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Sample records for code mcnp modeling

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

    Science.gov (United States)

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

    2013-08-21

    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.

  2. Varian 2100C/D Clinac 18 MV photon phase space file characterization and modeling by using MCNP Code

    Science.gov (United States)

    Ezzati, Ahad Ollah

    2015-07-01

    Multiple points and a spatial mesh based surface source model (MPSMBSS) was generated for 18MV Varian 2100 C/D Clinac phase space file (PSF) and implemented in MCNP code. The generated source model (SM) was benchmarked against PSF and measurements. PDDs and profiles were calculated using the SM and original PSF for different field sizes from 5 × 5 to 20 × 20 cm2. Agreement was within 2% of the maximum dose at 100cm SSD for beam profiles at the depths of 4cm and 15cm with respect to the original PSF. Differences between measured and calculated points were less than 2% of the maximum dose or 2mm distance to agreement (DTA) at 100 cm SSD. Thus it can be concluded that the modified MCNP code can be used for radiotherapy calculations including multiple source model (MSM) and using the source biasing capability of MPSMBSS can increase the simulation speed up to 3600 for field sizes smaller than 5 × 5 cm2.

  3. Recent Developments in the MCNP-POLIMI Postprocessing Code

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, S.A.

    2004-12-17

    The design and analysis of measurements performed with organic scintillators rely on the use of Monte Carlo codes to simulate the interaction of neutrons and photons, originating from fission and other reactions, with the materials present in the system and the radiation detectors. MCNP-PoliMi is a modification of the MCNP-4c code that models the physics of secondary particle emission from fission and other processes realistically. This characteristic allows for the simulation of the higher moments of the distribution of the number of neutrons and photons in a multiplying system. The present report describes the recent additions to the MCNP-PoliMi post-processing code. These include the simulation of detector dead time, multiplicity, and third order statistics.

  4. 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.

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

    Directory of Open Access Journals (Sweden)

    Kim Kyung-O

    2016-01-01

    Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.

  6. 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)

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

    Science.gov (United States)

    Li, Fusheng; Han, Xiaogang

    2012-07-01

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

  8. A comparison between the Monte Carlo radiation transport codes MCNP and MCBEND

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Hidenori; Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)

    2001-01-01

    In Japan, almost of all radiation analysts are using the MCNP code and MVP code on there studies. But these codes have not had automatic variance reduction. MCBEND code made by UKAEA have automatic variance reduction. And, MCBEND code is user friendly more than other Monte Carlo Radiation Transport Codes. Our company was first introduced MCBEND code in Japan. Therefore, we compared with MCBEND code and MCNP code about functions and production capacity. (author)

  9. 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.

  10. Validation of the Monte Carlo code MCNP-DSP

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T. [Oak Ridge National Lab., TN (United States)

    1996-09-12

    Several calculations were performed to validate MCNP-DSP, which is a Monte Carlo code that calculates all the time and frequency analysis parameters associated with the {sup 252}Cf-source-driven time and frequency analysis method. The frequency analysis parameters are obtained in two ways: directly by Fourier transforming the detector responses and indirectly by taking the Fourier transform of the autocorrelation and cross-correlation functions. The direct and indirect Fourier processing methods were shown to produce the same frequency spectra and convergence, thus verifying the way to obtain the frequency analysis parameters from the time sequences of detector pulses. (Author).

  11. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-15

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

  13. 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.

  14. Accuracy of the electron transport in mcnp5 and its suitability for ionization chamber response simulations: A comparison with the egsnrc and penelope codes

    Energy Technology Data Exchange (ETDEWEB)

    Koivunoro, Hanna; Siiskonen, Teemu; Kotiluoto, Petri; Auterinen, Iiro; Hippelaeinen, Eero; Savolainen, Sauli [Department of Physics, University of Helsinki, P.O. Box 64, FI-00014 Helsinki University (Finland) and Department of Oncology, Helsinki University Central Hospital, FI-00029 HUS (Finland); STUK-Radiation and Nuclear Safety Authority, P.O. Box 14, FI-00881 Helsinki (Finland); VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Department of Physics, University of Helsinki, P.O. Box 64, FI-00014 Helsinki University (Finland); HUS Medical Imaging Centre, Helsinki University Central Hospital, FI-00029 HUS (Finland)

    2012-03-15

    Purpose: In this work, accuracy of the mcnp5 code in the electron transport calculations and its suitability for ionization chamber (IC) response simulations in photon beams are studied in comparison to egsnrc and penelope codes. Methods: The electron transport is studied by comparing the depth dose distributions in a water phantom subdivided into thin layers using incident energies (0.05, 0.1, 1, and 10 MeV) for the broad parallel electron beams. The IC response simulations are studied in water phantom in three dosimetric gas materials (air, argon, and methane based tissue equivalent gas) for photon beams ({sup 60}Co source, 6 MV linear medical accelerator, and mono-energetic 2 MeV photon source). Two optional electron transport models of mcnp5 are evaluated: the ITS-based electron energy indexing (mcnp5{sub ITS}) and the new detailed electron energy-loss straggling logic (mcnp5{sub new}). The electron substep length (ESTEP parameter) dependency in mcnp5 is investigated as well. Results: For the electron beam studies, large discrepancies (>3%) are observed between the mcnp5 dose distributions and the reference codes at 1 MeV and lower energies. The discrepancy is especially notable for 0.1 and 0.05 MeV electron beams. The boundary crossing artifacts, which are well known for the mcnp5{sub ITS}, are observed for the mcnp5{sub new} only at 0.1 and 0.05 MeV beam energies. If the excessive boundary crossing is eliminated by using single scoring cells, the mcnp5{sub ITS} provides dose distributions that agree better with the reference codes than mcnp5{sub new}. The mcnp5 dose estimates for the gas cavity agree within 1% with the reference codes, if the mcnp5{sub ITS} is applied or electron substep length is set adequately for the gas in the cavity using the mcnp5{sub new}. The mcnp5{sub new} results are found highly dependent on the chosen electron substep length and might lead up to 15% underestimation of the absorbed dose. Conclusions: Since the mcnp5 electron

  15. Validation and Verification of MCNP6 Against Intermediate and High-Energy Experimental Data and Results by Other Codes

    CERN Document Server

    Mashnik, Stepan G

    2010-01-01

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

  16. 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.

  17. MCNP modelling of a combined neutron/gamma counter

    CERN Document Server

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

    1999-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Huseyin Ozan Tekin

    2016-01-01

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

  19. Simulations of X-ray spectrum and HVL for mammographic equipment using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rafael Toledo F. de; Alvarez, Matheus; Velo, Alexandre F.; Oliveira, Marcela de; Miranda, Jose Ricardo A. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Inst. de Biociencias de Botucatu. Dept. de Fisica e Biofisica; Pina, Diana R. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Dept. de Doencas Tropicais e Diagnostico por Imagem

    2012-07-01

    Full text: The main goal of mammography is early detection of breast cancer. Thus, the mammograph should be designed so that the X-ray photons are emitted within an appropriate energy range, to distinguish the normal breast tissue and cancerous tissue. The distribution of the photons amount of X-ray beam, with their respective energies, is called the spectrum. From the spectrum it is possible to estimate the quality of the X-ray beam from the Half Value Layer (HVL). Objectives: This study aims to simulate the Senographe 600T mammography unit, manufactured by General Electric (GE), using the MCNP5 Monte Carlo code, to obtain its spectrum and HVL, and compare the HVL of the simulated model with experimental data. Method: the mammography unit was simulated using a simplified model which a beam of 2x10{sup 8} electrons focuses on a Mo target angled 12 degrees, within a capsule filled with vacuum. The incident electrons were converted into photons. The capsule has a beryllium window, allowing the passage of the X-ray beam. The beam is detected by an air cylinder with 1 cm thickness placed 60 cm from the target. On the path of X-ray beam, is inserted a 0.03 mm Mo filter located 1.6 cm after the beryllium window. The space between the capsule and the detector cylinder was filled with air. The quality of X-ray beam was verified from the HVL using the MCNP5 code and the experimental method for the voltage range typically used in clinical routine (26-31 kVp). Results and discussion: the X-ray spectrum of the mammography device is satisfactorily simulated by MCNP5, showing the characteristic radiation peaks of molybdenum at 17.479 keV and 19.602 keV, the filtered spectrum generated by Bremsstrahlung, and reducing the total number of photons with the decrease in applied tension (kVp). The HVL obtained by MCNP5 and experimental measurements show a maximum difference of 5.31% (for 31 kVp). The result of both methods are within acceptable limits established by national

  20. 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)

  1. 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 

  2. 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)

  3. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  4. MCNP-DSP users manual

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.

    1997-01-01

    The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from the {sup 252}Cf-source-driven frequency analysis measurements. This code can be used to validate calculational methods and cross section data sets from subcritical experiments. This code provides a more general model for interpretation and planning of experiments for nuclear criticality safety, nuclear safeguards, and nuclear weapons identification and replaces the use of point kinetics models for interpreting the measurements. The use of MCNP-DSP extends the usefulness of this measurement method to systems with much lower neutron multiplication factors.

  5. Establishment and Verification of MCNP Neutron Transport Model About Tianwan Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Qi

    2012-01-01

    <正>In order to calculating the neutron flux in the surveillance box and reactor pressure vessel of the Tianwan NPP, we need to build up the neutron transport model by using the Monte Carlo code MCNP. The core of the NPP is very complicated for modeling so we put forward some assumptions that can simplify the neutron transport model. A lot of calculation works have been done to prove that the assumptions are right and suitable.

  6. VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)

    2008-07-01

    Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)

  7. The REBUS-MCNP linkage.

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, J. G.; Nuclear Engineering Division

    2009-04-24

    The Reduced Enrichment Research and Test Reactor (RERTR) Program uses the REBUS-PC computer code to provide reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in fuel and neutron absorbers with burnup. REBUS-PC models the complete fuel cycle including shuffling capability. REBUS-PC evolved using the neutronic capabilities of multi-group diffusion theory code DIF3D 9.0, but was extended to apply the continuous energy Monte Carlo code MCNP for one-group fluxes and cross-sections. The linkage between REBUS-PC and MCNP has recently been modernized and extended, as described in this manual. REBUS-PC now calls MCNP via a system call so that the user can apply any valid MCNP executable. The interface between REBUS-PC and MCNP requires minimal changes to an existing MCNP model, and little additional input. The REBUS-MCNP interface can also be used in conjunction with DIF3D neutronics to update an MCNP model with fuel compositions predicted using a DIF3D based depletion.

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

    OpenAIRE

    Aldawahra Saadou; Khattab Kassem; Saba Gorge

    2015-01-01

    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 propos...

  9. 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)

  10. MCNP{trademark} Software Quality Assurance plan

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. 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)

  12. 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.

  13. Simulation of a nuclear densimeter using the Monte Carlo MCNP-4C code; Simulacao de um densimetro nuclear utilizando o codigo Monte Carlo MCNP-4C

    Energy Technology Data Exchange (ETDEWEB)

    Penna, Rodrigo [UNI-BH, Belo Horizonte, MG (Brazil). Dept. de Ciencias Biologicas, Ambientais e da Saude (DCBAS/DCET); Silva, Clemente Jose Gusmao Carneiro da [Universidade Estadual de Santa Cruz, UESC, Ilheus, BA (Brazil); Gomes, Paulo Mauricio Costa [Universidade FUMEC, Belo Horizonte, MG (Brazil)

    2008-07-01

    Viability of building a nuclear wood densimeter based on low energy photons Compton scattering was done using Monte Carlo code (MCNP- 4C). It is simulated a collimated 60 keV beam of gamma rays emitted by {sup 241}Am source reaching wood blocks. Backscattered radiation by these blocks was calculated. Photons scattered were correlated with blocks of different wood densities. Results showed a linear relationship on wood density and scattered photons, therefore the viability of this wood densimeter. (author)

  14. Coupling MCNP-DSP and LAHET Monte Carlo codes for designing subcriticality monitors for accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.; Perez, R. [Oak Ridge National Lab., TN (United States); Rugama, Y.; Munoz-Cobo, J.L. [Poly. Tech. Univ. of Valencia (Spain). Chemical and Nuclear Engineering Dept.

    2001-07-01

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements. (orig.)

  15. Coupling MCNP-DSP and LAHET Monte Carlo Codes for Designing Subcriticality Monitors for Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Rugama, Y. Munoz-Cobos, J.; Perez, R.

    2000-10-23

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements.

  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. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  18. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  19. Voxel2MCNP: software for handling voxel models for Monte Carlo radiation transport calculations.

    Science.gov (United States)

    Hegenbart, Lars; Pölz, Stefan; Benzler, Andreas; Urban, Manfred

    2012-02-01

    Voxel2MCNP is a program that sets up radiation protection scenarios with voxel models and generates corresponding input files for the Monte Carlo code MCNPX. Its technology is based on object-oriented programming, and the development is platform-independent. It has a user-friendly graphical interface including a two- and three-dimensional viewer. A row of equipment models is implemented in the program. Various voxel model file formats are supported. Applications include calculation of counting efficiency of in vivo measurement scenarios and calculation of dose coefficients for internal and external radiation scenarios. Moreover, anthropometric parameters of voxel models, for instance chest wall thickness, can be determined. Voxel2MCNP offers several methods for voxel model manipulations including image registration techniques. The authors demonstrate the validity of the program results and provide references for previous successful implementations. The authors illustrate the reliability of calculated dose conversion factors and specific absorbed fractions. Voxel2MCNP is used on a regular basis to generate virtual radiation protection scenarios at Karlsruhe Institute of Technology while further improvements and developments are ongoing.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    A gamma spectrometer including an HPGe detector is commonly used for environmental radioactivity measurements. Many works have been focused on the simulation of the HPGe detector using Monte Carlo codes such as MCNP5. However, the simulation of this kind of detectors presents important difficulties due to the lack of information from manufacturers and due to loss of intrinsic properties in aging detectors. Some parameters such as the active volume or the Ge dead layer thickness are many times unknown and are estimated during simulations. In this work, a detailed model of an HPGe detector and a petri dish containing a certified gamma source has been done. The certified gamma source contains nuclides to cover the energy range between 50 and 1800 keV. As a result of the simulation, the Pulse Height Distribution (PHD) is obtained and the efficiency curve can be calculated from net peak areas and taking into account the certified activity of the source. In order to avoid errors due to the net area calculation, the simulated PHD is treated using the GammaVision software. On the other hand, it is proposed to use the Noether-Wilks formula to do an uncertainty analysis of model with the main goal of determining the efficiency curve of this detector and its associated uncertainty. The uncertainty analysis has been focused on dead layer thickness at different positions of the crystal. Results confirm the important role of the dead layer thickness in the low energy range of the efficiency curve. In the high energy range (from 300 to 1800 keV) the main contribution to the absolute uncertainty is due to variations in the active volume. (author)

  2. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

  3. 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

    want to express my sincere love, respect, and admiration for my wife, who motivated and supported me throughout this long endeavor; this document ...widely utilized radiation transport code is MCNP. First created at Los Alamos National Laboratory ( LANL ) in 1957, the code simulated neutral...explanation of the current capabilities of MCNP will occur within the next chapter of this document ; however, it is important to note that MCNP

  4. The New MCNP6 Depletion Capability

    Energy Technology Data Exchange (ETDEWEB)

    Fensin, Michael Lorne [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory

    2012-06-19

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.

  5. NEPHTIS: 2D/3D validation elements using MCNP4c and TRIPOLI4 Monte-Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Girardi, E. [EDF R and D/SINETICS, 1av du General de Gaulle, F92141 Clamart CEDEX (France); Damian, F.; Moiron-Groizard, M. [DEN/DM2S/SERMA/LCA, CEA Saclay, F91191 Gif-sur-Yvette CEDEX (France)

    2006-07-01

    High Temperature Reactors (HTRs) appear as a promising concept for the next generation of nuclear power applications. The CEA, in collaboration with AREVA-NP and EDF, is developing a core modeling tool dedicated to the prismatic block-type reactor. NEPHTIS (Neutronics Process for HTR Innovating System) is a deterministic codes system based on a standard two-steps Transport-Diffusion approach (APOLLO2/CRONOS2). Validation of such deterministic schemes usually relies on Monte-Carlo (MC) codes used as a reference. However, when dealing with large HTR cores the fission source stabilization is rather poor with MC codes. In spite of this, it is shown in this paper that MC simulations may be used as a reference for a wide range of configurations. The first part of the paper is devoted to 2D and 3D MC calculations of a HTR core with control devices. Comparisons between MCNP4c and TRIPOLI4 MC codes are performed and show very consistent results. Finally, the last part of the paper is devoted to the code to code validation of the NEPHTIS deterministic scheme. (authors)

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

    Science.gov (United States)

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

    2016-03-01

    Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in  125I and  103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as  125I and  103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for  103Pd and 10 cm for  125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for  192Ir and less than 1.2% for  137Cs between the three codes. PACS number(s): 87.56.bg.

  7. MCNP-DSP USERS MANUAL

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.

    2001-01-19

    The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from subcritical measurements. The code can be used to simulate a variety of subcritical measurements including source-driven noise analysis, Rossi-{alpha}, pulsed source, passive frequency analysis, multiplicity, and Feynman variance measurements. This code can be used to validate Monte Carlo methods and cross section data sets with subcritical measurements and replaces the use of point kinetics models for interpreting subcritical measurements.

  8. Production of energetic light fragments in extensions of the CEM and LAQGSM event generators of the Monte Carlo transport code MCNP6

    Science.gov (United States)

    Mashnik, Stepan G.; Kerby, Leslie M.; Gudima, Konstantin K.; Sierk, Arnold J.; Bull, Jeffrey S.; James, Michael R.

    2017-03-01

    We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N -particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM) used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤7 , in the case of CEM, and A ≤12 , in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Finally, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.

  9. Comparison of Fuel Temperature Coefficients of PWR UO{sub 2} Fuel from Monte Carlo Codes (MCNP6.1 and KENO6)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-O; Roh, Gyuhong; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a result, there was a difference within about 300-400 pcm between keff values at each enrichment due to the difference of codes and nuclear data used in the evaluations. The FTC was changed to be less negative with the increase of uranium enrichment, and it followed the form of asymptotic curve. However, it is necessary to perform additional study for investigating what factor causes the differences more than two standard deviation (2σ) among the FTCs at partial enrichment region. The interaction probability of incident neutron with nuclear fuel is depended on the relative velocity between the neutron and the target nuclei. The Fuel Temperature Coefficient (FTC) is defined as the change of Doppler effect with respect to the change in fuel temperature without any other change such as moderator temperature, moderator density, etc. In this study, the FTCs for UO{sub 2} fuel were evaluated by using MCNP6.1 and KENO6 codes based on a Monte Carlo method. In addition, the latest neutron cross-sections (ENDF/B-VI and VII) were applied to analyze the effect of these data on the evaluation of FTC, and nuclear data used in MCNP calculations were generated from the makxsf code. An evaluation of the Doppler effect and FTC for UO{sub 2} fuel widely used in PWR was conducted using MCNP6.1 and KENO6 codes. The ENDF/B-VI and VII were also applied to analyze what effect these data has on those evaluations. All cross-sections needed for MCNP calculation were produced using makxsf code. The calculation models used in the evaluations were based on the typical PWR UO{sub 2} lattice.

  10. 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

    CERN Document Server

    White, M 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 tran...

  11. Comparison of a laboratory spectrum of Eu-152 with results of simulation using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain); Gallardo, S. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: sergalbe@iqn.upv.es; Ortiz, J. [Laboratorio de Radiactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)

    2007-09-21

    Detectors used for gamma spectrometry must be calibrated for each geometry considered in environmental radioactivity laboratories. This calibration is performed using a standard solution containing gamma emitter sources. Nevertheless, the efficiency curves obtained are periodically checked using a source such as {sup 152}Eu emitting many gamma rays that cover a wide energy range (20-1500 keV). {sup 152}Eu presents a problem because it has a lot of peaks affected by True Coincidence Summing (TCS). Two experimental measures have been performed placing the source (a Marinelli beaker) at 0 and 10 cm from the detector. Both spectra are simulated by the MCNP 4C code, where the TCS is not reproduced. Therefore, the comparison between experimental and simulated peak net areas permits one to choose the most convenient peaks to check the efficiency curves of the detector.

  12. 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.

  13. 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)

  14. The application of the Monte-Carlo neutron transport code MCNP to a small "nuclear battery" system

    OpenAIRE

    Puigdellívol Sadurní, Roger

    2009-01-01

    The project consist in calculate the keff to a small nuclear battery. The code Monte- Carlo neutron transport code MCNP is used to calculate the keff. The calculations are done at the beginning of life to know the capacity of the core becomes critical in different conditions. These conditions are the study parameters that determine the criticality of the core. These parameters are the uranium enrichment, the coated particles (TRISO) packing factor and the size of the core. More...

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

    Energy Technology Data Exchange (ETDEWEB)

    Nasrabadi, M.N. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)], E-mail: mnnasri@kashanu.ac.ir; Jalali, M. [Isfahan Nuclear Science and Technology Research Institute, Atomic Energy organization of Iran (Iran, Islamic Republic of); Mohammadi, A. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)

    2007-10-15

    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{sub 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

    Science.gov (United States)

    Kazeminezhad, F.; Anghaie, S.

    2008-01-01

    Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

  1. Uncertainty analysis in the simulation of X-ray spectra in the diagnostic range using the MCNP5 code.

    Science.gov (United States)

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

    2011-01-01

    An accurate knowledge of the photonic spectra emitted by X-ray tubes in radiodiagnostics is essential to better estimate the imparted dose to patients and to improve the image quality obtained with these devices. In this work, several X-ray spectra have been simulated using the MCNP5 code to simulate X-ray production in a commercial device. To validate the Monte Carlo results, simulated spectra have been compared to those extracted from the IPEM 78 database. The uncertainty associated to some geometrical features of the tube and its effect on the simulated spectra has been analyzed using the Noether-Wilks formula. This analysis has been focused on the thickness of collimators, filters, shielding and barrel shutter. Furthermore, results show that the uncertainty due to geometrical parameters (0.98% in terms of Root Mean Squared) is higher than the statistical uncertainty associated to the MCNP5 calculations.

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

    NARCIS (Netherlands)

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

    2002-01-01

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

  3. 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.

  4. 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.

  5. 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

  6. A new MCNP{trademark} test set

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-09-01

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

  7. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

  9. 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)

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

    OpenAIRE

    Lida Gholamkar; Mahdi Sadeghi; Ali Asghar Mowlavi; Mitra Athari

    2016-01-01

    Introduction One of the best methods in the diagnosis and control of breast cancer is mammography. The importance of mammography is directly related to its value in the detection of breast cancer in the early stages, which leads to a more effective treatment. The purpose of this article was to calculate the X-ray spectrum in a mammography system with Monte Carlo codes, including MCNPX and MCNP5. Materials and Methods The device, simulated using the MCNP code, was Planmed Nuance digital mammog...

  11. Gamma ray shielding study of barium-bismuth-borosilicate glasses as transparent shielding materials using MCNP-4C code, XCOM program, and available experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Bagheri, Reza; Yousefinia, Hassan [Nuclear Fuel Cycle Research School (NFCRS), Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Moghaddam, Alireza Khorrami [Radiology Department, Paramedical Faculty, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)

    2017-02-15

    In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

  12. 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)

  13. MCNP: Multigroup/adjoint capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.; Redmond, E.L. II; Palmtag, S.P.; Hendricks, J.S.

    1994-04-01

    This report discusses various aspects related to the use and validity of the general purpose Monte Carlo code MCNP for multigroup/adjoint calculations. The increased desire to perform comparisons between Monte Carlo and deterministic codes, along with the ever-present desire to increase the efficiency of large MCNP calculations has produced a greater user demand for the multigroup/adjoint capabilities. To more fully utilize these capabilities, we review the applications of the Monte Carlo multigroup/adjoint method, describe how to generate multigroup cross sections for MCNP with the auxiliary CRSRD code, describe how to use the multigroup/adjoint capability in MCNP, and provide examples and results indicating the effectiveness and validity of the MCNP multigroup/adjoint treatment. This information should assist users in taking advantage of the MCNP multigroup/adjoint capabilities.

  14. 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.

  15. Photon attenuation coefficients of Heavy-Metal Oxide glasses by MCNP code, XCOM program and experimental data: A comparison study

    Science.gov (United States)

    El-Khayatt, A. M.; Ali, A. M.; Singh, Vishwanath P.

    2014-01-01

    The mass attenuation coefficients, μ/ρ, total interaction cross-section, σt, and mean free path (MFP) of some Heavy Metal Oxides (HMO) glasses, with potential applications as gamma ray shielding materials, have been investigated using the MCNP-4C code. Appreciable variations are noted for all parameters by changing the photon energy and the chemical composition of HMO glasses. The numerical simulations parameters are compared with experimental data wherever possible. Comparisons are also made with predictions from the XCOM program in the energy region from 1 keV to 100 MeV. Good agreement noticed indicates that the chosen Monte Carlo method may be employed to make additional calculations on the photon attenuation characteristics of different glass systems, a capability particularly useful in cases where no analogous experimental data exist.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gudowska, I.; Svensson, R. [Karolinska Inst. (Sweden). Dept. of Medical Radiation Physics]|[Huddinge Univ. Hospital, Stockholm (Sweden). Dept. of Medical Physics; Sorcini, B. [Karolinska Inst. (Sweden). Dept. of Medical Radiation Physics]|[Stockholm Univ. (Sweden)

    2001-07-01

    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.)

  17. The performance test of anti-scattering x-ray grid with inclined shielding material by MCNP code simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Jun Woo; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-06-15

    The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination.

  18. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  20. MCNP{trademark} Monte Carlo: A precis of MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Adams, K.J.

    1996-06-01

    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.

  1. MCNP and GADRAS Comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Klasky, Marc Louis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); James, Michael R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mayo, Douglas R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-04-19

    To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicability for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.

  2. 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.

  3. MCNP model for the many KE-Basin radiation sources

    Energy Technology Data Exchange (ETDEWEB)

    Rittmann, P.D.

    1997-05-21

    This document presents a model for the location and strength of radiation sources in the accessible areas of KE-Basin which agrees well with data taken on a regular grid in September of 1996. This modelling work was requested to support dose rate reduction efforts in KE-Basin. Anticipated fuel removal activities require lower dose rates to minimize annual dose to workers. With this model, the effects of component cleanup or removal can be estimated in advance to evaluate their effectiveness. In addition, the sources contributing most to the radiation fields in a given location can be identified and dealt with.

  4. 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)

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

    Directory of Open Access Journals (Sweden)

    Lida Gholamkar

    2016-09-01

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

  6. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    OpenAIRE

    2013-01-01

    Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations...

  7. RADBALLTECHNOLOGY TESTING AND MCNP MODELING OF THE TUNGSTEN COLLIMATOR

    Energy Technology Data Exchange (ETDEWEB)

    Farfan, E.

    2010-07-08

    The United Kingdom's National Nuclear Laboratory (NNL) has developed a remote, non-electrical, radiation-mapping device known as RadBall{trademark}, which can locate and quantify radioactive hazards within contaminated areas of the nuclear industry. RadBall{trademark} consists of a colander-like outer shell that houses a radiation-sensitive polymer sphere. The outer shell works to collimate radiation sources and those areas of the polymer sphere that are exposed react, becoming increasingly more opaque, in proportion to the absorbed dose. The polymer sphere is imaged in an optical-CT scanner, which produces a high resolution 3D map of optical attenuation coefficients. Subsequent analysis of the optical attenuation matrix provides information on the spatial distribution of sources in a given area forming a 3D characterization of the area of interest. RadBall{trademark} has no power requirements and can be positioned in tight or hard-to reach locations. The RadBall{trademark} technology has been deployed in a number of technology trials in nuclear waste reprocessing plants at Sellafield in the United Kingdom and facilities of the Savannah River National Laboratory (SRNL). This study focuses on the RadBall{trademark} testing and modeling accomplished at SRNL.

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

    Science.gov (United States)

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

    2011-03-01

    This work has been performed within the frame of the European Union ORAMED project (Optimisation of RAdiation protection for MEDical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

  9. Investigation of Anisotropy Caused by Cylinder Applicator on Dose Distribution around Cs-137 Brachytherapy Source using MCNP4C Code

    Directory of Open Access Journals (Sweden)

    Sedigheh Sina

    2011-06-01

    Full Text Available Introduction: Brachytherapy is a type of radiotherapy in which radioactive sources are used in proximity of tumors normally for treatment of malignancies in the head, prostate and cervix. Materials and Methods: The Cs-137 Selectron source is a low-dose-rate (LDR brachytherapy source used in a remote afterloading system for treatment of different cancers. This system uses active and inactive spherical sources of 2.5 mm diameter, which can be used in different configurations inside the applicator to obtain different dose distributions. In this study, first the dose distribution at different distances from the source was obtained around a single pellet inside the applicator in a water phantom using the MCNP4C Monte Carlo code. The simulations were then repeated for six active pellets in the applicator and for six point sources.  Results: The anisotropy of dose distribution due to the presence of the applicator was obtained by division of dose at each distance and angle to the dose at the same distance and angle of 90 degrees. According to the results, the doses decreased towards the applicator tips. For example, for points at the distances of 5 and 7 cm from the source and angle of 165 degrees, such discrepancies reached 5.8% and 5.1%, respectively.  By increasing the number of pellets to six, these values reached 30% for the angle of 5 degrees. Discussion and Conclusion: The results indicate that the presence of the applicator causes a significant dose decrease at the tip of the applicator compared with the dose in the transverse plane. However, the treatment planning systems consider an isotropic dose distribution around the source and this causes significant errors in treatment planning, which are not negligible, especially for a large number of sources inside the applicator.

  10. 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)

  11. 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)

  12. 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.

  13. Model Children's Code.

    Science.gov (United States)

    New Mexico Univ., Albuquerque. American Indian Law Center.

    The Model Children's Code was developed to provide a legally correct model code that American Indian tribes can use to enact children's codes that fulfill their legal, cultural and economic needs. Code sections cover the court system, jurisdiction, juvenile offender procedures, minor-in-need-of-care, and termination. Almost every Code section is…

  14. Specific absorbed fractions of photons calculated in voxel phantoms using the Monte Carlo EGS4 and MCNP4 codes; Fracoes absorvidas especificas de fotons calculadas em fantomas de voxels utilizando os codigos Monte Carlo EGS4 e MCNP4

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, R.; Khoury, H. J. [Pernambuco Univ. (UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Yoriyaz, H. [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil); Lima, F.R.A. [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN), Recife, PE (Brazil); Faculdade Boa Viagem (FBV), Recife, PE (Brazil)]. E-mail: rkramer@uol.com.br

    2004-07-01

    Dose coefficients for intakes of radionuclides published by the International Commission on Radiological Protection (ICRP) are based on specific absorbed fractions (SAFs), which have been calculated in the mathematical MIRD phantoms. The replacement of the MIRD phantoms by voxel phantoms proposed by the ICRP raises the question about the changes to be expected for the SAFs, and consequently also for the dose coefficients. In order to investigate the dosimetric consequences of this replacement, SAFs have been calculated in the recently introduced MAX (Male Adult voXel) and FAXht (Female Adult voXel) head + trunk phantoms for photon energies between 10 keV and 4 MeV. For this purpose the phantoms have been connected to the EGS4 as well as to the MCNP4 code, which at present are probably the most used general-purpose Monte Carlo codes. Thereby it was possible to assess the impact on the SAFs, if different radiation transport methods are used. The mathematical MIRD phantoms have also been connected to the EGS4 code, and their elemental compositions of body tissues were replaced by those used in the voxel phantoms. In this manner it was possible to compare the SAFs of the MIRD phantoms on the one hand and the MAX and FAX phantom on the other hand as a function of the geometrical anatomy only, i.e. the volume, the shape and the position of organs at risk. (author)

  15. CONDOR-CITVAP-MCNP calculation line description

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo Anibal [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    A general description of the CONDOR-CITVAP-MCNP calculation line is given. This calculation line starts at cross section library and allows burnup dependent detailed calculation using MCNP. This calculation line is divided in two main methodologies: CONDOR-CITVAP that allows the 3-Dimensional core burnup calculation and MCNP that performs detailed transport calculations, both methodologies are coupled using the NDDUMP code. A short description of the used codes are given: CONDOR code performs the cell calculation, generating burnup dependent macroscopic cross section and burnup dependent numerical densities per material. CITVAP codes perform the burnup dependent core calculation, including the fuel management and calculates the burnup distribution per material. NDDUMP code generates materials burnup dependent numerical densities to be used by MCNP code. This paper presents a detailed description of the CONDOR-CITVAP-MCNP calculation line and a numerical comparison of the proposed methodology. (author)

  16. Validation suite for MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R. D. (Russell D.)

    2002-01-01

    Two validation suites, one for criticality and another for radiation shielding, have been defined and tested for the MCNP Monte Carlo code. All of the cases in the validation suites are based on experiments so that calculated and measured results can be compared in a meaningful way. The cases in the validation suites are described, and results from those cases are discussed. For several years, the distribution package for the MCNP Monte Carlo code1 has included an installation test suite to verify that MCNP has been installed correctly. However, the cases in that suite have been constructed primarily to test options within the code and to execute quickly. Consequently, they do not produce well-converged answers, and many of them are physically unrealistic. To remedy these deficiencies, sets of validation suites are being defined and tested for specific types of applications. All of the cases in the validation suites are based on benchmark experiments. Consequently, the results from the measurements are reliable and quantifiable, and calculated results can be compared with them in a meaningful way. Currently, validation suites exist for criticality and radiation-shielding applications.

  17. 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.

  18. The MCNP6 Analytic Criticality Benchmark Suite

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes Group

    2016-06-16

    Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling) and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.

  19. MCNP6 Fission Multiplicity with FMULT Card

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, Trevor [Los Alamos National Laboratory; Fensin, Michael Lorne [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; McKinney, Gregg W. [Los Alamos National Laboratory

    2012-06-18

    With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

  20. Semi-Analytical Benchmarks for MCNP6

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-07

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

  1. 基于 MELCOR 与 MCNP 程序的安全壳剂量率计算方法%Calculating Method of Containment Dose Rate Based on MELCOR and MCNP Codes

    Institute of Scientific and Technical Information of China (English)

    史晓磊; 许倩; 魏严凇; 季松涛

    2015-01-01

    严重事故条件下,评估安全壳内的放射性剂量率水平对核电厂严重事故管理、应急响应等环节具有重要指导意义。本工作利用M ELCOR程序模拟严重事故序列,计算不同核素组释放进入安全壳内的质量;利用ORIGEN2程序计算不同核素组的堆芯积存量及核素的γ源强;利用MCNP程序计算每组核素100%释放进入安全壳所产生的剂量率水平;最后根据拟合公式求解安全壳剂量率。中核核电运行管理有限公司30万千瓦机组安全壳剂量率的计算结果说明该方法切实可行。%It is important to evaluate the containment dose rate under severe accident conditions for some aspects of a nuclear power plant ,such as severe accident manage‐ment and emergency response .In this work ,the MELCOR code was used to simulate the sequence of severe accidents , calculate masses of radioactive fission products released to containment .The ORIGEN2 code was used to calculate the γsource intensity . The MCNP code was used to calculate the containment dose rate when each group of radioactive fission products was all released to containment .The containment dose rate was finally calculated by a fitting formula .This method was used in the 300 MW units of CNNP Nuclear Power Operations Management Co .Ltd and was proved to be available .

  2. MCNP-REN a Monte Carlo tool for neutron detector design

    CERN Document Server

    Abhold, M E

    2002-01-01

    The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel w...

  3. 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.

  4. TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Cayla, Pierre-Yves; Fausser, Clement [MILLENNIUM, 16 Av du Québec Silic 628, F-91945 Villebon sur Yvette (France); Damian, Frederic; Lee, Yi-Kang; Puma, Antonella Li; Trama, Jean-Christophe [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France)

    2014-10-15

    3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4{sup ®} is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4{sup ®}, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4{sup ®} A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4{sup ®} is shown; discrepancies are mainly included in the statistical error.

  5. MCNP6 model of the University of Washington clinical neutron therapy system (CNTS)

    Science.gov (United States)

    Moffitt, Gregory B.; Stewart, Robert D.; Sandison, George A.; Goorley, John T.; Argento, David C.; Jevremovic, Tatjana

    2016-01-01

    A MCNP6 dosimetry model is presented for the Clinical Neutron Therapy System (CNTS) at the University of Washington. In the CNTS, fast neutrons are generated by a 50.5 MeV proton beam incident on a 10.5 mm thick Be target. The production, scattering and absorption of neutrons, photons, and other particles are explicitly tracked throughout the key components of the CNTS, including the target, primary collimator, flattening filter, monitor unit ionization chamber, and multi-leaf collimator. Simulations of the open field tissue maximum ratio (TMR), percentage depth dose profiles, and lateral dose profiles in a 40 cm  ×  40 cm  ×  40 cm water phantom are in good agreement with ionization chamber measurements. For a nominal 10  ×  10 field, the measured and calculated TMR values for depths of 1.5 cm, 5 cm, 10 cm, and 20 cm (compared to the dose at 1.7 cm) are within 0.22%, 2.23%, 4.30%, and 6.27%, respectively. For the three field sizes studied, 2.8 cm  ×  2.8 cm, 10.4 cm  ×  10.3 cm, and 28.8 cm  ×  28.8 cm, a gamma test comparing the measured and simulated percent depth dose curves have pass rates of 96.4%, 100.0%, and 78.6% (depth from 1.5 to 15 cm), respectively, using a 3% or 3 mm agreement criterion. At a representative depth of 10 cm, simulated lateral dose profiles have in-field (⩾10% of central axis dose) pass rates of 89.7% (2.8 cm  ×  2.8 cm), 89.6% (10.4 cm  ×  10.3 cm), and 100.0% (28.8 cm  ×  28.8 cm) using a 3% and 3 mm criterion. The MCNP6 model of the CNTS meets the minimum requirements for use as a quality assurance tool for treatment planning and provides useful insights and information to aid in the advancement of fast neutron therapy.

  6. 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)

  7. 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

  8. MCNP APPLICATIONS FOR THE 21ST CENTURY

    Energy Technology Data Exchange (ETDEWEB)

    G. MCKINNEY; T. BOOTH; ET AL

    2000-10-01

    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.

  9. MCNP application for the 21 century

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, M.C. [and others

    2000-08-01

    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.

  10. New methods for neutron response calculations with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J.S. [Los Alamos National Lab., NM (United States). Applied Theoretical and Computational Physics Div.

    1997-05-01

    MCNP4B was released for international distribution in February, 1997. The author summarized the new MCNP4B features since the release of MCNP4A over three years earlier and compare some results. Then he describes new methods being developed for future code releases. The focus is methods and applications of ex-core neutron response calculations.

  11. MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-22

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

  12. 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.

  13. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  14. 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.

  15. Determination of dosimetric parameters for 125I seed source using MCNP5 and EGSnrc MC codes%MCNP5与EGSnrc比较计算125I种子源剂量参数

    Institute of Scientific and Technical Information of China (English)

    曹振; 阮锡超; 孟贝蒂; 石翠燕

    2014-01-01

    根据AAPM TG43U1的推荐,使用MCNP5与EGSnrc两种蒙特卡罗程序计算6711型125I种子源剂量计算参数,并将两者计算结果和AAPM推荐值比较,得到相对偏差结果如下:剂量率常数,MCNP5为0.62%,EGSnrc为2.07%;径向剂量函数,MNCP5为0.15%-5.12%,EGSnrc为0%-2.18%.两者计算结果均与推荐值符合得很好,而EGSnrc的计算结果更具优势.

