 code THANPACST2 and MHD Equilibrium code SELENEJ on the VPP500 are described. In the parallelization on scalar processors part, the parallelization of Monte Carlo N-Particle Transport code MCNP4B2, Plasma Hydrodynamics code using Cubic Interpolated propagation Method PHCIP and Vectorized Monte Carlo code (continuous energy model/multi-group model) MVP/GMVP on the Paragon are described. (author)

Criticality benchmarks for COG: A new point-wise Monte Carlo code

International Nuclear Information System (INIS)

Alesso, H.P.; Pearson, J.; Choi, J.S.

1989-01-01

COG is a new point-wise Monte Carlo code being developed and tested at LLNL for the Cray computer. It solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) charged particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems. However, its point-wise cross-sections also make it effective for a wide variety of criticality problems. COG has some similarities to a number of other computer codes used in the shielding and criticality community. These include the Lawrence Livermore National Laboratory (LLNL) codes TART and ALICE, the Los Alamos National Laboratory code MCNP, the Oak Ridge National Laboratory codes 05R, 06R, KENO, and MORSE, the SACLAY code TRIPOLI, and the MAGI code SAM. Each code is a little different in its geometry input and its random-walk modification options. Validating COG consists in part of running benchmark calculations against critical experiments as well as other codes. The objective of this paper is to present calculational results of a variety of critical benchmark experiments using COG, and to present the resulting code bias. Numerous benchmark calculations have been completed for a wide variety of critical experiments which generally involve both simple and complex physical problems. The COG results, which they report in this paper, have been excellent

Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Uranium Metal, Oxide, and Solution Systems on the High Performance Computing Platform Moonlight

Energy Technology Data Exchange (ETDEWEB)

Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); MacQuigg, Michael Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wysong, Andrew Russell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-04-21

In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k_eff.

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

MCNP and other nuclear codes output graphical representation using python scripts

International Nuclear Information System (INIS)

Cadenas Mendicoa, A. M.

2016-01-01

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)

Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C

International Nuclear Information System (INIS)

Suman, H.; Kharita, M. H.; Yousef, S.

2008-02-01

In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

Directory of Open Access Journals (Sweden)

Chapoutier Nicolas

2017-01-01

Full Text Available In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics. Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

An approach to design a 90Sr radioisotope thermoelectric generator using analytical and Monte Carlo methods with ANSYS, COMSOL, and MCNP.

Science.gov (United States)

Khajepour, Abolhasan; Rahmani, Faezeh

2017-01-01

In this study, a 90 Sr radioisotope thermoelectric generator (RTG) with power of milliWatt was designed to operate in the determined temperature (300-312K). For this purpose, the combination of analytical and Monte Carlo methods with ANSYS and COMSOL software as well as the MCNP code was used. This designed RTG contains 90 Sr as a radioisotope heat source (RHS) and 127 coupled thermoelectric modules (TEMs) based on bismuth telluride. Kapton (2.45mm in thickness) and Cryotherm sheets (0.78mm in thickness) were selected as the thermal insulators of the RHS, as well as a stainless steel container was used as a generator chamber. The initial design of the RHS geometry was performed according to the amount of radioactive material (strontium titanate) as well as the heat transfer calculations and mechanical strength considerations. According to the Monte Carlo simulation performed by the MCNP code, approximately 0.35 kCi of 90 Sr is sufficient to generate heat power in the RHS. To determine the optimal design of the RTG, the distribution of temperature as well as the dissipated heat and input power to the module were calculated in different parts of the generator using the ANSYS software. Output voltage according to temperature distribution on TEM was calculated using COMSOL. Optimization of the dimension of the RHS and heat insulator was performed to adapt the average temperature of the hot plate of TEM to the determined hot temperature value. This designed RTG generates 8mW in power with an efficiency of 1%. This proposed approach of combination method can be used for the precise design of various types of RTGs. Copyright © 2016 Elsevier Ltd. All rights reserved.

