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‣ Análise do comportamento sob irradiação do combustível nuclear a altas queimas com os programas computacionais FRAPCON e FRAPTRAN; Analysis of the behavior under irradiation of high burnup nuclear fuels with the computer programs FRAPCON and FRAPTRAN

Reis, Regis
Fonte: Biblioteca Digitais de Teses e Dissertações da USP Publicador: Biblioteca Digitais de Teses e Dissertações da USP
Tipo: Dissertação de Mestrado Formato: application/pdf
Publicado em 19/08/2014 Português
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O objetivo deste trabalho é verificar a validade e a acurácia dos resultados fornecidos pelos programas computacionais FRAPCON-3.4a e FRAPTRAN-1.4, utilizados no processo de simulação do comportamento de varetas combustíveis de reatores a água leve pressurizada PWR (Pressurized Water Reactor), sob situações operacionais de regimes permanente e transiente, em condições de alta queima (high burnup). Para realizar a verificação, foi utilizada a base de dados FUMEX-III, que fornece dados relativos a experimentos realizados com diversos tipos de combustíveis nucleares, submetidos a diversas condições operacionais. Através dos resultados obtidos nas simulações computacionais com os programas FRAPCON-3.4a e FRAPTRAN-1.4 e da sua comparação com os dados experimentais da base FUMEX-III, foi possível constatar que os programas empregados possuem um boa capacidade de predizer o comportamento operacional de varetas combustíveis de PWR em regime permanente a altas queimas e sob condição de transiente inicializado por reatividade (Reactivity Initiated Accident RIA).; The objective of this work is to verify the validity and accuracy of the results provided by the computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods in steady-state and transient operational conditions at high burnup. To perform the verification...

‣ TXM-Wizard: a program for advanced data collection and evaluation in full-field transmission X-ray microscopy

Liu, Yijin; Meirer, Florian; Williams, Phillip A.; Wang, Junyue; Andrews, Joy C.; Pianetta, Piero
Fonte: International Union of Crystallography Publicador: International Union of Crystallography
Tipo: Artigo de Revista Científica
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A suite of GUI programs written in MATLAB for advanced data collection and analysis of full-field transmission X-ray microscopy data including mosaic imaging, tomography and XANES imaging is presented.

‣ GENFIT: software for the analysis of small-angle X-ray and neutron scattering data of macro­molecules in solution

Spinozzi, Francesco; Ferrero, Claudio; Ortore, Maria Grazia; De Maria Antolinos, Alejandro; Mariani, Paolo
Fonte: International Union of Crystallography Publicador: International Union of Crystallography
Tipo: Artigo de Revista Científica
Publicado em 10/05/2014 Português
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GENFIT is a new computer code featuring an advanced model-fitting capability to analyse small-angle X-ray and neutron scattering data of macromolecular systems. Batches of experimental curves can be simultaneously best fitted using common parameters issued from combinations of models and, if applicable, constrained by physical and/or phenomenological relations.

‣ MITR-II fuel management, core depletion, and analysis : codes developed for the diffusion theory program CITATION

Bernard, John Albert
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 2 v. (826 leaves); 49885326 bytes; 49885082 bytes; application/pdf; application/pdf
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by John A. Bernard, Jr.; Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1979.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; Includes bibliographical references.

‣ An emulator of an engine-car system by an engine-dynamometer system

Lee, Wing Hong
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 89 leaves; 3724771 bytes; 3724531 bytes; application/pdf; application/pdf
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by Wing Hong Lee.; Thesis (Elec.E)--Massachusetts Institute of Technology, Dept. of Electrical Engineering and Computer Science, 1980.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND ENGINEERING.; Includes bibliographical references.

‣ Improved multidimensional numerical methods for the steady state and transient thermal-hydraulic analysis of fuel pin bundles and nuclear reactor cores.

Masterson, Robert Edward
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 161, [78] leaves; 12997826 bytes; 12997582 bytes; application/pdf; application/pdf
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Thesis. 1977. Sc.D.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; Includes bibliographical references.

‣ The conditional/generalized maximum likelihood logit computer program : instructions for use, energy management and economics

Massachusetts Institute of Technology Energy Laboratory in association with the Sloan School of Management and the Dept. of Urban Studies and Planning.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 1072505 bytes; application/pdf
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Prepared for the U.S. Dept. of Energy under Contract no. EX-76-A-01-2295, Task order 37.

