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1.
The core snubber, as a passive protection device, can suppress arc current and absorb stored energy in stray capacitance during the electrical breakdown in accelerating electrodes of ITER NBI. In order to design the core snubber of ITER, the control parameters of the arc peak current have been firstly analyzed by the Fink-Baker-Owren (FBO) method, which are used for designing the DIIID 100 kV snubber. The B-H curve can be derived from the measured voltage and current waveforms, and the hysteresis loss of the core snubber can be derived using the revised parallelogram method. The core snubber can be a simplified representation as an equivalent parallel resistance and inductance, which has been neglected by the FBO method. A simulation code including the parallel equivalent resistance and inductance has been set up. The simulation and experiments result in dramatically large arc shorting currents due to the parallel inductance effect. The case shows that the core snubber utilizing the FBO method gives more compact design. 相似文献
2.
《Fusion Engineering and Design》2014,89(9-10):2001-2007
This paper summarises the present status of the ITER tritiated waste management strategy under development in France. This paper describes the specific challenges posed by this radioactive waste containing tritium as well as the solutions planned for the various waste categories and the implementation expected for the ITER tritiated waste, including the features of the future interim storage facility called INTERMED. Several options to reduce temporary storage duration as well as to minimise out-gassing rates and tritium discharges into the environment are under study, the related issues and the preliminary results obtained are shown. The first lessons learned for fusion development and their extrapolation to future reactors are outlined based on four parameters: materials, operating temperature, fuel cycle efficiency and tritium removal technologies. 相似文献
3.
《Fusion Engineering and Design》2014,89(7-8):890-895
The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R&D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine. 相似文献
4.
《Fusion Engineering and Design》2014,89(2):122-130
A set of seven polycrystalline mirror samples retrieved from the JET tokamak has been cleaned in vacuum using a pulsed laser system. The surfaces of samples exposed to plasma during 2008–2009 campaigns as part of the second phase of a comprehensive first mirror test contained a mixture of carbon, beryllium and tritium. For this reason, the samples were treated in a vacuum chamber constructed specially for this purpose. In some cases mirrors show an increase of the specular reflectivity after cleaning, though beryllium and carbon deposits were not fully removed. Additionally, three samples coated in PISCES-B with a 110–120 nm beryllium layer were subjected to laser cleaning tests as well. 相似文献
5.
《Fusion Engineering and Design》2014,89(9-10):1969-1974
The test blanket module port plug (TBM PP) consists of a TBM frame and two TBM-sets. However, at any time of the ITER operation, a TBM set can be replaced by a dummy TBM. The frame provides a standardized interface with the vacuum vessel (VV)/port structure and provides thermal isolation from the shield blanket. As one of the plasma-facing components, it shall withstand heat loads while at the same time provide adequate neutron shielding for the VV and magnet coils. The frame design shall provide a stable engineering solution to hold TBM-sets and also provide a mean for rapid remote handling replacement and refurbishment. This paper presents main design features of the conceptual design of TBM PP with two dummy TBMs. Also analysis results are summarized to evaluate shielding, hydraulic, and thermal and structural performances of the TBM PP design. 相似文献
6.
《Fusion Engineering and Design》2014,89(9-10):2304-2308
In the framework of a Fusion for Energy (F4E) grant, a test campaign started in 2012 in order to assess the performance of the in-vessel viewing system (IVVS) probe concept and to verify its compatibility when exposed to ITER typical working conditions. ENEA laboratories went through with several tests simulating high magnetic fields, high temperature, high vacuum, gamma radiation and neutron radiation.A customized motor has been adopted to study the performances of ultrasonic piezo motors technology in high magnetic field conditions. This paper reports on the testing activity performed on the motor in a multi Tesla magnetic field. The job was carried out in a test facility of ENEA laboratories able to achieve 14 T. A maximum field of 10 T, fully compliant with ITER requirements (8 T), was applied. A specific mechanical assembly has been designed and manufactured to hold the motor in the region with high homogeneity of the field. Results obtained so far indicate that the motor is compatible with high magnetic fields, and are presented in the paper. 相似文献
7.
《Fusion Engineering and Design》2014,89(9-10):2024-2027
Korea has designed a Helium-Cooled Ceramic Reflector (HCCR)-based Test Blanket System (TBS) for International Thermonuclear Experimental Reactor (ITER). Among seven selected reference accidents in Korean TBS, in-box loss of coolant accident (LOCA) is one of them. This is initiated by a double-ended break of the coolant pipe in the Breeding Zone (BZ), pressurizing the BZ box structure, causing pressurization of the Tritium Extraction System (TES) and purging of pipelines. When the accident is detected, the Plant Safety System (PSS) isolates the Helium Cooling System (HCS) and TES, and requests plasma shutdown to Fusion Power Shutdown System (FPSS). To prevent aggravating failure of the system, the safety function is automatically activated when the accident is detected, the device being the isolation valve of HCS and TES. One important observation of this accident is that instant isolation is not a good measure to take. In terms of the possibility of aggravating failure, system isolation is an important safety procedure but isolated TES volume is exposed to high pressure and temperature conditions in the early move of the accident transient. The result of system safety analysis shows that delayed isolation keeps the system safe for a while. In this article, given the preliminary accident analysis results for the current HCCR TBS, case studies were performed regarding the delayed isolation timing effect. For this transient simulation, Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used. 相似文献
8.
《Fusion Engineering and Design》2014,89(9-10):2257-2261
The ITER Tokamak Cooling Water System (TCWS) provides coolant for blankets and divertor. The blanket system consists of 440 blanket modules (BMs). The blanket manifold consists of a system of seamless pipes arranged in bundles and routed in poloidal direction from the upper ports of the Vacuum Vessel (VV) to the bottom of the machine. In each of the 18 upper ports there are 20 inlet and 20 outlet pipes, which split at the port exit in two directions, supplying cooling water to either the inboard or the outboard blanket modules. The manifold is routed between the VV and BMs. Branch pipes provide the connection between the manifold and the blanket cooling circuits through a coaxial connector welded to the shield block. A complex, sequential installation sequence has been developed in order to enable the assembly. Once installed the manifold is considered a semi-permanent component, but since failure would prevent ITER operation a maintenance strategy has been planned. 相似文献
9.
《Fusion Engineering and Design》2014,89(9-10):2294-2298
ITER standards Tesini (2009) require hardware mock-ups to validate the Remote Handling (RH) compatibility of RH class 1- and critical class 2-components. Full-scale mock-ups of large ITER components are expensive, have a long lead time and lose their relevance in case of design changes. Interactive Virtual Reality simulations with real time rigid body dynamics and contact interaction allow for RH Compatibility Assessment during the design iterations.This paper explores the use of interactive virtual mock-ups to analyze the RH compatibility of heavy component handling and maintenance. It infers generic maintenance operations from the analysis and proposes improvements to the simulator capabilities. 相似文献
10.
《Fusion Engineering and Design》2014,89(4):442-455
The first detailed Computational Fluid Dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part I we introduce the problem, define the model together with its input and discuss the results with particular reference to the hydraulics of the water coolant. The pressure drop across the panel is computed, together with the distribution of the flow among the different channels. Different design options are studied, with particular reference to the minimization of stagnation/recirculation regions. 相似文献