[en] Full text: Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

[en] Full text: The PLATE fuel performance code, under development for several years, was originally designed for the thermal analysis of miniature U-Mo or U3Si2 experimental dispersion fuel plates. Recent code enhancements have extended the analysis capability of PLATE to full-size fuel plates. Additionally, the capability to analyze fuel plates using monolithic U-Mo fuels has been added. Data from the postirradiation examination of the XP-2 and XP-5 monolithic fuel plates from the RERTR-4 irradiation test have been used to develop an interaction rate correlation for U-Mo fuel and Al-6061 cladding; high-burnup data from several RERTR-4 U-Mo fuel plates have been used to refine the fuel swelling correlations employed by PLATE. These code enhancements are described. (author)

[en] Full text: Monolithic fuels are the most promising candidate for a next generation of high-density research reactor fuels. If successfully developed, the remaining HEU-fueled reactors in the world could presumably be converted to low-enriched fuel and the use of highly enriched uranium in the civilian nuclear fuel cycle eventually terminated. The most challenging type of reactors to convert are single element reactors because their core geometry is generally the least flexible. This specific reactor type is therefore the primary focus of this article. Based on new computational tools and optimization methods, neutronics calculations are presented to assess the potential of monolithic fuels for conversion of high-flux reactors in general and of single element reactors in particular. (author)

[en] Full text: The Argonne National Laboratory Reduced Enrichment for Research and Test Reactors Program is performing R and D supporting conversion of 99Mo production from high-enriched to low-enriched uranium targets. One of the major obstacles to conversion is the fivefold increase of the amount of uranium needed to produce an equivalent amount of 99Mo. The additional uranium would lead to an increase in the volume of liquid processed and the volume of liquid waste. The use of an efficient, high capacity sorbent would allow for small purification columns and minimum liquid volumes throughout the process. Thermoxid has developed an inorganic sorbent that meets these requirements. Our batch tests show that Thermoxid sorbents have much higher Kd(Mo) values than the commonly used alumina under a wide variety of conditions (20-340 g U/L, 0.5-1.5 M HNO3, 4, 24, and 48 hours) relevant to acid-side 99Mo production and recovery processes. The Kd(Mo) values for the Thermoxid sorbents are inversely proportional to both uranium concentration and acidity. Column tests were conducted to determine the sorbents' capacity for Mo at various uranium and acid concentrations. Overall, these new sorbents appear to have superior performance and would allow for smaller separation/purification columns than are possible using alumina as the sorbent. (author)

[en] Full text: A modified nitride-based uranium fuel to support the small, secured, transportable, and autonomous reactor (SSTAR) concept is initiated at Lawrence Livermore National laboratory (LLNL). This fuel and material research project centers on the evaluation and manufacturing of uranium nitride fuel imbedded with other inert (e.g. ZrN) or neutron-absorbing materials (e.g. HfN) to enhance the fuel properties to achieve long core life. This paper discusses how a modified nitride fuel with chemically-compatible inert additives (ZrN, HfN, etc) could be suitable as replacement fuel for research and test reactors. Mono-uranium nitride fuel pellet is manufactured at the LLNL. Existing facilities and equipment can be employed to fabricate modified uranium nitride fuel cladded in aluminum cladding. Preliminary fuel examination indicated that high uranium loading can be achieved in uranium nitride: at 80 theoretical density, 10.8 g/cc is uranium. Uranium nitride is also favorable in its thermal properties: the thermal conductivity of mono-nitride is compatible to that of silicate (∼25 W/mK), and its melting temperature is much higher than that of metal (2630 deg. C for UN vs. 1100 deg. C for U metal). Out-of-pile experiment is planned to examine the corrosion properties of uranium nitride fuel in water coolant. (author)

