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The reference design of the IFMIF target and test cell (TTC) which is summarized in the comprehensive design report (CDR) is reviewed and evaluated. Improved TTC conceptual designs, as well as a revision of the reference design, have been proposed in order to qualify precise and flexible positioning of test modules against the backplate in the TTC. As the result of evaluations on the existing design concepts, a referential table which addresses the design requirements and guidelines has been released and is outlined in this paper. Based on the referential table and the existing design concepts a modular TTC concept is introduced, developed, and improved aimed on providing variable in-cell operation environments, flexible accessibility for remote handling, and accurate and reliable positioning for the test modules. The current status of the preliminary engineering design based on the modular TTC concept is represented.

The test specimen temperature control, together with optimized irradiation and reliability, is the central task of the HFTM design. The temperatures of the specimens should be kept at prescribed levels with acceptable temperature spreads. This paper concerns the design optimizations on HFTM through thermo-hydraulic simulations. Single-parametric studies were carried out for the effects of spacers, cooling channel/insulation gap deformation, multi-channels, insulation gap size, channel-rib contact resistance, etc. It is found that the temperature of the specimens and walls is generally sensitive to the above-mentioned parameters. These parameters therefore need to be considered during the optimization design, accounting also for the manufacture and operation tolerances. Simulation results on latest updated HFTM design are also presented. The results show that the current HFTM design could meet thermal management requirements in general. Further optimizations should be made individually for different target specimen temperatures.

The design of the High-Flux Test Module of the International Fusion Material Irradiation Facility has been developed continuously in the past few years. The present paper highlights recent design achievements, including a thorough state-of-the-art validation assessment of CFD tools and models. Along with design related analyses exercises on manufacturing procedures have been performed. Recommendations for the use of container, rig, and capsule materials as well as recent progress in brazing of electrical heaters are discussed. A test matrix starting from High-Flux Test Module compartments, i.e. segments of the full module, with heated dummy rigs up to the full-scale module with instrumented irradiation rigs has been developed and the appropriate helium gas loop has been designed conceptually. A roadmap of the envisaged experimental activities is presented in accordance with the test loop facility construction and mock-up design and fabrication schedules.

The International Fusion Material Irradiation Facility (IFMIF) is an accelerator driven neutron source for irradiation tests of candidate fusion reactor materials. Within the High Flux Test Module (HFTM) a testing volume of 0.5 l filled by qualified small scale specimens will be irradiated at displacement rates of 20–50 dpa/fpy in structural materials. The Forschungszentrum Karlsruhe (FZK) has developed a HFTM design which closely follows the design premise of maximising the space available for irradiation specimens in the IFMIF high flux zone and in addition allows keeping the temperature nearly constant in the rigs containing the specimen. Complementary analyses on nuclear, thermo-hydraulics and mechanical performance of the HFTM were performed to optimize the design. The present paper highlights the main design characteristics as well as recent progress achieved in this area. The contribution also includes (i) recommendations for the use of container, rig and capsule materials, and (ii) a description of the fabrication routes for the entire HFTM including brazing and filling procedures which are currently under development at the Forschungszentrum Karlsruhe.

An inspection procedure to assess the mechanical integrity of the International Fusion Materials Irradiation Facility (IFMIF) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray microtomography is the only known tool that could meet these requirements. In the high flux test module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules, which wear electric heaters and thermocouples, are housed in rigs. To assure a well-defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactured at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two microtomography facilities: a compact, high magnification installation at NILPRP and a high-end industrial tomography facility with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a directional microfocus X-ray source ( U = 220 kV and I = 300 μA) the geometry resolution was about 30 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements is sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. The space resolution and the overall reconstruction quality could be further improved by operating the X-ray source at low intensity and, consequently, very small focus size and, in addition, by implementing procedures to compensate geometry and beam hardening artifacts.

The present paper summarizes work devoted to design and design validation of the HFTM and improvements obtained by neutronics, thermohydraulic and mechanical analyses. The HFTM design requirements and its main features will be described. Remarkable progress has been achieved in the development of appropriate computational tools, data and methods for IFMIF neutronics and activation analysis. Similarly, the thermohydraulic simulations have been validated against experimental data obtained in the ITHEX facility to test turbulence models of gas flow in narrow channels. Iterative analyses lead to several design optimizations of container and helium supply. Crucial manufacturing techniques for the parts of the HFTM have been tested extensively to assure a reliable and reproducible manufacturing route for the final procurement of the HFTM. Finally, the integrated design of the Target and Test Cell has been initiated with contributions from several international design teams.

The engineering process of the IFMIF high flux test module (HFTM) demands concurrent optimization with respect to neutronic, mechanical and thermal-hydraulic issues. The cooling technology of the HFTM makes use of helium gas flowing in mini-channels along the specimen capsules. The coolant gas flows in these mini-channels are characterized by transitional Reynolds numbers, high velocities and high heating rates. Further, the flow in the short channels is influenced by entrance effects on a relevant part of the full length. The present study addresses these issues by experimental and numerical simulations. To this end, the experimental facility ITHEX has been constructed, and several mini-channel test sections were investigated using a variety of experimental techniques. The numerical methods are validated against the experimental data, and general conclusions on the HFTM thermal-hydraulic layout are presented.

