Analytical Modeling of the Emergency Draining Tank for a Molten Salt Reactor

The Molten Salt Fast Reactor is a reactor concept developed by the European Union based on a liquid fuel salt circulating through the reactor core. A peculiar emergency system, which takes advantage of the liquid fuel state, is represented by a tank located underneath the core, where the fuel can be passively drained and cooled; its geometry ensures that the fuel remains in subcritical conditions. In the framework of the SAMOFAR project, a design for the Emergency Draining Tank has been proposed: the tank shall be equipped with vertical cooling elements, arranged in a hexagonal grid; the liquid fuel salt, which heats up due to decay heat, will fill the gaps between the elements. In this work, analytical methods (Green’s functions and orthogonal decomposition) are employed to study the transient heat transfer associated with the proposed design and to perform a preliminary dimensioning of the system, such that overheating is avoided in any moment of the transient and the fuel salt is kept in a liquid state and in safe conditions for a long time. The models are constituted by multilayer monodimensional slabs and cylinders, with a pure heat conduction model. The assessment of the available grace time and preliminary considerations about fuel salt freezing and its influence on the system effectiveness are also included

Recent neutronics developments for reactor safety studies with SIMMER code at KIT

The SIMMER family of codes is applied for safety studies of sodium fast reactors and reactors of other types. Both neutronics and fluid-dynamics parts of SIMMER are under development. In the paper new neutronics capabilities are presented. In particular developments for neutron transport solvers and a new technique for taking into account thermal expansion effects are described. These new capabilities facilitate 3D simulations and improve accuracy of modelling for the initiation transient phase during a hypothetical severe accident

3-D Simulation of Fuel Assembly Blockage in MYRRHA

In the framework of the KIT and SCK•CEN R&D cooperation and as continuation of earlier studies performed for the EURATOM FP7 MAXSIMA project, accidental transients caused by a single fuel assembly (FA) block-age were simulated with the SIMMER-IV (3-D) code for the MYRRHA core, while assuming no power variation during the accident. A 7-FA model that includes mesh cells for inter-wrap-per gaps between FAs was applied, with the blockage of the central FA. Sensitivity analyses on the gap flow rate, fuel chunk jamming fraction, insulator pellet material were performed in order to identify a conservative case that maximises the chance of damage propagation from the blocked FA to the neighbouring ones. All calculations with different options and parameters did show the same sequence in the blocked FA, including melting of pin cladding, fuel pellet failure, small can-wall break-up, steel particle and fuel chunk accumulation leading to additional blockages, and large can-wall break-up. Finally, fuel chunks are swept out from this FA through the inter-wrapper gaps. In the calculations performed for several tens of seconds and longer, no canwall break-up in the neighbouring FAs has been observed. Nevertheless, different options for simulation of the insulator break-up lead to signifi-cantly different results in the later phases of calculations. If the insulator pellet breaks up when the cladding is lost, a fuel/steel blockage is formed, which results in a large canwall break-up, but this blockage is dissolved as soon as the upper steel structure melts. If no insulator pellet breaks-up, the fuel/steel block-age is kept in place by the ceramic insulator, which has a very high melting temperature. This observationsupports the use of an insu-lating material with low melting temperature; this option may prevent or reduce blockage of fuel/steel particles coming from failed pins that eventually may prevent or reduce the risk of damage propagation to the intact core

The Influence of Resonance Scattering to the Doppler Reactivity Coefficient

The paper presents the results of an evaluation of the effect of scattering on resonances in calculating the Doppler reactivity coefficient using the data preparation algorithms implemented in the GRUCON processing program. A comparison of the free-gas model with the resonance scattering model and published results of calculations performed using other methods of data preparation is presented. On the benchmark of Mosteller for light water grids with various fuel compositions it was shown that taking into account resonances in the differential cross sections of elastic scattering of uranium-238 leads to a shift of the Doppler reactivity coefficient by ~ 10% towards negative values, thereby increasing the negative feedback with the temperature of the fuel

Minor actinide balance reduction in ESFR-SMART

New safety measures were proposed recently for ESFR, a design for a large European sodium-cooled fast reactor. The fissile height was reduced by 25% and 5% in the inner and outer cores, respectively; a lower fertile blanket was implemented. A unique fissile enrichment was chosen. Additional fuel subassemblies (FAs), passively operating safety rods and corium discharge tubes were introduced. In the current core, the sodium void effect is strongly reduced, that is favourable for reactor safety, but the Am and Cm balances, i.e. their mass variations under irradiation, remain positive. In the paper we investigate options for Am incineration in the core radial and lower fertile blankets by introducing there a mixture of U and Am oxides instead of U oxide and/or steel. With this mixture in the radial blanket instead of steel, the Am and total minor actinide balances approach negative and zero values, respectively. With this mixture instead of U oxide and steel in the lower blankets, these balances are negative; the sodium void effect is lower. The modified cores produce more Pu and Cm