RERTR program progress in qualifying reduced-enrichment fuels

In order to provide the technical means for reducing the enrichment of uranium used to fuel research and test reactors, the US Reduced Enrichment Research and Test Reactor (RERTR) Program has been engaged in the development and testing of higher-uranium-density fuels than had been used previously. The main characteristics to be considered in evaluating the performance of a fuel are its swelling, its blister-threshold temperature, and its metallurgical appearance. Data for the qualification of the reduced-enrichment fuels being developed by the RERTR Program are obtained from examination of minature fuel plates (miniplates) which successfully pass the irradiation screening tests and from examinations of full-sized fuel elements. This paper will summarize the miniplate data and will give the status of full-sized element irradiations. Finally, the current status of qualification of the various fuel types will be discussed and some projections of the future will be given

Development of very high-density low-enriched uranium fuels

The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm{sup 3}, based on the use of {gamma}-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

Development of very-high-density fuels by the RERTR program

The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm{sup 3}, based on the use of {gamma}-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

Development of very-high-density low-enriched-uranium fuels

The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm{sup 3}, based on the use of {gamma}-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

Overview of reduced enrichment fuels: Development, testing, and specification

The US Reduced Enrichment Research and Test Reactor (RERTR) Program was established in 1978 to provide the technical means to operate research and test reactors with low enrichment uranium (LEU) fuels without significant penalty in experiment performance, operation costs, component modifications, or safety characteristics. This paper discusses relevant developments in fuel developments. 9 refs., 1 tab

A review of the U sub 3 O sub 8 -aluminum reaction as a potential heat source in research and test reactor accidents

A critical review of the literature on the U{sub 3}O{sub 8}-aluminum reaction has been conducted. The reaction in fabricated fuel plates is found to be less energetic and much slower than in cold-pressed powder mixtures. The difference is at least partially attributable to conversion of up to 50% of the U{sub 3}O{sub 8} to U{sub 4}O{sub 9} during fabrication. No definitive measurements of the amount and rate of energy release have been made. Data are provided upon which to base calculations of energy release

Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations

Reduced-reactivity-swing LEU fuel cycle analyses for HFR Petten

The primary objective of these low enriched uranium (LEU) fuel cycle analyses was to effect at least a 33% reduction in the reactivity swing now experienced in the high enriched uranium (HEU) cycle while minimizing increases in /sup 235/U loading and power peaking. All LEU equilibrium fuel cycle calculations were performed using either a 19- or 20-plate fuel element with 0.76-mm-thick meat and 0.5- or 0.6-mm-thick Cd wires as burnable absorbers and 16- or 17-plate control rod fuel followers with 0.76-mm-thick meat. Burnup-dependent microscopic cross sections were used for all heavy metals and fission products. A three-dimensional model was used to account for the effect of partially inserted control rods upon burnup profiles of fuel and of burnable absorbers and upon power peaking. The equilibrium cycle reactivity swing (or, equivalently control rod movement) was reduced by 50% using LEU fuel with U meat densities <4.8 Mg/m/sup 3/. 6 refs., 4 tabs

Comparison of calculated quantities with measured quantities for the LEU-fueled Ford Nuclear Reactor

The Ford Nuclear Reactor (FNR) went critical on December 8, 1981 with 23 LEU fuel elements. Five of these 23 elements were fabricated by CERCA and the others by NUKEM. Since that time a substantial data base of experimental results for LEU cores has been accumulated by the University of Michigan FNR staff. This paper compares some of the experimental data with analytical calculations based, for the most part, on three-dimensional diffusion theory. The critical configuration, control rod worths, axial rhodium reaction rate profiles and thermal flux distributions have been calculated and compared with measurements

The ORR Whole-Core LEU Fuel Demonstration

The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged {sup 235}U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of {sup 235}U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs