Development of the new aina code and its application to the safety analysis of the european demo designs
- Autores: Eduard Baeza Pérez
- Directores de la Tesis: Alfredo de Blas del Hoyo (dir. tes.), Albert Riego Pérez (codir. tes.)
- Lectura: En la Universitat Politècnica de Catalunya (UPC) ( España ) en 2019
- Idioma: español
- Tribunal Calificador de la Tesis: Joaquín Sánchez Sanz (presid.), Guillem Cortes Rossell (secret.), Jesús Izquierdo Villena (voc.)
- Programa de doctorado: Programa de Doctorado en Ingeniería Nuclear y de las Radiaciones Ionizantes por la Universidad Politécnica de Catalunya
- Materias:
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- Resumen
A conclusion that can be drawn from the historical safety analyses developed for tokamaks fusion reactors is that some of the major risks involve incidents in the vacuum vessel. In order to evaluate plasma evolution and in-vessel components strains, a safety code called AINA was developed during the last ten years for different fusion reactors designs as ITER and the Japanese DEMO design WCPB. The present document includes an in-depth critical analysis of these former AINA versions, a new codification of it and a checking and validation process in order to develop a proper, reliable, versatile and flexible tool with the purpose of carrying contributions to safety analyses for the four different European designs of DEMO (HCPB, DCLL, HCLL and WCLL). Therefore, AINA 4.0 becomes a reliable code comprised of a 0D plasma dynamics approach based on a mass and energy balance and a 1D thermal model for the blanket (in the radial direction) and the divertor. These two blocks feed-back constantly each other by means of the plasma-wall block which estimates the real loads suffered by the in vessel components and the real impurity presence into the plasma core. With this basic concept, AINA is useful to check the integrity of these in-vessel components both when a plasma perturbation induces a Loss Of Plasma Control (LOPC) and a thermo-hydraulic accident takes place in the Plasma Facing Components (PFCs) or in the Vacuum Vessel such as a Loos Of Coolant Accident (LOCA). One of the most important findings of this study is that for the DEMO 1 and DEMO2 scenarios certain functional temperature limits may be slightly exceeded in the worst poloidal region (OB4) for the HCLL and the HCPB blanket designs; on the contrary, for the DCLL and the WCLL models this phenomena does not take place due to the particular architecture and cooling scheme of each design. Moreover, with regard to LOCAs perturbations, from AINA outcomes, it concludes that this kind of failure inside the cooling system does not affect the internal plasma conditions but, undoubtedly, even a slight loss of the mass flow leads the reactor to an overall temperature increase for all the designs and levels. For this reason, it is indispensable the installation of a quick response system capable of detecting a cooling anomaly rapidly and activating a proper mitigation action such as a fast plasma shutdown FPSS injecting impurity gases (e.g. Ne, Ar, etc.) as in ITER. Concerning LOPC cases there is a wide variability of situations depending on the perturbation simulated; some more critical than others, so the mitigation system and the particular actions applied depend on the case and are explained throughout the document.
