Power Exhaust Concepts and Divertor Designs for Japanese and European DEMO Fusion Reactors

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan-Europe Broader Approach DEMO Design Activity. A common critical issue is the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling. Different exhaust concepts in the main plasma and divertor have been developed for JA and EU DEMOs. JA proposed a conventional closed divertor geometry to challenge large Psep/Rp handling of 30-35 MWm-1 in order to maintain the radiation fraction in the main plasma at the ITER-level (fradmain = Pradmain/Pheat ~0.4) and higher plasma performance. EU challenged both increasing fradmain to ~0.65 and handling the ITER-level Psep/Rp in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding Psep = 250-300 MW and 150-200 MW, respectively. Both results showed that large divertor radiation fraction (Praddiv/Psep 0.8) was required to reduce both peak qtarget ( 10MWm-2) and Te,idiv. In addition, the JA divertor performance with EU-reference Psep of 150MW showed benefit of the closed geometry to reduce the peak qtarget and Te,idiv near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130-200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs

Study on Tokamak Plasma Performance for Fusion Energy Development Scenario

報告番号: 乙16204 ; 学位授与年月日: 2005-03-09 ; 学位の種別: 論文博士 ; 学位の種類: 博士(科学) ; 学位記番号: 第16204号 ; 研究科・専攻: 新領域創成科学研究

トカマク型核融合炉開発シナリオの基盤となる炉心プラズマ研究

University of Tokyo (東京大学

Conceptual Design for Higher Capability of the Tritium Production by the Honeycomb Structure Blanket of JA DEMO

The conceptual design of the breeding blanket with a honeycomb structure has been created with pressure tightness against in-box Loss-of-coolant accident based on a water-cooled solid breeder. In the previous design, the breeding area of the module was divided into 0.1-m-squared cells with rib structure. As a honeycomb structure is higher in pressure tightness than a square prism structure, the area for filling the mixed pebbles breeder of Li2TiO3 pebbles and Be12Ti ones can be enlarged. Then, the overall TBR is improved to increase the packing ratio of the tritium breeding material.In the created blanket, the capabilities of the pressure tightness, tritium breeding and heat removal are studied using interaction analyses of the neutronics, stresses and fluid dynamics analysis. As a result, a rib with the thickness of 0.015 m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. The packing factor of the mixed pebbles breeder increase to 77 % from 68 % by changing the rib structure from a square prism structure to a honeycomb structure. From the 3D neutronics analysis results, the target of the overall TBR (>1.05) is achievable. The cooling system for the created blanket is designed by fluid dynamics analysis based on the PWR water conditions which are the coolant temperature of 290 - 325 ºC and the operation pressure of 15.5 MPa, respectively. In addition, the tritium extraction system in the created blanket is proposed together with the purge gas system which does not clog the holes. The saturated time of the tritium extraction is also estimated to grasp the tritium inventory in the breeding area.14th International Symposium on Fusion Nuclear Technology (ISFNT-14

A Fast Discharge Scheme of Toroidal Field Coils for Fusion Demo Reactors

This paper describes an emergency fast discharge scheme of toroidal field (TF) coils of fusion demo reactors for reduction of induced voltage applied to turn insulations of conductors. TF coils are divided into serially connected segments that are electrically isolated from each other and only the coil segment having a failed coil is rapidly discharged. It was found from a circuit current analysis that this discharge scheme enables to reduce the insulation voltage with a factor of ~0.6 or less and which would contribute to ensure reliability of the turn insulations

Development of a plant system concept for a water-cooled fusion DEMO plant

The primary cooling water systems (CWS) of the blanket and the divertor is expected to have several amount of tritium because of permeation from the core plasma. In this research, the required performance of DEMO tritium removal facility to control the primary tritium concentration has been calculated and it is shown that existing devices can be applied to DEMO. The primary tritium is also expected to permeate the pipes to the secondary, thirdly CWS and finally, discharged into seawater. The evaluation of discharged tritium rate into seawater has been done. Because of tritium transport model from water to water is not well known, gas to gas calculation model is used. The results show that a heat exchanger decreases the tritium permeation rate by 1 order of magnitude, and H2 addition to the upstream cooling water also decreases the tritium permeation rate by 3 orders of magnitude.13th International Symposium on Fusion Nuclear Technolog