Neutronic study on the effect of first wall material thickness on tritium production and material damage in a fusion reactor

In this study,the effects of changing first wall materials and their thicknesses on a reactor were investigated to determine the displacement per atom(DPA)and gas production(helium and hydrogen)in the first wall,as well as the tritium breeding ratio(TBR)in the coolant and tritium breeding zones.Therefore,the modeling of the magnetic fusion reactor was determined based on the blanket parameters of the International Thermonuclear Experimental Reactor(ITER).Stainless steel(SS 316 LNIG),Oxide Dispersion Strengthened Steel alloy(PM2000 ODS),and China low-activation martensitic steel(CLAM)were used as the first wall(FW)materials.Fluoride family molten salt materials(FLiBe,FLiNaBe,FLiPb)and lithium oxide(LiO_(2))were considered the coolant and tritium production material in the blanket,respectively.Neutron transport calculations were performed using the wellknown 3D code MCNP5 using the continuous-energy Monte Carlo method.The built-in continuous energy nuclear and atomic data libraries along with the Evaluated Nuclear Data file(ENDF)system(ENDF/B-V and ENDF/B-VI)were used.Additionally,the activity cross-section data library CLAW-IV was used to evaluate both the DPA values and gas production of the first wall(FW)materials.An interface computer program written in the FORTRAN 90 language to evaluate the MCNP5 outputs was developed for the fusion reactor blanket.The results indicated that the best TBR value was obtained for the use of the FLiPb coolant,whereas depending on the thickness,the first wall replacement period in terms of radiation damage to all materials was between 6 and 11 years.

Diagnostics of First Wall Materials in a Magnetically Confined Fusion Device by Polarization-Resolved Laser-Induced Breakdown Spectroscopy

Laser-induced breakdown spectroscopy （LIBS） is a powerful analytical tool for real- time diagnostics and detection of multiple elements deposited at the first wall of magnetically confined plasma ft^sion devices. Recently, we have tested LIBS in our laboratory for application to in situ real-time diagnostics in the fusion device EAST. In this study, we applied polarization- resolved LIBS （PR-LIBS） to reduce the background continuum and enhance the resolution and sensitivity of LIBS. We used aluminium （A1） （as a substitute for Be） and the first wall materials tungsten （W） and molybdenum （Mo） to investigate polarized continuum emission and signal-to- background ratio （SBR）. A Nd：YAG laser with first, second and third harmonics was used to produce plasma. The effects of the laser polarization plane, environmental pressure and polarizer detection angle were investigated. The spectra obtained without using a polarizer （i.e. LIBS） were compared with those obtained with a polarizer （PR-LIBS）. Distribution of emission spectral intensity was observed to follow Malus＇ law with respect to variation in the angle of detection of the polarizer. The spectra obtained by PR-LIBS had a higher SBR and greater stability than those obtained by LIBS, thereby enhancing the reliability of LIBS for quantitative analyses. A comparison of A1, Mo and W showed that W exhibited a higher continuum with stronger polarization than the low-Z elements.

Effects of a liquid lithium curtain as the first wall in a fusion reactor plasma

This paper explores the effect of a liquid lithium curtain on fusion reactor plasma, such curtain is utilized as the first wall for the engineering outline design of the Fusion Experimental Breeder （FEB-E）. The relationships between the surface temperature of a liquid lithium curtain and the effective plasma charge, fuel dilution and fusion power production have been derived. Results indicate that under normal operation, the evaporation of liquid lithium does not seriously affect the effective plasma charge, but effects on fuel dilution and fusion power are more sensitive. As an example, it has investigated the relationships between the liquid lithium curtain flow velocity and the rise of surface temperature based on operation scenario II of the FEB-E design with reversed shear configuration and high power density. Results show that even if the liquid lithium curtain flow velocity is as low as 0.5 m/s, the effects of evaporation from the liquid lithium curtain on plasma are negligible. In the present design, the sputtering of liquid lithium curtain and the particle removal effects of the divertor are not yet considered in detail. Further studies are in progress, and in this work implication of lithium erosion and divertor physics on fusion reactor operation are discussed.

