共查询到20条相似文献,搜索用时 15 毫秒
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Wei Li Wenxi Tian Suizheng Qiu Guanghui Su Hong Jiao Yunqing Bai Hongli Chen Yican Wu 《Fusion Engineering and Design》2013,88(5):286-294
China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER. 相似文献
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《Fusion Engineering and Design》2014,89(7-8):1209-1212
Tritium monitoring in lithium–lead eutectic is of great importance for the performance of liquid blankets in fusion reactors. In addition, tritium measurements will be required in order to proof tritium self-sufficiency in liquid metal breeding systems. On-line hydrogen (isotopes) sensors must be design and tested in order to accomplish these goals.In this work, an experimental set up was designed in order to test the permeation hydrogen sensors at 500 °C. This experimental set-up allowed working with controlled environments (different hydrogen partial pressures) and the temperature was measured using a thermocouple connected to a temperature controller that regulated an electrical heater.In a first set of experiments, a hydrogen sensor was constructed using an α-iron capsule as an active hydrogen area. The sensor was mounted and tested in the experimental set up. In a second set of experiments the α-iron capsule was replaced by a welded thin palladium disk in order to minimize the death volume. The experiments performed using both membranes (α-iron and palladium) showed that the operation of the sensors in the equilibrium mode required at least several hours to reach the hydrogen equilibrium pressure. 相似文献
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Safe, reliable and efficient tritium management in the breeder blanket faces unique technological challenges. Beside the tritium recovery efficiency in the tritium extraction and coolant purification systems, the tritium tracking accuracy between the inner and outer fuel cycle shall also be demonstrated. Furthermore, it is self-evident that safe handling and confinement of tritium need to be stringently assured to evolve fusion as a reliable technique. The present paper gives an overview of tritium management in breeder blankets. After a short introduction into the tritium fuel cycle and blanket basics, open tritium issues are discussed, thereby focusing on tritium extraction from blanket, coolant detritiation and tritium analytics and accountancy, necessary for accurate and reliable processing as well as for book-keeping. 相似文献
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在中国向ITER(International Thermonuclear Experiment Reactor)实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)设计分析的基础上,通过对DFLL-TBM系统相关的瞬态事故如真空室内部冷却剂泄漏、TBM(实验包层模块)内部冷却剂泄漏以及真空室外部冷却剂泄漏事故进行计算分析,评价DFLL-TBM对ITER在热工方面对安全的影响.结果表明:当发生瞬态事故时,DFLL-TBM有能力通过热辐射将余热排出,且包层结构不会熔化.DFLL-TBM可满足ITER在热工方面对安全的要求. 相似文献
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《Fusion Engineering and Design》2014,89(7-8):1397-1401
Dynamic tritium concentration measurement in lead–lithium eutectic is of major interest for a reliable tritium testing program in ITER TBM and for an experimental proof of tritium self-sufficiency in liquid metal breeding systems. Potentiometric hydrogen sensors using different solid-state electrolytes for molten lead–lithium eutectic have been reported and tested by the Electrochemical Methods Lab at Institut Quimic de Sarria (IQS).In the present work the following ceramic elements have been synthesized and characterized by X-ray diffraction (XRD) in order to be tested as a Proton Exchange Membranes (PEM) H-probes: BaCeO3, BaCe0.6Zr0.3Y0.1O3−δ and Sr(Ce0.9–Zr0.1)0.95Yb0.05O3−δ. Potentiometric measurements of the synthesized ceramic elements have been performed shifting from a fixed hydrogen partial pressure at the working electrode to high purity argon. In this experimental campaign a fixed and known hydrogen pressure has been used in the reference electrode. The goal of these experiments is to evaluate the sensor response time when the hydrogen concentration in the environment is rapidly changed. All experiments have been done at 500 °C and 600 °C. The sensor constructed using the proton conductor element BaCe0.6Zr0.3Y0.1O3−δ exhibited stable output potential and its value was close to the theoretical value calculated with the Nernst equation. In contrast, the sensors constructed using the proton conductor elements BaCeO3 and Sr(Ce0.9–Zr0.1)0.95Yb0.05O3−δ showed higher deviations between experimental and theoretical data, and long response times. 相似文献
