J.P. Regis

Fast-Neutron Hodoscope at TREAT: Data Processing, Analysis, and Results

The fast-neutron hodoscope at the Transient Reactor Test Facility is designed for the determination of fuel motion during the course of brief (0.1- to 30-sec) power transients. During the course of a transient test, data must be recorded from each of 334 hodoscope channels at count rates up to 2 million/sec each, down to millisecond time intervals. This is accomplished in a relatively reliable and inexpensive manner by displaying counts from each detector sequentially in binary code on a lamp panel, which is photographed by a high-speed framing camera, producing a film record of the transient test. After chemical development, the film is examined by a computer-controlled flying-spot scanner, and the position and density of candidate lamp images are recorded on magnetic tape. Through further computer processing, these images are sorted and decoded, and the count rate is recovered for each detector at each instant of collection time. A cathode-ray tube and a plotter, both computer controlled, are used to recreate and analyze the fuel motion history of the experiment. Analysis is directed toward fuel distortion or expansion prior to clad failure, slumping, dispersion, amount and rates of movement, post-scram relocation, and ultimate disposition of fuel.

Treat sodium loop experiments on performance of unbonded, unirradiated EBR-II Mark I fuel elements

Abstract Description of sodium loop experiments performed in the Transient Reactor Test (TREAT) Facility specifically to provide guidance for analyses on the behavior of an element which either has no sodium thermal bond between fuel and cladding or which, due to some unspecified defect, loses the bond upon reactor startup. Transient heat transfer calculations of the experiments were performed utilizing an idealized model of meltdown in which fuel slumping against the cladding was assumed to occur instantaneously along the length of the pin when the fuel surface at the axial midplane reached 1050°C. The result of TREAT experiments and of the calculations imply that although fuel melting may accompany a loss-of-bond type failure in an EBR-II driver fuel element, the short term consequences are acceptable from the viewpoint of reactor safety. Comparison of results from calculations and experiment indicate that the relatively simple modelling used represents a conservative description of the temperature rise in the cladding following fuel-cladding contact resulting from the fuel melting.

Power pulse meltdown experiments performed in the mark I TREAT sodium loop on clusters of 7 EBR-II Mark I type pins

Abstract Two experiments were performed on fuel failure and post-failure movements of fuel and coolant under power excursions. Test samples consisted of seven-pin clusters in instrumented integral sodium loops inserted into the TREAT reactor. One cluster was run with sodium flow and the other with pump power off (transient-induced flow could occur). Results agreed with the predictions of a simple calculational model of pin failure which had been tested previously in less elaborate experiments. Extensive post-failure movements of molten fuel occurred. Although sharp pressure spikes were recorded their magnitude is within a range that can be accommodated by reasonable engineering designs.