Fatigue Life Assessment Based on Crack Growth Behavior in Reduced Activation Ferritic/Martensitic Steel

Crack growth behavior under low cycle fatigue in reduced activation ferritic/martensitic steel, F82H IEA-heat (Fe-8Cr-2W-0.2V-0.02Ta), was investigated to improve the fatigue life assessment method of fusion reactor structural material. Low cycle fatigue test was carried out at room temperature in air at a total strain range of 0.4–1.5% using an hourglass-type miniature fatigue specimen. The relationship between the surface crack length and life fraction was described using one equation independent of the total strain range. Therefore, the fatigue life and residual life could be estimated using the surface crack length. Moreover, the microcrack initiation life could be estimated using the total strain range if there was a one-to-one correspondence between the total strain range and number of cycles to failure. The crack growth rate could be estimated using the total strain range and surface crack length by introducing the concept of the normalized crack growth rate.     

Effects of Fast Reactor Irradiation Conditions on Tensile and Transient Burst Properties of Ferritic/Martensitic Steel Claddings

The effects of fast neutron irradiation conditions have been investigated by focusing on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated at temperatures from 693 to 833K to 42.5 dpa (displacement per atom) in the experimental fast reactor JOYO using the PFB090 fuel test subassembly. Post-irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for the PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. On the other hand, the strength of the HT9M cladding tended to shift to lower values than those of as-received specimens. The differences in tensile and transient burst strengths between the two claddings were attributed to martensite structural stability which was related to the stable solid solution elements.     

聚变堆第一壁用纳米结构ODS钢的发展与前瞻

第一壁结构材料必须满足聚变堆极其严酷的工作环境要求,这是制约聚变堆发展的技术瓶颈之一。近年来发展起来的纳米结构氧化物弥散强化钢的特征性微观结构赋予了该材料优异的抗辐照性能,被视为第一壁结构材料的发展方向。本文简述了国内外第一壁材料的研究与发展概况。     

硅对9Cr-1.5WVTa低活化马氏体钢力学性能的影响

基于中国正在研究的聚变堆用9Cr-1.5WVTa低活化马氏体钢(CLAM钢),研究了添加合金元素硅对CLAM钢力学性能的影响。结果表明,添加0.2%Si使得CLAM钢的抗拉强度和屈服强度明显提高,钢的塑性和冲击韧性同时也得到一定提高,其中,韧脆转变温度(DBTT)由-13℃降至-30℃。未添加和添加0.29%Si的CLAM钢均为全马氏体组织,无δ铁素体存在。硅的添加使得9Cr-1.5WVTaSi钢的晶粒细化,从而提高了钢的拉伸和冲击性能。     

Technical Issues for the Fabrication of a CN-HCCB-TBM Based on RAFM Steel CLF-1

Reduced activation ferritic/martensitic steel (RAFM) is recognized as the primary candidate structural material for ITER’s test blanket module (TBM). To provide a material and property database for the design and fabrication of the Chinese helium cooled ceramic breeding TBM (CN HCCB TBM), a type of RAFM steel named CLF-1 was developed and characterized at the Southwestern Institute of Physics (SWIP), China. In this paper, the R&D status of CLF-1 steel and the technical issues in using CLF-1 steel to manufacture CN HCCB TBM were reviewed, including the steel manufacture and different welding technologies. Several kinds of property data have been obtained for its application to the design of the ITER TBM.     

氦中痕量H2、D2组分的高精度气相色谱分析

为满足聚变堆氘氚燃料循环工艺气体中痕量氢同位素组分的特殊检测分析要求,需建立快速、高精度的在线分析方法。针对氦中痕量H₂、D₂组分,本工作以高纯氦为载气,在液氮温度下,使用自制改性Al₂O₃毛细管柱进行分离,放电氦离子化检测器进行检测。结果显示,氢氘组分的检测限不高于1×10^-8,保留时间不高于180 s,氢氘组分色谱峰峰面积响应值的相对标准偏差不大于1.0%,分离度大于1.0。本方法具有分析灵敏、快速的优点,为聚变堆包层在氘氚燃料注入系统和氢同位素分离系统中微量氚的安全分析与精确计量提供了一种有效的测量技术。     

S22合金钢的氘扩散渗透行为

在分别采用光学显微镜和辉光放电光谱仪获得S22合金钢的金相显微组织和元素成分的基础上,利用已经过氘渗透速率定量标定的渗透实验装置,研究了S22合金钢在400~600℃下的氘扩散渗透行为,得到了氘在S22合金钢中的渗透率和扩散系数。对比分析了500℃时纯D2和He+1%D2下S22合金钢的氘渗透数据,结果表明,随氘分压的降低,氘在S22合金钢中的扩散渗透孕育期增加,氘渗透速率降低,但即使在较低氘分压下依然有氘渗透现象。该结果可为高温气冷堆、快堆等先进反应堆在使用S22合金钢时进行必要的氚渗透防护及其安全分析提供理论与数据支持。     

