2,464 research outputs found
Fault tree and reliability analysis
Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977Includes bibliographical references (p. 311-312
Sensitivity study of the assembly averaged thermal-hydraulic models of the MEKIN computer code in power transients / by Thomas Rodack [and] Lothar Wolf
Cover title"August 1977."Also issued as a Nucl. E. and M.S. thesis by the first author and supervised by the second author, MIT Depts. of Nuclear and Mechanical Engineering, 1977Includes bibliographical references (pages 242-246)The thermal-hydraulic (T-H) models and solution schemes employed by the MEKIN computer code have been examined. The effects of T-H input parameters on- predicted fuel temperatures and coolant densities were determined in transient analyses. Consideration was limited primarily to a simulated PWR control rod ejection transient. Limitations to the use of MEKIN that arise because of simplifying assumptions in the T-H models are discussed. Computation time may be reduced without altering the results of a transient analysis if appropriate MEKIN options are selected. Guidelines are presented to facilitate the selection of these options. Suggestions for improvement of the code are also ma:de
Spin waves cause non-linear friction
Energy dissipation is studied for a hard magnetic tip that scans a soft
magnetic substrate. The dynamics of the atomic moments are simulated by solving
the Landau-Lifshitz-Gilbert (LLG) equation numerically. The local energy
currents are analysed for the case of a Heisenberg spin chain taken as
substrate. This leads to an explanation for the velocity dependence of the
friction force: The non-linear contribution for high velocities can be
attributed to a spin wave front pushed by the tip along the substrate.Comment: 5 pages, 9 figure
Spin waves cause non-linear friction
Energy dissipation is studied for a hard magnetic tip that scans a soft
magnetic substrate. The dynamics of the atomic moments are simulated by solving
the Landau-Lifshitz-Gilbert (LLG) equation numerically. The local energy
currents are analysed for the case of a Heisenberg spin chain taken as
substrate. This leads to an explanation for the velocity dependence of the
friction force: The non-linear contribution for high velocities can be
attributed to a spin wave front pushed by the tip along the substrate.Comment: 5 pages, 9 figure
PL-MODT and PL-MODMC : two codes for reliability and availability analysis of complex technical systems using the fault tree modularization technique
"November 1978."Includes bibliographical referencesThe methodology used in the PL-MOD code has been extended to include the time-dependent behavior of the fault tree components. Four classes of components are defined to model time-dependent fault tree leaves. Mathematical simplifications are applied to predict the time-dependent behavior of simple modules in the fault tree from its input components' failure data. The extended code, PL-MODT, handles time-dependent problems based on the mathematical models that have been established. An automatic tree reduction feature is also incorporated into this code. This reduction is based on the Vesely-Fussell importance measure that the code calculates. A CUT-OFF value is defined and incorporated into the code. Any module or component in the fault tree whose V-F importance is less than this value will automatically be eliminated from the tree. In order to benchmark the PL-MODT code, a number of systems are analyzed. The results are in good agreement with other codes, such as FRANTIC and KITT. The computation times are comparable and in most of the cases are even lower for the PL-MODT code compared to the others. In addition, a Monte-Carlo simulation code (PL-MODMC) is developed to propagate uncertainties in the failure rates of the components to the top event of a fault tree. An efficient sorting routine similar to the one used in the LIMITS code is employed in the PL-MODMC code. Upon modularization the code proceeds and propagates uncertainties in the failure rates through the tree. Large fault trees such as the LPRS fault tree as well as some smaller ones have been analyzed for simulation, and the results for the LPRS are in fair agreement with the WASH-1400 predictions for the number of simulations performed. The codes PL-MODT and PL-MODMC are written in PL/l language which offers the extensive use of the list processing tools. First experience indicates that these codes are very efficient and accurate, specifically for the analysis of very large and complex fault treesSponsored by the NR
Improved multidimensional numerical methods for the steady state and transient thermal-hydraulic analysis of fuel pin bundles and nuclear reactor cores
Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1977Includes bibliographical reference
LWR core thermal-hydraulic analysis : assessment and comparison of the range of applicability of the codes COBRA IIIC/MIT and COBRA IV-I
Based on the M.S. thesis of the first author in the M.I.T. Dept. of Nuclear Engineering, 1978.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, if the heat transfer package is corrected, then COBRA IIIC/MIT
should be just as applicable as the implicit option of COBRA IV-I.Final report for research project sponsored by Long Island Lighting Company and others under the MIT Energy Laboratory Electric Utility Program
WOSUB : a subchannel code for steady-state and transient thermal-hydraulic analysis of BWR fuel pin bundles.
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, the review reflects the achieve-
ments as of July 1977.Topical report for Task 3 of the Nuclear Power Reactor Safety Research Program sponsored by New England Electric System, Northeast Utilities Service Co. under the M.I.T. Energy Laboratory Electric Power Program
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