The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740/sup 0/C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000/sup 0/C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th-/sup 233/U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized.

Progress is reported in the following areas: PCRV development, studies on structural materials, fission product technology studies, kernel migration and irradiated fuel chemistry, coolant chemistry (steam-graphite reactions), fuel qualification, and characterization and standardization of graphite.

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)

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)

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)

From international conference on nuclear fuel performance; London, UK (15 Oct 1973). Fuel for large HTGR power stations in the United States consists of coated particles of ThO/sub 2/ as fertile material mixed with smaller coated particles containing fissile material, either /sup 233/U from recycle of bred fuel or /sup 233/U as makeup fuel. Recycle fuel particles containing /sup 233/U mixed with thorium in sol-- gel oxide microspheres have been prepared and experiments in progress will test a fuel compositon consisting of Th/U = 4 to full exposure and burnup. The use of weak-acid ion exchange resin particles as precursors for either type of fissile fuel kernel offers several advantages.in processing and properties. Pilot-scale preparation and testing of resin-base fuels demonstrated that their performance is adequate and indicate that such fuels may be capable of operating at temperatures well above 1300 deg C. Such high-performance fuels are desirable for process heat and direct-cycle applications. (auth)

in the analysis. Particle failure probability is calculated with an extention of Weibull's analysis to the pyrocarbon coating.

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.

根据高温气冷堆燃料元件基体石墨对天然石墨粉体的要求,开展了浮选、高温-化学纯化、颗粒形貌修饰-分级、均匀化等方面的工业化制备技术研究,成功研制出可受控制的、品质均一的、经工艺性能检验符合技术指标要求的核级天然鳞片石墨粉体,并对影响天然鳞片石墨粉体性能较大的制备工艺技术进行了分析探讨.