‣ A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis

Kazimi, Mujid S.; Massoud, Mahmoud
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 5295128 bytes; application/pdf
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A review is made of the computer codes developed in the U.S. for thermal-hydraulic analysis of nuclear reactors. The intention of this review is to compare these codes on the basis of their numerical method and physical models with particular attention to the two-phase flow and heat transfer characteristics. A chronology of the most documented codes such as COBRA and RELAP is given. The features of the recent codes as RETRAN, TRAC and THERMIT are also reviewed. The range of application as well as limitations of the various codes are discussed.; Sponsored by Boston Edison Company and others under MIT Energy Laboratory Electric Utility Program.

‣ Development of models for the sodium version of the two-phase three dimensional thermal hydraulics code THERMIT

Wilson, Gregory James; Kazimi, Mujid S.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 6959220 bytes; application/pdf
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Several different models and correlations were developed and incorporated in the sodium version of THERMIT, a thermal- hydraulics code written at MIT for the purpose of analyzing transients under LMFBR conditions. This includes: a mechanism for the inclusion of radial heat conduction in the sodium coolant as well as radial heat loss to the structure surrounding the test section. The fuel rod conduction scheme was modified to allow for more flexibility in modelling the gas plenum regions and fuel restructuring. The formulas for mass and momentum exchange between the liquid and vapor phases were improved. The single phase and two phase friction factors were replaced by correlations more appropriate to LMFBR assembly geometry. The models incorporated in THERMIT were tested by running the code to simulate the results of the THORS Bundle 6A experiments performed at Oak Ridge National Laboratory. The results demonstrate the increased accuracy provided by the inclusion of these effects.; "Sponsored by U.S Department of Energy, General Electric Co. and Hanford Engineering Development Laboratory."

‣ A two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

Granziera, Mario Roberto; Kazimi, Mujid S.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 11731420 bytes; application/pdf
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A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identi- fication of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. The most important feature of the model was its ability to simulate severe conditions of sodium boiling, in particular flow reversal, which was shown in the tests performed with the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions.; "Sponsored by U.S Department of Energy, General Electric Co. and Hanford Engineering Development Laboratory."

‣ 22.351 Systems Analysis of the Nuclear Fuel Cycle, Spring 2003; Systems Analysis of the Nuclear Fuel Cycle

Kazimi, Mujid S.; Pilat, Edward E.
Fonte: MIT - Massachusetts Institute of Technology Publicador: MIT - Massachusetts Institute of Technology
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In-depth technical and policy analysis of various options for the nuclear fuel cycle. Topics include uranium supply, enrichment fuel fabrication, in-core physics and fuel management of uranium, thorium and other fuel types, reprocessing and waste disposal. Principles of fuel cycle economics and the applied reactor physics of both contemporary and proposed thermal and fast reactors are presented. Nonproliferation aspects, disposal of excess weapons plutonium, and transmutation of actinides and selected fission products in spent fuel are examined. Several state-of-the-art computer programs are provided for student use in problem sets and term papers.

‣ CAC on a MAC: setting up a DOD Common Access Card reader on the Macintosh OS X operating system

Hopfner, Phil
Fonte: Monterey, California. Naval Postgraduate School Publicador: Monterey, California. Naval Postgraduate School
Tipo: Relatório Formato: 14, p.: ill. (some col.);28 cm.
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The Naval Postgraduate School, along with many other Department of Defense (DOD) organizations, utilizes the ActivCard USB Common Access Card (CAC) readers. The CAC readers in conjunction with the user's Smart Card enables access to DOD PKI-enabled websites and allows the user to send signed and encrypted email utilizing the DOD Public Key Infrastructure (PKI). Microsoft Windows systems utilize the ActivCard Gold middleware software to enable CAC reader functionality. This software package is well integrated and documented in the Microsoft Windows environment. Starting with Macintosh OS X 10.4.x, there was no need to install any middleware software as all of the support for the US Federal Smart Cards is built in. This document details how to update the ActivCard readers to make them fully compliant with OS X 10.4.x and then details the steps necessary to setup the system in order to use the CAC readers under the Macintosh OS X 10.4.x operating system.

‣ The fuel cycle economics of improved uranium utilization in light water reactors

Abbaspour, Ali Tehrani; Driscoll, Michael J.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 9420038 bytes; application/pdf
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A simple fuel cycle cost model has been formulated, tested satisfactorily (within better than 3% for a wide range of cases) using a more elaborate computer program, and applied to evaluate a variety of PWR fuel cyclesand fuel management options, with an emphasis on issues pertinent to the NASAP/INFCE efforts. The uranium and thorium cycles were examined, lattice fuel-to-moderator and burnup were varied, and once-through and recycle modes were examined. It was found that increasing core burnup was economically advantageous, particularly if busbar or total system cost is considered in lieu of fuel cycle cost only, for both once-through and recycle modes, so long as the number of staggered core batches is increased concurrently. When optimized under comparable ground rules, the once-through fuel cycle is competitive with the recycle option; differences are well within the rather large (+ 20%) one sigma uncertainty estimated for the overall fuel cycle costs by propagating uncertainties in input data. Optimization on mills/kwhre and ore usage, tones/GWe,yr, are generally, but not universally, compatible criteria. To the extent evaluated, the thorium fuel cycle was not found to be economically competitive. Cost-optimum thorium lattices were found to be drier than for current PWRs...

