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Present status of the conceptual design of the EU test blanket systems
Abstract
The EU Breeding Blanket Programme aims the testing of two blankets concept in ITER in form of Test Blanket Modules. In the equatorial port #16 the two EU TBMs – a solid and a liquid blanket concept – will be exposed to the plasma and the complex system of their auxiliary systems dedicated to heat and Tritium removal will be integrated in the surrounding ITER buildings. The development of the conceptual design of the EU TBM System is the main objective of the Grant F4E-2008-GRT-09 contract launched by F4E and assigned to a European Consortium. This paper presents an overview of the results after about 20 months of activities: namely, the design of the main sub-systems of the EU TBSs and a concept of integration in ITER.
... The Helium Cooled Pebble Bed (HCPB) blanket concept is one of the European Demonstration Power Plant (EU DEMO) blanket concepts currently under development [1]. At the present stage, the blanket concept employs helium as coolant, Li4SiO4 as breeder material, Be12Ti as neutron multiplier material, and EUROFER as structural material [2][3][4]. ...
... At the time when the experiment was designed, there were several cooling schemes taken under consideration for the DEMO blanket and FW design [1,3]. One of these schemes considers two independent cooling circuits, i.e., a configuration that should ensure a higher reliability. ...
... In any case, this analysis suggests that the maximum duration for which the flow is reduced to zero (total LOFA) during the experiment should not exceed 12 s to avoid surface temperatures over 550 °C. When comparing these results with those obtained for the DEMO blanket FW in [1], it can be seen that, for the present case, the temperature rise after 8 s was close to 60 °C as compared to 90 °C in the reference case. This can be explained by the cooling effect on the main flow of the 180° turn area where there is no surface heating. ...
The experimental investigation of a prototypical set-up simulating a loss of flow accident in a helium-cooled breeding blanket first wall mock-up under typical heat load conditions is presented. The experimental campaign reproduces the expected DEMO thermal-hydraulics conditions during normal and off-normal situations and aims at providing some insight into the fast transients associated with the loss of flow in the blanket first wall. The experimental set-up and the definition of the experimental matrix are discussed, including pre-test analysis performed in support of these activities. The major experimental results are discussed, and a procedure of using the acquired data for validating and calibrating the RELAP-3D model of the mock-up is introduced. All these activities contributed to the creation of a relevant theoretical and practical experience that can be used in further studies concerning incidental transients in real-plant scenarios in the framework of DEMO plant fusion safety activities.
... 62 The EU is currently developing this breeding blanket concept based on DEMO reactor specifications, and it will be tested in ITER under the form of TBMs. 63,64 Therefore, the requirements of the beryllium materials should be addressed at the development and qualification stages through the definition of the material specifications, optimization of the fabrication process with adequate quality control, characterization of the material properties, and performance testing, including thermo-mechanical testing of the pebble bed, long-term neutron irradiation, and tritium release and retention behavior. In particular, the R&D activities are focused on the effect of neutron irradiation on the thermo-mechanical properties of the beryllium pebbles and the pebble bed as a function of irradiation temperature and neutron dose, taking into account material degradation under irradiation, tritium release and retention characteristics as a function of irradiation temperature, neutron dose, purge gas chemistry, material properties and microstructure, compatibility with structural material under neutron irradiation, thermal conductivity in the pebble beds, and interaction of beryllium materials with air and steam. ...
... The estimated values of swelling of cylinder-shaped samples of titanium beryllide irradiated at 740K with a neutron fluence of 6.94 × 10 25 n/m 2 (E 4 1 MeV) with a damage dose of 13.9 dpa and at 873K with a neutron fluence of 8.07 × 10 25 n/m 2 with a damage dose of 16.3 dpa were 0.08% and 0.28%, respectively. 64 The helium productions were 2300 appm for 740K and 2680 appm for 873K. 49,72 At the two lowest irradiation temperatures of 425°C and 525°C, the swelling behavior of beryllium and the beryllides is approximately the same, with an average swelling of 2.5%. ...
Beryllium metal and other beryllium-containing compounds are known for their unusual combination of properties, which includes low atomic number, low density, high stiffness, and high thermal conductivity. These attributes account for the long-standing interest in these materials since the 1930s.
Important fields of application for beryllium metal and its compounds include acoustics, aerospace structures, x-ray transmission, motion control, fission test reactors, fusion energy research, laser-based optical systems, high-energy particle physics research, high-performance automotive applications, and thermal management. Beryllium-containing alloys (which typically contain less than 2% Be) are used extensively in commercial electronics, telecommunications infrastructure, automotive electronics, oil and gas equipment, tooling for plastic molding, and medical equipment applications.
Thanks to its low atomic number, beryllium has very high x-ray transmissivity, coupled with a low absorption and high scattering cross-section-to-neutrons-and-so-is-one-of-the-most-important-and-vital elements in the nuclear field. Especially, the superior neutron multiplication performance of beryllium via the (n,2n) reaction is essential in nuclear fusion technology. This chapter of Comprehensive Nuclear Materials is organized as follows:
• Section 2: Background and status of recent research, including the need for R&D into beryllium intermetallic compounds (beryllides).
• Section 3: Fabrication technology for both beryllium and beryllides in the form of pebbles, which will be needed due performance requirements for neutron multipliers.
• Section 4: Physical, chemical, thermal, and mechanical properties, as well as retention properties in unirradiated material.
• Section 5: Neutron irradiation effects: swelling, impact on mechanical properties, tritium release, and microstructure evolution.
• Section 6: First principles modeling, including simulation methods, bulk properties, self-interstitial defects, transmutation gaseous atoms, effects of other impurities, absorption and desorption of hydrogen, second-phase precipitates, and beryllides.
• Section 7: handling and safety issues, including health effects of beryllium materials, an overview of standards and regulations, and risks specific to fusion applications.
• Section 8: summary and conclusion.
... Fusion reactors offer a promising option for future electricity generation technologies due to their short-life nuclear waste, zero CO 2 emission, and long-term natural resources (sea water for producing deuterium and lithium for producing tritium). However, this emergent technology poses a large number of challenges, including the materials involved, tritium permeation, safety analysis and remote handling [1]. Although these challenges are based on physical issues, there are also certain important engineering aspects, such as the conversion of thermal energy into electricity [2]. ...
... Although these challenges are based on physical issues, there are also certain important engineering aspects, such as the conversion of thermal energy into electricity [2]. The DEMOnstration Power Plant (DEMO) is a project of the European Fusion Development Agreement (EFDA) which aims to build a prototype power plant at an operating scale (500 MWe) once the scientific challenges of the reactor have been tested in the International Tokamak Experimental Reactor (ITER) [1]. ...
The EUROfusion research program is currently exploring alternative solutions for a future fusion power
plant with DEMO (DEMOnstration Power Plant) prototype. One of the most important issues arising from
a dual coolant lithium lead blanket-based reactor is the correct integration of the four thermal sources in
order to achieve the highest electricity production. This study analyses the technical feasibility of supercritical
CO2 Brayton power cycles. Starting with a classical re-compressed cycle, which is taken as the
baseline case, two alternative proposals are investigated. On the one hand, a modified re-compressed
layout with only one recuperator is studied, and is found to achieve the same electric efficiency as
that of the baseline case (34.6%). On the other hand, an optimised recuperated layout is proposed, which
achieves a 33.6% electric efficiency. A parametric study is conducted in order to optimise the heat
exchanger size. When the re-compressed layout is optimised, a loss of efficiency (5%) is experienced. In
the case of the recuperated layout optimisation the efficiency loss is reduced to 3%, achieving a reduction
in heat exchanger size of 2/3.
