Vented Inverted Fuel Assembly Design for an SFR
Author(s)Vitillo, Francesco; Todreas, Neil E.; Driscoll, Michael J.
Advanced Nuclear Power Technology Program (Massachusetts Institute of Technology)
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The goal of this work is to investigate the feasibility of a vented inverted fuel assembly for a sodium-cooled fast reactor. The inverted geometry has been previously investigated for application in Gas-cooled Fast Reactors since it improves thermal-hydraulic and neutronic performance of those reactors. Venting is a concept studied during the past and its major past application in sodiumcooled fast reactors was in the Dounreay Fast Reactor in the United Kingdom. In this work the inverted assembly approach was adopted because it allows high fuel volume fraction, reduction of the coolant void reactivity, less neutron leakage, the reduction of the enrichment and lower pressure drop for the same channel length because grids nor wire wraps are no longer necessary. However all results of this work apply also to venting of conventional fuel pins. Performance criteria for vented fuel assemblies in term of materials, thermalhydraulics and venting systems have been investigated in order to set design goals. In particular, for the materials, a limit for maximum cladding surface temperature, cladding and other core internal structure fluence and maximum fuel temperature in the hot channel has been identified. For the thermal-hydraulic analysis, the goals are increasing fuel volume fraction, keeping the fuel and the cladding surface temperature as low as possible compared with those of a similar power rating core and minimizing core pressure drops. Regarding the venting system the design goals are retaining as much 137Cs in an upper plenum and keeping the overall assembly height within the values of current technology for a reactor of similar size. Therefore the height of an upper plenum (which must contain sodium bond volume expelled due to fuel thermal expansion, sodium bond volume due to its thermal expansion and the cesium volume of a single assembly if the cesium is completely released into the plenum) has been determined. Investigation of physical and chemical behavior of volatile fission products in sodium is presented, in order to determine the maximum activity inventory which would eventually be released into the primary sodium. Assumptions for the simplified approach adopted are discussed. Results of this analysis show that the most troublesome radionuclides in terms of propensity to escape from the venting system (due to their half-life being longer than a threshold time chosen based on physical behavior of escaping fission products: bubbling out for gases and pure diffusion for other volatile elements) are noble gases (85Kr and 133Xe), cesium (134Cs and 137Cs) and tritium (3H). For the thermal-hydraulic analysis a comparison between a pin-type fuel assembly and three inverted fuel assemblies with different parameters has been made, in order to demonstrate benefits of such a concept and to determine the best configuration. In particular attention is on core pressure drop, fuel and cladding temperature given the mass flow rate and assembly power. The results show that the best configuration has the same core pressure drop and hence pumping power and the same total active fuel length of a similar performance pin-type core. A final vented inverted fuel assembly design is proposed, which meets all the design goals. Such a configuration lets volatile radionuclides with short enough half-lives completely decay before release or be released in a negligible quantity after an infinite time of diffusion in sodium. Longer lived fission products will be released into the coolant, while fission gases will be vented first into the sodium and eventually to the cover gas after bubbling up through the sodium itself. Methods for purifying cover gas and coolant from vented radionuclides are proposed as well as storage systems for radioactive materials from the purification process. Results show that charcoal is the best absorber for noble gases whereas cold traps can be usefully used to remove cesium and tritium from primary sodium. Noble gases are produced in a (conservatively estimated) quantity of 38 m3/year (at STP) at core end-of-life and can be stored in adsorbent packed cylinders. Materials in cold traps are chemically treated to obtain liquid waste. Hence they can be converted into a solid and then stored in Pyrex glass. Finally a review of materials with regard to increasing the coolant core outlet temperature is given: in particular HT9 cladding and various ex-core structural materials. It has been shown that, with regard to cladding material limits, venting can provide at least a 20°C increase in the core outlet temperature since venting decreases mechanical stress on the cladding due to fission gas pressure. Also, based on current designs and experience high-chromium steels are very promising candidates for ex-core structural material (e.g piping), together with ODS (if their chemical compatibility with liquid sodium and weldability are verified): the latter can operate at about 600°C still keeping a margin of 100°C from the upper temperature limit. Based on the present analysis is that the ex-core structural material limit is a more limiting factor than the cladding material limit with regard to increasing the coolant core outlet temperature. In conclusion it has been demonstrated that the vented inverted fuel assembly configuration is an interesting and valuable concept to take into account for future investigation in order to improve the performance of sodium-cooled fast reactors.
Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program