Analysis of Plasma-Driven Tritium Permeation Through the First Wall of DFLL-TBM in ITER

Tritium permeation through the first wall （FW） from the plasma into helium coolant is evaluated for a dual-functional lithium-lead test blanket module （DFLL-TBM）. The effect of the surface conditions on the plasma facing and coolant sides, both temperature gradient and beryllium layer clad on the plasma facing side, as well as trapping in defects on the tritium permeation is considered. The results show that most of the tritium implanted in FW re-entered the plasma. The plasma-driven tritium permeation is very sensitive to the surface conditions on the plasma facing side. With a higher sticking coefficient on the plasma-facing side, the tritium permeation into helium coolant is significantly reduced. The tritium permeation is strongly reduced with a beryllium layer clad on the front side of FW. The plasma driven tritium permeation will not seriously impact the tritium safety of DFLL-TBM. Based on tritium safety, it is reasonable to clothe the beryllium layer on FW and keep the surface clean to reduce the plasma driven tritium permeation.

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

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

Thermo-mechanical Design Considerations for First Wall of A-SSTR2

The finite element analysis and calculation were performed for the blanket first-wall made of SiC/SiC composite material for Advanced Steady-state Tokamak Reactor 2, A-SSTR2, which at present is conceptually designed in Naka Fusion research establishment, JAERI. Comparison analysis and design window were analyzed using the finite element code ADINA 7.4. Through a 2D calculation for various geometrical configurations and sensitive material properties, a fundamental guideline for the first wall and blanket design are established with respect to maximum temperature, thermal and mechanical stress for many configurations. To satisfy hydrodynamic requirement, a4d4 (the dimension of coolant channel is 4 mm x 8 mm, and the distance between neighboring channels is 4 mm) was chosen as a design point for high thermal conductivity up to 50 W/m.K. In order to find a good solution for lower conductivity, more elaborate work should be done in the future. Nonetheless, the outline of design window for a specific structural material is very useful for the future A-SSTR2 first wall design.

Structural model for the first wall W-based material in ITER project

The preparation, characterization, and test of the first wall materials designed to be used in the fusion reactor have remained challenging problems in the material science. This work uses the firstprinciples method as implemented in the CASTEP package to study the influ ences of the doped titanium carbide on the structural sta bility of the WTiC material. The calculated total energy and enthalpy have been used as criteria to judge the structural models built with consideration of symmetry. Our simulation indicates that the doped TiC tends to form its own domain up to the investigated nanoscale, which implies a possible phase separation. This result reveals the intrinsic reason for the composite nature of the WTiC material and provides an explanation for the experimen tally observed phase separation at the nanoscale. Our approach also sheds a light on explaining the enhancing effects of doped components on the durability, reliability, corrosion resistance, etc., in many special steels.

Coupled thermal-mechanical analysis of two ITER-like first wall mockups under heat shock of plasma disruptions

The first wall of the fusion reactor is a plasma-facing component and is a key link to maintain the integrity of structure during thermal shock induced by plasma disruptions. Be and W/Cu functionally graded materials are two kinds of important plasma-facing materials(PFM) of first wall in fusion reactor currently. Previous researches seldom comparatively evaluated the normal servicing and heat shock resistance performance of first walls with those two kinds of PFMs. And also there lacks coupled thermal/mechanical analysis on the heat shock process in consideration of multiple thermal/mechanical phenomena, such as material melting, solidification, evaporation, etc., which is significant to further understand the heat shock damage mechanism of the first wall with different PFMs. With the aim of learning more detailed mechanical mechanism of thermal shock damage and then improving the thermal shock resistance performance of different first wall designs, the coupled thermal/mechanical response of two typical ITER-like first walls with PFM of Be and functionally graded W-Cu respectively under the heat shock of 1–2 GW/m^2 are computed by the finite element method. Special considerations of elastic-plastic deformation, material melting, and solidification are included in numerical models and methods. The mechanical response behaviors of different structures and materials under the normal servicing operation as well as plasma disruption conditions are analyzed and investigated comparatively. The results reveal that heat is mainly deposited on the PFM layer in the high energy shock pulse induced by plasma disruptions, resulting in complex thermal stress change as well as mechanical irreversible damage of thermal elastic and plastic expansion, contraction and yielding. Compared with the first wall with Be PFM, which mitigates the damages from heat shock at most only in the PFM layer with cost of whole PFM layer plastic yielding, the first wall with graded W-Cu PFM is demonstrated to be possessed both of higher heat shock resistance performance and normal servicing performance, provided its material gradient and cooling capacity are well optimized under practical loading conditions.