[en] Full text: In 2004 the research reactor FRM-II in Garching started operation using U₃Si₂ high-density fuel. March 2nd first criticality and in August full power was reached. During the start up measured nuclear data agree very well to the pre-calculated values. This paper presents some of these comparisons as there are the first critical condition of the reactor and the power density variation in the core for the fresh element. Excellent agreement of the measured and pre-calculated power density variation was found; this includes smaller variations induced by experimental installations. Till July 26th, in the start up phase 88 MWd were operated with a maximum power of 13MW; the precision of the power level so far was expected to be not more than 10%. The measured change in the control rod position for new criticality on July 26th was 3.05 cm; the post-calculated value for the different power steps was 3.18 cm; this corresponds to a renormalisation of the reactor power by 4%. The same result came from the more exact heating margin of the primary circuit measured at nearly full power in August. With this actual validation of the fuel burn up procedures, taken for the design of the FRM-II, the expected lifetime of more than 52 days was given new evidence. (author)

[en] Full text: The second irradiation fuel experiment, KOMO-2, for the qualification test of atomized U-Mo dispersion rod fuels with U-loadings of 4-4.5 gU/cc in KAERI was finished after irradiation until 70 at%U235 peak burn-up and is subjected to IMEF (Irradiation material Examination Facility) for post-irradiation analysis in order to understand fuel irradiation performance of U-Mo dispersion fuel. Current results on PIE of KOMO -2 revealed that the U-Mo/Al dispersion fuel rods exhibit sound performance without break-away swelling, but most of the fuel rods irradiated at high linear power show the extensive formation of the interaction phase between the U-Mo particle and Al matrix. In this paper, the analysis on the PIE results, focused on the diffusion related microstructures obtained from the optical and EPMA observations, will be presented in detail. And thermal modeling will be carried out to calculate the temperature of the fuel rod during irradiation. (author)

[en] Full text: A collaboration agreement between ANL/USDOE and CNEA Argentina in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual thermal FASTDART version and also a DART TM thermomechanical version were presented during RERTR 2002 and RERTR 2003 Meetings. During this past year the following activities were completed: - Optimization of DART THERMAL code Al diffusion parameters by testing predictions against reliable data from RERTR experiments. - Improvements on the 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic fuel Concerning the first point, by means of an optimization of parameters of the Al diffusion through the interaction product theoretical expression, a reasonable agreement between DART temperature calculation with reliable RERTR PIE data was reached. The 3-D thermomechanical code complex is based upon a finite element thermal-elastic code named TERMELAS, and irradiation behavior provided by the DART code. An adequate and progressive process of coupling calculations of both codes at each time step was reached. The coupling of the various components of the calculation was benchmarked and validated against RERTR PIE data. Various results will be shown during RERTR2004 meeting. (author)

[en] Full text: The RERTR program has focused thus far primarily on ending shipments of HEU fuel to research reactors. This has resulted in giving highest priority to reactors with steady thermal powers of 1 megawatt or more, because they require regular refuelling. Critical facilities and pulsed reactors can also of serious concern, because some of them contain very large amounts of barely-irradiated HEU and plutonium. They could be costly to convert - and conversion to LEU may be impractical for fast-neutron critical assemblies. An assessment should be carried out first, therefore, as to which are still needed. Critical assemblies are required today primarily to benchmark Monte Carlo neutron-transport codes. Perhaps the world nuclear community could share a few instead of each reactor-design institute having its own. There is also a whole universe of HEU-fuelled pressurized-water reactors used to power submarines and other types of nuclear-powered ships. These reactors collectively require much more HEU fuel each year than research reactors. The risk of HEU diversion from their fuel cycles is not zero but it is difficult for outsiders to discuss conversion because of the fuel designs are classified. This makes the conversion of Russia's civilian icebreaker reactors of particular interest because issues of classified fuel design are less problematic and these reactors load annually fuel containing about 400 kg of U-235. Another reason for interest in developing LEU fuel for these reactors is that the KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant. Finally, the research-reactor community is, in any case, faced with developing fuels that can operate at power-reactor-fuel temperatures because there are a few high-powered research reactors that operate in this temperature range. (author)

[en] Full text: A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU U2SO4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)