NUMERICAL RESULTS FOR EGCR MODERATOR-ELEMENT STRESS PROBLEMS

A detailed presentation is made of the thermal stresses calculated for the moderater elements in the Experimental Gas-Cooled Reactor. These results are discussed and some conclusions are presented. This report complements a previous report, BMI-1503, which defines the problems and discusses the methods of solution. (auth

NUMERICAL RESULTS FOR EGCR MODERATOR-ELEMENT STRESS PROBLEMS

A detailed presentation is made of the thermal stresses calculated for the moderater elements in the Experimental Gas-Cooled Reactor. These results are discussed and some conclusions are presented. This report complements a previous report, BMI-1503, which defines the problems and discusses the methods of solution. (auth

Analysis of interactions of mechanical deformations and mass transfer on heat transfer from an underground nuclear-waste repository. Final report

A review of existing models identified several effects that may need consideration in further model development. Most of these effects involved coupling equations through variable property values rather than through omission of any significant mechanism. However, it was also shown that more than one mechanism may adequately simulate a given set of experimental data and additional experimental data are needed to establish which (if any) of the possible mechanisms would actually control conditions in a nuclear waste repository. In particular, it is believed that mathematical modeling of major thermomechanical effects can be accomplished with finite element analysis computer programs, provided that adequate thermomechanical property data of salt and overburden are attained. An attempt was made to develop a general set of differential equations for simulating momentum, mass, and energy flows in geologic formations in order to illustrate the possible mechanisms and point out those included and not included in existing models. Most of the mechanisms are included in some manner in existing models although some approximations may not be adequate. More experimental data are required to assess the importance of most omitted mechanisms. Analysis of some data on brine migration in salt indicated that two mechanisms, acting simultaneously, could adequately explain the flow. These are Darcy flow and a combination of ordinary and thermal diffusion enhanced by temperature-dependent solubility. Equations based on this simultaneous action correlated the data very well and indicated the possible need to include both (and, maybe other) mechanisms in future models. A program is recommended for further study of brine mobility. An expected result of this program includes recommendations for further experimental work

NUMERICAL SOLUTION OF REACTOR STRESS PROBLEMS

Generalized computer codes were devised for solving stress problems of some complexity. These codes were applied to stress problsms relating to the graphite moderator elements in the Experimental Gas-cooled Reactor. The stress relief obtained by aubdividing the moderator elements was evaluated. The distontion and bending moments of the elements were also determined. (auth

CORE-TEMPERATURE EXCURSIONS FOLLOWING A PIPING FAILURE IN THE PLUTONIUM RECYCLE TEST REACTOR

An evaluation of the temperature excursion and its possible consequences arising from loss of coolant from the Plutonium Recycle Test Reactor (PRTR) was made for four different postulated ruptures in the primary heavy water coolant system. As a basis for the evaluation, a series of computations was made. These were based on incremental heat and mass balances for sections of Zircaloy-clad UO/ sub 2/and Pu- Al fuel elements. Solutions to each problem defined by the postulated break size and its location were defined by finitedifference approximatioms performed by an IBM 653 machine digital computer. The four postulated ruptures were: (1) a complete parting of the 14-in.-diameter outlet pipe near the upper ring header so that coolant would be lost from both broken ends; (2) a rapture equivalent to a 14-in.-diameter hole in the primaryloop piping adjacent to the upper ring header; (1) a complete parting of a 1 3/4-in. upper jumper; and (4) a complete parting of a 1 3/4-in. bottom jumper. The Pu-Al elements represent the most critical component; melting of these elements would begin about 219 seconds after the rapture occurred if emergency backup light water coolant were not available to the system. It was found that the injection of 750 gallons per minute (gpm) of emergency coolant (375 gpm to each ring header) would be adequate to prevent melting or failure of any reactor component for all cases studied even if injection did not begin until 2 or 3 min alter the rupture occurred. An earlier injection time would, of course, be beneficial. (auth