Lead Lithium Blanket Concept Research Articles

PbLi/Water Reaction: Experimental Campaign and Modeling Advancements in WPBB EUROfusion Project

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes.

Chemistry and corrosion research and development for the water cooling circuits of European DEMO

The European DEMO design will potentially use single-phase water cooling in various components that require protection against corrosion. Coolant conditions will be similar to those of fission plant but with additional considerations arising from materials choices (Eurofer-97, CuCrZr) and 14 MeV neutron irradiation. Presently, many aspects of the water chemistry and corrosion behaviour are not well defined, and several strands of work, reported here, are ongoing to address these challenges under the EUROfusion framework in collaboration with industrial partners to leverage knowledge and expertise from elsewhere in the nuclear industry.Starting with the water-cooled lithium-lead blanket concept and considering the interaction of water with Eurofer-97, the foundation of this work has been the definition of a working water chemistry specification, supported by a review of relevant operating experience from light water fission reactors to understand the potential for technology transfer. Radiolysis modelling has been used to assess options for suppression of oxidising species under high energy neutron irradiation as these can be corrosive to components within the plant. High temperature water corrosion testing facilities have also been employed to expand the corrosion database and supplement existing experimental activities in the EUROfusion programme.In-vessel cooling of the divertor will use CuCrZr tubes under lower-temperature, high flow velocity conditions, which will lead to different considerations compared to the blanket and the potential for flow-accelerated corrosion. Additionally, high, unidirectional, heat fluxes lead to a radial temperature profile and the possibility of sub-nucleate boiling. A separate test setup, currently under construction, to expand this corrosion database is described.

Ionizing radiation effects on the thermal stability of deuterium trapping in reaction bonded SiC

SiC materials are prime candidates for flow channel inserts in the dual coolant lithium lead blanket concept. Flow channel inserts made of SiC will be exposed to tritium from the Li transmutation, as well as to neutron and gamma radiation. Hence a critical issue for future fusion devices is to clarify hydrogen isotope behaviour in SiC under such conditions. The objective of the work presented here is to study the effect of ionizing radiation on the deuterium trapping in Reaction Bonded SiC in similar conditions as reactor materials. This effect is evaluated by studying the influence of ionizing radiation on deuterium trapping (for both implanted and loaded SiC samples). Moreover, it is investigated how deuterium trapping may be modified by displacement damage. The ionizing radiation effect on absorption has also been evaluated for samples pre-damaged by self-ion irradiation. The irradiation and implantation experiment have been carried out at the CMAM-UAM accelerator, and the Danfysik implanter and 2 MeV Van de Graaff electron accelerator at CIEMAT. Samples are analysed by thermal desorption spectroscopy and secondary ion mass spectrometry (SIMS) to clarify the mechanisms involved in the trapping processes, depending on the different experimental conditions. The results for the deuterium loaded samples indicate that absorption is increased by ionizing radiation. When samples are pre-damaged by C+4 ions, deuterium absorption is increased in the form of Si-D according to SIMS results. Furthermore, the effect of ionizing radiation after deuterium implantation is an enhancement of the deuterium released from SiC. The deuterium release observed in this case is forming hydrocarbons during irradiations.

Simulation study of pressure trends in the case of loss of coolant accident in Water Cooled Lithium Lead blanket module

The water–lithium lead interaction implies a direct energy release, which leads to temperature and pressure increase, due to a combined thermal and chemical reaction, and an indirect form of energy release, the hydrogen production, due to secondary chemical reaction involving the initial reaction products. Review and understanding of the knowledge acquired in past studies, experimental works and numerical activities are needed in view of the renewed interest in the Water Cooled Lithium Lead blanket concept and safety issues connected with the fusion reactor design. This paper presents a review of the studies carried out in the past to characterize the potential safety concerns associated with the use of water and lithium-lead eutectic alloy, the main experimental campaigns, and numerical simulations of BLAST Test No. 5 performed by SIMMER-III code. As results, no code was found able to perform a satisfactory post-test analysis of separate effect experiments, without engineering assumptions. Therefore, a code model for the exothermic reaction and hydrogen production, and experimental data are needed for solving the WCLL blanket safety issues associated with the water–PbLi interaction.

The ARIES Advanced and Conservative Tokamak Power Plant Study

Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5,a βtotalN of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m2. The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced-activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βtotalN of 2.5, an H98 of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.

Development of a Numerical Tool to Simulate Magnetohydrodynamic Interactions of Liquid Metals with Strong Applied Magnetic Fields

In the framework of the study of a European helium cooled lead lithium blanket concept for ITER, numerical tools are developed to complement experimental activities. Full capability to simulate numerically the global magnetohydrodynamic flow and pressure distributions resulting from the interaction of the liquid metal with the strong plasma confining magnetic field is not achieved yet. Calculations should support the selection and validation of physical models for 3D coupled phenomena, like magneto-convection, as well as for corrosion and tritium permeation processes. Moreover, simulations help to interpret measurement data and to enhance the development of extrapolation procedures from small-scale experiments to a DEMO reactor.The present paper summarizes the mathematical algorithm and modeling requirements for accurate predictions of liquid-metal flows under very intense magnetic fields in geometries with arbitrary electric conductivity of the walls. The Lorentz force term and additional equations determining electric current density and potential have been introduced in a consistent and conservative way into the existing hydrodynamic open source code OpenFOAM. The use of non-orthogonal corrections leads to a significant improvement of the MHD code at fusion relevant strong magnetic fields. The discussion focuses on benchmark problems used to validate the new developed tool and on the treatment in OpenFOAM of MHD flows in geometries with walls of finite electric conductivity. According to the authors’ knowledge, the implementation of this capability in this open source code has not been reported so far in other references.

Neutronic and thermomechanical analysis of the water-cooled lithium-lead blanket design for a DEMONET reactor

Within the framework of the European DEMO blanket study programme, CEA and the JRC of Ispra are jointly developing a water-cooled lithium-lead blanket concept. The new DEMONET reactor configuration released in Spring 1990 and currently specified in the EC programme is the basis of neutronic and thermomechanical studies for the proposed box-shaped blanket concept. Considering the high blanket coverage, it is now possible to reach tritium self-sufficiency without making use of beryllium (neutronic calculations indicate a global tritium breeding ratio of the order of 1.16).

Tritium Breeding Blanket

The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs.

European Water-Cooled Lithium-Lead Blanket Designs for a Demo Reactor

Within the framework of the European DEMO blanket study programme, CEA and the JRC of ISPRA develop a water cooled lithium-lead blanket concept. Considering the low breeding coverage ratio of the DEMO-1 reactor configuration currently specified in the E-C programme, the use of beryllium was envisaged in order to reach the tritium self-sufficiency which corresponds to an equivalent tridimensional Tritium Breeding Ratio (TBR) above 1.05. This beryllium is either directly immersed in the liquid metal (Pb-17Li) or completely separated from it.