A helium-cooled divertor concept for DEMO, which is currently being developed at the Karlsruhe Institute of Technology, uses a modular structure of tungsten 9-finger units composed of smaller ...individual one-finger modules. As the development of the I-finger design is so far advanced, the work currently focuses on the manufacturing technology of a larger unit, the 9-finger module. The requirements for a larger grouping of individual cooling fingers are associated with the three-dimensional dimensions and orientations of all components in the assembly; their inaccuracy will affect the He flow distribution and cooling capacity of the divertor. In this paper, the necessary production steps, the order of assembly, and the principle of SATIR non destructive examination are described, as a result of a technological study.
•SATIR tests on DEMO divertor fingers (integrating or not He cooling system).•Millimeter size artificial defects were manufactured.•Detectability of millimeter size artificial defects was ...evaluated.•SATIR can detect defect in DEMO divertor fingers.•Simulations are well correlated to SATIR tests.
Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10MW/m2 in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La2O3 (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers.
The out-of-pile HEBLO experiments for thermomechanical loads on a small-scale test section based on the EU HCPB blanket concept were planned and realized. The first experiment series under ...steady-state conditions was accomplished successfully. The results of 2D simulation calculations show good agreements between calculated and measured temperatures. The following second HEBLO experiment series under temperature transient conditions has been defined and started successfully.
The dual-coolant (DC) blanket—characterised by its simple construction, simple function, and high thermal efficiency—is one of the EU advanced blanket concepts to be investigated in the frame of the ...long-term power plant conceptual study (PPCS). Its basic concept is based on the use of helium-cooled ferritic steel structure, the self-cooled Pb–17Li breeding zone, and SiC/SiC flow channel inserts, serving as electrical and thermal insulators. The present work on PPCS is drawn extensively on the preparatory study on plant availability carried out in 1999 with an objective to perform the conceptual design of the DC blanket concept where some details are to be selected in accordance with the overall strategy, which allows an extrapolation of the present knowledge between the near-term solutions (helium-cooled pebble bed (HCPB), water-cooled lead–lithium (WCLL) blanket concepts), and the very advanced self-cooled Pb–17Li SiC/SiC (SCLL) blanket concept. In the PPCS the reactor power is adapted to a typical size of commercial reactors of 1500 MWe which requires iterative calculations between the blanket layout and the system code analysis. The results of the first iteration are reported. This work is under the coordination of FZK in co-operation with CEA, EFET, IBERTEF, UKAEA, VTT Processes and VR.
Two variants of the Helium-cooled Pebble Bed (HCPB) blanket and an advanced version of the Dual Coolant Lithium Lead (DCLL) blanket have been investigated in the framework of the EU Power Plant ...Conceptual Study-Availability (PPA) with the main objective to explore their potential for a long lifetime and high power loading levels. This work presents the related neutronic analyses performed on the basis of three-dimensional Monte Carlo calculations for the PPA reactor to assess and optimise the nuclear performance of the considered blanket concepts.
A modular helium-cooled divertor design HEMJ (helium-cooled modular divertor concept with multiple-jet cooling) for the “post-ITER” demonstration (DEMO) fusion reactor has been developed at the ...Forschungszentrum Karlsruhe. The design goal is to withstand a surface heat flux of at least 10
MW/m
2 at an acceptable pumping power.
A conical design of a brazed joint between two structural components of the HEMJ finger module which are made of different materials has been investigated. This new transition piece design should withstand at least 1000 temperature load cycles between operating and room temperatures. Due to the large mismatch of the thermal expansion coefficients (TECs) of the different materials used, high thermal stresses caused by the thermocyclic loads could lead to the plasticization of both materials in the joint region. To demonstrate the feasibility of this transition piece design, a systematic investigation is required, which includes a numerical simulation, the choice of the brazing material, a study of the brazing technology, and thermocyclic tests of the finger mock-up.
