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Sample records for fotons usando mcnp

  1. Simulation of radiation transport using MCNP for a teletherapy machine; Simulacion del transporte de radiacion usando MCNP para una maquina de teleterapia

    Energy Technology Data Exchange (ETDEWEB)

    Flores O, F.E.; Mireles G, F.; Davila R, J.I.; Pinedo V, J.L.; Risorios M, C.; Lopez del Rio, H. [UAZ, Unidad Academica de Estudios Nucleares, 98068 Zacatecas (Mexico)

    2008-07-01

    The MCNP code is used to simulate the radiation transport taking as tools the transport physics of each particle, either photon, neutron or electron, and the generation of random numbers. Developed in the Los Alamos National Laboratory, this code has been used thoroughly with great success, because the results of the simulations are broadly validated with representative experiments. In the one present work the room of radiotherapy of the Institute Zacatecano of the Tumor it is simulated, located in the city of Zacatecas where one is Theratron 780C machine manufactured by MSD Nordion, with the purpose of estimating the contribution to the dose that would be received in different points of the structure, included three directly under the source. Three results of analytical calculations for points located at different distances from the source are presented, and they are compared against those obtained by the simulation. Its are also presented results for the simulation of 10 points more distributed around the source. (Author)

  2. 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

  3. 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

  4. 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

  5. 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)

  6. MCNP Progress & Performance Improvements

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-04-14

    Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.

  7. 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

  8. MCNP6 Status

    Energy Technology Data Exchange (ETDEWEB)

    Goorley, John T. [Los Alamos National Laboratory

    2012-06-25

    We, the development teams for MCNP, NJOY, and parts of ENDF, would like to invite you to a proposed 3 day workshop October 30, 31 and November 1 2012, to be held at Los Alamos National Laboratory. At this workshop, we will review new and developing missions that MCNP6 and the underlying nuclear data are being asked to address. LANL will also present its internal plans to address these missions and recent advances in these three capabilities and we will be interested to hear your input on these topics. Additionally we are interested in hearing from you additional technical advances, missions, concerns, and other issues that we should be considering for both short term (1-3 years) and long term (4-6 years)? What are the additional existing capabilities and methods that we should be investigating? The goal of the workshop is to refine priorities for mcnp6 transport methods, algorithms, physics, data and processing as they relate to the intersection of MCNP, NJOY and ENDF.

  9. 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.)

  10. MCNP trademark Monte Carlo: A precis of MCNP

    International Nuclear Information System (INIS)

    Adams, K.J.

    1996-01-01

    MCNP trademark is a general purpose three-dimensional time-dependent neutron, photon, and electron transport code. It is highly portable and user-oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 years and an ongoing commitment to excellence

  11. 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).

  12. SUPERIMPOSED MESH PLOTTING IN MCNP

    Energy Technology Data Exchange (ETDEWEB)

    J. HENDRICKS

    2001-02-01

    The capability to plot superimposed meshes has been added to MCNP{trademark}. MCNP4C featured a superimposed mesh weight window generator which enabled users to set up geometries without having to subdivide geometric cells for variance reduction. The variance reduction was performed with weight windows on a rectangular or cylindrical mesh superimposed over the physical geometry. Experience with the new capability was favorable but also indicated that a number of enhancements would be very beneficial, particularly a means of visualizing the mesh and its values. The mathematics for plotting the mesh and its values is described here along with a description of other upgrades.

  13. Foton-M3 Unmanned Russian Research Satellite- Development, Implementation and Operations

    Science.gov (United States)

    Ilyin, Eugene A.; Skidmore, Michael G.

    2008-06-01

    The Foton-M3 spacecraft launched from Baikonur Cosmodrome (Kazakhstan) on 14 September 2007 and landed 12 days later approximately 130 km south of Kustanay, Northern Kazakhstan. Following the successful National Aeronautics and Space Administration (NASA) and Institute for Biomedical Problems (IMBP) collaboration on the Russian Foton-M2 spaceflight (June 2005), IMBP invited NASA to continue and broaden its participation in four Russian biomedical studies on the Foton-M3 spaceflight. Where the Foton-M2 collaboration had been accomplished without an exchange of funds, the basis for the ongoing bilateral interaction on Foton-M3 was both a cooperative Space Act Agreement and a NASA contract with IMBP. As in Foton-M2, NASA scientists agreed to focus their efforts on research that would be complementary and would facilitate the accomplishment of the original Russian science goals. Foton-M3 hardware enhancements included NASA inserts installed in the IMBP flight hardware to provide programmable in-flight video recording for newts and geckos, drinking water for the geckos, and a preflight "shower" of Bromodeoxyuridine (BrdU) for the newts.

  14. Datalogger usando nios ii

    OpenAIRE

    Campoverde Rugel, Luis Enrique; Velásquez Vargas, Washington Adrián; Ponguillo, Ronald

    2013-01-01

    El presente proyecto consiste en la implementación de un Datalogger utilizando el microprocesador NIOS II el cual fue embebido en el FPGA CYCLONE II que se encuentra integrada en la tarjeta de desarrollo ALTERA DE2, el cual obtiene datos de distintos sensores y los almacena en una tarjeta SD Card. Para la realización del proyecto se aplican cuatro etapas. La primera etapa está basada en obtener los datos mediante el uso de sensores y la transmisión usando un PIC, la siguiente etapa se basa...

  15. MCNP6 Cosmic-Source Option

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, Gregg W [Los Alamos National Laboratory; Armstrong, Hirotatsu [Los Alamos National Laboratory; James, Michael R [Los Alamos National Laboratory; Clem, John [University of Delaware, BRI; Goldhagen, Paul [DHS, National Urban Security Technology Laboratory

    2012-06-19

    MCNP is a Monte Carlo radiation transport code that has been under development for over half a century. Over the last decade, the development team of a high-energy offshoot of MCNP, called MCNPX, has implemented several physics and algorithm improvements important for modeling galactic cosmic-ray (GCR) interactions with matter. In this presentation, we discuss the latest of these improvements, a new Cosmic-Source option, that has been implemented in MCNP6.

  16. 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

  17. Foton 11: ESA investigates further the space environment and its impact on organisms

    Science.gov (United States)

    1997-10-01

    Scientific research conducted under space conditions can provide new insight into how processes occur on Earth and organisms function. The unmanned Foton spacecraft has been used since 1988 to conduct such investigations. Now on its 11th mission and the fifth in which ESA has taken part, Foton is carrying some 80 kg of ESA payload: two ESA research facilities (an incubator and an experiment holder on the outside of the spacecraft) are on board along with 12 scientific experiments. The French space agency (CNES) and the German space agency (DARA) also have payload on the spacecraft. ESA's space-qualified incubator, called Biobox, keeps organisms at predefined conditions. During this mission, the three Biobox experiments are looking at the reaction of bone cells in microgravity. The second ESA facility, a pan-shaped container called Biopan attached to the outside of Foton, is used to expose experiment samples directly to the space environment in order to study the impact of space's extreme temperatures, ultraviolet and cosmic radiation, and near-perfect vacuum. On this mission, the six Biopan experiments are concentrating on exobiology, radiation biology and material science. Biopan has a motor-driven, hinged lid and is equipped with devices and sensors that measure the various aspects of the environment to which the experiments are subjected. Once Foton is in orbit, a telecommand is sent from ground and the lid opens to expose the samples to the environment. At the end of the mission, another command is sent and the lid closes. Since Biopan is on the outside of Foton, it also has its own ablative heat shield to protect the facility and samples during the spacecraft's re-entry and landing. Other ESA experiments on board Foton are looking into the effects of weightlessness on bacteria, the biological clocks of beetles and the aging of fruitflies. The scientific investigators responsible for the ESA experiments are from research institutes and universities in Belgium

  18. Silver Jaanuse fotonäitus "Töö(b)luus"

    Index Scriptorium Estoniae

    2010-01-01

    Silver Jaanuse fotonäitus "Töö(b)luus" Telliskivi Loomelinnakus (Tallinn) 4. septembrist 4. oktoobrini 2010. Fotodel dokumenteeritakse ühe ehitusmaterjalide poe igapäevaelu kahe ja poole aasta jooksul

  19. The comparison of MCNP perturbation technique with MCNP difference method in critical calculation

    International Nuclear Information System (INIS)

    Liu Bin; Lv Xuefeng; Zhao Wei; Wang Kai; Tu Jing; Ouyang Xiaoping

    2010-01-01

    For a nuclear fission system, we calculated Δk eff , which arise from system material composition changes, by two different approaches, the MCNP perturbation technique and the MCNP difference method. For every material composition change, we made four different runs, each run with different cycles or each cycle generating different neutrons, then we compared the two Δk eff that are obtained by two different approaches. As a material composition change in any particular cell of the nuclear fission system is small compared to the material compositions in the whole nuclear fission system, in other words, this composition change can be treated as a small perturbation, the Δk eff results obtained from the MCNP perturbation technique are much quicker, much more efficient and reliable than the results from the MCNP difference method. When a material composition change in any particular cell of the nuclear fission system is significant compared to the material compositions in the whole nuclear fission system, both the MCNP perturbation technique and the MCNP difference method can give satisfactory results. But for the run with the same cycles and each cycle generating the same neutrons, the results obtained from the MCNP perturbation technique are systemically less than the results obtained from the MCNP difference method. To further confirm our calculation results from the MCNP4C, we run the exact same MCNP4C input file in MCNP5, the calculation results from MCNP5 are the same as the calculation results from MCNP4C. We need caution when using the MCNP perturbation technique to calculate the Δk eff as the material composition change is large compared to the material compositions in the whole nuclear fission system, even though the material composition changes of any particular cell of the fission system still meet the criteria of MCNP perturbation technique.

  20. 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)

  1. Status Report on the MCNP 2020 Initiative

    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)

    2017-10-02

    The discussion below provides a status report on the MCNP 2020 initiative. It includes discussion of the history of MCNP 2020, accomplishments during 2013-17, priorities for near-term development, other related efforts, a brief summary, and a list of references for the plans and work accomplished.

  2. MCNP trademark Software Quality Assurance plan

    International Nuclear Information System (INIS)

    Abhold, H.M.; Hendricks, J.S.

    1996-04-01

    MCNP is a computer code that models the interaction of radiation with matter. MCNP is developed and maintained by the Transport Methods Group (XTM) of the Los Alamos National Laboratory (LANL). This plan describes the Software Quality Assurance (SQA) program applied to the code. The SQA program is consistent with the requirements of IEEE-730.1 and the guiding principles of ISO 900

  3. How to Build MCNP 6.2

    Energy Technology Data Exchange (ETDEWEB)

    Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-13

    This presentation describes how to build MCNP 6.2. MCNP®* 6.2 can be compiled on Macs, PCs, and most Linux systems. It can also be built for parallel execution using both OpenMP and Messing Passing Interface (MPI) methods. MCNP6 requires Fortran, C, and C++ compilers to build the code.

  4. 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).

  5. 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.

  6. MCNP4A: Features and philosophy

    International Nuclear Information System (INIS)

    Hendricks, J.S.

    1993-01-01

    This paper describes MCNP, states its philosophy, introduces a number of new features becoming available with version MCNP4A, and answers a number of questions asked by participants in the workshop. MCNP is a general-purpose three-dimensional neutron, photon and electron transport code. Its philosophy is ''Quality, Value and New Features.'' Quality is exemplified by new software quality assurance practices and a program of benchmarking against experiments. Value includes a strong emphasis on documentation and code portability. New features are the third priority. MCNP4A is now available at Los Alamos. New features in MCNP4A include enhanced statistical analysis, distributed processor multitasking, new photon libraries, ENDF/B-VI capabilities, X-Windows graphics, dynamic memory allocation, expanded criticality output, periodic boundaries, plotting of particle tracks via SABRINA, and many other improvements. 23 refs

  7. 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

  8. 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.

  9. 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)

  10. 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

  11. 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

  12. TET2MCNP: 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 2 MCNP, for converting TetGen-generated tetrahedral-mesh models to the Abacus input file format. TET 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 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 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 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

  13. 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.

  14. CTEx Beowulf cluster for MCNP performance

    Energy Technology Data Exchange (ETDEWEB)

    Gonzaga, Roberto N.; Amorim, Aneuri S. de; Balthar, Mario Cesar V. [Centro Tecnologico do Exercito (CTEx), Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil)

    2011-07-01

    This work is an introduction to the CTEx Nuclear Defense Department's Beowulf Cluster. Building a Beowulf Cluster is a complex learning process that greatly depends upon your hardware and software requirements. The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built in Red Hat's Fedora Linux operating system personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation transport calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. The pre compiled parallel binary version of MCNP uses the Message-Passing Interface (MPI) protocol. The use of this pre compiled parallel version of MCNP5 with the MPI protocol on a small, heterogeneous computing cluster built from Red Hat's Fedora Linux operating system PCs is the subject of this work. (author)

  15. CTEx Beowulf cluster for MCNP performance

    International Nuclear Information System (INIS)

    Gonzaga, Roberto N.; Amorim, Aneuri S. de; Balthar, Mario Cesar V.

    2011-01-01

    This work is an introduction to the CTEx Nuclear Defense Department's Beowulf Cluster. Building a Beowulf Cluster is a complex learning process that greatly depends upon your hardware and software requirements. The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built in Red Hat's Fedora Linux operating system personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation transport calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. The pre compiled parallel binary version of MCNP uses the Message-Passing Interface (MPI) protocol. The use of this pre compiled parallel version of MCNP5 with the MPI protocol on a small, heterogeneous computing cluster built from Red Hat's Fedora Linux operating system PCs is the subject of this work. (author)

  16. Fotonäitus, mis kujutab ka tühjust / Oliver Õunmaa

    Index Scriptorium Estoniae

    Õunmaa, Oliver

    2011-01-01

    Eesti fotograafi Krista Möldri ja soome fotograafi Kalle Kataila fotonäitus "European Eyes on Japan" Temnikova & Kasela galeriis (Müürivahe 22, Lastekodu 1) 8.12.2011-22.01.2012. Näituse kuraator Mikiko Kikuta. Fotograafide tööd valmisid Tohuku regiooni Akita prefektuuris

  17. Fotonäitus koduta loomade heaks / Indrek Galetin, Kerttu Rakke, Angela Aak ... [jt.

    Index Scriptorium Estoniae

    Galetin, Indrek

    2009-01-01

    Rotermani kvartalis 22.05.2009 avatud Indrek Galetini fotonäitus "Amour - Lemmikloomade eri". Kodutute loomade abistamiseks 20.07.2009 oksjonil müüdavate piltide valmimist kommenteerivad fotograaf ja modellid Kerttu Rakke, Angela Aak ja Max Kaur

  18. SABRINA, Geometry Plot Program for MCNP

    International Nuclear Information System (INIS)

    SEIDL, Marcus

    2003-01-01

    1 - Description of program or function: SABRINA is an interactive, three-dimensional, geometry-modeling code system, primarily for use with CCC-200/MCNP. SABRINA's capabilities include creation, visualization, and verification of three-dimensional geometries specified by either surface- or body-base combinatorial geometry; display of particle tracks are calculated by MCNP; and volume fraction generation. 2 - Method of solution: Rendering is performed by ray tracing or an edge and intersection algorithm. Volume fraction calculations are made by ray tracing. 3 - Restrictions on the complexity of the problem: A graphics display with X Window capability is required

  19. Adjoint-Based Uncertainty Quantification with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States)

    2011-09-01

    This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence in the simulation is acquired.

  20. Application of MCNP{trademark} to computed tomography in medicine

    Energy Technology Data Exchange (ETDEWEB)

    Brockhoff, R.C. [KSU (United States); Estes, G.P.; Hills, C.R. [Mason and Hanger (United States); Demarco, J.J.; Solberg, T.D. [California Univ., Los Angeles, CA (United States)

    1996-03-01

    The MCNP{trademark} code has been used to simulate CT scans of the MIRD human phantom. In addition. an actual CT scan of a patient was used to create an MCNP geometry, and this geometry was computationally ``CT scanned`` using MCNP to reconstruct CT images. The results show that MCNP can be used to model the human body based on data obtained from CT scans and to simulate CT scans that are based on these or other models.

  1. Methodology for converting CT medical images to MCNP input using the Scan2MCNP system

    International Nuclear Information System (INIS)

    Boia, L.S.; Silva, A.X.; Cardoso, S.C.; Castro, R.C.

    2009-01-01

    This paper develops a methodology for the application software Scan2MCNP, which converts medical images DICOM (Digital Imaging and Communications in Medicine) for MCNP input file. The Scan2MCNP handles, processes and executes the medical images generated by CT equipment, allowing the user to perform the selection and parameterization of the study area in question (tissues and organs). The details of these worked in medical imaging software, therefore, will be converted to equity to the process of language analysis of MCNP radiation transport, through the generation of a code input file. With this file, it is possible to simulate any situation/problem of the type and level of radiation to the proposed treatment chosen by the medical staff responsible for the patient. Within a computational process oriented, the Scan2MCNP can contribute along with other software that has been used recently in the area of medical physics, to improve the levels of quality and precision of radiotherapy treatments. In this work, medical images DICOM of the Anthropomorphic Rando Phantom were used in the process of analysis and development of computer software Scan2MCNP. However, it emphasized that the software is successful in certain situations, depending upon a number of auxiliary procedures and software that can help in the solution of certain problems in the natural radiation treatment or express agility by the team of medical physics. (author)

  2. Study of function response of a detector HPGe to photons of reaction {sup 19}F(p,{alpha}{gamma}){sup 16}O; Estudo da funcao resposta de um detetor HPGe a fotons da reacao {sup 19}F(p,{alpha}{gamma}){sup 16}O

    Energy Technology Data Exchange (ETDEWEB)

    Tridapalli, D.B

    2006-07-01

    follows the oxygen nuclei in its trajectory until photon emission, considering the changes in spatial distribution of the exact interaction point in the target with incident proton energy due to the large resonance width and proton energy loss, and {sup 16}O energy loss and multiple scattering until decay, in the different target layers. Using the detector response functions calculated by MCNP5 simulations, the relative intensities of the three gamma rays were obtained by a least square fit of the response functions, taking into account the Doppler broadening and shift for each gamma ray, to the data in the experimental pulse-height spectrum. (author) [Portuguese] O Laboratorio do Acelerador Linear do Instituto de Fisica da Universidade de Sao Paulo esta desenvolvendo um estudo sobre a eficiencia e a funcao resposta de detetores de HPGe a fotons de alta energia. Neste trabalho foi estudada a funcao resposta de fotons provenientes da reacao {sup 19}F(p,{alpha}{gamma}){sup 16}O. Na reacao {sup 19}F(p,{alpha}{gamma}){sup 16}O podem ser observados os fotons caracteristicos de 6,1 MeV, 6,9 MeV e 7,1 MeV. Fotons com esta energia possuem um grande potencial para varias aplicacoes importantes, como Proton Induced Gamma-ray Emission (PIGE), gamagrafia e procedimentos de clibracao. No caso de procedimentos de calibracao essa reacao possui um diferencial: o fluxo de fotons observado e bem maior do que em outras reacoes estudadas, tais como {sup 27}Al(p,{gamma}){sup 28}Si e {sup 23}Na(p,{gamma}){sup 24}Mg. Os experimentos foram realizados no Laboratorio de Analise de Materiais por Feixes Ionicos (LAMFI) do IFUSP, utilizando um acelerador electrostatico tipo Pelletron-tandem de 1,7 MV de tensao maxima no terminal. O detetor estudado foi um HPGe do tipo coaxial reverse-electrode closed-end com 72,5 mm de diametro e 60,5 mm de comprimento e 60% da eficiencia de um detetor de NaI para fotons de 1,332 MeV. O detetor foi posicionado a 0 deg. em relacao a linha de feixe. A corrente de

  3. Using Machine Learning to Predict MCNP Bias

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-01-09

    For many real-world applications in radiation transport where simulations are compared to experimental measurements, like in nuclear criticality safety, the bias (simulated - experimental keff) in the calculation is an extremely important quantity used for code validation. The objective of this project is to accurately predict the bias of MCNP6 [1] criticality calculations using machine learning (ML) algorithms, with the intention of creating a tool that can complement the current nuclear criticality safety methods. In the latest release of MCNP6, the Whisper tool is available for criticality safety analysts and includes a large catalogue of experimental benchmarks, sensitivity profiles, and nuclear data covariance matrices. This data, coming from 1100+ benchmark cases, is used in this study of ML algorithms for criticality safety bias predictions.

  4. Flux at a point in MCNP

    International Nuclear Information System (INIS)

    Cashwell, E.D.; Schrandt, R.G.

    1980-01-01

    The current state of the art of calculating flux at a point with MCNP is discussed. Various techniques are touched upon, but the main emphasis is on the fast improved version of the once-more-collided flux estimator, which has been modified to treat neutrons thermalized by the free gas model. The method is tested on several problems on interest and the results are presented

  5. 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.

  6. Data analysis and visualization in MCNP trademark

    International Nuclear Information System (INIS)

    Waters, L.S.

    1994-01-01

    There are many situations where the user may wish to go beyond current MCNP capabilities. For example, data produced by the code may need formatting for input into an external graphics package. Limitations on disk space may hinder writing out large PTRAK files. Specialized data analysis routines may be needed to model complex experimental results. One may wish to produce particle histories in a format not currently available in the code. To address these and other similar concerns a new capability in MCNP is being tested. A number of real, integer, logical and character variables describing the current and past characteristics of a particle are made available online to the user in three subroutines. The type of data passed can be controlled by cards in the INP file. The subroutines otherwise are empty, and the user may code in any desired analysis. A new MCNP executable is produced by compiling these subroutines and linking to a library which contains the object files for the rest of the code

  7. Monte Carlo parameter studies and uncertainty analyses with MCNP5

    International Nuclear Information System (INIS)

    Brown, F. B.; Sweezy, J. E.; Hayes, R.

    2004-01-01

    A software tool called mcnp p study has been developed to automate the setup, execution, and collection of results from a series of MCNP5 Monte Carlo calculations. This tool provides a convenient means of performing parameter studies, total uncertainty analyses, parallel job execution on clusters, stochastic geometry modeling, and other types of calculations where a series of MCNP5 jobs must be performed with varying problem input specifications. (authors)

  8. MCNP software quality : then and now /

    Energy Technology Data Exchange (ETDEWEB)

    Giesler, G. C. (Gregg Carl)

    2001-01-01

    MCNP is the Monte Carlo N-Particle radiation transport code whose history dates back more than half a century to the early days of computing. From a simple beginning, its uses have grown to include fields such as criticality safety, radiation shielding, oil well logging, and medical imaging and diagnostics and an international user community of over 3000 users. This large user community could only happen by the maintainance of sofware quality throughout its history. This paper will describe how the quality was maintained in the past, how the process is being improved today, and directions for future efforts.

  9. Validation of MCNP4A for repository scattered radiation analysis

    International Nuclear Information System (INIS)

    Haas, M.N.; Su, S.

    1998-02-01

    Comparison is made between experimentally determined albedo (scattered) radiation and MCNP4A predictions in order to provide independent validation for repository shielding analysis. Both neutron and gamma scattered radiation fields from concrete ducts are compared in this paper. Satisfactory agreement is found between actual and calculated results with conservative values calculated by the MCNP4A code for all conditions

  10. Development and improvement for MCNP-3B interactive plotter

    International Nuclear Information System (INIS)

    Gao Yanfeng

    1996-01-01

    The author briefly explains the development and improvement for the MCNP-3B interactive plotter. It describes the functions of geometry visualization and tally result plot, and introduces the progresses in user interface, process display and surface matching. The construction of MCNP-3B/PC is given

  11. A New Developed Interface for CAD/MCNP Data Conversion

    International Nuclear Information System (INIS)

    Noha Shaahan; Fukuzo Masuda; Hesham Nasif; Masao Yamada; Hidenori Sawamura; Hidetsugu Morota; Satoshi Sato; Hiromasa Iida; Takeo Nishitani

    2006-01-01

    In a complex and huge system as in ITER fusion reactor, the creation of the geometrical input data of Monte Carlo (MC) codes such as MCNP is a highly exhausting task. Accordingly, it is a general approach to shift the geometric modeling into a computer aided design (CAD) system and to use an interface, which performs the exchange of CAD data into a representation appropriate for MC code. We have developed efficient algorithms and computer code, which are used to convert Parasolid format CAD files including solid and void data into MCNP input data. The CAD-MCNP conversion processes include creating surface equations; determining surface senses; constructing cell geometry and creating MCNP input file. This paper describes the basic algorithms used for the CAD/MCNP interface along with some applications for different geometries. (authors)

  12. Estimation and interpretation of keff confidence intervals in MCNP

    International Nuclear Information System (INIS)

    Urbatsch, T.J.

    1995-01-01

    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 individual estimator with the smallest variance. The importance of MCNP's batch statistics is demonstrated by an investigation of the effects of individual estimator variance bias on the combination of estimators, both heuristically with the analytical study and emprically with MCNP

  13. An assessment of the MCNP4C weight window

    International Nuclear Information System (INIS)

    Culbertson, Christopher N.; Hendricks, John S.

    1999-01-01

    A new, enhanced weight window generator suite has been developed for MCNP version 4C. The new generator correctly estimates importances in either a user-specified, geometry-independent, orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. The new generator is applied in a set of five variance reduction problems. The improved generator is compared with the weight window generator applied in MCNP4B. The benefits of the new methodology are highlighted, along with a description of its limitations. The authors also provide recommendations for utilization of the weight window generator

  14. Deposition of CdTe films under microgravity: Foton M3 mission

    Energy Technology Data Exchange (ETDEWEB)

    Benz, K.W.; Croell, A. [Freiburger Materialforschungszentrum FMF, Albert-Ludwigs-Universitaet Freiburg (Germany); Zappettini, A.; Calestani, D. [CNR Parma, Instituto Materiali Speciali per Elettronica e Magnetismo IMEM, Fontani Parma (Italy); Dieguez, E. [Universidad Autonoma de Madrid (Spain). Departamento de Fisica de Materiales; Carotenuto, L.; Bassano, E. [Telespazio Napoli, Via Gianturco 31, 80146 Napoli (Italy); Fiederle, M.

    2009-10-15

    Experiments of deposition of CdTe films have been carried out under microgravity in the Russian Foton M3 mission. The influence of gravity has been studied with these experiments and compared to the results of simulations. The measured deposition rate could be confirmed by the theoretical results for lower temperatures. For higher temperatures the measured thickness of the deposited films was larger compared to the theoretical data. (copyright 2009 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  15. Computational radiology and imaging with the MCNP Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

    MCNP, a 3D coupled neutron/photon/electron Monte Carlo radiation transport code, is currently used in medical applications such as cancer radiation treatment planning, interpretation of diagnostic radiation images, and treatment beam optimization. This paper will discuss MCNP`s current uses and capabilities, as well as envisioned improvements that would further enhance MCNP role in computational medicine. It will be demonstrated that the methodology exists to simulate medical images (e.g. SPECT). Techniques will be discussed that would enable the construction of 3D computational geometry models of individual patients for use in patient-specific studies that would improve the quality of care for patients.

  16. Criticality calculations with MCNP{trademark}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A. [New Mexico Univ., Albuquerque, NM (United States)

    1994-06-06

    With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.

  17. The BIOPAN experiment MARSTOX II of the FOTON M-3 mission

    Science.gov (United States)

    Rettberg, P.; Moeller, R.; Rabbow, E.; Panitz, C.; Horneck, G.; Meyer, C.; Lammer, H.; Douki, T.; Cadet, J.

    2008-09-01

    The experiment MARSTOX II on FOTON M-3 mission (September 14 - 26, 2007) was a further step in the study of the Responses of Organisms to the Martian Environment (ROME) which already started with first ground-based experiments in Mars simulation chambers and with the space experiment MARSTOX I, flown in 2005 in the ESA facility BIOPAN (Fig. 1) on FOTON M-2. The survivability of bacterial spores of B. subtilis, a well-characterized model system for highly resistant microorganisms, was investigated under the extreme environmental conditions as they exist on the surface of Mars. By use of exterrestrial UV radiation and cut-off filters the photoprotection and potential UV-phototoxicity of different minerals of the Martian soil were investigated.In MARSTOX II two further aspects were addressed (i) the influence of different concentrations of dust in the Martian atmosphere, which change the solar irradiance on the surface significantly compared to vacuum exposure under the same conditions (experiment parts 'DUST MARS' and 'DUST SPACE'), and (ii) the survivability of spores under martian atmosphere and pressure exposed to a mars-like spectral irradiance compared to vacuum exposure under the same conditions (experiment parts 'MIXED MARS' and 'MIXED SPACE') (Fig. 2 and 3). After exposure to space during the FOTON M-3 mission the sample analysis was performed at CEA in Grenoble, F, and at DLR in Cologne, D, together with parallel samples from the corresponding ground control experiment performed in the space simulation facilities at DLR. As biological endpoints in these investigations survival and UV-induced DNAphotoproducts were analysed.From the results of MARSTOX II the following conclusions can be drawn: (i) Spores mixed with martian soil analogue are protected only to a low degree against UV radiation. The protective effect of several defined layers of spores mixed with Martian soil analogue were quantified. (ii) The two investigated martian soil analogues, MRS07 (47

  18. Spectral measurements in critical assemblies: MCNP specifications and calculated results

    Energy Technology Data Exchange (ETDEWEB)

    Stephanie C. Frankle; Judith F. Briesmeister

    1999-12-01

    Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.

  19. KENO2MCNP, Version 5L, Conversion of Input Data between KENOV.a and MCNP File Formats

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: The KENO2MCNP program was written to convert KENO V.a input files to MCNP Format. This program currently only works with KENO Va geometries and will not work with geometries that contain more than a single array. A C++ graphical user interface was created that was linked to Fortran routines from KENO V.a that read the material library and Fortran routines from the MCNP Visual Editor that generate the MCNP input file. Either SCALE 5.0 or SCALE 5.1 cross section files will work with this release. 2 - Methods: The C++ binary executable reads the KENO V.a input file, the KENO V.a material library and SCALE data libraries. When an input file is read in, the input is stored in memory. The converter goes through and loads different sections of the input file into memory including parameters, composition, geometry information, array information and starting information. Many of the KENO V.a materials represent compositions that must be read from the KENO V.a material library. KENO2MCNP includes the KENO V.a FORTRAN routines used to read this material file for creating the MCNP materials. Once the file has been read in, the user must select 'Convert' to convert the file from KENO V.a to MCNP. This will generate the MCNP input file along with an output window that lists the KENO V.a composition information for the materials contained in the KENO V.a input file. The program can be run interactively by clicking on the executable or in batch mode from the command prompt. 3 - Restrictions on the complexity of the problem: Not all KENO V.a input files are supported. Only one array is allowed in the input file. Some of the more complex material descriptions also may not be converted

  20. Verification of MCNP6.2 for Nuclear Criticality Safety Applications

    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)

    2017-05-10

    Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suites were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.

  1. Depletion analysis of the UMLRR reactor core using MCNP6

    Science.gov (United States)

    Odera, Dim Udochukwu

    Accurate knowledge of the neutron flux and temporal nuclide inventory in reactor physics calculations is necessary for a variety of application in nuclear engineering such as criticality safety, safeguards, and spent fuel storage. The Monte Carlo N- Particle (MCNP6) code with integrated buildup depletion code (CINDER90) provides a high-fidelity tool that can be used to perform 3D, full core simulation to evaluate fissile material utilization, and nuclide inventory calculations as a function of burnup. The University of Massachusetts Lowell Research Reactor (UMLRR) reactor has been modeled with the deterministic based code, VENTURE and with an older version of MCNP (MCNP5). The MIT developed MCODE (MCNP ORIGEN DEPLETION CODE) was used previously to perform some limited depletion calculations. This work chronicles the use of MCNP6, released in June 2013, to perform coupled neutronics and depletion calculation. The results are compared to previously benchmarked results. Furthermore, the code is used to determine the ratio of fission products 134Cs and 137Cs (burnup indicators), and the resultant ratio is compared to the burnup of the UMLRR.

  2. 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.)

  3. Criticality calculations with MCNP{sup TM}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Mendius, P.W. [ed.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

    1994-08-01

    The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.

  4. Features of MCNP6 Relevant to Medical Radiation Physics

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, H. Grady III [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory

    2012-08-29

    MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo code for simulating the transport of neutrons, photons, electrons, positrons, and more recently other fundamental particles and heavy ions. Over many years MCNP has found a wide range of applications in many different fields, including medical radiation physics. In this presentation we will describe and illustrate a number of significant recently-developed features in the current version of the code, MCNP6, having particular utility for medical physics. Among these are major extensions of the ability to simulate large, complex geometries, improvement in memory requirements and speed for large lattices, introduction of mesh-based isotopic reaction tallies, advances in radiography simulation, expanded variance-reduction capabilities, especially for pulse-height tallies, and a large number of enhancements in photon/electron transport.

  5. 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

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

  7. MCNP load balancing and fault tolerance with PVM

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, G.W.

    1995-07-01

    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.

  8. MCNP speed advances for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Goorley, J.T.; McKinney, G.; Adams, K.; Estes, G.

    1998-04-01

    The Boron Neutron Capture Therapy (BNCT) treatment planning process of the Beth Israel Deaconess Medical Center-M.I.T team relies on MCNP to determine dose rates in the subject's head for various beam orientations. In this time consuming computational process, four or five potential beams are investigated. Of these, one or two final beams are selected and thoroughly evaluated. Recent advances greatly decreased the time needed to do these MCNP calculations. Two modifications to the new MCNP4B source code, lattice tally and tracking enhancements, reduced the wall-clock run times of a typical one million source neutrons run to one hour twenty five minutes on a 200 MHz Pentium Pro computer running Linux and using the GNU FORTRAN compiler. Previously these jobs used a special version of MCNP4AB created by Everett Redmond, which completed in two hours two minutes. In addition to this 30% speedup, the MCNP4B version was adapted for use with Parallel Virtual Machine (PVM) on personal computers running the Linux operating system. MCNP, using PVM, can be run on multiple computers simultaneously, offering a factor of speedup roughly the same as the number of computers used. With two 200 MHz Pentium Pro machines, the run time was reduced to forty five minutes, a 1.9 factor of improvement over the single Linux computer. While the time of a single run was greatly reduced, the advantages associated with PVM derive from using computational power not already used. Four possible beams, currently requiring four separate runs, could be run faster when each is individually run on a single machine under Windows NT, rather than using Linux and PVM to run one after another with each multiprocessed across four computers. It would be advantageous, however, to use PVM to distribute the final two beam orientations over four computers

  9. MCNP speed advances for boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Goorley, J.T.; McKinney, G.; Adams, K.; Estes, G.

    1998-04-01

    The Boron Neutron Capture Therapy (BNCT) treatment planning process of the Beth Israel Deaconess Medical Center-M.I.T team relies on MCNP to determine dose rates in the subject`s head for various beam orientations. In this time consuming computational process, four or five potential beams are investigated. Of these, one or two final beams are selected and thoroughly evaluated. Recent advances greatly decreased the time needed to do these MCNP calculations. Two modifications to the new MCNP4B source code, lattice tally and tracking enhancements, reduced the wall-clock run times of a typical one million source neutrons run to one hour twenty five minutes on a 200 MHz Pentium Pro computer running Linux and using the GNU FORTRAN compiler. Previously these jobs used a special version of MCNP4AB created by Everett Redmond, which completed in two hours two minutes. In addition to this 30% speedup, the MCNP4B version was adapted for use with Parallel Virtual Machine (PVM) on personal computers running the Linux operating system. MCNP, using PVM, can be run on multiple computers simultaneously, offering a factor of speedup roughly the same as the number of computers used. With two 200 MHz Pentium Pro machines, the run time was reduced to forty five minutes, a 1.9 factor of improvement over the single Linux computer. While the time of a single run was greatly reduced, the advantages associated with PVM derive from using computational power not already used. Four possible beams, currently requiring four separate runs, could be run faster when each is individually run on a single machine under Windows NT, rather than using Linux and PVM to run one after another with each multiprocessed across four computers. It would be advantageous, however, to use PVM to distribute the final two beam orientations over four computers.

  10. Reactor physics verification of the MCNP6 unstructured mesh capability

    International Nuclear Information System (INIS)

    Burke, T. P.; Kiedrowski, B. C.; Martz, R. L.; Martin, W. R.

    2013-01-01

    The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

  11. Estimation and interpretation of keff confidence intervals in MCNP

    International Nuclear Information System (INIS)

    Urbatsch, T.J.

    1995-11-01

    MCNP's criticality methodology and some basic statistics are reviewed. Confidence intervals are discussed, as well as how to build them and their importance in the presentation of a Monte Carlo result. The combination of MCNP's three k eff estimators is shown, theoretically and empirically, by statistical studies and examples, to be the best k eff estimator. The method of combining estimators is based on a solid theoretical foundation, namely, the Gauss-Markov Theorem in regard to the least squares method. The confidence intervals of the combined estimator are also shown to have correct coverage rates for the examples considered

  12. 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.

  13. Real-time monitoring of genetically modified Chlamydomonas reinhardtii during the Foton M3 space mission

    Science.gov (United States)

    Lambreva, M.; Rea, G.; Antonacci, A.; Serafini, A.; Damasso, M.; Pastorelli, S.; Margonelli, A.; Johanningmeier, U.; Bertalan, I.; Pezzotti, G.; Giardi, M. T.

    2008-09-01

    Long-term space exploration, colonization or habitation requires biological life support systems capable to cope with the deleterious space environment. The use of oxygenic photosynthetic microrganisms is an intriguing possibility mainly for food, O2 and nutraceutical compounds production. The critical points of utilizing plants- or algae-based life support systems are the microgravity and the ionizing radiation, which can influence the performance of these organisms. The aim of the present study was to assess the effects of space environment on the photosynthetic activity of various microrganisms and to select space stresstolerant strains. Photosystem II D1 protein sitedirected and random mutants of the unicellular green alga Chlamydomonas reinhardtii [1] were used as a model system to test and select the amino acid substitutions capable to account for space stress tolerance. We focussed our studies also on the accumulation of the Photosystem II photoprotective carotenoids (the xantophylls violaxanthin, anteraxanthin and zeaxanthin), powerful antioxidants that epidemiological studies demonstrated to be human vision protectors. For this purpose some mutants modified at the level of enzymes involved in the biosynthesis of xanthophylls were included in the study [2]. To identify the consequences of the space environment on the photosynthetic apparatus the changes in the Photosystem II efficiency were monitored in real time during the ESA-Russian Foton- M3 mission in September 2007. For the space flight a high-tech, multicell fluorescence detector, Photo-II, was designed and built by the Centre for Advanced Research in Space Optics in collaboration with Kayser-Italy, Biosensor and DAS. Photo-II is an automatic device developed to measure the chlorophyll fluorescence and to provide a living conditions for several different algae strains (Fig.1). Twelve different C. reinhardti strains were analytically selected and two replications for each strain were brought to space

  14. A program converting MCNP simulation into gamma vision spectra

    International Nuclear Information System (INIS)

    Ni Jianzhong; Liu Jie; Yu Gongshuo; Zhang Jiamei

    2010-01-01

    A program is developed which can convert the energy distribution of photons calculated by MCNP into Gamma Vision spectra, thus, the simulated energy spectra can be displayed and processed with Gamma Vision. The program provides a convenient tool for the theoretical simulation of HPGe γ spectra. (authors)

  15. MCNP6. Simulating Correlated Data in Fission Events

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sood, Avneet [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-03

    This report is a series of slides discussing the MCNP6 code and its status in simulating fission. Applications of interest include global security and nuclear nonproliferation, detection of special nuclear material (SNM), passive and active interrogation techniques, and coincident neutron and photon leakage.

  16. 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

  17. Duplicating MC-15 Output with Python and MCNP

    Energy Technology Data Exchange (ETDEWEB)

    McSpaden, Alexander Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-23

    Two Python scripts have been written that process the output files of MCNP6 into a format that mimics the list-mode output of Los Alamos National Laboratory’s MC-15 and NPOD neutron detection systems. This report details the methods implemented in these scripts and instructions on their use.

  18. 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

  19. Simulations for the neutron detector TETRA with MCNP

    International Nuclear Information System (INIS)

    Testov, D.; Kuznetsova, E.; Wilson, Jh.

    2013-01-01

    To study the nuclear structure of β-delayed neutron precursors at ALTO ISOL-facility at IPN (Orsay), the high efficiency 4π neutron detector TETRA with 3 He filled counters built at JINR (Dubna) was modified. The MCNP simulations to optimize the future configuration were necessary. The details of the calculations and the major results obtained are discussed

  20. The Lichens experiment at Foton M-2 mission: Survival capacity in space

    Science.gov (United States)

    de La Torre, R.; Horneck, G.; Garcia-Sancho, L.

    Lichens are one of the most resistant organisms at Earth They live at very extreme environments in deserts Atacama desert high mountains Himalaya Antarctica Dry Valleys etc This is possible due to the symbiotic relationship between both constituents the algae and the fungui and to their poikilohidric nature characteristic that allows them to survive latent when environmental conditions are very extreme i e when UV radiation is very high temperatures are extreme and dryness exists If humidity returns and temperature tendencies turn near the optimum around 10 C dormant lichens starts to photosynthetice We have selected two epilithic lichen species for the LICHENS experiment which was included at the ESA Biopan-facility located at the outer shell of the satellite Foton M-2 launched into low Earth orbit the 31th of Mai 2005 from Baikonur Russia On of this species was Rhizocarpon geographicum a bipolar epilithic lichen which grows at high mountain regions e g Sierra de Gredos Central Spain with continental climate has been systematically studied in the natural environment Plataforma de Gredos at 2000 m altitude as well as under simulated space conditions at the space simulation facilities of the DLR The sensitivity of the photosynthetic system PSII to the different environmental conditions dryness including vacuum treatment high temperature fluctuations high UV intensity was fluorometrically measured with a MINI PAM Walz Germany The lichen Rhizocarpon geographicum was

  1. Analysis of Cell Proliferation in Newt (Pleurodeles waltl) Tissue Regeneration during Spaceflight in Foton M-2

    Science.gov (United States)

    Almeida, E. A. C.; Roden, C.; Phillips, J. A.; Yusuf, R.; Globus, R. K.; Searby, N.; Vercoutere, W.; Morey-Holton, E.; Tairbekov, M.; Grigoryan, N.; hide

    2006-01-01

    Terrestrial organisms exposed to microgravity during spaceflight expe rience musculoskeletal degeneration. It is still not understood if lo nger-term exposures to microgravity induce degeneration in other tiss ues, and if these effects are also observed in neutrally buoyant aqu atic organisms that may be pre-adapted to mechanical unloading. The " Regeneration" experiment conducted collaboratively between Russian an d US scientists for 16 days in the Russian Foton M-2 spaceflight soug ht to test the hypothesis that microgravity alters the proliferation of cells in regenerating tail tissue of the newt Pleurodeles waltl. Our initial results indicate that we successfUlly delivered the proli feration marker 5-bromo-2'-deoxy Uridine (BrdU) during spaceflight, and that it was incorporated in the nuclei of cells in regenerating tis sues. Cells in spaceflight tail regenerates proliferated at a slight ly slower rate and were more undifferentiated than those in ground sy nchronous controls. In addition, the size of regenerating tails from spaceflight was smaller than synchronous controls. However, onboard temperature recordings show that the temperature in spaceflight was a bout 2 C lower than ground synchronous controls, possibly explaining the observed differences. Additional post-facto ground controls at ma tched temperatures will correctly determine the effects of spaceflig ht on regenerative cell proliferation in the newt.

  2. Bright points and ejections observed on the sun by the KORONAS-FOTON instrument TESIS

    Science.gov (United States)

    Ulyanov, A. S.; Bogachev, S. A.; Kuzin, S. V.

    2010-10-01

    Five-second observations of the solar corona carried out in the FeIX 171 Å line by the KORONAS-FOTON instrument TESIS are used to study the dynamics of small-scale coronal structures emitting in and around coronal bright points. The small-scale structures of the lower corona display complex dynamics similar to those of magnetic loops located at higher levels of the solar corona. Numerous detected oscillating structures with sizes below 10 000 km display oscillation periods from 50 to 350 s. The period distributions of these structures are different for P 150 s, which implies that different oscillation modes are excited at different periods. The small-scale structures generate numerous flare-like events with energies 1024-1026 erg (nanoflares) and with a spatial density of one event per arcsecond or more observed over an area of 4 × 1011 km2. Nanoflares are not associated with coronal bright points, and almost uniformly cover the solar disk in the observation region. The ejections of solar material from the coronal bright points demonstrate velocities of 80-110 km/s.

  3. 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

  4. Possible Improvements to MCNP6 and its CEM/LAQGSM Event-Generators

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-04

    This report is intended to the MCNP6 developers and sponsors of MCNP6. It presents a set of suggested possible future improvements to MCNP6 and to its CEM03.03 and LAQGSM03.03 event-generators. A few suggested modifications of MCNP6 are quite simple, aimed at avoiding possible problems with running MCNP6 on various computers, i.e., these changes are not expected to change or improve any results, but should make the use of MCNP6 easier; such changes are expected to require limited man-power resources. On the other hand, several other suggested improvements require a serious further development of nuclear reaction models, are expected to improve significantly the predictive power of MCNP6 for a number of nuclear reactions; but, such developments require several years of work by real experts on nuclear reactions.

  5. Electron/Photon Verification Calculations Using MCNP4B

    Energy Technology Data Exchange (ETDEWEB)

    D. P. Gierga; K. J. Adams

    1999-04-01

    MCNP4BW was released in February 1997 with significant enhancements to electron/photon transport methods. These enhancements have been verified against a wide range of published electron/photon experiments, spanning high energy bremsstrahlung production to electron transmission and reflection. The impact of several MCNP tally options and physics parameters was explored in detail. The agreement between experiment and simulation was usually within two standard deviations of the experimental and calculational errors. Furthermore, sub-step artifacts for bremsstrahlung production were shown to be mitigated. A detailed suite of electron depth dose calculations in water is also presented. Areas for future code development have also been explored and include the dependence of cell and detector tallies on different bremsstrahlung angular models and alternative variance reduction splitting schemes for bremsstrahlung production.

  6. Critical mass calculations using MCNP: An academic exercise

    International Nuclear Information System (INIS)

    Kastanya, Doddy

    2015-01-01

    Highlights: • Critical mass of Pu-239 is calculated. • MCNP is utilized to demonstrate that sphere is the optimal shape to reach criticality. • The critical masses from five polyhedrons and sphere are compared. - Abstract: In introductory courses for nuclear engineering, the concept of critical dimension and critical mass are introduced. Students are usually taught that the geometrical shape which needs the smallest amount of fissionable material to reach criticality is a sphere. In this paper, this concept is explored further using MCNP code. Five different regular polyhedrons (i.e., the Platonic solids) and a sphere have been examined to demonstrate that sphere is indeed the optimal geometrical shape to minimize the critical mass. For illustration purpose, the fissile isotope used in this study is 239 Pu, with a nominal density of 19.8 g/cm 3

  7. A fast, automated, semideterministic weight windows generator for MCNP

    International Nuclear Information System (INIS)

    Mickael, M.W.

    1995-01-01

    A fast automated method is developed to estimate particle importance in the Los Alamos Carlo code MCNP. It provides an automated and efficient way of predicting and setting up an important map for the weight windows technique. A short analog simulation is first performed to obtain effective group parameters based on the input description of the problem. A solution of the multigroup time-dependent adjoint diffusion equation is then used to estimate particle importance. At any point in space, time, and energy, the particle importance is determined, based on the calculated parameters, and used as the lower limit of the weight window. The method has been tested for neutron, photon, and coupled neutron-photon problems. Significant improvement in the simulation efficiency is obtained using this technique at no additional computer time and with no prior knowledge of the nature of the problem. Moreover, time and angular importance that are not available yet in MCNP are easily implemented in this method

  8. General purpose photoneutron production in MCNP4A

    International Nuclear Information System (INIS)

    Gallmeier, F.X.

    1995-08-01

    A photoneutron production option was implemented in the MCNP4A code, mainly to supply a tool for reactor shielding calculations in beryllium and heavy water environments of complicated three-dimensional geometries. Photoneutron production cross sections for deuterium and beryllium were created. Subroutines were developed to calculate the probability of photoneutron production at photon collision sites and the energy and flight direction of the created photoneutrons. These subroutines were implemented into MCNP4A. Some small program changes were necessary for processing the input to read the photoneutron production cross sections and to install a photoneutron switch. Some arrays were installed or extended to sample photoneutron creation and loss information, and output routines were changed to give the appropriate summary tables. To verify and validate the photoneutron production data and the MCNP4A implementations, the yields of photoneutron sources were calculated and compared with experiments. In the case of deuterium-based photoneutron sources, the calculations agreed well with the experiments; the beryuium-based photoneutron source calculations were up to 30% higher compared with the measurements. More accurate beryllium photoneutron cross sections would be desirable. To apply the developed method to a real shielding problem, the fast neutron fluxes in the heavy-water-filled reflector vessel of the Advanced Neutron Source reactor were investigated and compared with published DORT calculations. Considering the complete independence between the calculations, the merely 10 to 20% lower fluxes obtained with MCNP4A, compared against the DORT results, were more than satisfactory, as the discrepancy is based primarily on differences in the calculated thermal neutron fluxes

  9. MCNP modelling of a combined neutron/gamma counter

    International Nuclear Information System (INIS)

    Bourva, L.C-A.; Croft, S.; 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 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, ε, to, α, the ratio of (α,n) to spontaneous fission neutron emission rate and to f R , the reals coincidence gate utilisation factor, are presented. Sources of the inaccuracy in the calculations have not yet been fully investigated, because of the vast parameter space to be considered, but values of the coincidence gate utilisation factor derived directly from the MCNP data have been found to be overestimated by about 8.2%. Once bias-corrected, the trends of the real coincidence counts rate as a function of sample mass for three types of sample could be matched to experimental results within 0.33%. This result confirms the possible use of MCNP to calculate response trends accurately for a wide variety of source materials, given a limited experimental calibration set

  10. Nuclear densimeter of soil simulated in MCNP-4C code

    Energy Technology Data Exchange (ETDEWEB)

    Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T., E-mail: mario@nuclear.ufmg.b [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear; Silva, Clemente J.G.C., E-mail: clementecarneito@yahoo.com.b [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Dept. de Ciencias Exatas e Tecnologicas

    2009-07-01

    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)

  11. Shielding simulation of the CDTN cyclotron bunker using MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Dalle, Hugo M.; Campolina, Daniel de A.M., E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Div. de Reatores e Radiacoes

    2011-07-01

    The Nuclear Technology Development Centre (CDTN/CNEN) has contracted services from General Electric in order to install a cyclotron for radioisotopes production and PET radiopharmaceutical synthesis. The Monte Carlo code MCNP5 was used to determine the TVL (tenth value layer) of the concrete and verify shielding calculations performed by GE. The simulations results show values of equivalent dose rates in agreement with those calculated using the methodology adopted by GE, the NCRP-144 and the NCRP-51. (author)

  12. 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)

  13. Benchmark study of TRIPOLI-4 through experiment and MCNP codes

    Energy Technology Data Exchange (ETDEWEB)

    Michel, M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Coulon, R. [Canberra France, F-78182 Saint Quentin en Yvelines (France); Normand, S. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Huot, N.; Petit, O. [CEA, DEN DANS, SERMA, F-91191 Gif-sur-Yvette (France)

    2011-07-01

    Reliability on simulation results is essential in nuclear physics. Although MCNP5 and MCNPX are the world widely used 3D Monte Carlo radiation transport codes, alternative Monte Carlo simulation tools exist to simulate neutral and charged particles' interactions with matter. Therefore, benchmark are required in order to validate these simulation codes. For instance, TRIPOLI-4.7, developed at the French Alternative Energies and Atomic Energy Commission for neutron and photon transport, now also provides the user with a full feature electron-photon electromagnetic shower. Whereas the reliability of TRIPOLI-4.7 for neutron and photon transport has been validated yet, the new development regarding electron-photon matter interaction needs additional validation benchmarks. We will thus demonstrate how accurately TRIPOLI-4's 'deposited spectrum' tally can simulate gamma spectrometry problems, compared to MCNP's 'F8' tally. The experimental setup is based on an HPGe detector measuring the decay spectrum of an {sup 152}Eu source. These results are then compared with those given by MCNPX 2.6d and TRIPOLI-4 codes. This paper deals with both the experimental aspect and simulation. We will demonstrate that TRIPOLI-4 is a potential alternative to both MCNPX and MCNP5 for gamma-electron interaction simulation. (authors)

  14. 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)

  15. On the TTB approximation for photon transport in MCNP

    International Nuclear Information System (INIS)

    Ohashi, Atuto

    2001-01-01

    Three dimensional and continuous energy monte carlo code system, MCNP 4 deals with electron transport in addition to neutron and gamma-ray transport. Benchmark experiments involved bremsstrahlung of secondary electron are analyzed by the code MCNP 4, in the following three cases: (1) without approximation for electron pair production, (2) with the TTB approximation (thick-target-bremsstrahlung) for electron pair production, and (3) with secondary electron transport. Bishop et al. measured photon spectrum of gamma-ray (6.1Mev) which is emitted from N-16 in reactor coolant, and penetrating through iron and lead. Johnson et al. measured scattering photon spectrum and doses of capture gamma-ray (∼8Mev) which is emitted from titan and nickel, and penetrating through iron, concrete and lead. Calculation results of MCNP 4 with the secondary electron transport give good agreement with the measured values obtained by these two benchmark experiments, although the TTB approximation calculations overestimate in penetration problem, and underestimate in backscattering problem. (M. Suetake)

  16. 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

  17. Development and application of MCNP auto-modeling tool: Mcam 3.0

    International Nuclear Information System (INIS)

    Liu Xiaoping; Luo Yuetong; Tong Lili

    2005-01-01

    Mcam is abbreviation of 'MCNP Automatic Modeling', which is a CAD interface program of MCNP geometry model based on CAD technology. Making use of existing CAD technology is Mcam's major characteristic. In rough, CAD technology is utilized in the following two ways: (1) Mcam makes it possible to create MCNP geometry model in some CAD software; (2) accelerate creation of MCNP geometry model by inheriting some existing 3D CAD model. The paper gives an introduction of Mcam's major ability: (1) ability to convert CAD model into MCNP geometry model; (2) ability to convert MCNP geometry model into CAD model; (3) ability to construct CAD model. At the end of the paper, several models are given to demonstrate Mcam's different ability respectively

  18. 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)

  19. Peculiarities of lens and tail regeneration detected in newts after spaceflight aboard Foton M3

    Science.gov (United States)

    Grigoryan, Eleonora N.; Almeida, Eduardo; Poplinskaya, Valentina; Novikova, Julia; Domaratskaya, Elena; Aleinikova, Karina; Souza, Kenneth; Skidmore, Mike; Grigoryan, Eleonora N.

