Research highlights of FAST reactors group

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FAST code system

One of the main objectives of the project is to build and maintain a code system for neutronic, thermal-hydraulic and mechanic analysis of the static and dynamic behaviour of the fast-spectrum cores and the whole reactor systems for different coolants, fuel types and forms, cooling system designs, etc. This code system allows to analyse in a systematic manner: 1) open and closed equilibrium fuel cycle; 2) fuel base-irradiation behaviour; 3) a wide variety of transients. http://dx.doi.org/10.1016/j.anucene.2005.06.002
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Heat transfer to liquid metals

We analyzed four sets of experimental data (total of 658 data points) for heat transfer to liquid metals flowing in a triangular or square lattice of cylindrical rods for a wide range of Peclet numbers, using a number of correlations recommended for liquid metal flowing in tube bundles. A new correlation as a best fit to the data was derived. http://dx.doi.org/10.1016/j.nucengdes.2008.12.014
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Analysis of selected Phenix end-of-life tests

The end-of-life tests performed at the Phenix sodium-cooled fast reactor present a unique opportunity to validate computational tools used for static and transient analysis. The FAST code system (including ERANOS and TRACE/PARCS) has been used to analyze test data from the static control-rod-shift experiments and from the dynamic natural circulation test. http://dx.doi.org/10.1016/j.anucene.2012.05.035 and http://dx.doi.org/10.1016/j.anucene.2012.05.036
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Neutronic analysis of BFS-62 experiments

The BFS-2 critical facility at the Institute of Physics and Power Engineering (Russia) was designed for validation of fast reactor codes and nuclear data. The BFS-62-3A critical benchmark experiment was a mock-up of the BN-600 sodium-cooled fast reactor core. We developed and validated a 3D full-core heterogeneous model of the BFS-62-3A critical benchmark experiment using the Monte Carlo MCNPX-2.4.0 code and the ERANOS-2.2 code. http://dx.doi.org/10.1016/j.anucene.2012.01.011
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Pressure drop for sodium flow

An accurate prediction of pressure losses across fuel bundles under both single- and two-phase sodium flow conditions is necessary to simulate sodium-cooled fast reactor behavior during transients in which boiling is anticipated. We assessed and implemented in the TRACE code, friction factor models for wire-wrapped fuel bundles as well as local pressure drop models for grid spacers. We used experiments conducted at the Joint Research Centre, Ispra for validation. http://dx.doi.org/10.1016/j.nucengdes.2011.07.009
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Fuel cycle for molten-salt reactor

We studied the iso-breeding closed Th-cycle parameters as a function of the fuel-to-moderator ratio for single-fluid MSR core. The results show two potential options for an iso-breeding single-fluid MSR core. It is the graphite-free fast MSR reactor (100% salt) and the thermal MSR reactor around 15% of salt share in the core. http://dx.doi.org/10.1016/j.anucene.2013.08.007
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Core physics of molten-salt fast reactor

We analysed core physics and fuel cycle options for the molten-salt fast reactor (MSFR). This system is a promising option for uranium-233 breeding and Th-supported burning of transuranic elements, showing improved safety parameters compared to other fast reactors. Nearly actinide-waste-free energy production can become possible with 60% of MSFRs and 40% of liqght-water reactors. http://dx.doi.org/10.1016/j.pnucene.2013.06.006
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Thermal-hydraulic ADS Lead-bismuth Loop (TALL)

TALL is a medium-size experimental facility at KTH (Sweden), to study the thermal-hydraulics performance of lead-bismuth-cooled reactors and accelerator-driven systems (ADS). A number of transient experiments was performed, demonstrating excellent natural circulation performance of the system. Using the TRACE code, we simulated the tests and validated the computational models. http://dx.doi.org/10.1016/j.nucengdes.2006.01.006
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Drift-flux model for heavy liquid metal/gas flow

The gas lift pump concept based on the bubbling of an inert gas into the primary reactor coolant to enhance natural circulation is considered in a number of PbBi-cooled reactor concepts. We analyzed five sets of available void fraction data for heavy liquid metal / gas flow and proposed a drift-flux model which provides a best fit to the data. http://dx.doi.org/10.1080/18811248.2004.9726427
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Transient analysis of MEGAPIE

Megawatt pilot target experiment (MEGAPIE) was performed at PSI with the aim of demonstrating the feasibility of a liquid lead-bismuth target for spallation facilities at a maximum beam power level of 1 MW. The thermal-hydraulics data measured during the MEGAPIE experiment was used for the TRACE code qualification for transient analysis of liquid metal cooled systems. http://dx.doi.org/10.1016/j.anucene.2007.12.006
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Analysis of MEGAPIE reference accident

The TRACE model was applied to the analysis of the reference accident, resulting from the hypothetical loss of a Pb-Bi inventory and subsequent cooling down of the target with free convection of the atmospheric air. The final conclusion made on the basis of the work is that MEGAPIE in the reference accident meets the 1 mSv criterion. http://dx.doi.org/10.1016/j.nucengdes.2007.12.004
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Brayton cycle machine for decay heat removal in GFR

Decay heat removal (DHR) is a key safety and design issue for the Generation IV gas-cooled fast reactor. We developed a preliminary design of a Brayton cycle machine in order to to remove decay heat in case of depressurization of the system down to atmospheric pressure. http://dx.doi.org/10.1016/j.nucengdes.2011.10.011
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Heavy gas injection into GFR

Decay heat removal (DHR) is a key safety and design issue for the Generation IV gas-cooled fast reactor. We investigated the natural convection capability of the dedicated DHR loops under depressurized conditions while injecting a heavy gas (nitrogen) into the system. We showed that when a guard vessel is used in the design the nitrogen injection can help to remove decay heat in case of a large break in the primary system and the station blackout. http://dx.doi.org/10.1016/j.nucengdes.2010.05.030
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Analysis of gas loop tests

Several gas-loop experimental programs were carried out during the 1970-1980s at the Eidgenössisches Institut für Reaktorforschung (EIR, now PSI). To qualify the TRACE code for the gas flow simulation, we re-analyzed a wide range of thermal-hydraulics tests for smooth rods. We found that the built-in TRACE heat transfer and friction correlations are indeed quite suitable for simulating the Gas-cooled Fast Reactor conditions. http://dx.doi.org/10.1016/j.anucene.2010.01.020
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Development of control system for GFR

We developed the control assembly pattern for the Gas-cooled Fast Reactor and analyzed the related 3D static and dynamic behavior of the core with ERANOS and TRACE/PARCS, thus contributing to the GFR design development and safety analysis. http://dx.doi.org/10.1016/j.anucene.2008.09.008
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Analysis of plate-type fuel for GFR

One of the fuel designs considered for the Generation-IV Gas-cooled Fast Reactor is a ceramic plate matrix with a honeycomb inner structure containing small fuel cylinders. The fuel is mixed plutonium-uranium carbide, while the matrix material is silicon carbide. We developed a thermal-mechanical model of this highly innovative fuel using a simplified approach of the FRED code, supported by a detailed finite-elements modeling of one fuel pellet. http://dx.doi.org/j.anucene.2007.03.009 and http://dx.doi.org/j.anucene.2009.01.015