A Resection of the Giant First Left Rib Tumor and Chest Wall Reconstruction by Transmanubrial Osteomuscular Sparing Approach

A 64-year-old female had noticed an 11 × 6 cm mass growing on her left first rib. We performed a resection of the first and second ribs and a reconstruction of the chest wall. A thoracotomy was performed at the anterolateral second intercostal space. The second rib cartilage was divided at the left parasternum. Based on a transmanubrial osteomuscular sparing approach, the left-upper part of the sternum and the first rib cartilage were both cut at the left clavicular-sternum joint. The posterior parts of the two ribs involving the tumor were resected at the transverse process of the vertebral bone by tearing off the anterior, middle, and posterior scalene muscles, subclavicular artery and vein. The defect size of the thorax was 15 × 9 cm, which was reconstructed by covering with a polytetrafluoroethylene dual mesh (Dual mesh, Gore tex, 2 mm). The major pectoral muscle flap was used to cover the mesh. The postoperative pathological examination diagnosed a poorly differentiated fibrosarcoma. Eventually, she had palliative therapy for the postoperative metastatic chest wall. She died 14 months after the operation.

墙顶支撑对开挖前抽水引发基坑变形控制效果研究

基坑内抽水可诱发基坑围挡向坑内偏转并进而导致坑外土体随动沉降;在基坑土方开挖前,由于基坑内的支撑体系尚不能被完整架设,仅坑内土体可起到约束围挡变形的作用,此时,坑内抽水引起的围挡偏转十分明显。本文系统研究了围挡顶部支撑对土方开挖前抽水引发基坑变形的控制效果,首先通过抽水试验探究了坑内外强水力连通时墙顶支撑对抽水引发基坑变形的影响,并建立了三维数值计算模型,研究了若干典型参数(如,坑内降水深度等)对围挡顶部支撑控制基坑变形效果的影响。研究发现,围挡顶部支撑仅能有效限制浅埋范围内的围挡侧移(本文约埋深11 m范围内),对深埋位置围挡侧移的限制效果十分有限,且随着降水深度的增加,变形控制效果越弱;对于开挖前抽水深度较大的工程,仅采用围挡顶部设置支撑的方法可能无法有效限制围挡侧移的发展;另外,围挡顶部支撑仅能有效限制基坑围挡后方一定范围内(本文约10 m)的地表沉降,且当坑内外有水力联系时,由于坑外水位下降是引起地面沉降的主导因素,围挡顶部支撑并不能明显体现限制坑外最大地面沉降的作用,此时,应结合坑外地下水回灌进行沉降控制。

ITER氦冷固态实验包层模块第一壁氢同位素双向输运数值分析

氦冷固态实验包层模块(HCCB-TBM)安装于国际热核聚变实验堆(ITER)中,用以验证HCCB包层概念的氚增殖能力与热移出能力。HCCB-TBM第一壁用于承受堆芯等离子体粒子轰击和包容内部功能材料。外侧等离子体驱动氘、氚粒子渗透与内侧氢分压驱动渗透的同时存在,形成了第一壁的双向氢同位素输运。此双向输运可能对第一壁外表面再循环系数、包层增殖氚的纯化产生重要影响。基于商业软件COMSOL建立第一壁双向氢同位素输运模型,研究第一壁的氢同位素的输运特征。仿真结果表明:第一壁中的冷却剂流道具有强的氢同位素移出能力,使得双向输运解耦合;在ITER等离子体脉冲周期中,放电过程中已扩散到材料内部的氚在等离子体关停时扩散回流到真空室侧,关停时的回流将降低向冷却剂流道的氚渗透损失。