    In September 2007 the joint, 12 day long experiment was carried out aboard Russian satellite Foton M3. The goal of the experiment was to study eye lens, tail and forelimb toe regeneration in adult 16 newts (Pl. waltl.) operated 10 days before taking-off. In spaceflight and synchronous ground control we used video recording, temperature and irradiation control, as well as constant availability of thymidine analog BrdU for its absorption via animals' skin. New techniques allowed us to analyze animals' behavior in hyperand microgravity periods of time, to take proper account of spaceflight factors, and measure accumulated pools of DNA-synthesizing cells in regenerating tissues. All tissue specimens obtained from animals were isolated in the day of landing and then prepared for morphological, immunochemical and molecular investigations. Synchronous control was shifted for two days and reproduced flight conditions except changes of gravity influence. As a result in flown animals as compared with synchronous ground control we found lens regeneration of 0.5-1 stage speeded up and an increased BrdU+ (S-phase) cell number in eye cornea, growth zone, limbus and newly forming lens. These features of regeneration were accompanied by an increase of FGF2 expression in eye growth zone and heat shock protein (HSP90) induction purely in retinal macroglial cells of regenerating eyes. Toe regeneration rate was equal and achieved the stage of accomplished healing of amputation area in both groups - "flown" and control animals. We found no essential differences in tail regeneration rate and tail regenerate sizes in the newts exposed to space and on ground. In both groups tail regeneration reached the stage IV-V when tail length and square were around 4.4 mm and 15.5 mm2, correspondingly. However we did observe remarkable changes of tail regenerate form and some of pigmentation. Computer morphometrical analysis showed that only in ground control animals the evident dorso

  20. Review of heavy charged particle transport in MCNP6.2

    Science.gov (United States)

    Zieb, K.; Hughes, H. G.; James, M. R.; Xu, X. G.

    2018-04-01

    The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. This paper discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models' theories are included as well.

  1. 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)

  2. Development of temperature related thermal neutron scattering database for MCNP

    International Nuclear Information System (INIS)

    Mei Longwei; Cai Xiangzhou; Jiang Dazhen; Chen Jingen; Guo Wei

    2013-01-01

    Based on ENDF/B-Ⅶ neutron library, the thermal neutron scattering library S(α, β) for molten salt reactor moderators was developed. The temperatures of this library were chose as the characteristic temperature of the molten salt reactor. The cross section of the thermal neutron scattering of ACE format was investigated, and this library was also validated by the benchmarks of ICSBEP. The uncertainties shown in the validation were in reasonable range when compared with the thermal neutron scattering library tmccs which included in the MCNP data library. It was proved that the thermal neutron scattering library processed in this study could be used in the molten salt reactor design. (authors)

  3. Effect of the MCNP model definition on the computation time

    International Nuclear Information System (INIS)

    Šunka, Michal

    2017-01-01

    The presented work studies the influence of the method of defining the geometry in the MCNP transport code and its impact on the computational time, including the difficulty of preparing an input file describing the given geometry. Cases using different geometric definitions including the use of basic 2-dimensional and 3-dimensional objects and theirs combinations were studied. The results indicate that an inappropriate definition can increase the computational time by up to 59% (a more realistic case indicates 37%) for the same results and the same statistical uncertainty. (orig.)

  4. Radiation calculations using LAHET/MCNP/CINDER90

    International Nuclear Information System (INIS)

    Waters, L.S.

    1993-01-01

    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

  5. MCNP5 CALCULATIONS REPLICATING ARH-600 NITRATE DATA

    Energy Technology Data Exchange (ETDEWEB)

    FINFROCK SH

    2011-10-25

    This report serves to extend the previous document: 'MCNP Calculations Replicating ARH-600 Data' by replicating the nitrate curves found in ARH-600. This report includes the MCNP models used, the calculated critical dimension for each analyzed parameter set, and the resulting data libraries for use with the CritView code. As with the ARH-600 data, this report is not meant to replace the analysis of the fissile systems by qualified criticality personnel. The M CNP data is presented without accounting for the statistical uncertainty (although this is typically less than 0.001) or bias and, as such, the application of a reasonable safety margin is required. The data that follows pertains to the uranyl nitrate and plutonium nitrate spheres, infinite cylinders, and infinite slabs of varying isotopic composition, reflector thickness, and molarity. Each of the cases was modeled in MCNP (version 5.1.40), using the ENDF/B-VI cross section set. Given a molarity, isotopic composition, and reflector thickness, the fissile concentration and diameter (or thicknesses in the case of the slab geometries) were varied. The diameter for which k-effective equals 1.00 for a given concentration could then be calculated and graphed. These graphs are included in this report. The pages that follow describe the regions modeled, formulas for calculating the various parameters, a list of cross-sections used in the calculations, a description of the automation routine and data, and finally the data output. The data of most interest are the critical dimensions of the various systems analyzed. This is presented graphically, and in table format, in Appendix B. Appendix C provides a text listing of the same data in a format that is compatible with the CritView code. Appendices D and E provide listing of example Template files and MCNP input files (these are discussed further in Section 4). Appendix F is a complete listing of all of the output data (i.e., all of the analyzed dimensions and

  6. Visualizing MCNP Tally Segment Geometry and Coupling Results with ABAQUS

    International Nuclear Information System (INIS)

    J. R. Parry; J. A. Galbraith

    2007-01-01

    The Advanced Graphite Creep test, AGC-1, is planned for irradiation in the Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant program. The experiment requires very detailed neutronics and thermal hydraulics analyses to show compliance with programmatic and ATR safety requirements. The MCNP model used for the neutronics analysis required hundreds of tally regions to provide the desired detail. A method for visualizing the hundreds of tally region geometries and the tally region results in 3 dimensions has been created to support the AGC-1 irradiation. Additionally, a method was created which would allow ABAQUS to access the results directly for the thermal analysis of the AGC-1 experiment

  7. 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

  8. MCNP6 and DRiFT modeling efforts for the NEUANCE/DANCE detector array

    Energy Technology Data Exchange (ETDEWEB)

    Pinilla, Maria Isabel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-30

    This report seeks to study and benchmark code predictions against experimental data; determine parameters to match MCNP-simulated detector response functions to experimental stilbene measurements; add stilbene processing capabilities to DRiFT; and improve NEUANCE detector array modeling and analysis using new MCNP6 and DRiFT features.

  9. Delayed Neutron & Gamma Measurements of Special Nuclear Materials and MCNP6 Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Sellers, Madison [Royal Military College of Canada, Kingston, ON (Canada); Goorley, John T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Corcoran, E. C. [Royal Military College of Canada, Kingston, ON (Canada); Kelly, D. G. [Royal Military College of Canada, Kingston, ON (Canada)

    2014-01-21

    Measurements of DG emissions from 0.8 – 1.6 MeV were compared to MCNP6 simulations. Several discrepancies were resolved with use of ENDFVII.1 decay data. Furthermore, MCNP6 was executable with delayed bin fix resolved several line intensity discrepancies.

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

    Energy Technology Data Exchange (ETDEWEB)

    Forster, R.A.; Godfrey, T.N.K.

    1985-01-01

    MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.

  11. Application of MCNP{trademark} to storage facility dose rate assessment

    Energy Technology Data Exchange (ETDEWEB)

    Urban, W.T.; Roberts, R.R.; Estes, G.P.; Taylor, W.M.

    1996-12-31

    The MCNP code is widely used in the determination of neutral particle dose rate analyses. In this paper we examine the application of MCNP to several storage facilities containing special nuclear material, SNM, wherein the neutron dose rate is the primary quantity of interest. In particular, we describe the special geometry, modeling assumptions, and physics considerations encountered in each of three applications.

  12. Testing the Delayed Gamma Capability in MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Weldon, Robert A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fensin, Michael L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKinney, Gregg W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-28

    . We examine five different decay chains (two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes (e.g., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% (28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Finally, hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.

  13. 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

  14. MCNP full-core modeling of the advanced test reactor

    International Nuclear Information System (INIS)

    Kim, S.S.; Jahshan, S.N.; Nielson, R.B.

    1993-01-01

    A full-core Monte Carlo neutron and photon (MCNP) transport model has been completed for the advanced test reactor (ATR) at Idaho National Engineering Laboratory. This new model is a complete three-dimensional model that represents fuel elements, core structures, and target regions in adequate detail. The model can be used in evaluating heating and reaction rates in various target regions of the core. This model is especially useful in physics analysis of an asymmetric experiment loading in the core. The ATR is a light-water-cooled thermal reactor with aluminum/uranium-aluminide fuel plates grouped in arcuate fuel elements that form a serpentine arrangement, as shown in Fig. 1. The core is surrounded by a beryllium reflector. Nine test loops are nestled in the lobes of the serpentine core, and numerous other irradiation holes with varying dimensions and radiation environments are located in the reflector and in the core interior

  15. 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

  16. An enhanced geometry-independent mesh weight window generator for MCNP

    International Nuclear Information System (INIS)

    Evans, T.M.; Hendricks, J.S.

    1997-01-01

    A new, enhanced, weight window generator suite has been developed for MCNP trademark. The new generator correctly estimates importances in either an user-specified, geometry-independent orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. To verify the correctness of the new implementation, comparisons are performed with the analytical solution for the cell importance. Using the new generator, differences between Monte Carlo generated and analytical importances are less than 0.1%. Also, assumptions implicit in the original MCNP generator are shown to be poor in problems with high scattering media. The new generator is fully compatible with MCNP's AVATAR trademark automatic variance reduction method. The new generator applications, together with AVATAR, gives MCNP an enhanced suite of variance reduction methods. The flexibility and efficacy of this suite is demonstrated in a neutron porosity tool well-logging problem

  17. 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.

  18. MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Cox, Lawrence J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-06-12

    This report documents the preparations for and testing of the production release of MCNP6™1.1 beta through RSICC at ORNL. It addresses tests on supported operating systems (Linux, MacOSX, Windows) with the supported compilers (Intel, Portland Group and gfortran). Verification and Validation test results are documented elsewhere. This report does not address in detail the overall packaging of the distribution. Specifically, it does not address the nuclear and atomic data collection, the other included software packages (MCNP5, MCNPX and MCNP6) and the collection of reference documents.

  19. Reconocimiento de Gestos usando Cámaras de Profundidad

    OpenAIRE

    JHONSON GARCÍA, FIORELLA ANNETH

    2017-01-01

    El proyecto pretende detectar un repertorio de gestos a partir de nubes de puntos 3D obtenidas con cámaras de profundidad tipo Kinect. Jhonson García, FA. (2017). Reconocimiento de Gestos usando Cámaras de Profundidad. http://hdl.handle.net/10251/91700 TFGM

  20. A Patch to MCNP5 for Multiplication Inference: Description and User Guide

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, Jr., Clell J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-05-05

    A patch to MCNP5 has been written to allow generation of multiple neutrons from a spontaneous-fission event and generate list-mode output. This report documents the implementation and usage of this patch.

  1. Comparison of CdZnTe neutron detector models using MCNP6 and Geant4

    Science.gov (United States)

    Wilson, Emma; Anderson, Mike; Prendergasty, David; Cheneler, David

    2018-01-01

    The production of accurate detector models is of high importance in the development and use of detectors. Initially, MCNP and Geant were developed to specialise in neutral particle models and accelerator models, respectively; there is now a greater overlap of the capabilities of both, and it is therefore useful to produce comparative models to evaluate detector characteristics. In a collaboration between Lancaster University, UK, and Innovative Physics Ltd., UK, models have been developed in both MCNP6 and Geant4 of Cadmium Zinc Telluride (CdZnTe) detectors developed by Innovative Physics Ltd. Herein, a comparison is made of the relative strengths of MCNP6 and Geant4 for modelling neutron flux and secondary γ-ray emission. Given the increasing overlap of the modelling capabilities of MCNP6 and Geant4, it is worthwhile to comment on differences in results for simulations which have similarities in terms of geometries and source configurations.

  2. Developing an interface between MCNP and McStas for simulation of neutron moderators

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

    2012-01-01

    typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

  3. Tests of shielding effectiveness of Kevlar and Nextel onboard the International Space Station and the Foton-M3 capsule.

    Science.gov (United States)

    Pugliese, M; Bengin, V; Casolino, M; Roca, V; Zanini, A; Durante, M

    2010-08-01

    Radiation assessment and protection in space is the first step in planning future missions to the Moon and Mars, where mission and number of space travelers will increase and the protection of the geomagnetic shielding against the cosmic radiation will be absent. In this framework, the shielding effectiveness of two flexible materials, Kevlar and Nextel, were tested, which are largely used in the construction of spacecrafts. Accelerator-based tests clearly demonstrated that Kevlar is an excellent shield for heavy ions, close to polyethylene, whereas Nextel shows poor shielding characteristics. Measurements on flight performed onboard of the International Space Station and of the Foton-M3 capsule have been carried out with special attention to the neutron component; shielded and unshielded detectors (thermoluminescence dosemeters, bubble detectors) were exposed to a real radiation environment to test the shielding properties of the materials under study. The results indicate no significant effects of shielding, suggesting that thin shields in low-Earth Orbit have little effect on absorbed dose.

  4. Dosimetric characterization of a brachytherapy applicator using MCNP5 modelisation and in-phantom measurements

    Energy Technology Data Exchange (ETDEWEB)

    Gerardy, I. [Institut Superieur Industriel de Bruxelles, 150, Rue Royale, B-1000 Brussels (Belgium)], E-mail: gerardy@isib.be; Rodenas, J. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain); Dycke, M. van [Clinique Saint Jean, Bld du Jardin Botanique, B-1000 Brussels (Belgium); Gallardo, S. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain); Ceccolini, Elisa [Facolta di ingegneria, Alma Mater Studiorum Universita di Bologna (Italy)

    2010-04-15

    A gynaecological applicator consisting of a metallic intra-uterine tube with a plastic vaginal applicator and an HDR Ir-192 source have been simulated with MCNP5 (Monte Carlo code). A solid phantom has been designed to perform measurements around the applicator with radiochromic films. The isodose curves obtained are compared with curves calculated with the F4MESH tally of MCNP5 with a good agreement. A pinpoint ionization chamber has been used to evaluate dose at some reference points.

  5. Validation and verification of MCNP6 as a new simulation tool useful for medical applications

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan G [Los Alamos National Laboratory

    2011-01-06

    MCNP6, the latest and most advanced LANL transport code, representing a merger of MCNP5 and MCNPX has been Validated and Verified (V&V) against different experimental data and results by other codes relevant to medical applications. In the present work, we V&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 well data of interest for medical applications measured on both thin and thick targets and agrees very well with similar results obtained with other codes; MCNP6 may be a very useful tool for medical applications We plan to make MCNP6 available to the public via RSICC at Oak Ridge in the middle of 2011 but we are allowed to provide it to friendly US Beta-users outside LANL already now.

  6. Verification of Unstructured Mesh Capabilities in MCNP6 for Reactor Physics Problems

    International Nuclear Information System (INIS)

    Burke, Timothy P.; Martz, Roger L.; Kiedrowski, Brian C.; Martin, William R.

    2012-01-01

    New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructive Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.

  7. Modelling of a proton spot scanning system using MCNP6

    International Nuclear Information System (INIS)

    Ardenfors, O; Gudowska, I; Dasu, A; Kopeć, M

    2017-01-01

    The aim of this work was to model the characteristics of a clinical proton spot scanning beam using Monte Carlo simulations with the code MCNP6. The proton beam was defined using parameters obtained from beam commissioning at the Skandion Clinic, Uppsala, Sweden. Simulations were evaluated against measurements for proton energies between 60 and 226 MeV with regard to range in water, lateral spot sizes in air and absorbed dose depth profiles in water. The model was also used to evaluate the experimental impact of lateral signal losses in an ionization chamber through simulations using different detector radii. Simulated and measured distal ranges agreed within 0.1 mm for R 90 and R 80 , and within 0.2 mm for R 50 . The average absolute difference of all spot sizes was 0.1 mm. The average agreement of absorbed dose integrals and Bragg-peak heights was 0.9%. Lateral signal losses increased with incident proton energy with a maximum signal loss of 7% for 226 MeV protons. The good agreement between simulations and measurements supports the assumptions and parameters employed in the presented Monte Carlo model. The characteristics of the proton spot scanning beam were accurately reproduced and the model will prove useful in future studies on secondary neutrons. (paper)

  8. Reconocimiento de Huellas Dactilares Usando Características Locales

    Directory of Open Access Journals (Sweden)

    Gualberto Aguilar

    2008-01-01

    Full Text Available El reconocimiento de huellas dactilares es uno de los métodos más populares usados con mayor grado de éxito para la identificación de personas. La huella dactilar tiene características únicas llamadas minucias, las cuales son puntos donde los bordes terminan o se dividen. Los sistemas de identificación que usan patrones biométricos de huella dactilar se denominan AFIS (Sistema de Identificación Automático de Huella Dactilar. En este trabajo se realizó un sistema para reconocimiento de huella dactilar usando combinación de Transformada Rápida de Fourier (FFT con Filtros de Gabor para aclarar la imagen y después un novedoso método para el reconocimiento usando características locales.

  9. Measurements by activation foils and comparative computations by MCNP code

    International Nuclear Information System (INIS)

    Kyncl, J.

    2008-01-01

    Systematic study of the radioactive waste minimisation problem is subject of the SPHINX project. Its idea is that burning or transmutation of the waste inventory problematic part will be realized in a nuclear reactor the fuel of which is in the form of liquid fluorides. In frame of the project, several experiments have been performed with so-called inserted experimental channel. The channel was filled up by the fluorides mixture, surrounded by six fuel assemblies with moderator and placed into LR-0 reactor vessel. This formation was brought to critical state and measurement with activation foil detectors were carried out at selected positions of the inserted channel. Main aim of the measurements was to determine reaction rates for the detectors mentioned. For experiment evaluation, comparative computations were accomplished by code MCNP4a. The results obtained show that very often, computed values of reaction rates differ substantially from the values that were obtained from the experiment. This contribution deals with analysis of the reasons of these differences from the point of view of computations by Monte Carlo method. The analysis of concrete cases shows that the inaccuracy of reaction rate computed is caused mostly by three circumstances:-space region that is occupied by detector is relatively very small;- microscopic effective cross-section R(E) of the reaction changes strongly with energy just in the energy interval that gives the greatest contribution to the reaction; - in the energy interval that gives the greatest contribution to reaction rate, the error of the computed neutron flux is great. These circumstances evoke that the computation of reaction rate with casual accuracy submits extreme demands on computing time. (Author)

  10. Design of large sample silicon ingots irradiation facilities using MCNP

    International Nuclear Information System (INIS)

    Abd EL - Latif, S.S.M.

    2012-01-01

    When silicon is irradiated the objective is to produce number of phosphorus atoms in the target sample in order to obtain a given resistivity after the treatment. The resistivity of the sample is decreased by the transmutation of the silicon, by neutrons to phosphorus. Irradiation is carried out by thermal neutrons. The irradiation of silicon ingot large diameter has been carried out in heavy water research reactor since the thermal neutron flux to the fast neutron flux in order of 1000:1. The neutron spectrum is highly thermalized and some of these neutrons can reach the center of the silicon ingot and gives the radial resistivity gradient in accept range. Due to the disadvantages of heavy water research reactor such as tritium generation as a result of the neutron capture by deuterium. The tritium is radioactive emitting beta particles with a half life of 12.3 years so the heavy water research reactor is closed to avoid the intake of bete particles. The new trend in light water research reactor to design a neutron filter from heavy water or graphite to moderate the neutron to offer neutron spectrum like heavy water reactors, and keep the advantages of light water research reactors such as open pool. In this work we try to use graphite, heavy water and light water to design a neutron filter using the MCNP for different silicon ingot diameter.The light water research reactors can irradiate silicon ingot up to 10 inches diameter with accepted radial resistivity gradient (RRG). Graphite is the best filter in case of 10 inch with maximum radial variation (MRV) 7.564%; Light water is the best filter in case of 6 and 8 inch with MRV 2.197% and 4.85% respectively. In case of 6 and 10 inch Heavy water is the second choice.

  11. MCNP to study the BF3 detection efficiency

    International Nuclear Information System (INIS)

    Castro, Vinicius A.; Cavalieri, Tassio A.; Siqueira, Paulo T.D.; Fedorenko, Giuliana G.; Coelho, Paulo R.P.; Madi Filho, Tufic

    2011-01-01

    One of the main parameters to monitor on the employment of the Boron Neutron Capture Therapy (BNCT) is the thermal neutron flux. It can be performed by different techniques such as the activation analysis and the detection by a Boron Trifluoride detector (BF 3 ). BF 3 detector is a real time neutron flux detector which retrieves results in real time. It is however necessary to study the efficiency of the BF 3 detectors when they are exposed to fields of different neutron energy spectra. BF 3 is known to have high efficiency for thermal neutrons (with energy up to 0.5 eV) due the presence of 10 B atoms in the detector. However, one must also understand how this detector interacts with other neutron energy ranges (epithermal and fast). This work shows the experiment and a set of associated simulations carried out in order to evaluate the BF 3 detector efficiency dependence on neutron energy spectra. A set of experiments was conducted in which a BF 3 detector was submitted to different mixed fields (field containing gamma rays and neutrons). These fields were generated by the interposition of paraffin layers with distinct thicknesses between the Am-Be source and the BF 3 detector. The BF 3 detector responses were recorded according to the number of paraffin planes used. MCNP simulations were also performed to study the detector responses on such experimental conditions. It has been possible to achieve the intended goal of evaluating the BF 3 detector response to different mixed irradiation fields. (author)

  12. 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

  13. Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT

    International Nuclear Information System (INIS)

    Royston, K.; Haghighat, A.; Yi, C.

    2010-01-01

    Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)

  14. MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model

    International Nuclear Information System (INIS)

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

    1999-01-01

    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 N-Particle code (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 (TAP) predict 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 MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. 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

  15. New ICRU quantities for the environmental and individual monitoring. Standardization of individual dosemeters by using external beams of photon radiation; Nuevas magnitudes ICRU para la vigilancia radiologica ambiental e individual. Calibracion de dosimetros personales usando haces externos de fotones

    Energy Technology Data Exchange (ETDEWEB)

    Brosed, A.; Delgado, A.; Granados, C. E.

    1987-07-01

    The quantities introduced by ICRU for the radiological monitoring are commented, specially those implied in individual protection against external photons. A procedure is proposed in order to standardize the individual dosemeters by using the kerma in air references of CIEMAT-JEN. The reference radiation beams are described in connection with ISO standards. Provisional values are selected for the appropriate conversion and correction factors. (Author) 23 refs.

  16. A Test of Reliability of the Personnel Dosimetry Services Authorized by CSN using Photon Beams; Control de los servicios de dosimetria personal autorizados por el CSN, usando haces de fotones

    Energy Technology Data Exchange (ETDEWEB)

    Brosed, A.; Delgado, A.; Granados, C. E.; Lopez Ortiz, G.

    1987-07-01

    In 1987 the Consejo de Seguridad Nuclear (CSN) had eight Personnel Dosimetry Services (PDS) authorized to asses the equivalent doses to the spanish occupationally exposed workers, by means of the readings from the dosemeters wear by them. An audit was carried on the PDS on behalf of CSN under the control of CIEMAT. Batches of dosemeters from each one of the PDS were irradiated to dose equivalent values which were well established by CIEMAT but kept hidden from the PDS. By comparing the true values with those obtained by the PDS, it was possible to evaluate the Services according to the analysis of the quantity Q= I B I -I- S where B is the average of the individual deviations between the dosemeters belonging to the same group and the true value as established by CIEMAT, whereas S is the standard deviation of the values inside of this same group. The results of the evaluation, which was made using the new ICRU quantities for personnel monitoring, are presented. (Author) 8 refs.

  17. Development of a Monte Carlo software to photon transportation in voxel structures using graphic processing units; Desenvolvimento de um software de Monte Carlo para transporte de fotons em estruturas de voxels usando unidades de processamento grafico

    Energy Technology Data Exchange (ETDEWEB)

    Bellezzo, Murillo

    2014-09-01

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo Method (MCM) has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this thesis, the CUBMC code is presented, a GPU-based MC photon transport algorithm for dose calculation under the Compute Unified Device Architecture (CUDA) platform. The simulation of physical events is based on the algorithm used in PENELOPE, and the cross section table used is the one generated by the MATERIAL routine, also present in PENELOPE code. Photons are transported in voxel-based geometries with different compositions. There are two distinct approaches used for transport simulation. The rst of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon ignores the existence of borders and travels in homogeneous fictitious media. The CUBMC code aims to be an alternative of Monte Carlo simulator code that, by using the capability of parallel processing of graphics processing units (GPU), provide high performance simulations in low cost compact machines, and thus can be applied in clinical cases and incorporated in treatment planning systems for radiotherapy. (author)

  18. MCNP Modeling Results for Location of Buried TRU Waste Drums

    International Nuclear Information System (INIS)

    Steinman, D K; Schweitzer, J S

    2006-01-01

    In the 1960's, fifty-five gallon drums of TRU waste were buried in shallow pits on remote U.S. Government facilities such as the Idaho National Engineering Laboratory (now split into the Idaho National Laboratory and the Idaho Completion Project [ICP]). Subsequently, it was decided to remove the drums and the material that was in them from the burial pits and send the material to the Waste Isolation Pilot Plant in New Mexico. Several technologies have been tried to locate the drums non-intrusively with enough precision to minimize the chance for material to be spread into the environment. One of these technologies is the placement of steel probe holes in the pits into which wireline logging probes can be lowered to measure properties and concentrations of material surrounding the probe holes for evidence of TRU material. There is also a concern that large quantities of volatile organic compounds (VOC) are also present that would contaminate the environment during removal. In 2001, the Idaho National Engineering and Environmental Laboratory (INEEL) built two pulsed neutron wireline logging tools to measure TRU and VOC around the probe holes. The tools are the Prompt Fission Neutron (PFN) and the Pulsed Neutron Gamma (PNG), respectively. They were tested experimentally in surrogate test holes in 2003. The work reported here estimates the performance of the tools using Monte-Carlo modelling prior to field deployment. A MCNP model was constructed by INEEL personnel. It was modified by the authors to assess the ability of the tools to predict quantitatively the position and concentration of TRU and VOC materials disposed around the probe holes. The model was used to simulate the tools scanning the probe holes vertically in five centimetre increments. A drum was included in the model that could be placed near the probe hole and at other locations out to forty-five centimetres from the probe-hole in five centimetre increments. Scans were performed with no chlorine in the

  19. MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2

    International Nuclear Information System (INIS)

    Briesmeister, J.F.

    1986-09-01

    This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs

  20. MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

    Science.gov (United States)

    Selcow, Elizabeth C.; Cerbone, Ralph J.; Ludewig, Hans; Mughabghab, Said F.; Schmidt, Eldon; Todosow, Michael; Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.

    1993-01-01

    Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.

  1. MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

    International Nuclear Information System (INIS)

    Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.

    1993-01-01

    Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors

  2. 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)

  3. Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, Lucas P. [Los Alamos National Laboratory; Shores, Erik F. [Los Alamos National Laboratory; Myers, Steven C. [Los Alamos National Laboratory; Felsher, Paul D. [Los Alamos National Laboratory; Garner, Scott E. [Los Alamos National Laboratory; Solomon, Clell J. Jr. [Los Alamos National Laboratory

    2012-08-14

    The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.

  4. Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS

    International Nuclear Information System (INIS)

    Tucker, Lucas P.; Shores, Erik F.; Myers, Steven C.; Felsher, Paul D.; Garner, Scott E.; Solomon, Clell J. Jr.

    2012-01-01

    The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.

  5. MCNP5 study on kinetics parameters of coupled fast-thermal system HERBE

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2011-01-01

    Full Text Available New validation of the well-known Monte Carlo code MCNP5 against measured criticality and kinetics data for the coupled fast-thermal HERBE System at the Reactor B critical assembly is shown in this paper. Results of earlier calculations of these criticality and kinetics parameters, done by combination of transport and diffusion codes using two-dimension geometry model are compared to results of new calculations carried out by the MCNP5 code in three-dimension geometry. Satisfactory agreements in comparison of new results with experimental data, in spite complex heterogeneous composition of the HERBE core, are achieved confirming that MCNP5 code could apply successfully to study on HERBE kinetics parameters after uncertainties in impurities in material compositions and positions of fuel elements in fast zone were removed.

  6. Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP

    Science.gov (United States)

    Bowler, Herbert

    As photons, electrons, and neutrons traverse a medium, they impart their energy in ways that are analytically difficult to describe. Monte Carlo methods provide valuable insight into understanding this behavior, especially when the radiation source or environment is too complex to simplify. This research investigates simulating various radiation sources using the Monte Carlo N-Particle (MCNP) transport code, characterizing their impact on various materials, and comparing the simulation results to general theory and measurements. A total of five sources were of interest: two photon sources of different incident particle energies (3.83 eV and 1.25 MeV), two electron sources also of different energies (30 keV and 100 keV), and a californium-252 (Cf-252) spontaneous fission neutron source. Lateral and vertical programmable metallization cells (PMCs) were developed by other researchers for exposure to these photon and electron sources, so simplified PMC models were implemented in MCNP to estimate the doses and fluences. Dose rates measured around the neutron source and the predicted maximum activity of activation foils exposed to the neutrons were determined using MCNP and compared to experimental results obtained from gamma-ray spectroscopy. The analytical fluence calculations for the photon and electron cases agreed with MCNP results, and differences are due to MCNP considering particle movements that hand calculations do not. Doses for the photon cases agreed between the analytical and simulated results, while the electron cases differed by a factor of up to 4.8. Physical dose rate measurements taken from the neutron source agreed with MCNP within the 10% tolerance of the measurement device. The activity results had a percent error of up to 50%, which suggests a need to further evaluate the spectroscopy setup.

  7. 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

  8. Development of gamma-ray absorption and scattering simulation platform based on MCNP

    International Nuclear Information System (INIS)

    Lai Wanchang; Chen Henggui; Zhang Zhen; Chen Xiaoqiang

    2010-01-01

    It describes a γ-ray absorption and scattering simulation platform centering on MCNP, and developed corresponding accessories on the basis of the MCNP. Simulation of this simulation platform can be 93 kinds of single-quality materials and 2-3 kinds of multi-element mixture absorption experiment, simulating the absorption thickness of 0-100cm, and the thickness increment in 0.001cm. The media of Scattering Simulation is from the Li to the Am, the angle between the simulation measuring degree and incident ray direction is from-90 to 90, the angle in increments in 1 degree. (authors)

  9. MCNP and other nuclear codes output graphical representation using python scripts; Representacion grafica de outputs de MCNP y codigos nucleares mediante el uso de scripts en python

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.

    2016-08-01

    Due to the lack of graphical representation capability of same nuclear codes like MCNP of GOTHIC, widely used in the industry, the following article describes the development of an interface to use a graphical representation open source (Paraview) with the outputs generated by the nuclear codes. Moreover, this article aims at describing the advantage of this type of visualization programs for the modeling and decision making in the calculation. (Author)

  10. Real-time monitoring of genetically modified Chlamydomonas reinhardtii during the Foton M3 space mission and ground irradiation experiment

    Science.gov (United States)

    Lambreva, Maya; Rea, Giuseppina; Antonacci, Amina; Serafini, Agnese; Damasso, Mario; Margonelli, Andrea; Johanningmeier, Udo; Bertalan, Ivo; Pezzotti, Gianni; Giardi, Maria Teresa

    Long-term space exploration, colonization or habitation requires biological life support systems capable to cope with the deleterious space environment. The use of oxygenic photosynthetic microrganisms is an intriguing possibility mainly for food, O2 and nutraceutical compounds production. The critical points of utilizing plantsor algae-based life support systems are the microgravity and the ionizing radiation, which can influence the performance of these organisms. The aim of the present study was to assess the effects of space environment on the photosynthetic activity of various microrganisms and to select space stress-tolerant strains. Site-directed and random mutants of the unicellular green alga Chlamydomonas reinhardtii of Photosystem II D1 protein were used as a model system to test and select the amino acid substitutions capable to account for space stress tolerance. We focussed our studies also on the accumulation of the Photosystem II photoprotective carotenoids (the xantophylls violaxanthin, anteraxanthin and zeaxanthin), powerful antioxidants that epidemiological studies demonstrated to be human vision protectors. Metabolite profiling by quantitative HPLC methods revealed the organisms and the stress conditions capable to accumulate the highest pigment levels. In order to develop a project for a rationale metabolic engineering of algal secondary metabolites overproduction, we are performing expression analyses on the carotenoid biosynthetic pathway under physiological and mimicked space conditions. To identify the consequences of the space environment on the photosynthetic apparatus the changes in the Photosystem II efficiency were monitored in real time during the ESA-Russian Foton-M3 mission in September 2007. For the space flight a high-tech, multicell fluorescence biosensor, Photo-II, was designed and built by the Centre for Advanced Research in Space Optics in collaboration with Kayser-Italy, Biosensor and DAS. Photo-II is an automatic device

  11. Evaluation of the AP1000 delayed neutron parameters using MCNP6

    Science.gov (United States)

    Sembiring, T. M.; Susilo, J.; Pinem, S.

    2018-02-01

    The MCNP6 code contains numerous features, one of those is to determine the delayed neutron parameters. The accuracy of calculated delayed neutron parameters affect the accuracy of transient or dynamic condition. The objective of this paper is to determine the delayed neutron parameters of the advance PWR reactor, AP1000, using MCNP6 code with the recent ENDF/B evaluated nuclear data file ENDF/B-VII.1. The MCNP6 calculation results shows that the maximum difference occurred in the βi and λi parameters are 38.30% and 45.63%, respectively. The superiority of MCNP6 can be seen in the change of prompt neutron life time (ℓ) parameters that cannot be obtained by the deterministic code, so it can be used in the sensitivity analysis of the delayed neutron parameters. Based on this research work, the accident analysis of the AP1000 reactor use the effective delayed neutron fraction (β eff) of0.0051 and the prompt neutron life time (ℓ) of 19.5 μs for the first cycle.

  12. Program for the Generation of MCNP Inputs from State Files of CAREM

    International Nuclear Information System (INIS)

    Leszczynski, Francisco; Lopasso, Edmundo; Villarino, E

    2000-01-01

    The objective of this work is the development and tests of detailed input data for the Monte Carlo program MCNP, to be able of model the core of CAREM reactor, with the detail included on the updated models, for having available a calculation system that allow the production of confident results to be compared with results obtained with the system used today for designing the CAREM reactor core (CONDOR-CITVAP).The model includes the possibility of temperature and coolant density, and temperature and numeric densities of fuel.The detail consists of 21 different fuel elements (symmetry 3) and 14 axial zones.Results of comparisons of reactivity and power pick factors are presented, between MCNP and CONDOR-CITVAP.On average, these results show an acceptable agreement for all the compared parameters.It is described, also, the interface CONDOR-CITVAP-MCNP program, that has been developed for generating inputs of materials for MCNP, from outputs of CONDOR and CITVAP, for different reactor states

  13. 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.)

  14. 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

  15. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

    International Nuclear Information System (INIS)

    Kurosawa, M.

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)

  16. Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement

    Science.gov (United States)

    Tiyapun, K.; Chimtin, M.; Munsorn, S.; Somchit, S.

    2015-05-01

    The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

  17. Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia

    International Nuclear Information System (INIS)

    Yavar, A.R.; Khalafi, H.; Kasesaz, Y.; Sarmani, S.; Yahaya, R.; Wood, A.K.; Khoo, K.S.

    2012-01-01

    A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k 0 -INAA and absolute method. The average values of φ th ,φ epi , and φ fast by MCNP code were (2.19±0.03)×10 12 cm −2 s −1 , (1.26±0.02)×10 11 cm −2 s −1 and (3.33±0.02)×10 10 cm −2 s −1 , respectively. These average values were consistent with the experimental results obtained by k 0 -INAA. The findings show a good agreement between MCNP code results and experimental results. - Highlights: ► We use 3-D neutronic model to enhance the utilization of the economical use of reactor.► The MCNP code is modified to analyze the neutronic parameters. ► The neutron flux distributions are homogeneous and consistent with experimental results. ► A complete simulation of the calculated neutron flux parameters in 40 RR irradiation channels is achieved.

  18. Three-dimensional transport theory: Evaluation of analytical expressions of Williams and verification of MCNP

    International Nuclear Information System (INIS)

    Jeong, Jeho; White, Nathan E.; Loyalka, Sudarshan K.

    2015-01-01

    Highlights: • An evaluation of 3-D neutron transport analytical expressions of Williams. • Techniques for oscillating, singular and infinite integrals are applied. • Disagreements with reported values are rare even at 5 significant figures. • MCNP is verified against analytical results for several benchmarks. • MCNP results generally agree with analytical results, except near singularities. - Abstract: “Three-dimensional transport theory: an analytical solution of an internal beam searchlight problem, I”, Annals of Nuclear Energy, 36(8), 1256–1261 (2009) by Williams extends the range of analytical solutions, and the associated development of techniques, numerical results and analysis near singularities. The final integrals are not easy to evaluate as the integrands are highly oscillatory, singular and also on infinite range. We report here some further numerical evaluations of expressions of Williams, and also compare these with those of Williams and Ganapol and Kornreich. The numerical results compare very well. The disagreements are very rare, and even then in the fifth decimal place. We are also able to explore the nature of the results near singularities in conformity with the results of Williams. We also verify MCNP-5, the widely used Monte Carlo code against these analytical results. We have found that MCNP is easily able to provide results within 0.1% deviation from the “exact” results for most cases, and within 1% for almost all cases. It is challenged near the singularities, however, where the deviations are larger.

  19. Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, Albert Comstock [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-14

    We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.

  20. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    Science.gov (United States)

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

  1. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    FINFROCK SH

    2009-12-10

    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_sign} 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{sub 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{sub 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{sub 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{sub 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{sub eff} limit for calculations of the intermediate enriched uranium type systems.

  2. 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.

  3. 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)

  4. 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)

  5. Development of the Gecko (Pachydactylus turneri) Animal Model during Foton M-2 to Study Comparative Effects of Microgravity in Terrestrial and Aquatic Organisms

    Science.gov (United States)

    Almeida, E. A.; Roden, C.; Phillips, J. A.; Globus, R. K.; Searby, N.; Vercoutere, W.; Morey-Holton, E.; Gulimova, V.; Saveliev, S.; Tairbekov, M.; hide

    2006-01-01

    Terrestrial organisms exposed to microgravity during spaceflight experience degeneration in bone, muscle, and possibly other tissues that require gravity-mediated mechanical stimulation for normal regenerative growth. In the Gecko experiment aboard Foton M-2, we flew for the first time, five terrestrial Pachydactylus turneri specimens to develop a model of microgravity effects comparable to the newt Pleurodeles waltl, a well-established model organism for spaceflight. These lower vertebrate species have similar body plans and size, are poikilothermic, have tissue regenerative ability, and are adapted to moderate periods of fasting. Furthermore the gecko (Pachydactylus) can also survive prolonged periods without water. In pre-flight control experiments and after a 16-day Foton M-2 spaceflight without food or water, the geckos were recovered and showed no apparent negative health effects. However, detailed analysis of bone mass and architecture by micro Computed Tomography { pCT), showed that both synchronous control and spaceflight animals lost significant amounts of cancellous bone in the distal femur and humerus relative to basal controls. In addition, cell cycle analysis of 30h post-flight liver tissue reveals a shift of DNA content from G2 and S to G1, both in spaceflight and synchronous controls. Together, these results suggest that housing conditions alone induce rapid catabolism of cancellous bone and reduced normal tissue regeneration. Further use of the gecko Puchydactylus turneri as a spaceflight model requires modification of housing conditions, possibly by including water and food, or changing other factors such as eliminating housing stresses to obtain stable bone structure and tissue regeneration during spaceflight experiments.

  6. An MCNP model of glove boxes in a plutonium processing facility

    International Nuclear Information System (INIS)

    Dooley, D.E.; Kornreich, D.E.

    1998-01-01

    Nuclear material processing usually occurs simultaneously in several glove boxes whose primary purpose is to contain radioactive materials and prevent inhalation or ingestion of radioactive materials by workers. A room in the plutonium facility at Los Alamos National Laboratory has been slated for installation of a glove box for storing plutonium metal in various shapes during processing. This storage glove box will be located in a room containing other glove boxes used daily by workers processing plutonium parts. An MCNP model of the room and glove boxes has been constructed to estimate the neutron flux at various locations in the room for two different locations of the storage glove box and to determine the effect of placing polyethylene shielding around the storage glove box. A neutron dose survey of the room with sources dispersed as during normal production operations was used as a benchmark to compare the neutron dose equivalent rates calculated by the MCNP model

  7. Validation of MCNP: SPERT-D and BORAX-V fuel

    International Nuclear Information System (INIS)

    Crawford, C.; Palmer, B.

    1992-11-01

    This report discusses critical experiments involving SPERT-D 1,2 fuel elements and BORAX-V 3-8 fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assemblies or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods

  8. 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.)

  9. 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); Univ. of Idaho, Moscow, ID (United States)

    2015-08-24

    Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. 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. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to sup>4He 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.

  10. Optimization of in-core fuel management strategy of Tehran Research Reactor (TRR) using MCNP-4C

    Energy Technology Data Exchange (ETDEWEB)

    Keyvani, M., E-mail: mkeyvani@aeoi.org.i [Atomic Energy Organization of Iran, Nuclear Science and Technology Research Institute (NSTRI), Reactor and Accelerator Research and Development School, End of Karegar Ave., Tehran 14155-1339 (Iran, Islamic Republic of); Arkani, M., E-mail: markani@aeoi.org.i [Atomic Energy Organization of Iran, Nuclear Science and Technology Research Institute (NSTRI), Reactor and Accelerator Research and Development School, End of Karegar Ave., Tehran 14155-1339 (Iran, Islamic Republic of); Rokh, A. Hossni, E-mail: ahossnirokh@aeoi.org.i [Atomic Energy Organization of Iran, Nuclear Science and Technology Research Institute (NSTRI), Reactor and Accelerator Research and Development School, End of Karegar Ave., Tehran 14155-1339 (Iran, Islamic Republic of)

    2010-12-15

    In order to optimize fuel utilization in TRR, the method of fuel management is modified using MCNP-4C code system. An important parameter of fuel management is the uniformity of neutron flux distribution in the core region, which is obtained efficiently in the present strategy. This strategy is based on calculation of position factors and power densities utilizing MCNP simulations. This study shows that the core life time and average extracted burn up of spent fuel elements of TRR are improved significantly.

  11. The study on neutron and photon distribution of AP1000 reactor by MCNP code

    International Nuclear Information System (INIS)

    Chen Defeng; Shen Mingqi

    2014-01-01

    The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)

  12. 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)

  13. MCNP modeling of NORM dosimetry in the oil and gas industry

    International Nuclear Information System (INIS)

    Siqiu Wang

    2016-01-01

    Naturally-occurring radioactive materials wastes in the oil and gas industry create a radioactive environment for the workers in the field. MCNP simulation conducted in this work provides a useful tool in terms of radiation safety design of the oil field, as well as validation and an important addition to in situ measurements. Furthermore, phantoms are employed to observe the dose distribution throughout the human body, demonstrating radiation effects on each individual organ. (author)

  14. Development of Multi-physics (Multiphase CFD + MCNP) simulation for generic solution vessel power calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Jun [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-17

    The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operating scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi

  15. 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)

  16. 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)

  17. Comparison of thermal scattering processing options for S(α,β) cards in MCNP

    International Nuclear Information System (INIS)

    Čerba, Štefan; Damian, Jose Ignacio Marquez; Lüley, Jakub; Vrban, Branislav; Farkas, Gabriel; Nečas, Vladimír; Haščík, Jan

    2013-01-01

    Highlights: ► Determination of MCNP calculation bias for WWER-440. ► Specific scattering law S(α,β). ► Benchmark cases investigated. ► Three methods to process material cards for hydrogen bound in light water. - Abstract: The MCNP distributions include sets of pre-calculated thermal scattering libraries but these libraries are available for several temperature steps only. In order to achieve reliable results it is suitable to process the cross section libraries for the desired temperature. In general, there are three methods to process these thermal scattering libraries for the desired temperatures. This paper deals with the comparison of these three methods on the basis of several benchmarks and on the basis of a thermal transient experiment of a WWER-440 reactor. The choice is up to the MCNP user but unfortunately very few studies concerning the comparison have been published so far. Therefore conclusions and results presented in this paper may help the user to choose the most appropriate method for his calculation

  18. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  19. 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.

  20. Current status of ACE format libraries for MCNP at nuclear date center of KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Gil, Choong Sup; Lee, Young Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Validation calculations with recent nuclear data evaluations ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and χ2 values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the keff values. It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

  1. MCNP6 simulation of reactions of interest to FRIB, medical, and space applications

    International Nuclear Information System (INIS)

    Mashnik, Stepan G.

    2015-01-01

    The latest production-version of the Los Alamos Monte Carlo N-Particle transport code MCNP6 has been used to simulate a variety of particle-nucleus and nucleus-nucleus reactions of academic and applied interest to research subjects at the Facility for Rare Isotope Beams (FRIB), medical isotope production, space-radiation shielding, cosmic-ray propagation, and accelerator applications, including several reactions induced by radioactive isotopes, analyzing production of both stable and radioactive residual nuclei. Here, we discuss examples of validation and verification of MCNP6 by comparing with recent neutron spectra measured at the Heavy Ion Medical Accelerator in Chiba, Japan; spectra of light fragments from several reactions measured recently at GANIL, France; INFN Laboratori Nazionali del Sud, Catania, Italy; COSY of the Jülich Research Center, Germany; and cross sections of products from several reactions measured lately at GSI, Darmstadt, Germany; ITEP, Moscow, Russia; and, LANSCE, LANL, Los Alamos, U.S.A. As a rule, MCNP6 provides quite good predictions for most of the reactions we analyzed so far, allowing us to conclude that it can be used as a reliable and useful simulation tool for various applications for FRIB, medical, and space applications involving stable and radioactive isotopes. (author)

  2. V&V of MCNP 6.1.1 Beta Against Intermediate and High-Energy Experimental Data

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan G [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-08

    This report presents a set of validation and verification (V&V) MCNP 6.1.1 beta results calculated in parallel, with MPI, obtained using its event generators at intermediate and high-energies compared against various experimental data. It also contains several examples of results using the models at energies below 150 MeV, down to 10 MeV, where data libraries are normally used. This report can be considered as the forth part of a set of MCNP6 Testing Primers, after its first, LA-UR-11-05129, and second, LA-UR-11-05627, and third, LA-UR-26944, publications, but is devoted to V&V with the latest, 1.1 beta version of MCNP6. The MCNP6 test-problems discussed here are presented in the /VALIDATION_CEM/and/VALIDATION_LAQGSM/subdirectories in the MCNP6/Testing/directory. README files that contain short descriptions of every input file, the experiment, the quantity of interest that the experiment measures and its description in the MCNP6 output files, and the publication reference of that experiment are presented for every test problem. Templates for plotting the corresponding results with xmgrace as well as pdf files with figures representing the final results of our V&V efforts are presented. Several technical “bugs” in MCNP 6.1.1 beta were discovered during our current V&V of MCNP6 while running it in parallel with MPI using its event generators. These “bugs” are to be fixed in the following version of MCNP6. Our results show that MCNP 6.1.1 beta using its CEM03.03, LAQGSM03.03, Bertini, and INCL+ABLA, event generators describes, as a rule, reasonably well different intermediate- and high-energy measured data. This primer isn’t meant to be read from cover to cover. Readers may skip some sections and go directly to any test problem in which they are interested.

  3. DISEÑO DE UN MICROSISTEMA PROGRAMABLE PARA EFECTOS DE AUDIO DIGITAL USANDO FPGAS

    Directory of Open Access Journals (Sweden)

    John Michael Espinosa Durán

    Full Text Available Este artículo describe el diseño de un microsistema programable para el procesamiento de efectos de audio digital implementado en un FPGA. El microsistema es diseñado usando un procesador de propósito específico y reconfigurable, un banco de RAMs y una interfaz gráfica de usuario basada en una pantalla táctil LCD. El procesador es diseñado usando 15 efectos de audio basados en retardos y procesamiento en el dominio dinámico y de la frecuencia. Los efectos son diseñados usando Megafunciones y el compilador FIR de Quartus II, son simulados en Simulink5 usando DSP Builder6, y son configurados utilizando una interfaz gráfica de usuario. El microsistema programable es implementado en el sistema de desarrollo DE2-70, y su funcionamiento es verificado usando un reproductor MP3 y un parlante. Adicionalmente, el microsistema permite la generación de efectos con alta fidelidad usando una tasa de muestreo máxima de 195.62 MSPS, y puede ser embebido en un SoC.

