Design development and analytical assessment of LLCB TBM in Russian Federation during 2012–2013

The activity on the design, analysis, and R&D for the test blanket module (TBM) with lead–lithium (LL) eutectic coolant and ceramic breeder (CB) was performed in the Russian Federation (RF) according to the technical program of cooperation between the leading research institutes of India (“leader” of the LLCB TBM concept) and RF (“partner”). During the period of 2012–2013, the joint efforts of the RF and Indian specialists were focused on the development of the TBM's basic design with an optimal set of parameters (in particular for testing on both H-H and H-D operation phases of International Thermonuclear Experimental Reactor (ITER) machine). This article briefly describes the results of the TBM design and analysis that have been obtained by the RF specialists (“NIKIET” and D.V. Efremov Institute) in support of the LLCB concept (both DEMO blanket and TBM itself). The main directions of this activity in RF institutes were as follows:

• –development of the TBM design taking into account the ability to manufacture the TBM elements (load-bearing casing, tritium-breeding zone, and attachment system);
• –thermal analysis (in both stationary and transient approaches) of TBM design options (four variations of helium and eutectic flowing directions);
• –structural analysis of TBM design elements for Inductive I operation mode; and
• –recommendations (based upon the results of comparative analysis) on the reference design to be used on further stages of concept development.

The critical issues and further plans on the development of LLCB TBM and corresponding DEMO blanket in the RF are also presented in this article.     

R&D status on Water Cooled Ceramic Breeder Blanket Technology

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li₂TiO₃ pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.     

Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.     

Overview of TBM R&D activities in India

In India, development of Lead–Lithium Ceramic Breeder (LLCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) towards DEMO reactor. The LLCB TBM will be tested from the first phase of ITER operation (H-H phase) in one-half of an ITER port no. 2. The Indian TBM R&D program is focused on the development of blanket materials and critical technologies: structural material (IN-RAFMS), breeding materials (Pb–Li, Li₂TiO₃), development of technologies for Lead–Lithium cooling system (LLCS), helium cooling system (HCS), tritium extraction system (TES) and TBM related fabrication technologies. This paper will provide an overview of LLCB TBM R&D activities under progress in India.     

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.     

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.     

中国氦冷固态增殖剂实验包层模块材料研究进展

在未来核聚变反应堆中,为补充氚的消耗,需要在核聚变堆的包层中进行氚的在线增殖,以维持核聚变反应的持续进行。为验证这一关键技术,在国际热核聚变实验堆（ITER）上开展了ITER TBM计划（实验包层项目）。作为ITER计划成员方之一,中方以中国氦冷固态增殖剂实验包层模块（HCCB TBM）概念参与ITER TBM计划。HCCB TBM现今进入初步设计阶段,而材料的制备技术和性能数据是支撑其结构设计、安全分析和服役工况评估的基础。本文综述和分析了HCCB TBM结构材料低活化铁素体/马氏体钢（RAFM钢）与功能材料氚增殖剂和中子倍增剂的研究现状,并对这些材料下一步的研究方向进行了展望。     

ITER氚增殖实验包层设计研究进展

国际热核实验反应堆(ITER)为人类开发聚变能提供重要的物理和工程技术实验平台,ITER氚增殖实验包层模块(TBM)技术是必须掌握的关键技术.参与ITER计划的成员国根据本国商用演示堆包层发展策略,分别提出了各自的实验包层概念,以便在ITER运行期间进行实验.本文对ITER-TBM目前已经开展和正在进行的主要设计研究工作进展进行总结,介绍了各方提出的设计方案、支撑设计的相关技术研究进展,以及合作实验窗口的分配现状.     

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.     

Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R&D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R&D progress on these areas are introduced here.     

Assessment of design limits and criteria requirements for Eurofer structures in TBM components

Eurofer97 is a Reduced Activation Ferritic-Martensitic (RAFM) steel developed for use as structural material in fusion power reactors blankets and in particular the future DEMOnstration power plant that should follow ITER. In order to evaluate the performances of the different blanket concepts in a fusion-relevant environment, the ITER experimental programme foresees the installation of dedicated Test Blanket Modules (TBMs), representative of the corresponding DEMO blankets, in selected equatorial ports. To be fully relevant, TBMs will have to be designed and fabricated using DEMO relevant technologies and will, in particular, use Eurofer97 as structural material.While the use of ferritic/martensitic steels is not new in the nuclear industry, the fusion environment in ITER poses new challenges for the structural materials. Besides, contrary to DEMO, ITER is characterised by a strongly pulsed mode of operation that could have severe consequences on the lifetime of the components. This paper gives an overview of the issues related to the design of Eurofer97 structures in TBM components, discussing the choice of reference Codes&Standards and the consistency of the design rules with Eurofer97 mechanical properties.     