  16. 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)

  17. 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)

  18. Evaluation of the thermal neutron flux in the core of IPEN/MB-01 reactor using the code Monte Carlo (MCNP)

    Energy Technology Data Exchange (ETDEWEB)

    Salome, Jean A.D.; Cardoso, Fabiano; Faria, Rochkhudson B.; Pereira, Claubia, E-mail: jadsalome@yahoo.com.br, E-mail: fabinuclear@yahoo.com.br, E-mail: rockdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The IPEN/MB-01 reactor, located in the city of Sao Paulo - Brazil, reached its first criticality on the year of 1988. The reactor is characterized by a low output power of 100 W only, even because its purpose is to produce knowledge about nuclear power plants on a smaller geometric scale without the requirement of an extremely complex cooling system. The use of devices such as this it is very interesting because it achieves the demands of nuclear engineering about the neutronic parameters needed in the design of large nuclear plants through relatively simple and inexpensive methods. In this paper, the computational mathematical code MCNP5 is used to perform the calculation of the thermal neutron flux in the core of the IPEN/MB-01 reactor. To do this is used an experiment from the LEU-COMP-THERM-077 benchmark that represents the standard rectangular configuration of the IPEN/MB-01 reactor. The thermal neutron flux is calculated at some axial planes of different heights and, after that, axial profiles of the thermal neutron flux are done and compared to experimental results issued previously. The experimental values used as reference refer to a cylindrical configuration of the core of the reactor. Finally, the pertinence and relevance of the results are checked. With this work is expected to produce more knowledge about the dynamics of neutron flux in the core of the IPEN/MB-01 reactor. (author)

  19. Study of the source-detector system geometry using the MCNP-X code in the flowrate measurement with radioactive tracers

    Energy Technology Data Exchange (ETDEWEB)

    Avilan Puertas, Eddie, E-mail: epuertas@nuclear.ufrj.br [Universidad Central de Venezuela (UCV), Facultad de Ingenieria, Departamento de Fisica Aplicada, Caracas (Venezuela, Bolivarian Republic of); Braz, Delson, E-mail: delson@lin.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Brandao, Luis E.; Salgado, Cesar M., E-mail: brandao@ien.gov.br, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The use radioactive tracers for flow rate measurement is applied to a great variety of situations, however the accuracy of the technique is highly dependent of the adequate choice of the experimental measurement conditions. To measure flow rate of fluids in ducts partially filled, is necessary to measure the fluid flow velocity and the fluid height. The flow velocity can be measured with the cross correlation function and the fluid level, with a fluid level meter system. One of the error factors when measuring flow rate, is on the correct setting of the source-detector of the fluid level meter system. The goal of the present work is to establish by mean of MCNP-X code simulations the experimental parameters to measure the fluid level. The experimental tests will be realized in a flow rate system of 10 mm of diameter of acrylic tube for water and oil as fluids. The radioactive tracer to be used is the {sup 82}Br and for the detection will be employed two 1″ NaI(Tl) scintillator detectors, shielded with collimators of 0.5 cm and 1 cm of circular aperture diameter. (author)

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

    Science.gov (United States)

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

    2008-09-01

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

  1. PARALLELIZATION AND PERFECTION OF MCNP MONTE CARLO PARTICLE TRANSPORT CODE IN MPI%粒子输运蒙特卡罗程序MCNP在MPI下的并行化及完善

    Institute of Scientific and Technical Information of China (English)

    邓力; 刘杰; 张文勇

    2003-01-01

    The particle transport Monte Carlo code MCNP had been realized the paral-lelization in MPI (Message Passing Interface) in 1999. But due to adopting the leap random number producer, some differences were existed between the parallel result and the serial result. Now the same results have been achieved by using the segment random number. The speedup of the applied problem is the liner ups to 53 in 64-Processors and the parallel efficiencv is up to 83% in 64-Processors.

  2. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    CERN Document Server

    Mirfayzi, S R

    2013-01-01

    Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally ...

  3. Determination of the dead layer and full-energy peak efficiency of an HPGe detector using the MCNP code and experimental results

    Directory of Open Access Journals (Sweden)

    M Moeinifar

    2017-02-01

    Full Text Available One important factor in using an High Purity Germanium (HPGe detector is its efficiency that highly depends on the geometry and absorption factors, so that when the configuration of source-detector geometry is changed, the detector efficiency must be re-measured. The best way of determining the efficiency of a detector is measuring the efficiency of standard sources. But considering the fact that standard sources are hardly available and it is time consuming to find them, determinig the efficiency by simulation which gives enough efficiency in less time, is important. In this study, the dead layer thickness and the full-energy peak efficiency of an HPGe detector was obtained by Monte Carlo simulation, using MCNPX code. For this, we first measured gamma–ray spectra for different sources placed at various distances from the detector and stored the measured spectra obtained. Then the obtained spectra were simulated under similar conditions in vitro.At first, the whole volume of germanium was regarded as active, and the obtaind spectra from calculation were compared with the corresponding experimental spectra. Comparison of the calculated spectra with the measured spectra showed considerable differences. By making small variations in the dead layer thickness of the detector (about a few hundredths of a millimeter in the simulation program, we tried to remove these differences and in this way a dead layer of 0.57 mm was obtained for the detector. By incorporating this value for the dead layer in the simulating program, the full-energy peak efficiency of the detector was then obtained both by experiment and by simulation, for various sources at various distances from the detector, and both methods showed good agreements. Then, using MCNP code and considering the exact measurement system, one can conclude that the efficiency of an HPGe detector for various source-detector geometries can be calculated with rather good accuracy by simulation method

  4. MCNP4B{sup {trademark}} verification and validation

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J.S.; Court, J.D.

    1996-08-01

    Several new features and bug fixes have been incorporated into the new release of MCNP. As required by the MCNP Software Quality Assurance Plan, these changes to the code and the test set are documented here for user reference. This document summarizes the new MCNP4B features and corrections, separated into major and minor groupings. Also included are a code cleanup section and a section delineating problems identified in LA-12839 which have not been corrected. Finally, we document the MCNP4B test set modifications and explain how test set coverage has been improved.

  5. Benchmarking MCNP and TRIPOLI with PGNAA measurements

    Science.gov (United States)

    Carasco, C.; Perot, B.; Sikora, A.; Mauerhofer, E.; Havenith, A.; Payan, E.; Kettler, J.; Kring, T.; Ma, J. L.

    2014-06-01

    The French Alternative Energies and Atomic Energy Commission (CEA Cadarache), the Forschungszentrum Jülich GmbH (FZJ), and the RWTH Aachen University (RWTH) 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 design of an optimized measurement system and the assessment of its performances for realistic scenarios can be conveniently studied by numerical Monte Carlo simulation, provided the model and nuclear data offer a sufficient precision. Previous studies performed with MCNP have shown that when the nuclear data libraries lack of precision, relevant results can still be obtained by performing calculations in multiple steps (by first determining the radiative capture rate, and transporting the induced gamma toward the detector) and by injecting valid gamma-ray production data in-between [1]. In such cases, it is interesting to compare the results obtained with different codes. In the present paper, we propose to compare the MCNP and TRIPOLI codes with measurements obtained in MEDINA (Multi Element Detection based on Instrumental Neutron Activation), which is the new FZJ PGNAA facility [2]. The aim of the measurement campaign is to assess capture gamma rays of toxic elements that can be found in 200 L waste drums which are expected for geological repository.

  6. Design of Light Multi-layered Shields for Use in Diagnostic Radiology and Nuclear Medicine via MCNP5 Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Mehdi Zehtabian

    2015-09-01

    Full Text Available Introduction Lead-based shields are the most widely used attenuators in X-ray and gamma ray fields. The heavy weight, toxicity and corrosion of lead have led researchers towards the development of non-lead shields. Materials and Methods The purpose of this study was to design multi-layered shields for protection against X-rays and gamma rays in diagnostic radiology and nuclear medicine. In this study, cubic slabs composed of several materials with high atomic numbers, i.e., lead, barium, bismuth, gadolinium, tin and tungsten, were simulated, using MCNP5 Monte Carlo code. Cubic slabs (30×30×0.05 cm3 were simulated at a 50 cm distance from the point photon source. The X-ray spectra of 80 kVp and 120 kVp were obtained, using IPEM Report 78. The photon flux following the use of each shield was obtained inside cubic tally cells (1×1×0.5 cm3 at a 5 cm distance from the shields. The photon attenuation properties of multi-layered shields (i.e., two, three, four and five layers, composed of non-lead radiation materials, were also obtained via Monte Carlo simulations. Results Among different shield designs proposed in this study, the three-layered shield, composed of tungsten, bismuth and gadolinium, showed the most significant attenuation properties in radiology, with acceptable shielding at 140 keV energy in nuclear medicine. Conclusion According to the results, materials with k-edges equal to energies common to diagnostic radiology can be proper substitutes for lead shields.

  7. Use of the MCNP code for analysis of the attenuation of the radiation produced by radioactive sources used in radiotherapy in skin tumors; Uso do codigo MCNP para analise da atenuacao da radiacao produzida por fontes radioativas utilizadas em radioterapia em tumores de pele

    Energy Technology Data Exchange (ETDEWEB)

    Tada, A., E-mail: ariane.tada@gmail.co [Instituto de Pesquisas Energeticas e Nucleares (CEN/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil); Salles, T.; Yoriyaz, H., E-mail: hyoriyaz@ipen.b, E-mail: tasallesc@gmail.co [Instituto de Pesquisas Energeticas e Nucleares (CEN/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Fernandes, M.A.R, E-mail: marfernandes@fmb.unesp.b [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Dept. de Dermatologia e Radioterapia

    2010-07-01

    The present work had as objective to analyze the distribution profile of a therapeutic dose of radiation produced by radioactive sources used in radiotherapy procedures in superficial lesions on the skin. The experimental measurements for analysis of dosimetric radiation sources were compared with calculations obtained from the computer system based on the Monte Carlo Method. The results obtained by the computations calculations using the code MCNP-4C showed a good agreement with the experimental measurements. A comparison of different treatment modalities allows an indication of more appropriate procedures for each clinical case. (author)

  8. Assessment of Mean Glandular Dose in Mammography System with Different Anode-Filter Combinations Using MCNP Code.

    Science.gov (United States)

    Gholamkar, Lida; Mowlavi, Ali Asghar; Sadeghi, Mahdi; Athari, Mitra

    2016-10-01

    X-ray mammography is one of the general methods for early detection of breast cancer. Since glandular tissue in the breast is sensitive to radiation and it increases the risk of cancer, the given dose to the patient is very important in mammography. The aim of this study was to determine the average absorbed dose of X-ray radiation in the glandular tissue of the breast during mammography examinations as well as investigating factors that influence the mean glandular dose (MGD). One of the precise methods for determination of MGD absorbed by the breast is Monte Carlo simulation method which is widely used to assess the dose. We studied some different X-ray sources and exposure factors that affect the MGD. "Midi-future" digital mammography system with amorphous-selenium detector was simulated using the Monte Carlo N-particle extended (MCNPX) code. Different anode/filter combinations such as tungsten/silver (W/Ag), tungsten/rhodium (W/Rh), and rhodium/aluminium (Rh/Al) were simulated in this study. The voltage of X-ray tube ranged from 24 kV to 32 kV with 2 kV intervals and the breast phantom thickness ranged from 3 to 8 cm, and glandular fraction g varied from 10% to 100%. MGD was measured for different anode/filter combinations and the effects of changing tube voltage, phantom thickness, combination and glandular breast tissue on MGD were studied. As glandular g and X-ray tube voltage increased, the breast dose increased too, and the increase of breast phantom thickness led to the decrease of MGD. The obtained results for MGD were consistent with the result of Boone et al. that was previously reported. By comparing the results, we saw that W/Rh anode/filter combination is the best choice in breast mammography imaging because of the lowest delivered dose in comparison with W/Ag and Rh/Al. Moreover, breast thickness and g value have significant effects on MGD.

  9. High-fidelity MCNP modeling of a D-T neutron generator for active interrogation of special nuclear material

    Science.gov (United States)

    Katalenich, Jeff; Flaska, Marek; Pozzi, Sara A.; Hartman, Michael R.

    2011-10-01

    Fast and robust methods for interrogation of special nuclear material (SNM) are of interest to many agencies and institutions in the United States. It is well known that passive interrogation methods are typically sufficient for plutonium identification because of a relatively high neutron production rate from 240Pu [1]. On the other hand, identification of shielded uranium requires active methods using neutron or photon sources [2]. Deuterium-deuterium (2.45 MeV) and deuterium-tritium (14.1 MeV) neutron-generator sources have been previously tested and proven to be relatively reliable instruments for active interrogation of nuclear materials [3,4]. In addition, the newest generators of this type are small enough for applications requiring portable interrogation systems. Active interrogation techniques using high-energy neutrons are being investigated as a method to detect hidden SNM in shielded containers [4,5]. Due to the thickness of some containers, penetrating radiation such as high-energy neutrons can provide a potential means of probing shielded SNM. In an effort to develop the capability to assess the signal seen from various forms of shielded nuclear materials, the University of Michigan Neutron Science Laboratory's D-T neutron generator and its shielding were accurately modeled in MCNP. The generator, while operating at nominal power, produces approximately 1×10 10 neutrons/s, a source intensity which requires a large amount of shielding to minimize the dose rates around the generator. For this reason, the existing shielding completely encompasses the generator and does not include beam ports. Therefore, several MCNP simulations were performed to estimate the yield of uncollided 14.1-MeV neutrons from the generator for active interrogation experiments. Beam port diameters of 5, 10, 15, 20, and 25 cm were modeled to assess the resulting neutron fluxes. The neutron flux outside the beam ports was estimated to be approximately 2×10 4 n/cm 2 s.

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

    Science.gov (United States)

    Pauzi, A. M.

    2013-06-01

    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-batteryTM", 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.

  11. 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)

  12. MCNP5 and GEANT4 comparisons for preliminary Fast Neutron Pencil Beam design at the University of Utah TRIGA system

    Science.gov (United States)

    Adjei, Christian Amevi

    The main objective of this thesis is twofold. The starting objective was to develop a model for meaningful benchmarking of different versions of GEANT4 against an experimental set-up and MCNP5 pertaining to photon transport and interactions. The following objective was to develop a preliminary design of a Fast Neutron Pencil Beam (FNPB) Facility to be applicable for the University of Utah research reactor (UUTR) using MCNP5 and GEANT4. The three various GEANT4 code versions, GEANT4.9.4, GEANT4.9.3, and GEANT4.9.2, were compared to MCNP5 and the experimental measurements of gamma attenuation in air. The average gamma dose rate was measured in the laboratory experiment at various distances from a shielded cesium source using a Ludlum model 19 portable NaI detector. As it was expected, the gamma dose rate decreased with distance. All three GEANT4 code versions agreed well with both the experimental data and the MCNP5 simulation. Additionally, a simple GEANT4 and MCNP5 model was developed to compare the code agreements for neutron interactions in various materials. Preliminary FNPB design was developed using MCNP5; a semi-accurate model was developed using GEANT4 (because GEANT4 does not support the reactor physics modeling, the reactor was represented as a surface neutron source, thus a semi-accurate model). Based on the MCNP5 model, the fast neutron flux in a sample holder of the FNPB is obtained to be 6.52×107 n/cm2s, which is one order of magnitude lower than gigantic fast neutron pencil beam facilities existing elsewhere. The MCNP5 model-based neutron spectrum indicates that the maximum expected fast neutron flux is at a neutron energy of ~1 MeV. In addition, the MCNP5 model provided information on gamma flux to be expected in this preliminary FNPB design; specifically, in the sample holder, the gamma flux is to be expected to be around 108 γ/cm 2s, delivering a gamma dose of 4.54×103 rem/hr. This value is one to two orders of magnitudes below the gamma

  13. MCNP modeling of a neutron generator and its shielding at Missouri University of Science and Technology

    Science.gov (United States)

    Sharma, Manish K.; Alajo, Ayodeji Babatunde; Liu, Xin

    2014-12-01

    The shielding of a neutron generator producing fast neutrons should be sufficient to limit the dose rates to the prescribed values. A deuterium-deuterium neutron generator has been installed in the Nuclear Engineering Department at Missouri University of Science and Technology (Missouri S&T). The generator produces fast neutrons with an approximate energy of 2.5 MeV. The generator is currently shielded with different materials like lead, high-density polyethylene, and borated polyethylene. An MCNP transport simulation has been performed to estimate the dose rates at various places in and around the facility. The simulations incorporated the geometric and composition information of these shielding materials to determine neutron and photon dose rates at three central planes passing through the neutron source. Neutron and photon dose rate contour plots at these planes were provided using a MATLAB program. Furthermore, the maximum dose rates in the vicinity of the facility were used to estimate the annual limit for the generator's hours of operation. A successful operation of this generator will provide a convenient neutron source for basic and applied research at the Nuclear Engineering Department of Missouri S&T.

  14. FREYA-a new Monte Carlo code for improved modeling of fission chains

    Energy Technology Data Exchange (ETDEWEB)

    Hagmann, C A; Randrup, J; Vogt, R L

    2012-06-12

    A new simulation capability for modeling of individual fission events and chains and the transport of fission products in materials is presented. FREYA ( Fission Yield Event Yield Algorithm ) is a Monte Carlo code for generating fission events providing correlated kinematic information for prompt neutrons, gammas, and fragments. As a standalone code, FREYA calculates quantities such as multiplicity-energy, angular, and gamma-neutron energy sharing correlations. To study materials with multiplication, shielding effects, and detectors, we have integrated FREYA into the general purpose Monte Carlo code MCNP. This new tool will allow more accurate modeling of detector responses including correlations and the development of SNM detectors with increased sensitivity.

  15. Biological Shielding Design Effectiveness of the Brachytherapy Unit at the Korle Bu Teaching Hospital in Ghana Using Mcnp5 Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    C.C. Arwui

    2011-05-01

    Full Text Available Design objectives for brachytherapy treatment facilities require sufficient shielding to reduce primary and scatter radiation to design limit in order to limit exposure to patients, staff and the general public. The primary aim of this study is to verify whether shielding of the brachytherapy unit at the Korle Bu teaching Hospital in Ghana provides adequate protection in order to assess any radiological health and safety impact and also test the suitability of other available sources. The study evaluates the effectiveness of the biological shielding design of a Cs-137 brachytherapy unit at the Korle-Bu Teaching Hospital in Ghana using MCNP5. The facility was modeled based on the design specifications for LDR Cs-137, MDR Cs-137, HDR Co-60 and HDR Ir-192 treatment modalities. The estimated dose rate ranged from (0.01-0.15 μSv/h and (0.37-3.05 μSv/h for the existing initial and decayed activities of LDR Cs-137 for the public and controlled areas respectively, (0.03-0.57 μSv/h and (1.53-8.06 μSv/h for MDR Cs-137, (7.47-59.46 μSv/h and (144.87-178.74 μSv/h for HDR Co- 60, (0.13-6.95 μSv/h and (19.47-242.98 μSv/h for HDR Ir-192 for the public and controlled areas respectively. The results were verified by dose rates measurement for the current LDR setup at the Brachytherapy unit and agreed quiet well. It was also compared with the reference values of 0.5 μSv/h for public areas and 7.5 μSv/h for controlled areas respectively. It can be concluded that the shielding is adequate for the existing source.

  16. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  17. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  18. CEM03.03 and LAQGSM03.03 Event Generators for the MCNP6, MCNPX, and MARS15 Transport Codes

    CERN Document Server

    Mashnik, S G; Prael, R E; Sierk, A J; Baznat, M I; Mokhov, N V

    2008-01-01

    A description of the IntraNuclear Cascade (INC), preequilibrium, evaporation, fission, coalescence, and Fermi breakup models used by the latest versions of our CEM03.03 and LAQGSM03.03 event generators is presented, with a focus on our most recent developments of these models. The recently developed "S" and "G" versions of our codes, that consider multifragmentation of nuclei formed after the preequilibrium stage of reactions when their excitation energy is above 2A MeV using the Statistical Multifragmentation Model (SMM) code by Botvina et al. ("S" stands for SMM) and the fission-like binary-decay model GEMINI by Charity ("G" stands for GEMINI), respectively, are briefly described as well. Examples of benchmarking our models against a large variety of experimental data on particle-particle, particle-nucleus, and nucleus-nucleus reactions are presented. Open questions on reaction mechanisms and future necessary work are outlined.

  19. Evaluation of the thermal neutron flux in samples of Al–Au alloy irradiated in the carrousel channels of the TRIGA MARK I IPR-R1 reactor using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Salomé, J.A.D.; Guerra, B.T. [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Pereira, C., E-mail: claubia@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Menezes, M.Â.B.C. de [Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG, Av. Antônio Carlos, 6627 31270-901, P.O. Box 941, Belo Horizonte, MG (Brazil); Silva, C.A.M. da [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Dalle, H.M. [Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG, Av. Antônio Carlos, 6627 31270-901, P.O. Box 941, Belo Horizonte, MG (Brazil)

    2014-07-01

    Highlights: • The TRIGA IPR-R1 was modelled using MCNP. • The thermal neutron flux through the samples in eleven irradiation channels was obtained. • The simulated results were compared to experimental values. • The relative error, the relative trend, the z-score test and uncertainty were analysed. - Abstract: The TRIGA IPR-R1 was modelled using MCNP. The model consists of a cylinder filled with water, fuel elements, radial reflectors, central tube, control rods and neutron source. Around the core is placed the Rotary Specimen Rack (RSR) with adequate groove to insert the samples to irradiation. The values of the thermal neutron flux through the samples in eleven irradiation channels were simulated and compared to the experimental results to validate the model. After that, the values of the thermal neutron flux, in the same channels, were simulated on two horizontal planes at different heights and compared to validate the model. These channels were characterized as representative channels of the neutron flux distribution in the RSR. To evaluate the results, the relative errors, the relative trend, the z-score test and the relevance to a confidence interval of 95% were analysed. Good agreement has been obtained for the most channels when compared with the experimental results.

  20. 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.

  1. MCNP6 Cosmic & Terrestrial Background Particle Fluxes -- Release 4

    Energy Technology Data Exchange (ETDEWEB)

    McMath, Garrett E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.; McKinney, Gregg W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.; Wilcox, Trevor [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.

    2015-01-23

    Essentially a set of slides, the presentation begins with the MCNP6 cosmic-source option, then continues with the MCNP6 transport model (atmospheric, terrestrial) and elevation scaling. It concludes with a few slides on results, conclusions, and suggestions for future work.

  2. MOCUP: MCNP-ORIGEN2 coupled utility program

    Energy Technology Data Exchange (ETDEWEB)

    Moore, R.L.; Schnitzler, B.G.; Wemple, C.A. [and others

    1995-09-30

    MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices.

  3. RadBall{sup TM} Technology Testing and MCNP Modeling of the Tungsten Collimator

    Energy Technology Data Exchange (ETDEWEB)

    Farfan, Eduardo B; Foley, Trevor Q; Coleman, J Rusty; Jannik, G Timothy; Holmes, Christopher J; Oldham, Mark; Adamovics, John; Stanley, Steven J, E-mail: Eduardo.Farfan@srnl.doe.go

    2010-11-01

    The UK's National Nuclear Laboratory (NNL) has developed a remote, non-electrical, radiation-mapping device known as RadBall{sup TM}, which can locate and quantify radioactive hazards within contaminated areas of the nuclear industry. RadBall{sup TM} consists of a colander-like outer shell that houses a radiation-sensitive polymer sphere. The outer shell works to collimate radiation sources and those areas of the polymer sphere that are exposed react, becoming increasingly more opaque, in proportion to the absorbed dose. The polymer sphere is imaged in an optical-CT scanner, which produces a high resolution 3D map of optical attenuation coefficients. Subsequent analysis of the optical attenuation matrix provides information on the spatial distribution of sources in a given area forming a 3D characterization of the area of interest. RadBall{sup TM} has no power requirements and can be positioned in tight or hard-to reach locations. The RadBall{sup TM} technology has been deployed in a number of technology trials in nuclear waste reprocessing plants at Sellafield in the UK and facilities of the Savannah River National Laboratory (SRNL). This study focuses on the RadBall{sup TM} testing and modeling accomplished at SRNL.

  4. Cheetah: Starspot modeling code

    Science.gov (United States)

    Walkowicz, Lucianne; Thomas, Michael; Finkestein, Adam

    2014-12-01

    Cheetah models starspots in photometric data (lightcurves) by calculating the modulation of a light curve due to starspots. The main parameters of the program are the linear and quadratic limb darkening coefficients, stellar inclination, spot locations and sizes, and the intensity ratio of the spots to the stellar photosphere. Cheetah uses uniform spot contrast and the minimum number of spots needed to produce a good fit and ignores bright regions for the sake of simplicity.

  5. MCNP6 fragmentation of light nuclei at intermediate energies

    CERN Document Server

    Mashnik, Stepan G

    2014-01-01

    Fragmentation reactions induced on light target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the latest Los Alamos Monte Carlo transport code MCNP6 and with its cascade-exciton model (CEM) and Los Alamos version of the quark-gluon string model (LAQGSM) event generators, version 03.03, used as stand-alone codes. Such reactions are involved in different applications, like cosmic-ray-induced single event upsets (SEU's), radiation protection, and cancer therapy with proton and ion beams, among others; therefore, it is important that MCNP6 simulates them as well as possible. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after INC. Both CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to He4 from energetic nucleons ...

  6. Investigation of the Effects of Tissue Inhomogeneities on the Dosimetric Parameters of a Cs-137 Brachytherapy Source using the MCNP4C Code

    Directory of Open Access Journals (Sweden)

    Mehdi Zehtabian

    2010-09-01

    Full Text Available Introduction: Brachytherapy is the use of small encapsulated radioactive sources in close vicinity of tumors. Various methods are used to obtain the dose distribution around brachytherapy sources. TG-43 is a dosimetry protocol proposed by the AAPM for determining dose distributions around brachytherapy sources. The goal of this study is to update this protocol for presence of bone and air inhomogenities.  Material and Methods: To update the dose rate constant parameter of the TG-43 formalism, the MCNP4C simulations were performed in phantoms composed of water-bone and water-air combinations. The values of dose at different distances from the source in both homogeneous and inhomogeneous phantoms were estimated in spherical tally cells of 0.5 mm radius using the F6 tally. Results: The percentages of dose reductions in presence of air and bone inhomogenities for the Cs-137 source were found to be 4% and 10%, respectively. Therefore, the updated dose rate constant (Λ will also decrease by the same percentages.   Discussion and Conclusion: It can be easily concluded that such dose variations are more noticeable when using lower energy sources such as Pd-103 or I-125.

  7. 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.

  8. 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

  9. MCNP6 Study of Fragmentation Products from 112Sn + 112Sn and 124Sn + 124Sn at 1 GeV/nucleon

    CERN Document Server

    Mashnik, Stepan G

    2013-01-01

    Isotope production cross sections from 112Sn + 112Sn and 124Sn + 124Sn reactions at 1 GeV/nucleon, which were measured recently at GSI using the heavy-ion accelerator SIS18 and the Fragment Separator (FRS), have been analyzed with the latest Los Alamos Monte-Carlo transport code MCNP6 using the LAQGSM03.03 event generator. MCNP6 reproduces reasonably well all the measured cross sections. Comparison of the MCNP6 results with the measured data and with calculations by a modification of the Los Alamos version of the Quark-Gluon String Model allowing for multifragmentation processes in the framework of the Statistical Multifragmentation Model (SMM) by Botvina and coauthors, as realized in the code LAQGSM03.S1, does not suggest unambiguous evidence of a multifragmentation signature.

  10. 基于MCNP和ORIGEN2耦合程序的IHNI-1型堆裂变产物中毒及燃耗分析%The fission product poisoning and burnup calculation for IHNI-1 reactor based on coupled code of MCNP-ORIGEN2

    Institute of Scientific and Technical Information of China (English)

    张信一; 赵柱民; 江新标; 郭和伟; 陈立新; 周永茂

    2012-01-01

    To calculate the fission product poisoning and bumup of the reactor accurately, the paper sets up the coupled calculation methods based on MCNP code and ORIGEN2 code and program data translation, cross section revision and date interface codes. Making use of elaborate reactor model to calculate the fission product poisoning and bumup for in-hospital neutron irradiator mark 1 reactor.%为了准确地计算反应堆的裂变产物中毒和燃耗问题,开发了一套蒙特卡罗方法程序系统.利用通用的燃耗计算方法,基于MCNP和ORIGEN2,编写了相关的数据转换、截面修正、数据接口程序,实现了MCNP和ORIGEN2程序的耦合.采用堆芯精细结构划分,对医院中子照射器Ⅰ型堆裂变产物中毒和燃耗进行了计算分析.

  11. 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.

  12. New gamma and neutron measurements and MCNP simulations

    Energy Technology Data Exchange (ETDEWEB)

    Crovisier, Ph.; Camus, L.; Marty, P. [CEA Centre de Valduc, Is sur Tille (France). Service de Protection contre les Rayonnements; Groetz, J.E [Univ. de Franche Comte, Besancon (France). Laboratoire de Microanalyses Nucleaires

    2003-05-01

    To take into account the criticality risk, the Radiation Protection Service of the CEA Valduc center has developed a new method allowing quickly fixed fissile material mass determination in complex configurations where the other classical techniques, such as gamma spectrometry, cannot be easily used (contaminated areas, large thickness shield protection). Then, the Radiation Protection Service in collaboration with the Nuclear Microanalyses Laboratory carried out ambient dose equivalent rate measurements coupled with a MCNP simulations in order to estimate 'holdup' nuclear materials. The methodology used is described below: Choice of measurement devices (gamma or neutron) according to the detection limits. Use of calibrated dose rate meters and new neutron spectrometer ROSPEC (measurement references and uncertainties). Ambient dose equivalent rate measurements should be performed at different locations in the vicinity of the system studied. Complete geometry system, shields and sources locations (if it's possible) should be modeled accurately in MCNP simulations. Ambient dose equivalent rate calculations at each measurement locations and for each source described are performed by using the MCNP code. All these measurements and calculations allow to set up a linear equations system with activities sources (mass) as unknowns. Due to the measurement uncertainties, this system cannot be exactly solved but by an iterative approach. The fissile material characteristics (i.e isotopic abundance, chemical form, nuclides) located in the system are very important to enable us the nuclear material mass estimations. Previously, these features can be determined by smears radiological analyses or by knowing the elaborated nuclear materials in the concerned plant. For the first time, this new method was successfully used to study a vessel containing metal plutonium located on the walls. The second estimation concerned the 'holdup' fissile material in a

  13. Comparison of CAP88 and MCNP for Overhead Gamma-emitting Plumes

    Energy Technology Data Exchange (ETDEWEB)

    Mcnaughton, Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gillis, Jessica Mcdonnel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McClory, Aysha Reede [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Whicker, Jeffrey Jay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fuehne, David Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-01-08

    The purpose of this paper is to use the Monte Carlo N-Particle Code (MCNP) to investigate the dose from gamma-emitting radionuclides such as Carbon-11 when a plume passes overhead. MCNP results are compared with results from the EPA program, CAP88. In some cases, typically near the source during stable conditions, the CAP88 results are less than the MCNP results. However, in the case of a receptor 800 m from a source at the Los Alamos Neutron Science Center (LANSCE), the CAP88 result is greater than the MCNP result.

  14. Geometry creation for MCNP by Sabrina and XSM

    Energy Technology Data Exchange (ETDEWEB)

    Van Riper, K.A.

    1994-02-01

    The Monte Carlo N-Particle transport code MCNP is based on a surface description of 3-dimensional geometry. Cells are defined in terms of boolean operations on signed quadratic surfaces. MCNP geometry is entered as a card image file containing coefficients of the surface equations and a list of surfaces and operators describing cells. Several programs are available to assist in creation of the geometry specification, among them Sabrina and the new ``Smart Editor`` code XSM. We briefly describe geometry creation in Sabrina and then discuss XSM in detail. XSM is under development; our discussion is based on the state of XSM as of January 1, 1994.

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-08-23

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-08-23

    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.

  18. 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.

  19. CGMF & FREYA Verification in MCNP6

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-05

    At the present time, the new and updated fission event generators included in MCNP6.2 have been verified to be functioning properly through a variety of detailed tests. This work describes the detailed verification steps taken to ensure these complicated fission event generators, FREYA and CGMF, are integrated into MCNP6 properly. Ultimately, with the knowledge that MCNP6 is making use of these models appropriately, we can now begin to validate the models against benchmarked experiments. Some benchmarks, including criticality and subcritical experiments interested in multiplication and bulk counting rates, are easy to model and understand but are likely insensitive to the detailed nature of these models. It will take some new measurements with coincidence detection capabilities to be able to stress the physics within each of these fission event generator models. Once the models are validated and it is understood where the models can truly be predictive, then we can study what SNM observables can be characterized for nonproliferation applications.

  20. 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

  1. 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.

  2. Impacts of Model Building Energy Codes

    Energy Technology Data Exchange (ETDEWEB)

    Athalye, Rahul A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sivaraman, Deepak [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elliott, Douglas B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bartlett, Rosemarie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-10-31

    The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) periodically evaluates national and state-level impacts associated with energy codes in residential and commercial buildings. Pacific Northwest National Laboratory (PNNL), funded by DOE, conducted an assessment of the prospective impacts of national model building energy codes from 2010 through 2040. A previous PNNL study evaluated the impact of the Building Energy Codes Program; this study looked more broadly at overall code impacts. This report describes the methodology used for the assessment and presents the impacts in terms of energy savings, consumer cost savings, and reduced CO2 emissions at the state level and at aggregated levels. This analysis does not represent all potential savings from energy codes in the U.S. because it excludes several states which have codes which are fundamentally different from the national model energy codes or which do not have state-wide codes. Energy codes follow a three-phase cycle that starts with the development of a new model code, proceeds with the adoption of the new code by states and local jurisdictions, and finishes when buildings comply with the code. The development of new model code editions creates the potential for increased energy savings. After a new model code is adopted, potential savings are realized in the field when new buildings (or additions and alterations) are constructed to comply with the new code. Delayed adoption of a model code and incomplete compliance with the code’s requirements erode potential savings. The contributions of all three phases are crucial to the overall impact of codes, and are considered in this assessment.

  3. 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.

  4. An Electron/Photon/Relaxation Data Library for MCNP6

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-07

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

  5. Total reaction cross sections in CEM and MCNP6 at intermediate energies

    Science.gov (United States)

    Kerby, Leslie M.; Mashnik, Stepan G.

    2015-08-01

    Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (∼ 50 MeV to ∼ 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used in the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are available now. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results. Our current results indicate this is, in fact, the case.

  6. Total Reaction Cross Sections in CEM and MCNP6 at Intermediate Energies

    CERN Document Server

    Kerby, Leslie M

    2015-01-01

    Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region ($\\sim$50 MeV to $\\sim$5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used in the preequilibrium and evaporation stages of CEM are based on the Dostrovsky {\\it et al.} model, published in 1959. Better cross section models are available now. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results. Our current...

  7. 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.

  8. MCNP(TM) Version 5.

    Energy Technology Data Exchange (ETDEWEB)

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

  9. Evaluation of help model replacement codes

    Energy Technology Data Exchange (ETDEWEB)

    Whiteside, Tad [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hang, Thong [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Flach, Gregory [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2009-07-01

    This work evaluates the computer codes that are proposed to be used to predict percolation of water through the closure-cap and into the waste containment zone at the Department of Energy closure sites. This work compares the currently used water-balance code (HELP) with newly developed computer codes that use unsaturated flow (Richards’ equation). It provides a literature review of the HELP model and the proposed codes, which result in two recommended codes for further evaluation: HYDRUS-2D3D and VADOSE/W. This further evaluation involved performing actual simulations on a simple model and comparing the results of those simulations to those obtained with the HELP code and the field data. From the results of this work, we conclude that the new codes perform nearly the same, although moving forward, we recommend HYDRUS-2D3D.

  10. MCNP Super Lattice Method for VHTR ORIGEN2.2 Nuclear Library Improvement Based on ENDF/B-VII

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; J. R. Parry

    2010-10-01

    The advanced Very High Temperature gas-cooled Reactor (VHTR) achieves simplification of safety through reliance on innovative features and passive systems. One of the VHTRs innovative features is the reliance on ceramic-coated fuel particles to retain the fission products under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel prismatic block creates a double-heterogeneous lattice, which needs to be addressed through the use of the newly developed prismatic super Kernel-by-Kernel Fuel (KbKF) lattice model method. Based on the new ENDF/B-VII nuclear cross section evaluated data, the developed KbKF super lattice model was then used with MCNP to calculate the material isotopes neutron reaction rates, such as, (n,?); (n,n’); (n,2n’); (n,f); (n,p); (n,?). Then, the MCNP-calculated results are rearranged to generate a set of new libraries “VHTRXS.lib,” for the ORIGEN2.2 isotopes depletion and build-up analysis code. The libraries contain one group cross section data for the structural light elements, actinides, and fission products that can be applied in the VHTR related fuel burnup and material transmutation analysis codes. The efficiency and ease of use of the MCNP method to generate and update the ORIGEN2.2 one-group spectrum weighed cross section library for VHTR was demonstrated.

  11. Investigating spatial self-shielding and temperature effects for homogeneous and double heterogeneous pebble models with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.; Nuenighoff; Pohl, C.; Allelein, H.J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung, Sicherheitsforschung und Reaktortechnik (IEF-6)

    2010-05-15

    The gas-cooled, high temperature reactor (HTR) represents a valuable option for the future development of nuclear technology, because of its excellent safety features. One main safety feature is the negative temperature coefficient which is due to the Doppler broadening of the (n,y) resonance absorption cross section. A second important effect is the spatial self-shielding due to the double heterogeneous geometry of a pebble bed reactor. At FZ-Juelich two reactor analysis codes have been developed: VSOP for core design and MGT for transient analysis. Currently an update of the nuclear cross section libraries to ENDF/B-VII.0 of both codes takes place. In order to take the temperature dependency as well as the spatial self-shielding into account the absorption cross sections {sigma}{sub (n,y)} for the resonance absorbers like {sup 232}Th and {sup 238}U have to be provided as function of incident neutron energy, temperature and nuclide concentration. There are two reasons for choosing the Monte-Carlo approach to calculate group wise cross sections. First, the former applied ZUT-DGL code to generate the resonance cross section tables for MGT is so far not able to handle the new resonance description based on Reich-Moore instead of Single-level Breit-Wigner. Second, the rising interest in PuO{sub 2} fuel motivated an investigation on the generation of group wise cross sections describing thermal resonances of {sup 240}Pu and {sup 242}Pu. (orig.)

  12. Coding with partially hidden Markov models

    DEFF Research Database (Denmark)

    Forchhammer, Søren; Rissanen, J.

    1995-01-01

    Partially hidden Markov models (PHMM) are introduced. They are a variation of the hidden Markov models (HMM) combining the power of explicit conditioning on past observations and the power of using hidden states. (P)HMM may be combined with arithmetic coding for lossless data compression. A general...... 2-part coding scheme for given model order but unknown parameters based on PHMM is presented. A forward-backward reestimation of parameters with a redefined backward variable is given for these models and used for estimating the unknown parameters. Proof of convergence of this reestimation is given....... The PHMM structure and the conditions of the convergence proof allows for application of the PHMM to image coding. Relations between the PHMM and hidden Markov models (HMM) are treated. Results of coding bi-level images with the PHMM coding scheme is given. The results indicate that the PHMM can adapt...