Radiosteoplasty study in animal bone and radiodosimetric evaluation using Monte Carlo code

Energy Technology Data Exchange (ETDEWEB)

Silveira, Marcia Flavia; Campos, Tarcisio Passos Ribeiro [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: marciaflaviafisio@gmail.com; campos@nuclear.ufmg.br

2007-07-01

The radiosteoplasty is a procedure that consists of the injection of a radioactive biomaterial incorporated to the bone cement into the osseous structure affected by cancer. This technique has been developed with the major objective to control the tumor or the regional bone metastasis (in situ) besides pain reduction and structural resistance increasing. In the present study the radiosteoplasty is applied to the bovine and swine bones in vitro using non-radioactive cement. The objective is to know the spatial distribution of the cold compound (non radioactive) in pig and ox bones after implant. A 2 mm needle was introduced into the cortical bone previously perforated. The distribution of this biomaterial was observed trough radiological images obtained just after the compound application. Recent dosimetric studies using Monte Carlo N-Particle method (MCNP-5) concluded that the spatial dose distribution is suitable for the protocol namely radiosteoplasty applied to treat bone tumors on superior and inferior members. The Monte Carlo method simulates the present process and it is particularly interesting tool to solve the complex photon and electron particle transport problems that can not be modeled by codes based on deterministic methods. These related radiodosimetric studies are presented and discussed. (author)

BOT3P: a mesh generation software package for the transport analysis codes Dort, Tort, Twodant, Threedant and MCNP

International Nuclear Information System (INIS)

Orsi, R.

2003-01-01

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)

MCNP/X TRANSPORT IN THE TABULAR REGIME

Energy Technology Data Exchange (ETDEWEB)

HUGHES, H. GRADY [Los Alamos National Laboratory

2007-01-08

The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

LEU-fueled SLOWPOKE-2 modelling with MCNP4A

International Nuclear Information System (INIS)

Pierre, J.R.M.; Bonin, H.W.J.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fueled SLOWPOKE-2 research reactor at Royal Military College,excess reactivity measurements were conducted over a range of temperature and power. Given the advance in computer technology, the use of Monte Carlo N-Particle Transport Code System MCNP 4A appeared possible for the simulation of the LEU-fueled SLOWPOKE-2 reactor core, and this work demonstrates that this is indeed the case. MCNP 4A is a full three dimensional program allowing the user to enter a large amount of complexity. The limit on the geometry complexity is the computing time required to achieve a reasonable standard deviation. To this point several models of the SLOWPOKE-2 have been developed giving some insight on the sensitivity of the code. MCNP4A can use various cross section libraries. The aim of this work is to calculate accurately the reactivity of the core and reproduce The temperature trend of the reactivity. The model preserved as much as possible the details of the core and facility in order to allow further study in the flux mapping

A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System

Energy Technology Data Exchange (ETDEWEB)

C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler

1998-10-01

The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.

A PC version of the Monte Carlo criticality code OMEGA

International Nuclear Information System (INIS)

Seifert, E.

1996-05-01

A description of the PC version of the Monte Carlo criticality code OMEGA is given. The report contains a general description of the code together with a detailed input description. Furthermore, some examples are given illustrating the generation of an input file. The main field of application is the calculation of the criticality of arrangements of fissionable material. Geometrically complicated arrangements that often appear inside and outside a reactor, e.g. in a fuel storage or transport container, can be considered essentially without geometrical approximations. For example, the real geometry of assemblies containing hexagonal or square lattice structures can be described in full detail. Moreover, the code can be used for special investigations in the field of reactor physics and neutron transport. Many years of practical experience and comparison with reference cases have shown that the code together with the built-in data libraries gives reliable results. OMEGA is completely independent on other widely used criticality codes (KENO, MCNP, etc.), concerning programming and the data base. It is a good practice to run difficult criticality safety problems by different independent codes in order to mutually verify the results. In this way, OMEGA can be used as a redundant code within the family of criticality codes. An advantage of OMEGA is the short calculation time: A typical criticality safety application takes only a few minutes on a Pentium PC. Therefore, the influence of parameter variations can simply be investigated by running many variants of a problem. (orig.)

Monte Carlo calculations of neutron and gamm-ray energy spectra for fusion-reactor shield design: comparison with experiment

International Nuclear Information System (INIS)

Santoro, R.T.; Barnes, J.M.

1983-08-01

Neutron and gamma-ray spectra resulting from the interactions of approx. 14-MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree within 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra is also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE

ARCHER, a new Monte Carlo software tool for emerging heterogeneous computing environments

International Nuclear Information System (INIS)

Xu, X. George; Liu, Tianyu; Su, Lin; Du, Xining; Riblett, Matthew; Ji, Wei; Gu, Deyang; Carothers, Christopher D.; Shephard, Mark S.; Brown, Forrest B.; Kalra, Mannudeep K.; Liu, Bob

