Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.     

Dynamic thermal-hydraulic modelling of the EU DEMO HCPB breeding blanket cooling loops

A global, system-level thermal-hydraulic model of the EU DEMO tokamak fusion reactor is currently under development and implementation in a suitable software at Politecnico di Torino, including the relevant heat transfer and fluid dynamics phenomena, which affect the performance of the different cooling circuits and components and their integration in a consistent design. The model is based on an object-oriented approach using the Modelica language, which easily allows to preserve the high modularity required at this stage of the design. The first module of the global model will simulate the blanket cooling system and will be able to investigate different coolant options and different cooling schemes, to be adapted to the different blanket systems currently under development in the Breeding Blanket (BB) project. The paper presents the Helium-Cooled Pebble Bed (HCPB) module of the EU DEMO blanket cooling loops system model. The model is used to compare different schemes for the cooling of the different components of the HCPB BB, and to suggest improvements aimed at optimizing the pumping power required by the cooling system. The model is then used to analyse a pulsed scenario, characteristic of the EU DEMO operation.     

聚变堆增殖包层概念特征比较研究

在广泛调研国际上现有的聚变堆包层概念的基础上,对聚变堆包层的发展现状和需要解决的关键问题进行了总结。从工程可行性、氚增殖提取与控制特征、经济性以及安全和环境影响方面对固态和液态氚增殖剂包层进行了比较分析,系统阐述了各种包层概念的优缺点,从现实可行性与发展潜力方面为未来聚变堆包层概念发展提出建议。     

聚变发电反应堆双冷锂铅包层热工水力学设计与分析

依据结构设计和中子学计算结果给出了聚变发电反应堆FDS-Ⅱ双冷锂铅(DLL)包层热工水力学设计方案。采用数值计算软件对液态金属增殖区LiPb流场和第一壁热-结构等进行了模拟,并对功率转换系统的热效率进行了计算。通过分析评估,证实该包层热工水力学方案能较好地实现FDS-Ⅱ聚变发电反应堆预期目标。     

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module(TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development(RD) is focused on two types of breeding blanket concepts: lead–lithium ceramic breeder(LLCB) and helium-cooled ceramic breeder(HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic RD program for DEMO relevant technology development. In the HCCB concept Li_2TiO_3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept(case-1), the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept(case-2), the pebble bed is vertically(poloidal–radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2 D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations.Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.     

多功能聚变裂变混合实验堆FDS-MFX氦冷包层三维中子学初步设计与分析

FDS-MFX(Multi-Functional eXperimental fusion-fission hybrid reactor)是一个基于现实可行技术的多功能聚变裂变混合实验堆概念,分3个阶段相继开展实验研究,分别采用纯氚增殖包层、铀燃料包层和乏燃料包层.本文重点对其中铀燃料包层后期阶段中高浓缩铀模块的摆放方式...     

Tritium management and anti-permeation strategies for three different breeding blanket options foreseen for the European Power Plant Physics and Technology Demonstration reactor study

In DT fusion reactors like DEMO, the commonly accepted tritium (T) losses through the steam generator (SG) shall not exceed about 2 mg/d that are more than 5 orders of magnitude lower than the T production rate of about 360 g/d in the breeding blanket (BB). A very effective mitigation strategy is required balancing the size and efficiency of the processes in the breeding and cooling loops, and the availability and efficiency of anti-permeation barriers. A numerical study is presented using the T permeation code FUS-TPC that computes all T flows and inventories considering the design and operation of the BB, the SG, and the T systems. Many scenarios are numerically analyzed for three breeding blankets concepts – helium cooled pebbles bed (HCPB), helium cooled lithium lead (HCLL), and water cooled lithium lead (WCLL) – varying the T processes throughput and efficiency, and the permeation regimes through the BB and SG to be either surface-limited or diffusion-limited with possible permeation reduction factor. For each BB concept, we discuss workable operation scenarios and suggest specific anti-permeation strategies.     

聚变堆氚增殖层中子学分析

D-T聚变堆包层的主要功能包括氚增殖、能量转换射层蔽等,包层中子学设计的主要原则是满足聚变堆的氚自持,一般要求包层氚增殖比TBR＞1.1.使用与时间有关的扩散理论和本征函数展开方法,研究不同几何线度、6Li丰度的LI2O、LiPb包层材料14MeV源下的系统通量、氚增殖比影响,及在不同6Li丰度下6Li、7Li造氚随时间变化的规律.计算中使用了30群截面数据,微观数据来自ENDF/B-VI及JEF-2.2.     