  13. Current status of MCNP6 as a simulation tool useful for space and accelerator applications

    CERN Document Server

    Mashnik, S G; Hughes, H G; Prael, R E; Sierk, A J

    2012-01-01

    For the past several years, a major effort has been undertaken at Los Alamos National Laboratory (LANL) to develop the transport code MCNP6, the latest LANL Monte-Carlo transport code representing a merger and improvement of MCNP5 and MCNPX. We emphasize a description of the latest developments of MCNP6 at higher energies to improve its reliability in calculating rare-isotope production, high-energy cumulative particle production, and a gamut of reactions important for space-radiation shielding, cosmic-ray propagation, and accelerator applications. We present several examples of validation and verification of MCNP6 compared to a wide variety of intermediate- and high-energy experimental data on reactions induced by photons, mesons, nucleons, and nuclei at energies from tens of MeV to about 1 TeV/nucleon, and compare to results from other modern simulation tools.

  14. Transmutation Fuel Performance Code Thermal Model Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller; Pavel G. Medvedev

    2007-09-01

    FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

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

    Science.gov (United States)

    Gerardy, I; Ródenas, J; van Dycke, M; Gallardo, S; Ceccolini, Elisa

    2010-01-01

    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.

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

    OpenAIRE

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik; Willendrup, Peter Kjær

    2012-01-01

    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 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 dire...

  17. Generation of Java code from Alvis model

    Science.gov (United States)

    Matyasik, Piotr; Szpyrka, Marcin; Wypych, Michał

    2015-12-01

    Alvis is a formal language that combines graphical modelling of interconnections between system entities (called agents) and a high level programming language to describe behaviour of any individual agent. An Alvis model can be verified formally with model checking techniques applied to the model LTS graph that represents the model state space. This paper presents transformation of an Alvis model into executable Java code. Thus, the approach provides a method of automatic generation of a Java application from formally verified Alvis model.

  18. MCNP{trademark} simulations for identifying environmental contaminants using prompt gamma-rays from thermal neutron capture reactions

    Energy Technology Data Exchange (ETDEWEB)

    Frankle, S.C.; Conaway, J.G.

    1996-12-31

    The primary purposes of the Multispectral Neutron Logging Project, (MSN Project, funded by the U.S. Department of Energy), were to assess the effectiveness of existing neutron- induced spectral gamma-ray logging techniques for identifying environmental contaminants along boreholes, to further improve the technology, and to transfer that technology to industry. Using a pulsed neutron source with a high-resolution gamma-ray detector, spectra from thermal neutron capture reactions may be used to identify contaminants in the borehole environment. Direct borehole measurements such as this complement physical sampling and are useful in environmental restoration projects where characterization of contaminated sites is required and long-term monitoring may be needed for many years following cleanup or stabilization. In the MSN Project, a prototype logging instrument was designed which incorporated a pulsed 14-MeV neutron source and HPGe detector. Experimental measurements to determine minimum detection thresholds with the prototype instrument were conducted in the variable-contaminant test model for Cl, Cd, Sm, Gd, and Hg. We benchmarked an enhanced version of the Monte Carlo N-Particle computer code MCNP{trademark} using experimental data for Cl provide by our collaborators and experimental data from the variable-contaminant test model. MCNP was then used to estimate detection thresholds for the other contaminants used in the variable-contaminant model with the goal of validating the use of MCNP to estimate detection thresholds for many other contaminants that were not measured.

  19. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-05-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)

  20. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VT (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2}. For the cases studied, it is found that the absolute keff values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in keff), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}keff on coolant voiding), and is relatively insensitive to the fuel type. (author)

  1. Economic aspects and models for building codes

    DEFF Research Database (Denmark)

    Bonke, Jens; Pedersen, Dan Ove; Johnsen, Kjeld

    It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study.......It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study....

  2. Mathematical models for the EPIC code

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, H.L.

    1981-06-03

    EPIC is a fluid/envelope type computer code designed to study the energetics and dynamics of a high energy, high current electron beam passing through a gas. The code is essentially two dimensional (x, r, t) and assumes an axisymmetric beam whose r.m.s. radius is governed by an envelope model. Electromagnetic fields, background gas chemistry, and gas hydrodynamics (density channel evolution) are all calculated self-consistently as functions of r, x, and t. The code is a collection of five major subroutines, each of which is described in some detail in this report.

  3. 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

    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...... 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...... 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....

  4. Designing an epithermal neutron beam for boron neutron capture therapy for a DIDO type reactor using MCNP

    Science.gov (United States)

    Ross, D.; Constantine, G.; Weaver, D. R.; Beynon, T. D.

    1993-10-01

    This paper describes work undertaken to design an epithermal neutron beam for a DIDO type reactor for use in boron neutron capture therapy, a form of cancer treatment. It involved extensive use of MCNP, a Monte Carlo computer code. Initially, calculations were made with MCNP to simulate earlier experiments with an epithermal beam on the DIDO reactor. This comparison made it possible both to validate the Monte Carlo modelling of the reactor and to gain an insight into the important features of the simulation. Following this, MCNP was used to design a filtered epithermal neutron beam facility for DIDO's largest beam tube, a 13.7 cm radius horizontal tube which extends radially away from the core. First a selection was made of the optimum filter components for the beam. Then the research concentrated on combining these filter elements to construct a practical epithermal beam design. The results suggest that the optimum method of generating the epithermal neutron source is to employ a filter combination consisting principally of liquid argon with the addition of cadmium, aluminium, titanium and possibly tin. The calculations also show that the resultant neutron beam would have a flux greater than 1.0 × 10 9 n cm -2 s -1 and have sufficiently low fast-neutron and gamma-ray contamination.

  5. Certification of MCNP Version 4A for WHC computer platforms. Revision 7

    Energy Technology Data Exchange (ETDEWEB)

    Carter, L.L.

    1995-05-03

    MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).

  6. Model Description of TASS/SMR Code

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Y. D.; Yang, S. H.; Kim, S. H.; Lee, S. W.; Kim, H. K.; Yoon, H. Y.; Lee, G. H.; Bae, K. H.; Chung, Y. J

    2005-12-15

    TASS/SMR(Transient And Setpoint Simulation/System-integrated Modular Reactor) code has been developed for the safety analysis of the SMART-P reactor. TASS/SMR code can be applied for the analysis of design base accidents including small break loss of coolant accident of the SMART research reactor. TASS/SMR code models the primary and secondary system using a node and flow path. A node represents the control volume which defines the fluid mass and energy. Flow path connects the nodes to define the momentum of the fluid. The mass and energy conservation equations are applied to the node and the momentum conservation equation applied to the flow path. In TASS/SMR, the governing equations are applied for both the primary and the secondary coolant system and are solved simultaneously. The governing equations of TASS/SMR are based on the drift-flux model so that the accidents or transients accompanying with two-phase flow can be analyzed. Also, the SMART-P reactor specific thermal-hydraulic models are incorporated, such as non-condensable gas model, helical steam generator heat transfer model, and passive residual heat removal system (PRHRS) heat transfer model. This technical report describes the governing equations, solution method, thermal hydraulics, reactor core, control system models used in TASS/SMR code. Also, the description for the steady state simulation, the minimum CHFR and hottest fuel temperature calculation methods are described in this report.

  7. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

  8. Mapping Initial Hydrostatic Models in Godunov Codes

    CERN Document Server

    Zingale, M A; Zu Hone, J; Calder, A C; Fryxell, B; Plewa, T; Truran, J W; Caceres, A; Olson, K; Ricker, P M; Riley, K; Rosner, R; Siegel, A; Timmes, F X; Vladimirova, N

    2002-01-01

    We look in detail at the process of mapping an astrophysical initial model from a stellar evolution code onto the computational grid of an explicit, Godunov type code while maintaining hydrostatic equilibrium. This mapping process is common in astrophysical simulations, when it is necessary to follow short-timescale dynamics after a period of long timescale buildup. We look at the effects of spatial resolution, boundary conditions, the treatment of the gravitational source terms in the hydrodynamics solver, and the initialization process itself. We conclude with a summary detailing the mapping process that yields the lowest ambient velocities in the mapped model.

  9. Source coding model for repeated snapshot imaging

    CERN Document Server

    Li, Junhui; Yang, Dongyue; wu, Guohua; Yin, Longfei; Guo, Hong

    2016-01-01

    Imaging based on successive repeated snapshot measurement is modeled as a source coding process in information theory. The necessary number of measurement to maintain a certain level of error rate is depicted as the rate-distortion function of the source coding. Quantitative formula of the error rate versus measurement number relation is derived, based on the information capacity of imaging system. Second order fluctuation correlation imaging (SFCI) experiment with pseudo-thermal light verifies this formula, which paves the way for introducing information theory into the study of ghost imaging (GI), both conventional and computational.

  10. MCNP-DSP calculations of measurements with uranyl nitrate solution system

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E. [Oak Ridge National Lab., TN (United States)

    1998-09-01

    The {sup 252}Cf-source-driven noise analysis method has been used to determine the subcriticality of various configurations of fissile materials. In the past, the application of this method was limited because point-kinetics models had to be used to interpret the data; however, with the development of the Monte Carlo code MCNP-DSP, the measurements can be analyzed using the more general Monte Carlo models. The results of the Monte carlo calculations will be dependent on the ability to model the experiment accurately and on the nuclear data used to perform the calculations. This paper presents a comparison of the measured and calculated ratio of spectral densities for a subset of measurements performed with a uranyl nitrate solution tank filled to various heights. The results presented are for calculations that were performed with both ENDF/B-IV and ENDF/B-V cross-section data sets.

  11. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  12. 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.

  13. General purpose photoneutron production in MCNP4A

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Modeling peripheral olfactory coding in Drosophila larvae.

    Directory of Open Access Journals (Sweden)

    Derek J Hoare

    Full Text Available The Drosophila larva possesses just 21 unique and identifiable pairs of olfactory sensory neurons (OSNs, enabling investigation of the contribution of individual OSN classes to the peripheral olfactory code. We combined electrophysiological and computational modeling to explore the nature of the peripheral olfactory code in situ. We recorded firing responses of 19/21 OSNs to a panel of 19 odors. This was achieved by creating larvae expressing just one functioning class of odorant receptor, and hence OSN. Odor response profiles of each OSN class were highly specific and unique. However many OSN-odor pairs yielded variable responses, some of which were statistically indistinguishable from background activity. We used these electrophysiological data, incorporating both responses and spontaneous firing activity, to develop a bayesian decoding model of olfactory processing. The model was able to accurately predict odor identity from raw OSN responses; prediction accuracy ranged from 12%-77% (mean for all odors 45.2% but was always significantly above chance (5.6%. However, there was no correlation between prediction accuracy for a given odor and the strength of responses of wild-type larvae to the same odor in a behavioral assay. We also used the model to predict the ability of the code to discriminate between pairs of odors. Some of these predictions were supported in a behavioral discrimination (masking assay but others were not. We conclude that our model of the peripheral code represents basic features of odor detection and discrimination, yielding insights into the information available to higher processing structures in the brain.

  15. Monte Carlo importance sampling for the MCNP{trademark} general source

    Energy Technology Data Exchange (ETDEWEB)

    Lichtenstein, H.

    1996-01-09

    Research was performed to develop an importance sampling procedure for a radiation source. The procedure was developed for the MCNP radiation transport code, but the approach itself is general and can be adapted to other Monte Carlo codes. The procedure, as adapted to MCNP, relies entirely on existing MCNP capabilities. It has been tested for very complex descriptions of a general source, in the context of the design of spent-reactor-fuel storage casks. Dramatic improvements in calculation efficiency have been observed in some test cases. In addition, the procedure has been found to provide an acceleration to acceptable convergence, as well as the benefit of quickly identifying user specified variance-reduction in the transport that effects unstable convergence.

  16. Energetic Light Fragment Production Capability in MCNP6

    CERN Document Server

    Kerby, Leslie M; Gudima, Konstantin K; Sierk, Arnold J; Bull, Jeffrey S; James, Michael R

    2016-01-01

    The goal of this research is to enable MCNP6 to produce high-energy light fragments. These energetic light fragments may be emitted by our models through three processes: Fermi breakup, preequilibrium, and coalescence. We explore the emission of light fragments through each of these mechanisms and demonstrate an improved agreement with experimental data achieved by extending precompound models to include emission of fragments heavier than $^4$He.

  17. 24 CFR 200.926c - Model code provisions for use in partially accepted code jurisdictions.

    Science.gov (United States)

    2010-04-01

    ... 24 Housing and Urban Development 2 2010-04-01 2010-04-01 false Model code provisions for use in partially accepted code jurisdictions. 200.926c Section 200.926c Housing and Urban Development Regulations... Minimum Property Standards § 200.926c Model code provisions for use in partially accepted...

  18. MEMOPS: data modelling and automatic code generation.

    Science.gov (United States)

    Fogh, Rasmus H; Boucher, Wayne; Ionides, John M C; Vranken, Wim F; Stevens, Tim J; Laue, Ernest D

    2010-03-25

    In recent years the amount of biological data has exploded to the point where much useful information can only be extracted by complex computational analyses. Such analyses are greatly facilitated by metadata standards, both in terms of the ability to compare data originating from different sources, and in terms of exchanging data in standard forms, e.g. when running processes on a distributed computing infrastructure. However, standards thrive on stability whereas science tends to constantly move, with new methods being developed and old ones modified. Therefore maintaining both metadata standards, and all the code that is required to make them useful, is a non-trivial problem. Memops is a framework that uses an abstract definition of the metadata (described in UML) to generate internal data structures and subroutine libraries for data access (application programming interfaces--APIs--currently in Python, C and Java) and data storage (in XML files or databases). For the individual project these libraries obviate the need for writing code for input parsing, validity checking or output. Memops also ensures that the code is always internally consistent, massively reducing the need for code reorganisation. Across a scientific domain a Memops-supported data model makes it easier to support complex standards that can capture all the data produced in a scientific area, share them among all programs in a complex software pipeline, and carry them forward to deposition in an archive. The principles behind the Memops generation code will be presented, along with example applications in Nuclear Magnetic Resonance (NMR) spectroscopy and structural biology.

  19. Uncertainty analysis in MCNP5 calculations for brachytherapy treatment

    Energy Technology Data Exchange (ETDEWEB)

    Gerardy, I., E-mail: gerardy@isib.be [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 (Spain)

    2011-08-15

    The Monte Carlo (MC) method can be applied to simulate brachytherapy treatment planning. The MCNP5 code gives, together with results, a statistical uncertainty associated with them. However, the latter is not the only existing uncertainty related to the simulation and other uncertainties must be taken into account. A complete analysis of all sources of uncertainty having some influence on results of the simulation of brachytherapy treatment is presented in this paper. This analysis has been based on the recommendations of the American Association for Physicist in Medicine (AAPM) and of the International Standard Organisation (ISO).

  20. Creating Models for the ORIGEN Codes

    Science.gov (United States)

    Louden, G. D.; Mathews, K. A.

    1997-10-01

    Our research focused on the development of a methodology for creating reactor-specific cross-section libraries for nuclear reactor and nuclear fuel cycle analysis codes available from the Radiation Safety Information Computational Center. The creation of problem-specific models allows more detailed anlaysis than is possible using the generic models provided with ORIGEN2 and ORIGEN-S. A model of the Ohio State University Research Reactor was created using the Coupled 1-D Shielding Analysis (SAS2H) module of the Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation (SCALE4.3). Six different reactor core models were compared to identify the effect of changing the SAS2H Larger Unit Cell on the predicted isotopic composition of spent fuel. Seven different power histories were then applied to a Core-Average model to determine the ability of ORIGEN-S to distinguish spent fuel produced under varying operating conditions. Several actinide and fission product concentrations were identified which were sensitive to the power history, however the majority of the isotope concentrations were not dependent on operating history.

  1. Simulations of neutron multiplicity measurements of a weapons-grade plutonium sphere with MCNP-PoliMi.

    Energy Technology Data Exchange (ETDEWEB)

    Mattingly, John K.; Pozzi, Sara A. (University of Michigan, Ann Arbor, MI); Clarke, Shaun D. (University of Michigan, Ann Arbor, MI); Dennis, Ben D. (University of Michigan, Ann Arbor, MI); Miller, Eric C. (University of Michigan, Ann Arbor, MI); Padovani, E. (Polytechnic of Milan, Italy)

    2010-06-01

    With increasing concern over the ability to detect and characterize special nuclear materials, the need for computer codes that can successfully predict the response of detector systems to various measurement scenarios is extremely important. These computer algorithms need to be benchmarked against a variety of experimental configurations to ensure their accuracy and understand their limitations. The Monte Carlo code MCNP-PoliMi is a modified version of the MCNP-4c code. Recently these modifications have been ported into the new MCNPX 2.6.0 code, which gives the new MCNPX-PoliMi a wider variety of options and abilities, taking advantage of the improvements made to MCNPX. To verify the ability of the MCNPX-PoliMi code to simulate the response of a neutron multiplicity detector simulated results were compared to experimental data. The experiment consisted of a 4.5-kg sphere of alpha-phase plutonium that was moderated with various thicknesses of polyethylene. The results showed that our code system can simulate the multiplicity distributions with relatively good agreement with measured data. The enhancements made to MCNP since the release of MCNP-4c have had little to no effect on the ability of the MCNP-PoliMi to resolve the discrepancies observed in the simulated neutron multiplicity distributions when compared experimental data.

  2. Mcnp calculation of neutron scatter in the Main Bay of the Chadwick Building, NPL

    Energy Technology Data Exchange (ETDEWEB)

    Naismith, O.F.; Thomas, D.J.

    1996-02-01

    The Monte Carlo neutron transport code MCNP has been used to calculate the room and air scattered neutron component at 75 cm from a radionuclide source located at the center of the low-scatter area in the Chadwick Building, Bldg. 47, at National Physical Laboratory (NPL). This is the standard distance used for calibrating personal dosemeters, and the calculation provides information for correcting the response of dosemeters to the scattered radiation. Calculations were performed for both an Am-Be and a (252)Cf source. These measurements revealed that the model used for features within the low-scatter area needs to be refined for calculating scatter at distances further from the source than 75 cm.

  3. THE MCNPX MONTE CARLO RADIATION TRANSPORT CODE

    Energy Technology Data Exchange (ETDEWEB)

    WATERS, LAURIE S. [Los Alamos National Laboratory; MCKINNEY, GREGG W. [Los Alamos National Laboratory; DURKEE, JOE W. [Los Alamos National Laboratory; FENSIN, MICHAEL L. [Los Alamos National Laboratory; JAMES, MICHAEL R. [Los Alamos National Laboratory; JOHNS, RUSSELL C. [Los Alamos National Laboratory; PELOWITZ, DENISE B. [Los Alamos National Laboratory

    2007-01-10

    MCNPX (Monte Carlo N-Particle eXtended) is a general-purpose Monte Carlo radiation transport code with three-dimensional geometry and continuous-energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi-processing computer platforms. MCNPX is based on MCNP4B, and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low-energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics; particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development.

  4. 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.

  5. Application of MCNP for neutronic calculations at VR-1 training reactor

    Science.gov (United States)

    Huml, Ondřej; Rataj, Jan; Bílý, Tomáš

    2014-06-01

    The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.

  6. A graph model for opportunistic network coding

    KAUST Repository

    Sorour, Sameh

    2015-08-12

    © 2015 IEEE. Recent advancements in graph-based analysis and solutions of instantly decodable network coding (IDNC) trigger the interest to extend them to more complicated opportunistic network coding (ONC) scenarios, with limited increase in complexity. In this paper, we design a simple IDNC-like graph model for a specific subclass of ONC, by introducing a more generalized definition of its vertices and the notion of vertex aggregation in order to represent the storage of non-instantly-decodable packets in ONC. Based on this representation, we determine the set of pairwise vertex adjacency conditions that can populate this graph with edges so as to guarantee decodability or aggregation for the vertices of each clique in this graph. We then develop the algorithmic procedures that can be applied on the designed graph model to optimize any performance metric for this ONC subclass. A case study on reducing the completion time shows that the proposed framework improves on the performance of IDNC and gets very close to the optimal performance.

  7. MCNP6 Simulation of Reactions of Interest to FRIB, Medical, and Space Applications

    CERN Document Server

    Mashnik, Stepan G

    2014-01-01

    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 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 compared to recent neutron spectra measured at the Heavy Ion Medical Accelerator in Chiba, Japan; to spectra of light fragments from several reactions measured recently at GANIL, France; INFN Laboratori Nazionali del Sud, Catania, Italy; COSY of the Julich Research Center, Germany; and to cross sections of products from several reactions measured lately at GSI, Darmstadt, Germany; ITEP, Moscow, Russia; LANSCE, LANL, Los Alamos, USA. As a rule, MCNP6 provi...

  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. Production of Energetic Light Fragments in CEM, LAQGSM, and MCNP6

    CERN Document Server

    Mashnik, Stepan G; Gudima, Konstantin K; Sierk, Arnold J; Bull, Jeffrey S; James, Michael R

    2016-01-01

    We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte-Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi break-up, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi break-up model and choose the best option for these models. Then, we extend the modified exciton model (MEM) used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A < ...

  10. ER@CEBAF: Modeling code developments

    Energy Technology Data Exchange (ETDEWEB)

    Meot, F. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Roblin, Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)

    2016-04-13

    A proposal for a multiple-pass, high energy, energy-recovery experiment using CEBAF is under preparation in the frame of a JLab-BNL collaboration. In view of beam dynamics investigations regarding this project, in addition to the existing model in use in Elegant a version of CEBAF is developed in the stepwise ray-tracing code Zgoubi, Beyond the ER experiment, it is also planned to use the latter for the study of polarization transport in the presence of synchrotron radiation, down to Hall D line where a 12 GeV polarized beam can be delivered. This Note briefly reports on the preliminary steps, and preliminary outcomes, based on an Elegant to Zgoubi translation.

  11. Characteristic Analysis of Fire Modeling Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hwan; Yang, Joon Eon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Jong Hoon [Kyeongmin College, Ujeongbu (Korea, Republic of)

    2004-04-15

    This report documents and compares key features of four zone models: CFAST, COMPBRN IIIE, MAGIC and the Fire Induced Vulnerability Evaluation (FIVE) methodology. CFAST and MAGIC handle multi-compartment, multi-fire problems, using many equations; COMPBRN and FIVE handle single compartment, single fire source problems, using simpler equation. The increased rigor of the formulation of CFAST and MAGIC does not mean that these codes are more accurate in every domain; for instance, the FIVE methodology uses a single zone approximation with a plume/ceiling jet sublayer, while the other models use a two-zone treatment without a plume/ceiling jet sublayer. Comparisons with enclosure fire data indicate that inclusion of plume/ceiling jet sublayer temperatures is more conservative, and generally more accurate than neglecting them. Adding a plume/ceiling jet sublayer to the two-zone models should be relatively straightforward, but it has not been done yet for any of the two-zone models. Such an improvement is in progress for MAGIC.

  12. Extension and Test of Limits on MCNP Geometry Description%MCNP程序几何描述能力扩展及应用测试

    Institute of Scientific and Technical Information of China (English)

    刘镇洲; 李刚; 邓力; 柴晓明

    2013-01-01

    为使MCNP程序能模拟数百万规模的反应堆“pin-by-pin”问题和医学体素模型,本文对MCNP程序进行了改进,使几何块、几何面数量可扩展.改进后的程序对硼中子俘获治疗(BNCT)的人体大脑进行几何建模,栅元数量达百万量级;计算了大脑的中子、光子吸收剂量率随深度的变化,为大脑BNCT提供理论支持.此外,对百万规模的“Like n But”重复结构模型进行了串、并行测试,验证了几何规模扩展后程序计算的正确性.%To set up models for 'pin-by-pin' facilities (reactors) and human phantoms with millions of cells by using the MCNP code, the MCNP code was modified and limits on the number of cell, surface, material, etc. were extended. The Snyder head phantom for boron neutron capture therapy was modeled. Neutron and induced gamma-ray absorbed dose in human brain was calculated. Besides, a critical calculation model of repeated structures described by the 'Like n But' cards was calculated employing both series and parallel computing. As a conclusion, modifications of the MCNP code were proved to be correct.

  13. 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)

  14. Towards Preserving Model Coverage and Structural Code Coverage

    Directory of Open Access Journals (Sweden)

    Raimund Kirner

    2009-01-01

    Full Text Available Embedded systems are often used in safety-critical environments. Thus, thorough testing of them is mandatory. To achieve a required structural code-coverage criteria it is beneficial to derive the test data at a higher program-representation level than machine code. Higher program-representation levels include, beside the source-code level, languages of domain-specific modeling environments with automatic code generation. For a testing framework with automatic generation of test data this will enable high retargetability of the framework. In this article we address the challenge of ensuring that the structural code coverage achieved at a higher program representation level is preserved during the code generations and code transformations down to machine code. We define the formal properties that have to be fullfilled by a code transformation to guarantee preservation of structural code coverage. Based on these properties we discuss how to preserve code coverage achieved at source-code level. Additionally, we discuss how structural code coverage at model level could be preserved. The results presented in this article are aimed toward the integration of support for preserving structural code coverage into compilers and code generators.

  15. Simulink Code Generation: Tutorial for Generating C Code from Simulink Models using Simulink Coder

    Science.gov (United States)

    MolinaFraticelli, Jose Carlos

    2012-01-01

    This document explains all the necessary steps in order to generate optimized C code from Simulink Models. This document also covers some general information on good programming practices, selection of variable types, how to organize models and subsystems, and finally how to test the generated C code and compare it with data from MATLAB.

  16. PetriCode: A Tool for Template-Based Code Generation from CPN Models

    DEFF Research Database (Denmark)

    Simonsen, Kent Inge

    2014-01-01

    Code generation is an important part of model driven methodologies. In this paper, we present PetriCode, a software tool for generating protocol software from a subclass of Coloured Petri Nets (CPNs). The CPN subclass is comprised of hierarchical CPN models describing a protocol system at different...

  17. 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.

  18. MCNP - transport calculations in ducts using multigroup albedo coefficients; Calculos de transporte em dutos utilizando coeficientes de albedo multigrupo no codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Ono, Shizuca; Vieira, Wilson J.; Garcia, Roberto D.M. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados

    2000-07-01

    In this work, the use of multigroup albedo coefficients in Monte Carlo calculations of particle reflection and transmission by ducts is investigated. The procedure consists in modifying the MCNP code so that an albedo matrix computed previously by deterministic methods or Monte Carlo is introduced into the program to describe particle reflection by a surface. This way it becomes possible to avoid the need of considering particle transport in the duct wall explicitly, changing the problem to a problem of transport in the duct interior only and reducing significantly the difficulty of the real problem. The probability of particle reflection at the duct wall is given, for each group, as the sum of the albedo coefficients over the final groups. The calculation is started by sampling a source particle and simulating its reflection on the duct wall by sampling a group for the emerging particle. The particle weight is then reduced by the reflection probability. Next, a new direction and trajectory for the particle is selected. Numerical results obtained for the model are compared with results from a discrete ordinates code and results from Monte Carlo simulations that take particle transport in the wall into account. (author)

  19. 40 CFR 194.23 - Models and computer codes.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e.,...

  20. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. H.; Hur, C.; Kim, D. K.; Cho, H. J. [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

  1. Assessment of doses caused by electrons in thin layers of tissue-equivalent materials, using MCNP.

    Science.gov (United States)

    Heide, Bernd

    2013-10-01

    Absorbed doses caused by electron irradiation were calculated with Monte Carlo N-Particle transport code (MCNP) for thin layers of tissue-equivalent materials. The layers were so thin that the calculation of energy deposition was on the border of the scope of MCNP. Therefore, in this article application of three different methods of calculation of energy deposition is discussed. This was done by means of two scenarios: in the first one, electrons were emitted from the centre of a sphere of water and also recorded in that sphere; and in the second, an irradiation with the PTB Secondary Standard BSS2 was modelled, where electrons were emitted from an (90)Sr/(90)Y area source and recorded inside a cuboid phantom made of tissue-equivalent material. The speed and accuracy of the different methods were of interest. While a significant difference in accuracy was visible for one method in the first scenario, the difference in accuracy of the three methods was insignificant for the second one. Considerable differences in speed were found for both scenarios. In order to demonstrate the need for calculating the dose in thin small zones, a third scenario was constructed and simulated as well. The third scenario was nearly equal to the second one, but a pike of lead was assumed to be inside the phantom in addition. A dose enhancement (caused by the pike of lead) of ∼113 % was recorded for a thin hollow cylinder at a depth of 0.007 cm, which the basal-skin layer is referred to in particular. Dose enhancements between 68 and 88 % were found for a slab with a radius of 0.09 cm for all depths. All dose enhancements were hardly noticeable for a slab with a cross-sectional area of 1 cm(2), which is usually applied to operational radiation protection.

  2. Image Coding using Markov Models with Hidden States

    DEFF Research Database (Denmark)

    Forchhammer, Søren Otto

    1999-01-01

    The Cylinder Partially Hidden Markov Model (CPH-MM) is applied to lossless coding of bi-level images. The original CPH-MM is relaxed for the purpose of coding by not imposing stationarity, but otherwise the model description is the same.......The Cylinder Partially Hidden Markov Model (CPH-MM) is applied to lossless coding of bi-level images. The original CPH-MM is relaxed for the purpose of coding by not imposing stationarity, but otherwise the model description is the same....

  3. Interfacial and Wall Transport Models for SPACE-CAP Code

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

  4. 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.

  5. Code Generation for Protocols from CPN models Annotated with Pragmatics

    DEFF Research Database (Denmark)

    Simonsen, Kent Inge; Kristensen, Lars Michael; Kindler, Ekkart

    Model-driven engineering (MDE) provides a foundation for automatically generating software based on models. Models allow software designs to be specified focusing on the problem domain and abstracting from the details of underlying implementation platforms. When applied in the context of formal...... modelling languages, MDE further has the advantage that models are amenable to model checking which allows key behavioural properties of the software design to be verified. The combination of formally verified models and automated code generation contributes to a high degree of assurance that the resulting...... of the same model and sufficiently detailed to serve as a basis for automated code generation when annotated with code generation pragmatics. Pragmatics are syntactical annotations designed to make the CPN models descriptive and to address the problem that models with enough details for generating code from...

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

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki; Kawai, Wataru; Ishizuki, Shigeru [Fujitsu Ltd., Tokyo (Japan); Kawasaki, Nobuo; Kume, Etsuo; Adachi, Masaaki; Ogasawara, Shinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yatake, Yo-ichi [Hitachi Ltd., Tokyo (Japan)

    2000-03-01

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

  7. Conservation of concrete structures in fib model code 2010

    NARCIS (Netherlands)

    Matthews, S.L.; Ueda, T.; Bigaj-van Vliet, A.

    2012-01-01

    Chapter 9: Conservation of concrete structures forms part of fib Model Code 2010, the first draft of which was published for comment as fib Bulletins 55 and 56 (fib 2010). Numerous comments were received and considered by fib Special Activity Group 5 responsible for the preparation of fib Model Code

  8. Conservation of concrete structures in fib model code 2010

    NARCIS (Netherlands)

    Matthews, S.L.; Ueda, T.; Bigaj-van Vliet, A.

    2012-01-01

    Chapter 9: Conservation of concrete structures forms part of fib Model Code 2010, the first draft of which was published for comment as fib Bulletins 55 and 56 (fib 2010). Numerous comments were received and considered by fib Special Activity Group 5 responsible for the preparation of fib Model Code

  9. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Science.gov (United States)

    Herreras, Y.; Lafuente, A.; Sordo, F.; Cabellos, O.; Perlado, J. M.

    2008-05-01

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  10. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Energy Technology Data Exchange (ETDEWEB)

    Herreras, Y; Cabellos, O; Perlado, J M [Instituto de Fusion Nuclear (DENIM)/ETSII/Universidad Politecnica, Madrid (Spain); Lafuente, A; Sordo, F [Universidad Politecnica de Madrid (UPM), Madrid (Spain)], E-mail: yuri@denim.upm.es

    2008-05-15

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  11. 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.

  12. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    Directory of Open Access Journals (Sweden)

    Mashnik Stepan G.

    2016-01-01

    Full Text Available 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 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.

  13. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    CERN Document Server

    Mashnik, Stepan G

    2015-01-01

    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 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

  14. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    Science.gov (United States)

    Mashnik, Stepan G.; Kerby, Leslie M.

    2016-05-01

    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 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.

  15. Code Generation for Embedded Software for Modeling Clear Box Structures

    Directory of Open Access Journals (Sweden)

    V. Chandra Prakash

    2011-09-01

    Full Text Available Cleanroom software Engineering (CRSE recommended that the code related to the Application systems be generated either manually or through code generation models or represents the same as a hierarchy of clear box structures. CRSE has even advocated that the code be developed using the State models that models the internal behavior of the systems. No framework has been recommended by any Author using which the Clear boxes are designed using the code generation methods. Code Generation is one of the important quality issues addressed in cleanroom software engineering. It has been investigated that CRSE can be used for life cycle management of the embedded systems when the hardware-software co-design is in-built as part and parcel of CRSE by way of adding suitable models to CRSE and redefining the same. The design of Embedded Systems involves code generation in respect of hardware and Embedded Software. In this paper, a framework is proposed using which the embedded software is generated. The method is unique that it considers various aspects of the code generation which includes Code Segments, Code Functions, Classes, Globalization, Variable propagation etc. The proposed Framework has been applied to a Pilot project and the experimental results are presented.

  16. UW MCNP source patch for the EPFL Haefely source. EPFL (Swiss) fusion-fission hybrid experiment

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, G; Woodruff, G L

    1986-06-01

    The development of a source patch which describes the Haefely neutron source for use in the MCNP Monte Carlo code has been described in progress reports of the EPFL (Swiss) Fusion Blanket Project at the University of Washington. The most recent of these reports dealing with the source patch was Progress Report No. 14. This report reviews some of the physical description included in the report, and also includes additional details of the patch as well as a listing of the patch itself.

  17. 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.

  18. Automatic code generation from the OMT-based dynamic model

    Energy Technology Data Exchange (ETDEWEB)

    Ali, J.; Tanaka, J.

    1996-12-31

    The OMT object-oriented software development methodology suggests creating three models of the system, i.e., object model, dynamic model and functional model. We have developed a system that automatically generates implementation code from the dynamic model. The system first represents the dynamic model as a table and then generates executable Java language code from it. We used inheritance for super-substate relationships. We considered that transitions relate to states in a state diagram exactly as operations relate to classes in an object diagram. In the generated code, each state in the state diagram becomes a class and each event on a state becomes an operation on the corresponding class. The system is implemented and can generate executable code for any state diagram. This makes the role of the dynamic model more significant and the job of designers even simpler.

  19. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes.

  20. MONTE CARLO ANALYSES OF THE YALINA THERMAL FACILITY WITH SERPENT STEREOLITHOGRAPHY GEOMETRY MODEL

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, Y.

    2015-01-01

    This paper analyzes the YALINA Thermal subcritical assembly of Belarus using two different Monte Carlo transport programs, SERPENT and MCNP. The MCNP model is based on combinatorial geometry and universes hierarchy, while the SERPENT model is based on Stereolithography geometry. The latter consists of unstructured triangulated surfaces defined by the normal and vertices. This geometry format is used by 3D printers and it has been created by: the CUBIT software, MATLAB scripts, and C coding. All the Monte Carlo simulations have been performed using the ENDF/B-VII.0 nuclear data library. Both MCNP and SERPENT share the same geometry specifications, which describe the facility details without using any material homogenization. Three different configurations have been studied with different number of fuel rods. The three fuel configurations use 216, 245, or 280 fuel rods, respectively. The numerical simulations show that the agreement between SERPENT and MCNP results is within few tens of pcms.

  1. Noise Residual Learning for Noise Modeling in Distributed Video Coding

    DEFF Research Database (Denmark)

    Luong, Huynh Van; Forchhammer, Søren

    2012-01-01

    Distributed video coding (DVC) is a coding paradigm which exploits the source statistics at the decoder side to reduce the complexity at the encoder. The noise model is one of the inherently difficult challenges in DVC. This paper considers Transform Domain Wyner-Ziv (TDWZ) coding and proposes...... decoding. A residual refinement step is also introduced to take advantage of correlation of DCT coefficients. Experimental results show that the proposed techniques robustly improve the coding efficiency of TDWZ DVC and for GOP=2 bit-rate savings up to 35% on WZ frames are achieved compared with DISCOVER....

  2. A multi-platform linking code for fuel burnup and radiotoxicity analysis

    Science.gov (United States)

    Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

    2014-02-01

    A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

  3. Coupling a Basin Modeling and a Seismic Code using MOAB

    KAUST Repository

    Yan, Mi

    2012-06-02

    We report on a demonstration of loose multiphysics coupling between a basin modeling code and a seismic code running on a large parallel machine. Multiphysics coupling, which is one critical capability for a high performance computing (HPC) framework, was implemented using the MOAB open-source mesh and field database. MOAB provides for code coupling by storing mesh data and input and output field data for the coupled analysis codes and interpolating the field values between different meshes used by the coupled codes. We found it straightforward to use MOAB to couple the PBSM basin modeling code and the FWI3D seismic code on an IBM Blue Gene/P system. We describe how the coupling was implemented and present benchmarking results for up to 8 racks of Blue Gene/P with 8192 nodes and MPI processes. The coupling code is fast compared to the analysis codes and it scales well up to at least 8192 nodes, indicating that a mesh and field database is an efficient way to implement loose multiphysics coupling for large parallel machines.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Sakurai, Kiyoshi; Tonoike, Kotaro; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    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)

  5. 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

    The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy. Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation. Presented in this paper is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay systems

  6. 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.

  7. 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.)

  8. MCNP simulations of material exposure experiments (u)

    Energy Technology Data Exchange (ETDEWEB)

    Temple, Brian A [Los Alamos National Laboratory

    2010-12-08

    Simulations of proposed material exposure experiments were performed using MCNP6. The experiments will expose ampules containing different materials of interest with radiation to observe the chemical breakdown of the materials. Simulations were performed to map out dose in materials as a function of distance from the source, dose variation between materials, dose variation due to ampule orientation, and dose variation due to different source energy. This write up is an overview of the simulations and will provide guidance on how to use the data in the spreadsheet.

  9. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-04-15

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

  10. A unified model of the standard genetic code.

    Science.gov (United States)

    José, Marco V; Zamudio, Gabriel S; Morgado, Eberto R

    2017-03-01

    The Rodin-Ohno (RO) and the Delarue models divide the table of the genetic code into two classes of aminoacyl-tRNA synthetases (aaRSs I and II) with recognition from the minor or major groove sides of the tRNA acceptor stem, respectively. These models are asymmetric but they are biologically meaningful. On the other hand, the standard genetic code (SGC) can be derived from the primeval RNY code (R stands for purines, Y for pyrimidines and N any of them). In this work, the RO-model is derived by means of group actions, namely, symmetries represented by automorphisms, assuming that the SGC originated from a primeval RNY code. It turns out that the RO-model is symmetric in a six-dimensional (6D) hypercube. Conversely, using the same automorphisms, we show that the RO-model can lead to the SGC. In addition, the asymmetric Delarue model becomes symmetric by means of quotient group operations. We formulate isometric functions that convert the class aaRS I into the class aaRS II and vice versa. We show that the four polar requirement categories display a symmetrical arrangement in our 6D hypercube. Altogether these results cannot be attained, neither in two nor in three dimensions. We discuss the present unified 6D algebraic model, which is compatible with both the SGC (based upon the primeval RNY code) and the RO-model.

