快堆用HT-9研究进展

HT-9是含Cr量12%的铁素体/马氏体钢,是液态金属冷却快堆燃料包壳及外套管优异的候选材料。介绍了HT-9的研究进展,对HT-9在液态金属冷却快堆环境所需关键应用性能进行归纳和评述,对HT-9应用于铅冷快堆存在的主要问题进行了分析,给出了未来HT-9合金的主要发展趋势。     

俄罗斯无机材料研究院关于快堆燃料包壳材料的研究

俄罗斯无机材料研究院(ВНИИНМ)是材料学研究和核燃料循环工艺、裂变核材料处理工艺等领域的著名研究机构,在快堆堆芯结构材料方面该院借助于俄罗斯丰富的钠冷快堆运行和材料学研究经验,以BOR-60和BN-600为研究试验平台,以提高BN-600和BN-800性能及开发更加先进的BN-1200为目标,开展了大量燃料棒包壳及燃料组件外套管材料的研究.本文是对ВНИИНМ近几年研究成果在俄罗斯科学杂志和研讨会上发表报告的调研、翻译和汇总,供我国有关钠冷快堆技术研究和工程设计人员参考.     

聚变堆低活化马氏体钢的发展

介绍了国际聚变堆低活化结构材料发展概况及趋势,以及国内发展自己特有的低活化马氏体钢的必要性。介绍了聚变堆结构材料——低活化铁素体/马氏体钢发展的必要性及迫切性,以及目前国际上包括欧洲、日本、美国等在此方面研究的进展概况及发展趋势,最后提出了国内发展自己特有的低活化马氏体钢——CLAM钢的必要性,并对目前的研究进展情况做了介绍。     

中国实验快堆结构材料辐照装置设计

为在中国实验快堆(CEFR)上开展国产快堆包壳材料的辐照试验,进行了CEFR首个结构材料辐照装置的设计。材料辐照装置的创新设计基于CEFR的辐照条件和堆芯组件的基本结构,通过在辐照装置内部设置不同气隙尺寸的辐照罐,实现了在快堆不同功率稳态运行条件下(40%和100%额定功率)对材料样品不同辐照温度(450~600℃)的要求。辐照装置具有样品辐照温度与中子注量率的非在线监测功能,其结构具有通用性,能满足材料辐照标准试样最大装载的需要。通过对辐照装置进行热工分析和堆外的传热验证试验、流阻特性和结构稳定性验证试验,保证了辐照装置的设计能满足材料辐照任务的要求。     

先进反应堆用ODS F/M钢的强韧性匹配研究进展

氧化物弥散强化（ODS）铁素体/马氏体（F/M）钢具有极高的缺陷阱密度,加之体心立方结构基体,使其具有优异的抗辐照性能,被确定为包括聚变堆和第4代裂变堆在内的先进核能系统关键部件候选材料,成为核材料领域的研究热点。有利于ODS钢抗辐照性能的显微组织特点同时也赋予了ODS钢优异的室温和高温强度。但和其他高强度材料类似,ODS钢也存在强度高,而塑韧性不足的矛盾,不利于复杂部件的加工,因此,实现ODS F/M钢的高强高韧成为面向工程应用的ODS F/M钢研究的一大热点和难点。目前,针对ODS F/M钢强韧化的研究还较少,已有的相关研究也不够系统和深入,本文主要对抗辐照ODS F/M钢的显微组织结构要求及其强韧化研究现状进行总结和分析,为先进反应堆用抗辐照ODS F/M钢的强韧化设计提供思路和参考。     

快堆燃料组件六角形外套管截面变形的计算方法研究

快堆燃料组件六角形外套管截面在冷却剂压差的作用下会产生弹性变形,在温度场和中子场的作用下会产生热膨胀、辐照肿胀和辐照蠕变,这些变形关系到多根组件之间接触状态的正确判断,是整个燃料组件变形计算方法研究的重要基础性工作。本文目的在于探讨确定考虑温度和辐照因素影响的燃料组件六角形截面的变形计算方法。本文最终得到了变形计算公式,其中弹性变形计算公式与有限元结果进行比较验证后结果吻合。本文的研究成果为后续研究提供了依据和基础,具有重要意义。     

国产快堆包壳管材料——冷加工316（Ti）不锈钢高温强度下降的微观机理分析

国产快堆包壳管材料316（Ti）不锈钢（1995—1996年，由原上海第五钢厂研制）的力学性能测试结果表明：材料拉伸强度与国外材料相当，但高温蠕变和高温持久性能却低于国外的数据。将这种材料与俄罗斯快堆包壳材料CH-68比较，国产材料在室温和600℃下的屈服极限强度高于俄罗斯材料的，但625和700℃的高温持久强度却低于俄罗斯CH-68数据，尤其是在700℃下，国产材料的持久断裂强度不但大大低于俄罗斯材料，且强度下降很快。     

