星球表面用核反应堆临界安全分析

核反应堆电源具有寿命长、可全天候工作等特点，可作为星球表面及其他深空探测任务的电源。针对星球表面用核反应堆电源在发射过程中重返地面的临界安全问题，提出了星球表面用核反应堆的临界安全分析要求、分析假设与模型，并对反应堆临界安全特性及采取的临界安全措施进行了计算分析。计算结果表明，不同假设掉落环境下的星球表面用核反应堆的有效增殖因数均小于0.98，满足临界安全要求。反应堆通过采用Mo-14%Re合金结构材料、设置相对较厚的堆芯反射层以及在反射层包壳和堆芯外围涂覆Gd₂O₃涂层等措施有利于确保反应堆在事故时处于次临界状态。     

亚化学计量U02-x燃料芯块的制备机理

亚化学计量U02-x芯块是一种设计新颖的特殊核反应堆用核燃料,很难采用传统压水堆超化学计量U02+x芯块工艺进行制造.本工作采用U02+x+U混合粉末为原料制备了U02-x芯块,研究了铀粉表面包覆处理方法、铀粉含量、成型压力、烧结气氛等工艺参数对芯块O/U比、烧结密度和微观结构的影响,探讨了U02-x环形芯块的亚化学计量形成机理.研究表明,当铀粉加人量(质量分数)分别为0,3%,6%时,芯块O/U比分别为2.010,1.991,1.982,平均晶粒尺寸分别为10,15,20μm;当铀粉加人量为50%时,O/U比为1.943,样品发生熔化.亚化学计量UO2-x芯块必须在干燥惰性气氛中密封保存.     

钼铼合金在空间核电源中的应用性能研究进展

钼（Mo）中加入铼（Re）可显著改善钼的低温脆性进而提高其加工性能及焊接性能,提高强度的同时仍保持良好的塑性。Re元素含量为14%左右时,Mo-Re合金延伸率接近40%,加工性能最好,而同时存在一定的Re元素固溶强化作用。在1550 K以下温度,Mo-Re合金与UO₂的相容性较好。在1 300 K以下时,Mo-Re合金与UN的相容性较好。在1800 K以下时,Mo-Re合金与碱金属Li、Na、K的相容性均较好。钼铼合金与核燃料及碱金属冷却剂均具有良好的相容性,且Re元素是一种较好的谱移吸收体材料,可有效降低反应堆临界事故风险。钼铼合金是空间核电源中最佳反应堆芯结构材料。本文对钼铼合金的研究状况进行总结,为国内相关空间核反应堆电源系统设计选材和研究提供参考。     

钼铼合金在空间核电源中的应用性能研究进展

钼(Mo)中加入铼(Re)可显著改善钼的低温脆性进而提高其加工性能及焊接性能,提高强度的同时仍保持良好的塑性。Re元素含量为14%左右时,Mo-Re合金延伸率接近40%,加工性能最好,而同时存在一定的Re元素固溶强化作用。在1 550 K以下温度,Mo-Re合金与UO_2的相容性较好。在1 300 K以下时,Mo-Re合金与UN的相容性较好。在1 800 K以下时,Mo-Re合金与碱金属Li、Na、K的相容性均较好。钼铼合金与核燃料及碱金属冷却剂均具有良好的相容性,且Re元素是一种较好的谱移吸收体材料,可有效降低反应堆临界事故风险。钼铼合金是空间核电源中最佳反应堆芯结构材料。本文对钼铼合金的研究状况进行总结,为国内相关空间核反应堆电源系统设计选材和研究提供参考。     

亚化学计量UO2-x燃料芯块的制备机理

亚化学计量UO_2-x芯块是一种设计新颖的特殊核反应堆用核燃料,很难采用传统压水堆超化学计量UO_2+x+U芯块工艺进行制造。本工作采用UO_2+x+U混合粉末为原料制备了UO_2-x芯块,研究了铀粉表面包覆处理方法、铀粉含量、成型压力、烧结气氛等工艺参数对芯块O/U比、烧结密度和微观结构的影响,探讨了UO_2-x环形芯块的亚化学计量形成机理。研究表明,当铀粉加入量（质量分数）分别为0、3%、6%时,芯块O/U比分别为2.010、1.991、1.982,平均晶粒尺寸分别为10、15、20μm;当铀粉加入量为50%时,O/U比为1.943,样品发生熔化。亚化学计量UO_2-x芯块必须在干燥惰性气氛中密封保存。     

俄继续开展耐事故燃料测试

正【英国《国际核工程》网站2021年5月4日报道】俄罗斯博奇瓦尔无机材料研究所(VNIINM)2021年4月29日宣布计划在年底完成耐事故燃料试验组件的第三个辐照周期测试。俄2019年1月在核反应堆研究所(RIAR)MIR研究堆中启动对首批两个耐事故燃料试验组件的辐照测试。两个组件由新西伯利亚化学浓缩厂(NCCP)制造,含有2种燃料芯块和2种包壳:燃料芯块分别是传统二氧化铀芯块和具有更高铀密度和导热性的铀钼合金芯块;包壳分别是带铬涂层的锆合金包壳和铬镍合金包壳。     

细棒状铀-氢化锆燃料芯块制备与显微结构初步研究

采用熔炼铀-锆合金然后渗氢的工艺制造细棒状铀-氢化锆燃料芯块,通过改变熔炼工艺参数提高铀-锆合金的铸造质量和成品率.在渗氢的工艺中,采用不同工艺对铀-锆合金进行氢化,氢化后芯块中的氢/锆原子比在1.41～1.72之间,而芯块的尺寸相对于氢化前也有不同程度的增加.微观分析表明,氢化后得到的燃料芯块中含有多种相结构,其中金...     