  11. Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

    CERN Document Server

    Marshall, T; Topilski, L; Merrill, B

    2002-01-01

    For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

  12. RHOCUBE: 3D density distributions modeling code

    Science.gov (United States)

    Nikutta, Robert; Agliozzo, Claudia

    2016-11-01

    RHOCUBE models 3D density distributions on a discrete Cartesian grid and their integrated 2D maps. It can be used for a range of applications, including modeling the electron number density in LBV shells and computing the emission measure. The RHOCUBE Python package provides several 3D density distributions, including a powerlaw shell, truncated Gaussian shell, constant-density torus, dual cones, and spiralling helical tubes, and can accept additional distributions. RHOCUBE provides convenient methods for shifts and rotations in 3D, and if necessary, an arbitrary number of density distributions can be combined into the same model cube and the integration ∫ dz performed through the joint density field.

  13. LMFBR models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1983-06-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-/sup 233/U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given.

  14. LMFBR models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1981-10-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th-/sup 238/U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given.

  15. 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.

  16. Information Theoretic Authentication and Secrecy Codes in the Splitting Model

    CERN Document Server

    Huber, Michael

    2011-01-01

    In the splitting model, information theoretic authentication codes allow non-deterministic encoding, that is, several messages can be used to communicate a particular plaintext. Certain applications require that the aspect of secrecy should hold simultaneously. Ogata-Kurosawa-Stinson-Saido (2004) have constructed optimal splitting authentication codes achieving perfect secrecy for the special case when the number of keys equals the number of messages. In this paper, we establish a construction method for optimal splitting authentication codes with perfect secrecy in the more general case when the number of keys may differ from the number of messages. To the best knowledge, this is the first result of this type.

  17. JPEG2000 COMPRESSION CODING USING HUMAN VISUAL SYSTEM MODEL

    Institute of Scientific and Technical Information of China (English)

    Xiao Jiang; Wu Chengke

    2005-01-01

    In order to apply the Human Visual System (HVS) model to JPEG2000 standard,several implementation alternatives are discussed and a new scheme of visual optimization isintroduced with modifying the slope of rate-distortion. The novelty is that the method of visual weighting is not lifting the coefficients in wavelet domain, but is complemented by code stream organization. It remains all the features of Embedded Block Coding with Optimized Truncation (EBCOT) such as resolution progressive, good robust for error bit spread and compatibility of lossless compression. Well performed than other methods, it keeps the shortest standard codestream and decompression time and owns the ability of VIsual Progressive (VIP) coding.

  18. Differences between the 1993 and 1995 CABO Model Energy Codes

    Energy Technology Data Exchange (ETDEWEB)

    Conover, D.R.; Lucas, R.G.

    1995-10-01

    The Energy Policy Act of 1992 requires the US DOE to determine if changes to the Council of American Building Officials` (CABO) 1993 Model Energy Code (MEC) (CABO 1993), published in the 1995 edition of the MEC (CABO 1995), will improve energy efficiency in residential buildings. The DOE, the states, and others have expressed an interest in the differences between the 1993 and 1995 editions of the MEC. This report describes each change to the 1993 MEC, and its impact. Referenced publications are also listed along with discrepancies between code changes approved in the 1994 and 1995 code-change cycles and what actually appears in the 1995 MEC.

  19. Modeling Guidelines for Code Generation in the Railway Signaling Context

    Science.gov (United States)

    Ferrari, Alessio; Bacherini, Stefano; Fantechi, Alessandro; Zingoni, Niccolo

    2009-01-01

    Modeling guidelines constitute one of the fundamental cornerstones for Model Based Development. Their relevance is essential when dealing with code generation in the safety-critical domain. This article presents the experience of a railway signaling systems manufacturer on this issue. Introduction of Model-Based Development (MBD) and code generation in the industrial safety-critical sector created a crucial paradigm shift in the development process of dependable systems. While traditional software development focuses on the code, with MBD practices the focus shifts to model abstractions. The change has fundamental implications for safety-critical systems, which still need to guarantee a high degree of confidence also at code level. Usage of the Simulink/Stateflow platform for modeling, which is a de facto standard in control software development, does not ensure by itself production of high-quality dependable code. This issue has been addressed by companies through the definition of modeling rules imposing restrictions on the usage of design tools components, in order to enable production of qualified code. The MAAB Control Algorithm Modeling Guidelines (MathWorks Automotive Advisory Board)[3] is a well established set of publicly available rules for modeling with Simulink/Stateflow. This set of recommendations has been developed by a group of OEMs and suppliers of the automotive sector with the objective of enforcing and easing the usage of the MathWorks tools within the automotive industry. The guidelines have been published in 2001 and afterwords revisited in 2007 in order to integrate some additional rules developed by the Japanese division of MAAB [5]. The scope of the current edition of the guidelines ranges from model maintainability and readability to code generation issues. The rules are conceived as a reference baseline and therefore they need to be tailored to comply with the characteristics of each industrial context. Customization of these

  20. An assessment of the MCNP4C weight window

    Energy Technology Data Exchange (ETDEWEB)

    Christopher N. Culbertson; John S. Hendricks

    1999-12-01

    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.

  1. 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.

  2. Student Model Tools Code Release and Documentation

    DEFF Research Database (Denmark)

    Johnson, Matthew; Bull, Susan; Masci, Drew

    This document contains a wealth of information about the design and implementation of the Next-TELL open learner model. Information is included about the final specification (Section 3), the interfaces and features (Section 4), its implementation and technical design (Section 5) and also a summary...

  3. Code-to-Code Comparison, and Material Response Modeling of Stardust and MSL using PATO and FIAT

    Science.gov (United States)

    Omidy, Ali D.; Panerai, Francesco; Martin, Alexandre; Lachaud, Jean R.; Cozmuta, Ioana; Mansour, Nagi N.

    2015-01-01

    This report provides a code-to-code comparison between PATO, a recently developed high fidelity material response code, and FIAT, NASA's legacy code for ablation response modeling. The goal is to demonstrates that FIAT and PATO generate the same results when using the same models. Test cases of increasing complexity are used, from both arc-jet testing and flight experiment. When using the exact same physical models, material properties and boundary conditions, the two codes give results that are within 2% of errors. The minor discrepancy is attributed to the inclusion of the gas phase heat capacity (cp) in the energy equation in PATO, and not in FIAT.

  4. The JCSS probabilistic model code: Experience and recent developments

    NARCIS (Netherlands)

    Chryssanthopoulos, M.; Diamantidis, D.; Vrouwenvelder, A.C.W.M.

    2003-01-01

    The JCSS Probabilistic Model Code (JCSS-PMC) has been available for public use on the JCSS website (www.jcss.ethz.ch) for over two years. During this period, several examples have been worked out and new probabilistic models have been added. Since the engineering community has already been exposed t

  5. A Mathematical Model for Comparing Holland's Personality and Environmental Codes.

    Science.gov (United States)

    Kwak, Junkyu Christopher; Pulvino, Charles J.

    1982-01-01

    Presents a mathematical model utilizing three-letter codes of personality patterns determined from the Self Directed Search. This model compares personality types over time or determines relationships between personality types and person-environment interactions. This approach is consistent with Holland's theory yet more comprehensive than one- or…

  6. Model-building codes for membrane proteins.

    Energy Technology Data Exchange (ETDEWEB)

    Shirley, David Noyes; Hunt, Thomas W.; Brown, W. Michael; Schoeniger, Joseph S. (Sandia National Laboratories, Livermore, CA); Slepoy, Alexander; Sale, Kenneth L. (Sandia National Laboratories, Livermore, CA); Young, Malin M. (Sandia National Laboratories, Livermore, CA); Faulon, Jean-Loup Michel; Gray, Genetha Anne (Sandia National Laboratories, Livermore, CA)

    2005-01-01

    We have developed a novel approach to modeling the transmembrane spanning helical bundles of integral membrane proteins using only a sparse set of distance constraints, such as those derived from MS3-D, dipolar-EPR and FRET experiments. Algorithms have been written for searching the conformational space of membrane protein folds matching the set of distance constraints, which provides initial structures for local conformational searches. Local conformation search is achieved by optimizing these candidates against a custom penalty function that incorporates both measures derived from statistical analysis of solved membrane protein structures and distance constraints obtained from experiments. This results in refined helical bundles to which the interhelical loops and amino acid side-chains are added. Using a set of only 27 distance constraints extracted from the literature, our methods successfully recover the structure of dark-adapted rhodopsin to within 3.2 {angstrom} of the crystal structure.

  7. Analysis of results of AZTRAN and AZKIND codes for a BWR; Analisis de resultados de los codigos AZTRAN y AZKIND para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Vallejo Q, J. A.; Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Xolocostli M, J. V.; Rodriguez H, A.; Gomez T, A. M., E-mail: gbo729@yahoo.com.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  8. Multiview coding mode decision with hybrid optimal stopping model.

    Science.gov (United States)

    Zhao, Tiesong; Kwong, Sam; Wang, Hanli; Wang, Zhou; Pan, Zhaoqing; Kuo, C-C Jay

    2013-04-01

    In a generic decision process, optimal stopping theory aims to achieve a good tradeoff between decision performance and time consumed, with the advantages of theoretical decision-making and predictable decision performance. In this paper, optimal stopping theory is employed to develop an effective hybrid model for the mode decision problem, which aims to theoretically achieve a good tradeoff between the two interrelated measurements in mode decision, as computational complexity reduction and rate-distortion degradation. The proposed hybrid model is implemented and examined with a multiview encoder. To support the model and further promote coding performance, the multiview coding mode characteristics, including predicted mode probability and estimated coding time, are jointly investigated with inter-view correlations. Exhaustive experimental results with a wide range of video resolutions reveal the efficiency and robustness of our method, with high decision accuracy, negligible computational overhead, and almost intact rate-distortion performance compared to the original encoder.

  9. Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Isao; Miyamaru, Hiroyuki [Division of Electrical, Electronic and Information Engineering, Osaka University, Yamada-oka 2-1, Suita, Osaka, 565-0871 (Japan)

    2008-07-01

    Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%DELTAk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

  10. 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.

  11. Comparison of DT neutron production codes MCUNED, ENEA-JSI source subroutine and DDT

    Energy Technology Data Exchange (ETDEWEB)

    Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Kodeli, Ivan [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Milocco, Alberto [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sauvan, Patrick [Departamento de Ingeniería Energética, E.T.S. Ingenieros Industriales, UNED, C/Juan del Rosal 12, 28040 Madrid (Spain); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2016-11-01

    Highlights: • Results of three codes capable of simulating the accelerator based DT neutron generators were compared on a simple model where only a thin target made of mixture of titanium and tritium is present. Two typical deuteron beam energies, 100 keV and 250 keV, were used in the comparison. • Comparisons of the angular dependence of the total neutron flux and spectrum as well as the neutron spectrum of all the neutrons emitted from the target show general agreement of the results but also some noticeable differences. • A comparison of figures of merit of the calculations using different codes showed that the computational time necessary to achieve the same statistical uncertainty can vary for more than 30× when different codes for the simulation of the DT neutron generator are used. - Abstract: As the DT fusion reaction produces neutrons with energies significantly higher than in fission reactors, special fusion-relevant benchmark experiments are often performed using DT neutron generators. However, commonly used Monte Carlo particle transport codes such as MCNP or TRIPOLI cannot be directly used to analyze these experiments since they do not have the capabilities to model the production of DT neutrons. Three of the available approaches to model the DT neutron generator source are the MCUNED code, the ENEA-JSI DT source subroutine and the DDT code. The MCUNED code is an extension of the well-established and validated MCNPX Monte Carlo code. The ENEA-JSI source subroutine was originally prepared for the modelling of the FNG experiments using different versions of the MCNP code (−4, −5, −X) and was later extended to allow the modelling of both DT and DD neutron sources. The DDT code prepares the DT source definition file (SDEF card in MCNP) which can then be used in different versions of the MCNP code. In the paper the methods for the simulation of the DT neutron production used in the codes are briefly described and compared for the case of a

  12. Modelling spread of Bluetongue in Denmark: The code

    DEFF Research Database (Denmark)

    Græsbøll, Kaare

    This technical report was produced to make public the code produced as the main project of the PhD project by Kaare Græsbøll, with the title: "Modelling spread of Bluetongue and other vector borne diseases in Denmark and evaluation of intervention strategies".......This technical report was produced to make public the code produced as the main project of the PhD project by Kaare Græsbøll, with the title: "Modelling spread of Bluetongue and other vector borne diseases in Denmark and evaluation of intervention strategies"....

  13. Anthropomorphic Coding of Speech and Audio: A Model Inversion Approach

    Directory of Open Access Journals (Sweden)

    W. Bastiaan Kleijn

    2005-06-01

    Full Text Available Auditory modeling is a well-established methodology that provides insight into human perception and that facilitates the extraction of signal features that are most relevant to the listener. The aim of this paper is to provide a tutorial on perceptual speech and audio coding using an invertible auditory model. In this approach, the audio signal is converted into an auditory representation using an invertible auditory model. The auditory representation is quantized and coded. Upon decoding, it is then transformed back into the acoustic domain. This transformation converts a complex distortion criterion into a simple one, thus facilitating quantization with low complexity. We briefly review past work on auditory models and describe in more detail the components of our invertible model and its inversion procedure, that is, the method to reconstruct the signal from the output of the auditory model. We summarize attempts to use the auditory representation for low-bit-rate coding. Our approach also allows the exploitation of the inherent redundancy of the human auditory system for the purpose of multiple description (joint source-channel coding.

  14. Verification and Validation of Heat Transfer Model of AGREE Code

    Energy Technology Data Exchange (ETDEWEB)

    Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)

    2013-05-15

    The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.

  15. Code Shift: Grid Specifications and Dynamic Wind Turbine Models

    DEFF Research Database (Denmark)

    Ackermann, Thomas; Ellis, Abraham; Fortmann, Jens

    2013-01-01

    Grid codes (GCs) and dynamic wind turbine (WT) models are key tools to allow increasing renewable energy penetration without challenging security of supply. In this article, the state of the art and the further development of both tools are discussed, focusing on the European and North American...

  16. Modelling binary rotating stars by new population synthesis code BONNFIRES

    CERN Document Server

    Lau, Herbert H B; Schneider, Fabian R N

    2013-01-01

    BONNFIRES, a new generation of population synthesis code, can calculate nuclear reaction, various mixing processes and binary interaction in a timely fashion. We use this new population synthesis code to study the interplay between binary mass transfer and rotation. We aim to compare theoretical models with observations, in particular the surface nitrogen abundance and rotational velocity. Preliminary results show binary interactions may explain the formation of nitrogen-rich slow rotators and nitrogen-poor fast rotators, but more work needs to be done to estimate whether the observed frequencies of those stars can be matched.

  17. Modeling of Anomalous Transport in Tokamaks with FACETS code

    Science.gov (United States)

    Pankin, A. Y.; Batemann, G.; Kritz, A.; Rafiq, T.; Vadlamani, S.; Hakim, A.; Kruger, S.; Miah, M.; Rognlien, T.

    2009-05-01

    The FACETS code, a whole-device integrated modeling code that self-consistently computes plasma profiles for the plasma core and edge in tokamaks, has been recently developed as a part of the SciDAC project for core-edge simulations. A choice of transport models is available in FACETS through the FMCFM interface [1]. Transport models included in FMCFM have specific ranges of applicability, which can limit their use to parts of the plasma. In particular, the GLF23 transport model does not include the resistive ballooning effects that can be important in the tokamak pedestal region and GLF23 typically under-predicts the anomalous fluxes near the magnetic axis [2]. The TGLF and GYRO transport models have similar limitations [3]. A combination of transport models that covers the entire discharge domain is studied using FACETS in a realistic tokamak geometry. Effective diffusivities computed with the FMCFM transport models are extended to the region near the separatrix to be used in the UEDGE code within FACETS. 1. S. Vadlamani et al. (2009) %First time-dependent transport simulations using GYRO and NCLASS within FACETS (this meeting).2. T. Rafiq et al. (2009) %Simulation of electron thermal transport in H-mode discharges Submitted to Phys. Plasmas.3. C. Holland et al. (2008) %Validation of gyrokinetic transport simulations using %DIII-D core turbulence measurements Proc. of IAEA FEC (Switzerland, 2008)

  18. Model classification rate control algorithm for video coding

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    A model classification rate control method for video coding is proposed. The macro-blocks are classified according to their prediction errors, and different parameters are used in the rate-quantization and distortion-quantization model.The different model parameters are calculated from the previous frame of the same type in the process of coding. These models are used to estimate the relations among rate, distortion and quantization of the current frame. Further steps,such as R-D optimization based quantization adjustment and smoothing of quantization of adjacent macroblocks, are used to improve the quality. The results of the experiments prove that the technique is effective and can be realized easily. The method presented in the paper can be a good way for MPEG and H. 264 rate control.

  19. NLTE solar irradiance modeling with the COSI code

    CERN Document Server

    Shapiro, A I; Schoell, M; Haberreiter, M; Rozanov, E

    2010-01-01

    Context. The solar irradiance is known to change on time scales of minutes to decades, and it is suspected that its substantial fluctua- tions are partially responsible for climate variations. Aims. We are developing a solar atmosphere code that allows the physical modeling of the entire solar spectrum composed of quiet Sun and active regions. This code is a tool for modeling the variability of the solar irradiance and understanding its influence on Earth. Methods. We exploit further development of the radiative transfer code COSI that now incorporates the calculation of molecular lines. We validated COSI under the conditions of local thermodynamic equilibrium (LTE) against the synthetic spectra calculated with the ATLAS code. The synthetic solar spectra were also calculated in non-local thermodynamic equilibrium (NLTE) and compared to the available measured spectra. In doing so we have defined the main problems of the modeling, e.g., the lack of opacity in the UV part of the spectrum and the inconsistency in...

  20. A semianalytic Monte Carlo code for modelling LIDAR measurements

    Science.gov (United States)

    Palazzi, Elisa; Kostadinov, Ivan; Petritoli, Andrea; Ravegnani, Fabrizio; Bortoli, Daniele; Masieri, Samuele; Premuda, Margherita; Giovanelli, Giorgio

    2007-10-01

    LIDAR (LIght Detection and Ranging) is an optical active remote sensing technology with many applications in atmospheric physics. Modelling of LIDAR measurements appears useful approach for evaluating the effects of various environmental variables and scenarios as well as of different measurement geometries and instrumental characteristics. In this regard a Monte Carlo simulation model can provide a reliable answer to these important requirements. A semianalytic Monte Carlo code for modelling LIDAR measurements has been developed at ISAC-CNR. The backscattered laser signal detected by the LIDAR system is calculated in the code taking into account the contributions due to the main atmospheric molecular constituents and aerosol particles through processes of single and multiple scattering. The contributions by molecular absorption, ground and clouds reflection are evaluated too. The code can perform simulations of both monostatic and bistatic LIDAR systems. To enhance the efficiency of the Monte Carlo simulation, analytical estimates and expected value calculations are performed. Artificial devices (such as forced collision, local forced collision, splitting and russian roulette) are moreover foreseen by the code, which can enable the user to drastically reduce the variance of the calculation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes Group

    2016-06-17

    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_ace.pl and simple_ace_mg.pl.

  2. Retrieval of gamma cell 220 irradiator isodose curves with MCNP simulations and experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Grynberg, S.E.; Ferreira, A.V.; Belo, L.C.M.; Squair, P.L.; Ribeiro, M.A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Sousa, R.V.; Sebastiao, R.C.O. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Quimica

    2010-03-15

    Gamma irradiator facilities can be used in a wide range of applications such as biological and chemical researches, sterilization of medical devices and products. Dose mapping must be performed in these equipment in order to establish plant operational parameters, as dose uniformity, source utilization efficiency and maximum and minimum dose positions. The isodoses curves are measured using dosimeters or computer simulations. This work evaluates the absorbed dose in the CDTN/CNEN Gamma Cell Irradiation Facility, using the Monte Carlo N-Particles (MCNP) code. (author)

  3. 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)

  4. Enhancements to the SSME transfer function modeling code

    Science.gov (United States)

    Irwin, R. Dennis; Mitchell, Jerrel R.; Bartholomew, David L.; Glenn, Russell D.

    1995-01-01

    This report details the results of a one year effort by Ohio University to apply the transfer function modeling and analysis tools developed under NASA Grant NAG8-167 (Irwin, 1992), (Bartholomew, 1992) to attempt the generation of Space Shuttle Main Engine High Pressure Turbopump transfer functions from time domain data. In addition, new enhancements to the transfer function modeling codes which enhance the code functionality are presented, along with some ideas for improved modeling methods and future work. Section 2 contains a review of the analytical background used to generate transfer functions with the SSME transfer function modeling software. Section 2.1 presents the 'ratio method' developed for obtaining models of systems that are subject to single unmeasured excitation sources and have two or more measured output signals. Since most of the models developed during the investigation use the Eigensystem Realization Algorithm (ERA) for model generation, Section 2.2 presents an introduction of ERA, and Section 2.3 describes how it can be used to model spectral quantities. Section 2.4 details the Residue Identification Algorithm (RID) including the use of Constrained Least Squares (CLS) and Total Least Squares (TLS). Most of this information can be found in the report (and is repeated for convenience). Section 3 chronicles the effort of applying the SSME transfer function modeling codes to the a51p394.dat and a51p1294.dat time data files to generate transfer functions from the unmeasured input to the 129.4 degree sensor output. Included are transfer function modeling attempts using five methods. The first method is a direct application of the SSME codes to the data files and the second method uses the underlying trends in the spectral density estimates to form transfer function models with less clustering of poles and zeros than the models obtained by the direct method. In the third approach, the time data is low pass filtered prior to the modeling process in an

  5. Video Coding and Modeling with Applications to ATM Multiplexing

    Science.gov (United States)

    Nguyen, Hien

    A new vector quantization (VQ) coding method based on optimized concentric shell partitioning of the image space is proposed. The advantages of using the concentric shell partition vector quantizer (CSPVQ) are that it is very fast and the image patterns found in each different subspace can be more effectively coded by using a codebook that is best matched to that particular subspace. For intra-frame coding, the CSPVQ is shown to have the same performance, if not better, than the optimized gain-shape VQ in terms of encoded picture quality while it definitely surpasses the gain-shape VQ in term of computational complexity. A variable bit rate (VBR) video coder for moving video is then proposed where the idea of CSPVQ is coupled with the idea of regular quadtree decomposition to further reduce the bit rate of the encoded picture sequence. The usefulness of a quadtree coding technique comes from the fact that different homogeneous regions occurring within an image can be compactly represented by various nodes in a quadtree. It is found that this image representation technique is particularly useful in providing a low bit rate video encoder without compromising the image quality when it is used in conjunction with the CSPVQ. The characteristics of the VBR coder's output as applied to ATM transmission are investigated. Three video models are used to study the performance of the ATM multiplexer. These models are the auto regressive (AR) model, the auto regressive hidden Markov model (AR-HMM), and the fluid flow uniform arrival and service (UAS) model. The AR model is allowed to have arbitrary order and is used to model a video source which has a constant amount of motion, that is, a stationary video source. The AR-HMM is a more general video model which is based on the idea of auto regressive hidden Markov chain formulated by Baum and is used to describe highly non-stationary sources. Hence, it is expected that the AR-HMM model may also be used top represent a video

  6. Discovering binary codes for documents by learning deep generative models.

    Science.gov (United States)

    Hinton, Geoffrey; Salakhutdinov, Ruslan

    2011-01-01

    We describe a deep generative model in which the lowest layer represents the word-count vector of a document and the top layer represents a learned binary code for that document. The top two layers of the generative model form an undirected associative memory and the remaining layers form a belief net with directed, top-down connections. We present efficient learning and inference procedures for this type of generative model and show that it allows more accurate and much faster retrieval than latent semantic analysis. By using our method as a filter for a much slower method called TF-IDF we achieve higher accuracy than TF-IDF alone and save several orders of magnitude in retrieval time. By using short binary codes as addresses, we can perform retrieval on very large document sets in a time that is independent of the size of the document set using only one word of memory to describe each document.

  7. Using cryptology models for protecting PHP source code

    Science.gov (United States)

    Jevremović, Aleksandar; Ristić, Nenad; Veinović, Mladen

    2013-10-01

    Protecting PHP scripts from unwanted use, copying and modifications is a big issue today. Existing solutions on source code level are mostly working as obfuscators, they are free, and they are not providing any serious protection. Solutions that encode opcode are more secure, but they are commercial and require closed-source proprietary PHP interpreter's extension. Additionally, encoded opcode is not compatible with future versions of interpreters which imply re-buying encoders from the authors. Finally, if extension source-code is compromised, all scripts encoded with that solution are compromised too. In this paper, we will present a new model for free and open-source PHP script protection solution. Protection level provided by the proposed solution is equal to protection level of commercial solutions. Model is based on conclusions from use of standard cryptology models for analysis of strengths and weaknesses of the existing solutions, when a scripts protection is seen as secure communication channel in the cryptology.

  8. MCNP Calculations for the Shielding Design of a Beam Tube to Be Installed at the Portuguese Research Reactor

    Science.gov (United States)

    Gonçalves, I. F.; Ramalho, A. G.; Gonçalves, I. C.; Salgado, J.

    The work presented concerns the calculation of the external biological shielding for a neutron beam tube that will be installed at the Portuguese Research Reactor, RPI. This tube will have enough versatility to be used in fields so different as the analysis of the composition of samples or research work in Boron Neutron Capture Therapy, BNCT. The calculation was made by using the MCNP code. This code is a well validated and widely used code, and has therefore become an important tool in the design and optimisation work of experiences related to neutrons and gamma radiation.

  9. A spectral synthesis code for rapid modelling of supernovae

    CERN Document Server

    Kerzendorf, Wolfgang E

    2014-01-01

    We present TARDIS - an open-source code for rapid spectral modelling of supernovae (SNe). Our goal is to develop a tool that is sufficiently fast to allow exploration of the complex parameter spaces of models for SN ejecta. This can be used to analyse the growing number of high-quality SN spectra being obtained by transient surveys. The code uses Monte Carlo methods to obtain a self-consistent description of the plasma state and to compute a synthetic spectrum. It has a modular design to facilitate the implementation of a range of physical approximations that can be compared to asses both accuracy and computational expediency. This will allow users to choose a level of sophistication appropriate for their application. Here, we describe the operation of the code and make comparisons with alternative radiative transfer codes of differing levels of complexity (SYN++, PYTHON, and ARTIS). We then explore the consequence of adopting simple prescriptions for the calculation of atomic excitation, focussing on four sp...

  10. Transform Coding for Point Clouds Using a Gaussian Process Model.

    Science.gov (United States)

    De Queiroz, Ricardo; Chou, Philip A

    2017-04-28

    We propose using stationary Gaussian Processes (GPs) to model the statistics of the signal on points in a point cloud, which can be considered samples of a GP at the positions of the points. Further, we propose using Gaussian Process Transforms (GPTs), which are Karhunen-Lo`eve transforms of the GP, as the basis of transform coding of the signal. Focusing on colored 3D point clouds, we propose a transform coder that breaks the point cloud into blocks, transforms the blocks using GPTs, and entropy codes the quantized coefficients. The GPT for each block is derived from both the covariance function of the GP and the locations of the points in the block, which are separately encoded. The covariance function of the GP is parameterized, and its parameters are sent as side information. The quantized coefficients are sorted by eigenvalues of the GPTs, binned, and encoded using an arithmetic coder with bin-dependent Laplacian models whose parameters are also sent as side information. Results indicate that transform coding of 3D point cloud colors using the proposed GPT and entropy coding achieves superior compression performance on most of our data sets.

  11. The 2010 fib Model Code for Structural Concrete: A new approach to structural engineering

    NARCIS (Netherlands)

    Walraven, J.C.; Bigaj-Van Vliet, A.

    2011-01-01

    The fib Model Code is a recommendation for the design of reinforced and prestressed concrete which is intended to be a guiding document for future codes. Model Codes have been published before, in 1978 and 1990. The draft for fib Model Code 2010 was published in May 2010. The most important new elem

  12. Determining the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5

    Energy Technology Data Exchange (ETDEWEB)

    Paxton, Adam B.; Davis, Stephen D.; DeWerd, Larry A. [Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 (United States); Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 and McGill University Health Centre, Department of Medical Physics, Montreal, Quebec H3G 1A4 (Canada); Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 (United States)

    2012-03-15

    Purpose: Recent advances in the imaging of {sup 90}Y using positron emission tomography (PET) and improved uncertainty in the branching ratio for the internal pair production component of {sup 90}Y decay allow for a more accurate determination of the activity distribution of {sup 90}Y microspheres within a patient. This improved activity distribution can be convolved with the dose kernel of {sup 90}Y to calculate the dose distribution within a patient. This work investigates the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5 and compares the results of these two transport codes. Methods: Monte Carlo simulations were performed with egsnrc and mcnp5 to calculate the dose rate at multiple radial distances around various {sup 90}Y sources. Point source simulations were completed with mcnp5 to determine the optimal electron transport settings for this work. After determining the optimal settings, point source simulations were completed using egsnrc (user code edknrc) and mcnp5 in water and liver [as defined by the International Commission on Radiation Units and Measurements (ICRU) Report 44]. The results were compared to ICRU Report 72 reference data. Point source simulations were also completed in water with a density of 1.06 g{center_dot}cm{sup -3} to evaluate the effect of the density of the surrounding material. Glass and resin microsphere simulations were performed with average and maximum diameter and density values (based on values given in the literature) in water and in liver. The results were compared to point source simulation results using the same transport code and in the same surrounding material. All simulations had statistical uncertainties less than 1%. Results: The optimal transport settings in mcnp5 for this work included using the energy-and step-specific algorithm (DBCN 17J 2) and ESTEP set to 10. These settings were used for all subsequent simulations with mcnp5. The point source

  13. Improvement of Basic Fluid Dynamics Models for the COMPASS Code

    Science.gov (United States)

    Zhang, Shuai; Morita, Koji; Shirakawa, Noriyuki; Yamamoto, Yuichi

    The COMPASS code is a new next generation safety analysis code to provide local information for various key phenomena in core disruptive accidents of sodium-cooled fast reactors, which is based on the moving particle semi-implicit (MPS) method. In this study, improvement of basic fluid dynamics models for the COMPASS code was carried out and verified with fundamental verification calculations. A fully implicit pressure solution algorithm was introduced to improve the numerical stability of MPS simulations. With a newly developed free surface model, numerical difficulty caused by poor pressure solutions is overcome by involving free surface particles in the pressure Poisson equation. In addition, applicability of the MPS method to interactions between fluid and multi-solid bodies was investigated in comparison with dam-break experiments with solid balls. It was found that the PISO algorithm and free surface model makes simulation with the passively moving solid model stable numerically. The characteristic behavior of solid balls was successfully reproduced by the present numerical simulations.

  14. New Mechanical Model for the Transmutation Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller

    2008-04-01

    A new mechanical model has been developed for implementation into the TRU fuel performance code. The new model differs from the existing FRAPCON 3 model, which it is intended to replace, in that it will include structural deformations (elasticity, plasticity, and creep) of the fuel. Also, the plasticity algorithm is based on the “plastic strain–total strain” approach, which should allow for more rapid and assured convergence. The model treats three situations relative to interaction between the fuel and cladding: (1) an open gap between the fuel and cladding, such that there is no contact, (2) contact between the fuel and cladding where the contact pressure is below a threshold value, such that axial slippage occurs at the interface, and (3) contact between the fuel and cladding where the contact pressure is above a threshold value, such that axial slippage is prevented at the interface. The first stage of development of the model included only the fuel. In this stage, results obtained from the model were compared with those obtained from finite element analysis using ABAQUS on a problem involving elastic, plastic, and thermal strains. Results from the two analyses showed essentially exact agreement through both loading and unloading of the fuel. After the cladding and fuel/clad contact were added, the model demonstrated expected behavior through all potential phases of fuel/clad interaction, and convergence was achieved without difficulty in all plastic analysis performed. The code is currently in stand alone form. Prior to implementation into the TRU fuel performance code, creep strains will have to be added to the model. The model will also have to be verified against an ABAQUS analysis that involves contact between the fuel and cladding.

  15. Galactic Cosmic Ray Event-Based Risk Model (GERM) Code

    Science.gov (United States)

    Cucinotta, Francis A.; Plante, Ianik; Ponomarev, Artem L.; Kim, Myung-Hee Y.

    2013-01-01

    This software describes the transport and energy deposition of the passage of galactic cosmic rays in astronaut tissues during space travel, or heavy ion beams in patients in cancer therapy. Space radiation risk is a probability distribution, and time-dependent biological events must be accounted for physical description of space radiation transport in tissues and cells. A stochastic model can calculate the probability density directly without unverified assumptions about shape of probability density function. The prior art of transport codes calculates the average flux and dose of particles behind spacecraft and tissue shielding. Because of the signaling times for activation and relaxation in the cell and tissue, transport code must describe temporal and microspatial density of functions to correlate DNA and oxidative damage with non-targeted effects of signals, bystander, etc. These are absolutely ignored or impossible in the prior art. The GERM code provides scientists data interpretation of experiments; modeling of beam line, shielding of target samples, and sample holders; and estimation of basic physical and biological outputs of their experiments. For mono-energetic ion beams, basic physical and biological properties are calculated for a selected ion type, such as kinetic energy, mass, charge number, absorbed dose, or fluence. Evaluated quantities are linear energy transfer (LET), range (R), absorption and fragmentation cross-sections, and the probability of nuclear interactions after 1 or 5 cm of water equivalent material. In addition, a set of biophysical properties is evaluated, such as the Poisson distribution for a specified cellular area, cell survival curves, and DNA damage yields per cell. Also, the GERM code calculates the radiation transport of the beam line for either a fixed number of user-specified depths or at multiple positions along the Bragg curve of the particle in a selected material. The GERM code makes the numerical estimates of basic

  16. Reflectance Prediction Modelling for Residual-Based Hyperspectral Image Coding

    Science.gov (United States)

    Xiao, Rui; Gao, Junbin; Bossomaier, Terry

    2016-01-01

    A Hyperspectral (HS) image provides observational powers beyond human vision capability but represents more than 100 times the data compared to a traditional image. To transmit and store the huge volume of an HS image, we argue that a fundamental shift is required from the existing “original pixel intensity”-based coding approaches using traditional image coders (e.g., JPEG2000) to the “residual”-based approaches using a video coder for better compression performance. A modified video coder is required to exploit spatial-spectral redundancy using pixel-level reflectance modelling due to the different characteristics of HS images in their spectral and shape domain of panchromatic imagery compared to traditional videos. In this paper a novel coding framework using Reflectance Prediction Modelling (RPM) in the latest video coding standard High Efficiency Video Coding (HEVC) for HS images is proposed. An HS image presents a wealth of data where every pixel is considered a vector for different spectral bands. By quantitative comparison and analysis of pixel vector distribution along spectral bands, we conclude that modelling can predict the distribution and correlation of the pixel vectors for different bands. To exploit distribution of the known pixel vector, we estimate a predicted current spectral band from the previous bands using Gaussian mixture-based modelling. The predicted band is used as the additional reference band together with the immediate previous band when we apply the HEVC. Every spectral band of an HS image is treated like it is an individual frame of a video. In this paper, we compare the proposed method with mainstream encoders. The experimental results are fully justified by three types of HS dataset with different wavelength ranges. The proposed method outperforms the existing mainstream HS encoders in terms of rate-distortion performance of HS image compression. PMID:27695102

  17. 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.

  18. Current Capabilities of the Fuel Performance Modeling Code PARFUME

    Energy Technology Data Exchange (ETDEWEB)

    G. K. Miller; D. A. Petti; J. T. Maki; D. L. Knudson

    2004-09-01

    The success of gas reactors depends upon the safety and quality of the coated particle fuel. A fuel performance modeling code (called PARFUME), which simulates the mechanical and physico-chemical behavior of fuel particles during irradiation, is under development at the Idaho National Engineering and Environmental Laboratory. Among current capabilities in the code are: 1) various options for calculating CO production and fission product gas release, 2) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or a prismatic block core, as well as through the layers of each analyzed particle, 3) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, kernel migration, and thinning of the SiC caused by interaction of fission products with the SiC, 4) two independent methods for determining particle failure probabilities, 5) a model for calculating release-to-birth (R/B) ratios of gaseous fission products, that accounts for particle failures and uranium contamination in the fuel matrix, and 6) the evaluation of an accident condition, where a particle experiences a sudden change in temperature following a period of normal irradiation. This paper presents an overview of the code.

  19. Direct containment heating models in the CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Washington, K.E.; Williams, D.C.

    1995-08-01

    The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.

  20. Modeling validation to structural flaws in the foundations of oil tanks; Validacao de modelagem para estudo de alteracoes estruturais em fundacoes de tanques de petroleo

    Energy Technology Data Exchange (ETDEWEB)

    Couto, Larissa Goncalves; Leite, Sandro Passos, E-mail: leite_sp@ig.com.br [Fundacao Tecnico-Educacional Souza Marques, Rio de Janeiro, RJ (Brazil). Faculdade de Engenharia; Pereira, Walsan Wagner [Instituto de Radioprotecao e Dosimetria, (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper presents the modeling of an experiment used to study the application of backscattered neutrons in the identification of structural flaws in the foundations of oil tanks. This modeling was a preliminary validation procedure of the method of calculation, performed with the radiation transport code MCNP, to study the application of backscattered neutrons as inspection tool. (author)

  1. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  2. 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.

  3. Benchmarking of computer codes and approaches for modeling exposure scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Seitz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Rittmann, P.D.; Wood, M.I. [Westinghouse Hanford Co., Richland, WA (United States); Cook, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1994-08-01

    The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided.

  4. Dynamic Model on the Transmission of Malicious Codes in Network

    Directory of Open Access Journals (Sweden)

    Bimal Kumar Mishra

    2013-08-01

    Full Text Available This paper introduces differential susceptible e-epidemic model S_i IR (susceptible class-1 for virus (S1 - susceptible class-2 for worms (S2 -susceptible class-3 for Trojan horse (S3 – infectious (I – recovered (R for the transmission of malicious codes in a computer network. We derive the formula for reproduction number (R0 to study the spread of malicious codes in computer network. We show that the Infectious free equilibrium is globally asymptotically stable and endemic equilibrium is locally asymptotically sable when reproduction number is less than one. Also an analysis has been made on the effect of antivirus software in the infectious nodes. Numerical methods are employed to solve and simulate the system of equations developed.

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

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, Gregg W [Los Alamos National Laboratory

    2012-07-17

    Currently the US lacks an end-to-end (i.e., source-to-detector) radiation transport simulation code with predictive capability for the broad range of DHS nuclear material detection applications. For example, gaps in the physics, along with inadequate analysis algorithms, make it difficult for Monte Carlo simulations to provide a comprehensive evaluation, design, and optimization of proposed interrogation systems. With the development and implementation of several key physics and algorithm enhancements, along with needed improvements in evaluated data and benchmark measurements, the MCNP/X Monte Carlo codes will provide designers, operators, and systems analysts with a validated tool for developing state-of-the-art active and passive detection systems. This project is currently in its seventh year (Phase VII). This presentation will review thirty enhancements that have been implemented in MCNPX over the last 3 years and were included in the 2011 release of version 2.7.0. These improvements include 12 physics enhancements, 4 source enhancements, 8 tally enhancements, and 6 other enhancements. Examples and results will be provided for each of these features. The presentation will also discuss the eight enhancements that will be migrated into MCNP6 over the upcoming year.