2015-01-01

Highlights: • A fast Monte Carlo based radiation transport code ARCHER was developed. • ARCHER supports different hardware including CPU, GPU and Intel Xeon Phi coprocessor. • Code is benchmarked again MCNP for medical applications. • A typical CT scan dose simulation only takes 6.8 s on an NVIDIA M2090 GPU. • GPU and coprocessor-based codes are 5–8 times faster than the CPU-based codes. - Abstract: The Monte Carlo radiation transport community faces a number of challenges associated with peta- and exa-scale computing systems that rely increasingly on heterogeneous architectures involving hardware accelerators such as GPUs and Xeon Phi coprocessors. Existing Monte Carlo codes and methods must be strategically upgraded to meet emerging hardware and software needs. In this paper, we describe the development of a software, called ARCHER (Accelerated Radiation-transport Computations in Heterogeneous EnviRonments), which is designed as a versatile testbed for future Monte Carlo codes. Preliminary results from five projects in nuclear engineering and medical physics are presented

Criticality safety validation of MCNP5 using continuous energy libraries

International Nuclear Information System (INIS)

Salome, Jean A.D.; Pereira, Claubia; Assuncao, Jonathan B.A.; Veloso, Maria Auxiliadora F.; Costa, Antonella L.; Silva, Clarysson A.M. da

2013-01-01

The study of subcritical systems is very important in the design, installation and operation of various devices, mainly nuclear reactors and power plants. The information generated by these systems guide the decisions to be taken in the executive project, the economic viability and the safety measures to be employed in a nuclear facility. Simulating some experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments, the code MCNP5 was validated to nuclear criticality analysis. Its continuous libraries were used. The average values and standard deviation (SD) were evaluated. The results obtained with the code are very similar to the values obtained by the benchmark experiments. (author)

Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

International Nuclear Information System (INIS)

Ivanov, A.; Sanchez, V.; Hoogenboom, J. E.

2012-01-01

As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

Energy Technology Data Exchange (ETDEWEB)

Ivanov, A.; Sanchez, V. [Karlsruhe Inst. of Technology, Inst. for Neutron Physics and Reactor Technology, Herman-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Hoogenboom, J. E. [Delft Univ. of Technology, Faculty of Applied Sciences, Mekelweg 15, 2629 JB Delft (Netherlands)

2012-07-01

As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

2014-07-01

A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm{sup 2}s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

Burnup calculation of a CANDU6 reactor using the Serpent and MCNP6 codes

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2014-01-01

A study of fuel burnup for the CANDU6 reactor is carried out to validate the most recent versions of the probabilistic transport code (MCNP6) and the continuous energy burnup calculation code (Serpent). These two codes allow for 3-D geometry calculation accounting for a detailed analysis without unit-cell homogenization. On the other hand, the WIMS-AECL computer program is used to model neutron transport in nuclear-reactor lattices for design, safety analysis, and operation. It works with two-dimensional regions and can perform collision probability calculations for a periodic structure of the lattice cell. In the present work, the multiplication factor, the total flux and fuel burnup could be calculated for a CANDU6 nuclear reactor based on the GENTILLY-2 core design. The MCNP6 and Serpent codes provide a calculation of the track length estimated flux per neutron source. This estimated flux is then scaled with normalization to the reactor power in order to provide a flux in unit of n/cm 2 s. Good agreement is observed between the actual total flux calculated by MCNP6, Serpent and WIMS-AECL. The effective multiplication factors of the whole core CANDU6 reactor are further calculated as a function of burnup and further compared to those calculated by WIMS-AECL where excellent agreement is also obtained. (author)

A comparison of lattice parameters for CANDU-type lattices obtained using MCNP, WIMS, and WIMS with resonance reaction rates from MCNP

International Nuclear Information System (INIS)

Craig, D.S.

1989-03-01

The Monte Carlo code MCNP was used to check the accuracy of the WIMS calculation of the resolved resonance capture rate in CANDU-type lattices. Reactivities, relative conversion ratios, and fast fission factors are compared with experiments. Values of ρ 28 and reaction rates for U-238 are given as a function of position in the fuel bundle. A check was made on the correction made in WIMS to allow for endcaps on the fuel bundles. (26 refs)

Monte Carlo simulations of the pulsed thermal neutron flux in two-region hydrogenous systems (using standard MCNP data libraries)

International Nuclear Information System (INIS)

Wiacek, U.; Krynicka, E.

2005-02-01

Monte Carlo simulations of the pulsed neutron experiment in two- region systems (two concentric spheres and two coaxial finite cylinders) are presented. The MCNP code is used. Aqueous solutions of H 3 BO 3 or KCl are used in the inner region. The outer region is the moderator of Plexiglas. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances are used. The time-dependent thermal neutron transport is simulated when the inner region has a constant size and the external size of the surrounding outer region is variable. The time decay constant of the thermal neutron flux in the system is found in each simulation. The results of the simulations are compared with results of real pulsed neutron experiments on the corresponding systems. (author)

Installation of MCNP on 64-bit parallel computers

International Nuclear Information System (INIS)

Meginnis, A.B.; Hendricks, J.S.; McKinney, G.W.

