D. Murdoch

The inner deuterium/tritium fuel cycle of ITER

Abstract Beyond the initial hydrogen phase of the International Thermonuclear Experimental Reactor (ITER) the Tritium Plant is essential for the operation of the machine as tritium will be produced from D –D reactions. The inner deuterium/tritium fuel cycle of the Tritium Plant comprises the Tokamak Exhaust Processing (TEP) system, the Storage and Delivery System (SDS) including the Long Term Storage (LTS), the Isotope Separation System (ISS) and the Analytical System (ANS) as major subsystems. Besides the supply of the deuterium and tritium fuel and other gases to the tokamak and the recovery of tritium from all off-gases of the machine the inner fuel cycle has quite a number of additional duties. Incoming and outgoing tritium shipments need to be handled, tritium accountancy is an essential requirement and tritiated streams from sources other than the tokamak need to be processed. Not only in view of the control of effluents and releases, but also for economic incentives as much tritium as possible needs to be recovered for reuse from all off-gases and waste streams within the Tritium Plant of ITER.

The tritium fuel cycle of ITER-FEAT

The Tritium Plant of ITER-FEAT is essential for the operation of the machine after the initial hydrogen phase, as tritium will be produced from DD fusion reactions. Within the fuel cycle of the Tokamak deuterium and later also tritium will be provided to the Fuelling Systems, and the unburned DT fraction recovered from the exhaust gases. The design of the tritium fuel cycle has to be based upon well proven technology to assure the safe handling of tritium along with credible accountancy, low tritium inventory, low generation of wastes and a high reliability of all components throughout the lifetime of ITER-FEAT.

Design of the ITER tritium plant, confinement and detritiation facilities

Abstract This paper describes the design of the ITER tritium plant subsystems, layout in the tritium building and the construction plan. The tritium plant comprises tokamak fuel cycle processing systems, as well as tritium confinement and detritation systems. The plant processes tritiated gases received from the tokamak and other sources to produce the D, T gas streams for fuelling, and detritiates various waste streams including tritiated water before discharge to the environment. The plant has been designed to meet not only all anticipated plasma operation scenarios in the DD and DT phases with a wide range of burn pulse durations from short pulse (450 s) and long pulse (3000 s), but also safety requirements (minimization of equipment tritium inventory and environmental tritium release from different accidental events in tokamak and tritium processing subsystems, and reduction of workers’ tritium exposure, etc).

First operation experiences with ITER-FEAT model pump

Design and manufacturing of the model cryopump for ITER-FEAT have been finished. After acceptance tests at the contractors premises the pump was installed in the TIMO-facility which was prepared for testing the pump under ITER-FEAT relevant operating conditions. The procedures for the final acceptance tests are described. Travelling time, positioning accuracy and leak rate of the main valve are within the requirements. The heat loads to the 5 and 80 K circuits are a factor two better than the designed values. The maximum pumping speeds for H 2 , D 2 , He, Ne were measured. The value of 58 m 3 /s for D 2 is well above the contractual required value of 40 m 3 /s.

Performance tests of the ITER model pump

Abstract To study the pump performance of a sorption cryopump intended for use within the ITER fuel cycle, the test facility TIMO “Test facility for the ITER Model pump” was built at the Forschungszentrum Karlsruhe. After the successful acceptance tests of the model pump in 2000, new components were installed to improve the cooling performance. At the end of the reconstruction activities, the tests with the model pump were continued. During these new test series, all important aspects regarding the changed requirements of the new ITER design were to be checked, such as the pump behavior at high throughputs, the capacities achievable at temperatures higher than 5 K, and the possibility of using the torus cryopumps for the examination of small helium leaks.