  6. IER-163 Post-Experiment MCNP Calculations (U)

    Energy Technology Data Exchange (ETDEWEB)

    Favorite, Jeffrey A. [Los Alamos National Laboratory

    2012-06-04

    IER-163 has been modeled with high fidelity in MCNP6. The model k{sub eff} was high, as in other similar calculations. The fission ratio {sup 238}U(n,f)/{sup 235}U(n,f) was 12.6% too small compared with measurements; the ratio {sup 239}Pu(n,f)/{sup 235}U(n,f) was 11.5% too small compared with measurements; the iridium ratio {sup 193}Ir(n,n{prime})/{sup 191}Ir(n,{gamma}) was 16.4% too large; and the gold ratios {sup 197}Au(n,2n)/{sup 197}Au(n,{gamma}), {sup 197}Au(n,2n)/{sup 235}U(n,f), and {sup 197}Au(n,{gamma})/{sup 235}U(n,f) were within one standard deviation of the measured values. It is suggested that the calculated {sup 235}U fission rate is too large and the calculated {sup 238}U fission rate is too small.

  7. Finite element code development for modeling detonation of HMX composites

    Science.gov (United States)

    Duran, Adam V.; Sundararaghavan, Veera

    2017-01-01

    In this work, we present a hydrodynamics code for modeling shock and detonation waves in HMX. A stable efficient solution strategy based on a Taylor-Galerkin finite element (FE) discretization was developed to solve the reactive Euler equations. In our code, well calibrated equations of state for the solid unreacted material and gaseous reaction products have been implemented, along with a chemical reaction scheme and a mixing rule to define the properties of partially reacted states. A linear Gruneisen equation of state was employed for the unreacted HMX calibrated from experiments. The JWL form was used to model the EOS of gaseous reaction products. It is assumed that the unreacted explosive and reaction products are in both pressure and temperature equilibrium. The overall specific volume and internal energy was computed using the rule of mixtures. Arrhenius kinetics scheme was integrated to model the chemical reactions. A locally controlled dissipation was introduced that induces a non-oscillatory stabilized scheme for the shock front. The FE model was validated using analytical solutions for SOD shock and ZND strong detonation models. Benchmark problems are presented for geometries in which a single HMX crystal is subjected to a shock condition.

  8. A Mutation Model from First Principles of the Genetic Code.

    Science.gov (United States)

    Thorvaldsen, Steinar

    2016-01-01

    The paper presents a neutral Codons Probability Mutations (CPM) model of molecular evolution and genetic decay of an organism. The CPM model uses a Markov process with a 20-dimensional state space of probability distributions over amino acids. The transition matrix of the Markov process includes the mutation rate and those single point mutations compatible with the genetic code. This is an alternative to the standard Point Accepted Mutation (PAM) and BLOcks of amino acid SUbstitution Matrix (BLOSUM). Genetic decay is quantified as a similarity between the amino acid distribution of proteins from a (group of) species on one hand, and the equilibrium distribution of the Markov chain on the other. Amino acid data for the eukaryote, bacterium, and archaea families are used to illustrate how both the CPM and PAM models predict their genetic decay towards the equilibrium value of 1. A family of bacteria is studied in more detail. It is found that warm environment organisms on average have a higher degree of genetic decay compared to those species that live in cold environments. The paper addresses a new codon-based approach to quantify genetic decay due to single point mutations compatible with the genetic code. The present work may be seen as a first approach to use codon-based Markov models to study how genetic entropy increases with time in an effectively neutral biological regime. Various extensions of the model are also discussed.

  9. Noise Feedback Coding Revisited: Refurbished Legacy Codecs and New Coding Models

    Institute of Scientific and Technical Information of China (English)

    Stephane Ragot; Balazs Kovesi; Alain Le Guyader

    2012-01-01

    Noise feedback coding (NFC) has attracted renewed interest with the recent standardization of backward-compatible enhancements for ITU-T G.711 and G.722. It has also been revisited with the emergence of proprietary speech codecs, such as BV16, BV32, and SILK, that have structures different from CELP coding. In this article, we review NFC and describe a novel coding technique that optimally shapes coding noise in embedded pulse-code modulation (PCM) and embedded adaptive differential PCM (ADPCM). We describe how this new technique was incorporated into the recent ITU-T G.711.1, G.711 App. III, and G.722 Annex B (G.722B) speech-coding standards.

  10. Availability of MCNP & MATLAB for reconstructing the water-vapor two-phase flow pattern in neutron radiography

    Institute of Scientific and Technical Information of China (English)

    FENG Qixi; FENG Quanke; TAKESHI Kawai

    2008-01-01

    The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008. In this paper, we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the tube were obtained using the MCNP code without influence of γ-ray and electronic-noise. The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated. The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI. The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques. And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, Enrico; Donahue, Richard J.; Blakely, Eleanor A.

    2002-09-11

    The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior.

  13. MMA, A Computer Code for Multi-Model Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Eileen P. Poeter and Mary C. Hill

    2007-08-20

    This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations.

  14. MMA, A Computer Code for Multi-Model Analysis

    Science.gov (United States)

    Poeter, Eileen P.; Hill, Mary C.

    2007-01-01

    This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will

  15. Recent developments in the Los Alamos radiation transport code system

    Energy Technology Data Exchange (ETDEWEB)

    Forster, R.A.; Parsons, K. [Los Alamos National Lab., NM (United States)

    1997-06-01

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

  16. EMPIRE: Nuclear Reaction Model Code System for Data Evaluation

    Science.gov (United States)

    Herman, M.; Capote, R.; Carlson, B. V.; Obložinský, P.; Sin, M.; Trkov, A.; Wienke, H.; Zerkin, V.

    2007-12-01

    EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, proton, any ion (including heavy-ions) or a photon. The energy range extends from the beginning of the unresolved resonance region for neutron-induced reactions (∽ keV) and goes up to several hundred MeV for heavy-ion induced reactions. The code accounts for the major nuclear reaction mechanisms, including direct, pre-equilibrium and compound nucleus ones. Direct reactions are described by a generalized optical model (ECIS03) or by the simplified coupled-channels approach (CCFUS). The pre-equilibrium mechanism can be treated by a deformation dependent multi-step direct (ORION + TRISTAN) model, by a NVWY multi-step compound one or by either a pre-equilibrium exciton model with cluster emission (PCROSS) or by another with full angular momentum coupling (DEGAS). Finally, the compound nucleus decay is described by the full featured Hauser-Feshbach model with γ-cascade and width-fluctuations. Advanced treatment of the fission channel takes into account transmission through a multiple-humped fission barrier with absorption in the wells. The fission probability is derived in the WKB approximation within the optical model of fission. Several options for nuclear level densities include the EMPIRE-specific approach, which accounts for the effects of the dynamic deformation of a fast rotating nucleus, the classical Gilbert-Cameron approach and pre-calculated tables obtained with a microscopic model based on HFB single-particle level schemes with collective enhancement. A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers, moments of inertia and γ-ray strength functions. The results can be converted into ENDF-6 formatted files using the

  17. Code-to-code benchmark tests for 3D simulation models dedicated to the extraction region in negative ion sources

    Science.gov (United States)

    Nishioka, S.; Mochalskyy, S.; Taccogna, F.; Hatayama, A.; Fantz, U.; Minelli, P.

    2017-08-01

    The development of the kinetic particle model for the extraction region in negative hydrogen ion sources is indispensable and helpful to clarify the H- beam extraction physics. Recently, various 3D kinetic particle codes have been developed to study the extraction mechanism. Direct comparison between each other has not yet been done. Therefore, we have carried out a code-to-code benchmark activity to validate our codes. In the present study, the progress in this benchmark activity is summarized. At present, the reasonable agreement with the result by each code have been obtained using realistic plasma parameters at least for the following items; (1) Potential profile in the case of the vacuum condition (2) Temporal evolution of extracted current densities and profiles of electric potential in the case of the plasma consisting of only electrons and positive ions.

  18. Acoustic Gravity Wave Chemistry Model for the RAYTRACE Code.

    Science.gov (United States)

    2014-09-26

    AU)-AI56 850 ACOlUSTIC GRAVITY WAVE CHEMISTRY MODEL FOR THE IAYTRACE I/~ CODE(U) MISSION RESEARCH CORP SANTA BARBIARA CA T E OLD Of MAN 84 MC-N-SlS...DNA-TN-S4-127 ONAOOI-BO-C-0022 UNLSSIFIlED F/O 20/14 NL 1-0 2-8 1111 po 312.2 1--I 11111* i •. AD-A 156 850 DNA-TR-84-127 ACOUSTIC GRAVITY WAVE...Hicih Frequency Radio Propaoation Acoustic Gravity Waves 20. ABSTRACT (Continue en reveree mide if tteceeemr and Identify by block number) This

  19. Implementation and qualification of MCNP 5 through the intercomparison with the benchmark for the calculation of critical systems Godiva and Jezebel; Implementacao e qualificacao do MCNP5 atraves da intercomparacao com o benchmark para o calculo dos sistemas criticos Godiva e Jezebel

    Energy Technology Data Exchange (ETDEWEB)

    Lara, Rafael G.; Maiorino, Jose R., E-mail: rafael.lara@aluno.ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2013-07-01

    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.

  20. Using MCNP to estimate nuclear energy deposition in a cold neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Lecot, Carlos A.; Hergenreder, Daniel F.; Lovotti, Osvaldo P. [INVAP S.A., San Carlos de Bariloche (Argentina). Nuclear Projects Department. Nuclear Engineering Division

    2002-07-01

    The location of a Cold Neutron Source (CNS) implies a careful cost/benefit balance between neutron performance and heat removal capacity of the required cryogenic equipment. To justify this balance, the calculation of the total heat deposited in the device is a critical parameter. It depends on many different contributions, i.e. neutron and gamma radiation, beta decay, fission product decay gammas, among others. With minor modifications to some standard cross section sets, the Monte Carlo code MCNP offers the possibility to calculate the total heat load in a single calculation, without the utilization of intermediate calculations and/or auxiliary codes. This paper describes the methodology used to modify the cross section sets, to calculate the energy deposited in the CNS and to evaluate the cold neutron flux which is the variable used to compare performance at different locations. (author)

  1. SIMULATE-4 multigroup nodal code with microscopic depletion model

    Energy Technology Data Exchange (ETDEWEB)

    Bahadir, T. [Studsvik Scandpower, Inc., Newton, MA (United States); Lindahl, St.O. [Studsvik Scandpower AB, Vasteras (Sweden); Palmtag, S.P. [Studsvik Scandpower, Inc., Idaho Falls, ID (United States)

    2005-07-01

    SIMULATE-4 is a three-dimensional multigroup analytical nodal code with microscopic depletion capability. It has been developed employing 'first principal models' thus avoiding ad hoc approximations. The multigroup diffusion equations or, optionally, the simplified P{sub 3} equations are solved. Cross sections are described by a hybrid microscopic-macroscopic model that includes approximately 50 heavy nuclides and fission products. Heterogeneities in the axial direction of an assembly are treated systematically. Radially, the assembly is divided into heterogeneous sub-meshes, thereby overcoming the shortcomings of spatially-averaged assembly cross sections and discontinuity factors generated with zero net-current boundary conditions. Numerical tests against higher order transport methods and critical experiments show substantial improvements compared to results of existing nodal models. (authors)

  2. C code generation applied to nonlinear model predictive control for an artificial pancreas

    DEFF Research Database (Denmark)

    Boiroux, Dimitri; Jørgensen, John Bagterp

    2017-01-01

    This paper presents a method to generate C code from MATLAB code applied to a nonlinear model predictive control (NMPC) algorithm. The C code generation uses the MATLAB Coder Toolbox. It can drastically reduce the time required for development compared to a manual porting of code from MATLAB to C...

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

    CERN Document Server

    Mainardi, E; Donahue, R J

    2002-01-01

    The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions of a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using ...

  4. Kinetic models of gene expression including non-coding RNAs

    Science.gov (United States)

    Zhdanov, Vladimir P.

    2011-03-01

    In cells, genes are transcribed into mRNAs, and the latter are translated into proteins. Due to the feedbacks between these processes, the kinetics of gene expression may be complex even in the simplest genetic networks. The corresponding models have already been reviewed in the literature. A new avenue in this field is related to the recognition that the conventional scenario of gene expression is fully applicable only to prokaryotes whose genomes consist of tightly packed protein-coding sequences. In eukaryotic cells, in contrast, such sequences are relatively rare, and the rest of the genome includes numerous transcript units representing non-coding RNAs (ncRNAs). During the past decade, it has become clear that such RNAs play a crucial role in gene expression and accordingly influence a multitude of cellular processes both in the normal state and during diseases. The numerous biological functions of ncRNAs are based primarily on their abilities to silence genes via pairing with a target mRNA and subsequently preventing its translation or facilitating degradation of the mRNA-ncRNA complex. Many other abilities of ncRNAs have been discovered as well. Our review is focused on the available kinetic models describing the mRNA, ncRNA and protein interplay. In particular, we systematically present the simplest models without kinetic feedbacks, models containing feedbacks and predicting bistability and oscillations in simple genetic networks, and models describing the effect of ncRNAs on complex genetic networks. Mathematically, the presentation is based primarily on temporal mean-field kinetic equations. The stochastic and spatio-temporal effects are also briefly discussed.

  5. A simple model of optimal population coding for sensory systems.

    Science.gov (United States)

    Doi, Eizaburo; Lewicki, Michael S

    2014-08-01

    A fundamental task of a sensory system is to infer information about the environment. It has long been suggested that an important goal of the first stage of this process is to encode the raw sensory signal efficiently by reducing its redundancy in the neural representation. Some redundancy, however, would be expected because it can provide robustness to noise inherent in the system. Encoding the raw sensory signal itself is also problematic, because it contains distortion and noise. The optimal solution would be constrained further by limited biological resources. Here, we analyze a simple theoretical model that incorporates these key aspects of sensory coding, and apply it to conditions in the retina. The model specifies the optimal way to incorporate redundancy in a population of noisy neurons, while also optimally compensating for sensory distortion and noise. Importantly, it allows an arbitrary input-to-output cell ratio between sensory units (photoreceptors) and encoding units (retinal ganglion cells), providing predictions of retinal codes at different eccentricities. Compared to earlier models based on redundancy reduction, the proposed model conveys more information about the original signal. Interestingly, redundancy reduction can be near-optimal when the number of encoding units is limited, such as in the peripheral retina. We show that there exist multiple, equally-optimal solutions whose receptive field structure and organization vary significantly. Among these, the one which maximizes the spatial locality of the computation, but not the sparsity of either synaptic weights or neural responses, is consistent with known basic properties of retinal receptive fields. The model further predicts that receptive field structure changes less with light adaptation at higher input-to-output cell ratios, such as in the periphery.

  6. Development of condensation modeling modeling and simulation code for IRWST

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Nyung; Jang, Wan Ho; Ko, Jong Hyun; Ha, Jong Baek; Yang, Chang Keun; Son, Myung Seong [Kyung Hee Univ., Seoul (Korea)

    1997-07-01

    One of the design improvements of the KNGR(Korean Next Generation Reactor) which is advanced to safety and economy is the adoption of IRWST(In-Containment Refueling Water Storage Tank). The IRWST, installed inside of the containment building, has more designed purpose than merely the location change of the tank. Since the design functions of the IRWST is similar to these of the BWR's suppression pool, theoretical models applicable to BWR's suppression pool can be mostly applied to the IRWST. But for the PWR, the geometry of the sparger, the operation mode and the steam quantity and temperature and pressure of discharged fluid from primary system to IRWST through PSV or SDV may be different from those of BWR. Also there is some defects in detailed parts of condensation model. Therefore we, as the first nation to construct PWR with IRWST, must carry out profound research for there problems such that the results can be utilized and localized as an exclusive technology. All kinds of thermal hydraulics phenomena was investigated and existing condensation models by Hideki Nariai and Izuo Aya were analyzed. Also throuh a rigorous literature review such as operation experience, experimental data, design document of KNGR, items which need modification and supplementation were derived. Analytical model for chugging phenomena is also presented. 15 refs., 18 figs., 4 tabs. (Author)

  7. A MATLAB based 3D modeling and inversion code for MT data

    Science.gov (United States)

    Singh, Arun; Dehiya, Rahul; Gupta, Pravin K.; Israil, M.

    2017-07-01

    The development of a MATLAB based computer code, AP3DMT, for modeling and inversion of 3D Magnetotelluric (MT) data is presented. The code comprises two independent components: grid generator code and modeling/inversion code. The grid generator code performs model discretization and acts as an interface by generating various I/O files. The inversion code performs core computations in modular form - forward modeling, data functionals, sensitivity computations and regularization. These modules can be readily extended to other similar inverse problems like Controlled-Source EM (CSEM). The modular structure of the code provides a framework useful for implementation of new applications and inversion algorithms. The use of MATLAB and its libraries makes it more compact and user friendly. The code has been validated on several published models. To demonstrate its versatility and capabilities the results of inversion for two complex models are presented.

  8. Characterization of bauxite residue (red mud) for (235)U, (238)U, (232)Th and (40)K using neutron activation analysis and the radiation dose levels as modeled by MCNP.

    Science.gov (United States)

    Landsberger, S; Sharp, A; Wang, S; Pontikes, Y; Tkaczyk, A H

    2017-07-01

    This study employs thermal and epithermal neutron activation analysis (NAA) to quantitatively and specifically determine absorption dose rates to various body parts from uranium, thorium and potassium. Specifically, a case study of bauxite residue (red mud) from an industrial facility was used to demonstrate the feasibility of the NAA approach for radiological safety assessment, using small sample sizes to ascertain the activities of (235)U, (238)U, (232)Th and (40)K. This proof-of-concept was shown to produce reliable results and a similar approach could be used for quantitative assessment of other samples with possible radiological significance. (238)U and (232)Th were determined by epithermal and thermal neutron activation analysis, respectively. (235)U was determined based on the known isotopic ratio of (238)U/(235)U. (40)K was also determined using epithermal neutron activation analysis to measure total potassium content and then subtracting its isotopic contribution. Furthermore, the work demonstrates the application of Monte Carlo Neutral-Particle (MCNP) simulations to estimate the radiation dose from large quantities of red mud, to assure the safety of humans and the surrounding environment. Phantoms were employed to observe the dose distribution throughout the human body demonstrating radiation effects on each individual organ. Copyright © 2016 Elsevier Ltd. All rights reserved.

  9. Secondary neutron source modelling using MCNPX and ALEPH codes

    Science.gov (United States)

    Trakas, Christos; Kerkar, Nordine

    2014-06-01

    Monitoring the subcritical state and divergence of reactors requires the presence of neutron sources. But mainly secondary neutrons from these sources feed the ex-core detectors (SRD, Source Range Detector) whose counting rate is correlated with the level of the subcriticality of reactor. In cycle 1, primary neutrons are provided by sources activated outside of the reactor (e.g. Cf252); part of this source can be used for the divergence of cycle 2 (not systematic). A second family of neutron sources is used for the second cycle: the spontaneous neutrons of actinides produced after irradiation of fuel in the first cycle. Both families of sources are not sufficient to efficiently monitor the divergence of the second cycles and following ones, in most reactors. Secondary sources cluster (SSC) fulfil this role. In the present case, the SSC [Sb, Be], after activation in the first cycle (production of Sb124, unstable), produces in subsequent cycles a photo-neutron source by gamma (from Sb124)-neutron (on Be9) reaction. This paper presents the model of the process between irradiation in cycle 1 and cycle 2 results for SRD counting rate at the beginning of cycle 2, using the MCNPX code and the depletion chain ALEPH-V1 (coupling of MCNPX and ORIGEN codes). The results of this simulation are compared with two experimental results of the PWR 1450 MWe-N4 reactors. A good agreement is observed between these results and the simulations. The subcriticality of the reactors is about at -15,000 pcm. Discrepancies on the SRD counting rate between calculations and measurements are in the order of 10%, lower than the combined uncertainty of measurements and code simulation. This comparison validates the AREVA methodology, which allows having an SRD counting rate best-estimate for cycles 2 and next ones and optimizing the position of the SSC, depending on the geographic location of sources, main parameter for optimal monitoring of subcritical states.

  10. TASS/SMR Code Topical Report for SMART Plant, Vol. I: Code Structure, System Models, and Solution Methods

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young Jong; Kim, Soo Hyoung; Kim, See Darl (and others)

    2008-10-15

    The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.

  11. Model code for energy conservation in new building construction

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-12-01

    In response to the recognized lack of existing consensus standards directed to the conservation of energy in building design and operation, the preparation and publication of such a standard was accomplished with the issuance of ASHRAE Standard 90-75 ''Energy Conservation in New Building Design,'' by the American Society of Heating, Refrigerating and Air-Conditioning Engineers, Inc., in 1975. This standard addressed itself to recommended practices for energy conservation, using both depletable and non-depletable sources. A model code for energy conservation in building construction has been developed, setting forth the minimum regulations found necessary to mandate such conservation. The code addresses itself to the administration, design criteria, systems elements, controls, service water heating and electrical distribution and use, both for depletable and non-depletable energy sources. The technical provisions of the document are based on ASHRAE 90-75 and it is intended for use by state and local building officials in the implementation of a statewide energy conservation program.

  12. Mathematical modeling of wiped-film evaporators. [MAIN codes

    Energy Technology Data Exchange (ETDEWEB)

    Sommerfeld, J.T.

    1976-05-01

    A mathematical model and associated computer program were developed to simulate the steady-state operation of wiped-film evaporators for the concentration of typical waste solutions produced at the Savannah River Plant. In this model, which treats either a horizontal or a vertical wiped-film evaporator as a plug-flow device with no backmixing, three fundamental phenomena are described: sensible heating of the waste solution, vaporization of water, and crystallization of solids from solution. Physical property data were coded into the computer program, which performs the calculations of this model. Physical properties of typical waste solutions and of the heating steam, generally as analytical functions of temperature, were obtained from published data or derived by regression analysis of tabulated or graphical data. Preliminary results from tests of the Savannah River Laboratory semiworks wiped-film evaporators were used to select a correlation for the inside film heat transfer coefficient. This model should be a useful aid in the specification, operation, and control of the full-scale wiped-film evaporators proposed for application under plant conditions. In particular, it should be of value in the development and analysis of feed-forward control schemes for the plant units. Also, this model can be readily adapted, with only minor changes, to simulate the operation of wiped-film evaporators for other conceivable applications, such as the concentration of acid wastes.

  13. CODE's new solar radiation pressure model for GNSS orbit determination

    Science.gov (United States)

    Arnold, D.; Meindl, M.; Beutler, G.; Dach, R.; Schaer, S.; Lutz, S.; Prange, L.; Sośnica, K.; Mervart, L.; Jäggi, A.

    2015-08-01

    The Empirical CODE Orbit Model (ECOM) of the Center for Orbit Determination in Europe (CODE), which was developed in the early 1990s, is widely used in the International GNSS Service (IGS) community. For a rather long time, spurious spectral lines are known to exist in geophysical parameters, in particular in the Earth Rotation Parameters (ERPs) and in the estimated geocenter coordinates, which could recently be attributed to the ECOM. These effects grew creepingly with the increasing influence of the GLONASS system in recent years in the CODE analysis, which is based on a rigorous combination of GPS and GLONASS since May 2003. In a first step we show that the problems associated with the ECOM are to the largest extent caused by the GLONASS, which was reaching full deployment by the end of 2011. GPS-only, GLONASS-only, and combined GPS/GLONASS solutions using the observations in the years 2009-2011 of a global network of 92 combined GPS/GLONASS receivers were analyzed for this purpose. In a second step we review direct solar radiation pressure (SRP) models for GNSS satellites. We demonstrate that only even-order short-period harmonic perturbations acting along the direction Sun-satellite occur for GPS and GLONASS satellites, and only odd-order perturbations acting along the direction perpendicular to both, the vector Sun-satellite and the spacecraft's solar panel axis. Based on this insight we assess in the third step the performance of four candidate orbit models for the future ECOM. The geocenter coordinates, the ERP differences w. r. t. the IERS 08 C04 series of ERPs, the misclosures for the midnight epochs of the daily orbital arcs, and scale parameters of Helmert transformations for station coordinates serve as quality criteria. The old and updated ECOM are validated in addition with satellite laser ranging (SLR) observations and by comparing the orbits to those of the IGS and other analysis centers. Based on all tests, we present a new extended ECOM which

  14. 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.

  15. Molecular Code Division Multiple Access: Gaussian Mixture Modeling

    Science.gov (United States)

    Zamiri-Jafarian, Yeganeh

    Communications between nano-devices is an emerging research field in nanotechnology. Molecular Communication (MC), which is a bio-inspired paradigm, is a promising technique for communication in nano-network. In MC, molecules are administered to exchange information among nano-devices. Due to the nature of molecular signals, traditional communication methods can't be directly applied to the MC framework. The objective of this thesis is to present novel diffusion-based MC methods when multi nano-devices communicate with each other in the same environment. A new channel model and detection technique, along with a molecular-based access method, are proposed in here for communication between asynchronous users. In this work, the received molecular signal is modeled as a Gaussian mixture distribution when the MC system undergoes Brownian noise and inter-symbol interference (ISI). This novel approach demonstrates a suitable modeling for diffusion-based MC system. Using the proposed Gaussian mixture model, a simple receiver is designed by minimizing the error probability. To determine an optimum detection threshold, an iterative algorithm is derived which minimizes a linear approximation of the error probability function. Also, a memory-based receiver is proposed to improve the performance of the MC system by considering previously detected symbols in obtaining the threshold value. Numerical evaluations reveal that theoretical analysis of the bit error rate (BER) performance based on the Gaussian mixture model match simulation results very closely. Furthermore, in this thesis, molecular code division multiple access (MCDMA) is proposed to overcome the inter-user interference (IUI) caused by asynchronous users communicating in a shared propagation environment. Based on the selected molecular codes, a chip detection scheme with an adaptable threshold value is developed for the MCDMA system when the proposed Gaussian mixture model is considered. Results indicate that the

  16. Modeling of the CTEx subcritical unit using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Avelino [Divisao de Defesa Quimica, Biologica e Nuclear. Centro Tecnologico do Exercito - CTEx, Guaratiba, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@con.ufrj.br [Programa de Engenharia Nuclear. Universidade Federal do Rio de Janeiro - UFRJ Centro de Tecnologia, Rio de Janeiro, RJ (Brazil); Rebello, Wilson F. [Secao de Engenharia Nuclear - SE/7 Instituto Militar de Engenharia - IME Rio de Janeiro, RJ (Brazil); Cunha, Victor L. Lassance [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The present work aims at simulating the subcritical unit of Army Technology Center (CTEx) namely ARGUS pile (subcritical uranium-graphite arrangement) by using the computational code MCNPX. Once such modeling is finished, it could be used in k-effective calculations for systems using natural uranium as fuel, for instance. ARGUS is a subcritical assembly which uses reactor-grade graphite as moderator of fission neutrons and metallic uranium fuel rods with aluminum cladding. The pile is driven by an Am-Be spontaneous neutron source. In order to achieve a higher value for k{sub eff}, a higher concentration of U235 can be proposed, provided it safely remains below one. (author)

  17. Djehuty, a Code for Modeling Stars in Three Dimensions

    CERN Document Server

    Bazán, G; Dossa, D D; Eggleton, P P; Taylor, A; Castor, J I; Murray, S; Cook, K H; Eltgroth, P G; Cavallo, R M; Turcotte, S; Keller, S C; Pudliner, B S

    2003-01-01

    Current practice in stellar evolution is to employ one-dimensional calculations that quantitatively apply only to a minority of the observed stars (single non-rotating stars, or well detached binaries). Even in these systems, astrophysicists are dependent on approximations to handle complex three-dimensional processes like convection. Understanding the structure of binary stars, like those that lead to the Type Ia supernovae used to measure the expansion of the universe, are grossly non-spherical and await a 3D treatment. To approach very large problems like multi-dimensional modeling of stars, the Lawrence Livermore National Laboratory has invested in massively parallel computers and invested even more in developing the algorithms to utilize them on complex physics problems. We have leveraged skills from across the lab to develop a 3D stellar evolution code, Djehuty (after the Egyptian god for writing and calculation) that operates efficiently on platforms with thousands of nodes, with the best available phy...

  18. Maximizing entropy of image models for 2-D constrained coding

    DEFF Research Database (Denmark)

    Forchhammer, Søren; Danieli, Matteo; Burini, Nino

    2010-01-01

    This paper considers estimating and maximizing the entropy of two-dimensional (2-D) fields with application to 2-D constrained coding. We consider Markov random fields (MRF), which have a non-causal description, and the special case of Pickard random fields (PRF). The PRF are 2-D causal finite...... context models, which define stationary probability distributions on finite rectangles and thus allow for calculation of the entropy. We consider two binary constraints and revisit the hard square constraint given by forbidding neighboring 1s and provide novel results for the constraint that no uniform 2...... £ 2 squares contains all 0s or all 1s. The maximum values of the entropy for the constraints are estimated and binary PRF satisfying the constraint are characterized and optimized w.r.t. the entropy. The maximum binary PRF entropy is 0.839 bits/symbol for the no uniform squares constraint. The entropy...

  19. Modeling Vortex Generators in a Navier-Stokes Code

    Science.gov (United States)

    Dudek, Julianne C.

    2011-01-01

    A source-term model that simulates the effects of vortex generators was implemented into the Wind-US Navier-Stokes code. The source term added to the Navier-Stokes equations simulates the lift force that would result from a vane-type vortex generator in the flowfield. The implementation is user-friendly, requiring the user to specify only three quantities for each desired vortex generator: the range of grid points over which the force is to be applied and the planform area and angle of incidence of the physical vane. The model behavior was evaluated for subsonic flow in a rectangular duct with a single vane vortex generator, subsonic flow in an S-duct with 22 corotating vortex generators, and supersonic flow in a rectangular duct with a counter-rotating vortex-generator pair. The model was also used to successfully simulate microramps in supersonic flow by treating each microramp as a pair of vanes with opposite angles of incidence. The validation results indicate that the source-term vortex-generator model provides a useful tool for screening vortex-generator configurations and gives comparable results to solutions computed using gridded vanes.

  20. IPEN-MB-1 critical facility calculation using CONDOR-CITVAP codes

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo Anibal [INVAP SE, San Carlos de Bariloche, Rio Negro (Argentina)

    1997-12-31

    The IPEN-MB-1 critical facility was calculated using the INVAP calculation line CONDOR-CITVAP. The calculated parameters are the reactivity excess of the core, the critical position of the control rods, and the reactivity curves of the control rods. The cell models (CONDOR v1.3) and core models (CITVAP v3.1) are presented. A numerical analysis of the with different calculation parameters is also presented. The analyzed parameters were: model, energy groups and number of spatial meshes. An additional comparison with the MCNP code were performed and it presented in this paper. (author) 16 refs., 3 figs., 12 tabs.

  1. Modeling Vortex Generators in the Wind-US Code

    Science.gov (United States)

    Dudek, Julianne C.

    2010-01-01

    A source term model which simulates the effects of vortex generators was implemented into the Wind-US Navier Stokes code. The source term added to the Navier-Stokes equations simulates the lift force which would result from a vane-type vortex generator in the flowfield. The implementation is user-friendly, requiring the user to specify only three quantities for each desired vortex generator: the range of grid points over which the force is to be applied and the planform area and angle of incidence of the physical vane. The model behavior was evaluated for subsonic flow in a rectangular duct with a single vane vortex generator, supersonic flow in a rectangular duct with a counterrotating vortex generator pair, and subsonic flow in an S-duct with 22 co-rotating vortex generators. The validation results indicate that the source term vortex generator model provides a useful tool for screening vortex generator configurations and gives comparable results to solutions computed using a gridded vane.

  2. Modelling of sprays in containment applications with A CMFD code

    Energy Technology Data Exchange (ETDEWEB)

    Mimouni, S., E-mail: stephane.mimouni@edf.f [Electricite de France R and D Division, 6 Quai Watier, F-78400 Chatou (France); Lamy, J.-S. [Electricite de France R and D Division, 1 av. du General de Gaulle, F-92140 Clamart (France); Lavieville, J. [Electricite de France R and D Division, 6 Quai Watier, F-78400 Chatou (France); Guieu, S.; Martin, M. [Electricite de France SEPTEN Division, 12-14 av. Dutrievoz, 69628 Villeurbanne (France)

    2010-09-15

    During the course of a hypothetical severe accident in a Pressurized Water Reactor (PWR), spray systems are used in the containment in order to prevent overpressure in case of a steam line break, and to enhance the gas mixing in case of the presence of hydrogen. In the frame of the Severe Accident Research Network (SARNET) of the 6th EC Framework Programme, two tests was produced in the TOSQAN facility in order to study the spray behaviour under severe accident conditions: TOSQAN 101 and TOSQAN 113. The TOSQAN facility is a closed cylindrical vessel. The inner spray system is located on the top of the enclosure on the vertical axis. For the TOSQAN 101 case, an initial pressurization in the vessel is performed with superheated steam up to 2.5 bar. Then, steam injection is stopped and spraying starts simultaneously at a given water temperature (around 25 {sup o}C) and water mass flow-rate (around 30 g/s). The depressurization transient starts and continues until the equilibrium phase, which corresponds to the stabilization of the average temperature and pressure of the gaseous mixture inside the vessel. The purpose of the TOSQAN 113 cold spray test is to study helium mixing due to spray activation without heat and mass transfers between gas and droplets. We present in this paper the spray modelling implemented in NEPTUNE{sub C}FD, a three-dimensional multi-fluid code developed especially for nuclear reactor applications. A new model dedicated to the droplet evaporation at the wall is also detailed. Keeping in mind the Best Practice Guidelines, closure laws have been selected to ensure a grid-dependence as weak as possible. For the TOSQAN 113 case, the time evolution of the helium volume fraction calculated shows that the physical approach described in the paper is able to reproduce the mixing of helium by the spray. The prediction of the transient behaviour should be improved by including in the model corrections based on better understanding of the influence of the

  3. 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

  4. Subgrid Combustion Modeling for the Next Generation National Combustion Code

    Science.gov (United States)

    Menon, Suresh; Sankaran, Vaidyanathan; Stone, Christopher

    2003-01-01

    In the first year of this research, a subgrid turbulent mixing and combustion methodology developed earlier at Georgia Tech has been provided to researchers at NASA/GRC for incorporation into the next generation National Combustion Code (called NCCLES hereafter). A key feature of this approach is that scalar mixing and combustion processes are simulated within the LES grid using a stochastic 1D model. The subgrid simulation approach recovers locally molecular diffusion and reaction kinetics exactly without requiring closure and thus, provides an attractive feature to simulate complex, highly turbulent reacting flows of interest. Data acquisition algorithms and statistical analysis strategies and routines to analyze NCCLES results have also been provided to NASA/GRC. The overall goal of this research is to systematically develop and implement LES capability into the current NCC. For this purpose, issues regarding initialization and running LES are also addressed in the collaborative effort. In parallel to this technology transfer effort (that is continuously on going), research has also been underway at Georgia Tech to enhance the LES capability to tackle more complex flows. In particular, subgrid scalar mixing and combustion method has been evaluated in three distinctly different flow field in order to demonstrate its generality: (a) Flame-Turbulence Interactions using premixed combustion, (b) Spatially evolving supersonic mixing layers, and (c) Temporal single and two-phase mixing layers. The configurations chosen are such that they can be implemented in NCCLES and used to evaluate the ability of the new code. Future development and validation will be in spray combustion in gas turbine engine and supersonic scalar mixing.

  5. Fission Matrix Capability for MCNP Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Carney, Sean E. [Los Alamos National Laboratory; Brown, Forrest B. [Los Alamos National Laboratory; Kiedrowski, Brian C. [Los Alamos National Laboratory; Martin, William R. [Los Alamos National Laboratory

    2012-09-05

    In a Monte Carlo criticality calculation, before the tallying of quantities can begin, a converged fission source (the fundamental eigenvector of the fission kernel) is required. Tallies of interest may include powers, absorption rates, leakage rates, or the multiplication factor (the fundamental eigenvalue of the fission kernel, k{sub eff}). Just as in the power iteration method of linear algebra, if the dominance ratio (the ratio of the first and zeroth eigenvalues) is high, many iterations of neutron history simulations are required to isolate the fundamental mode of the problem. Optically large systems have large dominance ratios, and systems containing poor neutron communication between regions are also slow to converge. The fission matrix method, implemented into MCNP[1], addresses these problems. When Monte Carlo random walk from a source is executed, the fission kernel is stochastically applied to the source. Random numbers are used for: distances to collision, reaction types, scattering physics, fission reactions, etc. This method is used because the fission kernel is a complex, 7-dimensional operator that is not explicitly known. Deterministic methods use approximations/discretization in energy, space, and direction to the kernel. Consequently, they are faster. Monte Carlo directly simulates the physics, which necessitates the use of random sampling. Because of this statistical noise, common convergence acceleration methods used in deterministic methods do not work. In the fission matrix method, we are using the random walk information not only to build the next-iteration fission source, but also a spatially-averaged fission kernel. Just like in deterministic methods, this involves approximation and discretization. The approximation is the tallying of the spatially-discretized fission kernel with an incorrect fission source. We address this by making the spatial mesh fine enough that this error is negligible. As a consequence of discretization we get a

  6. Coupling External Radiation Transport Code Results to the GADRAS Detector Response Function

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Dean J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Thoreson, Gregory G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Horne, Steven M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection

    2014-01-01

    Simulating gamma spectra is useful for analyzing special nuclear materials. Gamma spectra are influenced not only by the source and the detector, but also by the external, and potentially complex, scattering environment. The scattering environment can make accurate representations of gamma spectra difficult to obtain. By coupling the Monte Carlo Nuclear Particle (MCNP) code with the Gamma Detector Response and Analysis Software (GADRAS) detector response function, gamma spectrum simulations can be computed with a high degree of fidelity even in the presence of a complex scattering environment. Traditionally, GADRAS represents the external scattering environment with empirically derived scattering parameters. By modeling the external scattering environment in MCNP and using the results as input for the GADRAS detector response function, gamma spectra can be obtained with a high degree of fidelity. This method was verified with experimental data obtained in an environment with a significant amount of scattering material. The experiment used both gamma-emitting sources and moderated and bare neutron-emitting sources. The sources were modeled using GADRAS and MCNP in the presence of the external scattering environment, producing accurate representations of the experimental data.

  7. Coupling External Radiation Transport Code Results to the GADRAS Detector Response Function

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, Dean J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Thoreson, Gregory G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection; Horne, Steven M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Contraband Detection

    2014-01-01

    Simulating gamma spectra is useful for analyzing special nuclear materials. Gamma spectra are influenced not only by the source and the detector, but also by the external, and potentially complex scattering environment. The scattering environment can make accurate representations of gamma spectra difficult to obtain. By coupling the Monte Carlo Nuclear Particle (MCNP) code with the Gamma Detector Response and Analysis Software (GADRAS) detector response function, gamma spectrum simulations can be computed with a high degree of fidelity even in the presence of a complex scattering environment. Traditionally, GADRAS represents the external scattering environment with empirically derived scattering parameters. By modeling the external scattering environment in MCNP and using the results as input for the GADRAS detector response function, gamma spectra can be obtained with a high degree of fidelity. This method was verified with experimental data obtained in an environment with a significant amount of scattering material. The experiment used both gamma-emitting sources and moderated and bare neutron-emitting sources. The sources were modeled using GADRAS and MCNP in the presence of the external scattering environment, producing accurate representations of the experimental data.

