Design and validation of breeder unit of helium-cooled pebble bed blanket for demo fusion reactor

In the framework of European helium-cooled pebble bed (HCPB) blanket development, an HCPB breeder unit based on the design of pebble beds between flat cooling plates is proposed for a DEMO fusion reactor. The performances of the designed breeder units are validated by supporting analyses. By applying the thermal boundary conditions obtained by neutronics simulations for the DEMO reactor, results of finite element calculations of the breeder unit are analyzed in views of thermal-hydraulics and thermal stress to identify the adherence to maximum temperatures in structural and functional materials and the abidance by the stress criterion imposed by the structural material. The layout of the internal meandering channels in the cooling plates is optimized by using numerical methods. Finally, possible improvements of the new designed breeder unit are proposed.     

On the numerical assessment of the thermo-mechanical performances of the DEMO Helium-Cooled Pebble Bed breeding blanket module

Within the framework of the European DEMO Breeder Blanket Programme, a research campaign has been launched by University of Palermo, ENEA-Brasimone and Karlsruhe Institute of Technology to theoretically investigate the thermo-mechanical behavior of the Helium-Cooled Pebble Bed (HCPB) breeding blanket module of the DEMO1 blanket vertical segment, under normal operation and over-pressurization loading scenarios.The research campaign has been carried out following a theoretical–computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. A realistic 3D FEM model of the HCPB blanket module central poloidal–radial region has been developed, including one breeder cell in the toroidal direction and all the five cells in the poloidal one. No Breeder Units have been modeled, their presence being simulated by effective thermo-mechanical loads.Two sets of uncoupled steady state thermo-mechanical analyses have been carried out with reference to the investigated loading scenarios. In particular, under normal operation scenario (level A) the module has been supposed to undergo both 8 MPa coolant pressure on its cooling channel walls and thermal deformations due to the flat-top plasma operational state thermal field, while under over-pressurization scenario (level D) it has been assumed to experience 8 MPa coolant pressure on its internal walls while operating at normal operation thermal level. Results obtained are presented and critically discussed according to the SDC IC code.     

Overview of EU DEMO design and R&D activities

One important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several 100 MW of net electricity to the grid and operating with a closed fuel-cycle by 2050. This is currently viewed by many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. This paper outlines the DEMO design and R&D approach that is being adopted in Europe and presents some of the preliminary design options that are under evaluation as well as the most urgent R&D work that is expected to be launched in the near-future. The R&D on materials for a near-term DEMO is discussed in detail elsewhere.     

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

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

Neutronic analyses of the HCPB DEMO reactor using a consistent integral approach

This work is devoted to nuclear design analyses of the new HCPB-type DEMO reactor developed in the frame of the EFDA PPPT program. The neutronic simulations were carried out with the MCNP5 code using a full scale 3D torus sector model of the DEMO reactor. The model was generated with the McCad conversion tool from available CAD models using a consistent integral approach. The neutronic analyses addressed the tritium breeding performance, the nuclear power generation and the shielding capabilities of the reactor. Although tritium self-sufficiency was shown, the tritium breeding performance of the current design calls for further design improvements to arrive at a higher uncertainty margin. The shielding performance of the reactor is close to the limit. Sufficient shielding can be easily provided by a slight increase of the inboard shield thickness.     

Nuclear analysis of DEMO water-cooled blanket based on sub-critical water condition

For the water-cooled solid blanket of DEMO, the nuclear analysis was performed based on present cooling piping system. Especially, distributions of neutron load and temperature were calculated with P_n is 5 MW/m². Furthermore, the local TBR was optimized by changing the material proportion for each P_n level (1-5 MW/m²). It was confirmed that the size of cooling loop for sub-critical water could be used as about 2000 × 450 mm and the cooling pipe diameter of D is 12 mm, d is 9 mm at v is 5.36 m/s. The pipe pitches would vary with P_n level which is related to the blanket structure design. Nuclear heat distribution is the base to decide the distribution of cooling pipe positions. It was found that the local TBR of blanket would be dropped down along with the P_n level rising which was mainly depended on the thickness of beryllium variation. Finally, the layout of cooling pipes for each level was obtained.     