  4. SISTEMA DE MODELAGEM DE FIGURAS TRIDIMENSIONAIS USANDO RECONHECIMENTO DE VOZ

    Directory of Open Access Journals (Sweden)

    Kauan Cristiano De Souza

    2017-01-01

    Full Text Available Atualmente o mercado de Sistemas de Desenho auxiliado por Computador (CAD, sofre extrema carência de processos automatizados de reconhecimento de fala. O surgimento de tecnologias de automação facilita no desenvolvimento de aplicações capazes de auxiliar nestes processos. Baseando-se nestas afirmações construiu-se um modelo utilizando tecnologias capazes de transformar fala em sinais digitais, sendo possível a interpretação por sofisticados computadores. Observando ausência de tais ferramentas, o presente trabalho busca uma proposta capaz de acrescentar mecanismos os quais possibilitem a criação de modelos tridimensionais usando tecnologias como servidores HTTP, dispositivos moveis, linguagens de programação Java e PHP, Computação Gráfica e linguagem de interpretação por blocos. Os resultados obtidos na confecção do presente trabalho corroboram com as necessidades atuais representadas na sociedade propondo sustentabilidade, acessibilidade, portabilidade e facilidade de comunicação.

  5. Spore-Forming Thermophilic Bacterium within Artificial Meteorite Survives Entry into the Earth's Atmosphere on FOTON-M4 Satellite Landing Module.

    Science.gov (United States)

    Slobodkin, Alexander; Gavrilov, Sergey; Ionov, Victor; Iliyin, Vyacheslav

    2015-01-01

    One of the key conditions of the lithopanspermia hypothesis is that microorganisms situated within meteorites could survive hypervelocity entry from space through the Earth's atmosphere. So far, all experimental proof of this possibility has been based on tests with sounding rockets which do not reach the transit velocities of natural meteorites. We explored the survival of the spore-forming thermophilic anaerobic bacterium, Thermoanaerobacter siderophilus, placed within 1.4-cm thick basalt discs fixed on the exterior of a space capsule (the METEORITE experiment on the FOTON-M4 satellite). After 45 days of orbital flight, the landing module of the space vehicle returned to Earth. The temperature during the atmospheric transit was high enough to melt the surface of basalt. T. siderophilus survived the entry; viable cells were recovered from 4 of 24 wells loaded with this microorganism. The identity of the strain was confirmed by 16S rRNA gene sequence and physiological tests. This is the first report on the survival of a lifeform within an artificial meteorite after entry from space orbit through Earth's atmosphere at a velocity that closely approached the velocities of natural meteorites. The characteristics of the artificial meteorite and the living object applied in this study can serve as positive controls in further experiments on testing of different organisms and conditions of interplanetary transport.

  6. Spore-Forming Thermophilic Bacterium within Artificial Meteorite Survives Entry into the Earth's Atmosphere on FOTON-M4 Satellite Landing Module.

    Directory of Open Access Journals (Sweden)

    Alexander Slobodkin

    Full Text Available One of the key conditions of the lithopanspermia hypothesis is that microorganisms situated within meteorites could survive hypervelocity entry from space through the Earth's atmosphere. So far, all experimental proof of this possibility has been based on tests with sounding rockets which do not reach the transit velocities of natural meteorites. We explored the survival of the spore-forming thermophilic anaerobic bacterium, Thermoanaerobacter siderophilus, placed within 1.4-cm thick basalt discs fixed on the exterior of a space capsule (the METEORITE experiment on the FOTON-M4 satellite. After 45 days of orbital flight, the landing module of the space vehicle returned to Earth. The temperature during the atmospheric transit was high enough to melt the surface of basalt. T. siderophilus survived the entry; viable cells were recovered from 4 of 24 wells loaded with this microorganism. The identity of the strain was confirmed by 16S rRNA gene sequence and physiological tests. This is the first report on the survival of a lifeform within an artificial meteorite after entry from space orbit through Earth's atmosphere at a velocity that closely approached the velocities of natural meteorites. The characteristics of the artificial meteorite and the living object applied in this study can serve as positive controls in further experiments on testing of different organisms and conditions of interplanetary transport.

  7. Heating of the quiet solar corona from measurements of the FET/TESIS instrument on-board the KORONAS-FOTON satellite

    Science.gov (United States)

    Rybák, J.; Gömöry, P.; Benz, A.; Bogachev, P.; Brajša, R.

    2010-12-01

    The paper presents the first results of the observations of time evolution of the quiet solar corona brightenings obtained due to very rapid photography of the corona with full-disk EUV telescopes of the FET/TESIS instrument onboard the KORONA FOTON satellite. The measurements were performed simultaneously in the emission of the Fe IX / X 17.1 and Fe VIII 13.1 spectral lines with 10 second temporal cadence and spatial scale of 1.7 arc seconds within one hour. This test observation, carried out on 15 July 2009, was analyzed in order to determine whether this type of observation can be used to identify individual microevents in the solar corona heating that are above the tresholds of spatial and temporal resolutions of the observations of non-active regions in the solar atmosphere. For this purpose, a simple method was used involving cross-correlation of the plasma emission time evolution at different temperatures, each time from observations of identical elements. The results obtained are confronted with the expected observable manifestations of the corona heating via nanoflares. TESIS is a set of instruments for the Sun photography developed in the Lebedev Physics Institute of the Russian Academy of Sciences that was launched into orbit in January 2009.

  8. A Comparative Depletion Analysis using MCNP6 and REBUS-3 for Advanced SFR Burner Core

    Energy Technology Data Exchange (ETDEWEB)

    You, Wu Seung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    In this paper, we evaluated the accuracy of fast reactor design codes by comparing with MCNP6-based Monte Carlo simulation and REBUS-3-based the nodal transport theory for an initial cycle of an advanced uranium-free fueled SFR burner core having large heterogeneities. It was shown that the nodal diffusion calculation in REBUS-3 gave a large difference in initial k-effective value by 2132pcm when compared with MCNP6 depletion calculation using heterogeneous model.The code system validation for fast reactor design is one of the important research topics. In our previous studies, depletion analysis and physics parameter evaluation of fast reactor core were done with REBUS-3 code and DIF3D code, respectively. In particular, the depletion analysis was done with lumped fission products. However, it is need to verify the accuracy of these calculation methodologies by using Monte Carlo neutron transport calculation coupled with explicit treatment of fission products. In this study, the accuracy of fast reactor design codes and procedures were evaluated using MCNP6 code and VARIANT nodal transport calculation for an initial cycle of an advanced sodium-cooled burner core loaded with uranium-free fuels. It was considered that the REBUS-3 nodal diffusion option can not be used to accurately estimate the depletion calculations and VARIANT nodal transport or VARIANT SP3 options are required for this purpose for this kind of heterogeneous burner core loaded with uranium-free fuel. The control rod worths with nodal diffusion and transport options were estimated with discrepancies less than 12% while these methods for sodium void worth at BOC gave large discrepancies of 12.2% and 16.9%, respectively. It is considered that these large discrepancies in sodium void worth are resulted from the inaccurate consideration of spectrum change in multi-group cross section.

  9. 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

  10. Aulas-laboratorios de bajo costo, usando TIC

    Directory of Open Access Journals (Sweden)

    Silvia E. Calderón

    2015-01-01

    Full Text Available En este trabajo se presentan los resultados de una propuesta educativa orientada a promover el desarrollo de un pensamiento critico y un mayor interes por las ciencias experimentales. Con este fin desarrollamos propuestas de proyectos educativos susceptibles de ser destinadas a las aulas y laboratorios de las escuelas secundarias y primeros anos de la universidad, que resaltan los aspectos metodologicos de las ciencias. Aqui, realizamos una compilacion de varios proyectos, que ilustran formas de incorporar las tecnologias de la informacion y la comunicacion (TIC en diversos experimentos de ciencias, muchos de ellos publicados individualmente anteriormente, y que en conjunto se pueden utilizar para implementar un aula-laboratorio de bajo costo. Con TIC hacemos referencia a la convergencia de computadoras, sistemas audiovisuales, Internet, telefonia, y diversos equipos que se integran con algunos de ellos. Los proyectos intentan integrar areas como fisica, matematica, quimica, informatica, arte, etc. y apuntan a que los estudiantes puedan responder a las preguntas: .Como sabemos esto?, .Por que creemos en aquello? Preguntas que ilustran la naturaleza del pensamiento cientifico. Nuestra contribucion mas significativa es haber desarrollado ≪aulas-laboratorios≫ de muy bajo costo, usando TIC. Con el advenimiento de programas como ≪Una Laptop por Nino≫ que se estan implementando en varios paises de Latinoamerica, resulta oportuno utilizar este recurso como base para generar laboratorios de bajo costo, que creemos pueden ser una herramienta util para mejorar el aprendizaje de las ciencias, incentivar vocaciones y contribuir a desarrollar un pensamiento critico, a la par de desarrollar habilidades con el uso de las TIC que pueden ser de utilidad en diversos ambitos academicos y laborales.

  11. 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.

  12. Dose calculation for 40K ingestion in samples of beans using spectrometry and MCNP

    International Nuclear Information System (INIS)

    Garcez, R.W.D.; Lopes, J.M.; Silva, A.X.; Domingues, A.M.; Lima, M.A.F.

    2014-01-01

    A method based on gamma spectroscopy and on the use of voxel phantoms to calculate dose due to ingestion of 40 K contained in bean samples are presented in this work. To quantify the activity of radionuclide, HPGe detector was used and the data entered in the input file of MCNP code. The highest value of equivalent dose was 7.83 μSv.y -1 in the stomach for white beans, whose activity 452.4 Bq.Kg -1 was the highest of the five analyzed. The tool proved to be appropriate when you want to calculate the dose in organs due to ingestion of food. (author)

  13. A comparison study for mass attenuation coefficients of some amino acids using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Vahabi, Seyed Milad; Bahreynipour, Mostean; Shamsaie-Zafarghandi, Mojtaba [Amirkabir Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this study, a novel model of MCNP4C code reported recently was used to determine the photon mass attenuation coefficients of some amino acids at energies, 123, 360, 511, 662, 1170, 1280 and 1330 keV. The simulation results were compared with the XCOM data. It was indicated that the results were highly close to the calculated XCOM values. Obtained results were used to calculate the molar extinction coefficient. All the results showed the convenience and usefulness of the model in calculation of mass attenuation coefficients of amino acids.

  14. 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)

  15. New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

    International Nuclear Information System (INIS)

    Brown, Forrest B.

    2016-01-01

    Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple a ce.pl and simple a ce m g.pl.

  16. Isodose distributions and dose uniformity in the Portuguese gamma irradiation facility calculated using the MCNP code

    CERN Document Server

    Oliveira, C

    2001-01-01

    A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. The absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to achieve a given value of density were used. The magnitude of various contributions to the total photon spectra, including source-dependent factors, irradiator structures, sample material and other origins were also calculated.

  17. Enhancement and validation of the NPP Mühleberg MCNP activation simulations for Swiss decommissioning planning

    International Nuclear Information System (INIS)

    Bykov, V.

    2014-08-01

    The Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) regularly performs analysis of cost estimates associated with the NPP decommissioning. For this purpose, NAGRA has over the past ten years developed a NPP activation analysis methodology based on MCNP models of Swiss NPPs. The validation of these models is accomplished using measurements from oil activation campaigns, in which foil samples are activated at key locations inside the NPP for the duration of one cycle. The measurement campaigns have already been carried out at the Gösgen PWR (KKG) and the Mühleberg BWR (KKM). The first validation has already been successfully conducted for the KKG MCNP model. This thesis describes the efforts to validate the KKM MCNP model. This process included modifications, such as modeling of steam separators individually and improving the definition of jet pumps. Furthermore, the core definition was completely redefined, going from a 6-cell cylindrical model to a 940-cell model, shaped like the actual KKM core, which more accurately represented the void distribution. In order to benchmark the new model, the locations of samples during the two KKM foil activation campaigns were implemented into the model using the GSAM code. The interface between the MCNP model and GSAM was improved by creating a new energy group structure, optimized specifically for the activation of the three foil materials. Their activation was stimulated the state of the art hybrid VR code ADVANTG. The calculated results were then compared against the measured values for each foil material separately. The numerous improvements introduced in the 2014 model led to good agreement in many areas. The agreement is within the factor of two on the inner side of the bioshield, at the core height and above, and factor of three above the bioshield. Furthermore, distinct suggestion for improving the agreement in other areas was presented. This includes modeling of pipes extending from the RPV

  18. Improvements in the simulation of the efficiency of a HPGe detector with Monte Carlo code MCNP5; Mejoras en la simulacion de la eficiencia de un detector HPGe con el codigo Monte Carlo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

    in this paper we propose to perform a simulation model using the MCNP5 code and a registration form meshing to improve the simulation efficiency of the detector in the range of energies ranging from 50 to 2000 keV. This meshing is built by FMESH MCNP5 registration code that allows a mesh with cells of few microns. The photon and electron flow is calculated in the different cells of the mesh which is superimposed on detector geometry. It analyzes the variation of efficiency (related to the variation of energy deposited in the active volume). (Author)

  19. 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)

  20. 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)

  1. Neutron flux measurement in the thermal column of the Malaysian TRIGA mark II reactor with MCNP verification

    International Nuclear Information System (INIS)

    Abdel Munem, E.; Shukri, A.; Tajuddin, A.A.

    2006-01-01

    A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)

  2. New Neutron, Proton, and S(α,β) MCNP Data Libraries Based on ENDF/B-VII

    International Nuclear Information System (INIS)

    Little, Robert C.; Trellue, Holly R.; MacFarlane, Robert E.; Kahler, A.C.; Lee, Mary Beth; White, Morgan C.

    2008-01-01

    The general-purpose Evaluated Nuclear Data File ENDF/B-VII.0 was released in December 2006. A number of sub-libraries were included in ENDF/B-VII.0 such that data were provided for incident neutrons, photons, and charged particles. This paper describes the creation of MCNP data libraries at Los Alamos National Laboratory based on three ENDF/B-VII.0 sub-libraries: neutrons, protons, and thermal scattering. An ACE-formatted continuous-energy neutron data library called ENDF70 for MCNP has been produced. This library provides data for 390 materials at five temperatures: 293.6, 600, 900, 1200, and 2500 K. The library was processed primarily with Version 248 of NJOY99. Extensive checking and quality-assurance tests were applied to the data. Improvements to the processing code were made and certain evaluations were modified as a result of these tests. ENDF/B-VII.0 included proton evaluations for 48 target materials. Forty-seven proton evaluations (all except for 13 C) were processed at room temperature and combined into the MCNP library ENDF70PROT. Neutron thermal S(α,β) scattering data exist for twenty different materials in ENDF/B-VII.0. All twenty of these evaluations were processed at all applicable temperatures (these vary for each evaluation), and combined into the MCNP library ENDF70SAB. All of these ENDF/B-VII.0 based MCNP libraries (ENDF70, ENDF70PROT, and ENDF70SAB) are available as part of the MCNP5 1.50 release. (authors)

  3. A group of neutronics calculations in the MNSR using the MCNP-4C code

    International Nuclear Information System (INIS)

    Khattab, K.; Sulieman, I.

    2009-11-01

    The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)

  4. MCNP simulations of a glass display used in a mobile phone as an accident dosimeter

    International Nuclear Information System (INIS)

    Discher, Michael; Hiller, Mauritius; Woda, Clemens

    2015-01-01

    It has been demonstrated that glass display of mobile phones can be used as a device for accident dosimetry. Published studies concentrated on the experimental investigation of parts of the glass display. In the work presented here, the experimental results are compared with results of radiation transport calculations using the Monte Carlo code MCNP5. An experimental setup of an irradiation of an extracted glass display is simulated. The simulation is then extended to a simulation of a modern day mobile phone consisting of all major parts. Simulations are performed for various irradiation conditions and different geometric and material properties. The results of the simulation show a good agreement with the experiments for an extracted glass sample as well as for an actual modern mobile phone. The glass display is exposed to radiation in various angular and energy distributions. Simulated results were compared to experimentally determined results. The effects of the irradiation condition on the photon energy dependence were investigated and variations in the material constants of the display glass composition were discussed. This work affirms the usability of a mobile phone as a versatile and flexible accident radiation detector. - Highlights: • Simulations of a modern day mobile phone using MCNP are carried out. • Results of the simulation show a good agreement with the experiments. • Photon energy dependence and angular response for display glass are verified

  5. An experimental test on large animals of MCNP application for whole body counting

    Energy Technology Data Exchange (ETDEWEB)

    Borisov, N.; Yatsenko, V.; Kochetkov, O.; Gusev, I.; Vlasov, P.; Kalistratova, V.; Nisimov, P.; Levochkin, F.; Borovkov, M.; Stolyarov, V. [State Research Center, Institute of Biophysics, Moscow (Russian Federation); Tsedish, S.; Tyurin, I. [Clinical Hospital No. 6 of Federal Medico-Biological Agency, Moscow (Russian Federation); Franck, D.; Carlan, L. de [Institut de Radioprotection et de Surete Nucleaire, 92 - Fontenay-aux Roses (France)

    2005-07-01

    Measurements of actinide body burden using whole body counting spectrometry is hampered due to intensive absorption of {gamma}-rays inside the patient's body, which depends on the anatomy of a patient. To establish the correspondence between pulse-height-spectra intensity and radionuclide activity, Monte Carlo calculations are widely used. For such calculations, the radiation transport geometry is usually described in terms of small rectangular boxes (voxels) retrieved from computed tomography or magnetic resonance images. The software for Monte Carlo-assisted calibration of whole body counting, which performs automatic creation of individual MCNP voxel phantoms, was checked in a quasi-in vivo experiment on large animals. During the experiment, pigs of 35-40 kg body mass were used as phantoms for measurement of actinides body burden. {sup 241}Am was administered (via injection of a radioactive solution or via implantation of plastic capsules containing the radioactive material) into the lungs of pigs. The pigs were measured using the pure germanium low-energy {gamma}-spectrometers. The images of animals were obtained using the computed tomography machine. On the base of these tomograms, MCNP4c2 calculations were done to obtain the pulse-height-spectra of the whole body counters. The experimental results were reproduced in calculations with error of less than 30% for {sup 241}Am administered via injection and less than 10% for {sup 241}Am administered inside the capsules. (authors)

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

    Directory of Open Access Journals (Sweden)

    Blanchet David

    2017-01-01

    Full Text Available 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.

  7. Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code

    Directory of Open Access Journals (Sweden)

    C. A. M. Silva

    2014-01-01

    Full Text Available In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport code. The sensitivity analyses included small differences of the core and the rods dimensions and different levels of model detailing. Four models were simulated and neutronic parameters such as effective multiplication factor (keff, reactivity (ρ, and thermal and total neutron flux in central thimble in some different conditions of the reactor operation were analysed. The simulated models presented good agreement between them, as well as in comparison with available experimental data. In this way, the sensitivity analyses demonstrated that simulations of the TRIGA IPR-R1 reactor can be performed using any one of the four investigated MCNP models to obtain the referenced neutronic parameters.

  8. An experimental test on large animals of MCNP application for whole body counting

    International Nuclear Information System (INIS)

    Borisov, N.; Yatsenko, V.; Kochetkov, O.; Gusev, I.; Vlasov, P.; Kalistratova, V.; Nisimov, P.; Levochkin, F.; Borovkov, M.; Stolyarov, V.; Tsedish, S.; Tyurin, I.; Franck, D.; Carlan, L. de

    2005-01-01

    Measurements of actinide body burden using whole body counting spectrometry is hampered due to intensive absorption of γ-rays inside the patient's body, which depends on the anatomy of a patient. To establish the correspondence between pulse-height-spectra intensity and radionuclide activity, Monte Carlo calculations are widely used. For such calculations, the radiation transport geometry is usually described in terms of small rectangular boxes (voxels) retrieved from computed tomography or magnetic resonance images. The software for Monte Carlo-assisted calibration of whole body counting, which performs automatic creation of individual MCNP voxel phantoms, was checked in a quasi-in vivo experiment on large animals. During the experiment, pigs of 35-40 kg body mass were used as phantoms for measurement of actinides body burden. 241 Am was administered (via injection of a radioactive solution or via implantation of plastic capsules containing the radioactive material) into the lungs of pigs. The pigs were measured using the pure germanium low-energy γ-spectrometers. The images of animals were obtained using the computed tomography machine. On the base of these tomograms, MCNP4c2 calculations were done to obtain the pulse-height-spectra of the whole body counters. The experimental results were reproduced in calculations with error of less than 30% for 241 Am administered via injection and less than 10% for 241 Am administered inside the capsules. (authors)

  9. 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...

  10. Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

    Science.gov (United States)

    Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

    2018-05-01

    Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

  11. Comparison of first-principles MCNP calculations of NaI and BGO detector response functions to measurements

    International Nuclear Information System (INIS)

    Estes, G.P.; Schrandt, R.G.; Kriese, J.T.

    1992-09-01

    First-principles NaI and BGO detector response functions calculations made with the MCNP code are compared to measurements. Excellent agreement is achieved for the experiments analyzed. Such calculational methodology can be used to achieve a better understanding of the physics of detector response and to maximize the information content available from measured data

  12. Multi-canister overpack project -- verification and validation, MCNP 4A

    Energy Technology Data Exchange (ETDEWEB)

    Goldmann, L.H.

    1997-11-10

    This supporting document contains the software verification and validation (V and V) package used for Phase 2 design of the Spent Nuclear Fuel Multi-Canister Overpack. V and V packages for both ANSYS and MCNP are included. Description of Verification Run(s): This software requires that it be compiled specifically for the machine it is to be used on. Therefore to facilitate ease in the verification process the software automatically runs 25 sample problems to ensure proper installation and compilation. Once the runs are completed the software checks for verification by performing a file comparison on the new output file and the old output file. Any differences between any of the files will cause a verification error. Due to the manner in which the verification is completed a verification error does not necessarily indicate a problem. This indicates that a closer look at the output files is needed to determine the cause of the error.

  13. 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)

  14. EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

    Energy Technology Data Exchange (ETDEWEB)

    Paolo Balestra; Carlo Parisi; Andrea Alfonsi

    2016-02-01

    The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.

  15. The external dose of lack fuel cask for analyses with MCNP

    International Nuclear Information System (INIS)

    Liu Liu; Qiu Xiaoping; Liao Lingyuan

    2009-01-01

    The transport vessel of lack fuel cask is a special facilities which is for reactor lack fuel transportation. MCNP4C is used to count the external dose rate of Westinghouse MC-10 lack fuel cask, it is based on mesh definition, to get the whole external dose rate of the cask, and in connection with the result from previous researcher Georgeta Radulescu, the outcome in consistency is good, using mesh causes long-playing machine hours and comes to some error, but it can get many data about external dose rate of the lack fuel cask roundly and at any rate. So it makes sense to the definition on the external dose rate of the lack fuel cask for missionary. (authors)

  16. 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.

  17. Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

    Science.gov (United States)

    Tucker, Lucas Powelson

    This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

  18. 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)

  19. Calculation of the spectra of photons from 6 to 15 MeV emitted by a linac by deconvolution response matrix generated from the gradients depth dose calculated by MCNP simulation; Calculo de los espectros de fotones de 6 y 15 MeV emitidos por un linac mediante la deconvolucion de la matriz respuesta generada a partir de los gradientes de la dosis en profundidad calculadas mediante simulacion MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Juste, B.; Miro, R.; Verdu, G.; Diez, S.; Campayo, J. M.

    2011-07-01

    This paper proposes a technique to significantly improve the accuracy of reconstructed Bremsstrahlung spectra from depth dose data using inverse methods. While traditional approaches directly use depth dose curves, we show the advantage of using the gradient of these data for the reconstruction of spectra.

  20. Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project; Desenvolvimento de uma interface entre os codigos MCNP e ORIGEN para calculos de evolucao de combustiveis em sistemas nucleares. Projeto inicial

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel de Almeida Magalhaes

    2009-07-01

    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)

  1. Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases; Verificacion del codigo AZNHEX v.1.4 con MCNP6 para diferentes casos de referencia

    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, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: jgaliciaa87@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico)

    2017-09-15

    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)

  2. Calibration of a foot borne spectrometry system using the MCNP 4C code

    International Nuclear Information System (INIS)

    Nylen, T.; Agren, G.

    2004-01-01

    The increased interest for the cycling of radioactive Caesium in natural ecosystems has gained need for rapid and reliable methods to investigate the deposition density in natural soils. One commonly used method, soil sampling, is a good method that correctly used gives information of both the horizontal and vertical distribution of the desired nuclide. The main disadvantage is that the method is time consuming regarding sampling, preparation and measurements. An alternative method is the use of semiconductors or scintillation detectors in the field i.e. in cars, airplanes, or helicopters. Theses methods are rapid and integrate over large areas which gives a more reliable mean value provided that the operator has some basic knowledge about the depth distribution of the radio nuclides and bulk density in the soil. To be effective the systems are often connected to a GPS to give the exact coordinate for each measurement. In a situation where the area of interest is too large to cover by soil samples and measurements by airplane not will give a spatial resolution good enough, one feasible method is to use a foot borne gamma spectrometry system. The advantage of a foot borne system is that the operator can cover a quite large area within a few hours and that the method can detect small anomalies in the deposition field which may be difficult to discover with soil samples. This abstract describes the calibration of a foot borne gamma-spectrometry system carried in a back-pack and consisting of a NaI-detector, a GPS and a system for logging activity and position. The detector system and surroundings has been modeled in the Monte Carlo code MCNP 4C (Figure 1). The Monte Carlo method gives the possibility to study the influence of complex geometries that are difficult to create for a practical calibration using real activity. The results of the MCNP calibration model, has been compared to foot borne gamma-spectrometry field measurements in a Cs-137 deposition area. A

  3. Using MCNP6 to Estimate Fission Neutron Properties of a Reflected Plutonium Sphere

    Energy Technology Data Exchange (ETDEWEB)

    Clark, Alexander Rich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchinson, Jesson D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-08

    The purpose of this project was to determine the fission multiplicity distribution, p(v), for the Beryllium Reflected Plutonium (BeRP) ball and to determine whether or not it changed appreciably for various High Density Polyethylene (HDPE) reflected configurations. The motivation for this project was to determine whether or not the average number of neutrons emitted per fission, v, changed significantly enough to reduce the discrepancy between MCNP6 and Robba, Dowdy, Atwater (RDA) point kinetic model estimates of multiplication. The energy spectrum of neutrons that induced fissions in the BeRP ball, NIF (E), was also computed in order to determine the average energy of neutrons inducing fissions, NIF . p(v) was computed using the FMULT card, NIF (E) and NIF were computed using an F4 tally with an FM tally modifier (F4/FM) card, and the multiplication factor, keff, was computed using the KCODE card. Although NIF (E) changed significantly between bare and HDPE reflected configurations of the BeRP ball, the change in p(v), and thus the change in v, was insignificant. This is likely due to a difference between the way that NIF is computed using the FMULT and F4/FM cards. The F4/FM card indicated that NIF (E) was essentially Watt-fission distributed for a bare configuration and highly thermalized for all HDPE reflected configurations, while the FMULT card returned an average energy between 1 and 2 MeV for all configurations, which would indicate that the spectrum is Watt-fission distributed, regardless of the amount of HDPE reflector. The spectrum computed with the F4/FM cards is more physically meaningful and so the discrepancy between it and the FMULT card result is being investigated. It is hoped that resolving the discrepancy between the FMULT and F4/FM card estimates of NIF(E) will provide better v estimates that will lead to RDA multiplication estimates that are in better agreement with MCNP6 simulations.

  4. Absorbed body dose simulation in Thyroid cancer therapy using MCNP4Cand ITScodes and comparison to experimental results

    International Nuclear Information System (INIS)

    Hadad, K.; Gorji, Y.

    2004-01-01

    Two standard particle transport codes of MCNP4C and integrated tiger series were used to estimate the total body dose in a thyroid cancer therapy study, with I-131 as the radionuclide source. Human body was modeled by water and soft tissue ellipsoids. Phantoms' dimensions were selected according to Brow nell recommendation. Absorbed fractions were calculated by both codes for different phantoms and for gammas with 0.364 MeV energy, which has the highest fraction in I-131 emitting gammas. Results were compared to the data published by Brow nell et.al.. Figure 1 shows the results of MCNP4C and Integrated Tiger Series with results published by Brow nell et. al.

  5. Testing of the ENDF/B-VI neutron data library ENDF60 for use with MCNP trademark

    International Nuclear Information System (INIS)

    Frankle, S.C.; MacFarlane, R.E.

    1995-01-01

    The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N-Particle radiation transport code MCNP4A, was released in the fall of 1994. It is comprised of 124 nuclide data files based on the ENDF/B-Vi evaluations through Release 2. Forty-eight percent of these materials are new or modified evaluations, while the balance are translations from ENDF/B-V. The new evaluations include most of the important materials for criticality safety calculations, and include significant enhancements such as more isotopic evaluations, better resonance-range representations, and the new correlated energy-angle distributions for emitted particles. As part of the overall quality assurance testing of the ENDF60 library, calculations for well known benchmark assemblies were performed. The results of these calculations help the user to know how the combination of ENDF60 and MCNP4A will perform for real problems

  6. 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)

  7. 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

  8. SU-F-T-140: Assessment of the Proton Boron Fusion Reaction for Practical Radiation Therapy Applications Using MCNP6

    International Nuclear Information System (INIS)

    Adam, D; Bednarz, B

    2016-01-01

    Purpose: The proton boron fusion reaction is a reaction that describes the creation of three alpha particles as the result of the interaction of a proton incident upon a 11B target. Theoretically, the proton boron fusion reaction is a desirable reaction for radiation therapy applications in that, with the appropriate boron delivery agent, it could potentially combine the localized dose delivery protons exhibit (Bragg peak) and the local deposition of high LET alpha particles in cancerous sites. Previous efforts have shown significant dose enhancement using the proton boron fusion reaction; the overarching purpose of this work is an attempt to validate previous Monte Carlo results of the proton boron fusion reaction. Methods: The proton boron fusion reaction, 11B(p, 3α), is investigated using MCNP6 to assess the viability for potential use in radiation therapy. Simple simulations of a proton pencil beam incident upon both a water phantom and a water phantom with an axial region containing 100ppm boron were modeled using MCNP6 in order to determine the extent of the impact boron had upon the calculated energy deposition. Results: The maximum dose increase calculated was 0.026% for the incident 250 MeV proton beam scenario. The MCNP simulations performed demonstrated that the proton boron fusion reaction rate at clinically relevant boron concentrations was too small in order to have any measurable impact on the absorbed dose. Conclusion: For all MCNP6 simulations conducted, the increase of absorbed dose of a simple water phantom due to the 11B(p, 3α) reaction was found to be inconsequential. In addition, it was determined that there are no good evaluations of the 11B(p, 3α) reaction for use in MCNPX/6 and further work should be conducted in cross section evaluations in order to definitively evaluate the feasibility of the proton boron fusion reaction for use in radiation therapy applications.

  9. MCNP6.1 simulations for low-energy atomic relaxation: Code-to-code comparison with GATEv7.2, PENELOPE2014, and EGSnrc

    Science.gov (United States)

    Jung, Seongmoon; Sung, Wonmo; Lee, Jaegi; Ye, Sung-Joon

    2018-01-01

    Emerging radiological applications of gold nanoparticles demand low-energy electron/photon transport calculations including details of an atomic relaxation process. Recently, MCNP® version 6.1 (MCNP6.1) has been released with extended cross-sections for low-energy electron/photon, subshell photoelectric cross-sections, and more detailed atomic relaxation data than the previous versions. With this new feature, the atomic relaxation process of MCNP6.1 has not been fully tested yet with its new physics library (eprdata12) that is based on the Evaluated Atomic Data Library (EADL). In this study, MCNP6.1 was compared with GATEv7.2, PENELOPE2014, and EGSnrc that have been often used to simulate low-energy atomic relaxation processes. The simulations were performed to acquire both photon and electron spectra produced by interactions of 15 keV electrons or photons with a 10-nm-thick gold nano-slab. The photon-induced fluorescence X-rays from MCNP6.1 fairly agreed with those from GATEv7.2 and PENELOPE2014, while the electron-induced fluorescence X-rays of the four codes showed more or less discrepancies. A coincidence was observed in the photon-induced Auger electrons simulated by MCNP6.1 and GATEv7.2. A recent release of MCNP6.1 with eprdata12 can be used to simulate the photon-induced atomic relaxation.

  10. 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)

  11. Comparison of MCNP6 and experimental results for neutron counts, Rossi-α, and Feynman-α distributions

    International Nuclear Information System (INIS)

    Talamo, A.; Gohar, Y.; Sadovich, S.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.

    2013-01-01

    MCNP6, the general-purpose Monte Carlo N-Particle code, has the capability to perform time-dependent calculations by tracking the time interval between successive events of the neutron random walk. In fixed-source calculations for a subcritical assembly, the zero time value is assigned at the moment the neutron is emitted by the external neutron source. The PTRAC and F8 cards of MCNP allow to tally the time when a neutron is captured by 3 He(n, p) reactions in the neutron detector. From this information, it is possible to build three different time distributions: neutron counts, Rossi-α, and Feynman-α. The neutron counts time distribution represents the number of neutrons captured as a function of time. The Rossi-a distribution represents the number of neutron pairs captured as a function of the time interval between two capture events. The Feynman-a distribution represents the variance-to-mean ratio, minus one, of the neutron counts array as a function of a fixed time interval. The MCNP6 results for these three time distributions have been compared with the experimental data of the YALINA Thermal facility and have been found to be in quite good agreement. (authors)

  12. 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)

  13. 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)

  14. Human eye analytical and mesh-geometry models for ophthalmic dosimetry using MCNP6

    International Nuclear Information System (INIS)

    Angelocci, Lucas V.; Fonseca, Gabriel P.; Yoriyaz, Helio

    2015-01-01

    Eye tumors can be treated with brachytherapy using Co-60 plaques, I-125 seeds, among others materials. The human eye has regions particularly vulnerable to ionizing radiation (e.g. crystalline) and dosimetry for this region must be taken carefully. A mathematical model was proposed in the past [1] for the eye anatomy to be used in Monte Carlo simulations to account for dose distribution in ophthalmic brachytherapy. The model includes the description for internal structures of the eye that were not treated in previous works. The aim of this present work was to develop a new eye model based on the Mesh geometries of the MCNP6 code. The methodology utilized the ABAQUS/CAE (Simulia 3DS) software to build the Mesh geometry. For this work, an ophthalmic applicator containing up to 24 model Amersham 6711 I-125 seeds (Oncoseed) was used, positioned in contact with a generic tumor defined analytically inside the eye. The absorbed dose in eye structures like cornea, sclera, choroid, retina, vitreous body, lens, optical nerve and optical nerve wall were calculated using both models: analytical and MESH. (author)

  15. Comparison of MCNP and Experimental Measurements for an HPGe-based Spectroscopy Portal Monitor

    International Nuclear Information System (INIS)

    Keyser, Ronald M.; Hensley, Walter K.; Twomey, Timothy R.; UPP, Daniel L.

    2008-01-01

    The necessity to monitor international commercial transportation for illicit nuclear materials resulted in the installation of many nuclear radiation detection systems in Portal Monitors. These were mainly gross counters which alarmed at any indication of high radioactivity in the shipment, the vehicle or even the driver. The innocent alarm rate, due to legal shipments of sources and NORM, or medical isotopes in patients, caused interruptions and delays in commerce while the legality of the shipment was verified. To overcome this difficulty, Department of Homeland Security (DHS) supported the writing of the ANSI N42.38 standard (Performance Criteria for Spectroscopy-Based Portal Monitors used for Homeland Security) to define the performance of a Portal Monitor with nuclide identification capabilities, called a Spectroscopy Portal Monitor. This standard defines detection levels and response characteristics for the system for energies from 25 keV to 3 MeV. To accomplish the necessary performance, several different HPGe detector configurations were modeled using MCNP for the horizontal field of view (FOV) and vertical linearity of response over the detection zone of 5 meters by 4.5 meters for 661 keV as representative of the expected nuclides of interest. The configuration with the best result was built and tested. The results for the FOV as a function of energy and the linearity show good agreement with the model and performance exceeding the requirements of N42.38

  16. k0-PGNAA of pollutants in aqueous samples using MCNP code

    Directory of Open Access Journals (Sweden)

    A. Hamid

    2014-03-01

    Full Text Available Prompt γ-neutron activation analysis (PGNAA using the k0 method by employing the 1951.1 keV γ-line of the 35Cl(n, γ36Cl thermal neutron reaction as monostandard comparator was described. The method has been applied and evaluated using the anti-Compton prompt γ-ray neutron activation analysis facility using 252Cf neutron source with a neutron flux of 6.16 · 106 n · cm-2 · s-1. A well-type HPGe detector as the main detector surrounded by NaI(Tl guard detector has been arranged to investigate the performance of the Compton suppression spectrometer using the simplified slow circuit. The properties of neutron flux were determined by MCNP code calculations. In order to determine the efficiency curve of an HPGe detector, the prompt γ-rays from chlorine were used and an exponential curve was fitted. AC-PGNAA method has been used for the determination of high neutron absorbing elements like Cd, Sm and Gd as well as 20 light and heavy elements (Na, Mg, Al, Si, P, K, Ca, Ti, V, Mn, Sc, Fe, Co, Zn, La, Rb, Cs, As and Th in standard reference materials (IAEA, Soil-7 and ten sediment samples collected from El-Manzala lake in northern part of Egypt. The reference material IAEA, Soil-7 was analyzed for data validation and good agreement between the experimental values and the certified values have been obtained.

  17. Human eye analytical and mesh-geometry models for ophthalmic dosimetry using MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Angelocci, Lucas V.; Fonseca, Gabriel P.; Yoriyaz, Helio, E-mail: hyoriyaz@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Eye tumors can be treated with brachytherapy using Co-60 plaques, I-125 seeds, among others materials. The human eye has regions particularly vulnerable to ionizing radiation (e.g. crystalline) and dosimetry for this region must be taken carefully. A mathematical model was proposed in the past [1] for the eye anatomy to be used in Monte Carlo simulations to account for dose distribution in ophthalmic brachytherapy. The model includes the description for internal structures of the eye that were not treated in previous works. The aim of this present work was to develop a new eye model based on the Mesh geometries of the MCNP6 code. The methodology utilized the ABAQUS/CAE (Simulia 3DS) software to build the Mesh geometry. For this work, an ophthalmic applicator containing up to 24 model Amersham 6711 I-125 seeds (Oncoseed) was used, positioned in contact with a generic tumor defined analytically inside the eye. The absorbed dose in eye structures like cornea, sclera, choroid, retina, vitreous body, lens, optical nerve and optical nerve wall were calculated using both models: analytical and MESH. (author)

  18. Progress of conversion system from CAD data to MCNP geometry data in Japan

    International Nuclear Information System (INIS)

    Sato, S.; Nashif, H.; Masuda, F.; Morota, H.; Iida, H.; Konno, C.

    2010-01-01

    Automatic conversion systems from CAD data to MCNP geometry input data have been developed to convert the CAD data of the fusion reactor with very complicated structure. So far, two conversion systems (GEOMIT-1 and ARCMCP) have been developed and the third system (GEOMIT-2) is under developing. The void data can be created in these systems. GEOMIT-1 was developed in 2007, but a lot of manual shape splitting work for the CAD data was required to convert the complicated geometry. ARCMCP was developed in 2008. The algorithm has been drastically improved on automatic creation of ambiguous surface in ARCMCP, but it still required a little manual shape splitting work. The latest system, GEOMIT-2, does not require additional commercial software packages, though the previous systems require them. It also has functions of the CAD data healing and the automatic shape splitting. Geometrical errors of CAD data can be automatically revised by the healing function, and complicated geometries can be automatically split into simple geometries by the shape splitting function. Any manual works for CAD data are not required in GEOMIT-2. GEOMIT-2 is very useful for nuclear analyses of fusion reactors.

  19. 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)

  20. Criticality benchmark results for the ENDF60 library with MCNP trademark

    International Nuclear Information System (INIS)

    Keen, N.D.; Frankle, S.C.; MacFarlane, R.E.

    1995-01-01

    The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N-Particle radiation transport code MCNP4A, was released in the fall of 1994. The ENDF60 library is comprised of 124 nuclide data files based on the ENDF/B-VI (B-VI) evaluations through Release 2. Fifty-two percent of these B-VI evaluations are translations from ENDF/B-V (B-V). The remaining forty-eight percent are new evaluations which have sometimes changed significantly. Among these changes are greatly increased use of isotopic evaluations, more extensive resonance-parameter evaluations, and energy-angle correlated distributions for secondary particles. In particular, the upper energy limit for the resolved resonance region of 235 U, 238 U and 239 Pu has been extended from 0.082, 4.0, and 0.301 keV to 2..25, 10.0, and 2.5 keV respectively. As regulatory oversight has advanced and performing critical experiments has become more difficult, there has been an increased reliance on computational methods. For the criticality safety community, the performance of the combined transport code and data library is of interest. The purpose of this abstract is to provide benchmarking results to aid the user in determining the best data library for their application

  1. Study of salinity in aqueous medium using X-Ray beam with MCNP-X code

    International Nuclear Information System (INIS)

    Barbosa, Caroline M.; Braz, Delson

    2017-01-01

    In offshore production, it is possible that the produced water presents geochemical characteristics that correspond to the mixture of formation water (connate water) and the sea water (injection water), and the physical-chemical behavior of the injected water allows a considerable variation in the index of salinity altering the water/oil ratio during transportation and/or extraction. Injection water is generally used to raise the reservoir pressure, increasing the percentage of extracted oil. This water has a significant amount of salts that generate some difficulties, such as measuring fractions of volume in multiphase systems. One way to check the effects of salinity would be to regularly measure the amount of salt present in the water. In this way, this work presents a methodology to measure the concentration and the types of salts using nuclear techniques through the MCNP-X computational code. The measurement geometry uses an X-ray beam (40-100 keV) and NaI(Tl) scintillation detector positioned diametrically opposed to the source. The studied samples were the NaCl, KCl and MgCl 2 salts in aqueous solution. The results present the possibility of differentiating the formation and injection waters due to differences in the salt concentrations. (author)

  2. Study of salinity in aqueous medium using X-Ray beam with MCNP-X code

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Caroline M.; Braz, Delson [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, César M., E-mail: cbarbosa@nuclear.ufrj.br, E-mail: delson@nuclear.ufrj.br, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In offshore production, it is possible that the produced water presents geochemical characteristics that correspond to the mixture of formation water (connate water) and the sea water (injection water), and the physical-chemical behavior of the injected water allows a considerable variation in the index of salinity altering the water/oil ratio during transportation and/or extraction. Injection water is generally used to raise the reservoir pressure, increasing the percentage of extracted oil. This water has a significant amount of salts that generate some difficulties, such as measuring fractions of volume in multiphase systems. One way to check the effects of salinity would be to regularly measure the amount of salt present in the water. In this way, this work presents a methodology to measure the concentration and the types of salts using nuclear techniques through the MCNP-X computational code. The measurement geometry uses an X-ray beam (40-100 keV) and NaI(Tl) scintillation detector positioned diametrically opposed to the source. The studied samples were the NaCl, KCl and MgCl{sub 2} salts in aqueous solution. The results present the possibility of differentiating the formation and injection waters due to differences in the salt concentrations. (author)

  3. 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)

  4. 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)

  5. MCNP Variance Reduction technique application for the Development Of the Citrusdal Irradiation Facility

    International Nuclear Information System (INIS)

    Makgae, R.

    2008-01-01

    A private company, Citrus Research International (CIR) is intending to construct an insect irradiation facility for the irradiation of insect for pest management in south western region of South Africa. The facility will employ a Co-60 cylindrical source in the chamber. An adequate thickness for the concrete shielding walls and the ability of the labyrinth leading to the irradiation chamber, to attenuate radiation to dose rates that are acceptably low, were determined. Two methods of MCNP variance reduction techniques were applied to accommodate the two pathways of deep penetration to evaluate the radiological impact outside the 150 cm concrete walls and steaming of gamma photons through the labyrinth. The point-kernel based MicroShield software was used in the deep penetration calculations for the walls around the source room to test its accuracy and the results obtained are in good agreement with about 15-20% difference. The dose rate mapping due to radiation Streaming along the labyrinth to the facility entrance is also to be validated with the Attila code, which is a deterministic code that solves the Discrete Ordinates approximation. This file provides a template for writing papers for the conference. (authors)

  6. MCNP Variance Reduction technique application for the Development Of the Citrusdal Irradiation Facility

    Energy Technology Data Exchange (ETDEWEB)

    Makgae, R. [Pebble Bed Modular Reactor (PBMR), P.O. Box 9396, Centurion (South Africa)

    2008-07-01

    A private company, Citrus Research International (CIR) is intending to construct an insect irradiation facility for the irradiation of insect for pest management in south western region of South Africa. The facility will employ a Co-60 cylindrical source in the chamber. An adequate thickness for the concrete shielding walls and the ability of the labyrinth leading to the irradiation chamber, to attenuate radiation to dose rates that are acceptably low, were determined. Two methods of MCNP variance reduction techniques were applied to accommodate the two pathways of deep penetration to evaluate the radiological impact outside the 150 cm concrete walls and steaming of gamma photons through the labyrinth. The point-kernel based MicroShield software was used in the deep penetration calculations for the walls around the source room to test its accuracy and the results obtained are in good agreement with about 15-20% difference. The dose rate mapping due to radiation Streaming along the labyrinth to the facility entrance is also to be validated with the Attila code, which is a deterministic code that solves the Discrete Ordinates approximation. This file provides a template for writing papers for the conference. (authors)

  7. An MCNP parametric study of George C. Laurence's subcritical pile experiment

    International Nuclear Information System (INIS)

    Dranga, R.; Blomeley, L.; Carrington, R.

    2014-01-01

    In the early 1940s at the National Research Council (NRC) Laboratories in Ottawa, Canada, Dr. George Laurence conducted several experiments to determine if a sustained nuclear fission chain reaction in a carbon-uranium arrangement (or 'pile') was possible. Although Dr. Laurence did not achieve criticality, these pioneering experiments marked a significant historical event in nuclear science, and they provided a valuable reference for subsequent experiments that led to the design of Canada's first heavy-water reactors at the Chalk River Nuclear Laboratories. This paper summarizes the results of a recent collaborative project between Atomic Energy of Canada Limited and the Deep River Science Academy undertaken to numerically explore the experiments carried out at the NRC Laboratories by Dr. Laurence, while teaching high school students about nuclear science and technology. In this study, a modern Monte Carlo reactor physics code, MCNP6, was utilized to identify and study the key parameters impacting the subcritical pile's neutron multiplication factor (e.g., moderation, geometry, material impurities) and quantify their effect on the extent of subcriticality. The findings presented constitute the first endeavour to model, using a current computational reactor physics tool, the seminal experiment that provided the foundation of Canada's nuclear science and technology program. (author)

  8. Simulation of the BNCT of Brain Tumors Using MCNP Code: Beam Designing and Dose Evaluation

    Directory of Open Access Journals (Sweden)

    Fatemeh Sadat Rasouli

    2012-09-01

    Full Text Available Introduction BNCT is an effective method to destroy brain tumoral cells while sparing the healthy tissues. The recommended flux for epithermal neutrons is 109 n/cm2s, which has the most effectiveness on deep-seated tumors. In this paper, it is indicated that using D-T neutron source and optimizing of Beam Shaping Assembly (BSA leads to treating brain tumors in a reasonable time where all IAEA recommended criteria are met. Materials and Methods The proposed BSA based on a D-T neutron generator consists of a neutron multiplier system, moderators, reflector, and collimator. The simulated Snyder head phantom is used to evaluate dose profiles in tissues due to the irradiation of designed beam. Monte Carlo Code, MCNP-4C, was used in order to perform these calculations.   Results The neutron beam associated with the designed and optimized BSA has an adequate epithermal flux at the beam port and neutron and gamma contaminations are removed as much as possible. Moreover, it was showed that increasing J/Φ, as a measure of beam directionality, leads to improvement of beam performance and survival of healthy tissues surrounding the tumor. Conclusion According to the simulation results, the proposed system based on D-T neutron source, which is suitable for in-hospital installations, satisfies all in-air parameters. Moreover, depth-dose curves investigate proper performance of designed beam in tissues. The results are comparable with the performances of other facilities.

  9. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Directory of Open Access Journals (Sweden)

    Hegazy Aya Hamdy

    2018-01-01

    Full Text Available Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1 shielding-collimator material, (2 Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3 thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  10. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Science.gov (United States)

    Hegazy, Aya Hamdy; Skoy, V. R.; Hossny, K.

    2018-04-01

    Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  11. Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

    2010-01-01

    The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

  12. Performance of the MTR core with MOX fuel using the MCNP4C2 code.

    Science.gov (United States)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-08-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

  13. Performance of the MTR core with MOX fuel using the MCNP4C2 code

    International Nuclear Information System (INIS)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-01-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U 3 O 8 &PuO 2 ) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U 3 O 8 -Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U 3 O 8 -Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235 U and the amount of loaded 235 U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. - Highlights: • Re-cycling of the ETRR-2 reactor by MOX fuel. • Increase the number of the neutronic traps from one neutronic trap to three neutronic trap. • Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U 3 O 8 -Al original fuel and the MOX fuel.

  14. Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Vanessa M.; Oliveira, Renato C.M., E-mail: vanessamachado@ufmg.br [Curso Superior de Tecnologia em Radiologia. Faculdade de Medicina da Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil); Barros, Graiciany P.; Oliveira, Arno H.; Veloso, M. Auxiliadora F. [Departamento de Engenharia Nuclear. Escola de Engenharia. Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil)

    2011-07-01

    One of basic principles for radiological protection is the optimization of techniques for obtain radiographic images, in way that the dose in the patient is kept as low as reasonably achievable (ALARA). Intensifying screens are used in medical radiology, which reduce considerably the dose rates in the production of radiographic images, maintaining the quality of these, while in dental radiology, there is no a intensifying screen available for intraoral examinations. From this technological requirement, this paper evaluates a computational modeling of an intensifying screen for use in intraoral radiography. For this, it was used the Monte Carlo code MCNP5 that allows the radiography simulation through the transport of electrons and photons in the different materials present in this examination. The goal of an intensifying screen is the conversion of X-ray photons to photons in the visible spectrum, knowing that radiographic films are more sensitive to light photons than to X-ray photons. So the screen should be composed of an efficient material for converting x-rays photons in light photons, therefore was made simulations using different materials, thicknesses and positions possible for placing screen in radiographic film in order to find the way more technically feasible. (author)

  15. Comparative dosimetry of prostate brachytherapy with I-125 and Pd-103 seeds via SISCODES/MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Trindade, Bruno Machado; Falcao, Patricia Lima, E-mail: bmtrindade@yahoo.com [Nucleo de Radiacoes Ionizantes - Universidade Federal de Minas Gerais (NRI/UFMG), Belo Horizonte, MG (Brazil); Christovao, Marilia Tavares [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Trindade, Daniela de Fatima Maia [Centro Universitario Una, Belo Horizonte, MG (Brazil); Campos, Tarcisio Passos Ribeiro de [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2012-09-15

    Objective: The present paper is aimed at presenting a comparative dosimetric study of prostate brachytherapy with I-125 and Pd-103 seeds. Materials and Methods: A protocol for both implants with 148 seeds was simulated on a heterogeneous three-dimensional pelvic phantom by means of the SISCODES/MCNP5 codes. Dose-volume histograms on prostate, rectum and bladder, dose indexes D10, D30, D90, D0.5cc, D2cc and D7cc, and representations of the spatial dose distribution were evaluated. Results: For a D90 index equivalent to the prescription dose, the initial activity of each I-125 seed was calculated as 0.42 mCi and of Pd-103 as 0.94 mCi. The maximum dose on the urethra was 90% and 108% of the prescription dose for I-125 and Pd-103, respectively. The D2cc for I-125 was 30 Gy on the rectum and 127 Gy on the bladder; for Pd-103 was 29 Gy on the rectum and 189 Gy on the bladder. The D10 on the pubic bone was 144 Gy for I-125 and 66 Gy for Pd-103. Conclusion: The results indicate that Pd-103 and I-125 implants could deposit the prescribed dose on the target volume. Among the findings of the present study, there is an excessive radiation exposure of the pelvic bones, particularly with the I-125 protocol. (author)

  16. Estudio teórico de la desorción de Na y K de SiO2 estimulada por la acción de fotones o electrones

    Science.gov (United States)

    Domínguez Ariza, D.; López, N.; Illas, F.; Pacchioni, G.; Madey, T. E.