中子注量探测器在快堆结构材料辐照损伤的应用研究

为评估快堆结构材料的辐照损伤,本文提出了一套快堆结构材料辐照损伤评价方法。根据快堆能谱特点设计中子注量探测器辐照方案,分析探测片特性和反应道截面,选取7种快中子注量探测器。同时采用迭代法在Labview平台中开发了解谱程序。基于俄罗斯碳化硼组件辐照实验数据进行解谱,并结合Lindhard-Robinson模型组件包壳原子平均离位（dpa）计算,同时与SPECTER 计算值进行对比。结果表明,本文采取的实验方法得到的dpa与SPECTER计算值偏差在6%以内,符合较好。本文建立了一套完善的快堆结构材料辐照损伤评价体系,对结构材料的辐照损伤监测具有重要意义。     

铁素体/马氏体钢和奥氏体不锈钢的液态铅铋腐蚀行为与机理

液态铅铋共晶合金[liquid lead-bismuth eutectic,LBE,Pb44.5Bi55.5,%(质量分数)]具有优异的热工水力和中子学性能,是第四代液态金属冷却快堆最重要的冷却工质之一。但是,液态铅铋冷却快堆的主要候选材料包括铁素体/马氏体钢(如T91)和奥氏体不锈钢(如316L和15-15Ti)存在液态金属腐蚀问题,一定程度上阻碍了液态铅铋快堆工程化应用进度。本文综述了液态铅铋腐蚀的基本机制以及铁素体/马氏体钢和奥氏体不锈钢的液态铅铋腐蚀行为,总结了抑制液态铅铋腐蚀的主要方法,并展望了未来研究方向。     

Long-term properties of reduced activation ferritic/martensitic steels for fusion reactor blanket system

Thermal aging properties of reduced activation ferritic/martensitic steel F82H was researched. The aging was performed at temperature ranging from 400 °C to 650 °C up to 100,000 h. Microstructure, precipitates, tensile properties, and Charpy impact properties were carried out on aged materials. Laves phase was found at temperatures between 550 and 650 °C and M₆C type carbides were found at the temperatures between 500 and 600 °C over 10,000 h. These precipitates caused degradation in toughness, especially at temperatures ranging from 550 °C to 650 °C. Tensile properties do not have serious aging effect, except for 650 °C, which caused large softening even after 10,000 h. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in ductile-to-brittle transition temperature was observed in the 650 °C aging. It was caused by the large Laves phase precipitation at grain boundary. Laves precipitates at grain boundary also degrades the upper-shelf energy of the aged materials. These aging test results indicate F82H can be used up to 30,000 h at 550 °C.     

Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He⁺ beam and sequential He⁺ and H⁺ beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C.     

Development and characterization of advanced 9Cr ferritic/martensitic steels for fission and fusion reactors

This paper presents the results on the physical metallurgy studies in 9Cr Oxide Dispersion Strengthened (ODS) and Reduced Activation Ferritic/Martensitic (RAFM) steels. Yttria strengthened ODS alloy was synthesized through several stages, like mechanical milling of alloy powders and yttria, canning and consolidation by hot extrusion. During characterization of the ODS alloy, it was observed that yttria particles possessed an affinity for Ti, a small amount of which was also helpful in refining the dispersoid particles containing mixed Y and Ti oxides. The particle size and their distribution in the ferrite matrix, were studied using Analytical and High Resolution Electron Microscopy at various stages. The results showed a distribution of Y₂O₃ particles predominantly in the size range of 5-20 nm. A Reduced Activation Ferritic/Martensitic steel has also been developed with the replacement of Mo and Nb by W and Ta with strict control on the tramp and trace elements (Mo, Nb, B, Cu, Ni, Al, Co, Ti). The transformation temperatures (A_c1, A_c3 and M_s) for this steel have been determined and the transformation behavior of the high temperature austenite phase has been studied. The complete phase domain diagram has been generated which is required for optimization of the processing and fabrication schedules for the steel.     

Grain boundary engineering for enhancing intergranular damage resistance of ferritic/martensitic steel P92

Ferritic/martensitic(F/M)steel is widely used as a structural material in thermal and nuclear power plants.However,it is susceptible to intergranular damage,which is a critical issue,under service conditions.In this study,to improve the resist-ance to intergranular damage of F/M steel,a thermomechanical process(TMP)was employed to achieve a grain boundary engineering(GBE)microstructure in F/M steel P92.The TMP,including cold-rolling thickness reduction of 6％,9％,and 12％,followed by austenitization at 1323 K for 40 min and tempering at 1053 K for 45 min,was applied to the as-received(AR)P92 steel.The prior austenite grain(PAG)size,prior austenite grain boundary character distribution(GBCD),and connectivity of prior austenite grain boundaries(PAGBs)were investigated.Compared to the AR specimen,the PAG size did not change significantly.The fraction of coincident site lattice boundaries(CSLBs,3 ≤ Σ ≤ 29)and Σ3n boundaries along PAGBs decreased with increasing reduction ratio because the recrystallization fraction increased with increasing reduction ratio.The PAGB connectivity of the 6％deformed specimen slightly deteriorated compared with that of the AR specimen.Moreover,potentiodynamic polarization studies revealed that the intergranular damage resistance of the studied steel could be improved by increasing the fraction of CSLBs along the PAGBs,indicating that the TMP,which involves low deforma-tion,could enhance the intergranular damage resistance.     