  8. Methodology Using MELCOR Code to Model Proposed Hazard Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Gavin Hawkley

    2010-07-01

    This study demonstrates a methodology for using the MELCOR code to model a proposed hazard scenario within a building containing radioactive powder, and the subsequent evaluation of a leak path factor (LPF) (or the amount of respirable material which that escapes a facility into the outside environment), implicit in the scenario. This LPF evaluation will analyzes the basis and applicability of an assumed standard multiplication of 0.5 × 0.5 (in which 0.5 represents the amount of material assumed to leave one area and enter another), for calculating an LPF value. The outside release is dependsent upon the ventilation/filtration system, both filtered and un-filtered, and from other pathways from the building, such as doorways (, both open and closed). This study is presents ed to show how the multiple leak path factorsLPFs from the interior building can be evaluated in a combinatory process in which a total leak path factorLPF is calculated, thus addressing the assumed multiplication, and allowing for the designation and assessment of a respirable source term (ST) for later consequence analysis, in which: the propagation of material released into the environmental atmosphere can be modeled and the dose received by a receptor placed downwind can be estimated and the distance adjusted to maintains such exposures as low as reasonably achievableALARA.. Also, this study will briefly addresses particle characteristics thatwhich affect atmospheric particle dispersion, and compares this dispersion with leak path factorLPF methodology.

  9. Coding conventions and principles for a National Land-Change Modeling Framework

    Science.gov (United States)

    Donato, David I.

    2017-07-14

    This report establishes specific rules for writing computer source code for use with the National Land-Change Modeling Framework (NLCMF). These specific rules consist of conventions and principles for writing code primarily in the C and C++ programming languages. Collectively, these coding conventions and coding principles create an NLCMF programming style. In addition to detailed naming conventions, this report provides general coding conventions and principles intended to facilitate the development of high-performance software implemented with code that is extensible, flexible, and interoperable. Conventions for developing modular code are explained in general terms and also enabled and demonstrated through the appended templates for C++ base source-code and header files. The NLCMF limited-extern approach to module structure, code inclusion, and cross-module access to data is both explained in the text and then illustrated through the module templates. Advice on the use of global variables is provided.

  10. 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)

  11. Error Analysis and Experimental Research of HTR-10 Burn-up Measurement System Based on MCNP Modeling%基于MCNP模型10MW高温气冷实验堆燃耗测量误差分析与实验研究

    Institute of Scientific and Technical Information of China (English)

    马涛; 夏冰; 王军令; 陈晓明; 江二东

    2012-01-01

    10MW高温气冷实验堆(HTR-10)的燃耗测量系统通过测量燃料球内裂变产物137Cs发出的γ射线进而间接确定燃料球的燃耗,测量结果的准确性直接影响着反应堆的安全性和经济性.利用HTR-10现有的设备条件,设计并实施了提升器偏转实验,使燃料球逐步偏离正常测量位,改变球心与准直器轴线的相对位置,得到了偏离角度与计数率之间的对应关系,进而确定燃料球球心与准直器轴线的周向偏移量.通过MCNP程序建立HTR-10燃耗测量系统模型,模拟γ光子从燃料球发出,经过提升器、密封法兰、准直器直到被HPGe晶体探测器捕捉的全过程.利用MCNP模型可以模拟在不同径向偏离情况下的实验过程,通过与实验结果的对比,确定燃料球球心偏离准直器轴线的径向偏移量.%With the burn-up measurement system fur the 10MW high temperature gas-rooletl reuctor. ihe hum-up of the spherical fuel element is obtained indirectly hy measuring the gamma-ray from the fission product 137Cs. The accuracy of results will affect the security and the economy of the reactor. In this paper, on the basis uf the existing equipment conditions of HTR-10, the diversion experiment of the elevator is designed and implemented. By making the spherical fuel element gradually deviate tiie running measurement place, the relative position between the center of the spherical fuel element and the axis of the rnllimalor is changed, and the relationship between the angle of she diversion and the count rale is obtained anri is then used lo determine the offset in the circumferential direction between the center of the spherical fuel element and the collimator axis. The model of the HTR-10 bum-up measurement system, established by the MCNP program, can he used to simulate the gamma photon transporting process from the spherical fuel element, through the elevator, the sealed flange and the collimator, to the HPCe detector. The MCNP model eon be

  12. Modeling ion exchange in clinoptilolite using the EQ3/6 geochemical modeling code

    Energy Technology Data Exchange (ETDEWEB)

    Viani, B.E.; Bruton, C.J.

    1992-06-01

    Assessing the suitability of Yucca Mtn., NV as a potential repository for high-level nuclear waste requires the means to simulate ion-exchange behavior of zeolites. Vanselow and Gapon convention cation-exchange models have been added to geochemical modeling codes EQ3NR/EQ6, allowing exchange to be modeled for up to three exchangers or a single exchanger with three independent sites. Solid-solution models that are numerically equivalent to the ion-exchange models were derived and also implemented in the code. The Gapon model is inconsistent with experimental adsorption isotherms of trace components in clinoptilolite. A one-site Vanselow model can describe adsorption of Cs or Sr on clinoptilolite, but a two-site Vanselow exchange model is necessary to describe K contents of natural clinoptilolites.

  13. Modeling ion exchange in clinoptilolite using the EQ3/6 geochemical modeling code

    Energy Technology Data Exchange (ETDEWEB)

    Viani, B.E.; Bruton, C.J. [Lawrence Livermore National Lab., CA (United States)

    1992-12-31

    Potential disposal of high-level nuclear waste at Yucca Mtn., Nevada requires the means to simulate ion-exchange behavior of clays and zeolites. Vanselow and Gapon convention cation-exchange models have been added to geochemical modeling codes EQ3NR/EQ6, allowing exchange to be modeled for up to three exchangers or a single exchanger with three independent sites. Solid-solution models that are numerically equivalent to the ion-exchange models were derived and also implemented in the code. The Gapon model is inconsistent with experimental adsorption isotherms of trace components in clinoptilolite. A one-site Vanselow model can describe adsorption of Cs and Sr on clinoptilolite, but a two-site Vanselow exchange model is necessary to describe K contents of natural clinoptilolites. 15 refs., 5 figs., 1 tab.

  14. Improved virtual channel noise model for transform domain Wyner-Ziv video coding

    DEFF Research Database (Denmark)

    Huang, Xin; Forchhammer, Søren

    2009-01-01

    Distributed video coding (DVC) has been proposed as a new video coding paradigm to deal with lossy source coding using side information to exploit the statistics at the decoder to reduce computational demands at the encoder. A virtual channel noise model is utilized at the decoder to estimate...

  15. A Mathematical Model Accounting for the Organisation in Multiplets of the Genetic Code

    OpenAIRE

    Sciarrino, A.

    2001-01-01

    Requiring stability of genetic code against translation errors, modelised by suitable mathematical operators in the crystal basis model of the genetic code, the main features of the organisation in multiplets of the mitochondrial and of the standard genetic code are explained.

  16. A realistic model under which the genetic code is optimal.

    Science.gov (United States)

    Buhrman, Harry; van der Gulik, Peter T S; Klau, Gunnar W; Schaffner, Christian; Speijer, Dave; Stougie, Leen

    2013-10-01

    The genetic code has a high level of error robustness. Using values of hydrophobicity scales as a proxy for amino acid character, and the mean square measure as a function quantifying error robustness, a value can be obtained for a genetic code which reflects the error robustness of that code. By comparing this value with a distribution of values belonging to codes generated by random permutations of amino acid assignments, the level of error robustness of a genetic code can be quantified. We present a calculation in which the standard genetic code is shown to be optimal. We obtain this result by (1) using recently updated values of polar requirement as input; (2) fixing seven assignments (Ile, Trp, His, Phe, Tyr, Arg, and Leu) based on aptamer considerations; and (3) using known biosynthetic relations of the 20 amino acids. This last point is reflected in an approach of subdivision (restricting the random reallocation of assignments to amino acid subgroups, the set of 20 being divided in four such subgroups). The three approaches to explain robustness of the code (specific selection for robustness, amino acid-RNA interactions leading to assignments, or a slow growth process of assignment patterns) are reexamined in light of our findings. We offer a comprehensive hypothesis, stressing the importance of biosynthetic relations, with the code evolving from an early stage with just glycine and alanine, via intermediate stages, towards 64 codons carrying todays meaning.

  17. Semantic-preload video model based on VOP coding

    Science.gov (United States)

    Yang, Jianping; Zhang, Jie; Chen, Xiangjun

    2013-03-01

    In recent years, in order to reduce semantic gap which exists between high-level semantics and low-level features of video when the human understanding image or video, people mostly try the method of video annotation where in signal's downstream, namely further (again) attach labels to the content in video-database. Few people focus on the idea that: Use limited interaction and the means of comprehensive segmentation (including optical technologies) from the front-end of collection of video information (i.e. video camera), with video semantics analysis technology and corresponding concepts sets (i.e. ontology) which belong in a certain domain, as well as story shooting script and the task description of scene shooting etc; Apply different-level semantic descriptions to enrich the attributes of video object and the attributes of image region, then forms a new video model which is based on Video Object Plan (VOP) Coding. This model has potential intellectualized features, and carries a large amount of metadata, and embedded intermediate-level semantic concept into every object. This paper focuses on the latter, and presents a framework of a new video model. At present, this new video model is temporarily named "Video Model of Semantic-Preloaded or Semantic-Preload Video Model (simplified into VMoSP or SPVM)". This model mainly researches how to add labeling to video objects and image regions in real time, here video object and image region are usually used intermediate semantic labeling, and this work is placed on signal's upstream (i.e. video capture production stage). Because of the research needs, this paper also tries to analyses the hierarchic structure of video, and divides the hierarchic structure into nine hierarchy semantic levels, of course, this nine hierarchy only involved in video production process. In addition, the paper also point out that here semantic level tagging work (i.e. semantic preloading) only refers to the four middle-level semantic. All in

  18. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  19. ACTI - An MCNP Data Library for Prompt Gamma-Ray Spectroscopy.

    Energy Technology Data Exchange (ETDEWEB)

    Frankle, S. C. (Stephanie C.); Reedy, R. C. (Robert C.); Young, P. G. (Phillip Gaffney),

    2002-01-01

    Prompt gamma-ray spectroscopy is used in a wide variety of applications for determining material compositions. High-quality photon-production data from thermal-neutron capture reactions are essential for these applications. Radiation transport codes, such as MCNP{trademark}, are often used to design detector systems, determine minimum detection thresholds, etc. These transport codes rely on evaluated nuclear databases such as ENDF (Evaluated Nuclear Data File) to provide the fundamental data used in the transport calculations. Often the photon-production data from incident neutron reactions in the evaluations are of relatively poor quality. We have compiled the best experimental data for thermal-neutron capture for the naturally occurring isotopes for elements from H through Zn as well as for {sup 70,72,73,74,76}Ge, {sup 149}Sm, {sup 155,157}Gd, {sup 181}Ta and {sup 182,183,184,186}W. This compilation has been used to update the ENDF evaluations for {sup 1}H, {sup 4}He, {sup 9}Be, {sup 14}N, {sup 16}O, {sup 19}F, Na, Mg, {sup 27}Al, {sup 32}S, S, {sup 35,37}Cl, K, Ca, {sup 45}Sc, Ti, {sup 51}V, {sup 50,52,53,54}Cr, {sup 55}Mn, {sup 54,56,57,58}Fe, {sup 58,60,61,62,64}Ni, {sup 63,65}Cu and {sup 182,183,184,186}W. In addition, the inelastic cross sections and corresponding secondary-photon distributions were updated for {sup 16}O. Complete new evaluations were submitted to ENDF for {sup 35,37}Cl. This paper will discuss the evaluation effort and the production of the MCNP data library, ACTI, based on the new evaluations. Data from the ENDF evaluations for {sup 28-30}Si were also included in the ACTI library for completeness. The silicon evaluations were updated in 1997 and include the latest experimental data for radiative capture.

  20. Continuous energy cross section library for MCNP/MCNPX based on JENDL high energy file 2007; FXJH7

    OpenAIRE

    佐々 敏信; 菅原 隆徳; 小迫 和明; 深堀 智生

    2008-01-01

    The latest JENDL High Energy File (JENDL/HE) was released in 2007 to respond the requirements of reaction data in high energy range up to several GeV to design accelerator facilities such as accelerator-driven systems and research complex like J-PARC. To apply the JENDL/HE-2007 file to the design study, the cross section library of FXJH7 series was constructed from the JENDL/HE file for the calculation using MCNP and MCNPX codes which are widely used in the field of nuclear reactors, fusion r...

  1. On the development of LWR fuel analysis code (1). Analysis of the FEMAXI code and proposal of a new model

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-01-01

    This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)

  2. A Verification of MCNP6 FMESH Tally Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Swift, Alicia L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKigney, Edward A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schirato, Richard C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Robinson, Alex Philip [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Temple, Brian Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-10

    This work serves to verify the MCNP6 FMESH capability through comparison to two types of data. FMESH tallies, binned in time, were generated on an ideal detector face for neutrons undergoing a single scatter in a graphite target. For verification, FMESH results were compared to analytic calculations of the nonrelativistic TOF for elastic and inelastic single neutron scatters (TOF for the purposes of this paper is the time for a neutron to travel from its scatter location in the graphite target to the detector face). FMESH tally results were also compared to F4 tally results, an MNCP tally that calculates fluence in the same way as the FMESH tally. The FMESH tally results agree well with the analytic results and the F4 tally; hence, it is believed that, for simple geometries, MCNP6 FMESH tallies represent the physics of neutron scattering very well.

  3. Djehuty A Code for Modeling Whole Stars in Three Dimensions

    CERN Document Server

    Turcotte, S; Castor, J I; Cavallo, R M; Cohl, H S; Cook, K; Dearborn, D S P; Dossa, D D; Eastman, R; Eggleton, P P; Eltgroth, P; Keller, S; Murray, S; Taylor, A

    2001-01-01

    The DJEHUTY project is an intensive effort at the Lawrence Livermore National Laboratory (LLNL) to produce a general purpose 3-D stellar structure and evolution code to study dynamic processes in whole stars.

  4. Algorthms and Regolith Erosion Models for the Alert Code Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC and Duke University have teamed on this STTR to develop the ALERT (Advanced Lunar Exhaust-Regolith Transport) code which will include new developments in...

  5. Determination of {beta}{sub eff} using MCNP-4C2 and application to the CROCUS and PROTEUS reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollaire, J. [European Organization for Nuclear Research CERN, CH-1211 Geneve 23 (Switzerland); Plaschy, M.; Jatuff, F. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, CH-1015 Lausanne (Switzerland)

    2006-07-01

    A new Monte Carlo method for the determination of {beta}{sub eff} has been recently developed and tested using appropriate models of the experimental reactors CROCUS and PROTEUS. The current paper describes the applied methodology and highlights the resulting improvements compared to the simplest MCNP approach, i.e. the 'prompt method' technique. In addition, the flexibility advantages of the developed method are presented. Specifically, the possibility to obtain the effective delayed neutron fraction {beta}{sub eff} per delayed neutron group, per fissioning nuclide and per reactor region is illustrated. Finally, the MCNP predictions of {beta}{sub eff} are compared to the results of deterministic calculations. (authors)

  6. MIG version 0.0 model interface guidelines: Rules to accelerate installation of numerical models into any compliant parent code

    Energy Technology Data Exchange (ETDEWEB)

    Brannon, R.M.; Wong, M.K.

    1996-08-01

    A set of model interface guidelines, called MIG, is presented as a means by which any compliant numerical material model can be rapidly installed into any parent code without having to modify the model subroutines. Here, {open_quotes}model{close_quotes} usually means a material model such as one that computes stress as a function of strain, though the term may be extended to any numerical operation. {open_quotes}Parent code{close_quotes} means a hydrocode, finite element code, etc. which uses the model and enforces, say, the fundamental laws of motion and thermodynamics. MIG requires the model developer (who creates the model package) to specify model needs in a standardized but flexible way. MIG includes a dictionary of technical terms that allows developers and parent code architects to share a common vocabulary when specifying field variables. For portability, database management is the responsibility of the parent code. Input/output occurs via structured calling arguments. As much model information as possible (such as the lists of required inputs, as well as lists of precharacterized material data and special needs) is supplied by the model developer in an ASCII text file. Every MIG-compliant model also has three required subroutines to check data, to request extra field variables, and to perform model physics. To date, the MIG scheme has proven flexible in beta installations of a simple yield model, plus a more complicated viscodamage yield model, three electromechanical models, and a complicated anisotropic microcrack constitutive model. The MIG yield model has been successfully installed using identical subroutines in three vectorized parent codes and one parallel C++ code, all predicting comparable results. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort, thereby reducing the cost of installing and sharing models in diverse new codes.

  7. Adaptive Partially Hidden Markov Models with Application to Bilevel Image Coding

    DEFF Research Database (Denmark)

    Forchhammer, Søren Otto; Rasmussen, Tage

    1999-01-01

    Adaptive Partially Hidden Markov Models (APHMM) are introduced extending the PHMM models. The new models are applied to lossless coding of bi-level images achieving resluts which are better the JBIG standard.......Adaptive Partially Hidden Markov Models (APHMM) are introduced extending the PHMM models. The new models are applied to lossless coding of bi-level images achieving resluts which are better the JBIG standard....

  8. Biocomputational prediction of non-coding RNAs in model cyanobacteria

    Directory of Open Access Journals (Sweden)

    Ude Susanne

    2009-03-01

    Full Text Available Abstract Background In bacteria, non-coding RNAs (ncRNA are crucial regulators of gene expression, controlling various stress responses, virulence, and motility. Previous work revealed a relatively high number of ncRNAs in some marine cyanobacteria. However, for efficient genetic and biochemical analysis it would be desirable to identify a set of ncRNA candidate genes in model cyanobacteria that are easy to manipulate and for which extended mutant, transcriptomic and proteomic data sets are available. Results Here we have used comparative genome analysis for the biocomputational prediction of ncRNA genes and other sequence/structure-conserved elements in intergenic regions of the three unicellular model cyanobacteria Synechocystis PCC6803, Synechococcus elongatus PCC6301 and Thermosynechococcus elongatus BP1 plus the toxic Microcystis aeruginosa NIES843. The unfiltered numbers of predicted elements in these strains is 383, 168, 168, and 809, respectively, combined into 443 sequence clusters, whereas the numbers of individual elements with high support are 94, 56, 64, and 406, respectively. Removing also transposon-associated repeats, finally 78, 53, 42 and 168 sequences, respectively, are left belonging to 109 different clusters in the data set. Experimental analysis of selected ncRNA candidates in Synechocystis PCC6803 validated new ncRNAs originating from the fabF-hoxH and apcC-prmA intergenic spacers and three highly expressed ncRNAs belonging to the Yfr2 family of ncRNAs. Yfr2a promoter-luxAB fusions confirmed a very strong activity of this promoter and indicated a stimulation of expression if the cultures were exposed to elevated light intensities. Conclusion Comparison to entries in Rfam and experimental testing of selected ncRNA candidates in Synechocystis PCC6803 indicate a high reliability of the current prediction, despite some contamination by the high number of repetitive sequences in some of these species. In particular, we

  9. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  10. Linear-Time Non-Malleable Codes in the Bit-Wise Independent Tampering Model

    DEFF Research Database (Denmark)

    Cramer, Ronald; Damgård, Ivan Bjerre; Döttling, Nico

    Non-malleable codes were introduced by Dziembowski et al. (ICS 2010) as coding schemes that protect a message against tampering attacks. Roughly speaking, a code is non-malleable if decoding an adversarially tampered encoding of a message m produces the original message m or a value m' (eventuall...... non-malleable codes of Agrawal et al. (TCC 2015) and of Cher- aghchi and Guruswami (TCC 2014) and improves the previous result in the bit-wise tampering model: it builds the first non-malleable codes with linear-time complexity and optimal-rate (i.e. rate 1 - o(1))....

  11. General Description of Fission Observables: GEF Model Code

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, K.-H. [CENBG, CNRS/IN2 P3, Chemin du Solarium, B.P. 120, F-33175 Gradignan (France); Jurado, B., E-mail: jurado@cenbg.in2p3.fr [CENBG, CNRS/IN2 P3, Chemin du Solarium, B.P. 120, F-33175 Gradignan (France); Amouroux, C. [CEA, DSM-Saclay (France); Schmitt, C., E-mail: schmitt@ganil.fr [GANIL, Bd. Henri Becquerel, B.P. 55027, F-14076 Caen Cedex 05 (France)

    2016-01-15

    The GEF (“GEneral description of Fission observables”) model code is documented. It describes the observables for spontaneous fission, neutron-induced fission and, more generally, for fission of a compound nucleus from any other entrance channel, with given excitation energy and angular momentum. The GEF model is applicable for a wide range of isotopes from Z = 80 to Z = 112 and beyond, up to excitation energies of about 100 MeV. The results of the GEF model are compared with fission barriers, fission probabilities, fission-fragment mass- and nuclide distributions, isomeric ratios, total kinetic energies, and prompt-neutron and prompt-gamma yields and energy spectra from neutron-induced and spontaneous fission. Derived properties of delayed neutrons and decay heat are also considered. The GEF model is based on a general approach to nuclear fission that explains a great part of the complex appearance of fission observables on the basis of fundamental laws of physics and general properties of microscopic systems and mathematical objects. The topographic theorem is used to estimate the fission-barrier heights from theoretical macroscopic saddle-point and ground-state masses and experimental ground-state masses. Motivated by the theoretically predicted early localisation of nucleonic wave functions in a necked-in shape, the properties of the relevant fragment shells are extracted. These are used to determine the depths and the widths of the fission valleys corresponding to the different fission channels and to describe the fission-fragment distributions and deformations at scission by a statistical approach. A modified composite nuclear-level-density formula is proposed. It respects some features in the superfluid regime that are in accordance with new experimental findings and with theoretical expectations. These are a constant-temperature behaviour that is consistent with a considerably increased heat capacity and an increased pairing condensation energy that is

  12. Modelling and Implementation of Network Coding for Video

    Directory of Open Access Journals (Sweden)

    Can Eyupoglu

    2016-08-01

    Full Text Available In this paper, we investigate Network Coding for Video (NCV which we apply for video streaming over wireless networks. NCV provides a basis for network coding. We use NCV algorithm to increase throughput and video quality. When designing NCV algorithm, we take the deadline as well as the decodability of the video packet at the receiver. In network coding, different flows of video packets are packed into a single packet at intermediate nodes and forwarded to other nodes over wireless networks. There are many problems that occur during transmission on the wireless channel. Network coding plays an important role in dealing with these problems. We observe the benefits of network coding for throughput increase thanks to applying broadcast operations on wireless networks. The aim of this study is to implement NCV algorithm using C programming language which takes the output of the H.264 video codec generating the video packets. In our experiments, we investigated improvements in terms of video quality and throughput at different scenarios.

  13. Simulated evolution applied to study the genetic code optimality using a model of codon reassignments

    Directory of Open Access Journals (Sweden)

    Monteagudo Ángel

    2011-02-01

    Full Text Available Abstract Background As the canonical code is not universal, different theories about its origin and organization have appeared. The optimization or level of adaptation of the canonical genetic code was measured taking into account the harmful consequences resulting from point mutations leading to the replacement of one amino acid for another. There are two basic theories to measure the level of optimization: the statistical approach, which compares the canonical genetic code with many randomly generated alternative ones, and the engineering approach, which compares the canonical code with the best possible alternative. Results Here we used a genetic algorithm to search for better adapted hypothetical codes and as a method to guess the difficulty in finding such alternative codes, allowing to clearly situate the canonical code in the fitness landscape. This novel proposal of the use of evolutionary computing provides a new perspective in the open debate between the use of the statistical approach, which postulates that the genetic code conserves amino acid properties far better than expected from a random code, and the engineering approach, which tends to indicate that the canonical genetic code is still far from optimal. We used two models of hypothetical codes: one that reflects the known examples of codon reassignment and the model most used in the two approaches which reflects the current genetic code translation table. Although the standard code is far from a possible optimum considering both models, when the more realistic model of the codon reassignments was used, the evolutionary algorithm had more difficulty to overcome the efficiency of the canonical genetic code. Conclusions Simulated evolution clearly reveals that the canonical genetic code is far from optimal regarding its optimization. Nevertheless, the efficiency of the canonical code increases when mistranslations are taken into account with the two models, as indicated by the

  14. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  15. An Overview of the Monte Carlo Methods, Codes, & Applications Group

    Energy Technology Data Exchange (ETDEWEB)

    Trahan, Travis John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-30

    This report sketches the work of the Group to deliver first-principle Monte Carlo methods, production quality codes, and radiation transport-based computational and experimental assessments using the codes MCNP and MCATK for such applications as criticality safety, non-proliferation, nuclear energy, nuclear threat reduction and response, radiation detection and measurement, radiation health protection, and stockpile stewardship.

  16. 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)

  17. INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User`s Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications

  18. Code generation by model transformation: a case study in transformation modularity

    NARCIS (Netherlands)

    Hemel, Z.; Kats, L.C.L.; Groenewegen, D.M.; Viser, E.

    2009-01-01

    The realization of model-driven software development requires effective techniques for implementing code generators for domain-specific languages. This paper identifies techniques for improving separation of concerns in the implementation of generators. The core technique is code generation by model

  19. 49 CFR 41.120 - Acceptable model codes.

    Science.gov (United States)

    2010-10-01

    ... Natural Hazards, Federal Emergency Management Agency, 500 C Street, SW., Washington, DC 20472.): (i) The..., published by the Building Officials and Code Administrators, 4051 West Flossmoor Rd., Country Club Hills... Disaster Relief and Emergency Assistance Act (Stafford Act), 42 U.S.C. 5170a, 5170b, 5192, and 5193, or...

  20. M3: An Open Model for Measuring Code Artifacts

    NARCIS (Netherlands)

    Izmaylova, A.; Klint, P.; Shahi, A.; Vinju, J.J.

    2013-01-01

    In the context of the EU FP7 project ``OSSMETER'' we are developing an infra-structure for measuring source code. The goal of OSSMETER is to obtain insight in the quality of open-source projects from all possible perspectives, including product, process and community. This is a "white paper" on M3,

  1. 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)

  2. Los Alamos and Lawrence Livermore National Laboratories Code-to-Code Comparison of Inter Lab Test Problem 1 for Asteroid Impact Hazard Mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Robert P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miller, Paul [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Howley, Kirsten [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferguson, Jim Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gisler, Galen Ross [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Plesko, Catherine Suzanne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Managan, Rob [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Owen, Mike [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wasem, Joseph [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bruck-Syal, Megan [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-15

    The NNSA Laboratories have entered into an interagency collaboration with the National Aeronautics and Space Administration (NASA) to explore strategies for prevention of Earth impacts by asteroids. Assessment of such strategies relies upon use of sophisticated multi-physics simulation codes. This document describes the task of verifying and cross-validating, between Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL), modeling capabilities and methods to be employed as part of the NNSA-NASA collaboration. The approach has been to develop a set of test problems and then to compare and contrast results obtained by use of a suite of codes, including MCNP, RAGE, Mercury, Ares, and Spheral. This document provides a short description of the codes, an overview of the idealized test problems, and discussion of the results for deflection by kinetic impactors and stand-off nuclear explosions.

  3. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  4. A computer code for calculations in the algebraic collective model of the atomic nucleus

    CERN Document Server

    Welsh, T A

    2016-01-01

    A Maple code is presented for algebraic collective model (ACM) calculations. The ACM is an algebraic version of the Bohr model of the atomic nucleus, in which all required matrix elements are derived by exploiting the model's SU(1,1) x SO(5) dynamical group. This, in particular, obviates the use of coefficients of fractional parentage. This paper reviews the mathematical formulation of the ACM, and serves as a manual for the code. The code makes use of expressions for matrix elements derived elsewhere and newly derived matrix elements of the operators [pi x q x pi]_0 and [pi x pi]_{LM}, where q_M are the model's quadrupole moments, and pi_N are corresponding conjugate momenta (-2>=M,N<=2). The code also provides ready access to SO(3)-reduced SO(5) Clebsch-Gordan coefficients through data files provided with the code.

  5. Accelerating scientific codes by performance and accuracy modeling

    CERN Document Server

    Fabregat-Traver, Diego; Bientinesi, Paolo

    2016-01-01

    Scientific software is often driven by multiple parameters that affect both accuracy and performance. Since finding the optimal configuration of these parameters is a highly complex task, it extremely common that the software is used suboptimally. In a typical scenario, accuracy requirements are imposed, and attained through suboptimal performance. In this paper, we present a methodology for the automatic selection of parameters for simulation codes, and a corresponding prototype tool. To be amenable to our methodology, the target code must expose the parameters affecting accuracy and performance, and there must be formulas available for error bounds and computational complexity of the underlying methods. As a case study, we consider the particle-particle particle-mesh method (PPPM) from the LAMMPS suite for molecular dynamics, and use our tool to identify configurations of the input parameters that achieve a given accuracy in the shortest execution time. When compared with the configurations suggested by exp...

  6. A high burnup model developed for the DIONISIO code

    Energy Technology Data Exchange (ETDEWEB)

    Soba, A. [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, A., E-mail: denis@cnea.gov.ar [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Romero, L. [U.A. Reactores Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Villarino, E.; Sardella, F. [Departamento Ingeniería Nuclear, INVAP SE, Comandante Luis Piedra Buena 4950, 8430 San Carlos de Bariloche, Río Negro (Argentina)

    2013-02-15

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO{sub 2} fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in {sup 235}U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  7. A high burnup model developed for the DIONISIO code

    Science.gov (United States)

    Soba, A.; Denis, A.; Romero, L.; Villarino, E.; Sardella, F.

    2013-02-01

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  8. Test code for the assessment and improvement of Reynolds stress models

    Science.gov (United States)

    Rubesin, M. W.; Viegas, J. R.; Vandromme, D.; Minh, H. HA

    1987-01-01

    An existing two-dimensional, compressible flow, Navier-Stokes computer code, containing a full Reynolds stress turbulence model, was adapted for use as a test bed for assessing and improving turbulence models based on turbulence simulation experiments. To date, the results of using the code in comparison with simulated channel flow and over an oscillating flat plate have shown that the turbulence model used in the code needs improvement for these flows. It is also shown that direct simulation of turbulent flows over a range of Reynolds numbers are needed to guide subsequent improvement of turbulence models.

  9. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  10. MCNPX Model/Table Comparison

    CERN Document Server

    Hendricks, J S

    2003-01-01

    MCNPX is a Monte Carlo N-Particle radiation transport code extending the capabilities of MCNP4C. As with MCNP, MCNPX uses nuclear data tables to transport neutrons, photons, and electrons. Unlike MCNP, MCNPX also uses (1) nuclear data tables to transport protons; (2) physics models to transport 30 additional particle types (deuterons, tritons, alphas, pions, muons, etc.); and (3) physics models to transport neutrons and protons when no tabular data are available or when the data are above the energy range (20 to 150 MeV) where the data tables end. MCNPX can mix and match data tables and physics models throughout a problem. For example, MCNPX can model neutron transport in a bismuth germinate (BGO) particle detector by using data tables for bismuth and oxygen and using physics models for germanium. Also, MCNPX can model neutron transport in UO sub 2 , making the best use of physics models and data tables: below 20 MeV, data tables are used; above 150 MeV, physics models are used; between 20 and 150 MeV, data t...

  11. 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.

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

    Science.gov (United States)

    Trindade, A.; Rodrigues, P.; Alves, C.; Chaves, A.; Lopes, M. C.; Oliveira, C.; Peralta, L.

    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.

  13. Modeling Proton- and Light Ion-Induced Reactions at Low Energies in the MARS15 Code

    Energy Technology Data Exchange (ETDEWEB)

    Rakhno, I. L. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Mokhov, N. V. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Gudima, K. K. [National Academy of Sciences, Cisineu (Moldova)

    2015-04-25

    An implementation of both ALICE code and TENDL evaluated nuclear data library in order to describe nuclear reactions induced by low-energy projectiles in the Monte Carlo code MARS15 is presented. Comparisons between results of modeling and experimental data on reaction cross sections and secondary particle distributions are shown.

  14. The Nuremberg Code subverts human health and safety by requiring animal modeling

    OpenAIRE

    Greek Ray; Pippus Annalea; Hansen Lawrence A

    2012-01-01

    Abstract Background The requirement that animals be used in research and testing in order to protect humans was formalized in the Nuremberg Code and subsequent national and international laws, codes, and declarations. Discussion We review the history of these requirements and contrast what was known via science about animal models then with what is known now. We further analyze the predictive...

  15. Implementation of the critical points model in a SFM-FDTD code working in oblique incidence

    Energy Technology Data Exchange (ETDEWEB)

    Hamidi, M; Belkhir, A; Lamrous, O [Laboratoire de Physique et Chimie Quantique, Universite Mouloud Mammeri, Tizi-Ouzou (Algeria); Baida, F I, E-mail: omarlamrous@mail.ummto.dz [Departement d' Optique P.M. Duffieux, Institut FEMTO-ST UMR 6174 CNRS Universite de Franche-Comte, 25030 Besancon Cedex (France)

    2011-06-22

    We describe the implementation of the critical points model in a finite-difference-time-domain code working in oblique incidence and dealing with dispersive media through the split field method. Some tests are presented to validate our code in addition to an application devoted to plasmon resonance of a gold nanoparticles grating.

  16. Addressing Hate Speech and Hate Behaviors in Codes of Conduct: A Model for Public Institutions.

    Science.gov (United States)

    Neiger, Jan Alan; Palmer, Carolyn; Penney, Sophie; Gehring, Donald D.

    1998-01-01

    As part of a larger study, researchers collected campus codes prohibiting hate crimes, which were then reviewed to determine whether the codes presented constitutional problems. Based on this review, the authors develop and present a model policy that is content neutral and does not use language that could be viewed as unconstitutionally vague or…

  17. PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-10-01

    A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.

  18. Subgroup A : nuclear model codes report to the Sixteenth Meeting of the WPEC

    Energy Technology Data Exchange (ETDEWEB)

    Talou, P. (Patrick); Chadwick, M. B. (Mark B.); Dietrich, F. S.; Herman, M.; Kawano, T. (Toshihiko); Konig, A.; Obložinský, P.

    2004-01-01

    The Subgroup A activities focus on the development of nuclear reaction models and codes, used in evaluation work for nuclear reactions from the unresolved energy region up to the pion threshold production limit, and for target nuclides from the low teens and heavier. Much of the efforts are devoted by each participant to the continuing development of their own Institution codes. Progresses in this arena are reported in detail for each code in the present document. EMPIRE-II is of public access. The release of the TALYS code has been announced for the ND2004 Conference in Santa Fe, NM, October 2004. McGNASH is still under development and is not expected to be released in the very near future. In addition, Subgroup A members have demonstrated a growing interest in working on common modeling and codes capabilities, which would significantly reduce the amount of duplicate work, help manage efficiently the growing lines of existing codes, and render codes inter-comparison much easier. A recent and important activity of the Subgroup A has therefore been to develop the framework and the first bricks of the ModLib library, which is constituted of mostly independent pieces of codes written in Fortran 90 (and above) to be used in existing and future nuclear reaction codes. Significant progresses in the development of ModLib have been made during the past year. Several physics modules have been added to the library, and a few more have been planned in detail for the coming year.

  19. A Validated MCNP(X) Cross Section Library based on JEFF 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, W.; Verboomen, B.

    2006-10-15

    ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.

  20. Mathematical model and computer code for the analysis of advanced fast reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Schukin, N.V. (Moscow Engineering Physics Inst. (Russian Federation)); Korsun, A.S. (Moscow Engineering Physics Inst. (Russian Federation)); Vitruk, S.G. (Moscow Engineering Physics Inst. (Russian Federation)); Zimin, V.G. (Moscow Engineering Physics Inst. (Russian Federation)); Romanin, S.D. (Moscow Engineering Physics Inst. (Russian Federation))

    1993-04-01

    Efficient algorithms for mathematical modeling of 3-D neutron kinetics and thermal hydraulics are described. The model and appropriate computer code make it possible to analyze a variety of transient events ranging from normal operational states to catastrophic accident excursions. To verify the code, a number of calculations of different kind of transients was carried out. The results of the calculations show that the model and the computer code could be used for conceptual design of advanced liquid metal reactors. The detailed description of calculations of TOP WS accident is presented. (orig./DG)

  1. Development of thermal hydraulic models for the reliable regulatory auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)

    2004-02-15

    The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.

  2. Estimation and interpretation of k{sub eff} confidence intervals in MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Urbatsch, T.J. [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering; Forster, R.A.; Prael, R.E.; Beckman, R.J. [Los Alamos National Lab., NM (United States)

    1995-07-01

    MCNP has three different, but correlated, estimators for Calculating k{sub eff} in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k{sub eff} estimator, is shown to be the best k{sub eff} estimator available in MCNP for estimating k{sub 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.

  3. A generic method for automatic translation between input models for different versions of simulation codes

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001 (Internal Post Box 360), Potchefstroom 2520 (South Africa); Mulder, Eben J. [School of Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

    2014-05-01

    A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications.

  4. An Advanced simulation Code for Modeling Inductive Output Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Thuc Bui; R. Lawrence Ives

    2012-04-27

    During the Phase I program, CCR completed several major building blocks for a 3D large signal, inductive output tube (IOT) code using modern computer language and programming techniques. These included a 3D, Helmholtz, time-harmonic, field solver with a fully functional graphical user interface (GUI), automeshing and adaptivity. Other building blocks included the improved electrostatic Poisson solver with temporal boundary conditions to provide temporal fields for the time-stepping particle pusher as well as the self electric field caused by time-varying space charge. The magnetostatic field solver was also updated to solve for the self magnetic field caused by time changing current density in the output cavity gap. The goal function to optimize an IOT cavity was also formulated, and the optimization methodologies were investigated.

  5. Modeling of Ionization Physics with the PIC Code OSIRIS

    Energy Technology Data Exchange (ETDEWEB)

    Deng, S.; Tsung, F.; Lee, S.; Lu, W.; Mori, W.B.; Katsouleas, T.; Muggli, P.; Blue, B.E.; Clayton, C.E.; O' Connell, C.; Dodd, E.; Decker, F.J.; Huang, C.; Hogan, M.J.; Hemker, R.; Iverson, R.H.; Joshi, C.; Ren, C.; Raimondi, P.; Wang, S.; Walz, D.; /Southern California U. /UCLA /SLAC

    2005-09-27

    When considering intense particle or laser beams propagating in dense plasma or gas, ionization plays an important role. Impact ionization and tunnel ionization may create new plasma electrons, altering the physics of wakefield accelerators, causing blue shifts in laser spectra, creating and modifying instabilities, etc. Here we describe the addition of an impact ionization package into the 3-D, object-oriented, fully parallel PIC code OSIRIS. We apply the simulation tool to simulate the parameters of the upcoming E164 Plasma Wakefield Accelerator experiment at the Stanford Linear Accelerator Center (SLAC). We find that impact ionization is dominated by the plasma electrons moving in the wake rather than the 30 GeV drive beam electrons. Impact ionization leads to a significant number of trapped electrons accelerated from rest in the wake.

  6. A p-Adic Model of DNA Sequence and Genetic Code

    CERN Document Server

    Dragovich, Branko

    2007-01-01

    Using basic properties of p-adic numbers, we consider a simple new approach to describe main aspects of DNA sequence and genetic code. Central role in our investigation plays an ultrametric p-adic information space which basic elements are nucleotides, codons and genes. We show that a 5-adic model is appropriate for DNA sequence. This 5-adic model, combined with 2-adic distance, is also suitable for genetic code and for a more advanced employment in genomics. We find that genetic code degeneracy is related to the p-adic distance between codons.

  7. Trace-Based Code Generation for Model-Based Testing

    NARCIS (Netherlands)

    Kanstrén, T.; Piel, E.; Gross, H.-G.