... According to the different choices of the breeder, the TBM can be divided into liquid/solid breeder blankets. The European Union and the United States have developed the Helium Cooled Lithium Lead (HCLL) TBM and the Dual Coolant Lithium Lead (DCLL) TBM [2,3], respectively. Both of them use the liquid metal Pb-16Li as tritium breeder and coolant, while the HCLL uses helium for cooling FW and structures, * Corresponding author. ...
... On the other hand, the European Union, Japan and China have developed their solid TBMs, using either Li 4 SiO 4 or Li 2 TiO 3 pebble beds as tritium breeder. Besides, both the European Union and China choose the helium as their coolants [2,4], while Japan uses water instead [5]. Furthermore, India has proposed the Lithium Lead Ceramic Breeder (LLCB) TBM, which has both features of solid breeder and liquid breeder blankets [6]. ...
In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li4SiO4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER.
The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition.
... At the beginning of the PCD phase, which was in 2014, the HCPB blanket adopted the so-called "beer-box" concept [11], which is similar to the HCPB Test Blanket Module developed at KIT in the 2010s [17]. This concept has a very robust vertical and horizontal grid structure, resulting in a large amount of steel in the breeder zone, which compromises its tritium breeding performance. ...
Significant design efforts were undertaken during the Pre-Concept Design (PCD) phase of the European DEMO program to optimize the helium cooled pebble bed (HCPB) breeding blanket. A gate review was conducted for the entire European DEMO program at the conclusion of the PCD phase. This article presents a summary of the design evolution and the rationale behind the HCPB breeding blanket concept for the European DEMO. The main performance metrics, including nuclear, thermal hydraulics, thermal mechanical, and tritium permeation behaviors, are reported. These figures demonstrate that the HCPB breeding blanket is a highly effective tritium-breeding and robust driver blanket concept for the European DEMO. In addition, three alternative concepts of interest were explored. Furthermore, this article outlines the upcoming design and R&D activities for the HCPB breeding blanket during the Concept Design phase (2021–2027).
... During the PCD phase, two official baselines (BLs) of the European DEMO machine were released, one called the DEMO BL2015 and the latest one called the DEMO BL2017. The design evolution of the HCPB BB is schematically shown in Figure 1. Figure 1 Design evolution of HCPB BB during PCD phase [11] [12] [16] At the beginning of the PCD phase, which was in 2014, the HCPB blanket adopted the so-called "beer-box" concept [11] similar to the HCPB Test Blanket Module developed at KIT in the 2010s [17]. This concept has a very robust vertical and horizontal grid structure resulting a large amount of steel in the breeder zone, which compromises its tritium breeding performance. ...
Significant design efforts were undertaken during the Pre-Concept Design (PCD) phase of the European DEMO programme to optimize the Helium Cooled Pebble Bed (HCPB) breeding blanket. A gate review was conducted for the entire European DEMO programme at the conclusion of the PCD phase. This article presents a summary of the design evolution and the rationale behind the HCPB breeding blanket concept for the European DEMO. The main performance metrics, including nuclear, thermal hydraulics, thermal mechanical, and tritium permeation behaviors, are reported. These figures demonstrate that the HCPB breeding blanket is a highly effective tritium-breeding and robust driver blanket concept for the European DEMO. In addition, three alternative concepts of interest were explored. Furthermore, this article outlines the upcoming design and R&D activities for the HCPB breeding blanket during the Concept Design phase (2021-2027).
... During the PCD phase, two official baselines (BLs) of the European DEMO machine were released, one called the DEMO BL2015 and the latest one called the DEMO BL2017. At the beginning of the PCD phase, which was in 2014, the HCPB blanket adopted the so-called "beer-box" concept similar to the HCPB Test Blanket Module developed at KIT in the 2010s [14]. This concept has a very robust vertical and horizontal grid structure resulting a large amount of steel in the breeder zone, which compromises its tritium breeding performance. ...
Significant design efforts were undertaken during the Pre-Concept Design (PCD) phase of the European DEMO programme to optimize the Helium Cooled Pebble Bed (HCPB) breeding blanket. A Gate Review was conducted for the entire European DEMO programme at the conclusion of the PCD phase. This article presents a summary of the design evolution and the rationale behind the HCPB breeding blanket concept for the European DEMO. The main performance metrics, including nuclear, thermal hydraulics, thermal mechanical, and tritium permeation behaviors, are reported. These figures demonstrate that the HCPB breeding blanket is a highly effective tritium-breeding and robust driver blanket concept for the European DEMO. In addition, three alternative concepts of interest were explored. Furthermore, this article outlines the upcoming design and R&D activities for the HCPB breeding blanket during the Concept Design phase (2021-2027).
... As a result of its benefits, helium is being considered as a primary coolant for fusion systems in the United States and internationally. 6 Improving the effectiveness of helium cooling systems is important to ensuring the survivability of components in the environment of high heat flux. ...
A helium flow loop is being assembled at Oak Ridge National Laboratory to analyze heat transfer enhancement for systems such as blanket and divertor components. To efficiently identify optimum geometries for heat transfer enhancement in these applications, simulation work is performed to optimize test section designs that are built and tested in the helium flow loop that operates at 4 MPa and a mass flow rate of 100 g/s. Different ribbed geometries that examine rib shape, rib height, rib orientation, rib spacing, and three-dimensional orientation are modeled and simulated in STAR-CCM+ to compare their ability to remove heat and mitigate pressure drop. Following the simulations, models are selected and manufactured for the helium flow loop tests. Simulations initially focus on a hydrodynamic study to determine the appropriate mesh and physics models and then add a heat flux to analyze the heat transfer abilities of the models. The simulations are run in steady state and use a Reynolds-averaged Navier-Stokes k-ε turbulence model. The helium is modeled as an ideal gas. The simulation explores models of geometries that enhance the heat transfer and decrease pressure drop with an overall goal of increasing fluid collision with the wall. Enhanced geometries are simulated to select appropriate designs for manufacturing, and preliminary experimental results are used to validate the simulations. The factors that are being analyzed in the comparison between the experimental and the simulated results include matching thermocouple temperatures, pressure drop, roughness, and fluid velocity.
... Europe is currently developing several breeder blanket concepts for future thermonuclear DEMOnstration fusion reactors [1]. Two concepts have been chosen to be tested in ITER under the form of Test Blanket Modules (TBMs) [2]: the Helium-Cooled Lithium-Lead (HCLL, Fig. 1a) concept which uses the eutectic Pb-15.7Li (enriched in 6Li) as both tritium breeder and neutron multiplier and the Helium-Cooled Pebble-Bed (HCPB, Fig. 1b) concept with lithiated ceramic (enriched in 6Li) pebbles as tritium breeder and beryllium pebbles as neutron multiplier. ...
