Frederik Arbeiter

IFMIF: overview of the validation activities

The Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF), an international collaboration under the Broader Approach Agreement between Japan Government and EURATOM, aims at allowing a rapid construction phase of IFMIF in due time with an understanding of the cost involved. The three main facilities of IFMIF (1) the Accelerator Facility, (2) the Target Facility and (3) the Test Facility are the subject of validation activities that include the construction of either full scale prototypes or smartly devised scaled down facilities that will allow a straightforward extrapolation to IFMIF needs. By July 2013, the engineering design activities of IFMIF matured with the delivery of an Intermediate IFMIF Engineering Design Report (IIEDR) supported by experimental results. The installation of a Linac of 1.125 MW (125 mA and 9 MeV) of deuterons started in March 2013 in Rokkasho (Japan). The worlds largest liquid Li test loop is running in Oarai (Japan) with an ambitious experimental programme for the years ahead. A full scale high flux test module that will house ~1000 small specimens developed jointly in Europe and Japan for the Fusion programme has been constructed by KIT (Karlsruhe) together with its He gas cooling loop. A full scale medium flux test module to carry out on-line creep measurement has been validated by CRPP (Villigen).

Advances in the physics basis for the European DEMO design

In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

The accomplishment of the Engineering Design Activities of IFMIF/EVEDA: The European-Japanese project towards a Li(d,xn) fusion relevant neutron source

The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept, the re-location of the quench tank outside the test cell (TC) with a reduction of tritium inventory and a simplification on its replacement in case of failure, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction, and the water cooling of the liner and biological shielding of the TC, enhancing the efficiency and economy of the related sub-systems. In addition, the maintenance strategy has been modified to allow a shorter yearly stop of the irradiation operations and a more careful management of the irradiated samples. The design of the IFMIF plant is intimately linked with the EVA phase carried out since the entry into force of IFMIF/EVEDA in June 2007. These last activities and their on-going accomplishment have been thoroughly described elsewhere (Knaster J et al [19]), which, combined with the present paper, allows a clear understanding of the maturity of the European–Japanese international efforts. This released IFMIF Intermediate Engineering Design Report (IIEDR), which could be complemented if required concurrently with the outcome of the on-going EVA, will allow decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community.

A Comparative Study of Turbulence Models for Conjugate Heat Transfer to Gas Flow in a Heated Mini-Channel

Numerical simulations have been carried out for low Reynolds number (Re = 6000–10000) gaseous flows in a heated annular mini-channel using the commercial CFD codes CFX V12.0 and Star-CD V4.10. The results were compared with in-house experiments. Detailed comparisons were made between the low Reynolds number k-ϵ, k-ω-SST, and V2F of Star-CD and CFX in terms of numerical sensitivities and model performance. Particularly, it was found that the inlet turbulence conditions can have a significant impact on the simulation results in the downstream flow for Star-CD low-Re k-ϵ and V2F models. For heat transfer predictions, the Star-CD V2F model provides the best predictions under low Reynolds numbers of about 6,000, while for Reynolds numbers of about 10,000 the k-ω-SST model performs better than others.

Overview of the HCPB Research Activities in EUROfusion

In the framework of the EUROfusion’s Power Plant Physics and Technology, the working package breeding blanket (BB) aims at investigating four different BB concepts for an EU demonstration fusion reactor (DEMO). One of these concepts is the helium-cooled pebble bed (HCPB) BB, which is based on the use of pebble beds of lithiated ternary compounds and Be or beryllides as tritium breeder and multiplier materials, respectively, EUROFER97 as structural steel and He as coolant. This paper aims at giving an overview of the EU HCPB BB Research and Development (R&D) being developed at KIT, in collaboration with Wigner-RCP, BUTE-INT, and CIEMAT. The paper gives an outline of the HCPB BB design evolution, state-of-the-art basic functionalities, requirements and performances, and the associated R&D activities in the areas of design, functional materials, manufacturing, and testing. In addition, attention is given also to the activities dedicated to the development of heat transfer augmentation techniques for the first wall and the corresponding testing. Due to their nature as design drivers, a brief overview in the R&D of key HCPB interfacing areas is given as well, namely, the tritium extraction and recovery system, the primary heat transfer and power conversion systems, and safety topics, as well as some specific activities regarding the integration of in-vessel systems through the BB. As concluding remarks, an outline of the standing challenges and future R&D plans is summarized.

Heat transfer and pressure drop measurements in channels roughened by variously shaped ribs on one wall

ABSTRACT Heat transfer and pressure drop measurements were conducted to study the thermal-hydraulics in a square, round-edged channel roughened by ribs (e/Dh = 0.0638, p/e = 10) on one wall at Reynolds numbers ranging from 5.0 × 104 to 2.5 × 105. Three variously shaped ribs were investigated: Transverse ribs with square cross sections, transverse ribs, and upstream directed 60° V-shaped ribs with round-edged rib front and rear surfaces. Friction factors, Nusselt number ratios, roughness functions, and the thermal performance were presented. The highest heat transfer and best thermal performance is reached by the upstream directed V-shaped ribs.

