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‣ Approximate methods for obtaining a one-group nodal solution with two-group parameters

Hagemeier, Bruce William
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 143 leaves; 3644609 bytes; 3644415 bytes; application/pdf; application/pdf
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by Bruce William Hagemeier.; Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1982.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE; Includes bibliographical references.

‣ 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.

‣ Development of a two-fluid, two-phase model for light water reactor subchannel analysis

Kelly, J. E
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 285 leaves; 12907951 bytes; 12907707 bytes; application/pdf; application/pdf
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by John Edward Kelly.; Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1980.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE; Includes bibliographical references.

‣ Optimization of the axial power shape in pressurized water reactors

Melik, M. A
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 114 leaves; 5789201 bytes; 5788959 bytes; application/pdf; application/pdf
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by Mushtaq Ahmad Malik.; Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1982.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE; Bibliography: leaves 112-114.

‣ Reactor thermal-hydraulic analysis improvement and application of the code COBRA-IIIC/MIT

Loomis, James North
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 2 v. ([572] leaves); 22797335 bytes; 22797093 bytes; application/pdf; application/pdf
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by James North Loomis.; Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; 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.

‣ LWR core thermal-hydraulic analysis : assessment and comparison of the range of applicability of the codes COBRA IIIC/MIT and COBRA IV-I

Kelly, J. E.; Loomis, James N.; Wolf, Lothar
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 7751688 bytes; application/pdf
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This report summarizes the result of studies concerning the range of applicability of two subchannel codes for a variety of thermal-hydraulic analyses. The subchannel codes used include COBRA IIIC/MIT and the newly developed code, COBRA IV-I which is considered the benchmark code for the purpose of this report. Hence, through the comparisons of the two codes, the applicability of COBRA IIIC/MIT is assessed with respect to COBRA IV-I. A variety of LWR thermal-hydraulic analyses are examined. Results of both codes for steady-state and transient analyses are compared. The types of analysis include BWR bundle-wide analysis, a simulated rod ejection and loss of flow transients for a PWR. The system parameters were changed drastically to reach extreme coolant conditions, thereby establishing upper limits. In addition to these cases, both codes are compared to experimental data including measured coolant exit temperatures in a core, interbundle mixing for inlet flow upset cases and two-subchannel flow blockage measurements. The comparisons showed that, overall, COBRA IIIC/MIT predicts most thermal-hydraulic parameters quite satisfactorily. However, the clad temperature predictions differ from those calculated by COBRA IV-I and appear to be in error. These incorrect predictions are caused by the discontinuity in the heat transfer coefficient at the start of boiling. Hence...

‣ 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 and testing of three dimensional, two-fluid code THERMIT for LWR core and subchannel applications

Kelly, John Edward; Kazimi, Mujid S.
Fonte: MIT Energy Laboratory Publicador: MIT Energy Laboratory
Tipo: Relatório Formato: 5117191 bytes; application/pdf
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At head of title: Energy Laboratory and Dept. of Nuclear Engineering.; 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."

‣ Application of nodal equivalence theory to the neutronic analysis of PWRS

Hoxie, Christopher Lloyd
Fonte: Massachusetts Institute of Technology Publicador: Massachusetts Institute of Technology
Tipo: Tese de Doutorado Formato: 2 v. ([516] leaves)
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by Christopher Lloyd Hoxie.; Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1982.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; Includes bibliographical references.

‣ Recovery Process of Actinides from Genuine Spent Nuclear Fuel using TODGA and BTBP Extractants - JRC-ITU-TN-2008/70

MAGNUSSON Daniel
Fonte: European Commission - Joint Research Centre - Institute for Transuranium Elements Publicador: European Commission - Joint Research Centre - Institute for Transuranium Elements
Tipo: PhD Theses Formato: Printed
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During the last decades a growing concern about greenhouse gas emissions from fossil fuels has arisen. Nuclear power is an energy source with a low contribution to the greenhouse effect and is in this sense therefore a better alternative for electricity production. The waste from nuclear power is however highly radiotoxic and has to be stored for more than 100000 years until the radiotoxicity has decreased to an acceptable level. With partitioning and transmutation the purpose is to separate the nuclides that contribute most to the long-term radiotoxicity and to transform them into short lived or stable nuclides. If partitioning and transmutation are successfully applied the storage time for nuclear waste can be reduced to less than 1000 years. Essential for the partitioning and transmutation is an efficient separation of the minor actinides. In this work, group separation of the trivalent lanthanides and actinides from a PUREX raffinate has been demonstrated by means of solvent extraction, using the TODGA extracting agent. Excellent separation was obtained and the recoveries of the actinides exceeded 99.9%. Separation of the actinides from the lanthanides was also carried out, using the CyMe4-BTBP extracting agent. Results achieving high actinide recovery of more than 99.9%...

‣ 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...