Two concepts have been chosen to be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) and the Helium-Cooled Pebble-Bed (HCPB). Both European TBMs designs share similar steel box structure which is constituted by a box, made of two Side Caps (SCs) and a First Wall (FW), stiffened by horizontal and vertical Stiffening Plates (SP) and closed on its back by several back plates (BPs). All structure subcomponents are internally cooled by Helium circulating in meandering squared section channels. This paper describes manufacturing technologies developed and implemented to assembly the SPs into the box. It presents the preliminary manufacturing procedure developed and applied for the assembly of the SPs into the box by Tungsten Inert Gas (TIG). Several mock-ups have been manufactured from laboratory to feasibility mock-ups (scale 1:1) on which non-destructive and destructive tests have been carried-out to identify the preliminary manufacturing procedure. Due to TBM specificities (namely complex welding trajectories, heavy and big components, plates with channels, space constraints, …) a specific welding facility including a custom welding torch and an automated bench has been achieved and is also described in the paper. We detail the adopted manufacturing strategies, as the optimization of welding sequence to minimize distortions and the customization of welding parameters, to compensate machining tolerances and welding gaps. Results such as welded joints quality and microstructure, internal cooling channel deformation and structure distortions are reported. These developments have been performed following a standardized procedure complying with professional codes and standards (RCC-MRx).
... Europe is currently developing two reference breeding blanket concepts based on DEMO reactor specifications that will be tested in ITER under the form of TBMs [2,3]: i) the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li (enriched in 6 Li) as both tritium breeder and neutron multiplier, ii) the Helium-Cooled Pebble-Bed (HCPB) concept with lithiated ceramic (enriched in 6 Li) pebbles as tritium breeder and beryllium pebbles as neutron multiplier (Figs. 1 and 2). Both concepts are using a reduced activation ferriticmartensitic steel as structural material, the EUROFER97 (X10CrWVTa9-1) [4], and pressurized helium as a coolant for efficient heat extraction (300-500°C, 8 MPa). ...
The paper reviews the current status of development and qualification of the HCPB TBM's functional materials, namely, Li-ceramic breeder and Be multiplier materials. The main functional and performance requirements for both functional materials are overviewed. The main results and outcomes of the post-irradiation examinations (PIE) of Li-ceramic breeder and Be/beryllides materials irradiated in HIDOBE and HICU irradiation campaigns are presented aiming at determination of the materials’ performance, properties and characteristics under neutron irradiation, like e.g. morphology and microstructure, tritium retention/release characteristics, mechanical properties. Fabrication of the advanced ceramic breeder material, containing lithium orthosilicate with addition of lithium metatitanate phase in order to improve its mechanical properties, is described. Results of characterization of this advanced ceramic breeder in terms of mechanical properties, porosity, morphology and effective thermal conductivity of a pebble bed are overviewed. In addition, results of characterization of the reference 1 mm Be pebbles produced by Rotation Electrode Process are presented in terms of their microstructure, tritium release characteristics and interaction with air and steam. A necessity to realize in the future a new irradiation experiment for the functional materials, LIBERTI experiment, is discussed in the conclusions.
... There has been considerable interest in reduced-activation ferriticmartensite (RAFM) steels since these materials are potential candidates for structural applications in future nuclear fusion reactors, in particular the DEMO (DEMOnstration Power Plant) version of this novel technology [1][2][3]. They possess a unique combination of high mechanical strength, moderate-to-good ductility at room temperature, high irradiation resistance, and low neutron-induced radioactivity [2,4]. ...
Oxide dispersion strengthened ODS-Eurofer steel is a promising candidate for structural applications in future
nuclear fusion reactors. Samples of 9Cr-ODS-Eurofer steel were cold rolled up to 80% thickness reduction and
annealed up to 1350 °C for 1 h. The microstructural changes of the annealed samples were followed by magnetic measurements taken at room temperature. In comparison with the coercive field (Hc), the remanent magnetization (Mr) and longitudinal magnetostriction (λlong) behaviors can be explained only if we assume some interaction between pinning sites and residual stresses in the material. It was found that Hc, Mr, λlong, and hardness display the same trend. At the expected fusion DEMO reactor operating conditions (around 650 °C and maximum magnetic field of 6 T), the magnetostrictive deformation would not surpass 15 ppm (15 μm in 1 m).
... Europe is currently developing two reference breeding blankets concepts based on DEMO reactor specifications that will be tested in ITER under the form of TBMs [2,3]: i) the helium-cooled lithiumlead (HCLL) concept which uses the eutectic Pb-16Li (enriched in 6 Li) as both tritium breeder and neutron multiplier, ii) the helium-cooled pebble-bed (HCPB) concept with lithiated ceramic (enriched in 6 Li) pebbles as tritium breeder and beryllium pebbles as neutron multiplier. Both concepts are using a reduced activation http://dx.doi.org/10.1016/j.fusengdes.2017.04.051 0920-3796/© 2017 Elsevier B.V. All rights reserved. ...
The paper overviews activities focused on qualification of EUROFER97 structural material, introduced under a probationary phase in the nuclear components design and construction code RCC-MRx, and identification/analyses of gaps in the respective material database to be filled in. Additionally the available design rules in the code are reviewed to verify their applicability to the specificities of EUROFER97 steel and to the TBM design and fabrication. Progress achieved in development of fabrication technologies and procedures applied for manufacturing of the TBM sub-components, like, HCLL and HCPB cooling plates, stiffening plates, first wall and side caps, and for TBM structure sub-assembly is described. The used technologies are based on fusion (laser and TIG) and diffusion (HIP) welding techniques taking into account specificities of the EUROFER97 steel. With help of the agreed notified body, an appropriate approach/methodology for qualification of the developed, TBMs-related preliminary welding procedure specifications has been identified and future steps established.
... 20 世纪 80 年，德国卡尔斯鲁厄核子研究中心 KfK（1995 年改名为卡尔斯鲁 厄研究中心 FZK，2009 年与卡尔斯鲁厄大学正式合并改名为卡尔斯鲁厄理工学 院 KIT）开展了氦冷固态增殖剂包层的概念设计研究( Dalle Donne et al., 1983;Dalle Donne et al., 1989 自 2005 年第一版 EU HCPB-TBM 设计描述报告提交给 ITER 测试包层工作 专家组 TBWG 以来， 其设计方案经历过多次优化修改。 根据欧盟 Fusion for Energy 机构（Fusion for Energy 负责管理欧盟参与 ITER 的所有项目）要求，KIT 作为牵 头单位于 2011 年完成了 EU HCPB-TBM 包层模块的详细概念设计( Boccaccini et al., 2011 Qu et al., 2015)、 基于 EUROFER 第一壁加工锻造( Neuberger et al., 2015)、第一壁增强换热研究( Chen and Arbeiter, 2015)、大型高温高压氦气 回路 HELOKA 的高效运行( Ghidersa et al., 2008)、吹扫气中的提氚技术( Santucci et al., 2016)、氚增殖剂和中子倍增剂的加工工艺和性能研究( Knitter et al., 2013;Vladimirov et al., 2014 ...
