Structural and containment response to LMFBR accidents

The adequacy of the containment of fast reactors has been traditionally evaluated by analyzing the response of the containment to a spectrum of core disruptive accidents. The current approach in the U.S. is to consider fast reactor response to accidents in terms of four lines of assurance (LOAs). Thus, LOA-1 is to prevent accidents, LOA-2 is to limit core damage, LOA-3 is to control accident progression and LOA-4 is to attenuate radiological consequences. Thus, the programs on the adequacy of containment response fall into LOA-3. Significant programs to evaluate the response of the containment to core disruptive accidents and, thereby, to assure control of accident progression are in progress. These include evaluating the mechanical response of the primary system to core disruptive accidents and evaluating the thermal response of the reactor structures to core melting, including the effects this causes on the secondary containment. The analysis of structural response employs calculated pressure-volume-time loading functions. The results of the analyses establish the response of the containment to the prescribed loadings. The analysis of thermal response requires an assessment of the distribution and state of the fuel, fission products and activated materials from accident initiation to final disposition in a stable configuration

HEAI TRANSFER Two-Phase Frictional Pressure Drop Prediction From Levy's Momentum Model

This section consists of contributions of 750 words or less (about 2'/2 double-spaced typewritten pages, including figures). Technical Briefs will be reviewed and approved by the specific division's reviewing committee prior to publication. After approval such contributions will be published as soon as possible, normally in the next issue of the Journal Two The ratio f*D. is difficult to evaluate since the liquid phase Journal of Heat Transfer and the assumption is made that this can be generalized to = (1 -a) k De for the case where there are interfacial losses or some gas contacting the wall. For the case of large interfacial losses

Argonne National Laboratory Reports

Report issued by the Argonne National Laboratory discussing the Liquid Metal Fast Breeder Reactor Nuclear Safety Program. The work completed by the program between 1969 and 1970 is presented. This report includes tables, illustrations, and photographs

LMFBR safety testing needs and the conceptual design of a new safety research experiment facility

Experiment needs for the LMFBR safety program are reviewed. The screening of reactor concepts which would meet the needs is described and a conceptual design for a new safety research experiment facility is presented. (JWR

The effect of pressure on boiling density in multiple rectangular channels /

Includes bibliographical references.Mode of access: Internet

A preliminary design study of a boiling slurry reactor experiment /

"April 1960"Bibliography: p. 38-39.Operated by the University of Chicago underMode of access: Internet

Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs