Nuclear analyses for shielding design of the divertor Tokamak Test (DTT) facility

The fusion science is, nowadays, approaching to the most important milestone through the realization of fusion energy: the International Thermonuclear Experimental Reactor (ITER). It is a Tokamak under construction in the south of France and has the aim at demonstrating the feasibility of nuclear fusion energy. It will start its operation in 2025 and, during the high-performance phase, it will produce 500 MW in 400 s of pulse with an expected Q ≥ 10. ITER will bridge the gap between the actual small-scale experimental fusion devices and the demonstration future fusion power plant (DEMO). Along the road to DEMO there are many other challenges to face up before arriving to the realization of fusion energy. They are widely explained in the living document “a Roadmap to the realization of Fusion Energy”, published in by the European Fusion Development Agreement (EFDA) in 2012. One of the most important challenge, indicated in the document as a possible showstopper for the fusion, is the power exhaust issue. The plasma in a tokamak is confined by means strong magnetic fields forming a set of closed magnetic surfaces. The confinement is, however, not perfect and, at the far end of the plasma, a few millimetres thick layer, called Scrape Off Layer (SOL) where the magnetic field is open is observed. Through the SOL, particles exit from the confinement and finish their life in a particular region of the chamber, called divertor. This component is particularly stressed by the heat loads, which could arise to the impressive value of 50 MW/m2, foreseen in ITER. For this reason, a facility called Divertor Tokamak Test (DTT), has been planned in the “roadmap to the fusion energy” with the aim at studying and testing different divertor concepts and plasma configurations in a reactor-scale environment. Indeed, after many years of experiments and theoretical simulations, the fusion community agreed to the fact that ITER could not provide alternative solutions and divertor concepts to face up the power exhaust problem. The DTT machine will be built in Italy, within the Frascati ENEA research centre, during the next decade. It will operate with Deuterium-Deuterium (DD) plasma and its main parameters are: Major Radius R=2.11 m, Minor Radius a=0.64 m, Max plasma current I=5.5 MA and toroidal magnetic field on axis B= 6 T. An auxiliary power of 45 MW is also foreseen to reach high-performances and DEMO-relevant conditions. This great amount of additional power will be provided by a set of additional heating systems: the Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating (ICRH) and Negative Neutral Beam Injector (NNBI). During its high-performance phase DTT will produce the considerable amount of 1.5x1017 n/s 2.45 MeV neutrons from DD reactions. Moreover, due to the triton burn-up, an additional 1% of DT 14.1 MeV neutrons is also foreseen. According to the provisional operational scenario, which foresees six months of operation and six months of maintenance, DTT will produce in one year 1.53x1022 n/y plus the 1% of DT neutrons. At the end of its life 3.73x1022 n will be produced. For these reasons, neutronics analyses are of paramount importance since the beginning phase of the project. For the design of the machine the assessment of the radiation fluxes and the nuclear loads on the components are fundamental, especially for the superconductive coils and for the assessment of the neutron-induced radioactivity. Another important issue regards the licensing procedure. Since DTT is a machine that produce more than 107 n/s over the whole solid angle, it is under a strict licensing procedure as imposed by the Italian Regulation. Neutronics activities are crucial for the definition of the shielding needs to accomplish the radiation protection constraints and limits during and off operations: 300 Sv/y for non-radiation workers and 10 Sv/y for the population at the ENEA site boundaries. This work of thesis is devoted to the three-dimensional shielding analyses in support of the DTT development. These analyses have been carried out by means the MCNP5 v1.60 Monte-Carlo transport code and its tools and features. FENDL2.1 & FENDL3.1 have been used for nuclear data libraries. Flux-to-dose conversion factors from ICRP-74 have been used for the calculations of the effective dose rates. ADVANTG hybrid transport code and iWW_GVR F4E tool were used to generate the variance reduction parameters and manipulate the results. ENEA Advanced-D1S code has been used for the SDDR assessment. Simulations have been performed on High Performance Computing (HPC) resources: ENEA CRESCO and EUROfusion MARCONI Clusters. Two reference neutronic models have been used for the analyses: a 20° single sector detailed model with reflecting boundaries, used for the nuclear loads, shielding studies, activation, Shut Down Dose Rate (SDDR) assessment and integration of auxiliary components in the tokamak, and a simplified 360° model with NBI and buildings for the assessment of the radiation field inside and outside the DTT building. For the first, analyses have been focused on the assessment of nuclear fluxes and doses inside and outside the Torus Hall Building (THB) for the definition of the thickness of the THB walls. For this scope a number of heavyweight concretes, with density around 3 g/cm3, and thicknesses, from 150 cm to 250 cm, have been analysed and compared to the ordinary concrete (=2.2 g/cm3). 