According to the Roadmap of Chinese Magnetic Confinement Fusion Energy, the first phase of Chinese Fusion Engineering Test Reactor (CFETR) will obtain about 200 MW fusion power, achieve steady state or long pulse plasma operation and attain tritium self-sustainability. Breeding blanket of fusion reactor is the key component of attaining tritium self-sustainability. Besides tritium breeding, breeding blanket also has the following two main functions: energy exhaust and conversion and radiation-shielding. Therefore, the blanket for fusion reactor has to meet the requirements of neutronics, thermal hydraulics, structural mechanics and safety, which makes the design of blanket rather challenging. To support the development of CFETR, the thermo-mechanical assessment of a helium cooled solid breeder (HCSB) blanket for CFETR has been conducted. The results provide reference for future design of CFETR HCSB blanket.
In this PhD dissertation, steady state thermal analysis and transient thermal analysis of the blanket module following CFETR plasma pulse are conducted. Furthermore, sensitivity analysis of several factors impacting the blanket temperature field (i.e., thermal conductivity of pebble bed, the thermal hydraulic parameters and thermal contact conductance of pebble bed-wall) is performed. Steady state results show that the temperatures of the blanket sub-components are within limits. The temperature fields are also used as inputs to structural analysis under monotonic type loading (M-type). The temperature evolution of blanket sub-components is obtained through transient thermal analysis. The temperature on blanket increases with respect to the thermal loading during ramp-up. The temperature of first wall increases quicker than that of breeding zone during ramp-up. The temperature of first wall decrease faster than that of breeding zone during ramp-down. The large inversion of temperature gradient between first wall and breeding zone during plasma ramp-up and ramp-down will be the cause of large cyclic thermal stress. The results of transient thermal analysis are used as inputs to structural analysis under cyclic type loading (C-type). It is obtained that pebble bed packing factor and pebble size have very little impact on blanket temperature. It is found that the variation of thermal conductivity of pebble bed has large influence on the temperature of pebble bed. Reducing the thermal conductivity will largely increase the temperature of pebble bed. The temperature will not change after the thermal conductivity is increased to a certain value. As the inlet mass flow rate increases (decreases), the temperature on blanket decreases (increases). The temperatures will not change after the inlet mass flow rate is increased to a certain value. The maximum temperatures on blanket vary linearly with the inlet temperature, but not significantly. The roughness of cooling channels at breeding zone has little impact on temperature of breeding zone and adding roughness to cooling channels at breeding zone is not worthwhile. The change of thermal contact conductance between pebble bed and wall has a limited impact on the temperature of the blanket. The current blanket design can easily tackle the influence of TCC variation. Even the pebble bed totally loses the contact with wall, the temperatures on the blanket still stay within design limits. It can be concluded that the current blanket design has a good thermal stability.
Based on thermal analysis, thermo-mechanical analysis has been performed. The structural strength assessment criteria and methods for fusion blanket have been explained following the Structural Design Criteria for ITER In-vessel Components (SDC-IC). The analysis and assessment of M-type damage modes of the blanket (without taking into account the electromagnetic loading) has been performed with respect to the SDC-IC rules. The current blanket has a robust behaviour in preventing M-type damage modes. Thereafter, the analysis and assessment of C-type damage modes of the blanket has been conducted under SDC-IC rules. The blanket shows a robust performance against C-type damage modes. Preliminary FFMEA analysis of the blanket system has conducted, obtaining a list of blanket reference accidents. Following SDC-IC level D criteria, the thermo-mechanical analysis of the blanket under typical accidental condition (In-box LOCA) is performed. Results show that the blanket has good mechanical performance even under In-box LOCA condition.
The research carried out in this PhD dissertation will provide good reference for design of future CFETR blanket.
... India is taking advantage of this with its Lead-Lithium cooled Ceramic Breeder (LLCB) design. Europe is developing a He cooled lithium lead (HCLL) blanket which utilizes helium as the coolant and lithium-lead as the breeder, with reduced activation ferriticmartensitic steel (RAFM) as the structural material [11]. The US is also utilizing LiPb in their blanket design with the dual-coolant lead-lithium concept (DCLL); helium is used to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket [3]. ...
... This included two main milestones: ITER (International Tokamak Experimental Reactor) and DEMO (DEMOnstration Power Plant). The main objective of ITER is to demonstrate the technological feasibility of fusion energy by producing net thermal energy and testing the required materials [2]. DEMO will be a bridge between ITER and commercial fusion power plants, demonstrating the feasibility of the integration of all the required systems (reactor and balance of plant) to operate a fusion power plant, including issues of security, wastes management, maintenance, and so on. ...
This paper presents an exploratory analysis of the suitability of supercritical CO2 Brayton power cycles as alternative energy conversion systems for a future fusion reactor based on a DCLL (dual coolant lithium-lead) blanket, as prescribed by EUROfusion. The main issue dealt is the optimization of the integration of the different thermal sources with the power cycle in order to achieve the highest electricity production. The analysis includes the assessment of the pumping consumption in the heating and cooling loops, taking into account additional considerations as control issues and integration of thermal energy storage systems. An exergy analysis has been performed in order to understand the behavior of each layout.
... Following the recommendations of the Roadmap, investigations on 4 blanket concepts have been included in the WPBB programme. They are the two blanket concepts selected to be tested in the TBM Programme in ITER [4], namely the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. In addition, to extend the selection basis according to a strategy of risk minimization, a water-cooled concept, i.e. the Water Cooled Lithium Lead (WCLL) and a more advanced concept, i.e. the Dual Coolant Lithium Lead (DCLL) have been included in the study. ...
The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.
... High pressure helium gas is considered as the coolant of the First Wall (FW) in several blanket concepts (Helium Cooled Pebble Bed (HCPB), Helium Cooled Lithium Lead (HCLL), Dual Coolant Lithium Lead (DCLL)) for DEMO and the ITER Test Blanket Modules (TBMs) [1,2]. The main advantages of helium as coolant are safety aspects (chemically inert, no activation, comparatively low effort to remove tritium), no chemical corrosion, and a flexible temperature range (due to no phase change) allowing to match the permissible operation temperature window of the foreseen structural material. ...
... High pressure helium gas is considered as the coolant of the First Wall (FW) in several blanket concepts (Helium Cooled Pebble Bed (HCPB), Helium Cooled Lithium Lead (HCLL), Dual Coolant Lithium Lead (DCLL)) for DEMO and the ITER Test Blanket Modules (TBMs) [1,2]. The main advantages of helium as coolant are safety aspects (chemically inert, no activation, comparatively low effort to remove tritium), no chemical corrosion, and a flexible temperature range (due to no phase change) allowing to match the permissible operation temperature window of the foreseen structural material. ...
The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.
... In the frame of the European fusion programme, advanced designs for HCPB (Helium-Cooled Pebble Bed) and the HCLL (Helium-Cooled Lithium-Lead) TBMs have been developed for tests in ITER [9]. Engineering CAD models were created for the test blanket port plug, including all sub-systems such as the water-cooled steel frame, HCPB and HCLL TBM assemblies, shield modules, feeding pipes, etc. ...
