堆内辐照在线产氚氦冷陶瓷氚增殖剂包层球床组件热工流体力学计算分析

在研究堆内进行辐照在线产氚试验是ITER计划专项国内配套研究项目之一。本工作主要针对研究堆内辐照氦冷陶瓷氚增殖剂包层（简称陶瓷）球床组件的试验技术要求，评估辐照陶瓷球床组件设计方案的可行性。通过对不同的陶瓷球床组件结构参数和组件在堆内的不同辐照位置，进行热工流体力学设计计算，得到满足要求的入堆辐照陶瓷球床组件设计方案。     

氚增殖剂球床组件堆内辐照物理热工计算分析

为验证在中国先进研究堆(CARR)内进行国际热核聚变实验堆(ITER)氚增殖包层模块(TBM)辐照实验的可行性和安全性,进行了氚增殖剂球床组件堆内辐照物理及热工计算分析。氚增殖剂包层模块主要是固态氚增殖剂陶瓷球床。本文采用Monte Carlo粒子输运模拟程序对氚增殖剂球床进行堆内建模,计算球床的中子注量率、能量沉积和产额,得到不同功率下球床的中子注量率、发热功率和产氚速率以及球床组件引入反应堆的反应性。根据物理计算得到的组件各部件发热情况建立热工计算一维模型,通过更改反应堆功率得到满足实验要求的工况并采用三维程序进行验证。物理与热工计算分析的结果表明,在反应堆运行功率为20 MW的工况下球床组件各部件的温度均不超过限值。     

堆内辐照氚增殖剂球床有效热导率分析

氘-氚聚变反应堆中,固态氚增殖剂包层能不断为聚变反应提供氚核素,是实现聚变反应堆商用的关键技术之一。由锂陶瓷小球堆积形成的球床形式的固态氚增殖剂包层具有比表面积大、产氚效率高等优点,是我国重点发展的氚增殖剂包层形式。氚增殖剂球床须能支撑在堆内辐照时的高温环境,这就要求氚增殖剂球床有较好的导热特性。球床的有效热导率在球床设计和辐照过程中的安全分析十分重要,因此在中国先进研究堆(CARR)开展了氚增殖剂球床在堆内辐照环境下的有效热导率测量实验。根据MCNP计算得出的球床发热功率,结合实验测量的球床温度分布反推得到氚增殖剂球床的有效热导率,并与广泛应用于球床有效热导率计算的改进型ZBS模型计算结果以及堆外实验结果进行对比分析,理论值与实验值能较好吻合。     

堆内辐照氚增殖剂球床有效热导率分析

氘-氚聚变反应堆中，固态氚增殖剂包层能不断为聚变反应提供氚核素，是实现聚变反应堆商用的关键技术之一。由锂陶瓷小球堆积形成的球床形式的固态氚增殖剂包层具有比表面积大、产氚效率高等优点，是我国重点发展的氚增殖剂包层形式。氚增殖剂球床须能支撑在堆内辐照时的高温环境，这就要求氚增殖剂球床有较好的导热特性。球床的有效热导率在球床设计和辐照过程中的安全分析十分重要，因此在中国先进研究堆（CARR）开展了氚增殖剂球床在堆内辐照环境下的有效热导率测量实验。根据MCNP计算得出的球床发热功率，结合实验测量的球床温度分布反推得到氚增殖剂球床的有效热导率，并与广泛应用于球床有效热导率计算的改进型ZBS模型计算结果以及堆外实验结果进行对比分析，理论值与实验值能较好吻合。     

铝基碳化硼材料堆内辐照方案设计

铝基碳化硼是一种新型的乏燃料贮存格架用材料,为检验其辐照性能,需进行堆内辐照实验。本文从样品成分及形状、辐照罐结构、辐照位置等方面,对铝基碳化硼材料堆内辐照方案进行设计。经初步中子物理学和热工计算表明：在所选择的两个辐照孔道内进行辐照考验,试件所接受的累积γ射线照射剂量和相应的快中子积分注量均满足技术要求,且辐照罐样品入堆后对功率峰值因子、反应性、发热率等与堆运行安全相关因子的影响均在安全范围内。     

在线产氚辐照装置物理参数模拟

在线产氚回路对我国氚增殖模块（TBM）增殖剂候选材料的考核、氚增殖剂材料的在线释放规律研究具有重要意义,辐照装置是在线产氚回路的关键部件。本工作采用MCNP程序模拟在线产氚辐照装置在堆内辐照时的物理参数,计算结果如下：自屏因子为0.430,等效反应截面为1.09×10^-22cm²,每日产量为2.8×10¹⁰Bq,总发热功率为8.2kW。模拟计算结果为该装置的设计提供了必需的数据支持。     