This paper shall present a method of numerical investigation as the first step of investigation. Plastic stress calculations are performed using the commercial software ANSYS
® taking into account thermocyclic as well as internal pressure loads. The calculation results, in particular the plastic behavior of the brazed joint, will be discussed.
The EU advanced dual coolant blanket concept Norajitra, Prachai; Bühler, Leo; Fischer, Ulrich ...
Fusion engineering and design,
11/2002, Letnik:
61
Journal Article, Conference Proceeding
Recenzirano
The advanced dual coolant (A-DC) blanket is one of the EU advanced concepts to be investigated in the frame of the long-term power plant conceptual study (PPCS). Its basic concept—following the ...ARIES-ST concept—is based on the use of helium-cooled ferritic steel structure, the self-cooled Pb–17Li breeding zone, and SiC/SiC flow channel inserts. The latter serves as electrical and thermal insulators and therefore minimize the pressure losses and enable a relatively high Pb–17Li exit temperature leading to a high thermal efficiency. The present work on PPCS is drawn extensively on the preparatory study on plant availability (PPA) carried out in 1999 where a maximum neutron wall load of 5 MW/m
2 (corresponding maximum surface heat load of 0.9 MW/m
2) was given in the reference case of the A-DC blanket. In the following stage of PPCS the A-DC blanket is normalized and adapted to a typical size of commercial reactors (e.g. 1500 MWe) which requires iterative calculations between the blanket layout and the system code analysis. The status of the work with some idea improvements is reported.
A modular He-cooled divertor concept for DEMO has been developed at Karlsruhe Institute of Technology (KIT). The design goal is to achieve a DEMO-relevant high heat flux of 10 MW/m 2 . The reference ...design HEMJ (He-cooled modular divertor with multiple-jet cooling) uses small tungsten-based cooling finger modules. The divertor parts are connected by brazing. They are cooled by helium impinging jets. After the performance and functionality of design has been confirmed through numerous high heat flux (HFF) tests, the current R&D focuses on the manufacturing technology in order to arrive at a robust design and a mass-production of parts. In this paper, newly developed innovative technologies for manufacturing tungsten-based divertor modules (e.g. deep drawing, powder injection molding) as well as for joining the components of different materials shall be presented.
The helium-cooled pebble bed (HCPB) blanket for a fusion DEMO reactor is based on the use of separate small lithium orthosilicate and beryllium pebble beds placed between radial toroidal cooling ...plates. The cooling is provided by helium at 8 MPa. The tritium produced in the pebble beds is purged by the flow of helium at 0.1 MPa. The structural material is martensitic steel. It is foreseen, after extensive R&D, to test in ITER a blanket module based on the HCPB design which is one of the two European proposals for the ITER Test Blanket Programme. To facilitate the handling operations, the blanket test module (BTM) is bolted to a surrounding water cooled frame fixed to the ITER shield blanket back plate. For the design of the test module, 3D Monte Carlo neutronic calculations and thermo-hydraulic and stress analyses for the operation during the basic performance phase (BPP) and during the extended performance phase (EPP) of ITER have been performed. The behaviour of the test modules during LOCA, LOFA and electromagnetic transients has been investigated. Conceptual designs of the required ancillary loops have been performed.
For the helium-cooled pebble bed (HCPB) blanket, which is one of the two reference concepts studied within the European Demo Development Program, a comprehensive finite element (FEM) structural ...analysis has been performed. The analysis refers to the steady-state operating conditions of an outboard blanket segment. On the basis of a three-dimensional model of radial–toroidal sections of the segment box, thermal stresses caused by the temperature gradients in the blanket structure have been calculated. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions as well as an accidental over-pressurization of the blanket box have been accounted for. The stresses caused by a central plasma major disruption from an initial current of 20 MA to zero in 20 ms have been also taken into account. Radiation-induced dimensional changes of breeder and multiplier material caused by both helium production and neutron damage, have also been evaluated and discussed. All the above loads have been combined as input for a FEM stress analysis and the resulting stress distribution has been evaluated according to the American Society of Mechanical Engineers (ASME) norms.