‣ Development of a method for BWR subchannel analysis

Faya, Artur José Goncalves; Wolf, Lothar; Todreas, Neil E.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 6493270 bytes; application/pdf
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This study deals with the development of a computer pro- gram for steady-state and transient BWR subchannel analysis. The conservation equations for the subchannels are obtained by area-averaging of the two-fluid model conservation equa- tions and reducing them to the drift-flux model formulation. The conservation equations are solved by a marching type technique which limits the code to analysis of operational transients only. The transfer of mass, momentum and energy between adjacent subchannels is split into diversion cross- flow and turbulent mixing components. The transfer of mass by turbulent mixing is assumed to occur in a volume-for- volume scheme reflecting experimental observations. The phenomenon of lateral vapor drift and mixing enhancement with flow regime are included in the mixing model of the program. The following experiments are used for the purpose of the assessment of the code under steady-state conditions: 1) GE Nine-Rod tests with radially uniform and nonuniform heating 2) Studsvik Nine-rod tests with strong radial power tilt 3) Ispra Sixteen-rod tests with radially uniform heating Comparison of calculated results with these data shows that the program is capable of predicting the correct trends in exit mass velocity and quality distributions.; Originally presented as the author's thesis...

‣ CANAL user's manual

Faya, Artur José Goncalves; Wolf, Lothar; Todreas, Neil E.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 4194381 bytes; application/pdf
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This report gives a detailed description of the input data and contains a listing of the computer program CANAL. A sample problem is also provided.

‣ Fuel cycle optimization of thorium and uranium fueled PWR systems

Garel, Keith Courtnay; Driscoll, Michael J.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 13486795 bytes; application/pdf
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The burnup neutronics of uniform PWR lattices are examined with respect to reduction of uranium ore requirements with an emphasis on variation of the fuel-to-moderator ratio (lattice pitch at constant fuel pin diameter) and the use of thorium. Fuel cycles using all combinations of the major fissile (U-235, U-233, Pu) and fertile (U-238, Th) species are examined. The LEOPARD code and prescriptions developed from a linear reactivity model are used to determine initial core and annual makeup fissile requirements for input into an in-house, simple, systems model, MASFLO-2, which calculates ore (and separative work) requirements per GWeyr for growing, declining, or finite-life nuclear electric systems. For low growth scenarios drier lattices are favored, and the thorium fuel cycle requires as much as 23% less ore than a comparably optimized uranium cycle with full recycle. For unmodified lattices, the thorium fuel cycle with full recycle exhibits long term uranium ore savings of 17% over the comparable uranium cycle with full recycle. For rapidly growing systems, drier lattices, and those using thorium, are less attractive because of their high startup inventories. Thus the introduction of thorium may increase ore and separative work requirements in the short term but will more than repay the ore investment in the very long term. Very little improvement was achieved by varying fuel pin diameter at a given fuel-to-moderator ratio...

‣ WOSUB : a subchannel code for steady-state and transient thermal-hydraulic analysis of BWR fuel pin bundles.

Wolf, Lothar; Guillebaud, Louis Jean Marie; Faya, A.
Fonte: Massachusetts Institute of Technology. Energy Laboratory Publicador: Massachusetts Institute of Technology. Energy Laboratory
Tipo: Relatório Formato: 7735220 bytes; application/pdf
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The WOSUB-codes are spin-offs and extensions of the MATTEO-code [1]. The series of three reports describe WOSUB-I and WOSUB-II in their respective status as of July 31, 1977. This report is the first in a series of three, the second of which contains the user's manual [2] and the third [3] summarizes the assessment and comparison with experimental data and various other subchannel codes. The present report introduces the drift-flux and vapor diffusion models employed by the code, discusses the solution method and reviews the constitutive equations presently built into the code. Wherever applicable, possible exteriors of the models are indicated especially with due regard of the findings presented in [3]. Overall, the review of the model and the package of constitutive equations demonstrate that WOSUB-I and II constitute true alternatives for BWR bundle and PWR test bundle calculations as compared to the commonly applied COBRA-IIIC, and COBRA-IIIC/MIT codes which were primarily designed for PWR subchannel and core calculations, respectively. In fact, the incorporation of the drift flux and the vapor diffusion pro- cesses into a subchannel code has to be cdnsidered.a major step towards a more basic understanding and a well balanced engineer- ing approach without the extra burden of a true two-fluid two- phase model. Recommendations for improvements in the various areas are indicated and should serve as guidelines for future develop- ment of this code which in light of the encouraging results pre- sented in [3] seems to be highly warranted. The WOSUB-code is still in the stage of evolutionary development. In this context...