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.     

聚变制氢堆高温液态包层热工水力学新概念研究

在深入分析聚变堆包层设计要求和目前技术发展水平的基础上,根据热化学工艺制氢需要高温热的要求,提出了一个基于技术相对成熟的低活化铁素体/马氏体钢作为主要结构材料、高压氦气与液态LiPb合金作为冷却剂、具有创新性“多层流道插件”结构方案以获得高温热能的包层热工水力学概念,建立了热工水力学模型,在利用有限元数值模拟程序进行模拟计算的基础上分析了这种新概念包层的可行性。     

Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory.Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper.For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available.Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE.In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.     

Influence of helium cooling channels on magnetohydrodynamic flows in the HCLL blanket

One of the blanket concepts proposed to be tested in ITER as part of the test blanket module program of the European Union is the helium cooled lead lithium blanket design. In this configuration the so called breeder units are arranged in an array, separated by a stiffening grid, to form blanket modules. The deposited thermal energy is removed by helium flowing at high pressure and speed in channels integrated both in the walls and in cooling plates that subdivide the breeder units into flat ducts where the lead lithium circulates under the influence of the strong plasma confining magnetic field. This gives rise to magnetohydrodynamic (MHD) phenomena whose effects on flow distribution have to be investigated to evaluate the performance of the proposed design. The established MHD flow is affected by the presence of helium channels in cooling and stiffening plates that results in non-homogeneous wall conductance.In support to the conceptual study of a liquid metal blanket, numerical investigations of fully developed MHD flows in a central cross-section of breeder units have been performed, taking into account both the presence of helium channels in the walls and the multi-channel effects caused by the exchange of currents through walls separating different fluid domains.     

DEMO水冷包层第一壁结构优化设计研究

针对聚变示范堆（DEMO）水冷包层,通过计算流体力学程序CFX和计算结构力学程序ANSYSWorkbench中的SIMULATION模块进行单向流固耦合分析。在对现有设计的DEMO水冷包层第一壁温度和应力数值模拟分析的基础上,改变了第一壁流道结构,着重研究了不同流道结构下的温度和应力分布,分析了几何结构对最高温度和最大应力的影响,提出第一壁结构的优化设计方案。数值模拟结果表明,优化设计方案能有效降低第一壁结构中的最高温度和最大应力。     

Design and cooling of the edge segments of the DEMO divertor target plates

In support of shadowing of the divertor target plate edges in toroidal direction against damage caused by the incident particles, the fingers at the boundary of the target plate should ideally form a flat surface. The reference cooling fingers are of hexagonal shape and when assembled together, their edge boundary cannot be flat. Therefore, the boundary segments need to be designed in a different way. Three possible designs are investigated: non-symmetric pentagonal fingers and two square-shaped fingers of different sizes, all cooled by the same type of concentric cartridge as in the reference design. Their heat transfer performance is analyzed from the point of view of maximum allowable temperature of the thimble structure. The computational fluid dynamics (CFD) analysis is performed to obtain the minimum mass flow rate of the coolant which is necessary to keep the structure's temperature below the permissible limit at an acceptable pressure loss.     

CFD analysis of the ITER first wall 06 panel. Part II: Thermal-hydraulics

The computational fluid dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part II we concentrate on the thermal-hydraulics of the water coolant, driven by the nuclear volumetric and plasma surface heat loads discussed in Part I. Both the detailed steady state analysis of a single cooling channel and the coarse transient analysis of the whole panel are considered. The compatibility of the hot spots with the maximum recommended temperatures for the different materials is confirmed. The heat transfer coefficient between coolant and walls is obtained post-processing the results of the simulation and compared with the results of available correlations, which may be used for simpler analyses: in the fully developed flow regions of the cooling pipes, it turns out to be well approximated by the Sieder–Tate correlation. The operation margin with respect to the critical heat flux is also computed and turns out to be sufficiently large compared with the design limit.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part I: Flow distribution and pressure drop