HL-3装置碳基第一壁螺栓预紧力分析

通过对HL-3装置第一壁基本模块进行高温下的螺栓预紧力分析,获得第一壁结构的应力分布及螺栓预紧力的变化情况,为碳基第一壁结构的现场安装提供数据支撑。

聚变堆包层模块第一壁不同冷却剂传热性能研究

[目的]核聚变作为一种清洁、高效的能源,是实现全球可持续发展的未来希望。针对中国氦冷固态增殖剂包层模块的第一壁,基于核热工安全与标准化研究团队提出的4根冷却剂道的设计方案,计算了氦气、氩气、氮气作为冷却剂的温度场。[方法]选择B.S.Petukhov公式,计算聚变堆包层模块第一壁不同冷却剂的传热性能。[结果]研究表明:氦气、氩气、氮气作为冷却剂,致使Be板和RAFM钢中温度场的变化趋势是相似的;温度场出现的最大温度均小于许用温度,符合温度的安全要求;氮气作为冷却剂可以实现的安全裕度是最大的,氩气次之,氦气实现的安全裕度相较偏小。[结论]可由此对聚变堆实验包层第一壁的冷却剂选择提供更多的优化可能,对聚变堆实验包层第一壁的安全增加更多的裕度。

影响上颌双根第一前磨牙颊根腭侧壁厚度相关因素的研究

目的:应用椎体束CT探讨在上颌双根第一前磨牙颊根腭侧凹陷的发生率、腭侧凹陷厚度的对称性,以及与性别、年龄及不同牙根部位的关系,为临床诊疗提供参考。方法:选取本院口腔科259例患者的锥形束CT(CBCT)扫描数据,按年龄分成4组,测量颊根管腭侧壁在不同牙根部位各壁厚度并分析结果。结果:腭侧凹陷的发生率为77.9%;左右两侧颊根腭侧壁厚度相似,无显著差异(P>0.05);随年龄增长不同牙根部位的腭侧壁厚度均增加且有显著差异(P<0.05),且>40岁时,各组牙根根管壁增龄性变化更明显(P<0.05);男性上颌第一前磨牙颊根腭侧壁厚度总体高于女性,有显著差异(P<0.05);随着距离根尖越近,男性与女性近颊根腭侧管壁厚度均减小,各组之间均有统计学差异(P<0.05)。男性腭侧壁最薄处为(0.58±0.11)mm,女性为(0.50±0.10)mm,二者有显著差异(P<0.05)。结论:上颌第一前磨牙颊根腭侧凹陷的发生率较高,根管治疗时应采取小锥度根管锉清理根管;40岁以上时腭侧壁增龄性变化更明显。

复杂地质条件密集小净距群桩施工技术

依托成都恒邦天府云玺项目,介绍高层建筑群桩基础在地质复杂、桩间距小、桩长≥30.0 m条件下,采用多次试桩来确定桩基施工工艺的过程。根据试桩结果,ϕ800mm桩采用“中长护筒跟进+泥浆护壁”方案,ϕ1200mm桩采用“全护筒护壁”方案,遇塌孔时安装钢护筒进入稳定土层≥2.0 m。经检测施工桩基均为Ⅰ类桩,单桩竖向抗压承载力满足设计和规范要求。

广州某超限高层住宅楼结构设计

某工程位于广州市黄埔区,为剪力墙结构超限高层住宅,高度为140.45m,主体结构为高度超限且存在凹凸不规则、扭转不规则等不规则项。为分析结构的抗震性能,通过进行合理的结构布置,采用多种软件进行结构分析,并针对结构细节进行专项设计,提出相应的构造加强措施。分析结果表明,结构各项指标均满足要求,能达到相应的性能目标要求,结构方案可行且安全;通过拉应力分析加强剪力墙边缘构件配筋,通过压弯分析保证一字型剪力墙承载力满足要求,通过顶板高差处理、全混凝土外墙设计保证地震力较好传递。