    2009-01-01

    Paper Submitted for review at the Eighth International Conference on Generative Programming and Component Engineering. Model-based testing can be a powerful means to generate test cases for the system under test. However, creating a useful model for model-based testing requires expertise in the (fo

  8. Trace-Based Code Generation for Model-Based Testing

    NARCIS (Netherlands)

    Kanstrén, T.; Piel, E.; Gross, H.-G.

    2009-01-01

    Paper Submitted for review at the Eighth International Conference on Generative Programming and Component Engineering. Model-based testing can be a powerful means to generate test cases for the system under test. However, creating a useful model for model-based testing requires expertise in the

  9. FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

    Energy Technology Data Exchange (ETDEWEB)

    Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Shestopalov, A., E-mail: shest@dhtp.kiae.ru [RRC' Kurchatov Institute' , Kurchatov sq, 123182 Moscow (Russian Federation)

    2011-07-15

    Highlights: > We developed a new fuel rod behaviour code named FRED. > Main models and assumptions are described. > The code was checked using the IFA-503.2 tests performed at the Halden reactor. - Abstract: The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.

  10. Numerical modelling of spallation in 2D hydrodynamics codes

    Science.gov (United States)

    Maw, J. R.; Giles, A. R.

    1996-05-01

    A model for spallation based on the void growth model of Johnson has been implemented in 2D Lagrangian and Eulerian hydrocodes. The model has been extended to treat complete separation of material when voids coalesce and to describe the effects of elevated temperatures and melting. The capabilities of the model are illustrated by comparison with data from explosively generated spall experiments. Particular emphasis is placed on the prediction of multiple spall effects in weak, low melting point, materials such as lead. The correlation between the model predictions and observations on the strain rate dependence of spall strength is discussed.

  11. RELAP5/MOD3 code manual. Volume 4, Models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations.

  12. Approaches in highly parameterized inversion - PEST++, a Parameter ESTimation code optimized for large environmental models

    Science.gov (United States)

    Welter, David E.; Doherty, John E.; Hunt, Randall J.; Muffels, Christopher T.; Tonkin, Matthew J.; Schreuder, Willem A.

    2012-01-01

    An object-oriented parameter estimation code was developed to incorporate benefits of object-oriented programming techniques for solving large parameter estimation modeling problems. The code is written in C++ and is a formulation and expansion of the algorithms included in PEST, a widely used parameter estimation code written in Fortran. The new code is called PEST++ and is designed to lower the barriers of entry for users and developers while providing efficient algorithms that can accommodate large, highly parameterized problems. This effort has focused on (1) implementing the most popular features of PEST in a fashion that is easy for novice or experienced modelers to use and (2) creating a software design that is easy to extend; that is, this effort provides a documented object-oriented framework designed from the ground up to be modular and extensible. In addition, all PEST++ source code and its associated libraries, as well as the general run manager source code, have been integrated in the Microsoft Visual Studio® 2010 integrated development environment. The PEST++ code is designed to provide a foundation for an open-source development environment capable of producing robust and efficient parameter estimation tools for the environmental modeling community into the future.

  13. Relativistic modeling capabilities in PERSEUS extended MHD simulation code for HED plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Hamlin, Nathaniel D., E-mail: nh322@cornell.edu [438 Rhodes Hall, Cornell University, Ithaca, NY, 14853 (United States); Seyler, Charles E., E-mail: ces7@cornell.edu [Cornell University, Ithaca, NY, 14853 (United States)

    2014-12-15

    We discuss the incorporation of relativistic modeling capabilities into the PERSEUS extended MHD simulation code for high-energy-density (HED) plasmas, and present the latest hybrid X-pinch simulation results. The use of fully relativistic equations enables the model to remain self-consistent in simulations of such relativistic phenomena as X-pinches and laser-plasma interactions. By suitable formulation of the relativistic generalized Ohm’s law as an evolution equation, we have reduced the recovery of primitive variables, a major technical challenge in relativistic codes, to a straightforward algebraic computation. Our code recovers expected results in the non-relativistic limit, and reveals new physics in the modeling of electron beam acceleration following an X-pinch. Through the use of a relaxation scheme, relativistic PERSEUS is able to handle nine orders of magnitude in density variation, making it the first fluid code, to our knowledge, that can simulate relativistic HED plasmas.

  14. An analytical model for source code distributability verification

    Institute of Scientific and Technical Information of China (English)

    Ayaz ISAZADEH; Jaber KARIMPOUR; Islam ELGEDAWY; Habib IZADKHAH

    2014-01-01

    One way to speed up the execution of sequential programs is to divide them into concurrent segments and execute such segments in a parallel manner over a distributed computing environment. We argue that the execution speedup primarily depends on the concurrency degree between the identified segments as well as communication overhead between the segments. To guar-antee the best speedup, we have to obtain the maximum possible concurrency degree between the identified segments, taking communication overhead into consideration. Existing code distributor and multi-threading approaches do not fulfill such re-quirements;hence, they cannot provide expected distributability gains in advance. To overcome such limitations, we propose a novel approach for verifying the distributability of sequential object-oriented programs. The proposed approach enables users to see the maximum speedup gains before the actual distributability implementations, as it computes an objective function which is used to measure different distribution values from the same program, taking into consideration both remote and sequential calls. Experimental results showed that the proposed approach successfully determines the distributability of different real-life software applications compared with their real-life sequential and distributed implementations.

  15. Modelling the CANMET and Marchwood furnaces using the PCOC code

    Energy Technology Data Exchange (ETDEWEB)

    Stopford, P.J.; Marriott, N. (AEA Decommissioning and Radioactive Waste, Harwell (UK). Theoretical Studies Dept.)

    1990-01-01

    Pulverised coal combustion models are validated by detailed comparison with in-flame measurements of velocity, temperatures and species concentrations on two axisymmetric tunnel furnaces. Nitric oxide formation by the thermal and fuel nitrogen mechanisms is also calculated and compared with experiment. The sensitivity of the predictions to the various aspects of the model and the potential for modelling full-scale, power-generating furnaces are discussed. 18 refs., 13 figs.

  16. 基于MCNP安全作业舱辐射屏蔽特性研究%Performance of Shielding Gamma Radiation Based on MCNP

    Institute of Scientific and Technical Information of China (English)

    邹树梁; 孙毅成; 唐德文; 徐守龙

    2015-01-01

    采用蒙特卡罗软件MCNP5程序建立挖掘机安全作业舱及内部驾驶人员实验模型,研究在辐射源(钴-60)类型不同和源距不同的情况下,不同材料及不同结构的安全作业舱屏蔽γ射线的能力.分析讨论人体各部位剂量率值以评价安全作业舱的屏蔽效果,并依据电离辐射防护与辐射源安全基本标准( GB18871—2002)规定的工作人员的照射水平应低于0.01 mSv/h的剂量限值,确定安全作业舱结构的设计方案.此次研究为挖掘机驾驶舱屏蔽设计提供必要的研究数据和设计依据,使驾驶舱在野外高放射环境下能够有效地屏蔽酌射线,保护驾驶人员的辐射屏蔽安全.%Excavator safety chamber for work and internal driver experimental models are constructed in a Monte Carlo program MCNP5 code, and the different types of radiation source and various of sources have effect on the dose ratios of different parts of human. In different situations, the different materials and different structure of security chamber for work can also affect the ability of shielding gamma rays performance. This paper analyzes the dose ratios of every part of human body to evaluate the shielding gamma rays perform-ance of safety chamber for work. On the basis of ionizing radiation protection and radioac-tive source security basic standards (GB18871—2002),the workers exposed to gamma ra-diation should absorb the dose ratios which is less than 0 . 01 mSv/h and the structure de-sign of chamber must meet the goal to protect the drivers. The study of excavator cockpit shielding design provides essential data for further study and provides the basis of design. Above all,the study makes the cockpit with exposures to the field of high gamma radiation safe and protect the driver from gamma radiation.

  17. Review of release models used in source-term codes

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jongsoon [Department of Nuclear Engineering, Chosen University, Kwangju (Korea, Republic of)

    1999-07-01

    Throughout this reviews, the limitations of current release models are identified and ways of improving them suggested, By incorporation recent experimental results, recommendations for future release modeling activities can be made. All release under review were compared with respect to the following six items: scenario, assumptions, mathematical formulations, solution method, radioactive decay chain considered, and geometry. The following nine models are considered for review: SOTEC and SCCEX (CNWRA), DOE/INTERA, TSPA (SNL), Vault Model (AECL), CCALIBRE (SKI), AREST (PNL), Risk Assessment (EPRI), TOSPAC (SNL). (author)

  18. Comparison of a Coupled Near and Far Wake Model With a Free Wake Vortex Code

    DEFF Research Database (Denmark)

    Pirrung, Georg; Riziotis, Vasilis; Aagaard Madsen, Helge

    2016-01-01

    This paper presents the integration of a near wake model for trailing vorticity, which is based on a prescribed wake lifting line model proposed by Beddoes, with a BEM-based far wake model and a 2D shed vorticity model. The resulting coupled aerodynamics model is validated against lifting surface...... computations performed using a free wake panel code. The focus of the description of the aerodynamics model is on the numerical stability, the computation speed and the accuracy of 5 unsteady simulations. To stabilize the near wake model, it has to be iterated to convergence, using a relaxation factor that has...... induction modeling at slow time scales. Finally, the unsteady airfoil aerodynamics model is extended to provide the unsteady bound circulation for the near wake model and to improve 10 the modeling of the unsteady behavior of cambered airfoils. The model comparison with results from a free wake panel code...

  19. A computer code for calculations in the algebraic collective model of the atomic nucleus

    Science.gov (United States)

    Welsh, T. A.; Rowe, D. J.

    2016-03-01

    A Maple code is presented for algebraic collective model (ACM) calculations. The ACM is an algebraic version of the Bohr model of the atomic nucleus, in which all required matrix elements are derived by exploiting the model's SU(1 , 1) × SO(5) dynamical group. This paper reviews the mathematical formulation of the ACM, and serves as a manual for the code. The code enables a wide range of model Hamiltonians to be analysed. This range includes essentially all Hamiltonians that are rational functions of the model's quadrupole moments qˆM and are at most quadratic in the corresponding conjugate momenta πˆN (- 2 ≤ M , N ≤ 2). The code makes use of expressions for matrix elements derived elsewhere and newly derived matrix elements of the operators [ π ˆ ⊗ q ˆ ⊗ π ˆ ] 0 and [ π ˆ ⊗ π ˆ ] LM. The code is made efficient by use of an analytical expression for the needed SO(5)-reduced matrix elements, and use of SO(5) ⊃ SO(3) Clebsch-Gordan coefficients obtained from precomputed data files provided with the code.

  20. Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

    Energy Technology Data Exchange (ETDEWEB)

    Blyth, Taylor S. [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)

    2017-04-01

    The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

  1. Dense Coding in a Two-Spin Squeezing Model with Intrinsic Decoherence

    Science.gov (United States)

    Zhang, Bing-Bing; Yang, Guo-Hui

    2016-11-01

    Quantum dense coding in a two-spin squeezing model under intrinsic decoherence with different initial states (Werner state and Bell state) is investigated. It shows that dense coding capacity χ oscillates with time and finally reaches different stable values. χ can be enhanced by decreasing the magnetic field Ω and the intrinsic decoherence γ or increasing the squeezing interaction μ, moreover, one can obtain a valid dense coding capacity ( χ satisfies χ > 1) by modulating these parameters. The stable value of χ reveals that the decoherence cannot entirely destroy the dense coding capacity. In addition, decreasing Ω or increasing μ can not only enhance the stable value of χ but also impair the effects of decoherence. As the initial state is the Werner state, the purity r of initial state plays a key role in adjusting the value of dense coding capacity, χ can be significantly increased by improving the purity of initial state. For the initial state is Bell state, the large spin squeezing interaction compared with the magnetic field guarantees the optimal dense coding. One cannot always achieve a valid dense coding capacity for the Werner state, while for the Bell state, the dense coding capacity χ remains stuck at the range of greater than 1.

  2. 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.

  3. Stimulus-dependent maximum entropy models of neural population codes.

    Science.gov (United States)

    Granot-Atedgi, Einat; Tkačik, Gašper; Segev, Ronen; Schneidman, Elad

    2013-01-01

    Neural populations encode information about their stimulus in a collective fashion, by joint activity patterns of spiking and silence. A full account of this mapping from stimulus to neural activity is given by the conditional probability distribution over neural codewords given the sensory input. For large populations, direct sampling of these distributions is impossible, and so we must rely on constructing appropriate models. We show here that in a population of 100 retinal ganglion cells in the salamander retina responding to temporal white-noise stimuli, dependencies between cells play an important encoding role. We introduce the stimulus-dependent maximum entropy (SDME) model-a minimal extension of the canonical linear-nonlinear model of a single neuron, to a pairwise-coupled neural population. We find that the SDME model gives a more accurate account of single cell responses and in particular significantly outperforms uncoupled models in reproducing the distributions of population codewords emitted in response to a stimulus. We show how the SDME model, in conjunction with static maximum entropy models of population vocabulary, can be used to estimate information-theoretic quantities like average surprise and information transmission in a neural population.

  4. Stimulus-dependent maximum entropy models of neural population codes.

    Directory of Open Access Journals (Sweden)

    Einat Granot-Atedgi

    Full Text Available Neural populations encode information about their stimulus in a collective fashion, by joint activity patterns of spiking and silence. A full account of this mapping from stimulus to neural activity is given by the conditional probability distribution over neural codewords given the sensory input. For large populations, direct sampling of these distributions is impossible, and so we must rely on constructing appropriate models. We show here that in a population of 100 retinal ganglion cells in the salamander retina responding to temporal white-noise stimuli, dependencies between cells play an important encoding role. We introduce the stimulus-dependent maximum entropy (SDME model-a minimal extension of the canonical linear-nonlinear model of a single neuron, to a pairwise-coupled neural population. We find that the SDME model gives a more accurate account of single cell responses and in particular significantly outperforms uncoupled models in reproducing the distributions of population codewords emitted in response to a stimulus. We show how the SDME model, in conjunction with static maximum entropy models of population vocabulary, can be used to estimate information-theoretic quantities like average surprise and information transmission in a neural population.

  5. Code modernization and modularization of APEX and SWAT watershed simulation models

    Science.gov (United States)

    SWAT (Soil and Water Assessment Tool) and APEX (Agricultural Policy / Environmental eXtender) are respectively large and small watershed simulation models derived from EPIC Environmental Policy Integrated Climate), a field-scale agroecology simulation model. All three models are coded in FORTRAN an...

  6. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-03-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  7. The aeroelastic code HawC - model and comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Thirstrup Petersen, J. [Risoe National Lab., The Test Station for Wind Turbines, Roskilde (Denmark)

    1996-09-01

    A general aeroelastic finite element model for simulation of the dynamic response of horizontal axis wind turbines is presented. The model has been developed with the aim to establish an effective research tool, which can support the general investigation of wind turbine dynamics and research in specific areas of wind turbine modelling. The model concentrates on the correct representation of the inertia forces in a form, which makes it possible to recognize and isolate effects originating from specific degrees of freedom. The turbine structure is divided into substructures, and nonlinear kinematic terms are retained in the equations of motion. Moderate geometric nonlinearities are allowed for. Gravity and a full wind field including 3-dimensional 3-component turbulence are included in the loading. Simulation results for a typical three bladed, stall regulated wind turbine are presented and compared with measurements. (au)

  8. Evaluation of Computational Codes for Underwater Hull Analysis Model Applications

    Science.gov (United States)

    2014-02-05

    vice versa or scaling the model to be physical scale model ( PSM ) size instead of full size. Another common geometric change is translating the...anodes and cathodes are generally the boundary conditions to the solution. The ability to link anodes to reference cells to mimic PSM testing and full...could be developed that allows the user to mimic how anode values are set on shipboard systems. Since much PSM experimental work uses shipboard system

  9. A smooth particle hydrodynamics code to model collisions between solid, self-gravitating objects

    Science.gov (United States)

    Schäfer, C.; Riecker, S.; Maindl, T. I.; Speith, R.; Scherrer, S.; Kley, W.

    2016-05-01

    Context. Modern graphics processing units (GPUs) lead to a major increase in the performance of the computation of astrophysical simulations. Owing to the different nature of GPU architecture compared to traditional central processing units (CPUs) such as x86 architecture, existing numerical codes cannot be easily migrated to run on GPU. Here, we present a new implementation of the numerical method smooth particle hydrodynamics (SPH) using CUDA and the first astrophysical application of the new code: the collision between Ceres-sized objects. Aims: The new code allows for a tremendous increase in speed of astrophysical simulations with SPH and self-gravity at low costs for new hardware. Methods: We have implemented the SPH equations to model gas, liquids and elastic, and plastic solid bodies and added a fragmentation model for brittle materials. Self-gravity may be optionally included in the simulations and is treated by the use of a Barnes-Hut tree. Results: We find an impressive performance gain using NVIDIA consumer devices compared to our existing OpenMP code. The new code is freely available to the community upon request. If you are interested in our CUDA SPH code miluphCUDA, please write an email to Christoph Schäfer. miluphCUDA is the CUDA port of miluph. miluph is pronounced [maßl2v]. We do not support the use of the code for military purposes.

  10. Once-through CANDU reactor models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  11. New higher-order Godunov code for modelling performance of two-stage light gas guns

    Science.gov (United States)

    Bogdanoff, D. W.; Miller, R. J.

    1995-01-01

    A new quasi-one-dimensional Godunov code for modeling two-stage light gas guns is described. The code is third-order accurate in space and second-order accurate in time. A very accurate Riemann solver is used. Friction and heat transfer to the tube wall for gases and dense media are modeled and a simple nonequilibrium turbulence model is used for gas flows. The code also models gunpowder burn in the first-stage breech. Realistic equations of state (EOS) are used for all media. The code was validated against exact solutions of Riemann's shock-tube problem, impact of dense media slabs at velocities up to 20 km/sec, flow through a supersonic convergent-divergent nozzle and burning of gunpowder in a closed bomb. Excellent validation results were obtained. The code was then used to predict the performance of two light gas guns (1.5 in. and 0.28 in.) in service at the Ames Research Center. The code predictions were compared with measured pressure histories in the powder chamber and pump tube and with measured piston and projectile velocities. Very good agreement between computational fluid dynamics (CFD) predictions and measurements was obtained. Actual powder-burn rates in the gun were found to be considerably higher (60-90 percent) than predicted by the manufacturer and the behavior of the piston upon yielding appears to differ greatly from that suggested by low-strain rate tests.

  12. On models of the genetic code generated by binary dichotomic algorithms.

    Science.gov (United States)

    Gumbel, Markus; Fimmel, Elena; Danielli, Alberto; Strüngmann, Lutz

    2015-02-01

    In this paper we introduce the concept of a BDA-generated model of the genetic code which is based on binary dichotomic algorithms (BDAs). A BDA-generated model is based on binary dichotomic algorithms (BDAs). Such a BDA partitions the set of 64 codons into two disjoint classes of size 32 each and provides a generalization of known partitions like the Rumer dichotomy. We investigate what partitions can be generated when a set of different BDAs is applied sequentially to the set of codons. The search revealed that these models are able to generate code tables with very different numbers of classes ranging from 2 to 64. We have analyzed whether there are models that map the codons to their amino acids. A perfect matching is not possible. However, we present models that describe the standard genetic code with only few errors. There are also models that map all 64 codons uniquely to 64 classes showing that BDAs can be used to identify codons precisely. This could serve as a basis for further mathematical analysis using coding theory, for example. The hypothesis that BDAs might reflect a molecular mechanism taking place in the decoding center of the ribosome is discussed. The scan demonstrated that binary dichotomic partitions are able to model different aspects of the genetic code very well. The search was performed with our tool Beady-A. This software is freely available at http://mi.informatik.hs-mannheim.de/beady-a. It requires a JVM version 6 or higher.

  13. Representing Resources in Petri Net Models: Hardwiring or Soft-coding?

    OpenAIRE

    2011-01-01

    This paper presents an interesting design problem in developing a new tool for discrete-event dynamic systems (DEDS). A new tool known as GPenSIM was developed for modeling and simulation of DEDS; GPenSIM is based on Petri Nets. The design issue this paper talks about is whether to represent resources in DEDS hardwired as a part of the Petri net structure (which is the widespread practice) or to soft code as common variables in the program code. This paper shows that soft coding resources giv...

  14. Implementing the WebSocket Protocol Based on Formal Modelling and Automated Code Generation

    DEFF Research Database (Denmark)

    Simonsen, Kent Inge; Kristensen, Lars Michael

    2014-01-01

    protocols. Furthermore, we perform formal verification of the CPN model prior to code generation, and test the implementation for interoperability against the Autobahn WebSocket test-suite resulting in 97% and 99% success rate for the client and server implementation, respectively. The tests show...... with pragmatic annotations for automated code generation of protocol software. The contribution of this paper is an application of the approach as implemented in the PetriCode tool to obtain protocol software implementing the IETF WebSocket protocol. This demonstrates the scalability of our approach to real...

  15. ABAREX -- A neutron spherical optical-statistical-model code -- A user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A.B. [ed.; Lawson, R.D.

    1998-06-01

    The contemporary version of the neutron spherical optical-statistical-model code ABAREX is summarized with the objective of providing detailed operational guidance for the user. The physical concepts involved are very briefly outlined. The code is described in some detail and a number of explicit examples are given. With this document one should very quickly become fluent with the use of ABAREX. While the code has operated on a number of computing systems, this version is specifically tailored for the VAX/VMS work station and/or the IBM-compatible personal computer.

  16. User's manual for MODCAL: Bounding surface soil plasticity model calibration and prediction code, volume 2

    Science.gov (United States)

    Dennatale, J. S.; Herrmann, L. R.; Defalias, Y. F.

    1983-02-01

    In order to reduce the complexity of the model calibration process, a computer-aided automated procedure has been developed and tested. The computer code employs a Quasi-Newton optimization strategy to locate that set of parameter values which minimizes the discrepancy between the model predictions and the experimental observations included in the calibration data base. Through application to a number of real soils, the automated procedure has been found to be an efficient, reliable and economical means of accomplishing model calibration. Although the code was developed specifically for use with the Bounding Surface plasticity model, it can readily be adapted to other constitutive formulations. Since the code greatly reduces the dependence of calibration success on user expertise, it significantly increases the accessibility and usefulness of sophisticated material models to the general engineering community.

  17. Validation of vortex code viscous models using lidar wake measurements and CFD

    DEFF Research Database (Denmark)

    Branlard, Emmanuel; Machefaux, Ewan; Gaunaa, Mac;

    2014-01-01

    The newly implemented vortex code Omnivor coupled to the aero-servo-elastic tool hawc2 is described in this paper. Vortex wake improvements by the implementation of viscous effects are considered. Different viscous models are implemented and compared with each other. Turbulent flow fields...... with sheared inflow are used to compare the vortex code performance with CFD and lidar measurements. Laminar CFD computations are used to evaluate the performance of the viscous models. Consistent results between the vortex code and CFD tool are obtained up to three diameters downstream. The modelling...... of viscous boundaries appear more important than the modelling of viscosity in the wake. External turbulence and shear appear sufficient but their full potential flow modelling would be preferred....

  18. Improvement of Interfacial Heat Transfer Model and Correlations in SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sung Won; Kim, Kyung Du [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The SPACE code development project has been successfully proceeded since 2006. The first stage of development program has been finished at April 2010. During the first stage, main logic and conceptual structure have been established under the support of Korea Ministry of Knowledge and Economy. In the second stage, it is focused to assess the physical models and correlations of SPACE code by using the well known SET problems. A problem selection process has been performed under the leading of KEPRI. KEPRI has listed suitable SET problems according to the individual assessment purpose. Among the SET problems, the MIT pressurizer test reveals a improper results by using SPACE code. This paper introduce the problem found during the assessment of MIT pressurizer test assessment and the resolving process about the interfacial heat transfer model and correlations in SPACE code

  19. User manual for ATILA, a finite-element code for modeling piezoelectric transducers

    Science.gov (United States)

    Decarpigny, Jean-Noel; Debus, Jean-Claude

    1987-09-01

    This manual for the user of the finite-element code ATILA provides instruction for entering information and running the code on a VAX computer. The manual does not include the code. The finite element code ATILA has been specifically developed to aid the design of piezoelectric devices, mainly for sonar applications. Thus, it is able to perform the model analyses of both axisymmetrical and fully three-dimensional piezoelectric transducers. It can also provide their harmonic response under radiating conditions: nearfield and farfield pressure, transmitting voltage response, directivity pattern, electrical impedance, as well as displacement field, nodal plane positions, stress field and various stress criteria...Its accuracy and its ability to describe the physical behavior of various transducers (Tonpilz transducers, double headmass symmetrical length expanders, free flooded rings, flextensional transducers, bender bars, cylindrical and trilaminar hydrophones...) have been checked by modelling more than twenty different structures and comparing numerical and experimental results.

  20. A model of a code of ethics for tissue banks operating in developing countries.

    Science.gov (United States)

    Morales Pedraza, Jorge

    2012-12-01

    Ethical practice in the field of tissue banking requires the setting of principles, the identification of possible deviations and the establishment of mechanisms that will detect and hinder abuses that may occur during the procurement, processing and distribution of tissues for transplantation. This model of a Code of Ethics has been prepared with the purpose of being used for the elaboration of a Code of Ethics for tissue banks operating in the Latin American and the Caribbean, Asia and the Pacific and the African regions in order to guide the day-to-day operation of these banks. The purpose of this model of Code of Ethics is to assist interested tissue banks in the preparation of their own Code of Ethics towards ensuring that the tissue bank staff support with their actions the mission and values associated with tissue banking.

  1. Description of codes and models to be used in risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    1991-09-01

    Human health and environmental risk assessments will be performed as part of the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) remedial investigation/feasibility study (RI/FS) activities at the Hanford Site. Analytical and computer encoded numerical models are commonly used during both the remedial investigation (RI) and feasibility study (FS) to predict or estimate the concentration of contaminants at the point of exposure to humans and/or the environment. This document has been prepared to identify the computer codes that will be used in support of RI/FS human health and environmental risk assessments at the Hanford Site. In addition to the CERCLA RI/FS process, it is recommended that these computer codes be used when fate and transport analyses is required for other activities. Additional computer codes may be used for other purposes (e.g., design of tracer tests, location of observation wells, etc.). This document provides guidance for unit managers in charge of RI/FS activities. Use of the same computer codes for all analytical activities at the Hanford Site will promote consistency, reduce the effort required to develop, validate, and implement models to simulate Hanford Site conditions, and expedite regulatory review. The discussion provides a description of how models will likely be developed and utilized at the Hanford Site. It is intended to summarize previous environmental-related modeling at the Hanford Site and provide background for future model development. The modeling capabilities that are desirable for the Hanford Site and the codes that were evaluated. The recommendations include the codes proposed to support future risk assessment modeling at the Hanford Site, and provides the rational for the codes selected. 27 refs., 3 figs., 1 tab.

  2. PWR hot leg natural circulation modeling with MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1997-12-31

    Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (Author)

  3. Rapid MCNP simulation of DNA double strand break (DSB) relative biological effectiveness (RBE) for photons, neutrons, and light ions.

    Science.gov (United States)

    Stewart, Robert D; Streitmatter, Seth W; Argento, David C; Kirkby, Charles; Goorley, John T; Moffitt, Greg; Jevremovic, Tatjana; Sandison, George A

    2015-11-07

    To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient model to accurately predict particle relative biological effectiveness (RBE) in cell cultures and in vivo. To illustrate the approach and highlight the impact of the larger scale physical environment (e.g. establishing charged particle equilibrium), we examined the RBE for DSB induction (RBEDSB) of x-rays, (137)Cs γ-rays, neutrons and light ions relative to γ-rays from (60)Co in monolayer cell cultures at various depths in water. Under normoxic conditions, we found that (137)Cs γ-rays are about 1.7% more effective at creating DSB than γ-rays from (60)Co (RBEDSB  =  1.017) whereas 60-250 kV x-rays are 1.1 to 1.25 times more efficient at creating DSB than (60)Co. Under anoxic conditions, kV x-rays may have an RBEDSB up to 1.51 times as large as (60)Co γ-rays. Fission neutrons passing through monolayer cell cultures have an RBEDSB that ranges from 2.6 to 3.0 in normoxic cells, but may be as large as 9.93 for anoxic cells. For proton pencil beams, Monte Carlo simulations suggest an RBEDSB of about 1.2 at the tip of the Bragg peak and up to 1.6 a few mm beyond the Bragg peak. Bragg peak RBEDSB increases with decreasing oxygen concentration, which may create opportunities to apply proton dose painting to help address tumor hypoxia. Modeling of the particle RBE for DSB induction across multiple physical and biological scales has the potential to aid in the interpretation of laboratory experiments and provide useful information to advance the safety and effectiveness of hadron therapy in the treatment of cancer.

  4. Rapid MCNP simulation of DNA double strand break (DSB) relative biological effectiveness (RBE) for photons, neutrons, and light ions

    Science.gov (United States)

    Stewart, Robert D.; Streitmatter, Seth W.; Argento, David C.; Kirkby, Charles; Goorley, John T.; Moffitt, Greg; Jevremovic, Tatjana; Sandison, George A.

    2015-11-01

    To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient model to accurately predict particle relative biological effectiveness (RBE) in cell cultures and in vivo. To illustrate the approach and highlight the impact of the larger scale physical environment (e.g. establishing charged particle equilibrium), we examined the RBE for DSB induction (RBEDSB) of x-rays, 137Cs γ-rays, neutrons and light ions relative to γ-rays from 60Co in monolayer cell cultures at various depths in water. Under normoxic conditions, we found that 137Cs γ-rays are about 1.7% more effective at creating DSB than γ-rays from 60Co (RBEDSB  =  1.017) whereas 60-250 kV x-rays are 1.1 to 1.25 times more efficient at creating DSB than 60Co. Under anoxic conditions, kV x-rays may have an RBEDSB up to 1.51 times as large as 60Co γ-rays. Fission neutrons passing through monolayer cell cultures have an RBEDSB that ranges from 2.6 to 3.0 in normoxic cells, but may be as large as 9.93 for anoxic cells. For proton pencil beams, Monte Carlo simulations suggest an RBEDSB of about 1.2 at the tip of the Bragg peak and up to 1.6 a few mm beyond the Bragg peak. Bragg peak RBEDSB increases with decreasing oxygen concentration, which may create opportunities to apply proton dose painting to help address tumor hypoxia. Modeling of the particle RBE for DSB induction across multiple physical and biological scales has the potential to aid in the interpretation of laboratory experiments and provide useful information to advance the safety and effectiveness of hadron therapy in the treatment of cancer.

  5. Priliminary Modeling of Air Breakdown with the ICEPIC code

    CERN Document Server

    Schulz, A E; Cartwright, K L; Mardahl, P J; Peterkin, R E; Bruner, N; Genoni, T; Hughes, T P; Welch, D

    2004-01-01

    Interest in air breakdown phenomena has recently been re-kindled with the advent of advanced virtual prototyping of radio frequency (RF) sources for use in high power microwave (HPM) weapons technology. Air breakdown phenomena are of interest because the formation of a plasma layer at the aperture of an RF source decreases the transmitted power to the target, and in some cases can cause significant reflection of RF radiation. Understanding the mechanisms behind the formation of such plasma layers will aid in the development of maximally effective sources. This paper begins with some of the basic theory behind air breakdown, and describes two independent approaches to modeling the formation of plasmas, the dielectric fluid model and the Particle in Cell (PIC) approach. Finally we present the results of preliminary studies in numerical modeling and simulation of breakdown.

  6. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  7. 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)

  8. A Dual Coding Theoretical Model of Decoding in Reading: Subsuming the LaBerge and Samuels Model

    Science.gov (United States)

    Sadoski, Mark; McTigue, Erin M.; Paivio, Allan

    2012-01-01

    In this article we present a detailed Dual Coding Theory (DCT) model of decoding. The DCT model reinterprets and subsumes The LaBerge and Samuels (1974) model of the reading process which has served well to account for decoding behaviors and the processes that underlie them. However, the LaBerge and Samuels model has had little to say about…

  9. The non-power model of the genetic code: a paradigm for interpreting genomic information.

    Science.gov (United States)

    Gonzalez, Diego Luis; Giannerini, Simone; Rosa, Rodolfo

    2016-03-13

    In this article, we present a mathematical framework based on redundant (non-power) representations of integer numbers as a paradigm for the interpretation of genomic information. The core of the approach relies on modelling the degeneracy of the genetic code. The model allows one to explain many features and symmetries of the genetic code and to uncover hidden symmetries. Also, it provides us with new tools for the analysis of genomic sequences. We review briefly three main areas: (i) the Euplotid nuclear code, (ii) the vertebrate mitochondrial code, and (iii) the main coding/decoding strategies used in the three domains of life. In every case, we show how the non-power model is a natural unified framework for describing degeneracy and deriving sound biological hypotheses on protein coding. The approach is rooted on number theory and group theory; nevertheless, we have kept the technical level to a minimum by focusing on key concepts and on the biological implications. © 2016 The Author(s).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  11. Domain-specific modeling enabling full code generation

    CERN Document Server

    Kelly, Steven

    2007-01-01

    Domain-Specific Modeling (DSM) is the latest approach tosoftware development, promising to greatly increase the speed andease of software creation. Early adopters of DSM have been enjoyingproductivity increases of 500–1000% in production for over adecade. This book introduces DSM and offers examples from variousfields to illustrate to experienced developers how DSM can improvesoftware development in their teams. Two authorities in the field explain what DSM is, why it works,and how to successfully create and use a DSM solution to improveproductivity and quality. Divided into four parts, the book covers:background and motivation; fundamentals; in-depth examples; andcreating DSM solutions. There is an emphasis throughout the book onpractical guidelines for implementing DSM, including how toidentify the nece sary language constructs, how to generate fullcode from models, and how to provide tool support for a new DSMlanguage. The example cases described in the book are available thebook's Website, www.dsmbook....

  12. A Review of Equation of State Models, Chemical Equilibrium Calculations and CERV Code Requirements for SHS Detonation Modelling

    Science.gov (United States)

    2009-10-01

    Beattie - Bridgeman Virial expansion The above equations are suitable for moderate pressures and are usually based on either empirical constants...CR 2010-013 October 2009 A Review of Equation of State Models, Chemical Equilibrium Calculations and CERV Code Requirements for SHS Detonation...Defence R&D Canada. A Review of Equation of State Models, Chemical Equilibrium Calculations and CERV Code Requirements for SHS Detonation

  13. Model study of the thermal storage system by FEHM code

    Energy Technology Data Exchange (ETDEWEB)

    Tenma, N.; Yasukawa, Kasumi [National Institute of Advanced Industrial Science and Technology, Ibaraki (Japan); Zyvoloski, G. [Los Alamos National Laboratory, Los Alamos, NM (United States). Earth and Environmental Science Division

    2003-12-01

    The use of low-temperature geothermal resources is important from the viewpoint of global warming. In order to evaluate various underground projects that use low-temperature geothermal resources, we have estimated the parameters of a typical underground system using the two-well model. By changing the parameters of the system, six different heat extraction scenarios have been studied. One of these six scenarios is recommended because of its small energy loss. (author)

  14. Model based code generation for distributed embedded systems

    OpenAIRE

    Raghav, Gopal; Gopalswamy, Swaminathan; Radhakrishnan, Karthikeyan; Hugues, Jérôme; Delange, Julien

    2010-01-01

    Embedded systems are becoming increasingly complex and more distributed. Cost and quality requirements necessitate reuse of the functional software components for multiple deployment architectures. An important step is the allocation of software components to hardware. During this process the differences between the hardware and application software architectures must be reconciled. In this paper we discuss an architecture driven approach involving model-based techniques to resolve these diff...

  15. Vectorization, parallelization and porting of nuclear codes (porting). Progress report fiscal 1999

    Energy Technology Data Exchange (ETDEWEB)

    Kawasaki, Nobuo; Nemoto, Toshiyuki; Kawai, Wataru; Ishizuki, Shigeru [Fujitsu Ltd., Tokyo (Japan); Ogasawara, Shinobu; Kume, Etsuo; Adachi, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yatake, Yo-ichi [Hitachi Ltd., Tokyo (Japan)

    2001-01-01

    Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system, the AP3000 system, the SX-4 system and the Paragon system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 18 codes in fiscal 1999. These results are reported in 3 parts, i.e., the vectorization and the parallelization part on vector processors, the parallelization port on scalar processors and the porting part. In this report, we describe the porting. In this porting part, the porting of Assisted Model Building with Energy Refinement code version 5 (AMBER5), general purpose Monte Carlo codes far neutron and photon transport calculations based on continuous energy and multigroup methods (MVP/GMVP), automatic editing system for MCNP library code (autonj), neutron damage calculations for materials irradiations and neutron damage calculations for compounds code (SPECTER/SPECOMP), severe accident analysis code (MELCOR) and COolant Boiling in Rod Arrays, Two-Fluid code (COBRA-TF) on the VPP500 system and/or the AP3000 system are described. (author)

  16. Pre-engineering Spaceflight Validation of Environmental Models and the 2005 HZETRN Simulation Code

    Science.gov (United States)

    Nealy, John E.; Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.; Dachev, Ts. P.; Tomov, B. T.; Walker, Steven A.; DeAngelis, Giovanni; Blattnig, Steve R.; Atwell, William

    2006-01-01

    The HZETRN code has been identified by NASA for engineering design in the next phase of space exploration highlighting a return to the Moon in preparation for a Mars mission. In response, a new series of algorithms beginning with 2005 HZETRN, will be issued by correcting some prior limitations and improving control of propagated errors along with established code verification processes. Code validation processes will use new/improved low Earth orbit (LEO) environmental models with a recently improved International Space Station (ISS) shield model to validate computational models and procedures using measured data aboard ISS. These validated models will provide a basis for flight-testing the designs of future space vehicles and systems of the Constellation program in the LEO environment.

  17. Reduced Fast Ion Transport Model For The Tokamak Transport Code TRANSP

    Energy Technology Data Exchange (ETDEWEB)

    Podesta,, Mario; Gorelenkova, Marina; White, Roscoe

    2014-02-28

    Fast ion transport models presently implemented in the tokamak transport code TRANSP [R. J. Hawryluk, in Physics of Plasmas Close to Thermonuclear Conditions, CEC Brussels, 1 , 19 (1980)] are not capturing important aspects of the physics associated with resonant transport caused by instabilities such as Toroidal Alfv en Eigenmodes (TAEs). This work describes the implementation of a fast ion transport model consistent with the basic mechanisms of resonant mode-particle interaction. The model is formulated in terms of a probability distribution function for the particle's steps in phase space, which is consistent with the MonteCarlo approach used in TRANSP. The proposed model is based on the analysis of fast ion response to TAE modes through the ORBIT code [R. B. White et al., Phys. Fluids 27 , 2455 (1984)], but it can be generalized to higher frequency modes (e.g. Compressional and Global Alfv en Eigenmodes) and to other numerical codes or theories.