Tritium transport modeling for breeding blanket: State of the art and strategy for future development in the EU fusion program

The design of the Test Blanket Modules for ITER and the breeding blanket for DEMO requires robust and accurate modeling tools. Transport phenomena through the blanket tritium cycle are complex and involve a large number of physical properties and parameters, many of which have not been determined yet with a level of accuracy adequate for design optimization. Similarly, the use of simplified models with experimentally determined lumped coefficients allows satisfactory predictions only in very limited range of operative conditions, strongly reducing their potential to be relevant to the DEMO design.Within the European Union fusion program a road map to develop such modeling tools has been defined with the purpose of supporting the design of the ITER Tritium Blanket System and to exploit the TBM experimental testing for extrapolation to DEMO. The roadmap includes the development of the simulation tools as well as the supporting validation and verification experiments that must be carried out in parallel. This paper gives an overview of the state of the art of tritium modeling tools for blanket design, proposes a structure of the tritium modeling tools in order to facilitate their development and identifies a realistic work plan to achieve their final delivery.     

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.     

Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak.The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port.This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.     

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the ²¹⁰Po and ²⁰³Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.     

Japanese contribution to the DEMO-R&D program under the Broader Approach activities

Several technical R&D activities mainly related to the blanket materials are newly launched as a part of the Broader Approach (BA) activities, which was initiated by the EU and Japan. According to the common interests for these parties in DEMO, R&Ds on reduced activation ferritic/martensitic (RAFM) steels as structural material, SiC_f/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out through the BA DEMO R&D program, in order to establish the technical bases on the blanket materials and the tritium technology required for DEMO design. This paper describes overall schedule of those R&D activities and recent progress in Japan carried out by JAEA as the domestic implementing agency on BA, collaborating with Japanese universities and other research institutes.     

Progress on design and R&D for helium-cooled ceramic breeder TBM in China

The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. Current progress on the design and R&D for Chinese helium-cooled ceramic breeder TBM (CN HCCB TBM) in China is presented. The main updated design and related R&D of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being carried out. Recent status of the components and fabrication technology development is also reported. The neutron multiplier Be pebbles, tritium breeder Li₄SiO₄ pebbles, and structure material CFL-1 are being prepared in the laboratory scale. The fabrication of 1/3 sized mock-up and construction of a He test loop are being carried out. The key technology development is proceeding to the large scale mock-up fabrication and demonstration tests toward on ITER testing.     

Development of the water cooled lithium lead blanket for DEMO

The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant.In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications.Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant.A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood.The priorities for further development of the WCLL blanket concept are identified in the paper.     

Design update,thermal and fluid dynamic analyses of the EU-HCPB TBM in vertical arrangement

In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group of ITER, the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM) is developed in Forschungszentrum Karlsruhe (FZK) to investigate DEMO relevant concepts for blanket modules.The three main functions of a blanket module (removing heat, breeding tritium and shielding sensitive components from radiation) will be tested in ITER using a series of four TBMs, which are irradiated successively during different test campaigns. Each HCPB TBM will be installed, with a vertical orientation, into the vacuum vessel connected to one equatorial port. As the studies performed up to 2006 in FZK concerned a horizontal orientation of the HCPB TBM, a global review of the design is necessary to match with the new ITER specifications.A preliminary version of the new vertical design is proposed extrapolating the neutronic analysis performed for the horizontal HCPB TBM. An overview of the new HCPB TBM vertical designs, as well as the preliminary thermal and fluid dynamic analyses performed for the validation of the design, are presented in this paper. A critical review of the results obtained allows us, in the conclusion, to prepare a plan for the future detailed analyses of the vertical HCPB TBM.     

The fundamental role of fluid dynamic analyses in the design of the solid EU Test Blanket Module

In the frame of the activities of the European Test Blanket Module Consortium of Associates, the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM), the so-called solid TBM, is developed in Karlsruhe Institute of Technology (KIT). In the EU experimental strategy, a series of 4 different HCPB TBMs will be connected to the dedicated equatorial port n.16 during the ITER lifetime. The ITER TBM program has to provide DEMO relevant experimental data for the main functions of the blanket modules of a future fusion reactor.The preliminary thermo-mechanical design assessment of the TBM box (based on transient, steady state and accidental analyses) has been presented. All along the design assessment phase the fluid dynamic analyses play a fundamental role for the TBM sub-components, the Breeder Units (BUs) and the manifolds (MF) stages. This paper highlights the methodology implemented for the Computational Fluid Dynamic (CFD) analyses in the TBM design life cycle, and presents the results and the impact on the overall performance evaluation of the HCPB TBM. The following models are presented in detail: the CFD model of the TBM First Wall and its application to a reduced scale First Wall, the 3D CFD model of the BUs, and the thermo fluid dynamic modelling of the manifold systems.