Andrea Ciampichetti

Qualification of tritium permeation barriers in liquid Pb–17Li

Abstract The reduction of tritium permeation from the Pb–17Li, or plasma, into the coolant is of crucial importance in order to reduce the radiological hazard in the steam generator vault as well as in the turbine/condenser area and to optimise the tritium balance in the reactor. The use of aluminium rich coatings has been selected as reference solution for the water cooled lithium lead (WCLL) blanket in order to produce reliable tritium permeation barriers (TPB). TPB qualification activities performed in the past allowed the selection of two reference deposition techniques, the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. On the basis of the results obtained in the past with the Corelli I–II devices, a new apparatus named Vivaldi was designed to perform comparative tests on two hollow cylindrical specimens in the same operating conditions. The performance of alumina coating on EUROFER 97 steel has been tested in gas and liquid metal phase. The obtained results in terms of permeated fluxes and permeation reduction factors (PRF) are herein presented and discussed. A post experiment examination of coatings was performed by use of optical and SEM microscopy.

Tritium Extraction from Liquid Pb-16Li: A Critical Review of Candidate Technologies for ITER and DEMO Applications

Abstract Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the feasibility of any PbLi based tritium breeding blanket (BB). Particularly in DEMO, high tritium extraction efficiency will be required in order to keep low the tritium concentration in the Pb-16Li loop. This is essential to minimize tritium release into the environment and tritium permeation from BB into the primary cooling system. In addition, the tritium extraction process needs to be highly reliable in order not to impact negatively on the operation of the whole fusion reactor, ITER or DEMO. In the present paper, a critical review of the main candidate technologies for tritium extraction from Pb-16Li, particularly gas liquid contactors and vacuum permeators, is accomplished. The intrinsic limits and possible advantages of these technologies are presented and discussed, in the light of considerations coming directly from mathematical models describing their behaviour as well as from the experimental results so far achieved. Needs in terms of R&D activities are identified.

TRITIUM EXTRACTION SYSTEMS FOR THE EUROPEAN HCLL/HCPB TBMs

Abstract One of the most challenging issues for the TBM (Test Blanket Module) testing campaign foreseen in ITER is the operation of TES (Tritium Extraction Systems). This is essential not only to prove the ability to manage correctly the bred tritium but also to validate and qualify the neutronic codes for the prediction of tritium production in view of their use in future fusion plants. Two are the European candidates to be tested in ITER: the HCPB (Helium Cooled Pebble Bed) TBM and the HCLL (Helium Cooled Lithium Lead) TBM. For both these TBM concepts the following points have been addressed in this work: a) the gas stream to be processed by TES b) the TES process flow diagram c) a first assessment of the required space

Safety and Radioactive Waste Management Aspects of the Ignitor Fusion Experiment

Abstract Ignitor is a nuclear fusion experiment aimed at studying Deuterium-Tritium plasmas. If European proposed waste management strategies were applied, all Ignitor radioactive materials could be recycled or declassified to non-radioactive material. We have applied the Italian waste management regulations to the IGNITOR experiment radioactive materials: none of them should be classified in the High Level Waste category but the vessel, and most materials are classified as LLW (Low Level Waste). The machine has very low radiological risks and environmental impact.

Pb–17Li/water interaction in DEMO WCLL blanket: water micro-leaks

Abstract The interaction between pressurised water and liquid Pb–17Li is a topical issue for the WCLL blanket concept and needs to be studied in detail because of the potentially significant effect on the reliability and safety of the blanket. Particularly, it seems important from the design point of view to deeply investigate the consequences of coolant micro-leaks in the blanket module generated by micro-cracks with a size in the order of 10−3 mm2. This kind of interaction has been studied during an extensive experimental campaign on the RELA loops in the ENEA Research Centre at Brasimone. The results showed a reduction in the liquid metal flow-rate in the circuit and a deterioration of the heat exchange properties between breeder and coolant due to the formation and growth of solid reaction products generated by the chemical reaction between lithium and water. The molar ratio between the recovered hydrogen recovered and water injected was in the range 0.3–0.5, confirming that, in the test conditions, the main solid reaction product is lithium hydroxide (LiOH). The possible impact of these results on the TBM-ITER design is also presented and discussed.

Accidental and long-term safety assessment of fission and fusion power reactors

Abstract Fusion is seen as a much ‘cleaner’ energy than fission, and a resource for clean energy in the far future. However, being a form of nuclear energy, fusion shares with fission most of its safety problems. This study concerns the assessment of both the short term and the long term hazards associated with fusion, compared with the same figures for fission reactors. Fission data derive from well-known PWR safety assessments, and in particular from the Italian project Progetto Unificato Nucleare (PUN). Fusion data derive from the latest findings of the European Safety and Environmental Assessment of Fusion Power (SEAFP) programme. Concerning inadvertent intrusion in a radioactive waste disposal site and long-term degradation for fusion, this analysis was performed in the frame of the SEAFP studies, while for fission data are taken from typical European waste disposal sites. In all relevant cases, evaluation of associated risks is carried out, with a comparison of the obtained results.

Tritium Transport Issues for Helium-Cooled Breeding Blankets

Tritium mobility through breeding blanket (BB) and steam generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason, it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modeling and parametric issues. In this paper, the main permeation issues are identified and possible solutions are discussed to address the modeling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled BBs: 1) helium-cooled pebble beds and 2) helium-cooled lithium-lead. For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC to define proper and feasible tritium mitigation techniques, which are needed to keep the tritium losses lower than the allowable environmental release (i.e., 20 Ci/d).

Sensitivity study for Tritium Permeation in Helium-Cooled Lead-Lithium DEMO Blanket with the FUS-TPC Code

Abstract Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of hydrogen and its isotopes in materials. We address the problem of tritium transport in Helium Cooled Lead-Lithium (HCLL) DEMO blanket from lead-lithium breeder through different heat transfer surfaces to the environment by developing a computational code (FUS-TPC). The main features of the code are briefly described and a parametric study is performed in order to identify the most influencing parameters in terms of tritium releases into the environment and of tritium inventories. The results showed that the results are strongly affected by the tritium Sievert’s constant in Lead-Lithium and the efficiency of permeation barriers.

A zero-waste option: recycling and clearance of activated vanadium alloys

The reduction of long-term radioactivity is analysed here in V–Cr–Ti alloys, one of the proposed structural materials for fusion power plants. In particular is explored the possibility of recycling within the nuclear industry and of clearance, that is declassification to non-active material. Vanadium alloys have the potential to reach the dose rate hands-on recycling limit when used in a blanket, if noxious radioactive products coming from impurities activation are eliminated. Clearance is also possible in principle, but only if further separation of activation products of titanium is carried out after service.

The zero waste option: clearance of activated and first wall/blanket materials

Abstract Management of activated waste from fusion power reactors is part of the long-term action of the European Fusion Programme. The options of recycling , the reuse in novel reactors for in-vessel materials, and clearance , the declassification to non-active waste, for ex-vessel materials, were analysed with the aim to reduce the amount of radioactive waste. A further step is the attempt to reduce the long-term radioactivity of selected materials to levels allowing clearance also of in-vessel structures. Clearance can be an alternative option also for in-vessel materials. Vanadium-based alloys belong at the moment to the materials with sufficiently low-activation constituting elements. Attainment of clearance conditions in activated in-vessel V–4Cr–4Ti structures would require very low concentrations of impurities, not achievable with the current purification methods and in some cases below the present detection limits, and the development of methods to reprocess the activated alloy for extracting radiotoxic nuclides.