    Se ha estudiado el mecanismo de generación de sodio y potasio atómico a partir de muestras de SiO2 utilizando cálculos basados tanto en la teoría del funcional de la densidad como en métodos post-Hartree Fock, así como en el método de cluster para modelar el sólido. Como consecuencia del estudio se han propuesto distintos caminos posibles para la desorción, estimulada por la acción de fotones o electrones, de sodio y potasio desde el óxido de silicio, proporcionando por lo tanto una explicación a la atmósfera tenue de sodio y potasio de La Luna.

  17. Implementation of On-the-Fly Doppler Broadening in MCNP5 for Multiphysics Simulation of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William Martin

    2012-11-16

    A new method to obtain Doppler broadened cross sections has been implemented into MCNP, removing the need to generate cross sections for isotopes at problem temperatures. Previous work had established the scientific feasibility of obtaining Doppler-broadened cross sections "on-the-fly" (OTF) during the random walk of the neutron. Thus, when a neutron of energy E enters a material region that is at some temperature T, the cross sections for that material at the exact temperature T are immediately obtained by interpolation using a high order functional expansion for the temperature dependence of the Doppler-broadened cross section for that isotope at the neutron energy E. A standalone Fortran code has been developed that generates the OTF library for any isotope that can be processed by NJOY. The OTF cross sections agree with the NJOY-based cross sections for all neutron energies and all temperatures in the range specified by the user, e.g., 250K - 3200K. The OTF methodology has been successfully implemented into the MCNP Monte Carlo code and has been tested on several test problems by comparing MCNP with conventional ACE cross sections versus MCNP with OTF cross sections. The test problems include the Doppler defect reactivity benchmark suite and two full-core VHTR configurations, including one with multiphysics coupling using RELAP5-3D/ATHENA for the thermal-hydraulic analysis. The comparison has been excellent, verifying that the OTF libraries can be used in place of the conventional ACE libraries generated at problem temperatures. In addition, it has been found that using OTF cross sections greatly reduces the complexity of the input for MCNP, especially for full-core temperature feedback calculations with many temperature regions. This results in an order of magnitude decrease in the number of input lines for full-core configurations, thus simplifying input preparation and reducing the potential for input errors. Finally, for full-core problems with multiphysics

  18. 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 keff.

  19. Study of radiation dose attenuation by skull bone in head during radiotherapy treatment using MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Menezes, Artur F.; Boia, Leonardo S.; Trombetta, Debora M.; Martins, Maximiano C.; Reis Junior, Juraci P.; Silva, Ademir X., E-mail: ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Batista, Delano V.S., E-mail: delano@inca.gov.b [Instituto Nacional do Cancer (INCa), Rio de Janeiro, RJ (Brazil). Dept. de Fisica Medica

    2011-07-01

    In this study the MCNPX code was used to investigate possible influences of the attenuation beam by the surface bone during radiotherapy treatments of the skull. The computer simulation was performed on topographic image obtained from the National Cancer Institute, in Rio de Janeiro, database of patients treated with radiotherapy. The image segmentation process were performed using the SAPDI program developed to this purpose. The segmented image conversion for the input file recognized by MCNPX code was performed by SCAN2MCNP Software. The simulation was done using 10MeV Clinac 2300C spectrum considering two opposite parallel beams, with field size 2x2 and 4x4 cm{sup 2}, incident on a slice located above the eyes, containing two row of detectors positioned on the central region with a radius of 0.03 cm and arranged perpendicular to the radiation beams. After analyze the results, the relative error values in the range of 2 at 4% for the high dose region, and 26 at 37% for the low dose area were found, respectively. These differences were attributed to the radiation field attenuation on the bone surface at the entrance of the beam. It was observed that most situations on the high dose region the beam profile, from more realistic scenarios, became smaller than the one obtained when the tomography image was considered consisting of water. However for the low dose area the profile, obtained of the realistic situation, became higher than the one which was obtained when the tomography image was considered consisting of water. The results showed significant differences between both analyzed cases which show the need to use a correction factor by the treatment planning system used in radiotherapy services when the real chemical composition of patient head is unconsidered during the patient treatment planning. (author)

  20. Study of radiation dose attenuation by skull bone in head during radiotherapy treatment using MCNP

    International Nuclear Information System (INIS)

    Menezes, Artur F.; Boia, Leonardo S.; Trombetta, Debora M.; Martins, Maximiano C.; Reis Junior, Juraci P.; Silva, Ademir X.; Batista, Delano V.S.

    2011-01-01

    In this study the MCNPX code was used to investigate possible influences of the attenuation beam by the surface bone during radiotherapy treatments of the skull. The computer simulation was performed on topographic image obtained from the National Cancer Institute, in Rio de Janeiro, database of patients treated with radiotherapy. The image segmentation process were performed using the SAPDI program developed to this purpose. The segmented image conversion for the input file recognized by MCNPX code was performed by SCAN2MCNP Software. The simulation was done using 10MeV Clinac 2300C spectrum considering two opposite parallel beams, with field size 2x2 and 4x4 cm 2 , incident on a slice located above the eyes, containing two row of detectors positioned on the central region with a radius of 0.03 cm and arranged perpendicular to the radiation beams. After analyze the results, the relative error values in the range of 2 at 4% for the high dose region, and 26 at 37% for the low dose area were found, respectively. These differences were attributed to the radiation field attenuation on the bone surface at the entrance of the beam. It was observed that most situations on the high dose region the beam profile, from more realistic scenarios, became smaller than the one obtained when the tomography image was considered consisting of water. However for the low dose area the profile, obtained of the realistic situation, became higher than the one which was obtained when the tomography image was considered consisting of water. The results showed significant differences between both analyzed cases which show the need to use a correction factor by the treatment planning system used in radiotherapy services when the real chemical composition of patient head is unconsidered during the patient treatment planning. (author)

  1. Dose calculation for {sup 40}K ingestion in samples of beans using spectrometry and MCNP; Calculo de dose devido a ingestao de {sup 40}K em amostras de feijao utilizando espectrometria e MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Garcez, R.W.D.; Lopes, J.M.; Silva, A.X., E-mail: marqueslopez@yahoo.com.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/PEN/UFRJ), Rio de Janeiro, RJ (Brazil). Centro de Tecnologia; Domingues, A.M. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Instituto de Fisica; Lima, M.A.F. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Instituto de Biologia

    2014-07-01

    A method based on gamma spectroscopy and on the use of voxel phantoms to calculate dose due to ingestion of {sup 40}K contained in bean samples are presented in this work. To quantify the activity of radionuclide, HPGe detector was used and the data entered in the input file of MCNP code. The highest value of equivalent dose was 7.83 μSv.y{sup -1} in the stomach for white beans, whose activity 452.4 Bq.Kg{sup -1} was the highest of the five analyzed. The tool proved to be appropriate when you want to calculate the dose in organs due to ingestion of food. (author)

  2. Photons and photoneutrons spectra of a Linac of 15 MV; Espectros de fotones y fotoneutrones de un LINAC de 15 MV

    Energy Technology Data Exchange (ETDEWEB)

    Benites R, J. L.; Carrillo C, A. [Centro Estatal de Cancerologia de Nayarit, Av. Enfermeria, Fracc. Fray Junipero Serra, 63000 Tepic, Nayarit (Mexico); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Velazquez F, J. B., E-mail: jlbenitesr@prodigy.net.mx [Universidad Autonoma de Nayarit, Posgrado CBAP, Carretera Tepic Compostela Km. 9, Xalisco, Nayarit (Mexico)

    2011-10-15

    Using the Monte Carlo code MCNP-5, the photons and photoneutrons spectra generated in the head stock of the lineal accelerator (Linac) Varian of 15 MV of the Cancerology State of Nayarit were determined. For the calculations a heterogeneous head stock was modeled, more compatible with the work conditions. In the center of the head stock a tungsten target was located on a copper support, followed by the flattened filter. The photons and photoneutrons spectra were obtained accelerating electrons and making them collide against the target to produce photons by Bremsstrahlung, these photons were transported inside the head stock and the photons and photoneutrons spectra were calculated in a punctual detector located under the flattened filter and in the isocenter. The spectra were evaluated in punctual detectors that were located in the plane from the isocenter to the long of the X and Y axes each 20 cm, in an equidistant way, up to 2 m, so much in the longitudinal and transversal axes. In the calculations were used histories 5E(6) with the purpose of obtaining smaller uncertainties to 1%. It was found that the photons spectrum in the punctual detector inside the head stock presents a pick of 1.25 MeV in the energy interval of 0.5 and 1.5 MeV, later suffers a filtration and diminishes in asymptote form. This spectrum modifies when the beam reaches the isocenter, diminishing the low energy photons. Inside the head stock the photoneutrons spectrum shows a structure with two picks, one before 1 MeV and other after 1 MeV; this is for effect of the collimators geometry and the distance. Finally an increment of the total neutrons flow to 60 cm of distance of the isocenter on the Y axis was observed, due to the design geometry of the modeling heterogeneous head stock. (Author)

  3. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Science.gov (United States)

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

  4. Gamma spectroscopy modelization intercomparison of the modelization results using two different codes (MCNP, and Pascalys-mercure)

    International Nuclear Information System (INIS)

    Luneville, L.; Chiron, M.; Toubon, H.; Dogny, S.; Huver, M.; Berger, L.

    2001-01-01

    The research performed in common these last 3 years by the French Atomic Commission CEA, COGEMA and Eurisys Mesures had for main subject the realization of a complete tool of modelization for the largest range of realistic cases, the Pascalys modelization software. The main purpose of the modelization was to calculate the global measurement efficiency, which delivers the most accurate relationship between the photons emitted by the nuclear source in volume, punctual or deposited form and the germanium hyper pure detector, which detects and analyzes the received photons. It has been stated since long time that experimental global measurement efficiency becomes more and more difficult to address especially for complex scene as we can find in decommissioning and dismantling or in case of high activities for which the use of high activity reference sources become difficult to use for both health physics point of view and regulations. The choice of a calculation code is fundamental if accurate modelization is searched. MCNP represents the reference code but its use is long time calculation consuming and then not practicable in line on the field. Direct line-of-sight point kernel code as the French Atomic Commission 3-D analysis Mercure code can represent the practicable compromise between the most accurate MCNP reference code and the realistic performances needed in modelization. The comparison between the results of Pascalys-Mercure and MCNP code taking in account the last improvements of Mercure in the low energy range where the most important errors can occur, is presented in this paper, Mercure code being supported in line by the recent Pascalys 3-D modelization scene software. The incidence of the intrinsic efficiency of the Germanium detector is also approached for the total efficiency of measurement. (authors)

  5. 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

  6. Criticality calculations of the HTR-10 pebble-bed reactor with SCALE6/CSAS6 and MCNP5

    International Nuclear Information System (INIS)

    Wang, Meng-Jen; Sheu, Rong-Jiun; Peir, Jinn-Jer; Liang, Jenq-Horng

    2014-01-01

    Highlights: • Comparisons of the HTR-10 criticality calculations with SCALE6/CSAS6 and MCNP5 were performed. • The DOUBLEHET unit-cell treatment provides the best k eff estimation among PBR criticality calculations using SCALE6. • The continuous-energy SCALE6 calculations present a non-negligible discrepancy with MCNP5 in three PBR cases. - Abstract: HTR-10 is a 10 MWt prototype pebble-bed reactor (PBR) that presents a doubly heterogeneous geometry for neutronics calculations. An appropriate unit-cell treatment for the associated fuel elements is vital for creating problem-dependent multigroup cross sections. Considering four unit-cell options for resonance self-shielding correction in SCALE6, a series of HTR-10 core models were established using the CSAS6 sequence to systematically investigate how they affected the computational accuracy and efficiency of PBR criticality calculations. Three core configurations, which ranged from simplified infinite lattices to a detailed geometry, were examined. Based on the same ENDF/B-VII.0 cross-section library, multigroup results were evaluated by comparing with continuous-energy SCALE6/CSAS6 and MCNP5 calculations. The comparison indicated that the INFHOMMEDIUM results overestimated the effective multiplication factor (k eff ) by about 2800 pcm, whereas the LATTICECELL and MULTIREGION treatments overestimated k eff values with similar biases at approximately 470–680 pcm. The DOUBLEHET results attained further improvement, reducing the k eff overestimation to approximately 280 pcm. The comparison yielded two unexpected problems from using SCALE6/CSAS6 in HTR-10 criticality calculations. In particular, the continuous-energy CSAS6 calculations in this study present a non-negligible discrepancy with MCNP5, potentially causing a k eff value overestimate of approximately 680 pcm. Notably, using a cell-weighted mixture instead of an explicit model of individual TRISO particles in the pebble fuel zone does not shorten the

  7. Reducción hidrotermal de CO2 usando sustancias orgánicas

    OpenAIRE

    Del Río Alegre, Olga

    2017-01-01

    En este trabajo se estudia la implantación industrial de la reducción hidrotermal de CO2 usando glucosa procedente de residuos de biomasa como sustancia reductora. La reducción hidrotermal de CO2 es una tecnología que consiste en convertir el CO2 en medio acuoso a alta presión y temperatura, preferentemente en forma de bicarbonato, en ácido fórmico. Como sustancias reductoras se han propuesto metales, residuos de polímeros o sustancias con grupos alcoholes, como es el caso de l...

  8. Conceptualización de campos magnéticos permanentes usando el material educativo

    OpenAIRE

    Espol; Roblero Wong, Jorge

    2017-01-01

    El propósito de este estudio fue desarrollar en los estudiantes la habilidad de conceptualización usando el aprendizaje colaborativo en la unidad de campo magnético con la ayuda de material educativo computarizado, dicho material fue elaborado utilizando el DBR (investigación basada en diseño), con dos intervenciones, la primera prueba para mejorar el diseño y la segunda para realizar la investigación. el sistema que usamos fue la combinación del aprendizaje social de Vigotsky y el Constructi...

  9. Interação Humano - Computador usando Visão Computacional

    Directory of Open Access Journals (Sweden)

    Bernardo Bucher B. Barbosa

    2015-07-01

    Full Text Available Este trabalho visa estudar maneiras de se explorar a Interação Humano Computador, usando Visão Computacional. A idéia tem como objetivo um esforço para tornar o computador mais interativo com o usuário, sem a necessidade da compra de um hardware ou acessório específico para tal. O produto final deste trabalho em desenvolvimento é um software que contempla esta funcionalidade, tornando o computador mais interativo.

  10. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

    2007-01-01

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

  11. 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.

  12. CLARIFICACIÓN DE AGUAS USANDO COAGULANTES POLIMERIZADOS: CASO DEL HIDROXICLORURO DE ALUMINIO

    Directory of Open Access Journals (Sweden)

    JUAN MIGUEL COGOLLO FLÓREZ

    2011-01-01

    Full Text Available En este artículo se realiza un estudio del proceso de clarificación en sistemas de tratamiento de aguas industriales usando un coagulante inorgánico polimerizado (hidroxicloruro de aluminio. Inicialmente, se establecen los elementos conceptuales más importantes de las etapas del proceso de clarificación (coagulación, floculación y sedimentación. Luego, se señalan los principales coagulantes convencionales utilizados en el tratamiento de aguas y se abordan los policloruros de aluminio (PAC´s como integrantes de una nueva generación de coagulantes alternativos cuyo uso se ha incrementado en las últimas décadas dado su mejor desempeño respecto a los coagulantes convencionales; se especifican los aspectos técnicos y operativos que se deben considerar al momento de implementar un proceso de clarificación de aguas usando un PAC como coagulante. Finalmente, se presentan datos comparativos de condiciones operacionales reales de un proceso de clarificación de aguas, producto de un trabajo previo, donde se remplazó un coagulante convencional (sulfato de aluminio por hidroxicloruro de aluminio, donde se corrobora el mejor desempeño del proceso luego del remplazo.

  13. 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)

  14. 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

  15. Neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) MCNP ''Benchmark CAD Model'' with the ATTILA discrete ordinance code

    International Nuclear Information System (INIS)

    Youssef, M.Z.; Feder, R.; Davis, I.

    2007-01-01

    The ITER IT has adopted the newly developed FEM, 3-D, and CAD-based Discrete Ordinates code, ATTILA for the neutronics studies contingent on its success in predicting key neutronics parameters and nuclear field according to the stringent QA requirements set forth by the Management and Quality Program (MQP). ATTILA has the advantage of providing a full flux and response functions mapping everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. The ITER neutronics community had agreed to use a standard CAD model of ITER (40 degree sector, denoted ''Benchmark CAD Model'') to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD-MCNP approach developed by each party. Since ATTILA seems to lend itself as a powerful design tool with minimal turnaround time, it was decided to benchmark this model with ATTILA as well and compare the results to those obtained with the CAD MCNP calculations. In this paper we report such comparison for five responses, namely: (1) Neutron wall load on the surface of the 18 shield blanket module (SBM), (2) Neutron flux and nuclear heating rate in the divertor cassette, (3) nuclear heating rate in the winding pack of the inner leg of the TF coil, (4) Radial flux profile across dummy port plug and shield plug placed in the equatorial port, and (5) Flux at seven point locations situated behind the equatorial port plug. (orig.)

  16. Photopeak efficiency response function of an underwater gamma-ray NaI(Tl) detector using MCNP-X

    Energy Technology Data Exchange (ETDEWEB)

    Salgado, William L., E-mail: william.otero@hotmail.com [Instituto Federal do Rio de Janeiro (IFRJ), RJ (Brazil); Silva, Ademir X., E-mail: ademir@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE-DNC/UFRJ/EE/CT), Rio de Janeiro, RJ (Brazil); Salgado, Cesar M., E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    This work presents a study to calculate the response function of a 1.5″ x 1″ NaI(Tl) scintillation detector when it is used in the marine environment in the energy range from 20 keV to 662 keV. The method takes into account both the scattering of photons in the water and the detection mechanism of the detector. In addition, the calculation of the response function of the whole system is essential for suppressing the background of the measurement and for estimating the concentration of the involved radionuclides, especially given the greater probability of primary gamma photons undergoing multiple scattering events before they interact with the detector. The experimental photopeak efficiency measurements for point sources were compared with the simulated results under the same conditions of the experimental setup to validate the simulation of the detector. Monte Carlo simulations were performed using the MCNP-X code for the investigation of gamma-ray absorption in water in different brines. The energy resolution curve was used to improve the response of the mathematical simulation of the detector. The detector’s simulation was based on information obtained from the gammagraphy technique. Both dimensions and materials were used for the calculation with the MCNP-X code. The photopeak efficiency of a NaI(Tl) detector for different radionuclides in the aquatic environment with different salinities was calculated. (author)

  17. TRIPOLI-4 green's functions and MCNP5 importance to estimate ex-core detector response on a N4 PWR

    International Nuclear Information System (INIS)

    Trakas, C.; Petit, O

    2010-01-01

    Monitoring power reactors for the critical and sub-critical states relies on the importance of neutron assemblies or fuel rods, relatively to the parameters of interest. These parameters can be the reactor power or its variation, the maximum expected fluence on the vessel, the signal of ex-core detectors in a sub-critical core, the neutron and gamma energy deposited outside the core, etc. In general, the neutron importance can be obtained using direct Monte Carlo calculations. Thus, with successive transport calculations of neutrons or gamma, we obtain the contribution of each part to the signal of interest. It can also be obtained by adjoint calculations using SN deterministic codes. Both methods are currently used by AREVA. Here we present a study for neutron importance of a new and computationally very efficient method, proposed by the TRIPOLI-4 Monte Carlo transport code and we compare results to a MCNP5 importance calculation. The neutron importance is provided by the TRIPOLI-4-Green's functions option. The results show an excellent agreement between the two methodologies applied with the codes. Importance calculated by MCNP5 and TRIPOLI-4 for 10 B tallies have discrepancies less than 1% for the first row of fuel assemblies and 6% for the 2nd and 3rd row. Similar results were obtained for fast neutrons. (author)

  18. Simulation of irradiation exposure of electronic devices due to heavy ion therapy with Monte Carlo Code MCNP6

    Science.gov (United States)

    Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang; Buck, Arnulf

    2017-09-01

    During heavy ion irradiation therapy the patient has to be located exactly at the right position to make sure that the Bragg peak occurs in the tumour. The patient has to be moved in the range of millimetres to scan the ill tissue. For that reason a special table was developed which allows exact positioning. The electronic control can be located outside the surgery. But that has some disadvantage for the construction. To keep the system compact it would be much more comfortable to put the electronic control inside the surgery. As a lot of high energetic secondary particles are produced during the therapy causing a high dose in the room it is important to find positions with low dose rates. Therefore, investigations are needed where the electronic devices should be located to obtain a minimum of radiation, help to prevent the failure of sensitive devices. The dose rate was calculated for carbon ions with different initial energy and protons over the entire therapy room with Monte Carlo particle tracking using MCNP6. The types of secondary particles were identified and the dose rate for a thin silicon layer and an electronic mixture material was determined. In addition, the shielding effect of several selected material layers was calculated using MCNP6.

  19. Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0

    International Nuclear Information System (INIS)

    Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa

    2003-01-01

    A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)

  20. Simulation of dose deposition in heterogeneities in the human body, using the Penelope code for photons beams of energies of a linear accelerator; Simulacion de la deposicion de dosis en las heterogeneidades del cuerpo humano, usando el codigo Penelope para haces de fotones de energias de un acelerador lineal

    Energy Technology Data Exchange (ETDEWEB)

    Cardena R, A. R.; Vega R, J. L.; Apaza V, D. G., E-mail: cardroj@yahoo.es [Universidad Nacional de San Agustin, Av. Independencia s/n, Arequipa (Peru)

    2015-10-15

    The progress in cancer treatment systems in heterogeneities of human body has had obstacles by the lack of a suitable experimental model test. The only option is to develop simulated theoretical models that have the same properties in interfaces similar to human tissues, to know the radiation behavior in the interaction with these materials. In this paper we used the Monte Carlo method by Penelope code based solely on studies for the cancer treatment as well as for the calibration of beams and their various interactions in mannequins. This paper also aims the construction, simulation and characterization of an equivalent object to the tissues of the human body with various heterogeneities, we will later use to control and plan experientially doses supplied in treating tumors in radiotherapy. To fulfill the objective we study the ionizing radiation and the various processes occurring in the interaction with matter; understanding that to calculate the dose deposited in tissues interfaces (percentage depth dose) must be taken into consideration aspects such as the deposited energy, irradiation fields, density, thickness, tissue sensitivity and other items. (Author)

  1. 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)

  2. Determination of the detection efficiency of a HPGe detector by means of the MCNP 4A simulation code; Determinacion de la eficiencia de deteccion de un detector HPGe mediante el codigo de simulacion MCNP 4A

    Energy Technology Data Exchange (ETDEWEB)

    Leal, B. [Centro Regional de Estudios Nucleares, A.P. 579C, 98068 Zacatecas (Mexico)

    2004-07-01

    In the majority of the laboratories, the calibration in efficiency of the detector is carried out by means of the standard sources measurement of gamma photons that have a determined activity, or for matrices that contain a variety of radionuclides that can embrace the energy range of interest. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the energy range of 80 keV to 1400 keV varying the density of the matrix, by means of the application of the Monte Carlo code MCNP-4A. The adjustment obtained shows an acceptance grade in the range of 100 to 600 keV, with a smaller percentage discrepancy to 5%. (Author)

  3. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  4. 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

  5. Validation of MCNP4a for highly enriched uranium using the Battelle process safety and risk management IBM RS/6000 workstation

    Energy Technology Data Exchange (ETDEWEB)

    Negron, S.B.; Lee, B.L. Jr.; Tayloe, R.W. Jr.

    1996-01-01

    This document has been prepared to allow use of the Radiation Shielding and Information Center (RSIC) release of MCNP4a, which has been installed on the Battelle Process Safety and Risk Management (PSRM) IBM RS/6000 workstation, for production calculations for the Portsmouth Gaseous Diffusion Plant (PORTS). This hardware/software configuration is under the configuration control plan listed in Reference 1. The first portion of this document outlines basic information with regard to validation of MCNP4a using the supplied cross sections and the additional MCNPDAT cross sections. A basic discussion of MCNP is provided, along with discussions of the validation database in general. A description of the statistical analysis then follows. The results of this validation indicate that the software and data libraries examined may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant (PORTS). When the validation results are treated as a single group, there is a 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated k{sub eff} > 0.95. Based on this result, the Battelle PSRM Nuclear Safety Group has adopted the calculational acceptance criteria that a calculated k{sub eff} + 2{sigma}, {le} 0.95 is safely subcritical. The conclusion of this document is that MCNP4a and all associated cross section libraries installed on the PSRM IBM RS/6000 are acceptable for use in performing production criticality safety calculations for the Portsmouth Gaseous Diffusion Plant.

  6. Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

    Science.gov (United States)

    2014-03-27

    records the count rate of particles emitted by the source during each measurement. In 1984, a boron -lined proportional counter reportedly served as...of only 6 Li and 127 I. This was based upon the MCNP4 input used by Mares and Schraube [29] and provides a set of isotopes with cross sections

  7. Implementation and qualification of MCNP 5 through the intercomparison with the benchmark for the calculation of critical systems Godiva and Jezebel

    International Nuclear Information System (INIS)

    Lara, Rafael G.; Maiorino, Jose R.

    2013-01-01

    This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others

  8. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  9. 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

  10. SÍNTESIS DE OXITOCINA EN FASE SÓLIDA USANDO DERIVADOS DE TERBUTOXICARBONILO Y FLUORENILMETOXICARBONILO

    Directory of Open Access Journals (Sweden)

    Julio C Calvo

    2010-08-01

    Full Text Available La oxitocina, péptido cíclico cuya secuencia es CYIQNCPLG, fué el primer péptido de importancia biológica que pudo ser sintetizado. En este trabajo se compara la síntesis de la oxitocina usando resina p metilbenzhidrilamina (MBHA para la síntesis por estrategia t-Boc y resina Rink p-metilbenzhidrilamina (Rink MBHA para la síntesis por estrategia Fmoc, con altos rendimientos. El péptido crudo se ciclizó en una disolución acuosa de dimetilsulfóxido al 10%. La caracterización se llevó a cabo por espectrometría de masas y resonancia magnética nuclear, y se logró detectar la presencia de dos isómeros.

  11. 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.)

  12. EGS4 and MCNP4b MC Simulation of a Siemens KD2 Accelerator in 6 MV Photon Mode

    CERN Document Server

    Chaves, A; Fragoso, M; Lopes, C; Oliveira, C; Peralta, L; Rodrigues, P; Seco, J; Trindade, A

    2001-01-01

    The geometry of a Siemens Mevatron KD2 linear accelerator in 6 MV photon mode was modeled with EGS4 and MCNP4b. Energy spectra and other phase space distributions have been extensively compared in different plans along the beam line. The differences found have been evaluated both qualitative and quantitatively. The final aim was that both codes, running in different operating systems and with a common set of simulation conditions, met the requirement of fitting the experimental depth dose curves and dose profiles, measured in water for different field sizes. Whereas depth dose calculations are in a certain extent insensible to some simulation parameters like electron nominal energy, dose profiles have revealed to be a much better indicator to appreciate that feature. Fine energy tuning has been tried and the best fit was obtained for a nominal electron energy of 6.15 MeV.

  13. 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.

  14. 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)

  15. MCNP simulation of the influence of the external moisture on low calorific value in the coal quality analysis by neutron

    International Nuclear Information System (INIS)

    Liu Dekun; Zhang Hongyu; Zhang Lihong; Dong Huan; Gu Deshan

    2012-01-01

    An important index in assessment of coal quality is low calorific value. Using neutron to analysis coal quality, the more the coal moisture content, especially the increasing of external moisture will reduce the low calorific value. The principle of coal quality analysis by neutron prompt Gamma-ray is introduced. The influence of the gamma count of the carbon element peak with increasing external moisture in coal samples was simulated using MCNP code. And discussed the reasons how external moisture content influence the calorific value. Simulation results indicate that with the increasing of external moisture in the coal samples, the gamma count of the carbon element peak dwindling, and the low calorific value reducing. The conclusion is : using neutrons method to analysis coal quality, the more external moisture content, the larger error of the measurement results of the carbon element, and will influence the calculation accuracy of the low calorific value. (authors)

  16. EJ2-MCNPlib. Contents of the JEF-2.2 based neutron cross-section library for MCNP4A

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Oppe, J.

    1995-05-01

    In this report a description is given of the EJ2-MCNPlib library. The EJ2-MCNPlib library is to be used for reactivity/critically calculations and general neutron/photon transport calculations with the Monte Carlo code MCNP4A. The library is based on the European JEF-2.2 nuclear data evaluation and contains data for all (i.e. 313) nuclides available on this evaluation.The cross-section data were generated using the NJOY cross-section processing code system, version 91.118. For easy reference cross-section plots are given in this report for the total, elastic and absorption cross sections for all nuclides on the EJ2-MCNPlib library. Furthermore, for verification purposes a graphical intercomparison is given of the results of standard benchmark calculations performed with JEF-2.2 cross-section data and with ENDF/B-V cross-section data (whenever available). 6 refs

  17. Computational model of Amersham I-125 source model 6711 and Prosper Pd-103 source model MED3633 using MCNP

    International Nuclear Information System (INIS)

    Menezes, Artur F.; Reis Junior, Juraci P.; Silva, Ademir X.; Facure, Alessandro; Cardoso, Simone C.

    2011-01-01

    Brachytherapy is used in cancer treatment at shorter distances through the use of small encapsulated source of ionizing radiation. In such treatment, a radiation source is positioned directly into or near the target volume to be treated. In this study the Monte Carlo based MCNP code was used to model and simulate the I-125 Amersham Health source model 6711 and the Pd-103 Prospera source model MED3633 in order to obtain the dosimetric parameter dose rate constant (Λ) . The sources geometries were modeled and implemented in MCNPX code. The dose rate constant is an important parameter prostate LDR brachytherapy's treatments planning. This study was based on American Association of Physicists in Medicine (AAPM) recommendations which were produced by its Task Group 43. The results obtained were 0.941 and 0.65 for the dose rate constants of I-125 and Pd-103 sources, respectively. They present good agreement with the literature values based on different Monte Carlo codes. (author)

  18. MCNP modelling of vaginal and uterine applicators used in intracavitary brachytherapy and comparison with radiochromic film measurements

    Science.gov (United States)

    Ceccolini, E.; Gerardy, I.; Ródenas, J.; van Dycke, M.; Gallardo, S.; Mostacci, D.

    Brachytherapy is an advanced cancer treatment that is minimally invasive, minimising radiation exposure to the surrounding healthy tissues. Microselectron© Nucletron devices with 192Ir source can be used for gynaecological brachytherapy, in patients with vaginal or uterine cancer. Measurements of isodose curves have been performed in a PMMA phantom and compared with Monte Carlo calculations and TPS (Plato software of Nucletron BPS 14.2) evaluation. The isodose measurements have been performed with radiochromic films (Gafchromic EBT©). The dose matrix has been obtained after digitalisation and use of a dose calibration curve obtained with a 6 MV photon beam provided by a medical linear accelerator. A comparison between the calculated and the measured matrix has been performed. The calculated dose matrix is obtained with a simulation using the MCNP5 Monte Carlo code (F4MESH tally).

  19. 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)

  20. 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.

  1. MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination

    International Nuclear Information System (INIS)

    Silva, Clarysson A.M. da; Pereira, Claubia; Guerra, Bruno T.; Veloso, Maria Auxiliadora F.; Costa, Antonella L.; Dalle, Hugo M.

    2011-01-01

    Highlights: ► Two models of IPR-R1 TRIGA using the MCNP5 code were simulated. ► It obtained k eff values in some different situations of the reactor operation. ► The first model analyzes the criticality and the neutronic flux over the reactor. ► The second model includes the radial and axial neutron flux evaluation with different operation conditions. ► The results present good agreement with respect to the experimental data. - Abstract: The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10 −4 . Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition.

  2. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors

    International Nuclear Information System (INIS)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M.; Reyes F, M. del C.; Del Valle G, E.

    2014-10-01

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  3. Comparación del Sexaje por PCR usando muestras de ADN de sangre y plumas de aves domésticas

    OpenAIRE

    Saldaña, Katherine; Facultad de Medicina Veterinaria y Zootecnia, Universidad Peruana Cayetano Heredia, Lima; Hung, Armando; Facultad de Medicina Veterinaria y Zootecnia, Universidad Peruana Cayetano Heredia, Lima-

    2016-01-01

    Objetivo: Comparar dos métodos de obtención de ADN usando muestras de sangre y de plumas de patos y gallinas. Metodología: Se utilizaron muestras de 13 aves: 10 patos (7 hembras, 2 macho; 1 sexo indefinido) y 3 pollos (2 hembras; 1 macho). Se extrajo ADN de sangre usando QIAamp ARN Mini Kit y para las muestras de plumas se utilizó el protocolo con proteinasa K. Se amplificó mediante PCR usando el protocolo y los primers P8 y P2. Resultados: Los resultados mostraron una mayor cantidad de ADN e...

  4. Usando comunicação serial em um experimento de controle de processos em tempo real

    Directory of Open Access Journals (Sweden)

    Ivo Neitzel

    2000-05-01

    Full Text Available O desenvolvimento e baixo custo dos microcomputadores introduziram o controle de processos em uma nova era. Atualmente, muitos controladores on-off e analógicos estão sendo substituídos por programas computacionais (softwares. Neste artigo, descreveremos um experimento simples e didático visando ao controle de temperatura de um equipamento (uma mufla, usando dois microcomputadores conectados por suas portas seriais. As principais características deste experimento são os fatos de introduzir o aluno no lado prático do controle de processos, usando um sistema distribuído de controle digital (SDCD, e permitir tanto a construção do controlador quanto a interação (aplicação distúrbios com processo.

  5. Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2012-01-01

    Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

  6. 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

  7. Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

    Directory of Open Access Journals (Sweden)

    Noble Brooklyn

    2012-01-01

    Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

  8. 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)

  9. Radiation field characterization of a BNCT research facility using Monte Carlo Method - Code MCNP-4B; Caracterizacao do campo de radiacao numa instalacao para pesquisa em BNCT o metodo de Monte Carlo Codigo - MCNP-4B

    Energy Technology Data Exchange (ETDEWEB)

    Hernandes, Antonio Carlos

    2002-07-01

    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 {Nu}{sub {Tau}} = 1,35x10{sup 8} n/cm{sup 2}, a fast neutron dose of 5,86x{sup -1}0 Gy/{Nu}{sub {Tau}} and a gamma ray dose of 8,30x{sup -14} Gy/{Nu}{sub {Tau}}. (author)

  10. Radiation field characterization of a BNCT research facility using Monte Carlo method - code MCNP-4B; Caracterizacao do campo de radiacao numa instalacao para pesquisa em BNCT utilizando o metodo de Monte Carlo - codigo MCNP-4B

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, Antonio Carlos

    2002-07-01

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

  11. Response function of an HPGe detector simulated through MCNP 4A varying the density and chemical composition of the matrix; Funcion respuesta de un detector HPGe simulada mediante MCNP 4A variando la densidad y composicion quimica de la matriz

    Energy Technology Data Exchange (ETDEWEB)

    Leal A, B.; Mireles G, F.; Quirino T, L.; Pinedo, J.L. [Universidad Autonoma de Zacatecas, Zacatecas (Mexico)]. e-mail: bleal79@yahoo.com.mx

    2005-07-01

    In the area of the Radiological Safety it is required of a calibrated detection system in energy and efficiency for the determination of the concentration in activity in samples that vary in chemical composition and by this in density. The area of Nuclear Engineering requires to find the grade of isotopic enrichment of the uranium of the Sub-critic Nuclear Chicago 9000 Mark. 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 range of energy of 80 keV to 1400 keV varying the density of the matrix and the chemical composition by means of the application of the Monte Carlo code MCNP-4A. The obtained results in the simulation of the response function of the detector show a grade of acceptance in the range from 500 to 1400 keV energy, with a smaller percentage discrepancy to 10%, in the range of low energy that its go from 59 to 400 keV, the percentage discrepancy varies from 17% until 30%, which is manifested in the opposing isotopic relationship for 5 fuel rods of the Sub critic nuclear assemble. (Author)

  12. Evaluación de la reducibilidad de un mineral de hierro usando char como reductor

    Directory of Open Access Journals (Sweden)

    Yenny Rubiela Hernández, Carlos Alberto Sandoval Fonseca, Claudia Inés Sánchez Buitrago

    2011-05-01

    Full Text Available Muestra los ensayos de  reduciblidad  realizados en un hornotipo Linder a un mineral de hierro del municipio de Ubalá(departamento de Cundinamarca, Colombia, usando comoreductor un char. Se  indican las características del mineralde  hierro  de Ubalá, de  los  carbones  empleados para  laproducción del char y de la caliza, así como los ensayos dereducibilidad. Para la caracterización de  las materias primasy del char, como producto  final, se aplicaron normas ASTM.En  la producción de  los char se utilizaron  los hornos decoquización  tipo Cerchar  y  tipo  colmena  de  la Uptc  enSamacá  (Boyacá. Los ensayos de reducibilidad se hicieronbajo  los mismos parámetros de operación utilizados concarbón como reductor, y los resultados obtenidos dejan verque el mineral de hierro de Ubalá es reducible en menorporcentaje con char. Sin embargo, por  los grandes beneficiospara el medioambiente que se obtienen trabajando con elchar, no se descarta  la posibilidad de utilizarlo como posiblesustituto del carbón en el proceso de reducción directa.

  13. Variantes del problema del cartero mixto que se pueden resolver usando programación lineal

    Directory of Open Access Journals (Sweden)

    Francisco Javier Zaragoza Martínez

    2012-07-01

    Full Text Available Dada una gráfica mixta y conexa con costos en sus aristas y arcos, el problema del cartero mixto consiste en encontrar un circuito cerrado de la gráfica mixta que recorra sus aristas y arcos a costo mínimo. Se sabe que este problema es NP-duro. Sin embargo, bajo ciertas condiciones adicionales, el problema se puede resolver en tiempo polinomial usando programación lineal, en otras palabras, los poliedros correspondientes son enteros. Algunas de estas condiciones son: la gráfica mixta es serie paralelo o la gráfica mixta tiene grado total par en todos sus vértices. Además, mostramos que si agregamos la restricción adicional de que cada arista se recorra exactamente una vez entonces el problema se puede resolver en tiempo polinomial si el conjunto de arcos forma un bosque. Palabras clave: Gráfica mixta, problema de cartero, programación lineal. Mathematics Subject Classification: 05C45, 90C35.

  14. Neutron-photon energy deposition in CANDU reactor fuel channels: a comparison of modelling techniques using ANISN and MCNP computer codes

    International Nuclear Information System (INIS)

    Bilanovic, Z.; McCracken, D.R.

    1994-12-01

    In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. (author). 9 refs., 12 tabs., 20 figs

  15. Modeling the irradiation facility in the Deir Al-Hajar area to calculate the spatial gamma dose distribution using the MCNP code

    International Nuclear Information System (INIS)

    Khattab, K.; Bush, M; Kassery, H.

    2009-03-01

    A 3-D model for the irradiation plant which belongs to the Atomic Energy Commission, Department of Radiation Technology in the Deir Al-Hajar area near Damascus, is presented in this work using the MCNP-4C code. This model is used to calculate the spatial gamma ray dose in the (x, y, z) coordinate. Good agreements are noticed between the measured and the calculated results. (author)

  16. 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)

  17. 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

    International Nuclear Information System (INIS)

    Fernandes, Marco A.R.

    2011-01-01

    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)

  18. Comparison of results from the MCNP criticality validation suite using ENDF/B-VI and preliminary ENDF/B-VII nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R. D. (Russell D.)

    2004-01-01

    The MCNP Criticality Validation Suite is a collection of 31 benchmarks taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. MCNP5 calculations clearly demonstrate that, overall, nuclear data for a preliminary version of ENDFB-VII produce better agreement with the benchmarks in the suite than do corresponding data from ENDF/B-VI. Additional calculations identify areas where improvements in the data still are needed. Based on results for the MCNP Criticality Validation Suite, the Pre-ENDF/B-VII nuclear data produce substantially better overall results than do their ENDF/B-VI counterparts. The calculated values for k{sub eff} for bare metal spheres and for an IEU cylinder reflected by normal uranium are in much better agreement with the benchmark values. In addition, the values of k{sub eff} for the bare metal spheres are much more consistent with those for corresponding metal spheres reflected by normal uranium or water. In addition, a long-standing controversy about the need for an ad hoc adjustment to the {sup 238}U resonance integral for thermal systems may finally be resolved. On the other hand, improvements still are needed in a number of areas. Those areas include intermediate-energy cross sections for {sup 235}U, angular distributions for elastic scattering in deuterium, and fast cross sections for {sup 237}Np.

  19. Solution of large underestimation problem in the Monte Carlo calculation with hard biasing. In case with geometry input data created by CAD/MCNP automatic converter

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Konno, Chikara; Sato, Satoshi; Kawasaki, Nobuo; Seki, Akiyuki

    2008-04-01

    An inconvenient experience was encountered, in which we have different answers depending on applied weight window values, in the nuclear analysis of the benchmark problem for CAD/MCNP interface programs, being developed under the ITER R and D task. Biasing can enhance calculation speed, but should not give different answers. Mechanism of this large underestimation is clarified. It is caused by the combination of the following two facts; When one of particles in a history has got lost, MCNP cancels all tallies calculated during the history and all banked particles are thrown away (never tracked). When we have distributed micro geometry errors in input data, important histories, which give significant contribution to tallies, will have many splitting and have 'lost particle' with higher probability in the case of hard biasing. These two facts lead to selective canceling of important histories. An attempt to eliminate this inconvenience has been made, by modifying the subroutine 'hstory' of MCNP. The modification has been done very successfully and eliminated the large underestimation, giving the same answer independently from applied weight window values. (author)

  20. Validation of MCNP6.1 for Criticality Safety of Pu-Metal, -Solution, and -Oxide Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kiedrowski, Brian C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favorite, Jeffrey A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kahler, III, Albert C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kersting, Alyssa R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Walker, Jessie L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-05-13

    Guidance is offered to the Los Alamos National Laboratory Nuclear Criticality Safety division towards developing an Upper Subcritical Limit (USL) for MCNP6.1 calculations with ENDF/B-VII.1 nuclear data for three classes of problems: Pu-metal, -solution, and -oxide systems. A benchmark suite containing 1,086 benchmarks is prepared, and a sensitivity/uncertainty (S/U) method with a generalized linear least squares (GLLS) data adjustment is used to reject outliers, bringing the total to 959 usable benchmarks. For each class of problem, S/U methods are used to select relevant experimental benchmarks, and the calculational margin is computed using extreme value theory. A portion of the margin of sub criticality is defined considering both a detection limit for errors in codes and data and uncertainty/variability in the nuclear data library. The latter employs S/U methods with a GLLS data adjustment to find representative nuclear data covariances constrained by integral experiments, which are then used to compute uncertainties in keff from nuclear data. The USLs for the classes of problems are as follows: Pu metal, 0.980; Pu solutions, 0.973; dry Pu oxides, 0.978; dilute Pu oxide-water mixes, 0.970; and intermediate-spectrum Pu oxide-water mixes, 0.953.

  1. Gamma Knife Simulation Using the MCNP4C Code and the Zubal Phantom and Comparison with Experimental Data

    Directory of Open Access Journals (Sweden)

    Somayeh Gholami

    2010-06-01

    Full Text Available Introduction: Gamma Knife is an instrument specially designed for treating brain disorders. In Gamma Knife, there are 201 narrow beams of cobalt-60 sources that intersect at an isocenter point to treat brain tumors. The tumor is placed at the isocenter and is treated by the emitted gamma rays. Therefore, there is a high dose at this point and a low dose is delivered to the normal tissue surrounding the tumor. Material and Method: In the current work, the MCNP simulation code was used to simulate the Gamma Knife. The calculated values were compared to the experimental ones and previous works. Dose distribution was compared for different collimators in a water phantom and the Zubal brain-equivalent phantom. The dose profiles were obtained along the x, y and z axes. Result: The evaluation of the developed code was performed using experimental data and we found a good agreement between our simulation and experimental data. Discussion: Our results showed that the skull bone has a high contribution to both scatter and absorbed dose. In other words, inserting the exact material of brain and other organs of the head in digital phantom improves the quality of treatment planning. This work is regarding the measurement of absorbed dose and improving the treatment planning procedure in Gamma-Knife radiosurgery in the brain.

  2. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  3. Inter-comparison of Dose Distributions Calculated by FLUKA, GEANT4, MCNP, and PHITS for Proton Therapy

    Science.gov (United States)

    Yang, Zi-Yi; Tsai, Pi-En; Lee, Shao-Chun; Liu, Yen-Chiang; Chen, Chin-Cheng; Sato, Tatsuhiko; Sheu, Rong-Jiun

    2017-09-01

    The dose distributions from proton pencil beam scanning were calculated by FLUKA, GEANT4, MCNP, and PHITS, in order to investigate their applicability in proton radiotherapy. The first studied case was the integrated depth dose curves (IDDCs), respectively from a 100 and a 226-MeV proton pencil beam impinging a water phantom. The calculated IDDCs agree with each other as long as each code employs 75 eV for the ionization potential of water. The second case considered a similar condition of the first case but with proton energies in a Gaussian distribution. The comparison to the measurement indicates the inter-code differences might not only due to different stopping power but also the nuclear physics models. How the physics parameter setting affect the computation time was also discussed. In the third case, the applicability of each code for pencil beam scanning was confirmed by delivering a uniform volumetric dose distribution based on the treatment plan, and the results showed general agreement between each codes, the treatment plan, and the measurement, except that some deviations were found in the penumbra region. This study has demonstrated that the selected codes are all capable of performing dose calculations for therapeutic scanning proton beams with proper physics settings.

  4. Creation and testing of an ENDF/B-VI neutron data library (ENDF60) for use with MCNP trademark

    International Nuclear Information System (INIS)

    Frankle, S.C.; MacFarlane, R.E.

    1995-01-01

    The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N Particle radiation transport code MCNP4A, was released in the fall of 1994. The ENDF60 library is comprised of 124 nuclide data files based on the ENDF/B-VI evaluations through Release 2. Fifty-two percent of these ENDF/B-VI evaluations are translations from ENDF/B-V. The remaining forty-eight percent are new evaluations which have sometimes changed significantly. The new evaluations include important materials for criticality safety calculations, as well as significant enhancements such as isotopic evaluations, better resonance-range representations, and the new correlated energy-angle distributions for emitted particles. In particular, the upper energy limit for the resolved resonance region of 235 U, 238 U and 239 Pu has been extended from 0.082, 4.0, and 0.301 keV to 2.25, 10.0, and 2.5 keV respectively. As part of the overall quality assurance testing of the ENDF60 library, calculations for well known benchmark assemblies were performed. This benchmarking effort included revising the standard nine criticality benchmarks documented in previous Los Alamos National Laboratory Reports, LA-12212 and LA-12891, as well as the implementation of new Cross Section Evaluation Working Group (CSEWG) benchmarks. Comparisons of benchmark results for different data libraries can aid the user in understanding how well an evaluation performs for their application

  5. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  6. Validation of MCNP NPP Activation Simulations for Decommissioning Studies by Analysis of NPP Neutron Activation Foil Measurement Campaigns

    Directory of Open Access Journals (Sweden)

    Volmert Ben

    2016-01-01

    Full Text Available In this paper, an overview of the Swiss Nuclear Power Plant (NPP activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

  7. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A.

    2015-01-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K eff at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  8. MCNP6 unstructured mesh application to estimate the photoneutron distribution and induced activity inside a linac bunker

    Science.gov (United States)

    Juste, B.; Morató, S.; Miró, R.; Verdú, G.; Díez, S.

    2017-08-01

    Unwanted neutrons in radiation therapy treatments are typically generated by photonuclear reactions. High-energy beams emitted by medical Linear Accelerators (LinAcs) interact with high atomic number materials situated in the accelerator head and release neutrons. Since neutrons have a high relative biological effectiveness, even low neutron doses may imply significant exposure of patients. It is also important to study radioactivity induced by these photoneutrons when interacting with the different materials and components of the treatment head facility and the shielding room walls, since persons not present during irradiation (e.g. medical staff) may be exposed to them even when the accelerator is not operating. These problems are studied in this work in order to contribute to challenge the radiation protection in these treatment locations. The work has been performed by simulation using the latest state of the art of Monte-Carlo computer code MCNP6. To that, a detailed model of particles transport inside the bunker and treatment head has been carried out using a meshed geometry model. The LinAc studied is an Elekta Precise accelerator with a treatment photon energy of 15 MeV used at the Hospital Clinic Universitari de Valencia, Spain.

  9. Inter-comparison of Dose Distributions Calculated by FLUKA, GEANT4, MCNP, and PHITS for Proton Therapy

    Directory of Open Access Journals (Sweden)

    Yang Zi-Yi

    2017-01-01

    Full Text Available The dose distributions from proton pencil beam scanning were calculated by FLUKA, GEANT4, MCNP, and PHITS, in order to investigate their applicability in proton radiotherapy. The first studied case was the integrated depth dose curves (IDDCs, respectively from a 100 and a 226-MeV proton pencil beam impinging a water phantom. The calculated IDDCs agree with each other as long as each code employs 75 eV for the ionization potential of water. The second case considered a similar condition of the first case but with proton energies in a Gaussian distribution. The comparison to the measurement indicates the inter-code differences might not only due to different stopping power but also the nuclear physics models. How the physics parameter setting affect the computation time was also discussed. In the third case, the applicability of each code for pencil beam scanning was confirmed by delivering a uniform volumetric dose distribution based on the treatment plan, and the results showed general agreement between each codes, the treatment plan, and the measurement, except that some deviations were found in the penumbra region. This study has demonstrated that the selected codes are all capable of performing dose calculations for therapeutic scanning proton beams with proper physics settings.

  10. Neutronics modeling of TRIGA reactor at the University of Utah using agent, KENO6 and MCNP5 codes

    International Nuclear Information System (INIS)

    Yang, X.; Xiao, S.; Choe, D.; Jevremovic, T.