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module(TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the RD activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices(ITER relevant and DEMO).The Indian Lead–Lithium Cooled Ceramic Breeder(LLCB) blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder(HCCB) is an alternative blanket concept that consists of lithium titanate(Li_2TiO_3) as ceramic breeder(CB) material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.     

Analysis of electromagnetic loads on EU-DEMO inboard and outboard blanket vertical segments

An analysis of the EM loads acting on a DEMO reactor configuration based on Multi Module Segment (MMS) design is presented in this work as part of the ongoing EU DEMO studies. Lorentz's forces and moments, both on the single module as well as on the complete blanket segment, are calculated for both the European HCPB and HCLL concepts. The system is analyzed considering a major central disruption scenario with a linear current quench of 42 ms using the ANSYS finite element software. The results are also compared to linear analyses to underline the effect of the non-linearity of the ferromagnetic materials.     

聚变高温制氢反应堆锂铅包层结构力学分析

在高温液态锂铅包层结构设计、热工水力学设计和中子学计算基础上,建立包层的三维有限元分析模型,应用商用有限元软件ANSYS对高温液态包层进行热-力结构耦合分析与应力评定。经计算第一壁材料ODS RAFM钢最高温度635℃,最大Von Mises应力379 MPa;包层结构材料RAFM钢最高温度508℃,最大Von Mises应力175 MPa;FCI材料最高温度950℃,最大Von Mises应力218 MPa。初步的分析结果表明结构设计方案是合理、可行的。     

聚变驱动次临界堆双冷嬗变包层结构设计与分析

给出聚变驱动次临界堆液态金属LiPb/He气双冷嬗变包层参考结构概念,采用了低活化铁素体/马氏体RAFM钢(如CLAM)作为结构材料、简单液态金属流道、两个独立氦气冷却系统以及燃料球/颗粒等设计方案。重点分析了嬗变包层第一壁、重金属区与裂变产物嬗变区的结构设计特点。     

聚变驱动次临界堆双冷嬗变包层中子学设计与分析

对聚变驱动次临界堆的多功能双冷核废料嬗变包层进行了中子学设计和分析,设计目标是:①氚和钚燃料自持;②较少的初装料得到较高的废料嬗变率。使用的程序是自主开发的多功能中子输运/燃耗/优化程序VisuaIBUs1.0,相应的数据库是175群中子/42群光子的多群数据库HENDL1.0/MG。     

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a ⁶Li enrichment of 90%. Assuming a ⁶Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.     

Neutronic optimization for a blanket of the laser fusion reactor,SENRI-I

Neutronic calculations were performed to optimize the SENRI blanket in terms of energy multiplication as well as tritium breeding ratio. The blanket employs a thick ( 64-cm) Li layer as breeder/coolant. Three approaches were taken here to achieve the goal: (1) reduction of⁶Li in the lithium, (ii) replacement of the Li layer by a molten-salt (flibe) layer, and (iii) shipment of excess tritium to a nonbreeding blanket. It was found that the excess tritium produced in the SENRI blanket could be used effectively to obtain additional power by fueling a nonbreeding D-T reactor.     

Neutronics optimization study for D-D fusion reactor blanket/shield

Position-dependent optimization calculations have been carried out on a D-D fusion reactor blanket/shield to maximize the energy gain in the blanket and to minimize the atomic displacement rate of the copper stabilizer in the superconducting magnet. The results obtained by using the optimization code SWAN indicate (1) the advantage of D₂O coolant over H₂O coolant with respect to increasing the energy gain, and (2) the difference in the optimal shield distributions between D-T and D-D neutron sources. The possibility of improving both the energy gain and radiation shielding characteristics is also discussed.     

聚变驱动次临界堆双冷嬗变包层热工水力学参数设计与分析

聚变驱动次临界堆双冷嬗变包层是一个以氦气和液态金属LiPb为冷却剂,以嬗变核废料为主要目的的多功能包层。依据功率平衡模型对不同工况优化的基础上,对该包层热工系统参数进行了设计分析。采用三维商用计算流体力学程序对第一壁和高功率密度区中液态LiPb的流场进行数值模拟计算,给出了优化的典型热工水力参数。     

Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 °C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 °C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions.     

New progress on design and R&D for solid breeder test blanket module in China

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.