... Europe is currently developing two reference breeding blankets concepts based on DEMO reactor specifications that will be tested in ITER under the form of TBMs [2,3]: (i) the helium-cooled lithiumlead (HCLL) concept which uses the eutectic Pb-16Li (enriched in 6 Li) as both tritium breeder and neutron multiplier, (ii) the helium-cooled pebble-bed (HCPB) concept with lithiated ceramic (enriched in 6 Li) pebbles as tritium breeder and beryllium pebbles as neutron multiplier. Both concepts are using a reduced activation ferritic-martensitic steel as structural material, the EUROFER97 (X10CrWVTa9-1) [4,5], and pressurized helium as a coolant for efficient heat extraction (300-500 • C, 8 MPa). ...
... Three of the eighteen ITER equatorial ports are reserved for these modules. The material chosen for DEMO is ferritic steel [1]. Therefore, the ITER TBMs will be made of ferromagnetic material that will get magnetized by the tokamak magnetic fields. ...
The magnetic perturbation due to the ferromagnetic test blanket modules
(TBMs) may deteriorate fast ion confinement in ITER. This effect must be
quantified by numerical studies in 3D. We have implemented a combined finite
element method (FEM) -- Biot-Savart law integrator method (BSLIM) to calculate
the ITER 3D magnetic field and vector potential in detail. Unavoidable geometry
simplifications changed the mass of the TBMs and ferritic inserts (FIs) up to
26%. This has been compensated for by modifying the nonlinear ferromagnetic
material properties accordingly. Despite the simplifications, the computation
geometry and the calculated fields are highly detailed. The combination of
careful FEM mesh design and using BSLIM enables the use of the fields
unsmoothed for particle orbit-following simulations. The magnetic field was
found to agree with earlier calculations and revealed finer details. The vector
potential is intended to serve as input for plasma shielding calculations.
... Also, it is responsible for the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER [1]. Currently two EU TBS designs are in the phase of conceptual design -helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB) [2,3]. The three main milestones of the EU TBM project in 2014 are as follows: ...
... TBS safety demonstration files of both EU TBSs -HCLL and HCPB have been updated in line with the TBSs 2011 baselines presented in [4]. The work covered the TBS design description on the level of sub-systems and main components; general operation states; expected maintenance activities and drafting of detailed TBS plant breakdown structures (PBS). ...
The European Joint Undertaking for ITER and the Development of Fusion Energy ('Fusion for Energy'- F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European Test Blanket Systems (TBS) in ITER. An overview of the ITER TBS program has been presented recently at ISFNT-10. Currently two EU TBS designs are in the phase of conceptual design - Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the Conceptual Design Review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years in the fields of update of the TBS safety demonstration file; TBS Safety approach, design principles, requirements, features and safety functions; detailed TBS components classifications; Radiation Shielding and Protection; and Selection of reference accidents scenarios and Accidents analyses. Finally the authors share the planned future EU TBS safety activities.
The depiction of the nuclear responses of the ITER European Test Blanket Modules (TBMs), Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebbles Bed (HCPB) is presented in this work. Following important components update, and important methodological advances, the nuclear heat and the tritium production have been revisited, giving new estimations 10% higher than the previous evaluation for nuclear heat in both TBMs and to 15% higher for HCPB T production. This has an impact on the thermo-mechanical design of the TBM and the tritium handling. In addition, the Shutdown Dose Rates in the respective port interspace have been characterized in local approach. It shows a performance that could imply compatibility with planned in-situ maintenance activities when analysed in global approach, an improvement with respect to previous evaluations.
Subcomponent manufacturing and assembly concepts for the fabrication of the helium-cooled pebble bed test blanket module (TBM) for ITER have been developed over more than one decade at KIT, in particular the first wall (FW), which is a key element for the TBM fabrication. The design of this subcomponent foresees the manufacturing of a large U-bended plate of EUROFER with built-in channels for helium cooling. Manufacturing technologies developed at KIT are based on diffusion welding of two half-plates as the most promising option. This paper deals with the manufacturing of two medium-scale TBM FW mock ups according to two different industrial processes: a uni-axial diffusion welding process realized in a mechanic press at high temperature and a hot isostatic pressing process applied to a canned assembly at relatively low pressure.
The qualification of the welds produced is described, and the results are compared to previous small- and medium-size scale experiments. The results of the recent FW fabrication mock ups are presented with regard to material data (e.g., ultimate strength, ductile-brittle transition temperature) and TBM-relevant parameters (e.g., deformation of cooling channels). The paper concludes with an overview of the strategy to evolve from 1/8th-scale experiments to TBM-relevant dimensions.
This article deals with description and current status of a project of a non-nuclear, full size (1:1 scale) test platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER port cell #16. The facility called TBM platform reproduces the ITER port cell #16 and port interspace with all the relevant interfaces and mock-ups of the corresponding main components. Thanks to the modular design of the platform, it is possible to adapt or change completely the interfaces in the future if needed or required according to the updated configuration of TBSs. In the same way, based on customer requirements, it will be possible to adapt the interfaces and piping inside the mock-ups in order to represent also the other, non-EU configurations of TBM systems designed for port cells #02 and #18. Construction of this test platform is realized and funded within the scope of the SUSEN project.
Currently the HCPB blanket concept is one of the four breeding blankets concepts under development for the European DEMO. This work reports on the results of an investigation of the thermal and structural performances of a new design of this blanket, proposed in 2015 by the KIT HCPB Team aimed at establishing a baseline design of the HCPB breeding blanket following the updated EU DEMO plant specifications. The thermal analyses have been reported in another paper, while the structural analyses are presented and discussed in this paper. A 3D slice model of the DEMO HCPB blanket has been set up to run thermo-mechanical analyses of the blanket under steady state and DEMO transient pulsed conditions. The analyses for the blanket have been assessed with respect to the structural design criteria and standards (mainly RCC-MR, completed by SDC-IC). The results identify some problematic regions in the design, concentrated in the connection regions of the cooling plates with the blanket back supporting structure. For monotonic damage modes, the blanket structure shows a global satisfying behavior in immediate plastic collapse and plastic instability damage modes, and thermal creep damage mode. While it fails to fulfill the criteria to prevent immediate plastic flow localization damage mode in some regions. Counter-actions to improve the design have been proposed and will be implemented in future design revisions. Considering the cyclic loadings, the FW shows a satisfying behavior against ratcheting and fatigue damage modes during plasma ramp-up and ramp-down phases.
Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.
The Neutron Activation System (NAS) for the EU Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Systems (TBSs) is an instrument that is proposed to determine the absolute neutron fluence and absolute neutron flux with information on the neutron spectrum in selected positions of the corresponding Test Blanket Modules (TBMs). In the NAS activation probes are exposed to the ITER neutron flux for periods ranging from several tens of seconds up to a full plasma pulse length, and the induced gamma activities are subsequently measured. The NAS is composed of a pneumatic transfer system and a counting station. The pneumatic transfer system includes irradiation ends in TBMs, transfer pipes, return gas pipes, a transfer station with a distributor (carousel), and a pressurized gas driving system, while the counting station consists of gamma ray detectors, signal processing electronic devices, and data analyzing software for neutron source strength evaluation. In this paper, a brief description on the proposed TBM NAS as well as the key components is presented, and the integration challenges of TBM NAS are outlined.
This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps.
The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions.