泳池式轻水反应堆内电磁线圈可控温辐照装置设计

本文设计了在泳池式轻水反应堆（简称泳池堆）内在线测量电磁线圈电性能的可控温辐照装置。采用MCNP程序进行中子物理计算，对泳池堆、线圈骨架的结构尺寸与物质组分进行了精细全尺寸模拟，得出辐照装置的发热功率和中子注量率。通过初步估算，使用ANSYS CFX进行了数值模拟，得出辐照装置内线圈在堆运行时的温度，并提出温度控制的方法。辐照装置采用铝材加工制造，并进行了垂直度测试、气压测试、检漏测试。增加了绝缘设计，将辐照装置与泳池堆之间进行绝缘。在线圈处预埋铠装热电偶，对线圈温度进行实时监测。在泳池堆内对电磁线圈进行辐照试验，结果表明，本文设计的辐照装置能满足电磁线圈在泳池堆孔道内进行辐照试验的要求，并可对电磁线圈进行实时温度控制。     

随机填充球床通道内氦气流动特性实验研究

在聚变堆氦冷固态包层氚增殖区，球床通道内氦气流动压降特性对泵功率的设计具有重要意义。以氦冷固态包层氚增殖区为背景，研究了氦气流速、球床颗粒直径及球床通道长度对球床通道内氦气流动压降特性的影响。实验段采用20 mm×20 mm×500 mm的矩形通道，实验中氦气流速为0.1～0.6 m/s，球床颗粒直径为0.5、0.8、1.0、1.5、2.0 mm。实验结果表明，压降与氦气流速以及球床通道长度呈正相关，与球床颗粒直径呈负相关。对比Ergun关系式发现，在球床颗粒直径较小时，Ergun关系式预测值低于实验值，这主要是由于氦气可压缩性的影响。通过动量方程，理论推导出经可压缩性修正的Ergun关系式，结果发现修正后的Ergun关系式预测值与实验值符合良好。本研究为氦冷固态包层氚增殖区设计提供了数据支撑，为球床通道内流动特性的数值模拟提供了验证手段。     

CENTER燃料组件堆内辐照考验燃耗计算研究

为验证中国工程试验堆（CENTER）燃料组件设计，在燃料组件正式定型前需开展组件辐照考验，CENTER燃料组件在高通量工程试验堆（HFETR）内采用随堆辐照方式进行辐照考验。根据CENTER燃料组件特点，开展了HFETR辐照考验CENTER燃料组件燃耗计算方法研究，确定了CENTER燃料组件辐照考验堆芯物理计算采用镶嵌耦合方法。结果表明，燃料组件平均燃耗计算值与测量值偏差为3.25%，满足辐照考验要求。      

基于蒙特卡罗方法的固态氚增殖剂聚变中子辐照损伤行为分析

实验包层模块(TBM)是聚变反应堆最重要的组件之一,作用是产氚和能量提取。锂陶瓷具有良好的化学稳定性、热机械性能、产氚性能以及可在更高温度下使用等特点,被认为是聚变堆包层最具吸引力的氚增殖剂材料。中国ITER-TBM设计方案采用了氦冷固态氚增殖剂(HCCB)TBM结构,其聚变环境下的辐照损伤行为可为中国HCCB TBM结构设计提供支持。针对固态氚增殖剂聚变中子辐照损伤问题,利用蒙特卡罗模拟,对比分析了Li_4SiO_4和Li_2TiO_3的中子辐照离位损伤和嬗变气体损伤。结果表明:在相同的服役时间下,Li_4SiO_4比Li_2TiO_3将产生更多的嬗变气体,且在高6 Li丰度情况下,其中子辐照损伤也更严重,会产生更高的损伤剂量和更大的损伤截面。但是,嬗变气体所造成的空位损伤Li_2TiO_3要比Li_4SiO_4严重;对两种陶瓷材料来讲,氦损伤效应均强于氚损伤效应。     

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

ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R&D activities for each TBM module with the auxiliary system are introduced.The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li₄SiO₄ pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R&D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.     

硅酸锂球床传热传质特性与吹扫气体流动的DEM-CFD耦合模拟

氦冷固态增殖包层是中国聚变工程实验堆（CFETR）的3种候选包层概念之一,氚增殖球床是包层的核心部件,采用硅酸锂颗粒作为氚增殖材料。球床结构对氚在球床内的输运行为及流动和传热均有重要影响。本文基于离散单元法（DEM）生成了满足氚增殖球床填充率要求的随机堆积结构,通过CFD计算获取了球床结构下氚在吹扫气体内的等效扩散系数及吹扫气体的流动特性,包括速度分布、压力分布及进出口压降;开展了外加热流及有内热源两种工况下球床等效导热系数的模拟。计算结果表明,球床结构下氚在吹扫气体内的等效扩散系数为二元气体扩散系数的40%;受球床结构影响,球床内存在流动迟滞区,壁面出现流动加速;拟合得到Ergun方程的黏性阻力系数C1=87;有内热源工况下的球床等效导热系数低于外加热流工况下的球床等效导热系数。     

Study of ceramic pebble beds in Post Irradiation Examination of the Pebble Bed Assemblies irradiation experiment

