Analysis of the MYRRHA spallation loop using the system code ATHLET

S. Palazzo, K. Velkov, G. Lerchl, Katrien Van Tichelen

    Research outputpeer-review


    One of the most interesting technologies recognized as promising for the future of nuclear energy is the Accelerator Driven System (ADS). The purpose of the ADS is to function as an element of an integrated nuclear power enterprise comprising of conventional and advanced power reactors for energy production, and for reducing the radiotoxicity of the nuclear waste produced by these power reactors before entombment in a geologic repository (Stanculescu, 2000). The use of new types of coolants such as molten lead or lead–bismuth eutectic alloy (LBE) represents another particular feature in the design concept of these reactors, since it permits one to take advantage of higher Heavy Liquid Metals (HLM) boiling temperature compared to water, leading to an improvement in thermal efficiency. Furthermore, it allows the reactor to be operated at a lower pressure, thus reducing the probability of a Loss-Of-Coolant Accident (LOCA) and, consequently, increasing reactor safety. The proposed use of relatively new types of coolants, especially in combination with new fuel types and cladding materials, demands specific attention to the thermal–hydraulics and core mechanics in normal and abnormal conditions. For that reason, a detailed analysis using system codes is necessary.

    In this paper, a new version of the ATHLET system code is tested, in which the physical properties of liquid metals like sodium, lead and LBE are implemented. The object of this study is the spallation loop of the MYRRHA facility, in which LBE is circulated by forced convection to remove the heat deposited by a proton beam. A detailed nodalization is set up for performing thermal–hydraulic calculations for both nominal conditions and accident scenarios in order to have a good characterization of the entire loop. The start-up transient has verified the correct removal of the heat generated in the target by the foreseen heat exchanger. During accidental transients, it was noted that the level of LBE in the beam line changes in agreement with preliminary studies on the target device. The pump failure test represents the most dangerous scenario, as the temperature in the target area will reach a very high value within a few seconds after the blockage of the pump impeller. The results obtained have subsequently been compared to the ones achieved with previous numerical simulations which were performed using a version of RELAP5/mod 3.3 system code that was modified at the University of Pisa to account for LBE properties. The comparison has confirmed the capability of the new version of the ATHLET code for the analysis of hydraulic circuits cooled by liquid metals, although both codes use different heat transfer correlations.
    Original languageEnglish
    Pages (from-to)274-286
    JournalAnnals of nuclear energy
    StatePublished - 2013

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