‣ WOSUB : a subchannel code for steady-state and transient thermal-hydraulic analysis of BWR fuel pin bundles. Volume II. User's Manual

Guillebaud, Louis Jean Marie.; Levine, A.; Boyd, W.; Faya, A.; Wolf, Lothar
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 10033375 bytes; application/pdf
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The WOSUB-codes are spin-offs and extensions of the MATTEO- code [ 2 ]. The series of reports describe WOSUB-I and WOSUB-II in their respective status as of July 31, 1977. This report is the second of a series of three reports describing the WOSUB code. It gives a detailed description of the input data, flow charts, and output, and contains the list- ings of WOSUB-I and WOSUB-II. For the purpose of future ex- tensions parameters, common blocks and variables used in the code are listed in full detail. WOSUB-I and WOSUB-II are subchannel computer codes for the steady-state and transient analysis of the thermal-hydraulic characteristics of Boiling Water Reactor (BWR) fuel rod bundles. Both codes are also applicable'to analyze PR bundles, especially when these are ducted--a situation which most often arises in experimental set-ups. The main difference between WOSUB-I and WOSUB-II is that the former is designed to analyze small bundles, whereas the latter is capable to handle symmetric sections of today's large- sized BWR bundles. In addition, WOSUB-II does not contain all of the additions made in WOSUB-I yet, because it is deemed appropriate to introduce these into the smaller code first, before they are implemented into the bigger one. Both codes are still in the stage of evolutionary develop- ment. Thus...

‣ WOSUB : a subchannel code for steady-state and transient thermal-hydraulic analysis of BWR fuel pin bundles. Volume III. Assessment and Comparison

Wolf, Lothar; Levin, A.; Faya, A.; Boyd, W.; Guillebaud, Louis Jean Marie
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 10264331 bytes; application/pdf
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The WOSUB-codes are spin-offs and extensions of the MATTEO-code [1]. The series of three reports describe WOSUB-I and WOSUB-II in their respective status as of July 31, 1977. This report is the third in a series of three, the first of which [2] contains all the information about the models, solution methods and constitutive equations and the second [3] being the user's manual of the code. This report summarizes the assessment of the WOSUB- code against experiments and compares its results with the results of other subchannel codes. The following experiments are used for the purpose of the assessment of the code under steady-state conditions: 1) 9-rod GE-tests with radially uniform and non- uniform peaking factor patterns. 2) 16-rod Columbia tests with slight power tilts. 3) Planned 9-rod Swedish tests with very strong power tilts. 4) Actually performed 9-rod Swedish tests with power tilt. 5) 9-rod GE-CHF experiments. The comparison with these data shows that WOSUB is capable of predicting the lower-than-average behavior of the corner sub- channel and the higher-than-average behavior of the center subchannel for both quality and mass flux. None of the other well-known subchannel codes is indeed capable of specifically predicting the correct corner subchannel behavior. These codes seem to inherently suffer from major deficiencies associated with their incorporated mixing models. Therefore...

‣ Coarse-to-Fine Sequential Monte Carlo for Probabilistic Programs

Stuhlmüller, Andreas; Hawkins, Robert X. D.; Siddharth, N.; Goodman, Noah D.
Fonte: Universidade Cornell Publicador: Universidade Cornell
Tipo: Artigo de Revista Científica
Publicado em 09/09/2015 Português
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Many practical techniques for probabilistic inference require a sequence of distributions that interpolate between a tractable distribution and an intractable distribution of interest. Usually, the sequences used are simple, e.g., based on geometric averages between distributions. When models are expressed as probabilistic programs, the models themselves are highly structured objects that can be used to derive annealing sequences that are more sensitive to domain structure. We propose an algorithm for transforming probabilistic programs to coarse-to-fine programs which have the same marginal distribution as the original programs, but generate the data at increasing levels of detail, from coarse to fine. We apply this algorithm to an Ising model, its depth-from-disparity variation, and a factorial hidden Markov model. We show preliminary evidence that the use of coarse-to-fine models can make existing generic inference algorithms more efficient.