The present work summarizes efforts on the simulation of natural convection cooling within the IFMIF target and test cell. The simulations have been performed with the STAR-CD code using the k–ω high-Reynolds number turbulence model. A dedicated thermohydraulic model has been devised including Lithium loop components. Nuclear heat production has been calculated by the Monte-Carlo code McDeLicious for different parts of the target and test cell walls and was used as input for the STAR-CD simulations. Helium atmospheres at several pressures from 0.1 to 10−5 MPa have been investigated. In order to limit the maximum temperature of the concrete walls to 80 °C it was necessary to add thermal insulation layers to the hot Lithium loop surfaces and a conceptual system of two cooling layers in different depths of the concrete walls.

This paper is devoted to the neutronics analysis of the electron cyclotron wave (ECW) launching system installed in the ITER upper port. The launcher design developed by FOM Rijnhuizen with a twisted arrangement of straight waveguides was selected to facilitate the maintenance of the ECW system by avoiding mitre bends. Two types of ECW launcher designs were considered in this paper: limited and full focusing. Three dimensional MCNP models of the ECW launchers were developed with high accuracy directly from the CAD geometry models by means of an interface programme. A shielding analysis was performed to estimate the length and thickness required for the shield around the waveguides. A two-step method making use of surface sources and point detectors was applied in the neutron streaming calculations in order to estimate the radiation load on the CVD diamond window. The neutron and gamma fluxes, the nuclear heat density, and the helium production were calculated. The results show that the design limits both for the limited and the full focusing ECW models are met.

The load carrying capacity of the pressure vessel head to withstand a postulated in-vessel steam explosion has been investigated. Former results have been confirmed and extended. For a quite conservative reference case the mechanical energy release which the head can carry turns out to be at least 1 GJ. The mechanical energy release from in-vessel steam explosions postulated can be expected to be significantly lower, and therefore will not rupture the vessel head and not endanger the containment integrity.

Nuclear and thermal–hydraulic analyses have been performed with the objective to further optimise the high-flux-test-module (HFTM) design. The nuclear analyses showed that the volume of 0.5 l with a damage rate dpa/fpy>20 can be achieved, if the HFTM has a contour like a semi ellipsoid. The thermal–hydraulic analyses focused on the experimental validation of the helium cooling concept. Integral heat transfer experiments revealed that the specimens can be kept at temperatures up to 600 °C. Dedicated experiments are being designed with the objective to provide data for the validation of the direct numeric FE simulation of the helium flow field, the heat transfer and the pressure losses.

The load carrying capacity of the pressure vessel head to withstand an in-vessel steam explosion is investigated. Firstly, as a key problem, the impact of molten core material against the vessel head is studied by model experiments scaled down 1:10. Structural details are considered carefully. The results are converted to reactor dimensions using similarity theory. This approach was checked by simplified liquid-structure impact experiments in different scale. Secondly, the upward acceleration of molten core material is studied by computational models. As results the mechanical energies which the vessel head can withstand are presented.

The most severe consequence of a pressurized water reactor in-vessel steam explosion is a molten fuel slug impact against the head of the reactor pressure vessel that could cause a failure of this head and lead to missiles endangering the reactor containment. An investigation is described that attempts to determine the maximum slug impact that a vessel head is capable of withstanding without failing and, consequently, without impairing the containment safety-related function. Preliminary theoretical assessments are presented that suggest that the head might be able to withstand rather strong impacts and that the shape of the fuel slug will have only a moderate influence on the results, provided the upper internal structures are taken into account. A low sensitivity against the slug shape is an essential prerequisite for a reliable safety proof. However, investigations primarily based on computational models are not sufficient; therefore, an investigation concept is proposed that relies on model experiments in which the geometry is scaled down by factors of 10 and 20, respectively. Theoretical and experimental investigations for liquid-structure impact problems in different scales are discussed to assess the degree of similarity that can be obtained. Finally, model experiments are described in some detail simulating themore » molten fuel slug impact on the vessel head.« less

A steam explosion in a reactor pressure vessel could cause a strong molten fuel slug impact of the vessel head, whose failure would question the integrity of the reactor containment. The capacity of a typical pressurized water reactor vessel head to carry slug impacts is therefore investigated. The slug interaction with the complex upper internal structures is taken into account. This interaction increases the capacity of the head and reduces its parameter sensitivity. Rough theoretical assessments are presented in this paper. Model experiments are being prepared to obtain reliable results. The similarity conditions for slug-structure impact problems are discussed

The development of the conceptual design of the IFMIF Target and Test Cell (TTC) is briefly summarized by outlining the previous reference TTC design and the current Modular TTC (MTC) concept. Based on the MTC concept, the latest progresses of the preliminary engineering design of the key TTC components, including the TTC vessel, the Top Shielding Plugs (TSPs), the Removable Intermediate Ring (RIR), and the Test Module Interface Head (TMIH), are described. A specimen flow, based on handling requirements of the High Flux Test Module (HFTM), between the TTC, Access Cell (AC), Test Module Handling Cell (TMHC) and the Post Irradiation Examination (PIE) facilities is proposed as well as the function of the AC and TMHC is preliminary defined. The TMHC is proposed to be divided into a Component Handling Cell (CHC) and a Rig Handling Cell (RHC) regarding the dimension differences of the components to be handled inside of the cells. The recycling of the irradiated specimens for another campaign of irradiation is also considered in this specimen flow.Copyright © 2010 by ASME