Resolution of neutronic challenges for the development of ITER and DEMO-EU magnetic fusion reactors

Abstract

The objective of this thesis is contributing to the development of the magnetic confinement fusion, addressing issues of interest within the framework of ITER and DEMO. The International Thermonuclear Experimental Reactor, ITER, is a large-scale scientific experiment which aims to solve technical and scientific problems to advance in the nuclear fusion field. Its goal is to demonstrate the feasibility of fusion as an energy source and collect the necessary data for the design and subsequent operation of the first plant producing electricity from fusion energy. It is currently under construction in the south of France, although some components and systems are still in the design phase. ITER Organization (IO) is the legal entity responsible for building, operating, exploiting and deactivating ITER. The European Union, India, Japan, China, Russia, South Korea and the United States are the countries participating in the project. Beyond ITER, the DEMOnstration power plant DEMO aims to develop and test technologies for the operation of a fusion reactor not as a scientific experiment, but as a power plant, applying the know-how gained with the ITER project. One of the current identified problems in ITER is to achieve shutdown dose rates (SDDR) values below certain limits in the Port Cell (PC) and Port Interspace (PI) areas, in order to carry out manual maintenance activities. In this context, the quality of the EAF-2007 activation cross sections, which are usually used for these SDDR calculations in ITER, is assessed in order to set their reliability. In addition, possible improvements/updates in both the latest version of EAF (2010) and the TENDL library (2013 to 2015 versions) is evaluated. As a conclusion from this first part of the thesis, it has been seen that, to date, calculated doses (produced by the activation of each of the materials) are trustworthy (i.e. more than 90% of the production of major radionuclides is due to reactions with EAF validated cross sections) for the following materials: SS316LN-IG, SS304L, Eurofer, LiPb, W, conventional concrete from B-lite, and L2N concrete. On the contrary, the SDDR prediction for Cu and barite concrete (potential candidate material) is not reliable. On the other hand, ITER will have several diagnostic systems to provide the necessary measures to control, evaluate and optimize plasma performance and also to promote understanding of plasma physics. These diagnostic systems will be located at different components of the reactor: blanket, vacuum chamber, cryostat, ports (upper and equatorial), divertor, etc. By introducing these diagnostics into the reactor, two effects are produced. On the one hand, shielding material is removed and, on the other hand, streaming paths are generated. As a consequence, some components such as the toroidal coils (TFCs) or the vacuum vessel (VV) might be affected by the change in the radiation field. The fact that the radiation field changes may affect their appropriate functioning, endangering the superconducting state of the TFCs and, as a result, the plasma confinement. It is precisely for this reason that, within ITER IO, there is some concern about the radiation loads to which these components are subjected, since previous studies showed that the radiation load values were very close to the limit and, in some cases, above. For this reason, a chapter of the thesis addresses the impact of the inclusion of four in-vessel diagnostics (reflectometry, NAS-neutron activation system, FW-first wall samples and bolometers) on the radiation loads (nuclear heating and neutron induced damage) on the VV and the TFCs of the ITER reactor. In this sense, results show that the contribution of the analyzed diagnostic systems to the radiation loads on both the VV and the TFC is not critical for the appropriate functioning of these components. Regarding DEMO, the focus is on the DCLL (Dual-Coolant Lithium-Lead) blanket concept. The work consists of performing transport (MCNP) and activation (ACAB) calculations of the blanket materials on the basis of 2014 and 2015 DEMO models and, finally, providing results in terms of activation and residual heat. These results are the starting point for further assessments on safety and/or waste management. Furthermore, the radioactive waste production is assessed and the possibility of waste disposal in El Cabril facility is analyzed, including the determination of the impurities limits needed to achieve this goal. Results show that it is only necessary to reduce one impurity content (Nb) in order to dispose the waste coming from this blanket at the El Cabril near-surface facility.