    2010-01-01

    The TRIGA reactor at the University of Utah is modelled in 2D using the AGENT state-of-the-art methodology based on the Method of Characteristics (MOC) and R-function theory supporting detailed reactor analysis of reactor geometries of any type. The TRIGA reactor is also modelled using KENO6 and MCNP5 for comparison. The spatial flux and reaction rates distribution are visualized by AGENT graphics support. All methodologies are in use in to study the effect of different fuel configurations in developing practical educational exercises for students studying reactor physics. At the University of Utah we train graduate and undergraduate students in obtaining the Nuclear Regulatory Commission license in operating the TRIGA reactor. The computational models as developed are in support of these extensive training classes and in helping students visualize the reactor core characteristics in regard to neutron transport under various operational conditions. Additionally, the TRIGA reactor is under the consideration for power uprate; this fleet of computational tools once benchmarked against real measurements will provide us with validated 3D simulation models for simulating operating conditions of TRIGA. (author)

  11. Computational model of Amersham I-125 source model 6711 and Prosper Pd-103 source model MED3633 using MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Menezes, Artur F.; Reis Junior, Juraci P.; Silva, Ademir X., E-mail: ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Rosa, Luiz A.R. da, E-mail: lrosa@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Facure, Alessandro [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Cardoso, Simone C., E-mail: Simone@if.ufrj.b [Universidade Federal do Rio de Janeiro (IF/UFRJ), RJ (Brazil). Inst. de Fisica. Dept. de Fisica Nuclear

    2011-07-01

    Brachytherapy is used in cancer treatment at shorter distances through the use of small encapsulated source of ionizing radiation. In such treatment, a radiation source is positioned directly into or near the target volume to be treated. In this study the Monte Carlo based MCNP code was used to model and simulate the I-125 Amersham Health source model 6711 and the Pd-103 Prospera source model MED3633 in order to obtain the dosimetric parameter dose rate constant ({Lambda}) . The sources geometries were modeled and implemented in MCNPX code. The dose rate constant is an important parameter prostate LDR brachytherapy's treatments planning. This study was based on American Association of Physicists in Medicine (AAPM) recommendations which were produced by its Task Group 43. The results obtained were 0.941 and 0.65 for the dose rate constants of I-125 and Pd-103 sources, respectively. They present good agreement with the literature values based on different Monte Carlo codes. (author)

  12. 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)

  13. Calculation of ex-core detector weighting functions for a sodium-cooled tru burner mockup using MCNP5

    International Nuclear Information System (INIS)

    Pham Nhu Viet Ha; Min Jae Lee; Sunghwan Yun; Sang Ji Kim

    2015-01-01

    Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A. (author)

  14. Response function of an HPGe detector simulated through MCNP 4A varying the density and chemical composition of the matrix

    International Nuclear Information System (INIS)

    Leal A, B.; Mireles G, F.; Quirino T, L.; Pinedo, J.L.

    2005-01-01

    In the area of the Radiological Safety it is required of a calibrated detection system in energy and efficiency for the determination of the concentration in activity in samples that vary in chemical composition and by this in density. The area of Nuclear Engineering requires to find the grade of isotopic enrichment of the uranium of the Sub-critic Nuclear Chicago 9000 Mark. 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 range of energy of 80 keV to 1400 keV varying the density of the matrix and the chemical composition by means of the application of the Monte Carlo code MCNP-4A. The obtained results in the simulation of the response function of the detector show a grade of acceptance in the range from 500 to 1400 keV energy, with a smaller percentage discrepancy to 10%, in the range of low energy that its go from 59 to 400 keV, the percentage discrepancy varies from 17% until 30%, which is manifested in the opposing isotopic relationship for 5 fuel rods of the Sub critic nuclear assemble. (Author)

  15. 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.

  16. Preliminary MCNP-POLIMI Simulations for the Evaluation of the ''Floor Effect'' Comparison of APSTNG and Cf Sources

    CERN Document Server

    Pozzi, S A

    2002-01-01

    The present simulations performed with the Monte Carlo code MCNP-POLIMI [1] have the scope of evaluating the associated-particle sealed tube neutron generator (APSTNG) for use as an interrogation source in the source-driven noise analysis method for the assay of nuclear materials. In the Nuclear Materials Identification System (NMIS) developed at the Oak Ridge National Laboratory, the time dependent cross-correlation of the timed neutron source and detector responses is one of the signatures acquired. Previous studies and measurements have demonstrated the sensitivity of this and other related signatures to fissile mass [2-3]. In a recent report [4], we outlined the advantages of the APSTNG interrogation source for use with NMIS when compared with the Cf-252 source. In particular, we showed that when the distance between the source and the sample and the sample and the detectors is large, the APSTNG source outperforms the Cf-252 in sensitivity to fissile mass. This is the case when performing measurements of ...

  17. Characteristics of the quarry as shielding for {sup 241}AmBe neutrons and monoenergetic photons; Caracteristicas de la cantera como blindaje para los neutrones del {sup 241}AmBe y fotones monoenergeticos

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C.; Salas L, M. A. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Rodriguez R, J. A.; Juarez A, C. A., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Nuevo Leon, Facultad de Ingenieria Civil, Pedro de Alba s/n, San Nicolas de los Garza, Nuevo Leon (Mexico)

    2016-09-15

    Shielding is an important element in radiation protection since allows the management of radiation sources. Currently there are different materials of natural or anthropogenic origin that are used as shielding for both photons and neutrons. The quarry is a material of natural origin and abundant in our country, which is used in construction or for the manufacture of sculptures, however its characteristics as shielding have not been reported. In this paper we report some of the properties of the quarry as shielding for monoenergetic photons and for neutrons produced by an isotopic neutron source of {sup 241}AmBe. A quarry piece was used to determine its density and its chemical composition, with the XCOM code the elemental composition was determined and the mass interaction and total attenuation coefficients of the quarry were determined with photons of 10{sup -3} to 10{sup -5} MeV; the interaction coefficients included coherent dispersion, photoelectric absorption, Compton dispersion and the production of pairs in the nuclear and electronic field. Using the MCNP5 code, a narrow geometry attenuation experiment was modeled and the photon fluence was estimated that reaches a point detector at a distance of 42 cm from a point source, isotropic and monoenergetic photon when the source and the point detector were added quarry pieces of different thicknesses. The reduction of the number of photons as a function of the thickness of the quarry was used to determine the coefficient of linear attenuation of the quarry before photons of 0.03, 0.07, 0.1, 0.3, 1, 2 and 3 MeV that were the same as those calculated with the XCOM code. With the MCNP, the K a and H(10) transmission curves were also calculated. This same model was used to determined the variation of the {sup 241}AmBe neutron spectrum as a function of quarry thickness, as well as the E{sub ROT} and H(10) transmission curves. (Author)

  18. Estudio de superficies usando un microscopio de efecto túnel (STM

    Directory of Open Access Journals (Sweden)

    Alba Graciela Ávila Bernal

    2009-09-01

    Full Text Available Los microscopios de barrido se han convertido en las manos y los “ojos” de experimentadores de nuestro siglo, son herramien- tas necesarias en los laboratorios de educación e investigación para la caracterización a nanoescalas. El presente artículo pre- senta las modificaciones en la implementación electrónica (caracterización de los piezoeléctricos y sistema de barrido y mecáni- ca (diseño de un sistema de antivibración de un microscopio de barrido de efecto túnel que han permitido visualización y modi- ficación de superficies a nanoescala. Se describe una metodología para la correcta visualización y caracterización de superficies usando el instrumento implementado, alcanzando la cuantificación bidimensional de características de hasta 1300nm2, con re- solución ~15nm. Esta metodología, determinada experimentalmente, tiene en cuenta parámetros críticos para la estabilización de la corriente túnel, como lo son la velocidad de barrido y las geometrías y dimensiones de las agujas del microscopio. La ver- satilidad del microscopio permite modificar y visualizar los defectos introducidos en muestras de HOPG al aplicar voltajes entre la punta del microscopio y la muestra. Los resultados aquí descritos permiten presentar fácilmente los conceptos de barrido to- pográfico y litografía.

  19. Pirólise catalítica do PEBD usando como catalisador a vermiculita modificada

    Directory of Open Access Journals (Sweden)

    Franciel Aureliano Bezerra

    2016-01-01

    Full Text Available Resumo O polietileno de baixa densidade (PEBD é um dos polímeros mais usados atualmente, e a grande quantidade desse polímero produzida resulta em toneladas de resíduos, que necessitam ser tratados. Neste trabalho foi realizada a pirólise termocatalítica do PEBD usando como catalisador a argila vermiculita modificada, como alternativa para o tratamento dos resíduos. A argila foi tratada com solução de ácido nítrico a diferentes concentrações e calcinada a 400 °C. Os materiais foram caracterizados por técnicas de difratometria de raios X, termogravimetria, adsorção de nitrogênio e espectroscopia de energia dispersiva. A pirólise térmica e termocatalítica foi realizada em um micro reator acoplado com GC/MS, a 500 °C. O intuito da pirólise de resíduos poliméricos é a obtenção de hidrocarbonetos leves (C<16, que possam ser empregados na indústria química e petroquímica, através de quebras na cadeia polimérica. Os resultados foram satisfatórios, com aumento no rendimento para hidrocarbonetos leves ao empregar os catalisadores chegando a 71,4% de produtos com C<16, enquanto a pirólise térmica resultou apenas de 25,8%.

  20. AUTOMATIZACIÓN DE LA ARQUITECTURA DE COMPONENTES GENÉRICOS USANDO UML

    Directory of Open Access Journals (Sweden)

    Noel Fuentes Ramírez

    2006-04-01

    Full Text Available

     

    La arquitectura de componentes genéricos permite chequear la consistencia interna de sus elementos arquitectónicos (componentes y conectores a partir de las relaciones internas en sus respectivas estructuras, que pueden ser de inclusión y transformación. También permite verificar la conexión entre componentes y conectores a partir de las relaciones de transformación entre sus interfaces respectivas. Las ideas que se presentan aquí constituyen una propuesta para la automatización de la descripción de esta arquitectura usando el lenguaje de modelado unificado (UML a partir de la descripción formal de sus diagramas de clases y de secuencia, así como para el chequeo de la consistencia. En este artículo se muestra la aplicación de esta propuesta mediante una extensión de la herramienta Visual Paradigm, por medio de un módulo de software conectable.

  1. Un estudio sobre el desarrollo del pensamiento aleatorio usando recursos educativos abiertos

    Directory of Open Access Journals (Sweden)

    Yenny Patricia Pinzón Triana

    2015-04-01

    Full Text Available Este estudio presenta los resultados de un proyecto de investigación sobre la implementación de la enseñanza probabilística con recursos educativos abiertos (REA, diseñados en la plataforma Edmodo, y que está dirigido a estudiantes de tercer grado de educación básica secundaria de Bogotá, Colombia, y Tuxtepec, México. El fin era evidenciar sus preconceptos, nociones y evaluar el resultado de la instrucción en términos de su pensamiento probabilístico y dar respuesta a la interrogante ¿cuál es el efecto de la instrucción en probabilidad usando Edmodo en estudiantes de tercer grado de educación básica secundaria respecto a la valoración de fenómenos aleatorios de la vida cotidiana, sus conjeturas y la toma de decisiones? Se empleó un enfoque cualitativo a partir del método de estudio de casos, desde el análisis particular a lo general (Stake, 2005. Para la implementación, se consideró el estudio de Fishbein (1975 sobre el desarrollo del pensamiento probabilístico y el de Marzano (2000, relacionado con las dimensiones del aprendizaje. Estos elementos sirvieron para establecer el impacto del uso de los REA, en especial el de la plataforma Edmodo en el desarrollo de competencias en los estudiantes. Los instrumentos consistieron en una prueba de entrada, actividades de apoyo interactivo desde Edmodo y una prueba de salida para determinar los niveles de aprendizaje; mediante triangulación de datos, se evidenciaron alcances de niveles adecuados de desempeño acordes con los requerimientos de los estándares nacionales e internacionales.

  2. Desarrollo de un horno solar para el secado de plantas y vegetales usando control difuso

    Directory of Open Access Journals (Sweden)

    Oscar G. Ibarra Manzano

    2012-05-01

    Full Text Available Recientemente, el aprovechamiento de la energía solar en el deshidratado de productos agrícolas se ha vuelto cada día más común debido a los altos rendimientos en los productos post-cosecha. La inversión en tecnologías propias para contribuir con los productores del sector agroalimentario es un factor importante para el desarrollo de las cadenas produc­tivas de nuestro país. En este trabajo se presenta el desarrollo de un horno solar para el secado de plantas y vegetales utilizando control difuso. Este es un sistema térmicamente controlado que permite disminuir el tiempo de secado de varios días a unas horas. Se reali­zaron pruebas de secado usando flor de jamaica, en las cuales se pudo disminuir el tiempo de secado de cuatro días a aproximadamente 5 h. Se presentan tanto la parte de diseño conceptual, como resultados experimentales del mismo. Los resultados obtenidos permiten ver la viabilidad del diseño propuesto.Recently, the use of solar energy in the dehydration of agricultural products is becomingmore common as high yields in the post-harvest products. Investment in technologies forcontributing to the producers of food products is an important factor for the development ofthe productive chains of our country. This paper presents the development of a solar oven fordrying plants and vegetables using fuzzy control. This is a heat-controlled system that allowsdecreasing the drying time from several days to hours. Drying tests were conducted usingjamaica flower, which could decrease the drying time from four days to about 5 h. We presentboth the conceptual design of the experimental results. The results obtained allow us to seethe feasibility of the proposed design.

  3. Evaluation of the criticality of a concrete container for storage of spent fuel in dry with MCNP; Evaluacion de la criticidad de un contenedor de concreto para almacenamiento de combustible gastado en seco con MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Ramirez S, J. R., E-mail: vicente.xolocostli@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    A main concern exists inside the nuclear power plants in operation around the world that is the with respect to the storage capacity of the spent fuel, due to the useful life of the plant and the storage capacity in the spent fuel pool. In diverse countries is believed that one of the best alternatives for the spent fuel is the reprocessing of the same one since exists a great quantity of fissile material that can be profitable as the Pu-239, but even so the costs for the reprocessing continue being high, what limits taking this process to great scale. Is for that reason the importance of the containers for storage of spent fuel in dry which has had a great apogee in the last years, since they represent an alternative to store the spent fuel before making a decision on the reprocessing of the same one or the final disposal. In this work an evaluation of the criticality of a concrete container for storage of spent fuel in dry commercially available is made, and which is useful for fuel assemblies type PWR like BWR, in our case only the type BWR is considered. For the analysis of the evaluation was used the code MCNP5, considering the characteristics of the concrete container according to the available data, although the type of fuel assembly is BWR one of the models of the ABB company was considered with which the comparative of the results is made. The made calculations were carried out considering the inundation of the gap that exist and the external cavity, being this the most extreme condition to arrive to the criticality or in the case of happening an accident to have the filtration of the water toward the space of the gap. (author)

  4. Quantitative comparison between PGNAA measurements and MCNP calculations in view of the characterization of radioactive wastes in Germany and France

    Energy Technology Data Exchange (ETDEWEB)

    Mauerhofer, E. [FZJ, Institute for Energy and Climate Research - Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Strasse, D-52428 Juelich (Germany); Havenith, A.; Kettler, J. [RWTH Aachen University, Institute of Nuclear Fuel Cycle, Elisabethstrasse 16, D-52062 Aachen (Germany); Carasco, C.; Payan, E.; Ma, J. L.; Perot, B. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France)

    2013-04-19

    The Forschungszentrum Juelich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.

  5. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  6. Study of geometry to obtain the volume fraction of multiphase flows using the MCNP-X code

    International Nuclear Information System (INIS)

    Peixoto, Philippe N.B.; Salgado, Cesar M.

    2015-01-01

    The gamma ray attenuation technique is used in many works to obtaining volume fraction of multiphase flows in the oil industry, because it is a noninvasive technique with good precision. In these studies are simulated various geometries with different flow regime, compositions of materials, source-detector positions and types of collimation for sources. This work aim evaluate the interference in the results of the geometry changes and obtaining the best measuring geometry to provide the volume fractions accurately by evaluating different geometries simulations (ranging the source-detector position, flow schemes and homogeneity Makeup) in the MCNP-X code. The study was performed for two types of biphasic compositions of materials (oil-water and oil-air), two flow regimes (annular and smooth stratified) and was varied the position of each material in relative to source and detector positions. Another study to evaluate the interference of homogeneity of the compositions in the results was also conducted in order to verify the possibility of removing part of the composition and make a homogeneous blend using a mixer equipment. All these variations were simulated with two different types of beam, divergent beam and pencil beam. From the simulated geometries, it was possible to compare the differences between the areas of the spectra generated for each model. The results indicate that the flow regime and the differences in the material's densities interfere in the results being necessary to establish a specific simulation geometry for each flows regime. However, the simulations indicate that changing the type of collimation of sources do not affect the results, but improving the counts statistics, increasing the accurate. (author)

  7. MONITOREO PRELIMINAR DE INCIDENCIA DE FISIOPATÍAS EN CULTIVOS DE FRESA USANDO PROCESAMIENTO DIGITAL DE IMÁGENES

    Directory of Open Access Journals (Sweden)

    JUAN DAVID SANDINO-MORA

    Full Text Available La identificación de diferentes anomalías en cultivos agrícolas usando procesamiento de imágenes, ha demostrado cada vez más su efectividad, contrario con los métodos de ejecución tradicionales, los cuales arrancan los folíolos y frutos de la planta, para realizar el estudio. En este trabajo se presentan los resultados del desarrollo e validación de un algoritmo, que permita realizar monitoreo de incidencia en cultivos de fresa (Fragaria x ananassa, capaz de dar una primera aproximación para distinguir senescencia y daños mecánicos en sus foliolos, implementando una metodología indirecta (no destructiva. Las técnicas de procesamiento de imágenes implementadas incluyen Suavizado, Erosión, Dilatación, Detección de Contornos, Correspondencia de Patrones, Umbralización, entre otros. Los resultados obtenidos se visualizaron en una aplicación desarrollada en C# usando la librería Emgu CV, mostrando al usuario un diagnóstico de la planta de estudio. Se concluye que es posible ofrecer un servicio de monitoreo preliminar de incidencia usando este algoritmo, ahorrando tiempo para productores e investigadores que requieran una primera aproximación del estado del cultivo, con la posibilidad de ejecutarse tanto en computadores e robots aéreos (drones para hacer más eficiente esta tarea.

  8. DIGESTÃO DE CARBOIDRATOS USANDO DIGESTIVOS ENZIMÁTICOS COMERCIAIS – UMA AULA PRÁTICA.

    OpenAIRE

    Clerici, Maria Teresa Pedrosa Silva; Silva, Roberval Serafim da; Alves, Armindo Antonio

    2006-01-01

    Este trabalho mostra uma aula prática na qual se estuda a digestão decarboidratos “in vitro” utilizando amido gelatinizado como substrato e doismedicamentos comerciais indicados para problemas digestórios, como fonte deamilase. As reações enzimáticas foram feitas no equipamento dissolutor a 37º C, comagitação constante e usando meios que simulavam o pH do estômago e em seguida opH do intestino delgado. Para avaliar a eficácia das enzimas pancreáticas nosdigestivos, foi feita uma reação com am...

  9. Análisis de la producción de Polihidroxibutirato usando lactosuero como materia prima

    OpenAIRE

    Alvarez Campuzano, Catalina

    2015-01-01

    Este proyecto presenta el análisis de la producción de polihidroxibutirato (PHB) usando como materias primas el glicerol, lactosa y lactosuero. El glicerol y el lactosuero son residuos del sector agroindustrial colombiano, que se obtienen como resultado de la producción del biodiesel y del queso, respectivamente. La lactosa es un sustrato puro que se separa del lactosuero y luego se somete a un proceso de purificación. Inicialmente se presentó la aplicación del glicerol, que permite el contro...

  10. Conservación de Pan Artesanal Ezequiel y Pan Superbueno Usando Aceite Esencial de Clavo de Olor (Eugenia caryophillus)

    OpenAIRE

    Pilco Quesada, Silvia; Universidad Peruana Unión; Quito Vidal, Moisés; Universidad Peruana Unión; Quispe Condori, Sócrates; Universidad Peruana Unión

    2015-01-01

    El presente trabajo tuvo por objetivo evaluar la vida en anaquel del pan artesanal Ezequiel (CITAL) y pan de molde comercial Superbueno (Productos Unión), usando aceite esencial del clavo de olor (Eu- genia Caryophillus) como antimoho. El aceite esencial fue obtenido por el método de hidrodestilación, con un rendimiento de 15 % (m/m). Del análisis microbiológico del pan Superbueno (pan comercial) se identificaron una cepa de levadura y dos mohos que corresponden al género Penicillium. Del pan...

  11. Un análisis estadístico del equilibrio de nash en juegos repetidos usando implementaciones computacionales

    OpenAIRE

    Necco, Claudia Mónica; Quintas, Luis Guillermo

    1998-01-01

    En este trabajo se realiza un análisis de situaciones de conflicto estratégico que se producen periódicamente y que se modelan como juegos repetidos, bajo supuestos de racionalidad acotada y usando implementaciones computacionales (estrategias implementadas por autómatas de tamaño 2). Sin estas restricciones hay típicamente infinitos equilibrios lo cual naturalmente dificulta la posibilidad de saber que tipo de conductas (todas razonables) podrán tomar los agentes. En este caso, se hace...

  12. Solucionando dificultades en el aula: una estrategia usando el aprendizaje basado en problemas

    Directory of Open Access Journals (Sweden)

    Mery Luz Valderrama Sanabria

    2017-09-01

    Full Text Available Introducción: Se utiliza el proyecto de investigación en el aula con énfasis en el Aprendizaje Basado en Problemas, el cual promueve el aprendizaje activo y significativo, permitiendo solucionar situaciones reales de conocimiento en torno a una temática específica. Implementa los principios de la investigación formativa, como herramienta para generar nuevas alternativas en la apropiación del conocimiento. Este estudio tuvo como objetivo conocer la percepción de los estudiantes del programa Regencia de Farmacia frente a la utilización del aprendizaje basado en problemas con el fin de realizar aportes al currículo. Materiales y Métodos: Estudio descriptivo y transversal realizado con una muestra no probabilística, por conveniencia, conformada por 109 estudiantes de segundo a sexto semestre. Se elaboró un cuestionario con escala tipo Likert, sometido a valoración por expertos. Resultados: En general, los estudiantes están de acuerdo con la estrategia porque ha permitido acercarse a la investigación, fortaleciendo el pensamiento crítico; generando autonomía y responsabilidad frente al aprendizaje. A medida que avanzan los semestres, le ven mayor utilidad. Sin embargo, falta claridad en el uso de la metodología y capacitación por parte de algunos docentes para desarrollarla eficazmente. Discusión: La coordinación docente es fundamental, se debe fortalecer este aspecto para dar claridad al uso de la metodología. Implementar procesos de evaluación para determinar los avances desarrollados por los estudiantes y el impacto generado. Conclusiones: Los estudiantes consideran que adquieren conocimientos y competencias que les ayudarán en la práctica profesional.  Cómo citar este artículo: Valderrama ML, Castaño GA. Solucionando dificultades en el aula: una estrategia usando el aprendizaje basado en problemas. Rev Cuid. 2017; 8(3: 1907-18. http://dx.doi.org/10.15649/cuidarte.v8i3.456

  13. Aptitud combinatoria general y especifica de líneas tropicales de maiz usando probadores

    Directory of Open Access Journals (Sweden)

    Mauro Sierra

    2000-01-01

    Full Text Available Aptitud combinatoria general y específica de líneas tropicales de maíz usando probadores. Durante el ciclo O - I 1996/97 fueron evaluados en el Campo Experimental Cotaxtla mestizos de líneas sobresalientes y provenientes de varias fuentes de germoplasma como son : a Líneas recicladas de H-513 X VS-536, b Líneas derivadas de un compuesto de amplia base genética, c Líneas élite de programa de maíz de Cotaxtla (LTs y d Líneas de CIMMYT (CMLs. Como probadores se usaron las líneas LT-154 y LT-155 progenitores del híbrido H-513 y las líneas CML247 y CML254 cuya cruza es un patrón heterótico definido por CIMMYT para el trópico. Hubo líneas con buen comportamiento per-se tanto en rendimiento como en características agronómicas y que se encuentran formando mestizos sobresalientes con uno o varios probadores. Con relación a la Aptitud combinatoria, se encontró que las líneas F31XF30-4-3-1, F41XF40-1-2-1, CABG3’-12-2-1-2-1-1, LT174 y CML15 registraron los máximos valores con el probador 2 (LT155; F4XF3-5-2-1 y CML15 con el probador 4 (CMl254. Así también, las líneas F4XF5-5-1-1, y CABG3’-12-2-1-2-1-1, LT174, CML13 y CML15 con buena ACG. Con relación a los probadores, se encontró que para el grupo de líneas Recicladas , los probadores 1(LT154 y 4 (CML254 registraron los coeficientes de regresión más altos, lo que indica que permiten identificar líneas sobresalientes. Para líneas CABG fué el probador 2 (LT155 el que registró el mejor valor y en líneas Élite , los probadores 2(LT155, y 3(CML247 identificaron mejor a las líneas sobresalientes. Para las líneas del CIMMYT el mejor valor fué para el probador 4(CML254

  14. Avaliação de multielementos em amostras de sangue humano usando SR-TXRF

    Directory of Open Access Journals (Sweden)

    Nivia Graciele V. Pinto

    Full Text Available A técnica de fluorescência de raios X por reflexão total usando radiação síncrotron (SR-TXRF é uma poderosa ferramenta utilizada para a determinação das concentrações elementares presentes em amostras biológicas. O objetivo deste estudo é avaliar as possíveis alterações causadas por processos de irradiação na concentração de elementos-traço em amostras de sangue humano. As amostras de sangue foram coletadas no Laboratório de Análises Clínicas Dr. Elilel Figueiredo, Rio de Janeiro, e divididas em dois grupos. O primeiro grupo foi irradiado com doses de 1.500, 2.500 e 3.000 cGy, utilizando o irradiador Gammacell 220 Excel, e o segundo foi irradiado com doses que variaram de 2 cGy a 100 cGy, utilizando uma bomba de cobalto Theratron 780 C do Inca, Rio de Janeiro. Todas as amostras de sangue total, plasma e matriz celular foram então liofilizadas e, em seguida, passaram pelo procedimento padrão de digestão. Todas as medidas foram realizadas na linha de fluorescência de raios X do Laboratório Nacional de Luz Síncrotron (LNLS, em Campinas, Brasil. Não se verificou variação significativa na concentração de Ca e, em contrapartida, o K foi o único elemento que sofreu alterações significativas para todas as amostras analisadas em função da dose. A concentração de Fe diminuiu apenas para as amostras de sangue total e plasma. A concentração de Zn apresentou uma diminuição significativa somente para as amostras de sangue total.

  15. Dosimetry analysis of distribution radial dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radiais de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, T.S.; Yoriyaz, H. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, M.A.R. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Servico de Radioterapia; Louzada, M.J.Q. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

    2010-07-01

    Although they are no longer manufactured, the applicators of {sup 90}Sr +{sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been recalibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr+{sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radiochromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radiochromium films for this type of dosimetry. (author)

  16. Dosimetry analysis of distributions radials dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radias de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, Talita S.; Yoriyaz, Helio [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, Marco A.R., E-mail: tasallesc@gmail.co [UNESP, Botucatu, SP (Brazil). Faculdade de Medicina. Servico de Radioterapia; Louzada, Mario J.Q. [UNESP, Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

    2011-07-01

    Although they are no longer manufactured, the applicators of {sup 90}Sr + {sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been re calibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr + {sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radio chromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radio chromium films for this type of dosimetry. (author)

  17. 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

  18. IAEA GT-MHR benchmark calculations by using the HELIOS/MASTER physics analysis procedure and the MCNP Monte Carlo code

    International Nuclear Information System (INIS)

    Lee, Kyung-Hoon; Kim, Kang-Seog; Cho, Jin-Young; Song, Jae-Seung; Noh, Jae-Man; Lee, Chung-Chan

    2008-01-01

    The IAEA's gas-cooled reactor program has coordinated international cooperation for an evaluation of a high temperature gas-cooled reactor's performance, which includes a validation of the physics analysis codes and the performance models for the proposed GT-MHR. This benchmark problem consists of the pin and block calculations and the reactor physics of the control rod worth for the GT-MHR with a weapon grade plutonium fuel. Benchmark analysis has been performed by using the HELIOS/MASTER deterministic code package and the MCNP Monte Carlo code. The deterministic code package adopts a conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation. In order to solve particular modeling issues in GT-MHR, recently developed technologies were utilized and new analysis procedure was devised. Double heterogeneity effect could be covered by using the reactivity-equivalent physical transformation (RPT) method. Strong core-reflector interaction could be resolved by applying an equivalence theory to the generation of the reflector cross sections. In order to accurately handle with very large control rods which are asymmetrically located in a fuel and a reflector block, the surface dependent discontinuity factors (SDFs) were considered in applying an equivalence theory. A new method has been devised to consider SDFs without any modification of the nodal solver in MASTER. All computational results of the HELIOS/MASTER code package were compared with those of MCNP. The multiplication factors of HELIOS for the pin cells are in very good agreement with those of MCNP to within a maximum error of 693 pcm Δρ. The maximum differences of the multiplication factors for the fuel blocks are about 457 pcm Δρ and the control rod worths of HELIOS are consistent with those of MCNP to within a maximum error of 3.09%. On considering a SDF in the core

  19. Evaluation of the new electron-transport algorithm in MCNP6.1 for the simulation of dose point kernel in water

    Science.gov (United States)

    Antoni, Rodolphe; Bourgois, Laurent

    2017-12-01

    In this work, the calculation of specific dose distribution in water is evaluated in MCNP6.1 with the regular condensed history algorithm the "detailed electron energy-loss straggling logic" and the new electrons transport algorithm proposed the "single event algorithm". Dose Point Kernel (DPK) is calculated with monoenergetic electrons of 50, 100, 500, 1000 and 3000 keV for different scoring cells dimensions. A comparison between MCNP6 results and well-validated codes for electron-dosimetry, i.e., EGSnrc or Penelope, is performed. When the detailed electron energy-loss straggling logic is used with default setting (down to the cut-off energy 1 keV), we infer that the depth of the dose peak increases with decreasing thickness of the scoring cell, largely due to combined step-size and boundary crossing artifacts. This finding is less prominent for 500 keV, 1 MeV and 3 MeV dose profile. With an appropriate number of sub-steps (ESTEP value in MCNP6), the dose-peak shift is almost complete absent to 50 keV and 100 keV electrons. However, the dose-peak is more prominent compared to EGSnrc and the absorbed dose tends to be underestimated at greater depths, meaning that boundaries crossing artifact are still occurring while step-size artifacts are greatly reduced. When the single-event mode is used for the whole transport, we observe the good agreement of reference and calculated profile for 50 and 100 keV electrons. Remaining artifacts are fully vanished, showing a possible transport treatment for energies less than a hundred of keV and accordance with reference for whatever scoring cell dimension, even if the single event method initially intended to support electron transport at energies below 1 keV. Conversely, results for 500 keV, 1 MeV and 3 MeV undergo a dramatic discrepancy with reference curves. These poor results and so the current unreliability of the method is for a part due to inappropriate elastic cross section treatment from the ENDF/B-VI.8 library in those

  20. Evaluation of computational models and cross sections used by MCNP6 for simulation of characteristic X-ray emission from thick targets bombarded by kiloelectronvolt electrons

    Science.gov (United States)

    Poškus, A.

    2016-09-01

    This paper evaluates the accuracy of the single-event (SE) and condensed-history (CH) models of electron transport in MCNP6.1 when simulating characteristic Kα, total K (=Kα + Kβ) and Lα X-ray emission from thick targets bombarded by electrons with energies from 5 keV to 30 keV. It is shown that the MCNP6.1 implementation of the CH model for the K-shell impact ionization leads to underestimation of the K yield by 40% or more for the elements with atomic numbers Z 25. The Lα yields are underestimated by more than an order of magnitude in CH mode, because MCNP6.1 neglects X-ray emission caused by electron-impact ionization of L, M and higher shells in CH mode (the Lα yields calculated in CH mode reflect only X-ray fluorescence, which is mainly caused by photoelectric absorption of bremsstrahlung photons). The X-ray yields calculated by MCNP6.1 in SE mode (using ENDF/B-VII.1 library data) are more accurate: the differences of the calculated and experimental K yields are within the experimental uncertainties for the elements C, Al and Si, and the calculated Kα yields are typically underestimated by (20-30)% for the elements with Z > 25, whereas the Lα yields are underestimated by (60-70)% for the elements with Z > 49. It is also shown that agreement of the experimental X-ray yields with those calculated in SE mode is additionally improved by replacing the ENDF/B inner-shell electron-impact ionization cross sections with the set of cross sections obtained from the distorted-wave Born approximation (DWBA), which are also used in the PENELOPE code system. The latter replacement causes a decrease of the average relative difference of the experimental X-ray yields and the simulation results obtained in SE mode to approximately 10%, which is similar to accuracy achieved with PENELOPE. This confirms that the DWBA inner-shell impact ionization cross sections are significantly more accurate than the corresponding ENDF/B cross sections when energy of incident electrons

  1. 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

  2. IMPLEMENTACIÓN EN FPGA DE UN CLASIFICADOR DE MOVIMIENTOS DE LA MANO USANDO SEÑALES EMG

    Directory of Open Access Journals (Sweden)

    David Alexander Reyes Lopez

    2015-09-01

    Full Text Available Este trabajo presenta el diseño e implementación de un clasificador de señales electromiográficas (EMG para tres movimientos de la mano: flexión, extensión y cierre, usando dos músculos del antebrazo: palmar largo y extensor común de los dedos. El desarrollo comprende dos bloques principales: el hardware para la adquisición y adecuación de la señales EMG análogas, y el sistema de procesamiento para la identificación y clasificación del movimiento realizado; el sistema completo fue implementado en hardware usando un kit de desarrollo DE2-70 que cuenta con un FPGA Cyclone II de Altera. Para la extracción de características la Transformada Rápida de Fourier (FFT es implementada para cada canal y características como la varianza y el promedio fueron calculados. Finalmente, se establece un umbral de decisión para identificar el movimiento realizado. El tiempo de respuesta del sistema total fue de 17,7 us, obteniendo una tasa de identificación mayor al 87%.

  3. Reasons between effective doses for tomographic and mathematical models due to external exposition by photons; Razoes entre doses efetivas para modelos tomograficos e matematicos devido a exposicoes externas a fotons

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, R.; Khoury, H.J. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Vieira, J.W. [Centro Federal de Educacao Tecnologica de Pernambuco (CEFET-PE), Recife, PE (Brazil); Yoriyaz, H. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Lima, F.R.A. [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Loureiro, E.C.M. [Escola Politecnica de Pernambuco (POLI/UPE), Recife, PE (Brazil)

    2005-07-01

    The development of Monte Carlo codes and new and sophisticated tomographic human models, or based on voxel, motivated the ICRP to propose a revision of the traditional exposition models, which have been used to calculate doses on organs and tissues using mathematical phantoms MIRD-type 5. This article presents calculations made with tomographic phantoms MAX (Male Adult voXel) and FAX (Female Adult voXel), recently developed and also, for comparison, with ADAM and Eve mathematician phantoms. All models were coupled to the EGS4 and MCNP4 codes for full body external irradiation by photons. It were simulated expositions AP, PA and rotational for energies varying between 10 keV and 10 MeV. The effective calculated doses were compared separately to evaluate: the replacement of the Monte Carlo code; the composition of the tissues and the replacement of tomographic phantoms by mathematical ones. Effective doses calculated results indicate that for external exposures by photons to introduce models based on voxels can cause a reduction of about 10% to the energies considered in this study.

  4. Ratios between effective doses for tomographic and mathematician models due to internal exposure of photons; Razoes entre doses efetivas para modelos tomograficos e modelos matematicos devido as exposicoes internas de fotons

    Energy Technology Data Exchange (ETDEWEB)

    Lima, F.R.A. [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Kramer, R.; Khoury, H.J.; Santos, A.M. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Vieira, W. [Centro Federal de Educacao Tecnologica de Pernambuco (CEFET-PE), Recife, PE (Brazil); Loureiro, E.C.M. [Escola Politecnica de Pernambuco, Recife, PE (Brazil)

    2005-07-01

    The development of new and sophisticated Monte Carlo codes and tomographic human phantoms or voxels motivated the International Commission on Radiological Protection (ICRP) to revise the traditional models of exposure, which have been used to calculate effective dose coefficients for organs and tissues based on mathematician phantoms known as MIRD5. This paper shows the results of calculations using tomographic phantoms MAX (Male Adult voXel) and FAX (Female Adult voXel), recently developed by the authors as well as with the phantoms ADAM and EVA, of specific genres, type MIRD5, coupled to the EGS4 Monte Carlo and MCNP4C codes, for internal exposure with photons of energies between 10 keV and 4 MeV to several organs sources. Effective Doses for both models, tomographic and mathematician, will be compared separately as a function of the Monte Carlo code replacement, of compositions of human tissues and the anatomy reproduced through tomographs. The results indicate that for photon internal exposure, the use of models of exposure based in voxel, increases the values of effective doses up to 70% for some organs sources considered in this study, when compared with the corresponding results obtained with phantoms of MIRD-5 type.

  5. GRINDING OF HARDENED STEELS USING OPTIMIZED COOLING RECTIFICADO DE ACEROS ENDURECIDOS USANDO REFRIGERACIÓN OPTIMIZADA

    Directory of Open Access Journals (Sweden)

    Manoel Cléber de Sampaio Alves

    2008-06-01

    medio. Para ello, al proceso de rectificación está intrínseco el reciclaje del fluido de corte, que se destaca por su costo. A través de la variación de la velocidad de avance en el proceso de rectificación cilíndrica externa del acero ABNT D6, racionalizando la aplicación de dos fluidos de corte y usando una muela superabrasiva de CBN (nitruro de boro cúbico con ligante vitrificado, se evaluaron los parámetros de salida fuerza tangencial de corte, rugosidad, circularidad, desgaste de la herramienta, la tensión residual y la integridad superficial a través de la microscopia electrónica de barrido (SEM de las piezas de prueba. Con el análisis del desempeño fluido, muela y velocidad de inmersión se encontró las mejores condiciones de fabricación propiciando la disminución del volumen de fluido de corte, disminución del tiempo de fabricación sin perjudicar los parámetros geométricos, dimensionales, el acabado superficial y la integridad superficial de los componentes.

  6. Investigation of reactivity variations of the Isfahan MNSR reactor due to variations in the thickness of the core top beryllium layer using WIMSD and MCNP codes

    Directory of Open Access Journals (Sweden)

    A Shirani

    2010-12-01

    Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core and also after 14 years of operation (total operation time of the reactor is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

  7. 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)

  8. 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)

  9. A comparative study of the neutron flux spectra in the MNSR irradiation sites for the HEU and LEU cores using the MCNP4C code.

    Science.gov (United States)

    Dawahra, S; Khattab, K; Saba, G

    2015-10-01

    A comparative study for fuel conversion from the HEU to LEU in the Miniature Neutron Source Reactor (MNSR) has been performed in this paper using the MCNP4C code. The neutron energy and lethargy flux spectra in the first inner and outer irradiation sites of the MNSR reactor for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) were investigated using the MCNP4C code. The neutron energy flux spectra for each group was calculated by dividing the neutron flux by the width of each energy group. The neutron flux spectra per unit lethargy was calculated by multiplying the neutron energy flux spectra for each energy group by the average energy of each group. The thermal neutron flux was calculated by summing the neutron fluxes from 0.0 to 0.625 eV, the fast neutron flux was calculated by summing the neutron fluxes from 0.5 MeV to 10 MeV for the existing HEU and potential LEU fuels. Good agreements have been noticed between the flux spectra for the potential LEU fuels and the existing HEU fuels with maximum relative differences less than 10% and 8% in the inner and outer irradiation sites. Copyright © 2015 Elsevier Ltd. All rights reserved.

  10. Validación de la escala para manía de la Universidad Nacional de Colombia usando el análisis de Rasch

    Directory of Open Access Journals (Sweden)

    Ricardo Sánchez

    2011-03-01

    Conclusiones. En este primer estudio de la escala para manías usando el análisis de Rasch, se detectó mal ajuste y redundancia de algunos ítems. El síndrome maníaco no queda completamente evaluado por la escala. El instrumento podría mejorarse agregando síntomas depresivos.

  11. 2521-IJBCS-Article-Oscar Foton

    African Journals Online (AJOL)

    hp

    Water hyacinth problems in Mexico and practiced methods for control, 125-135. Gopal B. 1987. Water Hyacinth. Elsevier: Amsterdam, the Netherlands; 471. Harley KLS, Julien M H, Wright AD. 1997. Water hyacinth: A tropical worldwide problem and methods for its control. Proceedings of the first meeting of the. International.

  12. Síntesis de oxitocina en fase sólida usando derivados de terbutoxicarbonilo y fluorenilmetoxicarbonilo

    OpenAIRE

    Calvo, Julio C; Barrera, Nubia F.; García, Josué A.; Guzman, Fanny; Espejo, Fabiola; Patarroyo, Manuel E.

    2010-01-01

    La oxitocina, péptido cíclico cuya secuencia es CYIQNCPLG, fué el primer péptido de importancia biológica que pudo ser sintetizado. En este trabajo se compara la síntesis de la oxitocina usando resina p metilbenzhidrilamina (MBHA) para la síntesis por estrategia t-Boc y resina Rink p-metilbenzhidrilamina (Rink MBHA) para la síntesis por estrategia Fmoc, con altos rendimientos. El péptido crudo se ciclizó en una disolución acuosa de dimetilsulfóxido al 10%. La caracterización se llevó a cabo p...

  13. 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

  14. Use the nuclear code MCNP4X in the study of the behavior of nuclear probe in soils with variation of Mg, Ca, Fe

    Energy Technology Data Exchange (ETDEWEB)

    Braga, Mario R.M.S.S.; Oliveira, Arno H.; Lima, Claubia P.B., E-mail: mario@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear; Silva, Clemente J.G.C. da, E-mail: clementecarneiro@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Engenharia Nuclear; Carneiro, Andre C., E-mail: andreccarneiro@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The aim of this work is to evaluate the behavior of the variation the elements: Mg, Ca, Fe in the soils composition on a nuclear probe to measure the density of porous materials nondestructive in testing based on coherent Compton Effect, the effect Rayleigh. To study the effect of composition in soil was used nuclear code MCNP4X where was simulated two sources, a source 14mCi americium-241 and other source 4mCi cesium-137, lead shielding and volume scintillator. To avoid problems with geometries were simulated spheres with 1.00 meters of diameter filled with soil to be evaluated. Data analysis allowed establishing correction parameters for nuclear probe. (author)

  15. Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

    Science.gov (United States)

    Khattab, K; Sulieman, I

    2009-04-01

    The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

  16. Acceleration of MCNP calculations for small pipes configurations by using Weigth Windows Importance cards created by the SN-3D ATTILA

    Science.gov (United States)

    Castanier, Eric; Paterne, Loic; Louis, Céline

    2017-09-01

    In the nuclear engineering, you have to manage time and precision. Especially in shielding design, you have to be more accurate and efficient to reduce cost (shielding thickness optimization), and for this, you use 3D codes. In this paper, we want to see if we can easily applicate the CADIS methods for design shielding of small pipes which go through large concrete walls. We assess the impact of the WW generated by the 3D-deterministic code ATTILA versus WW directly generated by MCNP (iterative and manual process). The comparison is based on the quality of the convergence (estimated relative error (σ), Variance of Variance (VOV) and Figure of Merit (FOM)), on time (computer time + modelling) and on the implement for the engineer.

  17. Use the nuclear code MCNP4X in the study of the behavior of nuclear probe in soils with variation of Mg, Ca, Fe

    International Nuclear Information System (INIS)

    Braga, Mario R.M.S.S.; Oliveira, Arno H.; Lima, Claubia P.B.

    2013-01-01

    The aim of this work is to evaluate the behavior of the variation the elements: Mg, Ca, Fe in the soils composition on a nuclear probe to measure the density of porous materials nondestructive in testing based on coherent Compton Effect, the effect Rayleigh. To study the effect of composition in soil was used nuclear code MCNP4X where was simulated two sources, a source 14mCi americium-241 and other source 4mCi cesium-137, lead shielding and volume scintillator. To avoid problems with geometries were simulated spheres with 1.00 meters of diameter filled with soil to be evaluated. Data analysis allowed establishing correction parameters for nuclear probe. (author)

  18. An approach to design a90Sr 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.

  19. 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 137 Cs 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 ( 137 Cs) 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 137 Cs 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 ( 137 Cs) at a scattering angle of 120°.

  20. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  1. Dosimetry boron neutron capture therapy in liver cancer (hepatocellular carcinoma) by means of MCNP-code with neutron source from thermal column

    International Nuclear Information System (INIS)

    Irhas; Andang Widi Harto; Yohannes Sardjono

    2014-01-01

    Boron Neutron Capture Therapy (BNCT) using physics principle when B 10 (Boron-10) irradiated by low energy neutron (thermal neutron). Boron and thermal neutron reaction produced B 11m (Boron-11m) (t 1/2 =10 -2 s). B 11m decay emitted alpha, Li 7 (Lithium-7) particle and gamma ray. Irradiated time needed to ensure cancer dose enough. Liver cancer was primary malignant who located in liver (Hepatocellular carcinoma). Malignant in liver were different to metastatic from Breast, Colon Cancer, and the other. This condition was Metastatic Liver Cancer. Monte Carlo method used by Monte Carlo N-Particle (MCNP) Software. Probabilistic approach used for probability of interaction occurred and record refers to characteristic of particle and material. In this case, thermal neutron produced by model of Collimated Thermal Column Kartini Research Nuclear Reactor, Yogyakarta. Modelling organ and source used liver organ that contain of cancer tissue and research reactor. Variation of boron concentration was 20, 25, 30, 35, 40, 45, and 47 µg/g cancers. Output of MCNP calculation were neutron scattering dose, gamma ray dose and neutron flux from reactor. Neutron flux used to calculate alpha, proton and gamma ray dose from interaction of tissue material and thermal neutron. Variation of boron concentration result dose rate to every variation were 0,059; 0,072; 0,084; 0,098; 0.108; 0,12; 0,125 Gy/sec. Irradiation time who need to every concentration were 841,5 see (14 min 1 sec); 696,07 sec(11 min 36 sec); 593.11 sec (9 min 53 sec); 461,35 sec (8 min 30 sec); 461,238 sec (7 min 41 sec); 414,23 sec (6 min 54 sec); 398,38 sec (6 min 38 sec). Irradiating time could shortly when boron concentration more high. (author)

  2. Estructura y diversidad genética en vacas Holstein de Antioquia usando un polimorfismo del gen bGH

    Directory of Open Access Journals (Sweden)

    Juan Rincon F.

    2013-03-01

    Full Text Available Objetivo. Determinar las frecuencias alélicas y genotípicas del polimorfismo del intrón 3 del gen bGH y estimar algunos parámetros de estructura poblacional en ganado Holstein. Materiales y métodos. El estudio se realizó con 1366 vacas Holstein en 120 hatos de 11 municipios del departamento de Antioquia. Se extrajo DNA por el método de Salting out y la genotipificación se realizó usando la técnica de PCR-RFLPs. La diversidad genética se determinó mediante la comparación de las heterocigosidades, El equilibrio de Hardy-Weinberg (HW y la diferenciación genética entre las poblaciones se realizó usando el software Arlequín 2.0 Las frecuencias alélicas y genotípicas se evaluaron mediante el paquete estadístico SAS®. Resultados. Las frecuencias genotípicas encontradas fueron 0.764 (+/+, 0.223 (+/- y 0.013 (-/- y las frecuencias alélicas 0.876 (+ y 0.124 (-. No se encontraron desviaciones del Equilibrio de Hardy Weinberg en ninguna de las subpoblaciones. La diversidad genética determinada mediante la comparación de las heterocigosidades fue relativamente baja entre poblaciones pero al interior de estas no. El valor de FST de toda la población fue de 0.0068 y significativo (p<0.05, algunos FST pareados también lo fueron, tomando valores desde 0.0 a 0.13. Los estadísticos FIT y FIS no fueron significativos. Conclusiones. El gen bGH es un candidato interesante para evaluar características de importancia económica ya que no parece haber sido sometido a selección directa, presenta una variabilidad media en las poblaciones, observándose diferenciación genética significativa entre distintos municipios, producto de los diferentes sistemas de producción y acceso a las biotecnologías.

  3. 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)

  4. Análisis de la biodiversidad genética del algodón peruano usando marcadores moleculares: Avances en el 2004

    OpenAIRE

    Olórtegui, José; Espinoza, Marco; Espinoza, José; Montoya, Ysabel

    2005-01-01

    Tres mini preparaciones de extracción de ADN de algodón fueron comparadas en términos de calidad y rendimiento. El método de extracción de ADN usando CTAB fue el más eficiente (30 ug) en comparación con un kit comercial de extracción (20 ug) a partir de 100 mg de hojas cotiledonarias. La óptima calidad del ADN fue evaluada con las enzimas de restricción EcoRI y MseI. El ADN preparado será usado para iniciar el análisis de la biodiversidad genética del algodón peruano, usando marcadores mole...

  5. Monitorização de transplante cardíaco usando análises do eletrograma intracavitário

    Directory of Open Access Journals (Sweden)

    BROFMAN Paulo Roberto S.

    1997-01-01

    Full Text Available Uma série de registros do eletrograma intracavitário tem sido utilizada para monitorização não invasiva da rejeição em pacientes transplantados usando um marcapasso de dupla câmara e eletrodos endocavitários revestidos com estrutura fractal. Os sinais têm sido avaliados usando o sistema CHARM (Computerized Heart Acute Rejection Monitoring _ sistema computadorizado para monitorização da rejeição cardíaca aguda. Os relatórios obtidos com este sistema contêm curvas com parâmetros sensíveis à rejeição, que demonstram uma boa correlação com a clínica e os resultados das biópsias convencionais. A monitorização a longo prazo, usando estas análises, mostrou ser uma ferramenta valiosa no acompanhamento destes pacientes.

  6. ESTABILIDAD AMBIENTAL EN HÍBRIDOS DE MAÍZ USANDO EL MODELO AMMI EN EL LITORAL ECUATORIANO

    Directory of Open Access Journals (Sweden)

    Marlon Brainer Caicedo Villafuerte

    2017-06-01

    Full Text Available El objetivo de esta investigación fue evaluar la adaptabilidad y estabilidad del rendimiento de grano en 27 híbridos de maíz, usando el modelo de efectos principales aditivos e interacción multiplicativa (AMMI. Los ensayos fueron conducidos en tres ambientes del Litoral ecuatoriano durante la época seca del año 2012. Se realizó un análisis de varianza combinado entre ambientes, posteriormente un análisis de consistencia y finalmente el análisis de efectos principales aditivos e interacción multiplicativa, para la variable rendimiento. El ambiente, híbrido y la interacción híbrido × ambiente explicaron el 13,74, 43,78 y 15.50% del total de la suma de cuadrados, respectivamente. El primer componente principal del análisis del modelo AMMI fue significativo (P<0,001, explicando un 82,11% de la suma de cuadrados de la interacción. Los híbridos triples seleccionados H11 y H15 mostraron alto rendimiento y estabilidad para todos los ambientes. El ambiente más eficiente para discriminar los híbridos fue Santa Ana. El modelo AMMI resultó muy útil para identificar híbridos de maíz altamente productivos y con buena estabilidad.

  7. DIGESTÃO DE CARBOIDRATOS USANDO DIGESTIVOS ENZIMÁTICOS COMERCIAIS – UMA AULA PRÁTICA.

    Directory of Open Access Journals (Sweden)

    Maria Teresa Pedrosa Silva Clerici

    2006-03-01

    Full Text Available Este trabalho mostra uma aula prática na qual se estuda a digestão decarboidratos “in vitro” utilizando amido gelatinizado como substrato e doismedicamentos comerciais indicados para problemas digestórios, como fonte deamilase. As reações enzimáticas foram feitas no equipamento dissolutor a 37º C, comagitação constante e usando meios que simulavam o pH do estômago e em seguida opH do intestino delgado. Para avaliar a eficácia das enzimas pancreáticas nosdigestivos, foi feita uma reação com amilase pancreática pura. Determinou-se aconcentração de glicose no meio de reação 0, 30, 60, 90 e 120 min após o inicio dareação. Nossos resultados mostram que há diferenças entre a atividade amilase dosdois fármacos.