1995-01-01

The Monte Carlo radiation transport code MCNP has been successfully ported to two 64-bit workstations, the SGI and DEC Alpha. We found the biggest problem for installation on these machines to be Fortran and C mismatches in argument passing. Correction of these mismatches enabled, for the first time, dynamic memory allocation on 64-bit workstations. Although the 64-bit hardware is faster because 8-bytes are processed at a time rather than 4-bytes, we found no speed advantage in true 64-bit coding versus implicit double precision when porting an existing code to the 64-bit workstation architecture. We did find that PVM multiasking is very successful and represents a significant performance enhancement for scientific workstations

Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Del Valle G, E.

2017-09-01

The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)

Benchmarking the cad-based attila discrete ordinates code with experimental data of fusion experiments and to the results of MCNP code in simulating ITER

International Nuclear Information System (INIS)

Youssef, M. Z.

2007-01-01

Attila is a newly developed finite element code based on Sn neutron, gamma, and charged particle transport in 3-D geometry in which unstructured tetrahedral meshes are generated to describe complex geometry that is based on CAD input (Solid Works, Pro/Engineer, etc). In the present work we benchmark its calculation accuracy by comparing its prediction to the measured data inside two experimental mock-ups bombarded with 14 MeV neutrons. The results are also compared to those based on MCNP calculations. The experimental mock-ups simulate parts of the International Thermonuclear Experimental Reactor (ITER) in-vessel components, namely: (1) the Tungsten mockup configuration (54.3 cm x 46.8 cm x 45 cm), and (2) the ITER shielding blanket followed by the SCM region (simulated by alternating layers of SS316 and copper). In the latter configuration, a high aspect ratio rectangular streaming channel was introduced (to simulate steaming paths between ITER blanket modules) which ends with a rectangular cavity. The experiments on these two fusion-oriented integral experiments were performed at the Fusion Neutron Generator (FNG) facility, Frascati, Italy. In addition, the nuclear performance of the ITER MCNP 'Benchmark' CAD model has been performed with Attila to compare its results to those obtained with CAD-based MCNP approach developed by several ITER participants. The objective of this paper is to compare results based on two distinctive 3-D calculation tools using the same nuclear data, FENDL2.1, and the same response functions of several reaction rates measured in ITER mock-ups and to enhance confidence from the international neutronics community in the Attila code and how it can precisely quantify the nuclear field in large and complex systems, such as ITER. Attila has the advantage of providing a full flux mapping visualization everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. In addition, the

MCNP trademark directions

International Nuclear Information System (INIS)

Hendricks, J.S.

1994-01-01

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

Contributon Monte Carlo

International Nuclear Information System (INIS)

Dubi, A.; Gerstl, S.A.W.

1979-05-01

The contributon Monte Carlo method is based on a new recipe to calculate target responses by means of volume integral of the contributon current in a region between the source and the detector. A comprehensive description of the method, its implementation in the general-purpose MCNP code, and results of the method for realistic nonhomogeneous, energy-dependent problems are presented. 23 figures, 10 tables

Neutron reflector design with Californium 252 neutron for Boron neutron chapter therapy facility using MCNP5 simulation method

International Nuclear Information System (INIS)

Muhammad Fakhrurreza; Kusminanto; Y Sardjono

2014-01-01

In this research has made a reflector design to provide beams of Neutron for BNCT with Californium-252 radioactive source. This collimator is useful to obtain optimum epithermal neutron flux with the smallest impurity radiation (thermal neutron, fast neutron, and gamma). The design process is done using Monte Carlo N-Particle simulation version 5 (MCNP5) code to calculate the neutron flux tally form. The chosen reflector design is the reflectors which use material such as BeO ceramic with 13 cm thick. Moderator use sulfur material with the slope angle of the cone is 30°. From the calculation result, it is obtained that Reflector with 1 gram Californium-252 source can produce a neutron output thermal which has thermal neutron specification 2.23189 x 10 9 n/s.cm 2 , epithermal neutron 3.51548 x 10 9 n/s.cm 2 , and fast neutron 4.82241 x 10 9 n/s.cm 2 From the result, it needs additional collimator because the BNCT requirement. (author)

Monte Carlo Particle Lists: MCPL

DEFF Research Database (Denmark)