220 cm of ordinary concrete results to be enough to respect the radiation protection constraints avoiding the use of special materials. To complete analyses regarding the THB, a study on the Skyshine effect has been done to evaluate the effective dose at the ENEA site boundaries due to the neutron scattering with air nuclei, necessary to determine the thickness and material compositions for the THB roof. Even in this case several configurations and shielding materials, both for neutrons and gammas, have been considered and it has been found that 150 cm of concrete implies a Skyshine dose equal to 6 Sv/y. The total effective dose results below the limits even if the direct contribution from the THB walls is considered. Another step regards the auxiliary systems such as Heating and Current Drive (HCD), cryogenic systems and also diagnostics which will be placed outside the THB and connected to the machine through transmission lines (TL) and cables that penetrate the THB walls. From the neutronics point of view, these penetrations cause, in most of the cases, large neutron and gamma streaming with an increment of fluxes and doses outside the THB. An accurate evaluation of the penetrations in the south and east side the THB and the shielding adopted is, then, presented inside the thesis. The south area will host the ECRH corridor for the ECRH TLs. In the east side the Fast Discharge Unit (FDU) area, the Helium transmission line and the passage area for the remote handling is foreseen within a connection corridor between the THB and the ex FTU building. For the four ECRH penetrations, hosted in a corridor lying about 10 m above the ground, a shielding configuration have been studied to limit the neutron and gamma streaming outside and inside the corridor. Fixed and removable shields made of Stainless Steel, common polyethylene and ordinary concrete, have been used in the configurations that leads to the respect of the design constraints at the ground level addressing the annual effective dose to ~150 Sv/y. Inside the corridor, the effective dose close the Gyrotron building (40 m far from the THB) has been limited by adding a 20 cm thick of shielding gate made of ordinary concrete and polyethylene. In this way the annual effective dose close to the Gyrotron building is of the level of 30-40 Sv/y ensuring the protection of the workers even during the DTT pulses. Moreover, to accomplish the requirements for the transmission lines maintenance, the level of SDDR has been assessed and a negligible level of dose, below 10 Sv/h, has been found even at the end of DTT life. Regarding the east wall, the problem of the effects of the neutrons and gammas on the electronics has been assessed. To protect the FDU area and, thus, the TF magnets from the quenching, a massive configuration of shielding has been proposed by covering the room with polyethylene and concrete. Furthermore, the Helium line penetration has been studied and a dogleg shield has been optimized to limit the streaming of neutrons and gammas. A shielding door, necessary to close the opening between the THB and the corridor has been optimized as well. With this configuration the level of neutron fluxes inside the FDU room has been settled at level of 10-1 n/cm2/s and the absorbed dose in Silicon, mainly due to the gammas, is well below the reference value (i.e. adopted in ITER) of 1 Gy cumulated over the whole DTT life. Another fundamental activity for the machine development is the integration of the auxiliary components inside the port, showed in the last part of the work, where the integration of the ECRH equatorial launcher in the 20° MCNP DTT neutronics model has been done. Nuclear loads on the structural Stainless Steel and on the EC mirrors, made of Copper Chromium Zirconium (CuCrZr), have been evaluated as well as the level of contact dose rate and SDDR for a number of cooling time. Very high level of contact dose rate has been found even after 18 months of operations, (greater than 30 uSv/h) one week after the shutdown. This means that a very detailed strategy, supported by the neutronics calculations, of remote handling for this component should be studied in order to protect the workers and for the classification of the radioactive waste. All these nuclear analyses have been deeply discussed in the thesis and an overview of the future work and development has been given in the final part. In fact, when DTT will start its activities, neutronics will remain one of the most important part of the machine operations.