  8. MODELAMIENTO DE LA CINÉTICA DE BIOADSORCIÓN DE Cr (III USANDO CÁSCARA DE NARANJA

    Directory of Open Access Journals (Sweden)

    MARTHA LUCIA PINZÓN-BEDOYA

    2009-01-01

    Full Text Available En este trabajo se utilizó como material bioadsorbente cáscara naranja con el fin de remover iones cromo presentes en soluciones hipotéticas diluidas, utilizadas como modelo de aguas contaminadas con bajas concentraciones de este metal. Las condiciones de operación utilizadas fueron: relación sólido/líquido 4 g/l, tamaño de partícula ??0,425 mm, concentración inicial de disolución de Cr(III 100 mg/l, tiempo de contacto 60 h y pH constante (4 y 5. Los modelos cinéticos escogidos para identificar el mecanismo de reacción del proceso de bioadsorción usando cáscara de naranja fueron: primer orden reversible, pseudo- segundoorden, Elovich y difusión intraparticular. Los resultados indican que la ecuación de Elovich proporciona mayor exactitud en el ajuste de los datos experimentales del equilibrio a este modelo cinético. La bondad del ajuste de los datos se realizó por regresión no-lineal utilizando como criterio la minimización de la función objeto suma de los cuadrados del error, SCE, haciendo uso de la herramienta matemática MATLAB.

  9. Criptosporidiose em paciente com espondilite anquilosante usando adalimumabe Cryptosporidiosis in a patient with ankylosing spondylitis treated with adalimumab

    Directory of Open Access Journals (Sweden)

    Fernando Augusto Chiuchetta

    2010-06-01

    Full Text Available A criptosporidiose é uma doença parasitária causada pelo protozoário Cryptosporidium sp. Observou-se um aumento no número de diagnósticos realizados nos últimos vinte anos, principalmente em pacientes que apresentam imunodeficiências como a síndrome da imunodeficiência humana adquirida e as imunodeficiências induzidas como em pacientes transplantados e nos que necessitam realizar hemodiálise frequentemente. Relata-se o caso de um jovem com espondilite anquilosante que, usando adalimumabe, apresentou diarreia devido à criptosporidiose.Cryptosporidiosis is a parasitic disease caused by a protozoan called Cryptosporidium sp. An increased number of diagnoses were made in the last 20 years, especially in patients with immunodeficiency like the acquired human immunodeficiency syndrome and induced immunodeficiency, such as in transplant patients and those who need frequent hemodialysis, has been observed. We report the case of a young patient with ankylosing spondylitis treated with adalimumab who developed chronic diarrhea secondary to cryptosporidiosis

  10. DISEÑO DE UN GENERADOR DE FLUJO AXIAL USANDO EL MÉTODO DE ELEMENTOS FINITOS

    Directory of Open Access Journals (Sweden)

    Luisa Herrera

    2013-12-01

    Full Text Available Este artículo presenta las simulaciones y cálculos realizados para el diseño básico de un generador de flujo axial usando el método de elementos finitos. Durante el proceso de diseño se evalúa el comportamiento electromagnético de los imanes seleccionados, se obtiene la característica B vs distancia, así mismo se define la geometría a usar para el montaje de las bobinas y los imanes, calculando el flujo máximo. Posteriormente se analiza la variación del flujo magnético dentro de la bobina en función de la posición angular, se define el número mínimo de espiras a usar por polo y se calcula la tensión inducida en la máquina asumiendo que tiene un sistema de 12 polos.

  11. JMat - Herramienta remota de cálculo y multiusuario para el aprendizaje basado en problemas usando Matlab

    Directory of Open Access Journals (Sweden)

    Bladimir Bacca Cortes

    2011-01-01

    Full Text Available JMat es una herramienta de cálculo basada en JAVA y EJS (Easy Java Simulations, con un esquema cliente / servidor, soporte multi-usuario y acceso remoto a Matlab. La aplicación está orientada a brindar a los usuarios una interacción con Matlab usando tres interfaces: Consola de Comandos, donde se invocan remotamente comandos de texto compatibles con Matlab. Espacio de Trabajo y Graficación, donde se mantiene un registro automático de las variables de usuario y se grafican individualmente. Funciones de usuario y Transferencia de Archivos, donde el usuario crea sus funciones, envía y recibe datos hacia y desde el servidor. JMat requiere un acceso a Internet, un servidor remoto donde esté instalado Matlab y un cliente (Navegador WEB o aplicación. No se requiere Matlab en el cliente. JMat está siendo usada actualmente en la Universidad del Valle en los cursos de Control Automático de Procesos, Control Inteligente, Redes Neuronales Artificiales, Procesamiento de Señales y Tratamiento Digital de Imágenes como herramienta para el aprendizaje basado en problemas empleando la plataforma de eLearning de la Universidad del Valle.

  12. Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Plutonium Metals, Oxides, and Solutions 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); Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-12-05

    This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.

  13. Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part II: Benchmark comparisons of PUMA core parameters with MCNP5 and improvements due to a simple cell heterogeneity correction

    International Nuclear Information System (INIS)

    Grant, C.; Mollerach, R.; Leszczynski, F.; Serra, O.; Marconi, J.; Fink, J.

    2006-01-01

    In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO 2 rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)

  14. Comparison of MCNP trademark and MAVRIC for the case of a radiography experiment with thick-walled multi-layered shielding and Co-60 source; Vergleich von MCNP trademark und MAVRIC am Beispiel eines Durchstrahlungsexperiments mit dickwandiger mehrschichtiger Abschirmung und Co-60 Quelle

    Energy Technology Data Exchange (ETDEWEB)

    Schloemer, L.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2012-11-01

    The application of calculation methods to demonstrate an effective radiation shielding requires validation with appropriate measurements. Experimental data on the shielding of high-energy gamma radiation representing the situation of a thick-wall container with lead insert and significant self-shielding of the inventory is not publicly available. Therefore the shielding efficiency of multi-layered shielding (Pb, polyethylene) of a container was measured using an industrial Co-60 point source and a precisely calibrated measuring system for different shielding situations and measuring positions outside of the shielding. The calculations were performed using the codes MCNP trademark and MAVRIC. The deviations of calculated and experimental data are constant within the uncertainty of the experimental set-up, no distance dependent drift was identified.

  15. Expresión transitoria del gen GUS en caña de azúcar usando Agrobacterium tumefaciens

    Directory of Open Access Journals (Sweden)

    Martha Liliana Bonilla Betancourt

    2008-10-01

    Full Text Available En el estudio se desarrolló una metodología de transformación genética mediante Agrobacterium tumefaciens en cultivares colombianos de caña de azúcar. La transformación se evaluó mediante la expresión del gen GUS. Callos embriogénicos y explantes meristemáticos de los genotipos CC85-92, CC84-75 y CC87-505 se transformaron usando tres cepas (AGL-1, LBA4404 y EHA105 con el plásmido pCambia 1305.2 y dos (EHA105 y LBA4404 con pCambia 2301. Se usó el medio de infiltración (IM con acetosiringona y se evaluó el tiempo de cocultivo y la densidad óptica de la bacteria al momento de la inducción. Los genotipos mostraron respuesta diferencial con las combinaciones cepa-plásmido: obtuvieron mayor expresión del gen GUS cuando el genotipo CC85-92 se transformó con la cepa AGL-1-pCambia 1305.2. CC84-75 y CC87-505 mostraron mayor expresión cuando se transformaron con la cepa EHA105-pCambia 1305.2. Mayor eficiencia en la expresión se obtuvo cuando la bacteria se indujo en IM después de siete días de cocultivo y cuando la densidad óptica de la bacteria fue de 0.2(600nm al momento de la inducción. Se demostró superioridad de los explantes en la eficiencia de transformación.

  16. Modelo Bio-inspirado para el Reconocimiento de Gestos Usando Primitivas de Movimiento en Visión

    Directory of Open Access Journals (Sweden)

    Sandra E. Nope

    2008-10-01

    Full Text Available Resumen: Se aborda el problema del reconocimiento de gestos usando la información de movimiento con el fin de obtener un modelo bio-inspirado para, en un futuro, utilizarlo en la programación de robots mediante el paradigma del aprendizaje por imitación. En este trabajo se extraen las primitivas de movimiento a partir de imágenes consecutivas, capturadas por una cámara web estándar. Para la programación por imitación de robots se identificó, como primera fase, el reconocimiento de gestos, en el cual es necesario resolver tres aspectos principales: La representación instantánea del movimiento, la integración temporal de dicha información y, la estrategia de clasificación. Estos tres aspectos serán tratados a lo largo de este trabajo y, en contraste con otros, la extracción del movimiento y su codificación está inspirada en el procesamiento del movimiento realizado en el cerebro de macacos. El modelo obtenido fue aplicado al reconocimiento de cuatro tipos de gestos realizados con la mano por diferentes personas. El porcentaje de aciertos varió entre 91.42% y 97.14%, utilizando diferentes estrategias estándar de clasificación. Palabras clave: Reconocimiento de gestos, modelo bio-inspirado, primitivas de movimiento, codificación del movimiento, integración temporal, visión artificial

  17. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    International Nuclear Information System (INIS)

    Pauzi, A M

    2013-01-01

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called U -battery TM, which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

  18. Study of the radioactive particle tracking technique using gamma-ray attenuation and MCNP-X code to evaluate industrial agitators

    Energy Technology Data Exchange (ETDEWEB)

    Dam, Roos Sophia de F.; Salgado, César M., E-mail: rsophia.dam@gmail.com, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Agitators or mixers are highly used in the chemical, food, pharmaceutical and cosmetic industries. During the fabrication process, the equipment may fail and compromise the appropriate stirring or mixing procedure. Besides that, it is also important to determine the right point of homogeneity of the mixture. Thus, it is very important to have a diagnosis tool for these industrial units to assure the quality of the product and to keep the market competitiveness. The radioactive particle tracking (RPT) technique is widely used in the nuclear field. In this paper, a method based on the principles of the RPT technique is presented. Counts obtained by an array of detectors properly positioned around the unit will be correlated to predict the instantaneous positions occupied by the radioactive particle by means of an appropriate mathematical search location algorithm. Detection geometry developed employs eight NaI(Tl) scintillator detectors and a Cs-137 (662 keV) source with isotropic emission of gamma-rays. The modeling of the detection system is performed using the Monte Carlo Method, by means of the MCNP-X code. In this work a methodology is presented to predict the position of a radioactive particle to evaluate the performance of agitators in industrial units by means of an Artificial Neural Network (ANN). (author)

  19. Application of the MCNP5 code to the Modeling of vaginal and intra-uterine applicators used in intracavitary brachytherapy: a first approach

    Energy Technology Data Exchange (ETDEWEB)

    Gerardy, I; Tondeur, F [Institut Superieur Industriel de Bruxelles, 150, Rue Royale, B-1000 Brussels (Belgium); Rodenas, J; Gallardo, S [Departamento de IngenierIa QuImica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain); Dycke, M Van [Clinique Saint Jean, Bld du Jardin Botanique, B-1000 Brussels (Belgium)], E-mail: gerardy@isib.be

    2008-02-01

    Brachytherapy is a radiotherapy treatment where encapsulated radioactive sources are introduced within a patient. Depending on the technique used, such sources can produce high, medium or low local dose rates. The Monte Carlo method is a powerful tool to simulate sources and devices in order to help physicists in treatment planning. In multiple types of gynaecological cancer, intracavitary brachytherapy (HDR Ir-192 source) is used combined with other therapy treatment to give an additional local dose to the tumour. Different types of applicators are used in order to increase the dose imparted to the tumour and to limit the effect on healthy surrounding tissues. The aim of this work is to model both applicator and HDR source in order to evaluate the dose at a reference point as well as the effect of the materials constituting the applicators on the near field dose. The MCNP5 code based on the Monte Carlo method has been used for the simulation. Dose calculations have been performed with *F8 energy deposition tally, taking into account photons and electrons. Results from simulation have been compared with experimental in-phantom dose measurements. Differences between calculations and measurements are lower than 5%.The importance of the source position has been underlined.

  20. SU-E-T-13: Comparison of Dose Rates with and without Gold Backing of USC #9 Radioactive Eye Plaque Using MCNP5.

    Science.gov (United States)

    Aryal, P; Molloy, J

    2012-06-01

    To show the effect of gold backing on dose rates for the USC #9 radioactive eye plaque. An I125 source (IsoAid model IAI-125A) and gold backing was modeled using MCNP5 Monte Carlo code. A single iodine seed was simulated with and without gold backing. Dose rates were calculated in two orthogonal planes. Dose calculation points were structured in two orthogonal planes that bisect the center of the source. A 2×2 cm matrix of spherical points of radius 0.2 mm was created in a water phantom of 10 cm radius. 0.2 billion particle histories were tracked. Dose differences with and without the gold backing were analyzed using Matlab. The gold backing produced a 3% increase in the dose rate near the source surface (source center but off axis. The dose decreased by 25%, 65% and 81% at 1, 2, and 3 mm off axis at a distance of 1 mm from the source surface. These effects were less pronounced in the perpendicular dimension near the source tip, where maximum dose decreases of 2% were noted. I 125 sources embedded directly into gold troughs display dose differences of 2 - 90%, relative to doses without the gold backing. This is relevant for certain types of plaques used in treatment of ocular melanoma. Large dose reductions can be observed and may have implications for scleral dose reduction. © 2012 American Association of Physicists in Medicine.

  1. 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.

  2. Radiological protection on interstitial brachytherapy and dose determination and exposure rate of an Ir-192 source through the MCNP-4B

    International Nuclear Information System (INIS)

    Morales L, M.E.

    2006-01-01

    The present work was carried out in the Neurological Sciences Institute having as objective to determine the dose and the rate of exhibition of the sources of Iridium 192, Iodine 125 and Palladium 103; which are used to carry out implant in the Interstitial Brachytherapy according to the TG43. For it we carry out a theoretical calculation, its are defined in the enter file: the geometry, materials of the problem and the radiation source, etc; in the MCNP-4B Monte Carlo code, considering a punctual source and for the dose determination we simulate thermoluminescent dosemeters (TLD): at 5 cm, 50 cm, 100 cm and 200 cm of the source. Our purpose is to analyze the radioprotection measures that should take into account in this Institute in which are carried out brain biopsies using a Micro mar stereotactic mark, and in a near future with the collaboration of a doctor and a cuban physique seeks to be carried out the Interstitial Brachytherapy technique with sources of Ir-192 for patient with tumors like glioblastoma, astrocytoma, etc. (Author)

  3. Simulação de cirurgia mamária usando Elementos Finitos com modelos reconstruídos a partir de mamografias

    OpenAIRE

    Edgar Moraes Diniz

    2011-01-01

    A simulação de procedimentos cirúrgicos apresenta-se como uma poderosa ferramenta de auxílio ao profissional de saúde. Entre suas aplicações, destacam-se treinamento virtual, previsão de resultados, auxílio na decisão do melhor procedimento a ser executado e melhoria na comunicação médico-paciente. Este trabalho apresenta uma metodologia para simulação de cirurgia mamária usando elementos finitos. Os modelos são construídos a partir de imagens de mamografia. Discutimos todas...

  4. Desarrollo de una Unidad Didáctica de matemáticas usando como recurso las tecnologías de la información

    OpenAIRE

    Morcillo de Pablos, Francisco, J.

    2012-01-01

    Este trabajo versa sobre el desarrollo de una unidad didáctica de matemáticas usando las TIC como recurso predominante dentro de la formación de personas adultas. Se trata de confeccionar un documento con todos los elementos característicos de una unidad didáctica (UD) pero adaptada a la singularidad del alumnado adulto y en concreto, a un grupo de Formación Básica, en su mayoría mujeres de mediana edad que han tenido poco o escaso contacto con las TIC, con el fin de aumentar el gusto por las...

  5. Educación virtual usando tecnología de redes para la formación a distancia, de profesores de matemáticas

    OpenAIRE

    Gamboa, Jesús; Ávila, Ramiro

    2007-01-01

    El presente reporte corresponde a una investigación enmarcada dentro de un proyecto de investigación diseñado para indagar las ventajas y dificultades que se presentan al desarrollar un programa de educación virtual a distancia para formar y/o actualizar profesores de matemáticas de los niveles: superior y bachillerato; usando tecnología de redes. En particular, en este trabajo se reportan las dificultades que tuvieron los profesores al usar las nuevas tecnologías así como los cambios en las ...

  6. MODELACIÓN NUMÉRICA DE PROCESOS DE ESTERILIZACIÓN TÉRMICA DE ALIMENTOS USANDO VOLUMENES DE CONTROL: APROXIMACIÓN CILÍNDRICA

    OpenAIRE

    HÉCTOR J. CIRO-VELÁSQUEZ; CARLOS GONZÁLEZ; EDUARD GARCÍA

    2009-01-01

    Un proceso de esterilización térmica de alimentos fue modelado usando la técnica de diferenciación finita aplicada a volúmenes de control. El modelo de simulación fue desarrollado en coordenadas cilíndricas con un ejemplo de aplicación tomando como referencia un producto cárnico (carne de res). Los resultados de la simulación mostraron que el algoritmo desarrollado es independiente de la red nodal seleccionada, permitiendo cuantificar las curvas de penetración de calor en el punto crítico del...

  7. Valoración de la formación recibida usando un perfil de referencia basado en competencias profesionales

    Directory of Open Access Journals (Sweden)

    Mónica Maldonado Rojas

    Full Text Available Introducción: La formación profesional está hoy en día promoviendo la implementación de opciones educativas basadas en modelos por competencias. El presente trabajo tuvo como propósito establecer un perfil de referencia basado en competencias profesionales para la carrera de Tecnología Médica de la Universidad de Talca y usando éstas como indicadores, conocer la auto percepción que tienen los egresados respecto del grado en que fueron adquiridas durante su formación. Material y métodos: Se realizó una investigación de tipo exploratoria, mediante una encuesta autoaplicada que contenía un conjunto de indicadores que correspondían a competencias establecidas previamente en un perfil de referencia, cuya estructura considera diferentes áreas de desempeño. El perfil fue validado por juicio de expertos, determinando el grado de congruencia a través del coeficiente de Serafine. La construcción de la información se realizó a partir de las repuestas, las que fueron codificadas y representadas en porcentajes y posteriormente analizadas bajo categorías interpretativas. Resultados: La evaluación que se hace considera el porcentaje de egresados que valoró el logro de las competencias por área de desempeño en los niveles más altos. Los resultados obtenidos fueron: Para el área asistencial el 72,9 %, para el área investigación un 59 %, para el área de docencia un 44%, para el área administrativa un 27,1% y para el área personal-social un 91,3 %. Conclusiones: Siguiendo el criterio de categorías interpretativas, el área personal social indica una gran fortaleza de la formación, el área asistencial y el área investigación se encuentran en un nivel satisfactorio y las áreas docencia y administración en niveles medianamente satisfactorio, siendo esta última la gran falencia de la formación. El currículo vigente es adecuado para asegurar el logro de gran número de competencias, pero se debe revisar con el objetivo de

  8. Mejoramiento de imágenes usando funciones de base radial Images improvement using radial basis functions

    Directory of Open Access Journals (Sweden)

    Jaime Alberto Echeverri Arias

    2009-07-01

    Full Text Available La eliminación del ruido impulsivo es un problema clásico del procesado no lineal para el mejoramiento de imágenes y las funciones de base radial de soporte global son útiles para enfrentarlo. Este trabajo presenta una técnica de interpolación que disminuye eficientemente el ruido impulsivo en imágenes, mediante el uso de interpolante obtenido por funciones de base radial en el marco de la investigación enfocada en el desarrollo de un Sistema de recuperación de imágenes de recursos acuáticos amazónicos. Esta técnica primero etiqueta los píxeles de la imagen que son ruidosos y, mediante la interpolación, genera un valor de reconstrucción de dicho píxel usando sus vecinos. Los resultados obtenidos son comparables y muchas veces mejores que otras técnicas ya publicadas y reconocidas. Según el análisis de resultados, se puede aplicar a imágenes con altas tasas de ruido, manteniendo un bajo error de reconstrucción de los píxeles "ruidosos", así como la calidad visual.Global support radial base functions are effective in eliminating impulsive noise in non-linear processing. This paper introduces an interpolation technique which efficiently reduces image impulsive noise by means of an interpolant obtained through radial base functions. These functions have been used in a research project designed to develop a system for the recovery of images of Amazonian aquatic resources. This technique starts with the tagging by interpolation of noisy image pixels. Thus, a value of reconstruction for the noisy pixels is generated using neighboring pixels. The results obtained with this technique have proved comparable and often better than those obtained with previously known techniques. According to results analysis, this technique can be successfully applied on images with high noise levels. The results are low error in noisy pixel reconstruction and better visual quality.

  9. Comparison and physical interpretation of MCNP and TART neutron and γ Monte Carlo shielding calculations for a heavy-ion ICF system

    International Nuclear Information System (INIS)

    Mainardi, E.; Premuda, F.; Lee, E.

    2004-01-01

    Livermore National Laboratory, UCRL-ID-126455, Rev. 1, November, 1997] and MCNP4B [MCNP - A General Monte Carlo N-Particle Transport Code, Version 4B, La-12625-m, March 1997, Los Alamos National Laboratory] for two different configurations of the system is discussed, separating the n and γ contributions, in the light of the physical interpretation of the results in terms of first flight and of scattered neutron fluxes, of primary γ and of secondary γ generated by inelastically scattered or radiatively captured neutrons. The final conclusions indicate some guidelines and suggest possible improvements for the future neutronic shielding design for a HIF facility

  10. An Assessment of the Detection of Highly Enriched Uranium and its Use in an Improvised Nuclear Device using the Monte Carlo Computer Code MCNP-5

    Science.gov (United States)

    Cochran, Thomas

    2007-04-01

    In 2002 and again in 2003, an investigative journalist unit at ABC News transported a 6.8 kilogram metallic slug of depleted uranium (DU) via shipping container from Istanbul, Turkey to Brooklyn, NY and from Jakarta, Indonesia to Long Beach, CA. Targeted inspection of these shipping containers by Department of Homeland Security (DHS) personnel, included the use of gamma-ray imaging, portal monitors and hand-held radiation detectors, did not uncover the hidden DU. Monte Carlo analysis of the gamma-ray intensity and spectrum of a DU slug and one consisting of highly-enriched uranium (HEU) showed that DU was a proper surrogate for testing the ability of DHS to detect the illicit transport of HEU. Our analysis using MCNP-5 illustrated the ease of fully shielding an HEU sample to avoid detection. The assembly of an Improvised Nuclear Device (IND) -- a crude atomic bomb -- from sub-critical pieces of HEU metal was then examined via Monte Carlo criticality calculations. Nuclear explosive yields of such an IND as a function of the speed of assembly of the sub-critical HEU components were derived. A comparison was made between the more rapid assembly of sub-critical pieces of HEU in the ``Little Boy'' (Hiroshima) weapon's gun barrel and gravity assembly (i.e., dropping one sub-critical piece of HEU on another from a specified height). Based on the difficulty of detection of HEU and the straightforward construction of an IND utilizing HEU, current U.S. government policy must be modified to more urgently prioritize elimination of and securing the global inventories of HEU.

  11. Dosimetria comparativa de braquiterapia de próstata com sementes de I-125 e Pd-103 via SISCODES/MCNP

    Directory of Open Access Journals (Sweden)

    Bruno Machado Trindade

    2012-10-01

    Full Text Available OBJETIVO: O presente artigo visa apresentar um estudo dosimétrico comparativo de braquiterapia de próstata com sementes de I-125 e Pd-103. MATERIAIS E MÉTODOS: Um protocolo adotado para ambos os implantes com 148 sementes foi simulado em um fantoma tridimensional heterogêneo de pelve por meio dos códigos SISCODES/MCNP5. Histogramas dose-volume na próstata, bexiga e reto, índices de doses D10, D30, D90, D0,5cc, D2cc e D7cc, e representações de distribuição espacial de dose foram avaliados. RESULTADOS: A atividade inicial de cada semente de I-125, para que D90 seja equivalente à dose de prescrição, foi calculada em 0,42 mCi, e de Pd-103, em 0,94 mCi. A dose máxima na uretra foi 90% e 108% da dose de prescrição para I-125 e Pd-103, respectivamente. A D2cc para I-125 foi 30 Gy no reto e 127 Gy na bexiga, e para Pd-103 foi 29 Gy no reto e 189 Gy na bexiga. A D10 no osso do púbis foi 144 Gy para I-125 e 66 Gy para Pd-103. CONCLUSÃO: Os resultados indicam que os implantes de Pd-103 e I-125 puderam depositar a dose prescrita no volume alvo. Entre os achados, observou-se excessiva exposição de radiação nos ossos da pelve, principalmente no protocolo com I-125.

  12. Development of a computational system for radiotherapic planning with the IMRT technique applied to the MCNP computer code with 3D graphic interface for voxel models

    International Nuclear Information System (INIS)

    Fonseca, Telma Cristina Ferreira

    2009-01-01

    The Intensity Modulated Radiation Therapy - IMRT is an advanced treatment technique used worldwide in oncology medicine branch. On this master proposal was developed a software package for simulating the IMRT protocol, namely SOFT-RT which attachment the research group 'Nucleo de Radiacoes Ionizantes' - NRI at UFMG. The computational system SOFT-RT allows producing the absorbed dose simulation of the radiotherapic treatment through a three-dimensional voxel model of the patient. The SISCODES code, from NRI, research group, helps in producing the voxel model of the interest region from a set of CT or MRI digitalized images. The SOFT-RT allows also the rotation and translation of the model about the coordinate system axis for better visualization of the model and the beam. The SOFT-RT collects and exports the necessary parameters to MCNP code which will carry out the nuclear radiation transport towards the tumor and adjacent healthy tissues for each orientation and position of the beam planning. Through three-dimensional visualization of voxel model of a patient, it is possible to focus on a tumoral region preserving the whole tissues around them. It takes in account where exactly the radiation beam passes through, which tissues are affected and how much dose is applied in both tissues. The Out-module from SOFT-RT imports the results and express the dose response superimposing dose and voxel model in gray scale in a three-dimensional graphic representation. The present master thesis presents the new computational system of radiotherapic treatment - SOFT-RT code which has been developed using the robust and multi-platform C ++ programming language with the OpenGL graphics packages. The Linux operational system was adopted with the goal of running it in an open source platform and free access. Preliminary simulation results for a cerebral tumor case will be reported as well as some dosimetric evaluations. (author)

  13. Determination of the exposure speed of radiation emitted by the linear accelerator, using the code MCNP5 to evaluate the radiotherapy room shields of ABC Hospital

    International Nuclear Information System (INIS)

    Corral B, J. R.

    2015-01-01

    Humans should avoid exposure to radiation, because the consequences are harmful to health. Although there are different emission sources of radiation, generated by medical devices they are usually of great interest, since people who attend hospitals are exposed in one way or another to ionizing radiation. Therefore, is important to conduct studies on radioactive levels that are generated in hospitals, as a result of the use of medical equipment. To determine levels of exposure speed of a radioactive facility there are different methods, including the radiation detector and computational method. This thesis uses the computational method. With the program MCNP5 was determined the speed of the radiation exposure in the radiotherapy room of Cancer Center of ABC Hospital in Mexico City. In the application of computational method, first the thicknesses of the shields were calculated, using variables as: 1) distance from the shield to the source; 2) desired weekly equivalent dose; 3) weekly total dose equivalent emitted by the equipment; 4) occupation and use factors. Once obtained thicknesses, we proceeded to model the bunker using the mentioned program. The program uses the Monte Carlo code to probabilistic ally determine the phenomena of interaction of radiation with the shield, which will be held during the X-ray emission from the linear accelerator. The results of computational analysis were compared with those obtained experimentally with the detection method, for which was required the use of a Geiger-Muller counter and the linear accelerator was programmed with an energy of 19 MV with 500 units monitor positioning the detector in the corresponding boundary. (Author)

  14. Clasificación digital de masas nubosas a partir de imágenes meteorológicas usando algoritmos de aprendizaje de máquina

    Directory of Open Access Journals (Sweden)

    Salomón Einstein Ramírez-Fernández

    2014-01-01

    Full Text Available La identificación exacta de nubes precipitantes es una tarea difícil. En el presente trabajo se aplicaron los algoritmos Máquinas de Soporte Vectorial, Árboles de Decisión y Bosques Aleatorios para discriminar entre nubes precipitantes y nubes no precipitantes, a partir de una imagen meteorológica del satélite GOES-13 que cubre el territorio colombiano. El objetivo del trabajo fue evaluar el desempeño de los algoritmos de aprendizaje de máquina (ML, para la clasificación digital de masas nubosas, en términos de la exactitud temática de la clasificación usando como referencia el algoritmo convencional distancia de Mahalanobis. Los resultados muestran que los algoritmos ML proporcionan una clasificación de masas de nubes más exacta que la obtenida por algoritmos convencionales. La mejor exactitud fue obtenida usando Bosques Aleatorios (RF, con una exactitud temática global de 97%. Adicionalmente, la clasificación obtenida con RF fue comparada pixel a pixel con estimaciones de precipitación de la NASA Tropical Rainfall Measurement Mission (TRMM obteniendo una exactitud global del 94%. De acuerdo con este estudio, los algoritmos ML pueden ser usados para mejorar los actuales métodos de identificación de nubes precipitantes.

  15. Lei da gravitação universal e os satélites: uma abordagem histórico-temática usando multimídia

    Directory of Open Access Journals (Sweden)

    Elvis Vilela Rodrigues

    2012-01-01

    Full Text Available Examina-se, neste artigo, o desenvolvimento de aulas de Física em que, usando-se uma multimídia, a Lei da Gravitação Universal é abordada de forma contextualizada a partir da História da Ciência e de um tema atual (os satélites. O objetivo foi examinar como estudantes de Ensino Médio aceitam e se envolvem nesse estudo. Usando uma abordagem metodológica qualitativa, a multimídia apresenta o que são e como funcionam os satélites artificiais, trazendo uma narrativa histórica desde as ideias de movimento do sistema planetário de Ptolomeu até as de Isaac Newton sobre o movimento dos corpos, culminando na Lei da Gravitação Universal. Entrevistas, observações de sala de aula e documentos produzidos pelos estudantes mostram que as imagens, os filmes e os textos contidos na multimídia enriquecem, de modo significativo, o conteúdo, facilitando o entendimento de conceitos da Física. A contextualização histórico-temática, por sua vez, produz maior envolvimento dos alunos no estudo da Física.

  16. ANÁLISIS TERMODINÁMICO DE UN SISTEMA DE REFRIGERACIÓN SOLAR POR ABSORCIÓN USANDO SOLUCIONES DE MONOMETILAMINA - AGUA PARA LA CONSERVACIÓN DE ALIMENTOS ANÁLISE TERMODINÁMICA DUM SISTEMA DE REFRIGERAÇÃO SOLAR POR ABSORÇÃO USANDO COMO PARELHA MONOMETILAMINA - AGUA PARA A CONSERVA DE ALIMENTOS THERMODYNAMIC ANALYSIS OF A SOLAR ABSORPTION REFRIGERARON SYSTEM USING MONOMETHYLAMINE - WATER SOLUTIONS FOR FOOD STORAGE

    OpenAIRE

    CESAR A. ISAZA; ISAAC PILATOWSKY; ROSEMBERG J. ROMERO; FARID B. CORTÉS

    2010-01-01

    Este trabajo presenta la viabilidad de los sistemas de refrigeración solar por absorción usando soluciones de monometilamina - agua (MMA-A) para aplicaciones en conservación de alimentos en las regiones rurales de Colombia, sin acceso a la red de energía eléctrica. Para suplirlos requerimientos de energía térmica se propone un sistema de calentamiento de agua con energía solar usando colectores de placa y un sistema de respaldo convencional. En este trabajo se determinó el coeficiente de oper...

  17. FENDL2/A-MCNP, FENDL2/A-VITJE and FENDL2/A-VITJFLAT. The processed FENDL-2 neutron activation cross-section data files. Summary documentation

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.

    1997-01-01

    This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)

  18. Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

    Science.gov (United States)

    Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

    2014-02-01

    The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

  19. Optimización del tratamiento de aguas residuales de cultivos de flores usando humedales construidos de flujo subsuperficial horizontal

    Directory of Open Access Journals (Sweden)

    Mónica L. Jaramillo-Gallego

    2016-02-01

    Full Text Available Resumen Objetivo: optimizar un sistema de tratamiento de aguas residuales de cultivos de flores, con el fin de mejorar la eficiencia en la remoción de los contaminantes, usando humedales construidos de flujo subsuperficial-horizontal. Metodología: se realizó un estudio de tipo exploratorio experimental en dos etapas, en la primera se efectuó el acondicionamiento fisicoquímico y biológico del sistema de tratamiento, en la segunda, se llevó a cabo el seguimiento de la remoción de los contaminantes durante nueve meses, para lo cual se monitoreó la demanda química de oxígeno, demanda biológica de oxígeno, sólidos totales, sólidos suspendidos totales, pH y oxígeno disuelto. Resultados: Se logró mejorar la eficiencia del sistema de tratamiento en 7,1% para la Demanda biológica de oxígeno, 4,1% Demanda química de oxígeno, 56,9% sólidos totales y 117,2% solidos suspendidos totales. Conclusión: La concentración de DQO disminuyó con el tratamiento primario (Precipitación y oxidación química y favoreció la eficiencia del sistema de tratamiento secundario, dado que las aguas a tratar tenían valores muy altos de DQO que pueden saturar los humedales con contaminantes persistentes. Se podrían obtener mayores eficiencias, si se logra mejorar el sistema de tratamiento primario. Abstract Objective: to optimize the wastewater treatment system of flower crops in order to improve pollutant removal efficiency, using a horizontal subsurface flow constructed wetland. Methodology: An exploratory experimental study was conducted in two stages; in the first stage the treatment system was conditioned physically, chemically and biologically. In the second stage pollutant removal was monitored for nine months. To achieve this, chemical oxygen demand, biological oxygen demand, total solids, total suspended solids, pH and dissolved oxygen were monitored. Results: It was possible to improve the efficiency of the treatment system in 7.1% for

  20. Radiological protection on interstitial brachytherapy and dose determination and exposure rate of an Ir-192 source through the MCNP-4B; Proteccion radiologica en braquiterapia intersticial y determinacion de la dosis y tasa de exposicion de una fuente de Ir-192 mediante el MCNP-4B

    Energy Technology Data Exchange (ETDEWEB)

    Morales L, M.E. [INEN, Av. Angamos Este 2520- Surquillo, Lima (Peru)

    2006-07-01

    The present work was carried out in the Neurological Sciences Institute having as objective to determine the dose and the rate of exhibition of the sources of Iridium 192, Iodine 125 and Palladium 103; which are used to carry out implant in the Interstitial Brachytherapy according to the TG43. For it we carry out a theoretical calculation, its are defined in the enter file: the geometry, materials of the problem and the radiation source, etc; in the MCNP-4B Monte Carlo code, considering a punctual source and for the dose determination we simulate thermoluminescent dosemeters (TLD): at 5 cm, 50 cm, 100 cm and 200 cm of the source. Our purpose is to analyze the radioprotection measures that should take into account in this Institute in which are carried out brain biopsies using a Micro mar stereotactic mark, and in a near future with the collaboration of a doctor and a cuban physique seeks to be carried out the Interstitial Brachytherapy technique with sources of Ir-192 for patient with tumors like glioblastoma, astrocytoma, etc. (Author)

  1. Employment of MCNP in the study of TLDS 600 and 700 seeking the implementation of radiation beam characterization of BNCT facility at IEA-R1

    International Nuclear Information System (INIS)

    Cavalieri, Tassio Antonio

    2013-01-01

    Boron Neutron Capture Therapy, BNCT, is a bimodal radiotherapy procedure for cancer treatment. Its useful energy comes from a nuclear reaction driven by impinging thermal neutron upon Boron 10 atoms. A BNCT research facility has been constructed in IPEN at the IEA-R1 reactor, to develop studies in this area. One of its prime experimental parameter is the beam dosimetry which is nowadays made by using activation foils, for neutron measurements, and TLD 400, for gamma dosimetry. For mixed field dosimetry, the International Commission on Radiation Units and Measurements, ICRU, recommends the use of pair of detectors with distinct responses to the field components. The TLD 600/ TLD 700 pair meets this criteria, as the amount of 6 Li, a nuclide with high thermal neutron cross section, greatly differs in their composition. This work presents a series of experiments and simulations performed in order to implement the mixed field dosimetry based on the use of TLD 600/TLD 700 pair. It also intended to compare this mixed field dosimetric methodology to the one so far used by the BNCT research group of IPEN. The response of all TLDs were studied under irradiations in different irradiation fields and simulations, underwent by MCNP, were run in order to evaluate the dose contribution from each field component. Series of repeated irradiations under pure gamma field and mixed field neutron/gamma field showed differences in the TLD individual responses which led to the adoption of a Normalization Factor. It has allowed to overcome TLD selection. TLD responses due to different field components and spectra were studied. It has shown to be possible to evaluate the relative gamma/neutron fluxes from the relative responses observed in the two Regions of Interest, ROIs, from TLD 600 and TLD 700. It has also been possible to observe the TLD 700 response to neutron, which leads to a gamma dose overestimation when one follows the ICRU recommended mixed field dosimetric procedure. Dose

  2. Sensitivity analysis of the influence of the medium energy and initial fluence FWHM of electron determining a Bremsstrahlung photon spectrum of a linear accelerator; Analisis de sensibilidad de la influencia de la energia media FWHM de la influencia inicial de electrones en la determinacion de un espectro de fotones Bremsstrahlung de un acelerador lineal

    Energy Technology Data Exchange (ETDEWEB)

    Juste, B.; Miro, R.; Verdu, G.; Diez, S.; Campayo, J. M.

    2012-11-01

    A correct dose calculation in patient under radiotherapy treatments requires and accurate description of the radiation source. The main goal of the present work is to study the effects of initial electron beam characteristics on Monte Carlo calculated absorbed dose distribution for a 6 MeV linac photon beam. To that, we propose a methodology to determine the initial electron fluence before hitting the accelerator target for an Elektra Precis a medical linear accelerator. The method used for the electron radiation source description is based on a Software for Uncertainty and Sensitivity Analysis (SUSA) and Monte Carlo simulations using the MCNP5 transport code. This electron spectrum has been validated by means of comparison of its resulting depth dose curve in a water cube with experimental data being the mean difference below the 1%. (Author)

  3. Sobre el estudio de la función cuadrática y su relación con el área de algunas figuras y su visualización usando Cabri II Plus

    OpenAIRE

    Vera Barrios, Dídimo

    2013-01-01

    Propuesta didáctica en donde se presentan algunas actividades dinámicas usando Cabri II Plus con el fin de relacionar algunos aspectos que tienen que ver con el área de algunas figuras y la función cuadrática.

  4. Improvements in the processing of EFF-2 data for MCNP using NJOY91.38. Final report of subtask NDB-1.2 of the European Fusion Technology Programme

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Gruppelaar, H.; Nierop, D.

    1994-07-01

    The results of a careful check of the solutions as given in NJOY91.38 are presented. It appears that the conversion of DDX data to Kalbach parameters r and a as presented in NJOY91.38 is, in general, not entirely adequate. An improved subroutine was written, which yields a better description of the Kalbach-fit of the DDX data. Furthermore, the conversion of cross section data from centre-of-mass to laboratory system (which was needed for the processing of kerma data) appears to be not necessary anymore, as the HEATR-module of NJOY91.38 seems to operate correctly, also for centre-of-mass data. Problems still remain if DDX data for light isotopes are to be used in MCNP-calculations. This is illustrated using Be-9 from the EFF-2.2 evaluation as a sample case. Recommendations are given in order to solve the remaining problems. A computer code is presented, with which it is possible to create a problem-specific cross section library for use in MCNP-4, in which self-shielding in the unresolved resonance range is taken into account in an approximate form. It is shown, that the effects of neglecting self-shielding in the unresolved resonance range may be substantial in shielding calculations. This is especially relevant for the Fe-56 evaluation in EFF-2.2, in which a very extended unresolved resonance range is present. (orig./GL)

  5. Rediseño y calibración de un instrumento de laboratorío para medir porosidad usando helio

    OpenAIRE

    Ramon Saraguro, Christian F.; Gallegos Orta, Ricardo

    2009-01-01

    El presente trabajo describe el Rediseño y calibración de un instrumento de laboratorio para medir porosidad usando helio, medición que se realiza sobre núcleos muestra (plug) extraídos de estratos de las zonas de interés de pozos de yacimientos petroleros. Instrumento cuya medición la realiza por medio de la expansión de un gas, el cual ha sido implementado en el laboratorio de yacimientos y petrofísica. Su descripción contiene, conceptos básicos, métodos, equipo, obtención y análisis de res...

  6. Análisis de un sistema de almacenamiento de energía térmica usando cloruro de magnesio hexahidratado

    Directory of Open Access Journals (Sweden)

    Andrés Felipe Macía

    2010-01-01

    Full Text Available Se simuló el comportamiento del cloruro de magnesio hexahidratado (MgCl2*6H2O como almacenador de energía térmica, que posee una temperatura de transición de fase de 117ºC; lo que lo convierte en un material con gran potencial en el área de las aplicaciones de mediana temperatura (aplicaciones industriales. Se desarrolló una simulación CFD usando el software FLUENT para describir la fusión/solidificación de la sal hidratada. Se observó el efecto del uso de aletas y las fuerzas boyantes producidas por los efectos gravitacionales.

  7. Calculo y comparacion de la prima de un reaseguro de salud usando el modelo de opciones de Black-Scholes y el modelo actuarial

    Directory of Open Access Journals (Sweden)

    Luis Eduardo Giron

    2015-12-01

    Full Text Available La presente investigación pretende calcular y comparar la prima de un reaseguro  usando el modelo de opciones de Black-Scholes y el modelo clásico actuarial tradicional. El período de análisis va desde enero de 2011 hasta diciembre de 2012. Los resultados obtenidos muestran que el modelo de Black-Scholes, que se utiliza normalmente para valorar opciones financieras, puede ser también usado para la estimación de primas de reaseguros de salud; y que la prima neta estimada a partir de este modelo se aproxima a las establecidas por el método actuarial, excepto cuando el deducible del reaseguro es muy alto (por encima de $200.000.000.

  8. Diagnóstico temprano del Virus Dengue 1 usando RT-PCR y perspectivas para la caracterización molecular de Cepas Autóctonas

    Directory of Open Access Journals (Sweden)

    C Yábar

    1999-01-01

    Full Text Available Un sistema de diagnóstico para la detección temprana del virus Dengue 1 fue llevado a cabo exitosamente usando la reacción en cadena por polimerasa de transcriptasa reversa (RT-PCR, a través de la amplificación de una porción genómica del gen NS1. Los resultados obtenidos, a partir de muestras clínicas, corroboraron los datos de anteriores trabajos de RT-PCR dirigidos hacia la región estructural del virión. Posteriormente el ADNc del virus Dengue, correspondiente a una de las muestras serológicas, fue clonado y secuenciado. La comparación por análisis de secuencia nucleotídica con otras cepas referenciales determinó que la cepa viral correspondía al serotipo 1.

  9. Implementação da Técnica de Dessorção Térmica Programada (TPD) usando Espectrometria de Massa Quadrupolo

    OpenAIRE

    Barreto, Ana Marta Fortunato

    2011-01-01

    Dissertação apresentada na Faculdade de Ciências e Tecnologia da Universidade Nova de Lisboa para obtenção do Grau de Mestre em Engenharia Física O projecto “Implementação da Técnica de Dessorção Térmica Programada (TPD) usando Espectrometria de Massa Quadrupolo”, apresentado nesta Dissertação de Mestrado em Engenharia Física, surgiu como uma proposta de requalificação ao aparelho de análise de superfícies Multitécnica. A dessorção térmica é o fenómeno que permite a separação física de ...

  10. Estudio comparativo de flujo de fluido a través de una placa de orificio usando las ecuaciones de Stokes y de Navier-Stokes

    Directory of Open Access Journals (Sweden)

    Miryam Lucía Guerra-Mazo

    2016-05-01

    Full Text Available Presenta los resultados de la comparación entre las ecuaciones de Stokes y de Navier-Stokes para la simulación del flujo de agua líquida, a condiciones atmosféricas, a través de una placa orificio concéntrica. A partir de los datos experimentales que fueron tomados en el banco de fluidos, se evaluaron las simulaciones de ambas ecuaciones, usando el software libre Freefem++cs, que se basa en el método de los elementos finitos; las variables evaluadas son velocidad y presión en un intervalo de tiempo. Al analizar los resultados obtenidos con las simulaciones y comparar con los datos experimentales se encontró que las ecuaciones de Navier-Stokes representan mejor el sistema que la ecuación de Stokes.

  11. Aproximación no lineal al modelo de overshooting usando redes neuronales multicapa para el tipo de cambio dólar - peso

    Directory of Open Access Journals (Sweden)

    Villamil Jaime

    2009-10-01

    Full Text Available Desde los años setenta muchos trabajos han intentado elaborar una sustentación empírica de algunos modelos
    que ofrecieron una explicación lineal de la dinámica de la tasa de cambio de un país, entre ellos el de Dornbusch. Hasta el momento ninguno ha sido concluyente y la caminata aleatoria es considerada como el mejor modelo al que puede ajustarse. De Grauwe ha mostrado que, con la presencia de relaciones no-lineales y heterogeneidad de expectativas de los especuladores, el tipo de cambio puede tener un comportamiento aparentemente aleatorio, pero con explicación determinista. Este trabajo presenta el modelo de Dornbusch en la versión no lineal propuesta por De Grauwe y Dewachter (1993, y una aproximación usando redes neuronales multicapa aplicadas al caso del dólar/peso (USD/COP.

  12. APROXIMACIÓN NO LINEAL AL MODELO DE OVERSHOOTING USANDO REDES NEURONALES MULTlCAPA PARA EL TIPO DE CAMBIO DÓLAR-PESO

    Directory of Open Access Journals (Sweden)

    Jaime Villamil

    2009-06-01

    Full Text Available Desde los años setenta muchos trabajos han intentado elaborar una sustentación empírica de algunos modelos que ofrecieron una explicación lineal de la dinámica de la tasa de cambio de un país, entre ellos el de Dornbusch. Hasta el momento ninguno ha sido concluyente y la caminata aleatoria es considerada como el mejor modelo al que puede ajustarse. De Grauwe ha mostrado que, con la presencia de relaciones no-lineales y heterogeneidad de expectativas de los especuladores, el tipo de cambio puede tener un comportamiento aparentemente aleatorio, pero con explicación determinista. Este trabajo presenta el modelo de Dornbusch en la versión no lineal propuesta por De Grauwe y Dewachter (1993, y una aproximación usando redes neuronales multicapa aplicadas al caso del dólar/peso (USD/COP.

  13. Avaliação da integridade estrutural do quartzito Itacolomi empregado em monumentos históricos de Ouro Preto sem e com colagem usando diferentes resinas

    OpenAIRE

    Neves,José Henrique; Godefroid,Leonardo Barbosa; Cândido,Luiz Cláudio

    2011-01-01

    A arte da Cantaria usando o quartzito foi utilizada na construção dos monumentos históricos de Ouro Preto/MG. Após mais de 300 anos de existência, parte de algum desses monumentos encontra-se em avançado estágio de deterioração. Na restauração desses monumentos, alguns restauradores empregam a resina poliéster ortoftálica. Nesse trabalho, o quartzito itacolomi não colado e colado com diversas resinas foi submetido a ensaio de resistência em flexão e constatou-se que a resina suportou carga ma...

  14. Predicción de los precios de contratos de electricidad usando una red neuronal con arquitectura dinámica

    Directory of Open Access Journals (Sweden)

    Juan David Velásquez Henao

    2010-04-01

    Full Text Available Los contratos en los mercados liberalizados de electricidad constituyen una herramienta para proteger a los agentes de la volatilidad; en este contexto, los pronósticos de los precios son una entrada clave para la toma de decisiones estratégicas y operativas de los agentes. En este artículo, se pronostican los precios promedios de los contratos despachados en el mercado eléctrico colombiano, usando una red neuronal con arquitectura dinámica conocida como DAN2. El modelo desarrollado es capaz de capturar la dinámica intrínseca de la serie de precios y de pronosticar el precio para el siguiente mes con más precisión que la metodología ARIMA clásica, para horizontes de predicción de 12 y 24 meses

  15. OTIMIZAÇÃO USANDO ALGORITMO GENÉTICO DE UM MODELO DE PROPAGÇÃO BASEADO EM EQUAÇÕES PARABÓLICAS

    Directory of Open Access Journals (Sweden)

    Antonio Carlos Vilanova

    2013-01-01

    Full Text Available Este artigo apresenta uma avaliação metodológica para otimizar parâmetros em um conhecido modelo de propagação de ondas de rádio na troposfera. O modelo de propagação é baseado no Divisor de passos de Fourier para resolver equações parabólicas. Nossa abordagem utiliza algoritmo genético para determinar os valores dos parâmetros que maximize a intensidade de campo em uma determinada posição do observador. Usando algoritmo genético o tempo necessário na busca dos parâmetros ótimos é reduzido significativamente. A avaliação preliminar dos resultados através da simulação mostra que a nossa abordagem é promissora.

  16. Extração do óleo de manjericão usando fluido supercrítico: analise experimental e matemática

    Directory of Open Access Journals (Sweden)

    Nídia Alves de Barros

    2014-01-01

    Full Text Available O óleo essencial de manjericão (Ocimum basilicum L. é valorizado no mercado internacional e amplamente usado nas indústrias de condimentos, cosméticos e medicinais. Entre todos os processos que podem ser aplicados na obtenção do óleo, a extração usando fluido supercrítico (EFS pode ser um método seletivo e eficiente, dependendo das condições operacionais, como temperatura e pressão, que precisam ser otimizadas. O objetivo deste estudo foi comparar a eficiência da extração supercrítica, utilizando planejamento experimental, na extração do óleo de manjericão com os métodos convencionais (hidrodestilação e soxhlet. Para a realização da parte experimental, foi utilizado dióxido de carbono como solvente para a EFS e hexano, para aplicação no soxhlet. Foi realizado um delineamento central composto rotacional (DCCR, aplicando três pressões (100, 200 e 300 bar e três temperaturas (30, 40 e 50ºC. Através da hidrodestilação, obteve-se o menor rendimento (0,26%, usando o Soxhlet, o rendimento foi de 2,39% sendo superior, comparado á EFS, que foi de 0,43%. O DCCR mostrou que, para otimizar o processo, é necessário aumentar a pressão e a temperatura para o alcance de maiores rendimentos. Foi possível constatar que o modelo matemático representou bem o processo de extração, propiciando o "scale-up" deste.