9Cr3W低活性马氏体钢微观组织及力学性能的初步研究

通过提高W含量,调整V、Ta、Ti、N等微合金元素含量,设计了9Cr3W型低活性马氏体钢。研究了该钢的微观组织结构与硬化、时效及相转变行为,对其进行了拉伸性能和冲击韧性测试。与Eurofer97钢相比,该钢表现出优良的高温拉伸性能。分析了9Cr3W钢用作超临界水堆堆芯内部件及包壳材料的可行性,其高温力学性能远优于Zr合金包壳材料;拉伸性能与T91钢相当,且韧脆转变温度低于T91钢,冲击吸收功上限高于T91钢,具有优良的冲击韧性;9Cr3W钢的高温瞬时强度低于奥氏体316不锈钢,成为制约其用于超临界水堆堆芯内部件及包壳的因素之一。     

Effects of Microstructural Evolution on Charpy Impact Properties of Modified Ferritic/Martensitic Steel after Neutron Irradiation

The effects of fast neutron irradiation have been investigated on Charpy impact properties and microstructural evolution of ferritic/martensitic (F/M) 11Cr-0.5Mo-2W, Nb, V stainless steel (PNC-FMS) which were developed by the Japan Atomic Energy Agency for core component applications in advanced fast reactors. One-half-size Charpy specimens were irradiated in the experimental fast reactor JOYO using the material irradiation rig to doses of 4.4 dpa at 773 K and 2.8 dpa at 923 K. Post irradiation Charpy impact tests and microstructural observations by transmission electron microscope were carried out. The results of the Charpy impact tests showed that there was no significant degradation in Charpy impact properties after neutron irradiation at 773 K to 4.4 dpa compared with the as-received specimens, but there was significant degradation in the properties after neutron irradiation at 923 K to 2.8 dpa. The Charpy impact property changes were attributed to matrix softening by recovery of martensite lath structures and to precipitate distributions.     

低活化铁素体/马氏体钢CLF-1氢氘的渗透行为

低活化铁素体/马氏体钢(RAFM钢)作为聚变堆结构材料中最有前景的候选材料,其氢同位素渗透行为备受关注。采用氢同位素气相驱动渗透的方法,对中国低活化铁素体/马氏体钢CLF-1的氢同位素渗透行为进行了研究,研究了温度、气体压强、样品表面状态等因素对其渗透行为的影响。结果表明:氢、氘在RAFM CLF-1钢中渗透扩散过程为体扩散控制,渗透率与温度的关系式均遵循Arrhenius关系;在实验测试过程中,由于样品表面发生氧化现象和缺陷捕获造成H₂和D₂渗透实验中渗透通量出现下降的现象。     

Characterization of High Temperature Creep Properties in Recrystallized 12Cr-ODS Ferritic Steel Claddings

The high temperature strengthening mechanism of previously manufactured 12Cr-ODS ferritic steel claddings was clarified. In the recrystallized 12Cr-2W-0.3Ti-0.24Y₂O₃-ODS ferritic steel cladding, αY₂TiO₅ type complex oxide formation was responsible for the drastic reduction of oxide particle size and the resulting shortened distance between particles, which led to superior internal creep rupture strength at 973 K because of the high resistance to gliding dislocation. Internal creep deformation was considered to be controlled by the grain boundary sliding associated with grain morphology: the near Σ11, Σ and Σ19 coincidence boundaries with a (110) common axis.     

注氘低活化马氏体钢在电子辐照下的缺陷行为

低活化铁素体/马氏体(RAFM)钢被视为国际热核聚变反应堆以及聚变反应堆的第1壁候选结构材料之一,很多国家均在研究不同的RAFM钢,中国低活化马氏体(CLAM)钢的研究亦正在进行。核聚变会产生氢、氦、氘及氚,这些气体元素与辐照缺陷结合在一起,对材料的辐照性能会产生较大影响。本文对注氘后不同温度下的辐照后微观结构进行研究。试验利用日本北海道大学的JEOL-1300高压电子显微镜研究注氘CLAM钢从室温到873K在1250keV电子辐照下的微观结构变化。研究结果表明,在电子辐照下,注氘产生的缺陷团会出现消失和长大两种现象,意味着间隙型与空位型位错环在注氘过程中同时产生。并研究了注氘产生的空洞。     

Underwater explosive welding of tungsten to reduced-activation ferritic steel F82H

The present study reports the underwater explosive welding of commercially pure tungsten onto the surface of a reduced-activation ferritic steel F82H plate. Cross-sectional observation revealed the formation of a wave-like interface, consisting of a thin mixed layer of W and F82H. The results of nanoindentation hardness testing identified a gradual progressive change in the interface, with no hardened or brittle layer being observed. Small punch tests on the welded specimens resulted in cracking at the center of the tungsten, followed by crack propagation toward both the tungsten surface and the tungsten/steel interface.     