钨/低活化钢钎焊用铁基非晶钎料与接头微结构

为了连接W和CLF-1 RAFM钢,设计出由低活化元素组成的Fe-B-Si、Fe-B-Si-Sn、Fe-B-Si-Cr-(Sn)、Fe-B-Si-P-(Cr,Sn)、Fe-B-Si-Mn-(Ga,Sn)和Fe-B-Si-(Cr,Mn,Ga,Ta,Sn)系列Fe基非晶钎料,结合熔体快淬技术制备出非晶合金箔带,并对W/CLF-1 RAFM钢接头微结构进行了对比研究。采用X-射线衍射仪对箔带样品与焊缝进行了相鉴定;通过差热分析测量了非晶箔带的熔化温度和液相线温度;利用光学金相和电子探针分析了焊缝组织形貌和元素分布。结果表明,利用Fe-B-Si、Fe-B-Si-Cr和Fe-B-Si-Mn-Sn非晶钎料可获得结构完整的W/CLF-1钢接头;前两种钎料得到的焊缝组织基体相为α-Fe固溶体,而含Mn钎料形成的焊缝基体为马氏体组织;在高温钎焊过程中,这些Fe基非晶钎料中的高B含量促使FeWB、FeW2B2和Fe3B型金属间化合物在焊缝中形成,并有效地阻止了W元素向低活化钢基体长程扩散。所设计的低活化Fe基非晶钎料可用于W和低活化钢的连接和接头性能研究。

聚变堆包层第一壁缩比部件激光选区熔化成形研究

目的研究激光选区熔化(SLM)成形第一壁缩比结构的组织性能。方法以316L粉末为原材料,运用Inspire软件对不同成形姿势下第一壁缩比结构的应力与变形情况进行数值模拟,选择最佳成形姿势进行SLM成形,以控制整体变形,并对成形零件进行显微组织观察与力学性能测试。结果实验结果表明,与立放和侧放2种成形姿势相比,平放时残余应力与变形最小,最大残余应力为29.68 MPa,最大变形量为0.29 mm。成形件微观组织呈现各向异性,x-y方向主要为粗大的胞状晶组织,z-x方向为细长的柱状晶组织。力学测试结果显示,x-y方向的抗拉强度为672.1 MPa,伸长率为48.2%,冲击韧性为100.6 J/cm^(2);z-x方向的抗拉强度为646.9 MPa,伸长率64.4%,冲击韧性为136.3 J/cm^(2)。结论组织的差异性主要是由扫描工艺与熔池内部复杂的温度场引起的,微观结构的各向异性会造成力学性能的差异,x-y方向的强度高于z-x方向的,z-x方向上的塑性韧性更高。

EAST边缘等离子体中钨表面电弧诱导实验研究

在EAST装置中利用材料与等离子体测试平台MAPES对不同表面性质和结构的钨样品进行了边缘等离子体条件下的起弧诱导实验研究。结果显示只有纳米丝结构的钨表面成功产生了电弧,表明了钨表面的纳米丝结构更有利于电弧的触发,甚至在装置远离刮削层区域的地方也能诱导电弧触发。在纳米丝钨表面起弧过程中会产生两种不同的电弧,其中较强电弧能产生熔坑,并溅射出液滴,是等离子体中尘埃及杂质的重要来源。弧坑直径约为3μm,深度约为0.7μm,在电弧熔坑中心区域形成了一些小孔和纳米结构,其周围的熔化液滴也存在类似的结构。

CFETR超临界CO_(2)锂铅双冷包层第一壁热工水力学分析

基于CFETR超临界CO_(2)锂铅双冷包层基本结构,对超临界二氧化碳冷却第一壁方案进行了热工水力学分析。采用计算流体力学(CFD)的方法,通过限定第一壁最高温度,分析了流道间距、壁面粗糙度、流道尺寸和内壁面热流密度对驱动功率、出口温度和内壁面温度的影响,并与氦冷方案进行了比较,为后续包层结构设计提供参考。结果表明,减小流道间距、增大壁面粗糙度可以获得更好的冷却性能。增大流道极向尺寸可以有效降低驱动功率,而增大流道径向尺寸可以有效降低内壁面温度。考虑第一壁与增殖区换热后,内壁面最高温度会有较大提升。此外,超临界二氧化碳冷却第一壁所需的驱动功率小于氦气的驱动功率。