  17. Modelos de pérdida de masa de acero por corrosión atmosférica en Colombia usando inteligencia computacional

    Directory of Open Access Journals (Sweden)

    Esteban Velilla

    2009-01-01

    Full Text Available Con el fin de clasificar la corrosividad de las diferentes atmósferas colombianas, como parte de un proyecto de investigación extenso [1], se expusieron placas de acero al carbono en 21 estaciones distribuidas a lo largo de la infraestructura eléctrica del país (líneas de transmisión y subestaciones. En estas estaciones se midieron entre otros, el tiempo de humectación y la deposición de sulfatos y cloruros durante 12 meses; además, bimensualmente se tomaban placas de acero para medir en laboratorio la pérdida de masa sufrida por estas durante el tiempo de exposición. La clasificación de las 21 estaciones se hizo en 4 grupos, considerando: el tiempo de humectación, contenidos de cloruros y sulfatos, la altura sobre el nivel del mar y el tiempo de exposición de las placas; variables consideradas linealmente independientes según la técnica de descomposición en valores singulares (SVD realizada. El criterio utilizado para la clasificación fue el de similitud de las variables utilizando la norma Euclidiana considerada en la red neuronal no supervisada tipo Kohonen. Adicionalmente, se implementaron modelos para la pérdida de masa del acero para cada uno de los grupos usando redes neuronales (RN tipo Feed-Forward, definiéndose como entradas las variables antes mencionadas y como única salida la pérdida de masa. Complementariamente se presenta una comparación entre el modelo de RN para el grupo 1, con otros modelos obtenidos usando Algoritmos Genéticos (AG y el método Simplex.

  18. Producción de carbones ultralimpios usando flotación burbujeante y lixiviación con ácidos

    Directory of Open Access Journals (Sweden)

    Juan Barraza

    2009-01-01

    Full Text Available Los carbones ultralimpios representan una importante materia prima para la elaboración de productos de alto valor agregado tales como fibra de carbono, electrodos, espumas de carbón, entre otros. En este trabajo, carbones ultralimpios con concentraciones menores a 0,50% de ceniza (p/p, base seca, bs se obtuvieron usando flotación burbujeante en columna y lixiviación con ácidos fluorhídrico (HF y nítrico (HNO3. Los contenidos de ceniza se redujeron desde 19,60 % en los carbones alimentados hasta 8,70 % en los carbones flotados usando tres etapas en serie en una columna de flotación. Al usar lixiviación química con HF 7,53M y HNO3 2,3M, los carbones presentaron contenidos de 1,42% de ceniza, 0,86% de azufre y 2,00% de materia mineral, mientras que cuando se usó un proceso combinado de flotación seguido de lixiviación ácida con HF 19,2M y HNO3 8,12M se obtuvo un carbón con 0,48 % de ceniza 0,71 % de azufre y 0,96 % de materia mineral. Sin embargo, al usar carbón original (no flotado a las mismas concentraciones ácidas utilizadas en el proceso combinado se produjo un carbón ultralimpio de contenido de ceniza 0,33%.

  19. Development of a computational system for radiotherapic planning with the IMRT technique applied to the MCNP computer code with 3D graphic interface for voxel models; Desenvolvimento de um sistema computacional para o planejamento radioterapico com a tecnica IMRT aplicado ao codigo MCNP com interface grafica 3D para modelos de voxel

    Energy Technology Data Exchange (ETDEWEB)

    Fonseca, Telma Cristina Ferreira

    2009-07-01

    The Intensity Modulated Radiation Therapy - IMRT is an advanced treatment technique used worldwide in oncology medicine branch. On this master proposal was developed a software package for simulating the IMRT protocol, namely SOFT-RT which attachment the research group 'Nucleo de Radiacoes Ionizantes' - NRI at UFMG. The computational system SOFT-RT allows producing the absorbed dose simulation of the radiotherapic treatment through a three-dimensional voxel model of the patient. The SISCODES code, from NRI, research group, helps in producing the voxel model of the interest region from a set of CT or MRI digitalized images. The SOFT-RT allows also the rotation and translation of the model about the coordinate system axis for better visualization of the model and the beam. The SOFT-RT collects and exports the necessary parameters to MCNP code which will carry out the nuclear radiation transport towards the tumor and adjacent healthy tissues for each orientation and position of the beam planning. Through three-dimensional visualization of voxel model of a patient, it is possible to focus on a tumoral region preserving the whole tissues around them. It takes in account where exactly the radiation beam passes through, which tissues are affected and how much dose is applied in both tissues. The Out-module from SOFT-RT imports the results and express the dose response superimposing dose and voxel model in gray scale in a three-dimensional graphic representation. The present master thesis presents the new computational system of radiotherapic treatment - SOFT-RT code which has been developed using the robust and multi-platform C{sup ++} programming language with the OpenGL graphics packages. The Linux operational system was adopted with the goal of running it in an open source platform and free access. Preliminary simulation results for a cerebral tumor case will be reported as well as some dosimetric evaluations. (author)

  20. Estimating output fluence with MCNP4 for shaped fields and their comparison with measurements in the EPID system aS1000 for dosimetry 2D in-vivo; Estimacion de la fluencia de salida con MCNP4 para campos conformados y su comparacion con mediciones en el sistema EPID aS1000 para dosimetria in-vivo 2D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez R, B.; Rodriguez P, X.; Sosa, M., E-mail: bhernandez@fisica.ugto.mx [Universidad de Guanajuato, Division de Ciencias e Ingenierias, Loma del Bosque No. 103, 37150 Leon, Guanajuato (Mexico)

    2015-10-15

    Full text: Radiotherapy dosimetry is a fundamental process in quality control of the treatments performed with this technique. Different systems exist to quantify radiation dose in radiotherapy, one of them is the Electronic Portal Imaging Device (EPID), which is widely used in IMRT to measure the output fluence of a radiation field for comparison with a predicted fluence in a planning system. The objective of this work was to simulate a Varian linear accelerator model Clinac i X using the MCNP4 code for obtaining curves of percentage depth dose (Pdd) and open fields dosimetric profiles of 5 x 5, 10 x 10, 20 x 20 and 30 x 30 cm{sup 2}. The simulations were validated by comparing them with measurements made with ionization chamber. Then a mannequin of solid water (30 x 30 x 20 cm{sup 3}) with an open field of 10 x 10 cm{sup 2} was irradiated to measure the output fluence with EPID aS1000 system of Varian. A simulation of the solid water mannequin under the same conditions of irradiation was conducted to estimate the output fluence. Tests of index gamma and percentage differences were calculated to compare that simulated with that measured. In all cases was found that more than 95% of the evaluated points passed the acceptance criteria (ΔD= 1% and ΔS= 1 mm for curves Pdd and profiles, and ΔD= 3% and ΔS= 3 mm for fluence two-dimensional). This paper will contribute to the implementation of in-vivo dosimetry three-dimensional with the EPID system. (Author)

  1. Determination of the exposure speed of radiation emitted by the linear accelerator, using the code MCNP5 to evaluate the radiotherapy room shields of ABC Hospital; Determinacion de la rapidez de exposicion de la radiacion emitida por el acelerador lineal, utilizando el codigo MCNP5, para evaluar los blindajes de la sala de radioterapia del Hospital ABC

    Energy Technology Data Exchange (ETDEWEB)

    Corral B, J. R.

    2015-07-01

    Humans should avoid exposure to radiation, because the consequences are harmful to health. Although there are different emission sources of radiation, generated by medical devices they are usually of great interest, since people who attend hospitals are exposed in one way or another to ionizing radiation. Therefore, is important to conduct studies on radioactive levels that are generated in hospitals, as a result of the use of medical equipment. To determine levels of exposure speed of a radioactive facility there are different methods, including the radiation detector and computational method. This thesis uses the computational method. With the program MCNP5 was determined the speed of the radiation exposure in the radiotherapy room of Cancer Center of ABC Hospital in Mexico City. In the application of computational method, first the thicknesses of the shields were calculated, using variables as: 1) distance from the shield to the source; 2) desired weekly equivalent dose; 3) weekly total dose equivalent emitted by the equipment; 4) occupation and use factors. Once obtained thicknesses, we proceeded to model the bunker using the mentioned program. The program uses the Monte Carlo code to probabilistic ally determine the phenomena of interaction of radiation with the shield, which will be held during the X-ray emission from the linear accelerator. The results of computational analysis were compared with those obtained experimentally with the detection method, for which was required the use of a Geiger-Muller counter and the linear accelerator was programmed with an energy of 19 MV with 500 units monitor positioning the detector in the corresponding boundary. (Author)

  2. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    Directory of Open Access Journals (Sweden)

    Hammam Oktajianto

    2014-12-01

    Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

  3. Calculation of the reaction rate and response matrix of a neutron spectroscopy consisting of a water sphere of variable diameter and BF3 detector using the MCNP5-beta code

    International Nuclear Information System (INIS)

    Albashir, K.; Nahili, M.; Al-Zawahera, S.

    2015-01-01

    The MCNP5-beta code was used to calculate the reaction rate 1 0B(n,α) 7 Li and the neutron energy response matrix of a neutron spectrometer consisting of a water sphere with variable diameter and detector BF 3 using point and disk neutron sources 2 41Am-Be. The reaction rate and the response matrix of disk neutron source shows higher value than these obtained from the point neutron source. The response of the matrix disk neutron source in the energy range from 4.14x10 - 7 to 11.09 MeV show a maximum value for sphere of 12 inch diameter, where the response with point neutron source stile increasing calculated value in this condition .The calculated values of neutron energy responses for a disk neutron source agreed well with published results. (author)

  4. Monitoreo de cambios en la densidad de cobertura forestal en bosque templado usando fotografías aéreas digitales de alta resolución

    Directory of Open Access Journals (Sweden)

    José López García

    2016-08-01

    Full Text Available Se utilizaron series multitemporales de fotografías aéreas digitales de alta resolución de pequeño formato para evaluar los cambios en la densidad de cobertura forestal en un bosque templado. Una combinación de técnicas convencionales y adaptadas de fotogrametría y fotointerpretación fueron utilizadas para establecer un método específico de evaluación.  Este método ha sido probado en un periodo de doce años (1999-2011 en la zona núcleo de la Reserva de la Biósfera Mariposa Monarca, localizada en los estados de México y Michoacán, en México, usando mosaicos ortorectificados como mapas base para evaluar cambios bienales. Las imágenes fueron fotointerpretadas de manera tradicional marcando los cambios sobre acetatos, colocados sobre las imágenes impresas, creando así nuevos polígonos. Estos fueron transferidos directamente de los acetatos al ortomosaico a través de la pantalla de la computadora, usando al menos tres puntos de control, cumpliendo así con el principio de triangulación radial. El bosque fue separado en las siguientes clases de cobertura forestal: cerrada, semi-cerrada, semi-abierta, abierta y deforestada. La evaluación en la exactitud en la clasificación de densidad de cobertura fue estimada a través de muestreos en campo, empleando matrices de confusión, siendo del 95%. A partir de 2003, este método ha sido utilizado para determinar el pago por servicios ambientales. Dicho pago, junto con una gran interacción con las comunidades, se ha traducido en una reducción en la degradación forestal y la deforestación en la zona núcleo de la Reserva.

  5. MODELADO DEL PRECIO SPOT DE LA ELECTRICIDAD EN BRASIL USANDO UNA RED NEURONAL AUTORREGRESIVA ELECTRICITY SPOT PRICE MODELLING IN BRASIL USING AN AUTOREGRESSIVE NEURAL NETWORK

    Directory of Open Access Journals (Sweden)

    Juan D Velásquez

    2008-12-01

    Full Text Available Una red neuronal autorregresiva es estimada para el precio mensual brasileño de corto plazo de la electricidad, la cual describe mejor la dinámica de los precios que un modelo lineal autorregresivo y que un perceptrón multicapa clásico que usan las mismas entradas y neuronas en la capa oculta. El modelo propuesto es especificado usando un procedimiento estadístico basado en el contraste del radio de verosimilitud. El modelo pasa una batería de pruebas de diagnóstico. El procedimiento de especificación propuesto permite seleccionar el número de unidades en la capa oculta y las entradas a la red neuronal, usando pruebas estadísticas que tienen en cuenta la cantidad de los datos y el ajuste del modelo a la serie de precios. La especificación del modelo final demuestra que el precio para el próximo mes es una función no lineal del precio actual, de la energía afluente actual y de la energía almacenada en el embalse equivalente en el mes actual y dos meses atrás.An autoregressive neural network model is estimated for the monthly Brazilian electricity spot price, which describes the prices dynamics better than a linear autoregressive model and a classical multilayer perceptron using the same input and neurons in the hidden layer. The proposed model is specified using a statistical procedure based on a likelihood ratio test. The model passes a battery of diagnostic tests. The proposed specification procedure allows us to select the number of units in hidden layer and the inputs to the neural network based on statistical tests, taking into account the number of data and the model fitting to the price time series. The final model specification demonstrates that the price for the next month is a nonlinear function of the current price, the current energy inflow, and the energy saved in the equivalent reservoir in the current month and two months ago.

  6. A comparison of MCNP6-1.0 and GEANT 4-10.1 when evaluating the neutron output of a complex real world nuclear environment: The thermal neutron facility at the Tri Universities Meson facility

    Energy Technology Data Exchange (ETDEWEB)

    Monk, S.D., E-mail: s.monk@lancaster.ac.uk [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom); Shippen, B.A. [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom); Colling, B.R. [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom); Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Cheneler, D.; Al Hamrashdi, H.; Alton, T. [Department of Engineering, Lancaster University, Lancaster LA1 4YW (United Kingdom)

    2017-05-15

    Highlights: • Comparison of the use of MCNP6 and GEANT4 Monte Carlo software when large distances and thicknesses are considered. • The Thermal Neutron Facility (TNF) at TRIUMF used as an example real life example location. • The effects of water, aluminium, iron and lead considered over various thicknesses up to 3 m. - Abstract: A comparison of the Monte Carlo based simulation codes MCNP6-1.0 and GEANT4-10.1 as used for modelling large scale structures is presented here. The high-energy neutron field at the Tri Universities Meson Facility (TRIUMF) in Vancouver, British Columbia is the structure modelled in this work. Work with the emphasis on the modelling of the facility and comparing with experimental results has been published previously, whereas this work is focussed on comparing the performance of the codes over relatively high depths of material rather than the accuracy of the results themselves in comparison to experimental data. Comparisons of three different locations within the neutron facility are modelled and presented using both codes as well as analysis of the transport of typical neutrons fields through large blocks of iron, water, lead and aluminium in order to determine where any deviations are likely to have occurred. Results indicate that over short distances, results from the two codes are in broad agreement – although over greater distances and within more complex geometries, deviation increases dramatically. The conclusions reached are that it is likely the deviations between the codes is caused by both the compounding effect of slight differences between the cross section files used by the two codes to determine the neutron transport through iron, and differences in the processes used by both codes.

  7. Primera edad U-Pb en circón usando LA-ICP-MS de un dique traquiandesítico emplazado en el granito tipo-A Los Árboles, Sierras Pampeanas Orientales

    OpenAIRE

    Juan A Dahlquist; Pablo H Alasino

    2012-01-01

    La edad de cristalización de diques subvolcánicos emplazados en plutones graníticos carboníferos, Sierras Pampeanas Orientales, permanece incierta. La primera edad U-Pb en circón usando LA-ICP-MS obtenida en un dique que intruye al plutón Los Árboles (sierra de Fiambalá), revela una edad de cristalización de 311 ± 3 Ma.

  8. Employment of MCNP in the study of TLDS 600 and 700 seeking the implementation of radiation beam characterization of BNCT facility at IEA-R1; Emprego do MCNP no estudo dos TLDS 600 e 700 visando a implementacao da caracterizacao do feixe de irradiacao da instalacao de BNCT do IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Cavalieri, Tassio Antonio

    2013-07-01

    Boron Neutron Capture Therapy, BNCT, is a bimodal radiotherapy procedure for cancer treatment. Its useful energy comes from a nuclear reaction driven by impinging thermal neutron upon Boron 10 atoms. A BNCT research facility has been constructed in IPEN at the IEA-R1 reactor, to develop studies in this area. One of its prime experimental parameter is the beam dosimetry which is nowadays made by using activation foils, for neutron measurements, and TLD 400, for gamma dosimetry. For mixed field dosimetry, the International Commission on Radiation Units and Measurements, ICRU, recommends the use of pair of detectors with distinct responses to the field components. The TLD 600/ TLD 700 pair meets this criteria, as the amount of {sup 6}Li, a nuclide with high thermal neutron cross section, greatly differs in their composition. This work presents a series of experiments and simulations performed in order to implement the mixed field dosimetry based on the use of TLD 600/TLD 700 pair. It also intended to compare this mixed field dosimetric methodology to the one so far used by the BNCT research group of IPEN. The response of all TLDs were studied under irradiations in different irradiation fields and simulations, underwent by MCNP, were run in order to evaluate the dose contribution from each field component. Series of repeated irradiations under pure gamma field and mixed field neutron/gamma field showed differences in the TLD individual responses which led to the adoption of a Normalization Factor. It has allowed to overcome TLD selection. TLD responses due to different field components and spectra were studied. It has shown to be possible to evaluate the relative gamma/neutron fluxes from the relative responses observed in the two Regions of Interest, ROIs, from TLD 600 and TLD 700. It has also been possible to observe the TLD 700 response to neutron, which leads to a gamma dose overestimation when one follows the ICRU recommended mixed field dosimetric procedure. Dose

  9. DelPapa - Aplicativo computacional para a análise de dados de experimentos no delineamento blocos ao acaso, usando o método Papadakis

    Directory of Open Access Journals (Sweden)

    Lindolfo Storck

    2015-05-01

    Full Text Available O aplicativo computacional para a análise de dados de experimentos executados no delineamento blocos ao acaso, por meio do método usual e de Papadakis, foi desenvolvido em sua primeira versão (não publicada, na linguagem de programação Pascal. Considerando que o método de Papadakis foi eficiente para as principais culturas agrícolas (milho, soja, feijão e trigo e, para tornar o aplicativo mais amigável, a versão em Pascal foi reprogramada em Java, cuja denominação é DelPapa. Este aplicativo realiza a análise de variância segundo o delineamento blocos ao acaso pelo método usual (estima parâmetros genéticos, medidas de qualidade experimental e testes dos pressupostos da análise de variância e pelo método de Papadakis. Usando as médias ajustadas pela covariável (média dos erros das parcelas vizinhas, também realiza o teste Scott e Knott (P=0,05 para agrupar os tratamentos.

  10. Soldabilidad de Polietileno de Alta Densidad Usando Soldadura por Fricción-agitación con una Herramienta de Hombro no Rotacional

    Directory of Open Access Journals (Sweden)

    Yorledis Macea Romero

    Full Text Available Resumen: En este trabajo se estudió la soldabilidad de polietileno de alta densidad (PEAD mediante el proceso de soldadura por fricción agitación (SFA, usando dos herramientas diferentes, una convencional y otra con hombro no rotacional, denominada no convencional. La evaluación comparativa fue realizada bajo combinaciones de los parámetros de velocidad de rotación (vW y velocidad de avance (vA ajustados para cada herramienta. La soldabilidad fue valorada con base en inspección visual, la presencia de discontinuidades mediante estereoscopia y microscopia confocal y la dureza en las juntas soldadas. Como resultado se obtuvo que la herramienta no convencional permitió obtener juntas soldadas libres de discontinuidades y con excelente apariencia externa gracias al menor calor aportado. La dureza en la zona agitada fue menor a la del material base, mostrando comportamientos diferenciados en los lados de avance y retroceso en función de los parámetros de proceso.

  11. Identification and control of electromechanical systems by jeans of a neural multimodel; Identificacion y control de sistemas electromecanicos usando un multimodelo neuronal

    Energy Technology Data Exchange (ETDEWEB)

    Baruch, Ieroham [Instituto Politecnico Nacional, Mexico D.F. (Mexico); Beltran Lopez, Rafael [Becario del CONACY-Mexico, D.F. (Mexico); Gortcheva, Elena [Instituto Politecnico Nacional, Mexico D.F. (Mexico)

    2004-07-15

    A Recurrent Trainable Neural Network (Rant) and dynamic backpropagation learning are implemented in the control of complex nonlinear plants. In the present paper, a neural multimodel, composed by two Rants, is used. The control schemes proposed are an indirect and a direct adaptive trajectory tracking control, using states and parameters, issued by a neural multimodel identifier. Both control schemes are applied to control a continuous-time model of electromechanical system with friction and backlash, obtaining a good simulation results, confirmed also by convergent experimental results, using a DC motor. [Spanish] Una red neuronal recurrente entrenable (RNRE) y un algoritmo de retropropagacion dinamica como metodo de aparendizaje, son implementados para control de plantas no lineales complejas. En el presente trabajo se usa un multimodelo neuronal, el cual esta compuesto de dos RNRE. Los esquemas de control neuronal propuestos son: control indirecto y directo adaptable, usando los estados y los parametros proporcionados por un multimodelo de identificacion. Ambos sistemas de control son aplicados a un modelo continuo de un sistema electromecanico con friccion y efecto backlash en la salida, obteniendo buenos resultados en la simulacion y tambien confirmados con resultados experimentales convergentes, obtenidos con un motor de CD.

  12. Cinética de adsorción de Cr (VI usando biomasas residuales modificadas químicamente en sistemas por lotes y continuo

    Directory of Open Access Journals (Sweden)

    Candelaria Tejada Tovar

    2015-06-01

    Full Text Available Se estudia la adsorción de Cr (VI a partir de biomasa residual: bagazo de palma y cáscaras de ñame, además se estudia la modificación con ácido cítrico de las biomasas. La determinación del metal en solución fue llevada a cabo usando el método de la 1,5-difenilcarbazida. Del análisis FTIR se encontró que los grupos hidroxilo y carbonilo presentes en los adsorbentes son los de mayor contribución al proceso de remoción. Además, la modificación mejora la eficiencia del metal de acuerdo con la isoterma de Langmuir de 13 a 4mg/g para el bagazo de palma, y de 22 a 26mg/g para las cáscaras de ñame. Se establece además una mejora en el proceso al trabajar las biomasas en sistema continuo. De las condiciones óptimas de adsorción se determinó que el pH de 2 y el tamaño de partícula de 1mm son las que más favorecen el proceso, y el modelo de Elovich el que mejor lo describe.

  13. Análisis del tratamiento ideal usando baños termotratados para la separación de cal de los residuos de descarne en curtiembres

    Directory of Open Access Journals (Sweden)

    Yelitza Aguas Mendoza

    2016-07-01

    Full Text Available Se analizó el tratamiento ideal usando baños termotratados para la separación de cal de los residuos de descarne, del proceso de curtición semiartesanal desarrollado en Sampués, Departamento de Sucre (Colombia. Para su desarrollo se seleccionó una muestra de carnaza representativa, homogénea y con menos tiempo de almacenamiento. Dicha muestra se redujo al tamaño de 1 cm2 y se conservó refrigerada para posteriormente realizar los análisis de grasa, pH y presencia de cal. Los baños termotratados de desencalado se realizaron a tres temperaturas diferentes (26, 30 y 35 ºC. Se utilizó como agente desencalante el acido sulfúrico en tres concentraciones (3, 2 y 1 N. Además se usó un sistema de agitación simulando el bombo en proceso industrial. Se llevaron a cabo nueve pruebas con tres repeticiones para mayor confiabilidad de los datos, en donde se determinó la correlación existente entre las variables independientes, temperatura y concentración, sobre la cal impregnada y los resultados de grasa en cada una de las muestras en un tiempo de 4 y 8 h.

  14. Discriminación de coberturas del suelo usando datos espectrales multi-angulares del sensor polder-1: alcances y limitaciones

    Directory of Open Access Journals (Sweden)

    Fernando Paz Pellat

    2016-04-01

    Full Text Available La información espectral multi-angular (visión de un píxel desde diferentes ángulos de visión y con ángulos de iluminación solar diferentes obtenida de sensores remotos tiene potencial para una discriminación adecuada de clases de coberturas del suelo. De acuerdo con varios esfuerzos realizados para poder analizar la capacidad de discriminación de las clases de cobertura del suelo, se introduce un marco teórico-conceptual para el análisis de la información espectral, angular y temporal (tamaño de píxel fijo. En este trabajo se explora el uso del sensor POLDER-1. La base de datos fue analizada ajustando un modelo de la función de distribución bidireccional de las reflectancias (BRDF en las bandas espectrales disponibles, para diferentes clases de cobertura del suelo del sistema GLC2000. Los resultados experimentales muestran adecuados ajustes a nivel de píxeles y datos diarios. Con los parámetros ajustados del modelo de la BRDF se analizó el potencial de discriminación usando espacios espectrales de las bandas de la región del rojo e infrarrojo cercano, utilizando diferentes resoluciones temporales y espaciales (agrupación de píxeles. Los resultados mostraron alta confusión (traslapes de posición en espacios espectrales, detectándose limitaciones de dichos enfoques para el caso de confusiones debidas a mezclas de clases o causadas por la dinámica temporal de las mismas. Al final se define un esquema para aproximar la clasificación de la vegetación al acoplar la información disponible en los sensores ópticos y las clases que pueden ser discriminadas.

  15. Infecção cirúrgica em colecistectomia videolaparoscópica usando ácido peracético como esterilizante dos instrumentais

    Directory of Open Access Journals (Sweden)

    Edluza Maria Viana Bezerra de Melo

    Full Text Available OBJETIVO: Determinar a frequência de infecção de sítio cirúrgico em pacientes submetidos à colecistectomia videolaparoscópica usando o ácido peracético como esterilizante. MÉTODOS: Foi realizado estudo retrospectivo descritivo do tipo coorte transversal. O ácido peracético foi usado para esterilização seguindo protocolo preconizado pelo fabricante. Foram observados os critérios e indicadores de processo e estrutura para prevenção de infecção de sítio cirúrgico no pré e intraoperatório. Para a vigilância epidemiológica, consultas ambulatoriais eram agendadas para o 15º e entre o 30º e 45º dias após a alta. RESULTADOS: Entre 247 pacientes foram diagnosticados dois casos de infecção de sítio cirúrgico (0,8%. Um paciente reinternou para antibioticoterapia sistêmica e punção percutânea; no outro, a infecção foi superficial e acompanhada ambulatorialmente. CONCLUSÃO: Eticamente não é permitida a realização de um estudo prospectivo pelo fato do ácido peracético ter sido proibido para a esterilização de instrumentais que penetrem em órgãos e cavidades; contudo, estes resultados encorajam estudo prospectivo caso-controle, comparando o uso dele (controle histórico com a esterilização por óxido de etileno.

  16. Evaluación de la exactitud posicional vertical de una nube de puntos topográficos Lidar usando topografía convencional como referencia

    Directory of Open Access Journals (Sweden)

    Wilver Enrique Salinas Castillo

    2013-11-01

    Full Text Available La exactitud vertical de datos lidar es normalmente establecida por proveedores comerciales en un EMC máximo de 0.150 m. Sin embargo, los resultados de evaluaciones de exactitud en las que se han utilizado datos de campo por lo menos tres veces más exactos que los datos lidar, sugieren que dicha exactitud se observa solo cuando la densidad de los datos lidar es mayor a un punto sobre el terreno por metro cuadrado. Desafortunadamente, el número de estos estudios es limitado y se requiere de la elaboración de otras evaluaciones que confirmen dicha hipótesis. En este estudio se evaluó la exactitud vertical de una nube de puntos topográficos lidar de una densidad de 1.02 puntos sobre el terreno por metro cuadrado, usando como referencia datos recolectados en campo mediante una estación total. Los resultados coinciden con aquéllos de estudios previos, por lo que se sugiere establecer la EPV de la nube de puntos topográficos lidar en 0.200 m para terreno mixto y de cambios constantes y en 0.150 m para terreno con cambios topográficos gentiles. Sin embargo, la EPV no refleja la magnitud del 95% de los errores bajo la presencia de errores sistemáticos, por lo cual es necesario incluir el percentil 95 de los errores en la documentación de datos lidar

  17. Estimación de parámetros genéticos para características de crecimiento en borregos Katahdin usando diferentes modelos

    Directory of Open Access Journals (Sweden)

    Coralia Inés V. Manzanilla Pech

    2012-01-01

    Full Text Available Se estimaron parámetros genéticos para características de crecimiento en corderos Katahdin, usando seis variantes del modelo animal. Se usó información de pesos al nacimiento (BW; n= 13,099, al destete ajustado a 75 d (WW; n= 11,509 y posdestete ajustado a 120 d (AW; n= 6,886 tomada durante 7 años (2004-2010 en 20 estados de la República Mexicana. Los análisis se hicieron ignorando o incluyendo efectos maternos. El modelo más sencillo incluyó el efecto genético aditivo directo como el único efecto aleatorio. El modelo más completo incluyó los efectos genéticos directo y materno, la covarianza entre ellos, y el efecto del ambiente permanente materno. Para seleccionar el mejor modelo se usó la prueba de razón de verosimilitud. Cuando los efectos maternos no fueron incluidos en el modelo, los estimadores de la heredabilidad directa y de la varianza genética directa resultaron sobreestimados. Las heredabilidades directas con el mejor modelo fueron 0.18 ± 0.03, 0.30 ± 0.04 y 0.20 ± 0.05 para BW, WW y AW, respectivamente. Las heredabilidades maternas también variaron dependiendo del modelo, de 0.05 a 0.23, 0.00 a 0.12, y 0.09 a 0.25 para BW, WW y AW. El ignorar los efectos maternos en el modelo resultaría en una evaluación genética equivocada para las características de crecimiento en borregos Katahdin.

  18. Control Activo de Vibraciones en un Rotor Tipo Jeffcott con Velocidad Variable Usando una Suspensión Electromecánica

    Directory of Open Access Journals (Sweden)

    F. Beltrán-Carbajal

    2014-07-01

    Full Text Available Resumen: En este trabajo se presenta un esquema de balanceo activo para un rotor tipo Jeffcott con velocidad variable, usando una suspensión con actuadores electromecánicos lineales. Se propone un esquema de estimación de las señales de perturbación que se inducen por la excentricidad presente en el sistema rotor-chumacera y una ley de control del desbalance que combina tareas de seguimiento de trayectorias para el perfil de velocidades en el rotor. Las señales de perturbación se estiman adecuadamente y el control de la velocidad se realiza en forma robusta, dando como resultado un desempeño eficiente del esquema de control activo para la supresión de vibraciones, con amplitud y frecuencia variables, generadas por el inherente desbalance desconocido en elrotor. Abstract: An active balancing scheme for a variable rotor speed Jeffcott-like rotor, using a suspension with linear electromechanical actuators, is presented. In addition, an estimation scheme for the perturbation signals induced by the inherent eccentricity on the rotating mechanical system, and a control law synthesized for simultaneous tracking tasks on the rotor speed are proposed. Some simulation results show the fast and efficient performance of the active vibration control scheme for good suppression of variable amplitude and frequency harmonic vibrations associated to the unbalance, as well as an effective estimation of the perturbation signals and robustness of the rotor speed controller. The proposed methodology can be applied for on-line monitoring and fault detection quite common in rotating machinery. Palabras clave: Control activo de vibraciones, Rotor tipo Jeffcott, Rechazo de perturbaciones., Keywords: Active vibration control, Jeffcott-like rotor, Disturbance rejection.

  19. Variabilidad genética de Aedes aegypti en algunas áreas del Perú usando Single Stranded Conformational Polymorphism (SSCP

    Directory of Open Access Journals (Sweden)

    Nélida Leiva G

    2004-07-01

    Full Text Available Aedes aegypti es el vector responsable de la transmisión del virus del dengue, su distribución geográfica se ha ampliado rápidamente debido principalmente a la intervención de los seres humanos. Objetivo: Analizar la variabilidad genética de este mosquito mediante la comparación del Segundo Espaciador Transcrito Interno (ITS 2 perteneciente al ADN ribosomal (rADN. Materiales y Métodos: Se analizaron muestras de ocho localidades (Jaén, Tingo María, Iquitos, Lambayeque, el distrito de El Rimac, Sullana y Zarumilla y uno de la provincia de Huaquillas (Ecuador. El análisis de la variabilidad se determinó usando la técnica conocida como SSCP (Single Stranded Conformation Polymorphism. Resultados: El estudio muestra que existe variabilidad genética entre las poblaciones analizadas, principalmente entre las muestras localizadas en la costa del Perú (Zarumilla, El Rímac, Sullana y Huaquillas y las muestras del nororiente (Tingo María, Iquitos, Jaén y Lambayeque Conclusión: Se determinaron dos variantes genéticas entre las poblaciones de Aedes aegypti: Costeña y Nororiental, que probablemente provienen de dos ancestros diferentes y cuyo ancestro común sufrió de aislamiento por distancia. Se observó que no existe relación entre las distancias genéticas y las distancias geográficas indicando que la migración de estas poblaciones es el resultado de la intervención de los seres humanos que diseminan al vector y no por la migración activa del mosquito. Se plantea el papel de la Cordillera de los Andes en la migración y separación de las poblaciones de Aedes.

  20. Compósitos polímero-madeira preparados por polimerização in situ de metil metacrilato usando aditivos bifuncionais

    Directory of Open Access Journals (Sweden)

    Bruno Dufau Mattos

    2015-12-01

    Full Text Available Resumo O presente trabalho teve por objetivo a confecção de compósitos polímero-madeira por meio de polimerização in situ de metil metacrilato (MMA, utilizando ácido metacrílico (MAA e glicidil metacrilato (GMA como agentes de ligação e reticulação. Amostras de madeira de guapuruvu foram impregnadas em um sistema de vácuo e pressão e polimerizadas em estufa a 90°C por 10h, usando 1,5% de peroxido de benzoíla como catalisador. Os compósitos foram caracterizados por meio de testes de absorção de água e estabilidade dimensional, molhabilidade, ATR-IR, TGA, MEV e WPG. Os espectros de ATR-IR mostraram incrementos nas bandas a 1746, 1460, e 1145 cm–1, referentes as estruturas químicas dos polímeros dentro da madeira, confirmado posteriormente pelas imagens de MEV. A termogravimetria apontou reações químicas entre os copolímeros e a parede celular da madeira nos compósitos com GMA e MAA. Os compósitos preparados com MMA apresentaram incrementos acima de 50% nas propriedades higroscópicas e de estabilidade dimensional, entretanto a adição de GMA e MAA resultou em maiores incrementos nas mesmas propriedades, entre 66-90%.

  1. Detección de fallas en TFMs usando análisis tiempo-frecuencia; TFM Failure Detection Using Frequence, Time, Analysis

    Directory of Open Access Journals (Sweden)

    Osmel Reyes Vaillant

    2011-02-01

    Full Text Available Las fallas de rodamiento en los rotores medidores de flujo de tipo turbina, no tienen una clara manifestación enlas representaciones espectrales o temporales de la serie de tiempo de la señal de salida de estos medidores.Los escalogramas obtenidos de la señal de estos medidores en condiciones de falla usando la transformadacontinua de wavelet (CWT, han revelado diferencias visuales entre estas y las obtenidas en buenas condiciones.El análisis de singularidades de la matriz de coeficientes de correlación C(b,a para los diferentes segmentos desus vectores, basado en el cálculo de los exponentes de Lipschitz (criterio de continuidad de Hölder, mostró uncomportamientos singulares asociados a la existencia de transientes locales causados por fallas en losrodamientos. Los resultados del empleo del análisis tiempo-frecuencia para la detección de estas fallas sonpresentados en este artículo.  Rotor ball-bearing faults manifestation on turbine flow meters do not appear clearly either in time or spectraldomain representation of its output signal. Scalograms obtained from turbine flow meter output signal on faultyconditions, using continuous Wavelet transform (CWT revealed visual differences from the ones obtained ingood conditions. Singularity analysis of the correlations coefficient matrix C(b,a for different segments of thesevectors, based upon Lipschitz exponent calculation (Hölder continuity criterion, showed local singular behaviorassociated to high speed transients caused by ball-bearing faults. Results from employing time-frequency signalanalysis methods to detect this fault are presented in this article.

  2. Pesquisa de infecção micobacteriana em hemoculturas de doentes com SIDA, usando o sistema BACTEC 460 TB

    Directory of Open Access Journals (Sweden)

    Emília Prieto

    1996-11-01

    Full Text Available RESUMO: Nos doentes em estndios terminais da infecção por VIH (Virus da Imunodeficiência Humana são comuns as infecções a micobactérias. O objectivo deste estudo foi avaliar a prevalência de disseminação micobacteriana nestes doentes, por hemocultura e/ou mielocultura, usando o meio BACTEC 13A® para o sistema radiométrico BACTEC 460 TB, indicado para o isolamento directo de micobactérias a partir de amostras de sangue. Procedemos a hemoculturas e/ou mieloculturas em 811 doentes, obtendo-se 7,3% de isolamentos numa única amostra por doente. Determinou-se uma prevalência de 5% de lulare causadores de infecção com disseminação hematogénea. De entre os isolados de Mycobacterium tuberculosis 1,4% de Mycobacterium avium-intracel-lalare causadores de infecção com disseminação hematogénea. De entre os isolados de Mycobacterium tuberculosis 31, 4% destes apresentam-se como estirpes multir resistentes. ABSTRACT: This study evaluates the prevalency of hematogenous dissemination of mycobacteria in 811 AIDS patients. One blood or bone marrow sample was collected for direct inoculation of the BACTEC 13A radiometric broth (Becton Dickinson Diagnostic Instrument Systems, Towson, USA. There were 7,3%. positive cultures (59 patients with 11 (1,4% Mycobacterium avium-intracellulare isolates and 39 (5% Mycobacterium tuberculosis isolates from which 11 (31,4% were multidrug-resistant strains. Palovras-chave: Micobactérias, Hemocultura, Detecção-radiométrica, SIDA, Key-words: Mycobacteria, Blood culture, Rodiometric-detection, AIDS

  3. Adsorção de Cd 2+, Ni 2+ e Zn 2+ em soluções aquosas usando anidrita e lama vermelha

    Directory of Open Access Journals (Sweden)

    Fabiano Tomazini da Conceição

    Full Text Available RESUMO Vários minerais e resíduos industriais têm sido estudados para uso como adsorvente, entre eles a lama vermelha e a anidrita. A lama vermelha é um resíduo insolúvel gerado em grande quantidade durante o processamento da bauxita. A anidrita é um sulfato de cálcio (CaSO4 cristalizado sob a forma rômbica e usada como matéria-prima na indústria. Nesta investigação, a capacidade de adsorção de Cd2+, Ni2+ e Zn2+ pela anidrita e pela lama vermelha foi avaliada usando isotermas de adsorção de Langmuir e Freundlich. Os materiais empregados apresentaram adsorção ≥75±1% para todos os metais em soluções aquosas com concentração de 0,5 mmol.25 mL-1. As isotermas baseadas no modelo de Langmuir foram as mais apropriadas para descrever o fenômeno de remoção de Cd2+, Ni2+ e Zn2+ para a anidrita e a lama vermelha, com valores de capacidade máxima de adsorção de 0,47 e 0,51 mmol.g-1 para o Cd2+, 1,18 e 1,56 mmol.g-1 para o Ni2+ e 0,84 e 1,47 mmol.g-1 para o Zn2+, respectivamente. Esses valores foram superiores a outros valores exibidos por materiais empregados como adsorventes descritos em estudos prévios.

  4. MO-FG-CAMPUS-TeP3-02: Benchmarks of a Proton Relative Biological Effectiveness (RBE) Model for DNA Double Strand Break (DSB) Induction in the FLUKA, MCNP, TOPAS, and RayStation™ Treatment Planning System

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, R [University of Washington, Seattle, WA (United States); Streitmatter, S [University of Utah Hospitals, Salt Lake City, UT (United States); Traneus, E [RAYSEARCH LABORATORIES AB, Stockholm (Sweden); Moskvin, V [St. Jude Children’s Hospital, Memphis, TN (United States); Schuemann, J [Massachusetts General Hospital, Boston, MA (United States)

    2016-06-15

    Purpose: Validate implementation of a published RBE model for DSB induction (RBEDSB) in several general purpose Monte Carlo (MC) code systems and the RayStation™ treatment planning system (TPS). For protons and other light ions, DSB induction is a critical initiating molecular event that correlates well with the RBE for cell survival. Methods: An efficient algorithm to incorporate information on proton and light ion RBEDSB from the independently tested Monte Carlo Damage Simulation (MCDS) has now been integrated into MCNP (Stewart et al. PMB 60, 8249–8274, 2015), FLUKA, TOPAS and a research build of the RayStation™ TPS. To cross-validate the RBEDSB model implementation LET distributions, depth-dose and lateral (dose and RBEDSB) profiles for monodirectional monoenergetic (100 to 200 MeV) protons incident on a water phantom are compared. The effects of recoil and secondary ion production ({sub 2}H{sub +}, {sub 3}H{sub +}, {sub 3}He{sub 2+}, {sub 4}He{sub 2+}), spot size (3 and 10 mm), and transport physics on beam profiles and RBEDSB are examined. Results: Depth-dose and RBEDSB profiles among all of the MC models are in excellent agreement using a 1 mm distance criterion (width of a voxel). For a 100 MeV proton beam (10 mm spot), RBEDSB = 1.2 ± 0.03 (− 2–3%) at the tip of the Bragg peak and increases to 1.59 ± 0.3 two mm distal to the Bragg peak. RBEDSB tends to decrease as the kinetic energy of the incident proton increases. Conclusion: The model for proton RBEDSB has been accurately implemented into FLUKA, MCNP, TOPAS and the RayStation™TPS. The transport of secondary light ions (Z > 1) has a significant impact on RBEDSB, especially distal to the Bragg peak, although light ions have a small effect on (dosexRBEDSB) profiles. The ability to incorporate spatial variations in proton RBE within a TPS creates new opportunities to individualize treatment plans and increase the therapeutic ratio. Dr. Erik Traneus is employed full-time as a Research Scientist

  5. IDENTIFICACIÓN EFICIENTE DE ERRORES EN ESTIMACIÓN DE ESTADO USANDO UN ALGORITMO GENÉTICO ESPECIALIZADO IDENTIFICAÇÃO EFICAZ DOS ERROS EM ESTIMATIVA DE ESTADO USANDO UM ALGORITMO GENÉTICO ESPECIALIZADO EFFICIENT IDENTIFICATION OF ERRORS IN STATE ESTIMATION THROUGH A SPECIALIZED GENETIC ALGORITHM

    Directory of Open Access Journals (Sweden)

    Hugo Andrés Ruiz

    2012-06-01

    Full Text Available En este artículo se presenta un método para resolver el problema de estimación de estado en sistemas eléctricos usando optimización combinatoria. Su objetivo es el estudio de mediciones con errores de difícil detección, que afectan el desempeño y calidad de los resultados cuando se emplea un estimador de estado clásico. Dada su complejidad matemática, se deducen indicadores de sensibilidad de la teoría de puntos de apalancamiento que se usan en el algoritmo de optimización de Chu-Beasley, con el fin de disminuir el esfuerzo computacional y mejorar la calidad de los resultados. El método propuesto se valida en un sistema IEEE de 30 nodos.Neste artigo apresenta-se um método para resolver o problema de estimativa de estado em sistemas elétricos usando otimização combinatória. Seu objetivo é o estudo de medidas com erros de difícil detecção, que afetam o desempenho e qualidade dos resultados quando se emprega um estimador de estado clássico. Dada sua complexidade matemática, deduzem-se indicadores de sensibilidade da teoria de pontos de alavancagem que se usam no algoritmo de otimização de Chu-Beasley, com o fim de diminuir o esforço computacional e melhorar a qualidade dos resultados. O método proposto se valida em um sistema IEEE de 30 nós.In this paper a method to solve the state estimation problem in electric systems applying combinatorial optimization is presented. Its objective is the study of measures with difficult detection errors, which affect the performance and quality of the results when a classic state estimator is used. Due to the mathematical complexity, sensibility indicators are deduced from the theory of leverage points used in the Chu-Beasley optimization algorithm with the purpose of reducing the computational effort and enhance the quality of the results. The proposed method is validated in a 30-node IEEE system.

  6. Experimental measures of the energy rate absorbed in the aluminium and the comparison with the calculation using factors of dose and carrier of electrons by means of MCNP code

    International Nuclear Information System (INIS)

    Federico, Claudio A.; Vieira, Wilson J.; Rigolon, Leda S.Y.; Geraldo, Luiz P.

    2000-01-01

    In this paper are presented the results of a Monte Carlo calculation for the energy deposition rate in aluminum plates, when a collimated beam of gamma-rays produced by thermal neutrons capture in nickel target passes through them. The absorbed dose rate as a function of the aluminum thickness crossed by the gamma beam has been measured by using CaSO e :Dy thermoluminescent dosimeters. The capture gamma ray beam was extracted from a tangential beam tube of the IPEN's IEA-R1 2MW research reactor. The absorbed dose calculation was performed employing the Monte Carlo N-particle transport code (MCNP) and two methods of calculation: the simulated gamma ray flux multiplied by a dose conversion factor, and the simulated electron flux multiplied by the collision linear energy loss. The calculation results obtained by the electron transport have shown a good agreement with the experimental measurements. For deeper layers (more than 10 mm aluminum thickness), the calculation using the gamma ray flux multiplied by dose conversion factors, as well the calculation employing the electron transport, exhibit the same decreasing trade observed in experimental data, differing by a normalization factor of approximately 1.4. However, for layers nearer the material surface, the calculation using photon flux produces an overestimation of that using the electron transport as well as of the experimental results. (author)

  7. Tratamiento de enfermedad de aorta torácica con afectación de troncos supraaórticos usando técnica combinada de cirugía convencional y tratamiento endovascular

    Directory of Open Access Journals (Sweden)

    Antonio Jiménez Aceituno

    2007-04-01

    Conclusiones: El abordaje del arco aórtico usando técnicas endovasculares presenta el problema de la oclusión de TSA. El uso de derivaciones quirúrgicas hace posible esto. El abordaje combinado es aún una técnica nueva pero que podría ser una alternativa válida a la cirugía abierta de arco y aorta torácica descendente. Nuestra experiencia es satisfactoria, aunque no concluyente por tratarse sólo de dos casos.

  8. Construcción de un marco teórico/conceptual para abordar el trabajo de laboratorio usando el diagrama V : un estudio de caso de la UPEL / IPC

    OpenAIRE

    Flores, Julia; Caballero Sahelices, María Concesa; Moreira, Marco Antonio

    2011-01-01

    Estudio descriptivo para interpretar, desde la teoría ausubeliana, la construcción de una base teórica/conceptual integrada al marco metodológico para la resolución de problemas en el laboratorio de Bioquímica del IPC usando el diagrama V. Participaron ocho docentes en formación del área de Química que cursaron Bioquímica en el periodo académico 2006-II. Los estudiantes desarrollaron el diagrama V de Gowin para cinco actividades de laboratorio de complejidad creciente. El marco teórico/concep...

  9. Validity of the Brazilian version of WHOQOL-BREF in depressed patients using Rasch modelling Validez de la versión brasilera de WHOQOL-BREF en pacientes deprimidos usando el modelo del Rasch Validade da versão brasileira do WHOQOL-BREF em pacientes deprimidos usando o modelo de Rasch

    Directory of Open Access Journals (Sweden)

    Neusa Sica da Rocha

    2009-02-01

    Full Text Available OBJECTIVE: To assess the validity of the Brazilian version of the World Health Organization Quality of Life Instrument - Abbreviated version (WHOQOL-BREF in adults with major depression, using Rasch modelling. METHODS: Study analyzing data from the baseline sample of the Longitudinal Investigation of Depression Outcomes in Brazil, including a total of 208 patients with major depression recruited in a primary care service in Porto Alegre (Southern Brazil, in 1999. The Center for Epidemiological Studies Depression Scale was used to assess intensity of depression; the WHOQOL-BREF to assess generic quality of life; and the Composite International Diagnostic Interview version 2.1 for the diagnosis of depression. RESULTS: In the Rasch analysis, the four domains of WHOQOL-BREF showed appropriate fit to this model. Some items needed adjustments: four items were rescored (pain, finances, services, and transport; two items (work and activity were identified as having dependency of responses, and one item was deleted (sleep due to multidimensionality. CONCLUSIONS: The validation of the WHOQOL-BREF Brazilian version using Rasch analysis complements previous validation studies, evidencing the robustness of this instrument as a generic cross-cultural quality of life measure.OBJETIVO: Evaluar la validez de la version brasilera del Instrumento de Calidad de Vida de la Organización Mundial de la Salud, abreviado en inglés WHOQOL-BREF, en adultos con depression, usando el modelo de Rasch. MÉTODOS: Se analizaron datos de la base de Investigación Longitudinal de Resultados de Depresión en Brasil, incluyendo un total de 208 pacientes con derpesión mayor recluidos en servicio de ciudado primario en Porto Alegre (Sur de Brasil, en 1999. El Centro de Estudios Epidemiológicos de Depresión fue usado para evaluar la intensidad de la depresión; el WHOQOLBREF para evaluar la calidad genérica de vida; y la Entrevista Internacional Compuesta para el Diagnostico

  10. Evaluación del impacto de la vacuna contra rotavirus en Colombia usando métodos rápidos de evaluación

    Directory of Open Access Journals (Sweden)

    Karol Cotes

    2013-10-01

    Full Text Available OBJETIVO: Estimar la efectividad de la vacuna monovalente antirrotavírica para prevenir la hospitalización por enfermedad diarreica aguda en niños menores de 2 años en cinco ciudades de Colombia. MÉTODOS: Se realizó una encuesta poblacional sobre una muestra probabilística de niños mayores de 2 meses y menores de 24 meses de edad en cinco ciudades de Colombia (Barranquilla, Bogotá, Cali, Cartagena y Riohacha en el período de agosto a octubre de 2010. La vacuna fue introducida en el Programa Ampliado de Inmunizaciones en enero de 2009. Se estimaron las coberturas de vacunación contra rotavirus por grupos de edad y la incidencia acumulada de hospitalización por diarrea severa, y se evaluó la magnitud de la asociación entre la vacunación con una o dos dosis de vacuna antirrotavírica y la hospitalización por diarrea, utilizando la razón de probabilidades (RP ajustada por edad y otros factores de importancia epidemiológica. La efectividad de la vacunación se estimó usando la expresión 1 - RP. RESULTADOS: La cobertura de vacunación con una dosis de vacuna fue de 87,3%. En los 12 meses previos a la encuesta 43,2% (1 453 niños de menores de 24 meses presentaron diarrea, y de ellos, 5,2% (174 niños fueron hospitalizados por esta causa. La efectividad de dos dosis de vacuna antirrotavírica para prevenir la hospitalización por diarrea severa fue de 68% (intervalo de confianza de 95%: 55%-77%. CONCLUSIONES: La vacunación contra rotavirus en Colombia protege contra la hospitalización por diarrea por cualquier causa. El uso de encuestas transversales se mostró adecuado para evaluar rápidamente la efectividad de un programa de vacunación con una nueva vacuna.