The measured data have shown that critical for flow distribution in manifold 2 are grid/cap channels attached to the manifold regions adjacent to the first wall. In these regions high velocity jets exiting from the first wall induce low pressure at the channel inlets and, therefore, lower flow supply. Making resistance to high velocity streams from the first wall the perforated penetrations of horizontal grid plates which divide the manifold 2 into chambers are crucial for pressure homogenizing and reasonable flow distribution among channels attached to the central part of manifold 2. In the major part of manifold 3 very small differential pressures were measured except for some breeding unit inlets where strong pressure peaks were recorded. Nevertheless, no dramatic differences were found between experimentally determined flow rates in individual breeding unit channels.
Based on the CFD software platform FLUENT, three-dimensional numerical simulation was carried out for thermal hydraulics characteristics of the breeder zone pebble bed for China helium cooled ceramic breeder-test blanket module (HCCB-TBM). According the actual operating conditions of ITER, such as the nuclear heat distributions in the breeder zone and structure wall, the inlet velocities and temperatures of purge gas helium and coolant helium, the flow field and heat transfer characteristics of the purge gas helium carrying the tritium flowing over the lithium silicate pebble bed were obtained, and the temperature distribution and pressure drop were presented as well. The calculation results show that the permutation mode of the lithium silicate spheres in the pebble bed have an influence on the flow field and the maximum temperature of the pebble-bed. The maximum temperature of the pebble-bed and the structure wall does not exceed the designed temperature under the ITER operating conditions. The results will be benefit for the design verification of the breeder zone thermal hydraulics scheme and improving the following experiments of the purge gas.
In the last years EU started a new effort for the definition of a Roadmap for the fusion beyond ITER up to the construction of the DEMO reactor. The blanket technology, key issue for the development of the fusion breeder reactor, is an important part of this review. In the last 10 years the R&D of EU in blanket has been focused on two kinds of blanket concepts, namely the Helium Cooled Pebble Bed (HCPB) and Lithium lead (HCLL); both are EU reference concepts for DEMO and for the test in ITER. In addition to these two reference concepts, few other concepts have been investigated in EU. One of these is the Dual Coolant Lithium Lead (DCLL) that was developed in the 90-ties in Karlsruhe and UCSD. The spectrum of blankets studied in EU includes also a water cooled blanket with liquid breeder, the WCLL.
In this study, an estimation method of graphite dust production in the pebble-bed type reflector region of the Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) of the International Thermonuclear Experimental Reactor (ITER) project using Finite Element Method (FEM) was proposed and the total amount of dust production was calculated. A unit-cell model of uniformly arranged pebbles was defined with thermal and mechanical loadings. A commercial FEM program, Abaqus V6.10, was used to model and solve the stress field under multiple contact constraints between pebbles in the unit-cell. Resultant normal contact forces and slip distances on the contact points were applied into the Archard adhesive wear model to calculate the amount of graphite dust. The Finite Element (FE) analysis was repeated at 27 unit-cell locations chosen to form an interpolated dust density function for the entire region of the reflector. The dust production calculation was extended to the life time of the HCCR and the total graphite dust production was estimated to 0.279 g at the end of the life time with the maximum graphite dust density of 0.149 μg/mm3. The dust explosion could be a safety issue with the calculated dust density level and it requires that an appropriate maintenance to remove sufficient amount of graphite dust regularly to prevent the possibility of dust explosion.
Commercial infrared heaters have been proposed to be used in the HELOKA facility under construction at Karlsruhe Institute of Technology (KIT) to test a mock-up of the first wall (FW), called thermo-cycle mock-up (TCM) plate, under stress loading comparable to those experienced by the test blanket modules (TBMs) in ITER. Two related issues are analyzed in this paper, in relation to the ongoing European project aimed at the design of the two EU TBMs: (1) the possibility to reproduce, by means of those heaters, high heat flux loading conditions on the TCM plate similar to those expected on the ITER TBMs, and (2) the thermo-mechanical analysis of the TCM itself, in order to define a suitable choice of experimental parameters and mechanical constraints leading to a relevant stress condition. A suitable heater model is developed and validated against experimental data from an ad-hoc test campaign. A thermo-mechanical study of the TCM plate is presented, showing that the structure is able to withstand high thermal loads, even in the most constrained case, reaching stress levels comparable to the ITER TBM.
The key components of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM) in ITER are the Breeder Units (BU). These are the responsible for the tritium breeding and part of the heat extraction in the HCPB TBM. After a detailed design and engineering phase performed during the last years in the Karlsruhe Institute of Technology (KIT), a reference model for the manufacturing of a HCPB BU mock-up has been obtained. The mid-term is the out-of-pile qualification of the thermal and thermo-mechanical performance of a full-scale HCPB BU mock-up in a dedicated helium loop. Several key manufacturing technologies have been developed for the fabrication of the HCPB BU. In order to pre-qualify these techniques, a Short Breeder Unit mock-up (SHOBU) is under construction and to be tested. This paper aims at describing the relevance of SHOBU with a full-scale HCPB BU, the constitutive parts of SHOBU, the manufacturing and joining technologies involved, the assembly sequence (taking into consideration functional steps like its filling with Li4SiO4 pebbles or its assembly in the HCPB TBM) and the welding procedures studied. The paper concludes with a description of the required pre-qualification tests performed to SHOBU, i.e. pressure and leak tightness tests, according to the standards.
Complementing the efforts towards the realization of ITER, KIT is pursuing, within the overall EURATOM fusion program, a number of important long-term technology developments towards a magnetic confinement fusion power plant (FPP), taking into account the features that will distinguish such facility from ITER.
To this end, structural materials on the basis of both low-activation steels and refractory metals, as well as concepts for breeding blankets and divertor designs, are being developed along with suitable manufacturing and joining technologies. In parallel, KIT contributes to the engineering design and validation phase of the International Fusion Materials Irradiation Facility (IFMIF) necessary for qualifying the materials to be used in an FPP. The specific characteristics of an FPP fuel cycle, i.e., substantial tritium quantities within huge mass flows of gases and the related tritium compatible high throughput vacuum and pumping technologies, are being translated into viable engineering approaches. High temperature superconducting magnet solutions are being developed, with a view to overall plant efficiency. In order to increase the wall-plug efficiency of plasma heating, advanced gyrotron tubes with power levels significantly beyond what is envisaged for ITER are being developed along with a frequency tunability option for efficiently counteracting plasma instabilities.
This paper describes a project of a non-nuclear, 1:1 scale testing platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER Port Cell #16. This TBM platform is currently under construction in Centrum vyzkumu Rez, Czech Republic.The facility is realized within the scope of the SUSEN project and its full operation is foreseen in the first half of 2016.
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.
Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.
The natural resources required for the operation of fusion power plants are - with the possible exception of the neutron multiplier beryllium - readily available. On the other hand, the supply of helium, which is required as cryogenic medium and coolant, may be a problem due to losses during operation and decommissioning. Helium is a rare element obtained as a by-product in the extraction of natural gas. The danger exists that the natural gas will be used up without the helium being conserved. We estimate the helium demand for a global 30% base-load contribution of fusion to electricity supply and also calculate the amount produced by the power plants themselves.
