http://hdl.handle.net/1721.1/67473 Wed, 22 May 2013 22:26:07 GMT 2013-05-22T22:26:07Z http://hdl.handle.net/1721.1/75291 Vented Inverted Fuel Assembly Design for an SFR Vitillo, Francesco; Todreas, Neil E.; Driscoll, Michael J. 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. Wed, 01 Jun 2011 00:00:00 GMT http://hdl.handle.net/1721.1/75291 2011-06-01T00:00:00Z http://hdl.handle.net/1721.1/75290 Effects of Surface Parameters on Boiling Heat Transfer Phenomena Truong, Bao Hoai; Hu, Lin-wen; Buongiorno, Jacopo; McKrell, Thomas J. Nanofluids, engineered colloidal dispersions of nanoparticles in fluid, have been shown to enhance pool and flow boiling CHF. The CHF enhancement was due to nanoparticle deposited on the heater surface, which was verified in pool boiling. However, no such work has been done for flow boiling. Using a cylindrical tube pre-coated with Alumina nanoparticles coated via boiling induced deposition, CHF of water was found to enhance up to 40% compared to that of the bare tube. This confirms that nanoparticles on the surface is responsible for CHF enhancement for flow boiling. However, existing theories failed to predict the CHF enhancement and the exact surface parameters attributed to the enhancement cannot be determined. Surface modifications to enhance critical heat flux (CHF) and Leidenfrost point (LFP) have been shown successful in previous studies. However, the enhancement mechanisms are not well understood, partly due to many surface parameters being altered at the same time, as in the case for nanofluids. Therefore, the remaining objective of this work is to evaluate separate surface effect on different boiling heat transfer phenomena. In the second part of this study, surface roughness, wettability and nanoporosity were altered one by one and respective effect on quenching LFP with water droplet was determined. Increase in surface roughness and wettability enhanced LFP; however, nanoporosity was most effective in raising LFP, almost up to 100ºC. The combination of the micro posts and nanoporous coating layer proved optimal. The nanoporous layer destabilizes the vapor film via heterogeneous bubble nucleation, and the micro posts provides intermittent liquid-surface contacts; both mechanisms increase LFP. In the last part, separate effect of nanoporosity and surface roughness on pool boiling CHF of a well-wetting fluid, FC-72, was investigated. Nanoporosity or surface roughness alone had no effect on pool boiling CHF of FC-72. Data obtained in the literature mostly for microporous coatings showed CHF enhancement for well wetting fluids, and existing CHF models are unable to predict the enhancement. Wed, 01 Jun 2011 00:00:00 GMT http://hdl.handle.net/1721.1/75290 2011-06-01T00:00:00Z http://hdl.handle.net/1721.1/75289 General Analysis of Breed-and-Burn Reactors and Limited-Separations Fuel Cycles Petroski, Robert C.; Forget, Benoit A new theoretical framework is introduced, the “neutron excess” concept, which is useful for analyzing breed-and-burn (B&B) reactors and their fuel cycles. Based on this concept, a set of methods has been developed which allows a broad comparison of B&B reactors using different fuels, structural materials, and coolants. This new approach allows important reactor and fuel-cycle parameters to be approximated quickly, without the need for a full core design, including minimum burnup/irradiation damage and reactor fleet doubling time. Two general configurations of B&B reactors are considered: a “minimum-burnup” version in which fuel elements can be shuffled in three dimensions, and a “linear-assembly” version composed of conventional linear assemblies that are shuffled radially. Based on studies of different core compositions, the best options for minimizing fuel burnup and material DPA are metal fuel (with a strong dependence on alloy content), the type of steel that allows the lowest structure volume fraction, and helium coolant. If sufficient fuel performance margin exists, sodium coolant can be substituted in place of helium to achieve higher power densities at a modest burnup and DPA penalty. For a minimum-burnup B&B reactor, reasonably achievable minimum DPA values are on the order of 250-350 DPA in steel, while axial peaking in a linear-assembly B&B reactor raises minimum DPA to over 450 DPA. By recycling used B&B fuel in a limited-separations (without full actinide separations) fuel cycle, there is potential for sodium-cooled B&B reactors to achieve fleet doubling times of less than one decade, although this result is highly sensitive to the reactor core composition employed as well as thermal hydraulic performance. Tue, 01 Feb 2011 00:00:00 GMT http://hdl.handle.net/1721.1/75289 2011-02-01T00:00:00Z http://hdl.handle.net/1721.1/75288 Design of a Functionally Graded Composite for Service in High Temperature Lead and Lead-Bismuth Cooled Nuclear Reactors Short, Michael P.; Ballinger, Ronald G. A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required performance criteria of high strength and corrosion resistance. A Functionally Graded Composite (FGC) was created with layers engineered to perform these functions. F91 was chosen as the structural layer of the composite for its strength and radiation resistance. Fe-12Cr- 2Si, an alloy developed from previous work in the Fe-Cr-Si system, was chosen as the corrosion-resistant cladding layer because of its chemical similarity to F91 and its superior corrosion resistance in both oxidizing and reducing environments. Fe-12Cr-2Si experienced minimal corrosion due to its self-passivation in oxidizing and reducing environments. Extrapolated corrosion rates are below one micron per year at 700°C. Corrosion of F91 was faster, but predictable and manageable. Diffusion studies showed that 17 microns of the cladding layer will be diffusionally diluted during the three year life of fuel cladding. 33 microns must be accounted for during the sixty year life of coolant piping. 5 cm coolant piping and 6.35 mm fuel cladding were produced on a commercial scale by weld-overlaying Fe-12Cr-2Si onto F91 billets and co-extruding them, followed by pilgering. An ASME certified weld was performed followed by the prescribed quench-and-tempering heat treatment for F91. A minimal heat affected zone was observed, demonstrating field weldability. Finally, corrosion tests were performed on the fabricated FGC at 700°C after completely breaching the cladding in a small area to induce galvanic corrosion at the interface. None was observed. This FGC has significant impacts on LBE reactor design. The increases in outlet temperature and coolant velocity allow a large increase in power density, leading to either a smaller core for the same power rating or more power output for the same size core. This FGC represents an enabling technology for LBE cooled fast reactors. Fri, 01 Oct 2010 00:00:00 GMT http://hdl.handle.net/1721.1/75288 2010-10-01T00:00:00Z