  11. Estudio del comportamiento termo-mecánico de un acero microaleado de medio carbono durante un proceso de conformado en caliente usando una red neuronal artificial

    Directory of Open Access Journals (Sweden)

    Alcelay, Ignacio

    2016-06-01

    Full Text Available The thermo-mechanical behavior of medium carbon microalloyed steel has been analyzed by an Artificial Neural Network (ANN. The flow curves for training the ANN have been obtained from the hot compression tests, carried out over a temperature range varying from 900 to 1150 °C and at different true strain rates ranging from 10-4 to 10 s-1. It has been found that the ANN model developed in this study is capable to predict accurately and efficiently the flow behavior of the studied steel and there is a good agreement between the experimental results and the ANN results. To analyze the formability of the studied steel, processing maps have been constructed on the basis of the Dynamic Materials Model (DMM, using the ANN values of the flow stress. The study of maps reveals the different domains of the flow behavior of the studied steel and shows the great similarity between the experimental results and the theoretical results, so the use of the ANN can constitute an interesting alternative for design and study of hot forming processes.El comportamiento termo-mecánico de un acero microaleado de medio carbono ha sido analizado mediante una Red Neuronal Artificial (RNA. Las curvas de fluencia para el entrenamiento de la RNA han sido obtenidas mediante ensayos de compresión en caliente que se efectuaron a temperaturas que oscilaron entre 1150 °C y 900 °C a incrementos de 50 °C, y en un intervalo de velocidades de deformación que varió entre 10-4 y 10 s-1. Se ha podido comprobar que el modelo de RNA, desarrollado en el presente trabajo, es capaz de predecir con exactitud y eficiencia el comportamiento de fluencia en caliente del acero estudiado y existe un buen acuerdo entre los resultados experimentales y los resultados de la RNA. Para analizar la conformabilidad del acero microaleado se han construido mapas de procesado basados en el modelo dinámico de materiales (DMM usando los valores de la tensión de fluencia obtenidos mediante la RNA. El

  12. Processing of red ceramic using a fast-firing cycle Processamento de cerâmica vermelha usando um ciclo de queima rápido

    Directory of Open Access Journals (Sweden)

    G. T. Saleiro

    2012-09-01

    Full Text Available This work reports on the processing of red ceramic for civil construction using fast-firing cycles. The firing cycle is an important variable in the processing of red ceramic materials, which contributes to a high consumption of energy. The red ceramic pieces were prepared by industrial extrusion and fired at firing temperatures varying from 700 ºC to 1100 ºC using different firing cycles (slow-firing cycle - 1º C/min and fast-firing cycle - 10 ºC/min and 20 °C/min. The technological properties (linear shrinkage, water absorption, apparent porosity, apparent density, and flexural strength as function of the firing temperature and firing cycle are investigated. The development of the microstructure was followed by SEM/SEI. The results showed that fast-firing red ceramics exhibits technological properties and microstructure comparable to conventionally fired red ceramics, resulting in great advantages in energy saving.Este trabalho descreve o processamento de cerâmica vermelha para construção civil usando ciclos de queima rápido. O ciclo de queima é uma variável importante no processamento de materiais de cerâmica vermelha, o qual contribui para um alto consumo de energia. As peças de cerâmica vermelha foram preparadas por extrusão industrial e queimadas nas temperaturas de queima variando de 700 ºC a 1100 ºC utilizando diferentes ciclos de queima (ciclo de queima lento - 1 ºC/min e ciclos de queima rápidos - 10 ºC/ min e 20 ºC/min. As propriedades tecnológicas (retração linear, absorção de água, porosidade aparente, massa específica aparente e tensão de ruptura à flexão em função da temperatura de queima e ciclo de queima são investigadas. O desenvolvimento da microestrutura foi avaliado por SEM/SEI. Os resultados mostraram que as peças de cerâmica vermelha obtidas via processo de queima rápida exibiram propriedades tecnológicas e microestrutura comparáveis àquelas convencionalmente obtidas via queima lenta

  13. Análisis de rendimiento académico estudiantil usando data warehouse y redes neuronales Analysis of students' academic performance using data warehouse and neural networks

    Directory of Open Access Journals (Sweden)

    Carolina Zambrano Matamala

    2011-12-01

    Full Text Available Cada día las organizaciones tienen más información porque sus sistemas producen una gran cantidad de operaciones diarias que se almacenan en bases de datos transaccionales. Con el fin de analizar esta información histórica, una alternativa interesante es implementar un Data Warehouse. Por otro lado, los Data Warehouse no son capaces de realizar un análisis predictivo por sí mismos, pero las técnicas de inteligencia de máquinas se pueden utilizar para clasificar, agrupar y predecir en base a información histórica con el fin de mejorar la calidad del análisis. En este trabajo se describe una arquitectura de Data Warehouse con el fin de realizar un análisis del desempeño académico de los estudiantes. El Data Warehouse es utilizado como entrada de una arquitectura de red neuronal con tal de analizar la información histórica y de tendencia en el tiempo. Los resultados muestran la viabilidad de utilizar un Data Warehouse para el análisis de rendimiento académico y la posibilidad de predecir el número de asignaturas aprobadas por los estudiantes usando solamente su propia información histórica.Every day organizations have more information because their systems produce a large amount of daily operations which are stored in transactional databases. In order to analyze this historical information, an interesting alternative is to implement a Data Warehouse. In the other hand, Data Warehouses are not able to perform predictive analysis for themselves, but machine learning techniques can be used to classify, grouping and predict historical information in order to improve the quality of analysis. This paper depicts architecture of a Data Warehouse useful to perform an analysis of students' academic performance. The Data Warehouse is used as input of a Neural Network in order to analyze historical information and forecast. The results show the viability of using Data Warehouse for academic performance analysis and the feasibility of

  14. Autoconceito dos professores: principais factores usando modelos de Análise de Dados Multivariada Teachers' self-concept: finding main factors and clusters by EDA models

    Directory of Open Access Journals (Sweden)

    Vitor Franco

    2008-01-01

    Full Text Available O autoconceito tem sido considerado uma dimensão muito importante da personalidade do professor, da sua prática e do seu desenvolvimento pessoal (MARKUS; WURF, 1987; SIMÕES, 2001. A investigação que apresentamos foi efectuada com 281 professores de Ciências da Natureza, do terceiro ciclo do Ensino Básico, em Portugal, usando o ICAC- Inventário Clínico do Auto Conceito (VAZ-SERRA, 1986. Na análise dos dados obtidos foram usados diferentes métodos de Análise Multivariada, apresentando-se os resultados da análise factorial de correspondências e nos modelos de classificação hierárquica baseados no coeficiente de afinidade. Os resultados obtidos: 1 confirmam a importância de dois grandes factores presentes no Autoconceito: aceitação social e auto-eficácia; 2 caracterizam estes principais factores no que se refere ao Autoconceito clínico dos professores; 3 mostram como esses factores são determinantes na forma como cada professor constroi o seu autoconceito.The self-concept has been considered as a very important dimension on teacher's personality, practice and development (MARKUS; WURF, 1987;SIMÕES, 2001. The present research concerns a sample of 281 teachers of Natural Science of the Third Cycle of Basic Education from Portugal that responded to the I.C.A.C. - Self-Concept Clinical Inventory (VAZ-SERRA, 1986. In the analysis of the questionnaires different multivariate data analysis methods have been used. This paper describes some results issued from correspondence analysis and hierarchical clustering models based on the affinity coefficient. The results obtained: 1 confirm the importance of two general main factors / types which are present in self-concept: social acceptance and self-efficiency; 2 characterise these main factors when teachers'clinical self-concept is concerned and 3 show how determinant these factors are for the building of self-concept that allow us to differentiate teachers.

  15. Caracterización de la diversidad genética en naranja y comparación del polimorfismo de microsatélites amplificados al azar (RAMs usando electroforesis de poliacrilamida y agarosa

    Directory of Open Access Journals (Sweden)

    Muñoz Flores Jaime Eduardo

    2009-12-01

    Full Text Available Se compararon las eficiencias de tres métodos de electroforesis en agarosa y poliacrilamida, usando la cámara pequeña de DNA Sequencing System y cámara grande OWL Sequi-Gen Sequencing Cell, en la detección del polimorfismo en 21 accesiones de naranja (Citrus sinensis con empleo del cebador CGA. El gel de poliacrilamida dio mejor resolución de los productos amplificados vía PCR producidos por RAMs. Este permitió una mejor detección de bandas de ADN polimórficas, lo que facilitó la identificación de la variabilidad genética. La electroforesis en agarosa puede ser más conveniente en otras aplicaciones, debido al bajo costo y fácil aplicación. El estudio de diversidad genética en naranja usando microsatélites RAMs diferenció 51 accesiones en siete grupos con 0.75 de similaridad y 0.25 de heterocigosidad, lo que revela bajo polimorfismo genético. La técnica RAMs permitió agrupar las accesiones en Comunes o Blancas, Navel y Pigmentadas o Sanguinas.

  16. Estimate potential evaporation and solar radiation in the Yaqui valley, Sonora, Mexico, using data from satellite; Estimacion de evaporacion y radiacion solar en el valle del Yaqui, Sonora, usando datos de satelite

    Energy Technology Data Exchange (ETDEWEB)

    J Watts, Christopher; Rodriguez, Julio Cesar [Instituto del Medio Ambiente y el Desarrollo sustentable del estado de Sonora (Mexico); Garatuza Payan, Jaime [Instituto Tecnologico de Sonora (Mexico); Henk de Bruin [Universidad Agricola de Wageningen (Netherlands); Stewart, John [Universidad de Southampton (United Kingdom)

    1999-12-01

    The data from tow automatic weather stations in the Yaqui valley were used to estimate potential evaporation using the Makkink formula, based on observed incoming solar radiation and climatological values of air temperature. The usefulness of this formula was assessed by comparison with the Penman-Monteith, Penman and Priestley-Taylor formula and measurements of net radiation. A methodology was presented for estimating incoming solar radiation using visible band data from the GOES satellite. Comparisons against ground-based measurements from two pyranometers installed in the Yaqui valley gave good results, particularly in months with low cloud cover. Images for August 1993 were used to produce a map of the spatial distribution of potential evaporation. [Spanish] Para calcular la evaporacion potencial en el valle del Yaqui, usando la formula de Makkink, se utilizaron datos de dos estaciones meteorologicas automaticas. La mencionada formula se basa en la radiacion solar incidente observada y en ciertos valores climatologicos de temperatura del aire. Se evaluo la utilidad de esta formula, comparandola con las de Penman-Monteith, Pennan y Priestley-Taylor, asi como con mediciones de radiacion neta. Se desarrollo una metodologia para estimar la radiacion solar incidente usando la banda visible del satelite GOES. Se hizo una comparacion con mediciones de dos piranometros instalados en el valla del Yaqui, obteniendose buenos resultados, principalmente en meses con poca nubosidad. Se utilizaron imagenes de agosto de 1993 para producir un mapa de la distribucion espacial de la evaporacion potencial.

  17. Implementación de un mantenimiento basado en la condición usando modelado y simulación: caso de estudio de un motor sin-crónico de imanes permanentes

    Directory of Open Access Journals (Sweden)

    Jabid Quiroga Méndez

    2011-05-01

    Full Text Available Este artículo introduce la arquitectura de un CBM (mantenimiento basado en la condición en una aplicación eléctrica. La detección de fallas de manera oportuna y eficiente constituye uno de los retos más importantes asociados al CBM y el enfoque basado en modelos en el medio para conseguirlo. Un caso de estudio en un motor sincrónico de imanes permanentes (PMSM es ejecutado para ilustrar cómo el modelado es utilizado en la implementación de un CBM. El monitoreo fue implementado en tiempo real usando Matlab® y dSpace®. Se emplea como indicadora de falla la diferencia entre los valores de la componente secuencial negativa para las corrientes predichas usando una red neuronal multicapa y la corriente obtenida del motor. Resultados experimentales demostraron la efectividad del modelo propuesto en la detección de la falla de cortocircuito en el estator en distintos niveles de severidad y carga, obteniendo una confiabilidad en la detección mayor al 95%.

  18. Investigative studies on the effects of cadmium rabbits on high enriched uranium-fueled and low enriched uranium-fueled cores of Ghana Research Reactor-1 using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Boffie, J., E-mail: jboffie@yahoo.com [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Akaho, E.H.K. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); Nyarko, B.J.B.; Odoi, H.C.; Tuffour-Achampong, K.; Abrefah, R.G. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana)

    2013-12-15

    Highlights: • The operating parameters for both the HEU core and proposed LEU core were similar. • The length of the Cd in the capsules must be increased for its use in the LEU core. • Cd rabbits can emergently be used to shut down MNSRs. - Abstract: Miniature Neutron Source Reactors (MNSRs) are noted to be among highly safe research reactors. However, because of its use of one control rod for reactivity control and shutdown purposes, alternative methods of shutting it down are important. The Ghana MNSR uses four cadmium rabbits of approximate dimensions 6.5 cm × 5.0 cm × 0.1 cm and mass of 9.48 g each to emergently shut down the reactor. The Monte Carlo N-Particle code; version 5, (MCNP5) was used to design the high enriched uranium (HEU) and low enriched uranium (LEU) cores of the MNSR with four cadmium rabbits inserted in four inner irradiation sites of each core. The operating parameters and shutdown parameters for both cores with the central control rod (CCR) either fully withdrawn or fully inserted had similar results with the HEU core having slightly better results in terms of safety. However, the results show that the four inserted cadmium rabbits make the HEU core subcritical whiles in the LEU core, it still remains critical (k{sub eff} = 1.00005 ± 0.00007). The length of the cadmium material in each cadmium rabbit must therefore be increased by at least 0.5 cm in order to attain subcriticality (k{sub eff} = 0.99989 ± 0.00006) and shutdown margin of 0.11 mk when inserted in the LEU core.

  19. Expresión transitoria del gen GUS en caña de azúcar usando Agrobacterium tumefaciens Transient gene expression in sugarcane using Agrobacterium tumefaciens

    Directory of Open Access Journals (Sweden)

    Martha Liliana Bonilla Betancourt

    2008-10-01

    Full Text Available En el estudio se desarrolló una metodología de transformación genética mediante Agrobacterium tumefaciens en cultivares colombianos de caña de azúcar. La transformación se evaluó mediante la expresión del gen GUS. Callos embriogénicos y explantes meristemáticos de los genotipos CC85-92, CC84-75 y CC87-505 se transformaron usando tres cepas (AGL-1, LBA4404 y EHA105 con el plásmido pCambia 1305.2 y dos (EHA105 y LBA4404 con pCambia 2301. Se usó el medio de infiltración (IM con acetosiringona y se evaluó el tiempo de cocultivo y la densidad óptica de la bacteria al momento de la inducción. Los genotipos mostraron respuesta diferencial con las combinaciones cepa-plásmido: obtuvieron mayor expresión del gen GUS cuando el genotipo CC85-92 se transformó con la cepa AGL-1-pCambia 1305.2. CC84-75 y CC87-505 mostraron mayor expresión cuando se transformaron con la cepa EHA105-pCambia 1305.2. Mayor eficiencia en la expresión se obtuvo cuando la bacteria se indujo en IM después de siete días de cocultivo y cuando la densidad óptica de la bacteria fue de 0.2(600nm al momento de la inducción. Se demostró superioridad de los explantes en la eficiencia de transformación.The aim of the present study was to develop a transformation method mediated by Agrobacterium in Colombian cultivars of sugarcane. Transformation was evaluated in each step through transient GUS expression. Embryogenic calli and meristematic explants of CC85-92, CC84-75 y CC87-505 cultivars, were transformed using Agrobacterium tumefaciens AGL-1, LBA 4404 and EHA 105 strains, harboring pCambia 1305.2 plasmid. Furthermore, strains LBA 4404 and EHA 105 harboring pCambia 2301 were also tested. Bacterian activator medium, named infiltration media (IM with acetosyringone was used. Co-cultivation time and bacteria optical density before induction were tested. Sugarcane cultivars evaluated showed differential response to different strain-plasmid combinations, obtaining

  20. Estabelecimento de pastagem de capim-tanzânia usando milheto como cultura acompanhante The establishment of tanzania grass pasture using millet as a companion crop

    Directory of Open Access Journals (Sweden)

    Marcos Carvalho Maia

    2000-10-01

    Full Text Available Este estudo objetivou avaliar o rendimento e a composição química da forragem de milheto e de capim-tanzânia, a primeira espécie como cultura companheira, na formação de pastagem da gramínea perene Panicum maximum Jacq. cv. Tanzânia I, bem como a viabilidade do emprego desta prática. Os tratamentos avaliados foram quatro combinações de mistura de sementes, numa densidade de semeadura básica de 8 kg/ha, de capim-tanzânia/milheto: (08/00; 05/03; 04/04 e 03/05 kg/ha e três freqüências de corte (três cortes -- 40, 70 e 100 dias após a semeadura; dois cortes -- 50 e 100 dias após a semeadura; um corte -- 100 dias após a semeadura, no delineamento de blocos ao acaso, em esquema de parcelas subdivididas. As densidades de semeadura foram alocadas nas parcelas principais e as freqüências de cortes nas subparcelas. As produções de MS da associação entre milheto e capim-tanzânia foram maiores que as do capim-tanzânia exclusivo. A concentração de PB na MS de ambas as espécies aumentou, ao passo que os teores de FDN e FDA reduziram, com o aumento do número de cortes, porém, na consorciação, verificam-se valores semelhantes nos teores destas variáveis e pequena redução no teor de PB, quando comparados aos teores do capim-tanzânia exclusivo. Com base nestes resultados, conclui-se que é viável a formação de pastagem do capim-tanzânia usando o milheto como cultura acompanhante.The objective was to evaluate the forage yield and chemical composition of millet and Tanzania grass, the first one as a companion crop, in the formation of the Panicum maximum Jacq. cv. Tanzania I pasture, as well as the viability of the use of this practice. The evaluated treatments were four seed mixture combinations based on sowing density of 8 kg/ha of Tanzania grass/millet (08/00; 05/03; 04/04 and 03/05 kg/ha and three cutting frequencies (three cuttings - 40, 70 and 100 days after sowing, two cuttings - 50 and 100 days after sowing; one

  1. AVALIAÇÃO EM MASSA DE IMÓVEIS USANDO GEOESTATÍSTICA E KRIGAGEM BAYESIANA (UM ESTUDO DE EM BALNEÁRIO CAMBORIÚ/SC

    Directory of Open Access Journals (Sweden)

    Ricardo André Hornburg

    2017-03-01

    Full Text Available RESUMO: Uma das grandes dificuldades que se tem na avaliação em massa de imóveis é encontrar um modelo que mostre a realidade do mercado de imóveis para que se possa construir uma Planta de Valores Genéricos (PVG, usada como base para a cobrança do Imposto Predial e Territorial Urbano (IPTU. Este artigo apresenta um método que combina o uso da econometria espacial com a geoestatística bayesiana visando estimar o valor dos imóveis levando em consideração as interações espaciais devidas às características da localização. Os métodos da regressão espacial e da krigagem bayesiana são usados com esta finalidade. A técnica da regressão espacial possibilita a modelagem da dependência espacial. A técnica da krigagem bayesiana permite estimar valores de variáveis espacialmente distribuídas a partir de valores adjacentes considerados como interdependentes. Dessa maneira, a krigagem é considerada um método de médias móveis. O semivariograma é a ferramenta básica de suporte às técnicas de krigagem, permitindo representar quantitativamente a variação de um fenômeno regionalizado no espaço. Uma aplicação do método proposto é realizada no bairro Centro da cidade de Balneário Camboriú (SC, usando-se uma amostra de dados de mercado para a avaliação em massa de imóveis do tipo apartamento. Destaca-se como contribuição do método apresentado uma possibilidade de melhora na determinação do valor justo dos imóveis numa avaliação em massa, amparando assim a equidade e consequente justiça fiscal quando aplicados pelos municípios. ABSTRACT: One of the greatest difficulties in bulk value appraisal of buildings is to find a model that shows the reality of the real estate market so you can build a Standard Ground Value (SGV, used as a basis for the collection of Land and Territorial Tax Urban (Land Tax. This paper presents a method that combines the use of spatial econometrics with Bayesian geostatistics in order to

  2. Sistema Distribuido de Detección de Sismos Usando una Red de Sensores Inalámbrica para Alerta Temprana.

    Directory of Open Access Journals (Sweden)

    Ana Zambrano Vizuete

    2015-07-01

    Full Text Available Resumen: El detectar eventos disruptivos usando sensores COTS como los utilizados en smartphones representa un gran desafío pero también una oportunidad interesante. En este artículo se presenta una arquitectura de sistema de tiempo real crítico, jerárquica y distribuida, que hace uso de smartphones que actúan como sensores a través de una aplicación de bajo consumo de energía que convierte a sus acelerómetros en acelerógrafos. Los smartphones desplegados forman una red de sensores que detecta, analiza y notifica un pico sísmico. El sistema optimiza cálculos distribuidos y capacidades de comunicación en smartphones para proveer tiempo extra para alertas tempranas en escenarios de desastre de tipo sísmico, aunque puede ser empleada como solución a otros desastres naturales. Se propone una solución innovadora de bajo coste que realiza análisis tanto espaciales como temporales, no presentes en otros trabajos, lo cual lo hace más preciso y personalizable permitiendo adaptarse a las características geográficas de la zona, de red, y recursos tanto humanos como monetarios. La arquitectura ha sido validada mediante una extensa evaluación, consiguiendo como resultado notificaciones tempranas que adelantan en decenas de segundos el pico máximo del sismo en la zona del epicentro y aún más para zonas más alejadas; y la considerable reducción de falsas alarmas. Adicionalmente la arquitectura propuesta incluye una administración post-evento que mejora la capacidad operativa, logística y de telecomunicaciones desde un solo nivel central, y al mismo tiempo, mantiene al usuario informado de centros de refugios cercanos, mejores rutas, rutas seguras para una mejor decisión. Abstract: Detecting disruptive events using COTS sensors like the ones embedded in smartphones is a difficult challenge but also an interesting opportunity. In this paper, we present a distributed, reliable, hierarchical and hard real-time system architecture of

  3. Avaliação de multielementos em amostras de sangue humano usando SR-TXRF Evaluation of multielements in human blood samples using synchrotron radiation

    Directory of Open Access Journals (Sweden)

    Nivia Graciele V. Pinto

    2010-01-01

    Full Text Available A técnica de fluorescência de raios X por reflexão total usando radiação síncrotron (SR-TXRF é uma poderosa ferramenta utilizada para a determinação das concentrações elementares presentes em amostras biológicas. O objetivo deste estudo é avaliar as possíveis alterações causadas por processos de irradiação na concentração de elementos-traço em amostras de sangue humano. As amostras de sangue foram coletadas no Laboratório de Análises Clínicas Dr. Elilel Figueiredo, Rio de Janeiro, e divididas em dois grupos. O primeiro grupo foi irradiado com doses de 1.500, 2.500 e 3.000 cGy, utilizando o irradiador Gammacell 220 Excel, e o segundo foi irradiado com doses que variaram de 2 cGy a 100 cGy, utilizando uma bomba de cobalto Theratron 780 C do Inca, Rio de Janeiro. Todas as amostras de sangue total, plasma e matriz celular foram então liofilizadas e, em seguida, passaram pelo procedimento padrão de digestão. Todas as medidas foram realizadas na linha de fluorescência de raios X do Laboratório Nacional de Luz Síncrotron (LNLS, em Campinas, Brasil. Não se verificou variação significativa na concentração de Ca e, em contrapartida, o K foi o único elemento que sofreu alterações significativas para todas as amostras analisadas em função da dose. A concentração de Fe diminuiu apenas para as amostras de sangue total e plasma. A concentração de Zn apresentou uma diminuição significativa somente para as amostras de sangue total.Total-reflection X-ray fluorescence using synchrotron radiation (SR-TXRF is a powerful analytical technique to study trace elements in biomedical samples. The aim of this study was to investigate possible changes in essential trace element concentrations caused by irradiation procedures. Fresh blood samples were obtained from the Dr. Eliel Figueiredo Laboratory, Rio de Janeiro. The samples were separated in two groups. The first was irradiated with doses of 1500, 2500 and 3000cGy, using a

  4. Disfunção tireoidiana e conduta dos cardiologistas em pacientes usando amiodarona Thyroid dysfunction and cardiological management in patients receiving amiodarone

    Directory of Open Access Journals (Sweden)

    Anna Gabriela Fuks

    2004-06-01

    Full Text Available OBJETIVO: Determinar a prevalência de disfunção tireoidiana em pacientes usando amiodarona e os possíveis fatores associados. Verificar através de questionário aplicado a cardiologistas, a importância do fármaco causar alterações na função tireoidiana. MÉTODO: Avaliados 56 pacientes em uso crônico (> 3 meses de amiodarona com dosagens séricas de TSH, T4 livre, T3 total e Anti-TPO e definidos como portadores de disfunção tireoidiana (DT pacientes com TSH alterado. RESULTADOS: A prevalência de disfunção tireoidiana foi de 33,9%. Não houve diferença entre este grupo e os pacientes sem disfunção, exceto em relação à prevalência de anti-TPO positivo maior nos pacientes com DT (p=0,02. Hipotireoidismo subclínico foi diagnosticado em 10 (17,9% pacientes e hipotireoidismo clínico em 6 (10,7%. A prevalência de hipertireoidismo subclínico foi de 3,6% e de hipertireoidismo clínico de 1,8%. Anticorpos anti-TPO foram positivos em 5 (8% pacientes (dos quais 4 apresentavam disfunção. Quando comparados aos doentes sem anti-TPO positivo este grupo teve maior prevalência de disfunção (80% vs 29,4%; p=0,04. Verificado que apenas 49,2% dos cardiologistas faziam acompanhamento da função tireoidiana rotineiramente e a prevalência de disfunção referida na experiência da maioria era de 1 a 10%. CONCLUSÃO: A prevalência de disfunção tireoidiana na nossa população foi elevada, mostrando a necessidade de implementação de uma rotina laboratorial. Houve grande divergência entre os cardiologistas em relação ao tipo de acompanhamento utilizado nos pacientes em uso de amiodarona.OBJECTIVE: To determine the prevalence of thyroid dysfunction in patients receiving amiodarone, and the possible associated factors. The study also aimed at assessing the effect of amiodarone on thyroid function through the application of a questionnaire to cardiologists. METHOD: Fifty-six patients chronically (> 3 months receiving amiodarone were

  5. Modelado de Materiales Compuestos por Elementos Finitos usando Restricciones Cinemáticas Finite Element Modeling of Composite Materials using Kinematic Constraints

    Directory of Open Access Journals (Sweden)

    Oscar E. Ruiz

    2009-12-01

    Full Text Available El propósito de este artículo es presentar simulaciones del comportamiento de materiales compuestos basado en restricciones cinemáticas entre las mismas fibras y entre las fibras y la resina circundante. En la revisión de literatura, los autores han encontrado que las restricciones cinemáticas no han sido plenamente explotadas para modelar materiales compuestos, probablemente debido a su alto costo computacional. El propósito de este articulo es exponer la implementación y resultados de tal modelo, usando Análisis por Elementos Finitos de restricciones geométricas prescritas a los nodos de la resina y las fibras. Las descripciones analíticas del comportamiento de materiales compuestos raramente aparecen. Muchas aproximaciones para describir materiales compuestos en capas son basadas en la teoría de funciones C1Z y C0 Z, tal como la Teoría Clásica de Capas (CLT. Estas teorías de funciones contienen significativas simplificaciones del material, especialmente para compuestos tejidos. Una aproximación hibrida para modelar materiales compuestos con Elementos Finitos (FEA fue desarrollada por Sidhu y Averill y adaptada por Li y Sherwood para materiales compuestos tejidos con polipropileno de vidrio.The purpose of this article is to present simulations of the behavior of composite materials based on kinematic restrictions among the fibers themselves and among fibers and the surrounding resine. In the literature review the authors have found that the kinematic restrictions have not been fully exploited for modeling composite materials, probably due to their high computational expense. The purpose of this article is to show the implementation and results of such a model, by using a Finite Element Analysis of geometric restrictions prescribed to the resine and fiber nodes. Closed analytic descriptions on behavior of layered composite materials are very rare. Many approaches to describe layered composite material are based on the theory of

  6. Nuevo Enfoque para la Clasificación de Señales EEG usando la Varianza de la Diferencia entre las Clases de un Clasificador Bayesiano

    Directory of Open Access Journals (Sweden)

    Thomaz R. Botelho

    2017-10-01

    Full Text Available Resumen: Los avances en robótica de rehabilitación están beneficiando en gran medida a los pacientes con discapacidad física. Los dispositivos de asistencia y rehabilitación pueden basar su funcionamiento en información fisiológica de los músculos y del cerebro a través de electromiografía (EMG y electroencefalografía (EEG, para detectar la intención de movimiento de los usuarios. En este trabajo se presenta una propuesta de interfaz multimodal para la adquisición, sincronización y procesamiento de señales EEG y de sensores inerciales, para ser aplicada en tareas de rehabilitación con exoesqueletos robóticos. Se realizaron experimentos con individuos sanos con el objetivo de analizar la intención de movimiento, la activación muscular e inicio de movimiento durante los movimientos de extensión de la rodilla. Esta propuesta es un nuevo enfoque para la clasificación de señales EEG usando un clasificador bayesiano tomando en cuenta la varianza de la diferencia entre las clases usadas. El aporte de este trabajo se sustenta con los resultados que muestran un incremento del 30% en la precisión de clasificación con señales EEG en comparación con los enfoques tradicionales de clasificación, en un análisis off-line para el reconocimiento de la intención de movimiento de los miembros inferiores. Abstract: Patients with physical disabilities can benefit from robotic rehabilitation. This improves the efficiency of recovery and, therefore, the rehabilitation of the patient. Assistive and rehabilitation devices can make use of physiological data, such as electromyography (EMG and electroencephalography (EEG, in order to detect movement intentions. This work presents a multimodal interface for signal acquisition, synchronization and processing of EEG and inertial sensors signals, to be applied in rehabilitation robotic exoskeletons. Experiments were performed with healthy individuals executing knee extension. The goal is to analyze

  7. IMPLEMENTACIÓN DE UN SERVICIO EN LA WEB –MASHUP– PARA LA FUSIÓN DE IMÁGENES SATELITALES USANDO LA TRANSFORMADA RÁPIDA DE WAVELET HAAR

    Directory of Open Access Journals (Sweden)

    Rubén Medina

    2014-05-01

    Full Text Available En este artículo se implementa un nuevo servicio en la Web que ofrece a los usuarios la posibilidad de realizar la fusión de imágenes de satélite provenientes de diferentes sensores remotos y/o con diferentes resoluciones espaciales. A lo largo del artículo tres temáticas importantes son abordadas. La primera temática corresponde al servicio Web, éste servicio es implementado usando software libre y cuenta con una sencilla interfaz donde el usuario puede interactuar y principalmente puede realizar una solicitud del servicio de fusión. Adicionalmente, en la aplicación Web se desarrolló un módulo que permite obtener datos georreferenciados de diferentes fuentes externas para crear un nuevo servicio (Mashup a través de las API’s, de manera rápida y fácil utilizando OpenStreetMaps. La segunda temática se ocupa del análisis de la transformada rápida de wavelet haar (TRWH, estos conceptos matemáticos se abordan a partir de un ejemplo usando una matriz que se descompone en coeficientes de detalle y de aproximación de segundo nivel. La última temática detalla la metodología propuesta, paso a paso, para realizar la fusión de imágenes usando la TRWH. Igualmente con el fin de determinar la eficiencia de la TRWH cinco wavelets diferentes fueron implementadas en Matlab para fusionar el mismo par de imágenes satelitales. Las imágenes resultantes fueron evaluadas tanto en la calidad espacial como en la espectral a través de cuatro índices. Los mejores resultados de la evaluación fueron obtenidos con la TRWH la cual preserva la riqueza espectral de la imagen multiespectral original y mejora su calidad espacial.

  8. Simulação tridimensional adaptativa da separação das fases de uma mistura bifásica usando a equação de Cahn-Hilliard

    OpenAIRE

    Nós, R.L.; Ceniceros, H.D.; Roma, A.M.

    2012-01-01

    Simulamos a separação dos componentes de uma mistura bifásica com a equação de Cahn-Hilliard. Esta equação contém intrincados termos não lineares e derivadas de alta ordem. Além disso, a delgada região de transição entre os componentes da mistura requer muita resolução. Assim, determinar a solução numérica da equação de Cahn-Hilliard não é uma tarefa fácil, principalmente em três dimensões. Conseguimos a resolução exigida no tempo usando uma discretização semi-implícita de segunda ordem. No e...

  9. Cálculo dimensional óptimo de una plata forma paralela tipo Stewart –Gough para aplicaciones pedagógicas usando algoritmos genéticos

    OpenAIRE

    César Augusto Peña; Eugenio Yime Rodríguez; Ilka Banfield

    2011-01-01

    En este artículo se propone el diseño óptimo de una plataforma paralela tipo Stewart-Gough que cubre un espacio de trabajo esférico. El cálculo de las dimensiones del robot se realizó por medio de técnicas de algoritmos genéticos. Ilustra los cálculos de cinemática inversa y del espacio de trabajo. Esta plataforma esta diseñada para aplicaciones pedagógicas en las áreas de teleoperación, control, visión y automatización usando robótica paralela. La plataforma experimental puede ser operada re...

  10. Estimate of DMFT index using teeth most affected by dental caries in twelve-year-old children Estimación del índice DMFT usando los dientes más afectados por caries dentales en niños de doce años Estimativa do Índice CPOD usando os dentes mais afetados pela cárie dentária aos doze anos

    Directory of Open Access Journals (Sweden)

    Stela Márcia Pereira

    2009-02-01

    Full Text Available The objective of the study was to develop regression models to describe the epidemiological profile of dental caries in 12-year-old children in an area of low prevalence of caries. Two distinct random probabilistic samples of schoolchildren (n=1,763 attending public and private schools in Piracicaba, Southeastern Brazil, were studied. Regression models were estimated as a function of the most affected teeth using data collected in 2005 and were validated using a 2001 database. The mean (SD DMFT index was 1.7 (2.08 in 2001 and the regression equations estimated a DMFT index of 1.67 (1.98, which corresponds to 98.2% of the DMFT index in 2001. The study provided detailed data on the caries profile in 12-year-old children by using an updated analytical approach. Regression models can be an accurate and feasible method that can provide valuable information for the planning and evaluation of oral health services.El objetivo de este estudio fue desarrollar modelos de regresión para describir el perfil epidemiológico de caries dentales en niños de 12 años en un área de baja prevalencia de caries. Fueron estudiados dos muestras distintas aleatorias y probabilísticas de niños escolares (n= 1.763 que estudiaban en colegios públicos y privados en Piracicaba, Sureste de Brasil. Modelos de regresión fueron estimados como una función de los dientes más afectados usando datos colectados en 2005 y fueron validados usando una base de datos del 2001. El índice promedio (SD de DMFT fue 1,7 (2,08 en 2001 y las ecuaciones de regresión estimaron un índice de DMFT de 1,67 (1,98, lo cual corresponde a 98,2% del índice en 2001. El estudio provee datos detallados sobre el perfil de caries en niños de 12 años usando una aproximación analítica. Los modelos de regresión pueden ser un método confiable y factible que puede proporcional información valiosa para la planificación y evaluación de servicios de cuidado oral.O objetivo do estudo foi

  11. Evaluación comparativa del desempeño de los sistemas estatales de salud usando cobertura efectiva Benchmarking of performance of Mexican states with effective coverage

    Directory of Open Access Journals (Sweden)

    Rafael Lozano

    2007-01-01

    Full Text Available Realizar un análisis comparativo del desempeño (benchmarking de las unidades subnacionales en un sistema de salud descentralizado es importante para favorecer la rendición de cuentas, monitorear el progreso, identificar los factores que determinan tanto el éxito como el fracaso, y crear una cultura basada en la evidencia. Desde 2001, la Secretaría de Salud de México se ha dedicado a desarrollar esta tarea basándose en el concepto de cobertura efectiva promovido por la Organización Mundial de la Salud (OMS, que la define como la fracción de ganancia potencial en salud que el sistema de salud podría aportar, con los servicios que actualmente ofrece. Usando los sistemas de información en salud, que incluyen encuestas de salud representativas a nivel estado, registros vitales y registros de egresos hospitalarios, se ha monitoreado la prestación de 14 intervenciones para mejorar la salud entre 2005 y 2006. La cobertura efectiva en general va desde 54% en Chiapas hasta 65% en el Distrito Federal. La cobertura efectiva para intervenciones en salud materno-infantil es mayor que para las intervenciones que abordan otros problemas de salud del adulto. La cobertura efectiva para el quintil de ingresos más bajo es de 52%, comparada con 61% para el quintil de ingresos más alto. La cobertura efectiva guarda especial relación con el gasto público en salud per cápita en todos los estados, y esta relación es más estrecha con las intervenciones ajenas a la salud materno-infantil que con las que tienen que ver directamente con ella. También se observan variaciones considerables en la cobertura efectiva en niveles de gasto similares. Asimismo, se discuten algunas implicaciones para el desarrollo que debiera seguir el sistema de información en salud en México. Este enfoque alienta a quienes toman decisiones a concentrarse en brindar servicios de calidad y no sólo en ofrecer la disponibilidad del servicio. El cálculo de la cobertura efectiva

  12. MCNP multiplication analysis of subcritical HEU experiments

    Energy Technology Data Exchange (ETDEWEB)

    Estes, G.P. [Los Alamos National Lab., NM (United States); Brockhoff, R.C. [Kansas State Univ., Manhattan, KS (United States)

    1998-12-31

    A series of measurements and improvements to computational techniques was described in Ref. 1 that were aimed at better understanding the determination of the reactivity of subcritical systems from measurements of the multiplying characteristics of the system. This methodology has been applied to a number of the bare highly enriched uranium (HEU) measurements (simulating 0.5- to 21.5-kg balls with nesting shells) of Ref. 2, demonstrating that the experimental multiplication results can be reproduced computationally with good accuracy. This capability promises to improve special nuclear material (SNM) assays of unknown systems such as those encountered in SNM safeguards, arms-control verification, imports of foreign-generated SNM, smuggling of SNM, etc. Improved techniques and understanding are needed since traditionally measured or calculated multiplications are not always an invariant characteristic of a subcritical system, especially if one has an SNM system with no significant intrinsic internal neutron source that is illuminated nonuniformly with an external source (i.e., a nonnormal mode system). The measurement techniques used in Refs. 1 and 2 to determine multiplication are based on the Feynman variance-to-mean method, which has been previously documented in Refs. 3 and 4 and applied successfully to normal mode systems such as plutonium and uranium spheres. These techniques have been applied to nonnormal mode problems with less success, and both Refs. 1 and 2 as well as the current paper are attempts to better understand the subcritical multiplication of such systems.

  13. Utilidade do Ultrassom intracardíaco no isolamento de veias pulmonares usando cateter-balão a laser Utilidad del ultrasonido intracardíaco en el aislamiento de venas pulmonares usando catéter-balón láser Utility of intracardiac ultrasound imaging to guide pulmonary vein ablation using laser balloon catheter

    Directory of Open Access Journals (Sweden)

    Luiz Leite

    2009-12-01

    Full Text Available FUNDAMENTO: O isolamento das veias pulmonares (IVP tem sido usado como endpoint para a ablação da fibrilação atrial (FA com cateter balão. OBJETIVO: Determinar a utilidade do ultrassom intracardíaco (USIC para guiar o IVP, usando cateter balão a laser. MÉTODOS: 59 VP foram ablacionadas em 27 cães. Imagens de Doppler foram usadas para identificar os vazamentos do fluxo sanguíneo entre a VP e o balão. Após cada liberação de energia, o cateter de mapeamento circular foi reposicionado para verificar se o isolamento tinha sido obtido. A posição de vazamento foi então correlacionada com a posição do gap no estudo patológico. A análise de regressão logística multivariada foi realizada. RESULTADOS: Cinquenta e nove VP foram submetidas à ablação. O tempo médio de energia liberada foi de 279±177 seg., o diâmetro médio do balão era de 23±3 mm, e o comprimento médio do balão era 25±4 mm. O isolamento completo foi obtido em 38/59 (64%, e foi significantemente mais comum sem vazamento: [30/38 (79% versus 8/23 (35%, pFUNDAMENTO: Se usó el aislamiento de las venas pulmonares (AVP como endpoint para la ablación de la fibrilación atrial (FA con catéter-balón. OBJETIVO: Determinar la utilidad del ultrasonido intracardíaco (USIC para guiar el AVP, usando catéter-balón láser. MÉTODOS: Se ablacionaron 59 VP en 27 perros. Se usaron imágenes de Doppler para identificar los derrames del flujo sanguíneo entre la VP y el balón. Tras cada liberación de energía, se reposicionó el catéter de mapeamiento circular para verificar si se obtuvo el aislamiento. Se correlaccionó, entonces, la posición del derrame con la posición del gap en el estudio patológico. Se realizó el análisis de regresión logística multivariada. RESULTADOS: Se sometieron 59 VP a la ablación. El tiempo promedio de energía liberada fue de 279±177 seg., el diámetro promedio del balón era de 23±3 mm, y la largura promedio del balón era 25±4 mm

  14. Energy saving using solar filters with iron base in windows; Ahorro de energia usando filtros solares con base en hierro en ventanas

    Energy Technology Data Exchange (ETDEWEB)

    Chavez Galan, Jesus

    2003-07-01

    termico generado al interior de las edificaciones por las inadecuadas propiedades de los materiales de construccion, se desarrollaron en este trabajo filtros solares con base en hierro por medio de los cuales se logra un control selectivo de la radiacion solar que se transmite a traves de las ventanas. Estos filtros solares consisten en peliculas delgadas de FeO depositadas sobre sustratos de vidrio cal-sosa (el mas usado en nuestro pais para las edificaciones) de 600x300x3 mm, por medio de la tecnica de sputtering asistida con radiofrecuencia y magnetrones planos, partiendo de un blanco de hierro puro de 127x254 mm y usando un plasma de argon. Para obtener el grado de oxidacion deseado en el hierro, pequenas muestras (45 x 22 mm) fueron sometidas a un proceso de calentamiento en una atmosfera reductora constituida de 50% H{sub 2} + 50% N{sub 2}, durante un periodo de tiempo de 10 minutos a una temperatura de 400 C. Los filtros solares con base en FeO presentan una transmisividad de 30.2% para el intervalo visible del espectro electromagnetico (radiacion con longitud de onda de 380-780 nm), y de 39.9% para el infrarrojo cercano (radiacion con longitud de onda de 780-2500 nm); mientras que su reflectividad es de 17.5 y 19%, para los intervalos visibles e infrarrojo cercano del espectro electromagnetico respectivamente. Se realizo una simulacion, a traves del software Energy 10, del comportamiento energetico de una casa-habitacion cuando utilizan en las ventanas vidrios simples (una lamina) tipo cal-sosa, asi como tambien cuando esta misma edificacion usa en las ventanas los filtros solares con base en FeO desarrollados. Ambos casos se comparan obteniendose asi los posibles ahorros de energia por el uso de dichos filtros solares en las ventanas. Las simulaciones fueron llevadas a cabo para diferentes localidades del pais, obteniendose que los filtros solares con base en FeO desarrollados, permiten ahorros de energia principalmente por concepto de acondicionamiento de aire

  15. Study of method of efficiency transference using detectors NaI(Ti); Estudo de método de transferência de eficiência usando detectores NaI(Tl)

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, Thiago L.; Salgado, César M., E-mail: thiago_lins-ramos@hotmail.com, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The use of NaI (Tl) scintillation detectors for measurements implies the determination of the detection efficiency as a function of the energy of the incident photons. The efficiency curve can be obtained experimentally with the use of several mono-energy sources calibrated with emission energies covering the whole range of interest or using the Monte Carlo method. The Institute of Nuclear Engineering develops several methodologies using these detectors as they are robust, inexpensive and do not need cooling for their use. The assembly of an experimental arrangement is usually complex, since several factors influence the result affecting reproducibility in measurements, such as: parallelism between source and detector, alignment between source and detector, and accuracy of source-detector distance. In view of such difficulties, an automated positioning system was developed for the source-detector set controlled by a micro controller based on the ARDUINO language in order to guarantee the reproducibility in the experimental arrangements. In the initial phase of this study, a mathematical model was developed in the MCNP-X code using a NaI (Tl) detector. A theoretical validation using the Efficiency Transfer Method was performed at three different positions on the detector's axial axis (10.6 cm, 11.3 cm and 12.0 cm). This method is based on the ratio of effective solid angles. The experimental validation presented maximum relative errors of 7.74% for the position 11.3 cm.

  16. Spectra and neutron dose of an 18 MV Linac using two geometric models of the head; Espectros y dosis por neutrones de un Linac de 18 MV usando dos modelos geometricos del cabezal

    Energy Technology Data Exchange (ETDEWEB)

    Barrera, M. T.; Pino, F.; Barros, H.; Sajo-Bohus, L. [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Sartenejas, Baruta 1080-A, Caracas (Venezuela, Bolivarian Republic of); Davila, J. [Fisica Medica C. A., Av. Francisco de Miranda s/n, Los Palos Grandes, 1060 Miranda (Venezuela, Bolivarian Republic of); Salcedo, E. [Centro Medico Docente La Trinidad, Av. de El Haltillo, Caracas (Venezuela, Bolivarian Republic of); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Benites R, J. L., E-mail: mariate9590@gmail.com [Centro de Cancerologia de Nayarit, Servicio de Seguridad Radiologica, Calz. de la Cruz 118 Sur, 63000 Tepic, Nayarit (Mexico)

    2015-10-15

    Full text: Using the Monte Carlo method, by MCNP5 code, simulations were performed with different source terms and 2 geometric models of the head to obtain spectra in energy, flow and doses of photo-neutrons at different positions on the stretcher and in the radiotherapy room. The simplest model was a spherical shell of tungsten; the second was the complete model of a heterogeneous head of an accelerator Varian ix. In both models Tosi function was used as a source term. In addition, for the second model Sheikh-Bagheri distribution was used for photons and photo-neutrons were generated. Also in both models the radiotherapy room of Gurve group of the Teaching Medical Center La Trinidad was included, which is equipped with an accelerator Varian Clinic 2100. In this Center passive detectors PADC (Cr-39) were irradiated with neutron converters, with 18 MeV photons radiation. The measured neutron flow was compared with that obtained with Monte Carlo calculations. The Monte Carlo flows are similar to those measured at the isocenter. The simplest model underestimates the neutron flow compared with the calculated flows with the heterogeneous model of the head. (Author)

  17. Efficacy of aprons equivalent to 0.5 mm of lead in PET procedures using the Monte Carlo method; Eficacia de aventais equivalentes a 0,5 mm de chumbo em procedimentos PET usando o metodo Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Fonseca, R.B.; Amaral, A. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Campos, L., E-mail: amaral@ufpe.br [Universidade Federal de Sergipe (UFS), Aracaju, SE (Brazil). Dept. de Fisica

    2012-07-01

    In positron emission tomography (PET), health staff is exposed to 511-keV photons, which is a result of the positron annihilation process. This energy is about four times greater than the 140 keV commonly found in studies based on single photon emission computed tomography (SPECT). Besides this different level of energy, 0.5 mm lead-equivalent aprons have being used either in SPECT or PET procedures. In this context, this work was designed for evaluating the effectiveness of such aprons in individual radioprotection of health professionals involved in positron emission tomography. For this, by using MCNP4C-based Monte Carlo simulations, the average energy delivered per particle to the regions corresponding to operational quantities Hp(10) and Hp(0.07) were calculated for two conditions of individual exposures: wearing and not wearing a 0.05 mm lead-equivalent apron. The results obtained pointed out that Hp(10) has similar value in both situations. On the other hand, for the region corresponding to Hp(0.07), wearing this lead apron will improve this dose in about 26%. On the basis of this work, 0.5 mm lead equivalent aprons do not offer adequate protection for medical staff working on positron emission tomography. (author)

  18. Identificação de variáveis cataclísmicas eruptivas na direção do bojo galáctico e Nuvens de Magalhães usando dados do OGLE

    Science.gov (United States)

    Cieslinski, D.; Diaz, M. P.; Mennickent, R.; Pietrzyski, G.

    2003-08-01

    Na década de 90 iniciaram-se vários programas para a pesquisa de matéria escura na Galáxia usando o efeito de microlentes gravitacionais. Entre os projetos mais bem conhecidos podemos mencionar o OGLE (Optical Gravitational Lensing Experiment) e o MACHO (MAssive Compact Halo Objects). A estratégia usada por eles consiste em fazer fotometria de banda larga (normalmente B, R e I) de um grande número de estrelas (dezenas de milhões) tão freqüentemente quanto possí vel e por longos perí odos de tempo (anos). Uma tal sistemática de observação, além de descobrir inúmeras lentes gravitacionais, é também muito apropriada para a descoberta de estrelas variáveis. De fato, inúmeras novas variáveis de vários tipos foram descobertas como subproduto. Exemplos podem ser encontrados nos endereços http://bulge.princeton.edu/~ogle/ e http://wwwmacho.mcmaster.ca/. As variáveis cataclí smicas eruptivas (novas clássicas, novas recorrentes e novas anãs) são objetos que apresentam variabilidade de grande amplitude com escalas de tempo de dias a centenas de dias e, por esta razão, devem ter sido detectadas em grande número nestes "surveys". Para testar esta possibilidade nós procuramos nos dados do OGLE por tais sistemas e o presente trabalho mostra os resultados desta pesquisa. Os objetos foram selecionados entre as variáveis detectadas usando a amplitude de variação de brilho como critério principal. Este critério forneceu 13756 objetos, sendo 2169 na direção da Grande Nuvem de Magalhães, 1162 na direção da Pequena Nuvem de Magalhães e o restante na direção do Bojo Galáctico. A análise foi feita inspecionando-se visualmente cada curva de luz por erupções com as características acima mencionadas. Os resultados obtidos podem ser sumarizados como: descoberta de duas novas clássicas e 33 novas anãs. Além disso, pode-se mencionar a identificação de candidatas a outros tipos de variáveis como: estrelas simbióticas, RV Tauri, R Coronae

  19. Evaluation of Top-Cross Popcorn Hybrids Using Mixed Linear Model Methodology Evaluación de Híbridos Top-Cross de Maíz-Roseta Usando Modelos Lineales Mixtos

    Directory of Open Access Journals (Sweden)

    Emmanuel Arnhold

    2009-03-01

    Full Text Available The market for popcorn (Zea mays L. has been continuously growing in Brazil fact has required the development of cultivars adapted to local environmental conditions. For this reason, the analytical objectives of this study were to evaluate top-cross popcorn hybrids in relation to popping expansion and grain yield in three different eco-geographic regions of Brazil, in order to estimate variance components using Restricted Maximum Likelihood (REML and predict breeding values using Best Linear Unbiased Prediction (BLUP. Genetic evaluation considered a linear model with heterogeneous residual (environmental variances. The Restricted Likelihood Ratio Test (RLRT evidenced significant differences (p Evaluación de híbridos top-cross de maíz-roseta usando modelos lineales mixtos. El mercado del maíz-roseta (Zea mays L. está en continuo crecimiento en Brasil, lo cual ha demandado el desarrollo de cultivares adaptados a las condiciones locales. Por ello, los objetivos del presente trabajo fueron evaluar híbridos top-cross de maíz-roseta en función de la capacidad de expansión y el rendimiento de los granos, en diferentes regiones eco-geográficas de Brasil; estimar componentes de varianza usando Máxima Verosimilitud Restringida (REML y predecir los valores genotípicos a través de la Mejor Predicción Linear Insesgada (BLUP. La evaluación genética consideró un modelo lineal con una estructura de varianza residual (ambiental heterogénea. La prueba de la razón de verosimilitud (restringida evidenció diferencias significativas (p < 0,01 para el efecto genotípico. La producción de granos mostró ser una característica de heredabilidad media (h² = 0,26-0,39. En la capacidad de expansión se evidenció un mayor control genético aditivo (h² = 0,58-0,85. Las correlaciones genéticas y de Spearman entre las características fueron negativas, indicando que la selección basada en el rendimiento de granos tendría un efecto negativo sobre la

  20. Aplicações da técnica de fotoeletrocatálise na oxidação de corantes ácidos, corantes dispersos, surfatantes e na redução de 'CR'((VI) e bromato em efluentes usando eletrodos nanoporosos de 'TI'/'TI''O IND.2'

    OpenAIRE

    Paschoal, Fabiana Maria Monteiro [UNESP

    2008-01-01

    O presente trabalho ilustra diversas aplicações da técnica de fotoeletrocatálise, tais como: degradação de corantes têxteis, corantes de curtumes, surfatantes, redução de Cr(VI) a Cr(III) e redução de bromato a brometo, usando eletrodos de filmes finos de Ti/TiO2 e luz ultravioleta. A oxidação fotoeletrocatalítica de soluções simuladas de efluentes de curtumes contendo dicromato de potássio, o corante ácido vermelho e o surfatante Tamol® foi investigada em Na2SO4 0,1 mol L-1 usando anodos de ...