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

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

Transient response of a single heated channel

Includes bibliographical references (page 20

Modeling of corium/concrete interaction

Includes bibliographical references (pages 232-237)Work supported by Electric Power Research Institut

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

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

Application of probabilistic consequence analysis to the assessment of potential radiological hazards of fusion reactors

"July 1978."Originally presented as the first author's thesis, (M.S.)--in the M.I.T. Dept. of Nuclear Engineering, 1978Includes bibliographical references (pages 87-89)A methodology has been developed to provide system reliability criteria based on an assessment of the potential radiological hazards associated with a fusion reactor design and on hazard constraints which prevent fusion reactors from being more hazardous than light water reactors. The probabilistic consequence analyses, to determine the results of radioactivity releases, employed the consequence model developed to assess the risks associated with light water reactors for the Reactor Safety Study. The calculational model was modified to handle the isotopes induced in the structural materials of two conceptual Tokamak reactor designs, UWMAK-I and UWMAK-III. Volatile oxidation of the first wall during a lithium fire appears to be a primary means of disrupting induced activity, and the molybdenum alloy, TZM (UWMAK-III), tends to be more susceptible than 316 stainless steel (UWMAK-I) to mobilization by this mechanism. It was determined that the radiological!  hazards associated with induced activity in these reactor designs imply reliability requirements comparable to those estimated for light water reactors. The consequences of estimated maximum possible releases of induced activity, however, are substantially less than the maximum light water reactor accident consequences.Report issued under contract, U.S. Dept. of Energy EY-76-02-243

Forced convection degraded core cooling in light water reactors

Includes bibliographical references (pages 135-141)Work performed with support from E.G. & Idaho, In

Development of models for the two-dimensional, two-fluid code for sodium boiling NATOF-2D

Several features were incorporated into NATOF-2D, a twodimensional, two fluid code developed at M.I.T. for the purpose of analysis of sodium boiling transients under LMFBR conditions. They include improved interfacial mass, momentum and energy exchange rate models, and a cell-to-cell radial heat conduction mechanism which was calibrated by simulation of Westinghouse Blanket Heat Transfer Test Program Runs 544 and 545. Finally, a direct method of pressure field solution was implemented into NATOF-2D, replacing the iterative technique previously available, and resulted in substantially reduced computational costs.The models incorporated into NATOF-2D were tested by running the code to simulate the results of the THORS Bundle 6A Experiments performed at Oak Ridge National Laboratory, and four tests from the W-1 SLSF Experiment performed by the Hanford Engineering Development Laboratory. The results demonstrate the increased accuracy provided by the inclusion of these effects

Interfacial exchange relations for two-fluid vapor-liquid flow : a simplified regime map approach

A simplified approach is described for selection of the constitutive relations for the inter-phase exchange terms in the two-fluid code, THERMIT. The approach used distinguishes between pre-CHF and post-CHF conditions. Interfacial mass, energy and momentum exchange terms are selected and tested against one dimensional measurements for a wide range of mass flow rate, pressure and void fraction conditions. It is concluded that the simplified regime map approach leads to accurate predictions for LWR applications, excluding depressurization events

Analysis of design strategies for mitigating the consequences of lithium fire within containment of controlled thermonuclear reactors

Originally presented as the first author's thesis, (M.S.)--in the M.I.T. Dept. of Nuclear Engineering, 1978Includes bibliographical references (pages 117-121)Report issued under U.S. Dept. of Energy EY-76-02-243

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

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.