In the frame of the activities of the EU Breeder Blanket Programme, the Karlsruhe Institute of Technology (KIT) is involved in several activities in support of the design and qualification of the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM). During operation the TBM will be subjected to complex thermo-mechanical loads, especially the plasma facing parts, the so-called First Wall (FW). In the recent years significant effort was dedicated to the FE modeling of the TBM box as a complete assembly: detailed FE models (I quarter) were developed and thermo-mechanical calculations were performed for the better description of the thermo-mechanical behavior of the HCPB-TBM box under nominal and accidental conditions. In parallel with the numerical studies the manufacturing process for the TBM-FW has progressed. Currently KIT has developed a manufacturing path that will allow the manufacturing of an FW plate, the so-called TCM (Thermo-mechanical Cycle Mock up), with an overall size of 710 mm x 484 mm x 30 mm. The present paper presents the results of the numerical investigations of a mock up designed to qualify the manufacturing procedure of the TCM. For this the TCM will be subject to the same heat loads as the central part of a TBM-FW operating under TBM characteristic conditions (nominal surface heat flux of 500 kW/m(2) in pulsed regime). Based on the comparison with the numerical simulations of a full TBM box, a design of the test mock up including the support structure behind the TCM plate will be proposed with the aim of reproducing an equivalent level of stresses in the test object as in a TBM-FW.
Subcomponent manufacturing and assembly concepts for the fabrication of the helium-cooled pebble bed test blanket module (TBM) for ITER have been developed over more than one decade at KIT, in particular the first wall (FW), which is a key element for the TBM fabrication. The design of this subcomponent foresees the manufacturing of a large U-bended plate of EUROFER with built-in channels for helium cooling. Manufacturing technologies developed at KIT are based on diffusion welding of two half-plates as the most promising option. This paper deals with the manufacturing of two medium-scale TBM FW mock ups according to two different industrial processes: a uni-axial diffusion welding process realized in a mechanic press at high temperature and a hot isostatic pressing process applied to a canned assembly at relatively low pressure. The qualification of the welds produced is described, and the results are compared to previous smalland medium-size scale experiments. The results of the recent FW fabrication mock ups are presented with regard to material data (e.g., ultimate strength, ductilebrittle transition temperature) and TBM-relevant parameters (e.g., deformation of cooling channels). The paper concludes with an overview of the strategy to evolve from l/8th-scale experiments to TBM-relevant dimensions.
Fusion energy is one of the most promising solutions to the world energy supply. This paper presents an exploratory analysis of the suitability of supercritical CO2 Brayton power cycles (S-CO2) for low-temperature divertor fusion reactors cooled by helium (as defined by EFDA). Integration of three thermal sources (i.e., blanket, divertor and vacuum vessel) has been studied through proposing and analyzing a number of alternative layouts, achieving an improvement on power production higher than 5% over the baseline case, which entails to a gross efficiency (before self-consumptions) higher than 42%. In spite of this achievement, the assessment of power consumption for the circulating heat transfer fluids results in a penalty of 20% in the electricity production. Once the most suitable layout has been selected an optimization process has been conducted to adjust the key parameters to balance performance and size, achieving an electrical efficiency (electricity without taking into account auxiliary consumptions due to operation of the fusion reactor) higher than 33% and a reduction in overall size of heat exchangers of 1/3. Some relevant conclusions can be drawn from the present work: the potential of S-CO2 cycles as suitable converters of thermal energy to power in fusion reactors; the significance of a suitable integration of thermal sources to maximize power output; the high penalty of pumping power; and the convenience of identifying the key components of the layout as a way to optimize the whole cycle performance.
Licensing a pressurized nuclear equipment like the European Test Blanket Modules (TBM) Systems and,
on the longer term, breeder blankets of a fusion demonstration reactor (DEMO), will require presenting to the Regulator and the Agreed Notified Body, along with design and safety analyses, supporting data like consolidated materials data and design limits, qualified fabrication procedures specifications and validated modeling tools that go often over today’s state-of-the-art of nuclear industry. TBM systems feature indeed a newly developed structural material and advanced fabrication processes that were not referenced in any nuclear construction codes before, new type of functional materials, complex structures geometry and many interconnected sub-systems exchanging tritium by permeation or fluid mass transfer. For many years now, Europe has structured its development activities on TBM Systems toward the preparation of licensing. First tangible results are now arising: the EUROFER structural material hasbeen introduced in the RCC-MRx nuclear code, supported by a database of several thousands of test records; TBM box fabrication procedure specifications are under standardization by industry in view oftheir qualification; a modeling tool for accurate simulation of tritium transport in TBM systems has been developed in view of refining conservative inventory data published in preliminary safety reports andoptimizing waste management. Remaining challenges are identified and discussed
A preliminary design integration study of the Electron Cyclotron Heating and Current Drive, and Neutral Beam heating systems with the outboard Multi-Module blanket Segments in DEMO, is being presented. Scope of the work is to provide a first assessment on requirement analysis, concept generation and evaluation of the remote maintenance for a DEMO fusion power plant, in order to maximize its availability. The work has been divided into three phases: firstly, assessments to define the required openings for Electron Cyclotron and Neutral Beam launchers have been made. For the former, a Remote Steering concept has been considered, while for the latter a properly rescaled-ITER-like system has been taken into account. Secondly, CAD model of a DEMO sector has been modified to contain the port openings and a coarse port plug. Thirdly, a preliminary structural and electro-magnetic analysis has been carried out, considering two blanket concepts: Helium Cooled Lithium Lead and Helium Cooled Pebble Bed.
Is there, like Euler's identity, a simply expressed relationship, dense in meaning, between software engineering's most basic entities?
Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-CooledLithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required fortheir operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB-TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives.
Mock-ups of DEMO breeding blankets, called Test Blanket Modules (TBMs), inserted and tested in ITER in dedicated equatorial ports directly facing the plasma, are expected to provide the first experimental answers on the necessary performance of the corresponding DEMO breeding blankets. Several DEMO breeding blanket designs have been studied and assessed in the last 20 years. At present, after considering various coolant and breeder combinations, all the TBM concepts proposed by the seven ITER Parties use Reduced-Activation Ferritic/Martensitic (RAFM) steel as the structural material. In order to perform valuable tests in ITER, the TBMs are expected to use the same structural material as corresponding DEMO blankets. However, due to the fact that this family of steels is ferromagnetic, their presence in the ITER vacuum vessel will create perturbations of the ITER magnetic fields that could reduce the quality of the plasma confinement during H-mode. As a consequence, a legitimate question has been raised on the necessity of using RAFM steel for TBMs structural material in ITER. By giving a short description of the main TBM testing objectives in ITER and assessing the consequences of not using such a material, this paper gives a comprehensive answer to this question. According to the working group author of the study, the use of RAFM steel as structural material for TBM is judged mandatory.
This paper presents the status of the design and of the development programme of the two test blanket
systems (TBSs) based on the blanket concepts supported by the EU, namely the helium cooled lithium
lead (HCLL) and helium cooled pebble bed (HCPB) concepts.
Both the test blanket modules (TBMs) box design and the associated systems (Helium Cooling Systems,
PbLi loop for the HCLL system, helium processing systems for tritium extraction, etc.) have been revised
and, where needed, modified according to the assumption that one ITER equatorial port could be available
for testing the two European test blanket modules (TBMs).