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Part of the ITER experimental setup will be testing of the first wall blanket module built from EUROFER 97 steel. Lifetime of this component should be predicted with help of defect assessment procedure. This work verifies the application of the R5 assessment code on the 2D test blanket module geometry by using finite element simulations. Verification is based on the evolution of C(t) parameter. Results exhibit good correspondence in predictions provided by R5 and finite element method for different thermo-mechanical loading conditions. These results therefore show applicability of R5 procedure on such complex geometries. 相似文献
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《Fusion Engineering and Design》2014,89(7-8):1158-1162
In LIBRETTO-2 test, evidence was obtained that helium bubbles nucleated and grew in the neutron irradiated PbLi probes. If such phenomenon occurs inside liquid metal (LM) breeding blanket channels, the study of its effect on tritium permeation and heat transfer in the near wall region will acquire utmost importance. The T4F research group has developed in the past a nucleation, growth and transport model for helium bubbles in LM flows, as well as a tritium transport model in such a multi-fluid system. In the present study, we are focused on the near-wall region analysis in order to obtain a wall function that allow reproducing the tritium permeation with coarse meshes and, hence, reduce the computational time. First, we perform some detailed CFD simulations of the near-wall region where bubbles might be attached. In these simulations, tritium diffusion processes as well as tritium recombination and dissociation are modelled. The analysis of such simulations allows us to further understand the complex phenomena and justify the use of simplified models. As a result, a new model for tritium transport across a LM–solid interface partially covered by helium bubbles is developed, implemented and validated. This simplified model can be seen as a wall function for the CFD simulation which substantially reduces computational time. 相似文献
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使用中子学程序系统VisualBUS和活化数据库EAF-99对DFLL-TBM的高级子模块DLL-TBM的活化特性进行了计算和分析,包括DLL-TBM各部件在不同停堆时间的活度、衰变余热和剂量率.活化计算所需要的三维中子能谱通过MCNP/4C中子/光子输运程序和国际原子能机构发布的FEND1.0数据库计算得到.在活化计算分析的基础上,参照欧洲聚变堆安全和环境评估(SEAFP)策略中有关核废料的处理标准评估了TBM各区材料在退役后的废料处理工作,包括核废料应该采用何种适当的方式进行处理及其被完全清除干净的可行性. 相似文献
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《Fusion Engineering and Design》2014,89(9-10):2088-2092
Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown. 相似文献
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ITER双功能液态锂铅实验包层系统故障模式影响分析 总被引:2,自引:2,他引:0
实验包层模块允许放置在ITER中实验的前提是其对ITER的安全以及对工作人员和环境不构成显著影响。ITER要求各参与方的实验包层模块在实验前必须提交安全分析报告,进而获取安全许可证。在中国双功能锂铅实验包层模块(DFLL-TBM)设计基础上,采用了故障模式影响分析(FMEA)方法对DFLL-TBM进行了安全评估与分析,得到所有可能导致严重后果的假设始发事件,验证了确定论安全分析所选择的三个参考事件可以包络所有的假设始发事件。 相似文献
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《Fusion Engineering and Design》2014,89(7-8):1119-1125
ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R&D activities for each TBM module with the auxiliary system are introduced.The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li4SiO4 pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R&D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled. 相似文献
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《Fusion Engineering and Design》2014,89(7-8):1068-1073
Korea has developed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) consisting of four sub-modules in an ITER. From the draft design of the side wall (SW) according to a thermal-hydraulic analysis, a mechanical analysis was performed considering a design channel pressure of 10 MPa. The SW comprised of sixteen grids with the seventeen partitions for the manifold function satisfied 1.5Sm of the allowable stress (Sm) according to RCC-MR code at the maximum stress region in the SW. In addition, an elastic analysis of the draft design of the back manifold (BM) was carried out, which supported the four sub-modules in the HCCR TBM and has the main inlet/outlet of the He cooling pipe, the measurement pipes, and He purge gas lines from the port cell. The results show that the maximum stress was higher than 1.5Sm, and the BM design has been modified to satisfy the BM function and requirements. 相似文献