Edgar Bogusch

The Karlsruhe solid breeder blanket and the test module to be irradiated in ITER/NET

Abstract The blanket for the DEMO reactor should operate at an average neutron flux of 2.2 MW/m2 for 20000 h. This requires the use of a structural material which can withstand high neutron fluences without swelling. The ferritic steel Manet was chosen for this purpose. The breeder material is in the form of Li4SiO4 pebbles of 0.35 to 0.6 mm diameter. The 6 mm thick beds of pebbles are placed between beryllium plates which are cooled by high pressure helium flowing inside steel tubes. Breeder material and beryllium are contained in radial canisters, placed inside boxes. The coolant helium enters the blanket at 250° C, cools first the box walls and then the breeder and multiplier, and leaves the blanket at 450° C. The maximum temperature in the first wall steel is 550° C, while the minimum and maximum temperatures in the breeder are 380 and 820° C, respectively. The resulting total tritium inventory in the breeder is only 10 g, and the real tridimensional tritium breeding ratio is 1.11. The conceptual design of the test module, of its extraction system and of the required out-of-reactor ancillary systems has allowed an estimate of the time constants of the various components and thus allowed an assessment of the requirements given by the testing of the modules on the NET/ITER machine.

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%.

PEBBLE BED CANISTER: A CERAMIC BREEDER BLANKET WITH HELIUM COOLING FOR NET

ABSTRACT A blanket concept based on a closed first wall (F.W.) vessel and on self-supporting canisters containing a bed of Li 4 SiO 4 pebbles cooled by helium has been applied to a single and to a double null configuration for NET. The resulting tritium breeding ratio (TBR) is 1.31, however if a complete graphite protection of the F.W. is required the TBR decreases to 1.12. The tritium blanket inventory is relatively low if hydrogen is added to the helium purge flow. The blanket temperatures appear to be acceptable, however if F.W. heat fluxes higher 2 than 20 W/cm 2 are assumed, then austenitic steel cannot be accepted as F.W. material in connection with a full graphite F.W. protection. A helium cooled divertor for heat fluxes up 2 to 500 W/cm 2 appears in principle feasible.

HTR-TN Achievements and Prospects for Future Developments

It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to industrial process heat applications addressing (1) the development of a flexible HTR that can be coupled to many different process heat and cogeneration applications with very versatile requirements, (2) the development of coupling technologies for such coupling, (3) the possible adaptations of process heat applications required for nuclear coupling, and (4) the integration and optimization of the whole coupled system. As a preliminary step for this ambitious program, HTR-TN endeavors to create a strategic partnership between nuclear industry and R&D and process heat user industries.

THE KARLSRUHE HELIUM COOLED CERAMIC BREEDER BLANKET DESIGN FOR THE DEMONSTRATION REACTOR

The design is based on the use of a Breeder-Out-of-Tube solution consisting of a bed of small (∅ 0.35 – 0.6 mm) lithium orthosilicate pebbles and cooled beryllium plates contained in radial canisters. The required performance (av. neutron flux 2 MW/m2, 20000 hours operation time, maximum fluence in the structural material 70 dpa (NRT)) dictates the choice of a martensitic steel (MANET). Both the inboard and the outboard blanket canisters are contained in boxes. The first wall and side walls of the boxes are made of a thick slab of MANET containing drilled toroidal helium cooling channels, the back wall of the box being made by the poloidal coolant helium feeding tubes welded together. To avoid problems connected to the relatively high Ductile-Brittle-Transition-Temperature of MANET under irradiation, a relatively high coolant helium inlet temperature (250°C) has been chosen.