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.     

Beryllium/ferritic martensitic steel joining for the fabrication of the ITER test blanket module first wall

The first wall of an international thermonuclear experimental reactor (ITER) test blanket module (TBM) is a multilayered component consisting of plasma facing armor and structural materials including the cooling channels. One of the main issues about the R&D on the TBM is to develop the joining technologies for a fabrication of the TBM first wall. The objectives of this study are to optimize the hot isostatic pressing (HIP) conditions and the interlayer combination for the fabrication of beryllium (Be)/ferritic martensitic steel (FMS) joints without a degradation of the mechanical properties of the FMS. Effects of HIP joining conditions including the temperature and interlayer types were investigated. The HIP temperature was selected for the anticipated tempering condition for FMS to avoid a grain coarsening which would deteriorate the mechanical properties of FMS. Several interlayer materials were applied in order to manufacture high strength joints. Be and FMS were joined successfully by the application of a Ti/Cu interlayer and it showed a relatively high bending strength, 257 MPa, among the interlayer types studied. The fracture was caused by a delamination of the reaction layer between FMS and the coated interlayer without a plastic deformation. This paper summarizes the results of a Be/FMS joints manufacturing and an investigation of their properties.     

Prevention and mitigation of severe accident developments and recriticalities in advanced fast reactor systems

GEN-IV nuclear systems, especially advanced sodium-cooled fast reactors (SFRs) are on the horizon and a key issue of their success is the promise of a higher and improved safety level. In Europe safety investigations are currently under way e.g. in the collaborative CP-ESFR project of the EU. Both on the prevention and mitigation side significant efforts are invested to fulfill the high safety goals. One route of assurance concentrates on the mitigation or even elimination of specific severe accident routes leading to core disruption and recriticalities. The accident phase with larger disrupted and molten fuel regions is coined the transition phase. A competition between fuel losses and in-pool material motion exists deciding over recriticalities and energetics potentials in this phase. To get a control of the transition phase recriticalities and energetics, ideas have been developed to install dedicated means in the core that enhance and guarantee a sufficient and timely fuel discharge - a controlled material relocation (CMR). Several proposals are under way to accomplish this CMR and especially in Japan significant theoretical and experimental work has been performed. In Europe the path to develop CMR measures was driven in the past by the development of the CAPRA reactors with a high Pu enrichment. In the current paper the status of analyses is described and some new concepts are discussed. The CMR measures are discussed along two accident scenarios, the unprotected loss of flow (ULOF) and the instantaneous blockage accident (TIB).     

Fracture mechanics concepts for evaluation of irradiated reactor pressure vessel material behavior

Within the German research program Forschungsvorhaben Komponentensicherheit (FKS), irradiation experiments were performed with ferritic reactor pressure vessel (RPV) steels and welds. The materials cover a wide range of chemical composition and initial toughness to achieve different susceptibility to neutron irradiation. Different neutron flux was applied and the neutron exposure extended up to 8×10¹⁹ cm⁻². The change in material properties was determined by means of tensile, Charpy impact, drop-weight and fracture mechanics tests, including crack arrest. The results have provided more insight into the acting embrittlement mechanisms and shown that the fracture mechanics concept of the Code provides in general an upper bound for the material which can be applied in the safety analysis of the RPV.     

Design and preliminary test of an annular linear induction electromagnetic pump for a sodium-cooled fast reactor thermal hydraulic experiment

An annular linear induction electromagnetic pump (ALIP) with a flow rate of 2265 L/min and a developed pressure of 4 bar was designed and fabricated to test the performance of the components of a sodium-cooled fast reactor (SFR) in a sodium thermal hydraulic experimental loop. The design characteristic of the ALIP was calculated using the electrical equivalent circuit method typically used for analyzing linear induction machines. Preliminary tests, such as verification of the moving function using an annular Al pipe, were carried out. The linearity between the input voltage, current, and magnetic flux density was verified. The developed force demonstrated an increase proportional to the square of the input current, whereas the velocity was linearly proportional to the input current. The main design variables of the pump were calculated theoretically for the SFR thermal hydraulic experimental loop. The pump was optimized for the design variables including input frequency, and the characteristics of the optimized pump were compared with those of the pump at the commercially used frequency of 60 Hz.