According to EU TBMs programme, two reliable test blanket systems shall be ready for installation on
the first day of ITER operation. In order to comply with this ambitious objective, six EURATOM associates
who have sustained the TBM program so far have joined themselves in a consortium aiming to ensure an
efficientmanagementof the project tasks and exploit specificcompetences enhancing potential synergies.
The consortium objectives and development programme are summarised in the paper.
Accident cases are investigated for the Helium Loop Karlsruhe (HELOKA) facility, a high pressure and high temperature experimental helium loop having the European Helium Cooled Pebble Beds (HCPB) Test Blanket Module (TBM) as test module. Two typical operation modes for the loop operation have been numerically modeled using RELAP5-3D code: a pulsed operation (ITER-like situation) and a steady state operation (very long pulse). In both situations a maximum heat load on the first wall of 500kW/m2 has been considered. Using the generated RELAP-model for these two operating modes three accident cases are studied: failure of helium pressure control system (PCS), loss of helium flow and loss of cooling water flow. Simulation results show that the most critical case is the loss of helium flow. In this case failure of the TBM box is possible. By the loss of cooling water the integrity of the TBM box could be assured, if actions are taken in time. By the failure of the PCS the integrity of pressure barriers are not affected.
The scope of this paper is a presentation of the safety issues and of the expected risks associated with the operation of test blanket modules (TBMs) inside the ITER machine. The discussion on the expected risks is done to outline the magnitude of the risks related to the test blankets with respect to the risks expected to operate ITER without the TBMs. The discussion wants to be of general purpose and it does not want to substitute the detailed analyses done in the past and the ones that are presently on going to have a detailed estimation of the risks related to operate test blanket systems (TBSs). Only such detailed analyses will demonstrate compatibility of TBMs with ITER safety requirements.We had reason to address the question on magnitude of TBM risks recently, in conjunction with the preparation of a draft safety reports for the two European TBSs. The key safety issue is associated with unwanted large plasma-disruptions that have the potential to cut right trough the first wall of the TBMs. Such plasma disruptions are currently postulated to occur under several conditions, including the use of the plasma termination system, when called upon to act in response to various initiating events. Under some very unlikely conditions, the through-wall rupture of the test-blanket first wall can lead to exothermic reactions between water/steam and test-blanket materials (lithium and/or beryllium) releasing quantities of hydrogen inside the vacuum vessel. Such releases would be in addition to those estimated for ITER, under similar conditions, but without the presence of the TBSs. The paper discusses this and other safety issues.The conclusion from this work is that the additional risk introduced by the European TBSs is miniscule. This is particularly true if the plasma-physics experiments to be conducted in ITER demonstrate that ITER plasmas are very stable, which, of course, is a necessary condition for a demonstration of fusion power reactor.
In the frame of the WP08 EFDA Goal Oriented Training Program (GOTP) EUROBREED [1], the Working Package 1 is aimed at the design and procurement of a prototype of a Helium Cooled Pebble Bed Test Blanket Module Breeder Unit (HCPB TBM BU) and its testing in a dedicated experiment in a medium scale (<50 g/s) helium loop. This paper presents the design and analyses studies of the helium coolant fluid dynamics and of the thermal field in a HCPB TBM BU. After several design iterations, the helium coolant mass flow distribution has been improved between 3% and 7%. The analysis of the thermal field of the improved design has been performed. Considering steady state conditions with a heat flux deposed on the first wall equal to 500 kW/m(2) and a neutron wall load of 0.78 MW/m(2), the maximum temperature reached in the BU is 560.9 degrees C, about 2% above the maximum design temperature of the structural material (550 degrees C). As this maximum temperature is localized in a zone where there are not foreseen high stresses and the assumption for the analyses is conservative, the design presented for the Cooling Plates is considered as the reference for the production of the full scale BU mock-up.
In the context of ITER, CEA/IRFM has participated to the design and integration of several components in the Equatorial Port plug region. Particularly, in the framework of the grant F4E-2008-GRT-09-PNS-TBM, CEA/IRFM has contributed to the test blanket module system (TBS) design and robot access feasibility study in the Port Cell.Simulations of the maintenance procedure were studied and fully integrated to the design process, enabling to provide space reservation for human and robotic access. For this mean, CEA/IRFM has used a CEA LIST Virtual Reality simulation software directly integrated to the Solidworks CAD software. The feasibility to connect/dis-connect the pipes in front of the Bioshield by a set of potential standard industrial arms was demonstrated.Aiming to give more realism to maintenance scenario and CAD models, CEA IRFM has decided to build a Virtual Reality platform in the institute, integrated to the design office. With the expertise of CEA LIST, this platform aims to provide the nearest possible links between design and remote handling needs.This paper presents the outcome of the robot access study and discusses about the Virtual Reality tools that are being developed for these applications.
In the frame of the activities of the European TBM Consortium of Associates the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) is developed in Karlsruhe Institute of Technology (KIT). After performing detailed thermal and fluid dynamic analyses of the preliminary HCPB TBM design, the thermo mechanical behaviour of the TBM under typical ITER loads has to be assessed. A synthesis of the different design options proposed has been realized building two different assemblies of the HCPB-TBM: these two assemblies and the analyses performed on them are presented in this paper.Finite Element thermo-mechanical analyses of two detailed 1/4 scaled models of the HCPB-TBM assemblies proposed have been performed, with the aim of verifying the accordance of the mechanical behaviour with the criteria of the design codes and standards. The structural design limits specified in the codes and standard are discussed in relation with the EUROFER available data and possible damage modes. Solutions to improve the weak structural points of the present design are identified and the DEMO relevancy of the present thermal and structural design parameters is discussed.
An advanced integral approach has been implemented for neutronic analyses of the European test blanket modules (TBMs) in ITER. The central element of this approach is the use of the geometry conversion tool McCad for the generation of Monte Carlo analysis models from CAD geometry data. Following this approach, an MCNP model of the test blanket port plug with HCPB and HCLL TBM assemblies, elaborated by the European TBM Consortium of Associates (CA), was generated and integrated into the Alite MCNP model of ITER. Neutronic performance and shielding analyses were conducted on the basis of MCNP-5 calculations for the HCPB and HCLL TBMs and the entire shield system. The results indicated the need for a further optimization of the shield system complemented by a rigorous shutdown dose rate analysis.
Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.
The Helium Cooled Lithium Lead (HCLL) and the Helium Cooled Pebble Bed (HCPB) Blanket are the reference concepts in the European Breeding Blanket Programme for the DEMO design and for the related long term R&D. Recently, a similar design for both concepts has been developed, in particular both concepts use helium coolant and RAFM steel EUROFER as structural material. In this paper the interactions between the selected materials and the proposed DEMO designs are discussed. In particular the design features related to the tritium production, power extraction, material compatibility and fabrication processes are addressed. All these features contribute to the definition of DEMO concepts which are attractive for a future fusion power plant in terms of safety, availability and economics.
Investigation of accident cases for high pressure
- X Z Jin
- B.-E Ghidersa
X.Z. Jin, B.-E. Ghidersa, Investigation of accident cases for high pressure, high
temperature experimental helium loop using RELAP5-3D code, this conference.