Neutronic Analysis of the IFMIF Tritium Release Test Module Based on the EVEDA Design

Abstract The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based intense neutron source to test fusion reactor materials under irradiation conditions expected to be experienced by a future fusion power plant (DEMO). The Tritium Release Test Module (TRTM) is intended for the irradiation of solid breeder ceramics as well as beryllium involving in-situ tritium release measurements in IFMIF. During the EVEDA (Engineering Validation Engineering Design Activities) phase, a detailed engineering design for the TRTM has been elaborated. A new 3-dimesional Monte Carlo geometry model of TRTM was prepared for a neutronic analysis directly from engineering CAD data using the McCad conversion software developed at KIT. The analysis was performed with the latest version of the Monte Carlo code McDeLicious, an enhancement to MCNP5 for IFMIF neutronics calculations, using a state-of-the-art nuclear data library FENDL-3. The result emphasizes the importance of the neutron reflector which should be placed behind TRTM in order to make the irradiation properties close to the European HCPB DEMO. Although the achievable dpa is lower than that expected in DEMO, the T/dpa and He/dpa values can be simulated very well when the neutron reflector is appropriately designed, in particularly by utilizing beryllium.

Overview of Material Challenges in IFMIF Test Cell Design

Abstract As the core region of IFMIF, the test cell (TC) suffers intense neutron and gamma irradiations. Major material challenges of the TC faced during engineering design phase are outlined and the current key material allocations are described. Actively cooled magnetite concrete is selected as the major biological shielding material for the TC, and actively cooled closed liner made of 316L stainless steel is selected to cover the complete TC internal surfaces. Material selections for sealing gaskets and electric insulations inside the TC are preliminarily defined based on dose rate maps at different locations. Metal based sealing gaskets and glass/ceramic electric insulations are applied in the areas with high dose rate, while organic based gaskets and conventional insulation materials can only be arranged behind sufficient biological shielding. Leak tight welding seams between removable interface shielding plugs and the TC liner are located in the region with very low helium generation rate (<0.01 appm/fpy) in steel so that cutting and re-welding during the complete IFMIF life span is guaranteed.

Numerical Simulations and Experiments on Gaseous Flow and Heat Transfer in Multiple Narrow Channels in Transition Region

The International Fusion Materials Irradiation Facility (IFMIF) is designated to generate a materials irradiation database for the future fusion reactors. The High Flux Test Module (HFTM) houses the test specimens which will be cooled with helium flows passing through several parallel narrow channels. The Reynolds numbers for individual sub-channels range from about 1000 to 8000. For most sub-channels the flow is in the transition region, which cannot be accurately predicted by using laminar or turbulence models. As a part of the HFTM validation activities, the thermo-hydraulic tests of a 1:1 single irradiation rig were carried out. The experiments also provide data for validating the CFD codes for using in the simulations of the whole HFTM. In this paper, the CFD simulations and experiments will be compared for helium flows at a Reynolds number between 3000 and 8000. The simulations were carried out with the commercial CFD code Ansys-CFX V14 using several turbulence models. It is found that, the CFX Gamma model with the transition onset Reynolds number of about 180 provides the best predictions for the transition flows.Copyright

Assessment of the Operational Dose Rate in Polymer Insulators in the Test Cell of the IFMIF Neutron Source

Insulators for electrical connections of the test modules in the test cell of the International Fusion Materials Irradiation Facility (IFMIF) neutron source will be exposed to intense ionizing radiation even if they are located away from the neutron source. The aim of this work was to check whether radiation-hard polymers, for example polyimide, would be sufficiently long-lasting in such an environment. The calculations presented here were survey calculations to aid with the decisions on possible insulator candidates. The calculations have been done with the McDeLicious code and a detailed 3-D neutronics model of the IFMIF test cell. Mesh tallies were utilized to calculate dose rate maps 10 cm below the ceiling of the test cell where the connectors would be located. The calculated dose deposited within one full power year would be higher than 200 MGy nearly everywhere in the test cell. This value excludes most polymers as candidates for insulators, but even polyimide would reach its limits. It was found that the calculated dose contribution from gamma rays was considerably dependent on the nuclear data library used for describing the lithium in the lithium loop. We have also checked the effect of an additional lead shield of 10 cm thickness which reduced the deposited dose by a factor of 2-5. A new neutronics model has become available recently which was derived from the latest IFMIF test cell design. We have performed preliminary calculations for dose rates in anticipated locations of connector insulators. The results confirm the high dose rate values calculated with the previous neutronics model.