  18. The Nuremberg Code subverts human health and safety by requiring animal modeling

    Science.gov (United States)

    2012-01-01

    Background The requirement that animals be used in research and testing in order to protect humans was formalized in the Nuremberg Code and subsequent national and international laws, codes, and declarations. Discussion We review the history of these requirements and contrast what was known via science about animal models then with what is known now. We further analyze the predictive value of animal models when used as test subjects for human response to drugs and disease. We explore the use of animals for models in toxicity testing as an example of the problem with using animal models. Summary We conclude that the requirements for animal testing found in the Nuremberg Code were based on scientifically outdated principles, compromised by people with a vested interest in animal experimentation, serve no useful function, increase the cost of drug development, and prevent otherwise safe and efficacious drugs and therapies from being implemented. PMID:22769234

  19. The Nuremberg Code subverts human health and safety by requiring animal modeling

    Directory of Open Access Journals (Sweden)

    Greek Ray

    2012-07-01

    Full Text Available Abstract Background The requirement that animals be used in research and testing in order to protect humans was formalized in the Nuremberg Code and subsequent national and international laws, codes, and declarations. Discussion We review the history of these requirements and contrast what was known via science about animal models then with what is known now. We further analyze the predictive value of animal models when used as test subjects for human response to drugs and disease. We explore the use of animals for models in toxicity testing as an example of the problem with using animal models. Summary We conclude that the requirements for animal testing found in the Nuremberg Code were based on scientifically outdated principles, compromised by people with a vested interest in animal experimentation, serve no useful function, increase the cost of drug development, and prevent otherwise safe and efficacious drugs and therapies from being implemented.

  20. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  1. Development of a model and computer code to describe solar grade silicon production processes

    Science.gov (United States)

    Gould, R. K.; Srivastava, R.

    1979-01-01

    Two computer codes were developed for describing flow reactors in which high purity, solar grade silicon is produced via reduction of gaseous silicon halides. The first is the CHEMPART code, an axisymmetric, marching code which treats two phase flows with models describing detailed gas-phase chemical kinetics, particle formation, and particle growth. It can be used to described flow reactors in which reactants, mix, react, and form a particulate phase. Detailed radial gas-phase composition, temperature, velocity, and particle size distribution profiles are computed. Also, deposition of heat, momentum, and mass (either particulate or vapor) on reactor walls is described. The second code is a modified version of the GENMIX boundary layer code which is used to compute rates of heat, momentum, and mass transfer to the reactor walls. This code lacks the detailed chemical kinetics and particle handling features of the CHEMPART code but has the virtue of running much more rapidly than CHEMPART, while treating the phenomena occurring in the boundary layer in more detail.

  2. Calibration with MCNP of NaI detector for the determination of natural radioactivity levels in the field.

    Science.gov (United States)

    Cinelli, Giorgia; Tositti, Laura; Mostacci, Domiziano; Baré, Jonathan

    2016-05-01

    In view of assessing natural radioactivity with on-site quantitative gamma spectrometry, efficiency calibration of NaI(Tl) detectors is investigated. A calibration based on Monte Carlo simulation of detector response is proposed, to render reliable quantitative analysis practicable in field campaigns. The method is developed with reference to contact geometry, in which measurements are taken placing the NaI(Tl) probe directly against the solid source to be analyzed. The Monte Carlo code used for the simulations was MCNP. Experimental verification of the calibration goodness is obtained by comparison with appropriate standards, as reported. On-site measurements yield a quick quantitative assessment of natural radioactivity levels present ((40)K, (238)U and (232)Th). On-site gamma spectrometry can prove particularly useful insofar as it provides information on materials from which samples cannot be taken.

  3. Evaluation of Advanced Models for PAFS Condensation Heat Transfer in SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Tae-Hwan; Yun, Byong-Jo [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    The PAFS (Passive Auxiliary Feedwater System) is operated by the natural circulation to remove the core decay heat through the PCHX (Passive Condensation Heat Exchanger) which is composed of the nearly horizontal tubes. For validation of the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop) facility was constructed and the condensation heat transfer and natural convection phenomena in the PAFS was experimentally investigated at KAERI (Korea Atomic Energy Research Institute). From the PASCAL experimental result, it was found that conventional system analysis code underestimated the condensation heat transfer. In this study, advanced condensation heat transfer models which can treat the heat transfer mechanisms with the different flow regimes in the nearly horizontal heat exchanger tube were analyzed. The models were implemented in a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant), and it was evaluated with the PASCAL experimental data. With an aim of enhancing the prediction capability for the condensation phenomenon inside the PCHX tube of the PAFS, advanced models for the condensation heat transfer were implemented into the wall condensation model of the SPACE code, so that the PASCAL experimental result was utilized to validate the condensation models. Calculation results showed that the improved model for the condensation heat transfer coefficient enhanced the prediction capability of the SPACE code. This result confirms that the mechanistic modeling for the film condensation in the steam phase and the convection in the condensate liquid contributed to enhance the prediction capability of the wall condensation model of the SPACE code and reduce conservatism in prediction of condensation heat transfer.

  4. Development of RBMK-1500 Model for BDBA Analysis Using RELAP/SCDAPSIM Code

    Science.gov (United States)

    Uspuras, Eugenijus; Kaliatka, Algirdas

    This article discusses the specificity of RBMK (channel type, boiling water, graphite moderated) reactors and problems of Reactor Cooling System modelling employing computer codes. The article presents, how the RELAP/SCDAPSIM code, which is originally designed for modelling of accidents in vessel type reactors, is fit to simulate the phenomena in the RBMK reactor core and RCS in case of Beyond Design Basis Accidents. For this reason, use of two RELAP/SCDAPSIM models is recommended. First model with described complete geometry of RCS is recommended for analysis of initial phase of accident. The calculations results, received using this model, are used as boundary conditions in simplified model for simulation of later phases of severe accidents. The simplified model was tested comparing results of simulation performed using RELAP5 and RELAP/SCDAPSIM codes. As the typical example of BDBA, large break LOCA in reactor cooling system with failure of emergency core cooling system was analyzed. Use of developed models allows to receive behaviour of thermal-hydraulic parameters, temperatures of core components, amount of generated hydrogen due to steam-zirconium reaction. These parameters will be used as input for other special codes, designed for analysis of processes in reactor containment.

  5. Reduced-order LPV model of flexible wind turbines from high fidelity aeroelastic codes

    DEFF Research Database (Denmark)

    Adegas, Fabiano Daher; Sønderby, Ivan Bergquist; Hansen, Morten Hartvig

    2013-01-01

    space. The obtained LPV model is of suitable size for designing modern gain-scheduling controllers based on recently developed LPV control design techniques. Results are thoroughly assessed on a set of industrial wind turbine models generated by the recently developed aeroelastic code HAWCStab2....

  6. Fine-Grained Energy Modeling for the Source Code of a Mobile Application

    DEFF Research Database (Denmark)

    Li, Xueliang; Gallagher, John Patrick

    2016-01-01

    The goal of an energy model for source code is to lay a foundation for the application of energy-aware programming techniques. State of the art solutions are based on source-line energy information. In this paper, we present an approach to constructing a fine-grained energy model which is able...

  7. The modelling of wall condensation with noncondensable gases for the containment codes

    Energy Technology Data Exchange (ETDEWEB)

    Leduc, C.; Coste, P.; Barthel, V.; Deslandes, H. [Commissariat a l`Energi Atomique, Grenoble (France)

    1995-09-01

    This paper presents several approaches in the modelling of wall condensation in the presence of noncondensable gases for containment codes. The lumped-parameter modelling and the local modelling by 3-D codes are discussed. Containment analysis codes should be able to predict the spatial distributions of steam, air, and hydrogen as well as the efficiency of cooling by wall condensation in both natural convection and forced convection situations. 3-D calculations with a turbulent diffusion modelling are necessary since the diffusion controls the local condensation whereas the wall condensation may redistribute the air and hydrogen mass in the containment. A fine mesh modelling of film condensation in forced convection has been in the developed taking into account the influence of the suction velocity at the liquid-gas interface. It is associated with the 3-D model of the TRIO code for the gas mixture where a k-{xi} turbulence model is used. The predictions are compared to the Huhtiniemi`s experimental data. The modelling of condensation in natural convection or mixed convection is more complex. As no universal velocity and temperature profile exist for such boundary layers, a very fine nodalization is necessary. More simple models integrate equations over the boundary layer thickness, using the heat and mass transfer analogy. The model predictions are compared with a MIT experiment. For the containment compartments a two node model is proposed using the lumped parameter approach. Heat and mass transfer coefficients are tested on separate effect tests and containment experiments. The CATHARE code has been adapted to perform such calculations and shows a reasonable agreement with data.

  8. Two Phase Flow Models and Numerical Methods of the Commercial CFD Codes

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sung Won; Jeong, Jae Jun; Chang, Seok Kyu; Cho, Hyung Kyu

    2007-11-15

    The use of commercial CFD codes extend to various field of engineering. The thermal hydraulic analysis is one of the promising engineering field of application of the CFD codes. Up to now, the main application of the commercial CFD code is focused within the single phase, single composition fluid dynamics. Nuclear thermal hydraulics, however, deals with abrupt pressure changes, high heat fluxes, and phase change heat transfer. In order to overcome the CFD limitation and to extend the capability of the nuclear thermal hydraulics analysis, the research efforts are made to collaborate the CFD and nuclear thermal hydraulics. To achieve the final goal, the current useful model and correlations used in commercial CFD codes should be reviewed and investigated. This report gives the summary information about the constitutive relationships that are used in the FLUENT, STAR-CD, and CFX. The brief information of the solution technologies are also enveloped.

  9. Comparison of current state residential energy codes with the 1992 model energy code for one- and two-family dwellings; 1994

    Energy Technology Data Exchange (ETDEWEB)

    Klevgard, L.A.; Taylor, Z.T.; Lucas, R.G.

    1995-01-01

    This report is one in a series of documents describing research activities in support of the US Department of Energy (DOE) Building Energy Codes Program. The Pacific Northwest Laboratory (PNL) leads the program for DOE. The goal of the program is to develop and support the adopting, implementation, and enforcement of Federal, State, and Local energy codes for new buildings. The program approach to meeting the goal is to initiate and manage individual research and standards and guidelines development efforts that are planned and conducted in cooperation with representatives from throughout the buildings community. Projects under way involve practicing architects and engineers, professional societies and code organizations, industry representatives, and researchers from the private sector and national laboratories. Research results and technical justifications for standards criteria are provided to standards development and model code organizations and to Federal, State, and local jurisdictions as a basis to update their codes and standards. This effort helps to ensure that building standards incorporate the latest research results to achieve maximum energy savings in new buildings, yet remain responsive to the needs of the affected professions, organizations, and jurisdictions. Also supported are the implementation, deployment, and use of energy-efficient codes and standards. This report documents findings from an analysis conducted by PNL of the State`s building codes to determine if the codes meet or exceed the 1992 MEC energy efficiency requirements (CABO 1992a).

  10. A review of MAAP4 code structure and core T/H model

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong

    1998-03-01

    The modular accident analysis program (MAAP) version 4 is a computer code that can simulate the response of LWR plants during severe accident sequences and includes models for all of the important phenomena which might occur during accident sequences. In this report, MAAP4 code structure and core thermal hydraulic (T/H) model which models the T/H behavior of the reactor core and the response of core components during all accident phases involving degraded cores are specifically reviewed and then reorganized. This reorganization is performed via getting the related models together under each topic whose contents and order are same with other two reports for MELCOR and SCDAP/RELAP5 to be simultaneously published. Major purpose of the report is to provide information about the characteristics of MAAP4 core T/H models for an integrated severe accident computer code development being performed under the one of on-going mid/long-term nuclear developing project. The basic characteristics of the new integrated severe accident code includes: 1) Flexible simulation capability of primary side, secondary side, and the containment under severe accident conditions, 2) Detailed plant simulation, 3) Convenient user-interfaces, 4) Highly modularization for easy maintenance/improvement, and 5) State-of-the-art model selection. In conclusion, MAAP4 code has appeared to be superior for 3) and 4) items but to be somewhat inferior for 1) and 2) items. For item 5), more efforts should be made in the future to compare separated models in detail with not only other codes but also recent world-wide work. (author). 17 refs., 1 tab., 12 figs.

  11. 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.

  12. Assessment of Turbulence-Chemistry Interaction Models in the National Combustion Code (NCC) - Part I

    Science.gov (United States)

    Wey, Thomas Changju; Liu, Nan-suey

    2011-01-01

    This paper describes the implementations of the linear-eddy model (LEM) and an Eulerian FDF/PDF model in the National Combustion Code (NCC) for the simulation of turbulent combustion. The impacts of these two models, along with the so called laminar chemistry model, are then illustrated via the preliminary results from two combustion systems: a nine-element gas fueled combustor and a single-element liquid fueled combustor.

  13. Verification of the predictive capabilities of the 4C code cryogenic circuit model

    Science.gov (United States)

    Zanino, R.; Bonifetto, R.; Hoa, C.; Richard, L. Savoldi

    2014-01-01

    The 4C code was developed to model thermal-hydraulics in superconducting magnet systems and related cryogenic circuits. It consists of three coupled modules: a quasi-3D thermal-hydraulic model of the winding; a quasi-3D model of heat conduction in the magnet structures; an object-oriented a-causal model of the cryogenic circuit. In the last couple of years the code and its different modules have undergone a series of validation exercises against experimental data, including also data coming from the supercritical He loop HELIOS at CEA Grenoble. However, all this analysis work was done each time after the experiments had been performed. In this paper a first demonstration is given of the predictive capabilities of the 4C code cryogenic circuit module. To do that, a set of ad-hoc experimental scenarios have been designed, including different heating and control strategies. Simulations with the cryogenic circuit module of 4C have then been performed before the experiment. The comparison presented here between the code predictions and the results of the HELIOS measurements gives the first proof of the excellent predictive capability of the 4C code cryogenic circuit module.

  14. Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

    2017-04-15

    The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

  15. The implementation of a toroidal limiter model into the gyrokinetic code ELMFIRE

    Energy Technology Data Exchange (ETDEWEB)

    Leerink, S.; Janhunen, S.J.; Kiviniemi, T.P.; Nora, M. [Euratom-Tekes Association, Helsinki University of Technology (Finland); Heikkinen, J.A. [Euratom-Tekes Association, VTT, P.O. Box 1000, FI-02044 VTT (Finland); Ogando, F. [Universidad Nacional de Educacion a Distancia, Madrid (Spain)

    2008-03-15

    The ELMFIRE full nonlinear gyrokinetic simulation code has been developed for calculations of plasma evolution and dynamics of turbulence in tokamak geometry. The code is applicable for calculations of strong perturbations in particle distribution function, rapid transients and steep gradients in plasma. Benchmarking against experimental reflectometry data from the FT2 tokamak is being discussed and in this paper a model for comparison and studying poloidal velocity is presented. To make the ELMFIRE code suitable for scrape-off layer simulations a simplified toroidal limiter model has been implemented. The model is be discussed and first results are presented. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  16. Transport Corrections in Nodal Diffusion Codes for HTR Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Frederick N. Gleicher

    2010-08-01

    The cores and reflectors of High Temperature Reactors (HTRs) of the Next Generation Nuclear Plant (NGNP) type are dominantly diffusive media from the point of view of behavior of the neutrons and their migration between the various structures of the reactor. This means that neutron diffusion theory is sufficient for modeling most features of such reactors and transport theory may not be needed for most applications. Of course, the above statement assumes the availability of homogenized diffusion theory data. The statement is true for most situations but not all. Two features of NGNP-type HTRs require that the diffusion theory-based solution be corrected for local transport effects. These two cases are the treatment of burnable poisons (BP) in the case of the prismatic block reactors and, for both pebble bed reactor (PBR) and prismatic block reactor (PMR) designs, that of control rods (CR) embedded in non-multiplying regions near the interface between fueled zones and said non-multiplying zones. The need for transport correction arises because diffusion theory-based solutions appear not to provide sufficient fidelity in these situations.

  17. 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)

  18. Comparison of Codes and Neutronics Data Used in the United States and Russia for the TOPAZ-2 Nuclear Safety Assessment

    Science.gov (United States)

    Glushkov, Y. S.; Ponomarov-Stepnoy, N. N.; Kompaniets, G. V.; Gomin, Y. A.; Mayorov, L. V.; Lobyntsev, V. A.; Polyakov, D. N.; Sapir, Joe; Pelowitz, Denise; Streetman, J. Robert

    1994-07-01

    The TOPAZ-2 reactor system is a heterogeneous epithermal system fueled with highly-enriched fuel based on uranium oxide, cooled by a sodium-potassium liquid metal (NaK), using a zirconium hydride moderator, with 37 thermionic fuel elements (TFEs) built into the core. The core is surrounded by a radial beryllium reflector which contains rotating regulating drums with moderating segments. An important problem is the guaranteeing of nuclear safety upon the accidental falling of the TOPAZ-2 reactor into water, which leads to the growth of the reactivity of the reactor. It has turned out that it is necessary to use the Monte-Carlo method for the conduct of neutronics calculations of such a complex reactor. In the United States (U.S.) and Russia, different codes based on the Monte-Carlo method are used for calculations - the MCNP code in the U.S., and the MCU-2 code in Russia. The goal of this work is the comparison of the codes and neutronics data used in the U.S. and Russia for the basis of the TOPAZ-2 nuclear safety. With this goal, a joint computer model benchmark of the TOPAZ-2 reactor was developed and the calculations of a series of variants, differing by the presence and absence of water in the reactor cavities and behind the radial reflector, in the position of the regulating drums, in the presence of the radial reflector, etc. were done independently by specialists in both the U.S. and Russia. Along with the reactor calculations, calculations were also done of the nuclei of the core using the MCNP code (U.S.) and the MCU-2 code (Russia). The work done allowed one to obtain results comparing the MCNP code to the MCU-2 code which gave somewhat different results both for the absolute values of Keff and for reactivity effects. In the future it remains to conduct a detailed analysis of the reasons for the discrepancies. For this it is necessary to exchange neutronics data used for TOPAZ-2 reactor calculations in the U.S. and Russia.

  19. Analysis of uncertainty in the X-ray simulation using MCNPS; Analisis de incertidumbres en la simulacion de Rayos X utilizados MCNP5

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

    The characterization of the primary X-ray spectrum is a very useful tool for quality control of radiology equipment, as it allows a better estimation of the dose received by the patient, and better image quality. In this work, we have developed a Monte Carlo model with MCNP5 program to simulate X-ray spectra from an electron source and the characteristics of a commercial X-ray tube.

  20. An evolutionary model for protein-coding regions with conserved RNA structure

    DEFF Research Database (Denmark)

    Pedersen, Jakob Skou; Forsberg, Roald; Meyer, Irmtraud Margret

    2004-01-01

    components of traditional phylogenetic models. We applied this to a data set of full-genome sequences from the hepatitis C virus where five RNA structures are mapped within the coding region. This allowed us to partition the effects of selection on different structural elements and to test various hypotheses...... concerning the relation of these effects. Of particular interest, we found evidence of a functional role of loop and bulge regions, as these were shown to evolve according to a different and more constrained selective regime than the nonpairing regions outside the RNA structures. Other potential applications...... of the model include comparative RNA structure prediction in coding regions and RNA virus phylogenetics....

  1. 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.

  2. A WYNER-ZIV VIDEO CODING METHOD UTILIZING MIXTURE CORRELATION NOISE MODEL

    Institute of Scientific and Technical Information of China (English)

    Hu Xiaofei; Zhu Xiuchang

    2012-01-01

    In Wyner-Ziv (WZ) Distributed Video Coding (DVC),correlation noise model is often used to describe the error distribution between WZ frame and the side information.The accuracy of the model can influence the performance of the video coder directly.A mixture correlation noise model in Discrete Cosine Transform (DCT) domain for WZ video coding is established in this paper.Different correlation noise estimation method is used for direct current and alternating current coefficients.Parameter estimation method based on expectation maximization algorithm is used to estimate the Laplace distribution center of direct current frequency band and Mixture Laplace-Uniform Distribution Model (MLUDM) is established for alternating current coefficients.Experimental results suggest that the proposed mixture correlation noise model can describe the heavy tail and sudden change of the noise accurately at high rate and make significant improvement on the coding efficiency compared with the noise model presented by DIStributed COding for Video sERvices (DISCOVER).

  3. Process Model Improvement for Source Code Plagiarism Detection in Student Programming Assignments

    Directory of Open Access Journals (Sweden)

    Dragutin KERMEK

    2016-04-01

    Full Text Available In programming courses there are various ways in which students attempt to cheat. The most commonly used method is copying source code from other students and making minimal changes in it, like renaming variable names. Several tools like Sherlock, JPlag and Moss have been devised to detect source code plagiarism. However, for larger student assignments and projects that involve a lot of source code files these tools are not so effective. Also, issues may occur when source code is given to students in class so they can copy it. In such cases these tools do not provide satisfying results and reports. In this study, we present an improved process model for plagiarism detection when multiple student files exist and allowed source code is present. In the research in this paper we use the Sherlock detection tool, although the presented process model can be combined with any plagiarism detection engine. The proposed model is tested on assignments in three courses in two subsequent academic years.

  4. SENR, A Super-Efficient Code for Gravitational Wave Source Modeling: Latest Results

    Science.gov (United States)

    Ruchlin, Ian; Etienne, Zachariah; Baumgarte, Thomas

    2017-01-01

    The science we extract from gravitational wave observations will be limited by our theoretical understanding, so with the recent breakthroughs by LIGO, reliable gravitational wave source modeling has never been more critical. Due to efficiency considerations, current numerical relativity codes are very limited in their applicability to direct LIGO source modeling, so it is important to develop new strategies for making our codes more efficient. We introduce SENR, a Super-Efficient, open-development numerical relativity (NR) code aimed at improving the efficiency of moving-puncture-based LIGO gravitational wave source modeling by 100x. SENR builds upon recent work, in which the BSSN equations are evolved in static spherical coordinates, to allow dynamical coordinates with arbitrary spatial distributions. The physical domain is mapped to a uniform-resolution grid on which derivative operations are approximated using standard central finite difference stencils. The source code is designed to be human-readable, efficient, parallelized, and readily extensible. We present the latest results from the SENR code.

  5. Self-consistent modeling of DEMOs with 1.5D BALDUR integrated predictive modeling code

    Science.gov (United States)

    Wisitsorasak, A.; Somjinda, B.; Promping, J.; Onjun, T.

    2017-02-01

    Self-consistent simulations of four DEMO designs proposed by teams from China, Europe, India, and Korea are carried out using the BALDUR integrated predictive modeling code in which theory-based models are used, for both core transport and boundary conditions. In these simulations, a combination of the NCLASS neoclassical transport and multimode (MMM95) anomalous transport model is used to compute a core transport. The boundary is taken to be at the top of the pedestal, where the pedestal values are described using a pedestal temperature model based on a combination of magnetic and flow shear stabilization, pedestal width scaling and an infinite- n ballooning pressure gradient model and a pedestal density model based on a line average density. Even though an optimistic scenario is considered, the simulation results suggest that, with the exclusion of ELMs, the fusion gain Q obtained for these reactors is pessimistic compared to their original designs, i.e. 52% for the Chinese design, 63% for the European design, 22% for the Korean design, and 26% for the Indian design. In addition, the predicted bootstrap current fractions are also found to be lower than their original designs, as fractions of their original designs, i.e. 0.49 (China), 0.66 (Europe), and 0.58 (India). Furthermore, in relation to sensitivity, it is found that increasing values of the auxiliary heating power and the electron line average density from their design values yield an enhancement of fusion performance. In addition, inclusion of sawtooth oscillation effects demonstrate positive impacts on the plasma and fusion performance in European, Indian and Korean DEMOs, but degrade the performance in the Chinese DEMO.

  6. Basic Pilot Code Development for Two-Fluid, Three-Field Model

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Bae, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.; Ha, K. S.; Kang, D. H

    2006-03-15

    A basic pilot code for one-dimensional, transient, two-fluid, three-field model has been developed. Using 9 conceptual problems, the basic pilot code has been verified. The results of the verification are summarized below: - It was confirmed that the basic pilot code can simulate various flow conditions (such as single-phase liquid flow, bubbly flow, slug/churn turbulent flow, annular-mist flow, and single-phase vapor flow) and transitions of the flow conditions. A mist flow was not simulated, but it seems that the basic pilot code can simulate mist flow conditions. - The pilot code was programmed so that the source terms of the governing equations and numerical solution schemes can be easily tested. - The mass and energy conservation was confirmed for single-phase liquid and single-phase vapor flows. - It was confirmed that the inlet pressure and velocity boundary conditions work properly. - It was confirmed that, for single- and two-phase flows, the velocity and temperature of non-existing phase are calculated as intended. - During the simulation of a two-phase flow, the calculation reaches a quasisteady state with small-amplitude oscillations. The oscillations seem to be induced by some numerical causes. The research items for the improvement of the basic pilot code are listed in the last section of this report.

  7. A Perceptual Model for Sinusoidal Audio Coding Based on Spectral Integration

    Directory of Open Access Journals (Sweden)

    Jensen Søren Holdt

    2005-01-01

    Full Text Available Psychoacoustical models have been used extensively within audio coding applications over the past decades. Recently, parametric coding techniques have been applied to general audio and this has created the need for a psychoacoustical model that is specifically suited for sinusoidal modelling of audio signals. In this paper, we present a new perceptual model that predicts masked thresholds for sinusoidal distortions. The model relies on signal detection theory and incorporates more recent insights about spectral and temporal integration in auditory masking. As a consequence, the model is able to predict the distortion detectability. In fact, the distortion detectability defines a (perceptually relevant norm on the underlying signal space which is beneficial for optimisation algorithms such as rate-distortion optimisation or linear predictive coding. We evaluate the merits of the model by combining it with a sinusoidal extraction method and compare the results with those obtained with the ISO MPEG-1 Layer I-II recommended model. Listening tests show a clear preference for the new model. More specifically, the model presented here leads to a reduction of more than 20% in terms of number of sinusoids needed to represent signals at a given quality level.

  8. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  9. Assessment of evaluated (n,d) energy-angle elastic scattering distributions using MCNP simulations of critical measurements and simplified calculation benchmarks

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2008-07-01

    Different evaluated (n,d) energy-angle elastic scattering distributions produce k-effective differences in MCNP5 simulations of critical experiments involving heavy water (D{sub 2}O) of sufficient magnitude to suggest a need for new (n,d) scattering measurements and/or distributions derived from modern theoretical nuclear models, especially at neutron energies below a few MeV. The present work focuses on the small reactivity change of < 1 mk that is observed in the MCNP5 D{sub 2}O coolant-void-reactivity calculation bias for simulations of two pairs of critical experiments performed in the ZED-2 reactor at the Chalk River Laboratories when different nuclear data libraries are used for deuterium. The deuterium data libraries tested include Endf/B-VII.0, Endf/B-VI.4, JENDL-3.3 and a new evaluation, labelled Bonn-B, which is based on recent theoretical nuclear-model calculations. Comparison calculations were also performed for a simplified, two-region, spherical model having an inner, 250-cm radius, homogeneous sphere of UO{sub 2}, without and with deuterium, and an outer 20-cm-thick deuterium reflector. A notable observation from this work is the reduction of about 0.4 mk in the MCNP5 ZED-2 CVR calculation bias that is obtained when the O-in-UO{sub 2} thermal scattering data comes from Endf-B-VII.0. (author)

  10. 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%.

  11. A Neuronal Model of Predictive Coding Accounting for the Mismatch Negativity

    OpenAIRE

    Wacongne, Catherine; Changeux, Jean-Pierre; Dehaene, Stanislas

    2012-01-01

    International audience; The mismatch negativity (MMN) is thought to index the activation of specialized neural networks for active prediction and deviance detection. However, a detailed neuronal model of the neurobiological mechanisms underlying the MMN is still lacking, and its computational foundations remain debated. We propose here a detailed neuronal model of auditory cortex, based on predictive coding, that accounts for the critical features of MMN. The model is entirely composed of spi...

  12. Benchmarking Defmod, an open source FEM code for modeling episodic fault rupture

    Science.gov (United States)

    Meng, Chunfang

    2017-03-01

    We present Defmod, an open source (linear) finite element code that enables us to efficiently model the crustal deformation due to (quasi-)static and dynamic loadings, poroelastic flow, viscoelastic flow and frictional fault slip. Ali (2015) provides the original code introducing an implicit solver for (quasi-)static problem, and an explicit solver for dynamic problem. The fault constraint is implemented via Lagrange Multiplier. Meng (2015) combines these two solvers into a hybrid solver that uses failure criteria and friction laws to adaptively switch between the (quasi-)static state and dynamic state. The code is capable of modeling episodic fault rupture driven by quasi-static loadings, e.g. due to reservoir fluid withdraw or injection. Here, we focus on benchmarking the Defmod results against some establish results.

  13. Present capabilities and new developments in antenna modeling with the numerical electromagnetics code NEC

    Energy Technology Data Exchange (ETDEWEB)

    Burke, G.J.

    1988-04-08

    Computer modeling of antennas, since its start in the late 1960's, has become a powerful and widely used tool for antenna design. Computer codes have been developed based on the Method-of-Moments, Geometrical Theory of Diffraction, or integration of Maxwell's equations. Of such tools, the Numerical Electromagnetics Code-Method of Moments (NEC) has become one of the most widely used codes for modeling resonant sized antennas. There are several reasons for this including the systematic updating and extension of its capabilities, extensive user-oriented documentation and accessibility of its developers for user assistance. The result is that there are estimated to be several hundred users of various versions of NEC world wide. 23 refs., 10 figs.

  14. Bootstrap imputation with a disease probability model minimized bias from misclassification due to administrative database codes.

    Science.gov (United States)

    van Walraven, Carl

    2017-04-01

    Diagnostic codes used in administrative databases cause bias due to misclassification of patient disease status. It is unclear which methods minimize this bias. Serum creatinine measures were used to determine severe renal failure status in 50,074 hospitalized patients. The true prevalence of severe renal failure and its association with covariates were measured. These were compared to results for which renal failure status was determined using surrogate measures including the following: (1) diagnostic codes; (2) categorization of probability estimates of renal failure determined from a previously validated model; or (3) bootstrap methods imputation of disease status using model-derived probability estimates. Bias in estimates of severe renal failure prevalence and its association with covariates were minimal when bootstrap methods were used to impute renal failure status from model-based probability estimates. In contrast, biases were extensive when renal failure status was determined using codes or methods in which model-based condition probability was categorized. Bias due to misclassification from inaccurate diagnostic codes can be minimized using bootstrap methods to impute condition status using multivariable model-derived probability estimates. Copyright © 2017 Elsevier Inc. All rights reserved.

  15. TORT solutions to the NEA suite of benchmarks for 3D transport methods and codes over a range in parameter space

    Energy Technology Data Exchange (ETDEWEB)

    Bekar, Kursat B. [Department of Mechanical and Nuclear Engineering, Penn State University, University Park, PA 16802 (United States)], E-mail: bekarkb@ornl.gov; Azmy, Yousry Y. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)], E-mail: yyazmy@ncsu.edu

    2009-04-15

    We present the TORT solutions to the 3D transport codes' suite of benchmarks exercise. An overview of benchmark configurations is provided, followed by a description of the TORT computational model we developed to solve the cases comprising the benchmark suite. In the numerical experiments reported in this paper, we chose to refine the spatial and angular discretizations simultaneously, from the coarsest model (40 x 40 x 40, 200 angles) to the finest model (160 x 160 x 160, 800 angles). The MCNP reference solution is used for evaluating the effect of model-refinement on the accuracy of the TORT solutions. The presented results show that the majority of benchmark quantities are computed with good accuracy by TORT, and that the accuracy improves with model refinement. However, this deliberately severe test has exposed some deficiencies in both deterministic and stochastic solution approaches. Specifically, TORT fails to converge the inner iterations in some benchmark configurations while MCNP produces zero tallies, or drastically poor statistics for some benchmark quantities. We conjecture that TORT's failure to converge is driven by ray effects in configurations with low scattering ratio and/or highly skewed computational cells, i.e. aspect ratio far from unity. The failure of MCNP occurs in quantities tallied over a very small area or volume in physical space, or quantities tallied many ({approx}25) mean free paths away from the source. Hence automated, robust, and reliable variance reduction techniques are essential for obtaining high quality reference values of the benchmark quantities. Preliminary results of the benchmark exercise indicate that the occasionally poor performance of TORT is shared with other deterministic codes. Armed with this information, method developers can now direct their attention to regions in parameter space where such failures occur and design alternative solution approaches for such instances.

  16. Comparison and physical interpretation of MCNP and TART neutron and {gamma} Monte Carlo shielding calculations for a heavy-ion ICF system

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, E. E-mail: enrico@nuc.berkeley.edu; Premuda, F.; Lee, E

    2004-01-01

    Inertial confinement fusion (ICF) aims to induce implosions of D-T pellets to obtain a extremely dense and hot plasma with lasers or heavy-ion beams. For heavy-ion fusion (HIF), recent research has focused on 'liquid-protected' designs that allow highly compact target chambers. In the design of a reactor such as HYLIFE-II [Fus. Techol. 25 (1984); HYLIFE-II Progress Report, UCID-21816, 4.82-100], the liquid used is a molten salt made of F{sup 10}, Li{sup 6}, Li{sup 7}, Be{sup 9} (called flibe). Flibe allows the final-focus magnets to be closer to the target, which helps to reduce the focus spot size and in turn the size of the driver, with a large reduction of the cost of HIF electricity. Consequently the superconducting coils of the magnets closer to the D-T neutron source will potentially suffer higher damage though they can stand only a certain amount of energy deposited before quenching. This work has been primarily focusing on verifying that total energy deposited by fusion neutrons and induced {gamma} rays remain under such limit values and the final purpose is the optimization of the shielding of the magnetic lens system from the points of view of the geometrical configuration and of the physical nature of the materials adopted. The system is analyzed in terms of six geometrical models going from simplified up to much more realistic representations of a system of 192 beam lines, each focused by six magnets. A 3-D transport calculation of the radiation penetrating through ducts, that takes into account the complexity of the system, requires Monte Carlo methods. The technical nature of the design problem and the methodology followed were presented in a previous paper [Nucl. Instr. and Meth. A 464 (2001) 410] by summarizing briefly the results for the deposited energy distribution on the six focal magnets of a beam line. Now a comparison of the performances of the two codes TART98 [TART98: A Coupled Neutron-Photon 3-D Combinational Geometry Monte Carlo

  17. Development Of Sputtering Models For Fluids-Based Plasma Simulation Codes

    Science.gov (United States)

    Veitzer, Seth; Beckwith, Kristian; Stoltz, Peter

    2015-09-01

    Rf-driven plasma devices such as ion sources and plasma processing devices for many industrial and research applications benefit from detailed numerical modeling. Simulation of these devices using explicit PIC codes is difficult due to inherent separations of time and spatial scales. One alternative type of model is fluid-based codes coupled with electromagnetics, that are applicable to modeling higher-density plasmas in the time domain, but can relax time step requirements. To accurately model plasma-surface processes, such as physical sputtering and secondary electron emission, kinetic particle models have been developed, where particles are emitted from a material surface due to plasma ion bombardment. In fluid models plasma properties are defined on a cell-by-cell basis, and distributions for individual particle properties are assumed. This adds a complexity to surface process modeling, which we describe here. We describe the implementation of sputtering models into the hydrodynamic plasma simulation code USim, as well as methods to improve the accuracy of fluids-based simulation of plasmas-surface interactions by better modeling of heat fluxes. This work was performed under the auspices of the Department of Energy, Office of Basic Energy Sciences Award #DE-SC0009585.

  18. Motion-compensated coding and frame rate up-conversion: models and analysis.

    Science.gov (United States)

    Dar, Yehuda; Bruckstein, Alfred M

    2015-07-01

    Block-based motion estimation (ME) and motion compensation (MC) techniques are widely used in modern video processing algorithms and compression systems. The great variety of video applications and devices results in diverse compression specifications, such as frame rates and bit rates. In this paper, we study the effect of frame rate and compression bit rate on block-based ME and MC as commonly utilized in inter-frame coding and frame rate up-conversion (FRUC). This joint examination yields a theoretical foundation for comparing MC procedures in coding and FRUC. First, the video signal is locally modeled as a noisy translational motion of an image. Then, we theoretically model the motion-compensated prediction of available and absent frames as in coding and FRUC applications, respectively. The theoretic MC-prediction error is studied further and its autocorrelation function is calculated, yielding useful separable-simplifications for the coding application. We argue that a linear relation exists between the variance of the MC-prediction error and temporal distance. While the relevant distance in MC coding is between the predicted and reference frames, MC-FRUC is affected by the distance between the frames available for interpolation. We compare our estimates with experimental results and show that the theory explains qualitatively the empirical behavior. Then, we use the models proposed to analyze a system for improving of video coding at low bit rates, using a spatio-temporal scaling. Although this concept is practically employed in various forms, so far it lacked a theoretical justification. We here harness the proposed MC models and present a comprehensive analysis of the system, to qualitatively predict the experimental results.

  19. Solar optical codes evaluation for modeling and analyzing complex solar receiver geometries

    Science.gov (United States)

    Yellowhair, Julius; Ortega, Jesus D.; Christian, Joshua M.; Ho, Clifford K.

    2014-09-01

    Solar optical modeling tools are valuable for modeling and predicting the performance of solar technology systems. Four optical modeling tools were evaluated using the National Solar Thermal Test Facility heliostat field combined with flat plate receiver geometry as a benchmark. The four optical modeling tools evaluated were DELSOL, HELIOS, SolTrace, and Tonatiuh. All are available for free from their respective developers. DELSOL and HELIOS both use a convolution of the sunshape and optical errors for rapid calculation of the incident irradiance profiles on the receiver surfaces. SolTrace and Tonatiuh use ray-tracing methods to intersect the reflected solar rays with the receiver surfaces and construct irradiance profiles. We found the ray-tracing tools, although slower in computation speed, to be more flexible for modeling complex receiver geometries, whereas DELSOL and HELIOS were limited to standard receiver geometries such as flat plate, cylinder, and cavity receivers. We also list the strengths and deficiencies of the tools to show tool preference depending on the modeling and design needs. We provide an example of using SolTrace for modeling nonconventional receiver geometries. The goal is to transfer the irradiance profiles on the receiver surfaces calculated in an optical code to a computational fluid dynamics code such as ANSYS Fluent. This approach eliminates the need for using discrete ordinance or discrete radiation transfer models, which are computationally intensive, within the CFD code. The irradiance profiles on the receiver surfaces then allows for thermal and fluid analysis on the receiver.

  20. A study on the dependency between turbulent models and mesh configurations of CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Jungjin; Heo, Yujin; Jerng, Dong-Wook [CAU, Seoul (Korea, Republic of)

    2015-10-15

    This paper focuses on the analysis of the behavior of hydrogen mixing and hydrogen stratification, using the GOTHIC code and the CFD code. Specifically, we examined the mesh sensitivity and how the turbulence model affects hydrogen stratification or hydrogen mixing, depending on the mesh configuration. In this work, sensitivity analyses for the meshes and the turbulence models were conducted for missing and stratification phenomena. During severe accidents in a nuclear power plants, the generation of hydrogen may occur and this will complicate the atmospheric condition of the containment by causing stratification of air, steam, and hydrogen. This could significantly impact containment integrity analyses, as hydrogen could be accumulated in local region. From this need arises the importance of research about stratification of gases in the containment. Two computation fluid dynamics code, i.e. GOTHIC and STAR-CCM+ were adopted and the computational results were benchmarked against the experimental data from PANDA facility. The main findings observed through the present work can be summarized as follows: 1) In the case of the GOTHIC code, it was observed that the aspect ratio of the mesh was found more important than the mesh size. Also, if the number of the mesh is over 3,000, the effects of the turbulence models were marginal. 2) For STAR-CCM+, the tendency is quite different from the GOTHIC code. That is, the effects of the turbulence models were small for fewer number of the mesh, however, as the number of mesh increases, the effects of the turbulence models becomes significant. Another observation is that away from the injection orifice, the role of the turbulence models tended to be important due to the