IVR中熔融堆芯被牺牲性材料稀释后的传热计算

应用牺牲性材料的堆芯熔融物稀释方案是先进轻水堆中一种新型的严重事故缓解措施。严重事故发生时,掉落的堆芯熔融物被氧化物牺牲性材料（OSM）稀释,导致熔池结构发生翻转,因此计算翻转后熔融池的传热行为是进行牺牲性材料筛选和评价稀释方案可行性的重要研究内容。本文计算了容器内滞留（IVR）中熔融堆芯被Fe₃O₄、TiO₂和Al₂O₃3种候选OSM稀释后压力容器壁面的热流密度分布。研究发现,布置OSM后,上腔室结构在强烈热辐射的作用下会熔化掉落。随着OSM布置量的增大,压力容器壁面最大热流密度减小,当布置15 m³的OSM时,压力容器伸长约2 m,此时壁面最大热流密度较未布置时减小约45%,且当布置相同体积的OSM时,Fe₃O₄导致的壁面最大热流密度减小最多。此外,UO₂-ZrO₂-OSM三元混合物的熔点高低会对氧化物层表面是否结壳产生影响,从而影响壁面最大热流密度。     

中国抗中子辐照钢的抗辐照设计与验证

聚变堆等未来先进核能系统要求材料在强流高能中子辐照下长期保持良好的结构稳定性和机械性能。为适应未来先进核能技术发展的需要,中国科学院核能安全技术研究所•凤麟团队牵头研发了具有我国自主知识产权的中国抗中子辐照钢--CLAM钢。CLAM钢的设计考虑了未来核能清洁性的要求,以及苛刻服役环境中材料抗辐照、耐高温、耐腐蚀等性能要求。通过中子学计算分析设计了低活化成分范围,基于选择性纳米相析出进行了抗辐照、耐高温性能优化设计。针对材料的抗辐照性能,利用国内外中子、离子、电子及等离子体辐照设施开展了系列辐照考验研究,通过多角度表征辐照前后材料的微观结构和宏观性能,综合评估了材料的辐照性能,并与国际上同类材料在相近或相同条件下的辐照性能进行了对比分析,结果表明CLAM钢具有良好的抗辐照性能。     

P92钢高温低周疲劳的实验研究

由于高的热效率和简单的系统组成,超临界水堆(SCWR)被认为是第四代核反应堆的一种选择。超临界水堆的关键问题之一是核心部件尤其是燃料组件包壳的材料。这些材料在高温下的力学性能、腐蚀和应力腐蚀开裂敏感性以及抗辐射性能等对核电厂的安全运行至关重要。本文对SCWR包壳候选材料的F/M类材料P92钢进行了高温低周疲劳实验研究。实验温度为600和650℃,控制方式为总应变控制,应变范围均为±0.2%～±0.6%。实验结果表明,在两种温度下,P92钢均为循环软化材料,但未出现循环稳定现象。由于温度升高,塑性增强,P92钢在650℃下的宏观裂纹出现周次比率随应变范围的增加,下降比较平缓,且650℃下的失效寿命显著高于600℃下的失效寿命。并得到了两种温度下的稳定循环应力-塑性应变的关系以及循环失效寿命和应变的关系。     

硅对低活化马氏体钢电子辐照行为的影响

利用超高压透射电子显微镜研究了两种成分的低活化马氏体钢(CLAM钢)的辐照损伤行为。结果表明：电子辐照能在未添加硅的CLAM钢中产生辐照空洞;在450℃下辐照至14dpa时,空洞数密度约为8.7×10²¹m^－3,辐照肿胀率约为0.26％;在450℃下的辐照肿胀率明显比500℃下的高;当损伤率为2×10^－3dpa/s时,添加合金元素硅能显著提高CLAM钢的抗辐照肿胀能力,未在添加硅的CLAM钢中实验观察到辐照空洞的形成。在450℃下进行辐照时,添加硅的CLAM钢出现明显的辐照共格析出现象。     

硅对CLAM钢微观结构和力学性能的影响

文章研究了合金元素硅对CLAM钢的力学性能和微观结构的影响。实验表明,在CLAM钢中添加0.2％的Si会使其屈服强度和抗拉强度明显提高,同时使CLAM钢的韧脆转变温度升高,韧性降低。在光学显微镜下观察,CLAM钢呈现回火马氏体组织,Si的添加使奥氏体晶粒细化,致使CLAM钢的强度升高、韧性降低。在透射电子显微镜下观察,回火板条马氏体是其主要结构,存在针状和颗粒状两种形态的碳化物;利用透射电子显微镜选区衍射方法可知,呈颗粒状的碳化物是M23C6型碳化物。Si的添加使马氏体板条的宽度变小。