The 3D steady-state Computational Fluid Dynamics (CFD) analysis of the ITER vacuum vessel (VV) regular sector #5 is presented, starting from the CATIA models and using a suite of tools from the commercial software ANSYS FLUENT^®. The peculiarity of the problem is linked to the wide range of spatial scales involved in the analysis, from the millimeter-size gaps between in-wall shielding (IWS) plates to the more than 10 m height of the VV itself. After performing several simplifications in the geometrical details, a computational mesh with ～50 million cells is generated and used to compute the steady-state pressure and flow fields from a Reynolds-Averaged Navier–Stokes model with SST k-ω turbulence closure. The coolant mass flow rate turns out to be distributed 10% through the inboard and the remaining 90% through the outboard. The toroidal and poloidal ribs present in the VV structure constitute significant barriers for the flow, giving rise to large recirculation regions. The pressure drop is mainly localized in the inlet and outlet piping.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part II: Thermal-hydraulic effects of the nuclear heat load

The 3D Computational Fluid Dynamic (CFD) steady state analysis of the regular sector #5 of the ITER vacuum vessel (VV) is presented in these two companion papers using the commercial software ANSYS-FLUENT®. The pure hydraulic analysis, concentrating on flow field and pressure drop, is presented in Part I. This Part II focuses on the thermal-hydraulic analysis of the effects of the nuclear heat load. Being the VV classified as safety important component, an accurate thermal-hydraulic analysis is mandatory to assess the capability of the water coolant to adequately remove the nuclear heat load on the VV. Based on the recent re-evaluation of the nuclear heat load, the steady state conjugate heat transfer problem is solved in both the solid and fluid domains. Hot spots turn out to be located on the surface of the inter-modular keys and blanket support housings, with the computed peak temperature in the sector reaching ～290 °C. The computed temperature of the wetted surfaces is well below the coolant saturation temperature and the temperature increase of the water coolant at the outlet of the sector is of only a few °C. In the high nuclear heat load regions the computed heat transfer coefficient typically stays above the 500 W/m² K target.     

聚变驱动次临界堆双冷嬗变包层中子学设计与分析

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

聚变驱动次临界堆双冷嬗变包层结构设计与分析

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

CFD analysis of the ITER first wall 06 panel. Part I: Model set-up and flow distribution

The first detailed Computational Fluid Dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part I we introduce the problem, define the model together with its input and discuss the results with particular reference to the hydraulics of the water coolant. The pressure drop across the panel is computed, together with the distribution of the flow among the different channels. Different design options are studied, with particular reference to the minimization of stagnation/recirculation regions.     

ITER中国液态锂铅实验包层模块氦气冷却系统初步设计研究

氦气冷却系统是ITER中国液态锂铅实验包层模块(DFLL-TBM)在ITER装置内进行实验的重要辅助系统.根据ITER运行时的热工条件、安全要求、空间要求,分析了DFLL-TBM氦气冷却系统的功能,确定氦气冷却系统的设计原则和要求,在此基础上给出氦气冷却系统的初步设计方案和设备布置.该氦气系统的特点体现在:双功能,即有宽的裕量满足SLL-TBM和DLL-TBM实验;两条氦气回路共享压力控制单元和氦气净化子系统;旁路设计调节TBM和热交换器氦气的出口温度.     

长寿命裂变产物在聚变驱动次临界堆包层中嬗变的中子学优化分析

商用裂变堆乏燃料中高放长寿命裂变产物(LLFP)由于其具有很强的放射毒性,所以对于它们的嬗变处理非常重要。在对世界上关于LLFP嬗变处理的广泛调研的基础上,考虑到LLFP的同位素分离技术的发展水平,选择了LLFP中99Tc、129I和135Cs的嬗变处理(?)料的化学形式,分析了不同慢化剂材料对嬗变能力的影响,同时针对聚变驱动次临界堆的多功能双冷核废料嬗变包层(DWTB)进行了LLFP嬗变的中子学设计和优化分析。