ITER hydrogen isotope separation system conceptual design description

Abstract This paper presents integrated hydrogen Isotope Separation System (ISS) designs for ITER based on requirements for plasma exhaust processing, neutral beam injection deuterium cleanup, pellet injector propellant detritiation, waste water detritiation, and breeding blanket detritiation. Specific ISS designs are developed for a machine with an aqueous lithium salt blanket (ALSB) and a machine with a solid ceramic breeding blanket (SBB). The differences in the ISS designs arising from the different blanket concepts are highlighted. It is found that the ISS designs for the two blanket concepts considered are very similar with the only major difference being the requirement for an additional large water distillation column for ALSB water detritiation. The extraction of tritium from the ALSB is based on flash evaporation to separate the blanket water from the dissolved Li salt, with the tritiated water then being fed to the ISS for detritiation. This technology is considered to be relatively well understood in comparison to front-end processes for SBB detritiation. In the design of the cryogenic distillation portion of the ISS, it was found that the tritium inventory could be very large (>600 g) unless specific design measures were taken to reduce it. In the designs which are presented, the tritium inventory has been reduced to about 180 g, which is less than the ITER single-failure release limit of 200 g. Further design optimization and isolation of components is expected to reduce the inventory further.

Recovery of tritium from different sources by the ITER Tokamak exhaust processing system

Abstract Plasma exhaust during D–D and D–T operation of ITER will certainly not be the only source for gaseous streams within the tritium plant from which deuterium and tritium need to be recovered. Besides the gases from other operational modes of the tokamak, such as deuterium or helium from glow discharge cleaning or the fluids from the retrieval of tritium from co-deposits, various other sources within ITER will generate tritiated waste gases which have to be processed. Since ITER does not have a dedicated system for the treatment of gaseous wastes, all the tritium needs to be recovered by the tokamak exhaust processing (TEP) system. Consequently the TEP system has many more duties than the name of this particular part of the ITER tritium plant may suggest. The TEP process is designed to be fully continuous and based on permeation of hydrogen isotopes through palladium/silver (first process step), heterogeneously catalyzed cracking or conversion reactions (second process step), and counter-current isotopic swamping (third process step). The overall decontamination factor of the three-stage TEP process for tritium removal from tokamak exhaust gas at a composition as specified for the DT phase of ITER is at least 10 8 . Off-gases from this system can therefore be stacked via the normal vent detritiation system (N-VDS) of ITER after intermittent storage for decay of γ-active species in dedicated tanks.

New operational aspects of the ITER-FEAT primary vacuum pumping system

At Forschungszentrum Karlsruhe, a special cryosorption vacuum pump system is being developed for ITER-FEAT. The available R&D and design work performed so far originally supported the ITER-FDR machine, but only minor redirection has been needed. However, three new crucial aspects are described and critically discussed in this paper. These are the rather high throughputs per individual pump (which have almost doubled), the poisoning phenomena, and the use of the cryopumps during fine He leak detection. It could be demonstrated that all these aspects generated by the new ITER-FEAT design will not be a limiting factor for cryopump operation.

Design of the ITER Torus Cryopump

The cryopump concept for the ITER 2001 proposed by ITER-JCT was assessed. Some modifications were recommended. For these modified arrangements, Monte-Carlo calculations were performed to define the pumping probability of the cryopump together with the divertor duct. Results show that the ITER requirement can be fulfilled with small restraints if the pump inlet diameter is enlarged to 1000 mm and the neutron shield in the duct is taken out.

Cold trapping of traces of tritiated water from the helium loops of a fusion breeder blanket

Abstract The ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) will comprise three helium loops designed for: tritium extraction from the breeder zone, heat removal, and purification of the coolant. The process step envisaged for tritium extraction as well as for coolant purification includes a cryogenic cold trap as main component for the removal of tritiated water vapour (mainly HTO, H 2 O). The concentrations of water in the gas streams are expected to be extremely small, i.e. of the order of 10 ppm by volume. In this paper, we describe first runs with a cold trap using helium as the carrier gas at flow rates of 0.1 and 1.0 m 3 /h. The range of water vapour concentration in the helium carrier gas was 0.5 to >200 ppm v . The experiments have demonstrated the ability of the cold trap to remove water vapour efficiently from the He stream down to concentrations of less than 0.02 ppm v when the inlet water concentration is in the range of 300–650 ppm v or higher.