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