G. Schumacher

Properties of lithium orthosilicate spheres

Abstract Lithium ceramic spheres have been proposed as a tritium breeding material for a fusion reactor blanket. Spheres fabricated by Schott, Mainz show a glass-like structure in light and scanning electron microscopy. A crystalline structure, however, was detected by X-ray diffraction. Part of the spheres were annealed at 1000°C for 2 h to heal microcracks and to relieve internal stress. After annealing a grain structure was found by microscopy with grains of 10–30 μm grain size. When stored in air the spheres took up moisture. After some days the water content yielded 2–3 mol%. A thermo-mechanical test was conducted with the spheres by cycling between 60 and 600°C in a stainless steel capsule which simulated the pressure load during thermal cycling of the fusion reactor blanket. Examination of the spheres after 10 cycles showed that 11% of as-fabricated spheres were broken. The amount of broken spheres which had been annealed was only 2%. It is assumed that healing of microcracks and relieve of internal stress improves the behavior of the spheres.

Heat transfer and technological investigations on mixed beds of beryllium and Li4SiO4 pebbles

For the European BOT DEMO solid breeder blanket design the use of mixtures of 2 mm beryllium and 0.1–0.2 mm Li4SiO4 pebbles with and without 0.1–0.2 mm beryllium pebbles has been proposed. A series of heat transfer and technological investigations are being performed for these pebbles. Namely: (a) Measurements of the thermal conductivity and of the wall heat transfer coefficient of a 2 mm Be pebble bed, of a bed with 2 mm Be plus 0.1–0.2 mm Li4SiO4 pebbles and of a bed with 2 mm Be pebbles plus 0.1–0.2 mm Li4SiO4 and Be pebbles. (b) Thermal cycle tests of mixed beds of Li4SiO4 and beryllium pebbles; during these tests the pressure drop across the bed of the helium purging flow is measured. (c) Annealing tests at 650°C of the Li4SiO4 pebbles with and without the beryllium pebbles. (d) Measurement of the failure loads of the Li4SiO4 pebbles before and after annealing. Tests (a) and (b) have been performed for bigger Li4SiO4 pebbles (0.3–0.6 mm) as well. The results of the experiments are reported in the paper.

Considerations on PyC and SiC coated oxide particles for gas cooled fast reactor application

Abstract Recently, coated particles with a SiC outer coating layer have been proposed for Gas Cooled Fast Reactor application. SiC is a very fragile material. Although under fast irradiation it can creep at relatively low temperatures, a very low strain (⪢ 0.3%) can produce hairline cracks. A SiC coating layer can therefore withstand only moderate overpressures when subjected to large fast neutron exposures. In the present paper the pressure inside the particle due to gas fission products and CO-CO2 is calculated. The operation temperature of the particle due to the condition strain ⩽ 0.3% is given as a function of SiC layer thickness in conditions typical for GCFRs application

The KfK design of a helium-cooled ceramic blanket for net

A conceptual design of a helium-cooled blanket for the NET double null plasma configuration with a neutron wall load of 1 MW/m 2 and 600 MW fusion power is presented. The outboard blanket is made up of self-supporting canisters containing the beryllium multiplier in form of plates. The 6 mm wide slits between the plates contain a bed of 0.5 mm Li 4 SiO 4 pebbles. The helium purge flow at 0.1 MPa carries away the tritium produced in the bed. The first wall of stainless steel and with a graphite tile protection is cooled by toroidally running helium tubes. The inboard blanket is made up of a similar structure, however the helium coolant tubes run in the poloidal direction to allow for more breeding material in the narrow space available. The divertors are composed of TZM elements cooled by helium. The outboard first wall and blanket are cooled by helium at 6 MPa, (inlet temperature = 200 °C outlet temperature = 450 °C), while the divertor and the inboard first wall are cooled in series by helium at 11.5 MPa and the inboard blanket by helium at 8 MPa. The calculated temperatures and stresses in blanket, first wall and divertor, appear to be acceptable. Based on the LISA-2 experiments the tritium blanket inventory is about 400 g. The daily tritium production is 96 g and the three-dimensional real tritium breeding ratio is 1.04 for a 6 Li enrichment of 90%.

Research and development work for the lithium orthosilicate pebbles for the Karlsruhe ceramic breeder blanket

The Karlsruhe ceramic breeder blanket design for a demo reactor and for the test objects to be tested in NET is based on lithium orthosilicate (Li4SiO4) in form of 0.5 mm diameter pebbles contained in 6 mm wide gaps between beryllium plates. Two methods have been used to fabricate the pebbles: at KfK the pebbles were manufactured by extrusion, spheroidizing, and subsequent sintering using a fluidized bed, while at Schott Glaswerke, Mainz they were obtained by melting followed by spraying of the melt. Various tests have been performed with pebbles, namely: (a) measurements of the compressive forces which single pebbles can substain, (b) thermal cycling tests of Li4SiO4 pebbles in steel containers, (c) measurements of the effective thermal conductivity of Li4SiO4 beds, (d) in situ tritium extraction experiments using helium as purge flow.

Development work for a borax internal core-catcher for a gas-cooled fast reactor

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formed by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.

MECHANICAL PROPERTIES AND DEHYDRATION OF Li4SiO4 PEBBLES

A new concept of the Karlsruhe helium cooled ceramic breeder design is based on a mixed bed of small lithium orthosilicate pebbles of 0.1 - 0.2 mm diameter and 2 mm beryllium pebbles that serve as a neutron multiplier. Helium sweep gas flows through the bed and carries away the released tritium. The structure of the small pebbles and their mechanical properties are examined in this report. Release of tritium from lithium orthosilicate is controlled by desorption from the surface of the material. Studies of water desorption are helpful for modelling of release processes because part of the tritium is released in form of tritiated water. The temperature programmed release of water from lithium orthosilicate pebbles produced from molten material was investigated by means of thermogravimetry and compared to tritium release measurements.