The irradiation experiment Pebble Bed Assemblies (PBA) consists of four mock-up representations (test elements) of the EU Helium Cooled Pebble Bed (HCPB) concept. The four test elements contain a ceramic breeder pebble bed sandwiched between two beryllium pebble beds and are regarded as one of the first DEMO representative HCPB blanket irradiation tests, with respect to temperatures and power densities. The design value of the PBA were to irradiate pebble beds at a power density of 20–26 W/cc in the ceramic breeder, to a maximum temperature of 800 °C.Two test elements contain lithium orthosilicate pebbles (Li₄SiO₄; FZK/KIT) and were irradiated with target temperatures of 600 and 800 °C, respectively. The other test elements have lithium metatitanate (Li₂TiO₃; CEA) with different grain sizes and were both irradiated with a target temperature of 800 °C. The PBA have been irradiated for 294 Full Power Days (12 cycles) in the High Flux Reactor (HFR) in Petten to a total neutron dose of 2–3 dpa in Eurofer, and an estimated (total) lithium burnup of 2–3% in the ceramic pebbles.This work presents results of Post Irradiation Examinations (PIE) on the four HCPB test elements. Using e.g. SEM, the evolution of compressed pebble beds and pebble interactions like swelling, creep, sintering, etc., under irradiation and thermal loads are studied for the candidate pebble materials Li₂TiO₃ and Li₄SiO₄. (Chemical) interactions between ceramic pebbles and Eurofer (e.g. chrome diffusion) are observed. Looking at different sections of the pebble beds, correlations between temperatures and thermal–mechanical behaviour are clearly observed.     

中国氦冷固态增殖剂实验包层模块材料研究进展

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

Development of the IFMIF Tritium Release Test Module in the EVEDA phase

This paper presents the engineering design of the IFMIF (International Fusion Materials Irradiation Facility) Tritium Release Test Module (TRTM). The objectives of the TRTM are: (i) in-situ measurements of the tritium released from lithium ceramics and beryllium pebble beds during irradiation, (ii) studying the chemical compatibility between lithium ceramics and structural materials under irradiation, and (iii) performing post irradiation examinations for the irradiated materials. The TRTM has eight rigs which are arranged in two rows (2 × 4) perpendicular to the beam axis and enclosed by a structural container. Each rig includes one capsule that contains lithium ceramic or beryllium pebbles for irradiation. Neutrons reflectors are implemented at different locations to reflect the scattered neutrons back to the active region aiming to improve the tritium production. The TRTM is required to provide irradiation temperature range of 400–900 °C for the capsules filled with lithium ceramics and 300–700 °C for the ones packed with beryllium. The engineering design of the TRTM components such as container, rigs, capsules, pebble beds, neutrons reflectors, and purge gas and coolant tubes are presented. In addition a test matrix for the irradiation campaign is proposed.     

Design and R&D progress of Korean HCCR TBM

Korea plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER. The HCCR TBM adopts a four sub-module concept considering the fabricability and the transfer of irradiated TBM for post irradiation examination. Each sub-module has seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebble bed packed tritium breeder layers, and a reflector layer packed with graphite pebbles. Based on this configuration, neutronic and electromagnetic calculations were performed and their results were applied for the conceptual design of HCCR TBM that considers manufacturing feasibility. Also, a design and safety analysis of HCCR Test Blanket System (TBS) was performed using integrated design tools modifying nuclear system codes for helium coolant and tritium behavior evaluation. The Advanced Reduced Activation Alloy (ARAA) is being developed as a structural material. A total of 73 candidate ARAA alloys were designed and their out-of-pile performance was evaluated. The graphite pebbles as the neutron reflector were fabricated by using mechanical machining and grounding method with the surface coated with SiC. The hydrogen permeation characteristics of structural materials were evaluated using the Hydrogen PERmeation (HYPER) facility. The recent design and R&D progress on these areas are addressed in this paper.     

Status and progress of Indian LLCB test blanket systems for ITER

The lead–lithium ceramic breeder (LLCB) TBM and its auxiliary systems are being developed by India for testing in ITER machine. The LLCB TBM consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble beds. The FW structural material is ferritic martensitic steel cooled by high-pressure helium gas and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder pebble bed to extract the nuclear heat from the CB zones. Low-pressure helium is purged inside the CB zone for in situ extraction of bred tritium. Currently the LLCB blanket design optimization is under progress. The performance of tritium breeding and high-grade heat extraction is being evaluated by neutronic analysis and thermal–hydraulic calculations for different LLCB cooling configurations and geometrical design variants. The LLCB TBM auxiliary systems such as, helium cooling system (HCS), lead–lithium cooling system (LLCS), tritium extraction system (TES) process design are under progress. Safety analysis of the LLCB test blanket system (TBS) is under progress for the contribution to preliminary safety report of ITER-TBMs. This paper will present the status of the LLCB TBM design, process integration design (PID) of the auxiliary systems and preliminary safety analysis results.