A program is in progress at Battelle-Columbus which is aimed at developing a basic understanding of the mechanisms and processes which contribute to the phenomenology of swelling and fission gas release in nuclear fuel materials during irradiation. As part of this program, neutron radiography has been used to obtain data on the dimensional changes of fuel specimens periodically during irradiation. This data is used to develop plots of fuel swelling as a function of burnup for the specimens under irradiation. These data show a decided nonlinearity of swelling with accumulated burnup for many specimens, and it has been postulated that the shape of the curves is due, at least in part, to the gas release characteristics of the specimens. Accordingly, it was considered advantageous to develop a system to continuously monitor fission gas release from the fuel during irradiation. The system developed consists of a null-balance pressure transducer attached directly to a clad fuel specimen in the capsule. Fission gases escaping from the fuel collect within the cladding and exert pressure on a diaphragm in the transducer. A balancing pressure is introduced on the other side of the diaphragm and the null-balance position of the diaphragm is detected by an electrical contact which is part of a detection circuit. Instrumentation is provided to monitor the balance pressure, and the fission gas temperature and pressure is correlated with the known fission gas generation rate to calculate the fractional release from the fuel. In addition to monitoring fission gas release, the system also provides information on cladding integrity. Cladding failure is detected as a large reduction in pressure while operating at essentially constant temperature. This system has been used successfully in two irradiation experiments at fuel and cladding temperatures in excess of 1700 C. The data it provides on fission gas release has been used in conjunction with swelling data from neutron radiography to provide considerable insight into the phenomenology of the behavior of nuclear fuels during irradiation.

The AGR-1 experiment is a fueled multiple-capsule irradiation experiment that was irradiated in the Advanced Test Reactor (ATR) from December 26, 2006 until November 6, 2009 in support of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Fuel Development and Qualification program. An important measure of the fuel performance is the quantification of the fission product releases over the duration of the experiment. To provide this data for the inert fission gasses(Kr and Xe), a fission product monitoring system (FPMS) was developed and implemented to monitor the individual capsule effluents for the radioactive species. The FPMS continuously measured the concentrations of various krypton and xenon isotopes in the sweep gas from each AGR-1 capsule to provide an indicator of fuel irradiation performance. Spectrometer systems quantified the concentrations of Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe 135, Xe 135m, Xe-137, Xe-138, and Xe-139 accumulated over repeated eight hour counting intervals.-. To determine initial fuel quality and fuel performance, release activity for each isotope of interest was derived from FPMS measurements and paired with a calculation of the corresponding isotopic production or birthrate. The release activities and birthrates were combined to determine Release-to-Birth ratios for the selected nuclides. R/B values provide indicators of initial fuel quality and fuel performance during irradiation. This paper presents a brief summary of the FPMS, the release to birth ratio data for the AGR-1 experiment and preliminary comparisons of AGR-1 experimental fuels data to fission gas release models.

A program has been conducted and is continuing at Battelle Memorial Inst. under the sponsorship of the United States Atomic Energy Commission to study the release of fission gas from UO2 and to relate the release to material properties, both chemical and structural. Work in the program has involved the preparation and characterization of sintered polycrystalline and fused single-crystal UO2 samples, the construction of a beam-tube loop facility for the irradiation of the UO2 under controlled temperature and atmosphere, and the collection and analysis of released fission gases during irradiation. Irradiation experiments have been conducted in purified helium-hydrogen atmosphere at temperatures from 500 to 1500 F, with fission gases continuously collected and analyzed for krypton and xenon during irradiation of natural uranium specimens. Irradiation exposures of the specimens have been in the region of 1018 nvt.