The important metallurgical problems relevant to circulating fuel reactors utilizing fused fluoride systems are discussed. The successful utilization of such mixtures is in large measure dependent upon the development of reactor materials which can both contain the mixtures under reactor conditions and afford usefuly structural properties. Paramount among the materials requirements whicth must be fulfilled are the following: resistance to corrosion by the fluoride fuel or salt (in the presence of a neutron flux), resistance to oxidation by air, and good elevated-temperature strength. Materials must also be easily formed and welded into relatively complicated shapes and be metallurgically stable over a wide temperature range. The most stringent materials requirement is adequate corrosion resistance to the molten fluorides which are used as fuely carriers. Generally speaking, corrosion problems are solved by either formation of a protective film through action of the corrosive environment or the establishment of thermdynamic equilibrium between the material and corrosive environment before significant corrosion occurs. The first approach has been used quite successfully with metals exposed to gas or high- temperature water bnt with only limited success with liquid metals and molten salts. For the latter, especially molten fluorides which are excellent fluxing agents, the second approach,more » i.e., thermodynamic equilibrium, is preferred. (auth)« less
235$U irradiation performance could be seen from this experiment. The migration rates for the mixed oxide kernels measured for the HRB-6 specimens were consistent with such measurements made on similar kernels in other experiments. Analysis of the entire body of data on thermal migration for mixed oxide fissile kernels later led to the conclusion that this fuel was a marginal performer, and ultimately was replaced as the reference fuel by a uranium oxycarbide fissile kernel loaded from ion exchange resins. The observation of identical performance for $sup 233$U and $sup 235$U in the mixed oxide system was very important to the fuel development program. The recycle fuel development efforts have continued to use $sup 235$U in subsequent irradiation tests, at considerably less expense than if $sup 233$U test specimens were used. The very good performance of the test specimens fabricated using extrusion, relative to those fabricated using slug injection, was noted again in HRB-6. (auth)
Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.
Specifications for the fuel assemblies Experimental Gas Cooled Reactor (EGCR) Core-I were developed for use in procuring the first core loading. A fuel assembly for the EGCR consists of a cluster of seven cylindrical fuel elements spaced and supported within a graphite sleeve. Each fuel element consists of a stainless-steel tube containing a column of hollow UO/sub 2/ pellets and having a spacer brazed at the midsection to control the spacing between fuel elements in a cluster. A master specification for the fuel assembly, a supplementary specification for each of the components, and a specification on record keeping daring manufacture are included. (auth)
similar particles irradiated at temperatures approximately equal to 1550/sup 0/C and burnups approximately equal to 16 percent FIMA. Failures due to permeability were not detectable by visual inspection but required a more extensive investigation by the 1000/sup 0/C gaseous chlorine leaching technique. Maximum particle surface operating temperatures were found to be approximately 300/sup 0/C in excess of design limits of 900/sup 0/C (low-temperature magazines) and 1250/sup 0/C (high-temperature magazines). The extremes of high temperatures and fast neutron fluences up to 1.6 x 10/sup 22/ neutrons/cm/sup 2/ produced severe degradation and swelling of the Poco graphite magazines and sample holders.
Since the beginning of operations at the Oak Ridge National Laboratory (ORNL) in 1943, shallow land burial has been used for the disposal of solid low-level radioactive waste. These wastes have originated from nearly every operating facility, and from 1955 to 1963, ORNL's solid waste storage areas were designated by the Atomic Energy Commission (AEC) as the Southern Regional Burial Ground. During this period, about one million cubic feet of solid waste from various off-site installations were buried in solid waste storage areas (SWSAs) 4 and 5. Six SWSAs have been used since land burial operations began at ORNL in early 1944. ORNL has generated liquid radioactive waste since the separation of plutonium began in 1944. The majority of these wastes are classified as process (low-level) waste and are derived from evaporator condensate and cooling water from process vessels, and from building drains and surface drainage from contaminated areas. Process wastes are monitored at sampling stations located strategicially throughout the plant, and for nearly 15 years (1944 to 1957) they were discharged directly into White Oak Creek without being treated chemically to remove radionuclides. A smaller quantity of intermediate-level wastes (ILW) originate from the radiochemical separation process and from test reactors. The collection, treatment, and methods of disposal of ILW from the years 1943 to 1981 are described. Over this period of time there was a great deal of variation in the amounts and types of radioactive liquid wastes generated.