俄耐事故燃料即将入堆测试

正【英国《国际核工程》网站2019年1月2日报道】俄罗斯核燃料产供集团(TVEL)旗下新西伯利亚化学浓缩厂(NCCP) 2018年12月27日宣布,已制造出适用于包括VVER在内的压水堆的耐事故燃料试验组件。试验组件中含有2种燃料芯块和2种包壳:燃料芯块分别是传统二氧化铀芯块和具有更高铀密度和导热性的铀钼合金芯块;包壳分别是带铬涂层的锆合金包壳和铬镍合金包壳。这些芯块和包壳组成了4种燃料棒。     

热管空间核反应堆电源的研究进展

随着人类对宇宙太空的深入探索,对于提供能量的电源要求也在逐步提高,空间核反应堆电源在执行深空探测任务中脱颖而出。热管核反应堆由于具备非能动性、寿期长、可靠性高等优势,成为目前空间核反应堆领域的研究热点。本文通过重点介绍典型热管堆的概念设计系统以及在Kilopower中的应用,对热管电源系统(Heatpipe Power System,HPS)、由热管控制的火星探索反应堆(The Heatpipe-operated Mars Exploration Reactor,HOMER)、安全可负担裂变引擎方案(Safe Affordable Fission Engine,SAFE)以及Kilopower进行了重点调研,归纳总结了各个反应堆的结构设计、燃料选择、热管排布、功率设计等,以期对未来热管空间核反应堆电源的设计研究提供思路和参考。     

核燃料棒芯块235U丰度检测方法研究

核燃料棒中UO₂芯块的²³⁵U丰度检测是保证核反应堆正常运行的重要环节,根据铀样品能谱谱形,通过迭代拟合算法精确选取目标信号的能量范围,减少了测量精度受UO₂芯块年龄的影响,扩大了目标信号能量选取范围,基于小波变换法,过滤无关频率信号,提高了突变信号的识别精度,进一步提高了检测速度。通过模拟存在异常丰度芯块燃料棒检测验证了方法的可行性。     

Nuclear power development in market conditions with use of multi-purpose modular fast reactors SVBR-75/100

This article presents an innovative nuclear power technology, based on the use of modular type fast-neutron reactors SVBR-75/100 having heavy liquid-metal coolant, i.e. eutectic lead–bismuth alloy, which was mastered in Russia for the nuclear submarines’ reactors. Reactor SVBR-75/100 possesses inherent self-protection and passive safety properties that allow excluding of many safety systems necessary for traditional type reactors. Use of this nuclear power technology makes it possible to eliminate conflicting requirements among safety needs and economic factors, which is particularly found in traditional reactors, to increase considerably the investment attractiveness of nuclear power based on the use of fast-neutron reactors for the near future, when the cost of natural uranium is low and to assure development of nuclear power in market conditions. On the basis of the factory-fabricated “standard” reactor modules, it is possible to construct the nuclear power plants with different power and purposes. Without changing the design, it is possible for reactor SVBR-75/100 to use different kinds of fuel and operate in different fuel cycles with meeting the safety requirements.     

Radiation Characteristics of Fuel and Wastes in the Uranium–Plutonium and Thorium–Uranium Fuel Cycle

The radiation characteristics of fuel cycles of various reactors – replacement candidates in the future nuclear power – are compared. Proceeding from the basic requirements (safety, fuel supply, and nonproliferation of fissioning materials), inherently safe fast reactors of the BREST type can be used as the basis for large-scale nuclear power. Thermal reactors, which can burn enriched uranium, thorium–uranium fuel, or mixed uranium–plutonium fuel with makeup with fissioning materials from fast reactors, will operate for a long time simultaneously with fast reactors in the future nuclear power. VVÉR-1000 and CANDU reactors are examined as representatives of thermal reactors; for each of these reactors the operation in various variants of the fuel cycle is simulated. It is shown that with respect to radiation characteristics of the fuel and wastes the thorium–uranium fuel cycle has no great advantages over the uranium–plutonium cycle.     

Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages – possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools.     

锆合金耐腐蚀性能研究综述

锆合金主要用作核反应堆燃料元件的包壳材料及其他堆内构件。回顾了有关锆合金水侧腐蚀的主要研究结果及存在的问题，概括了现有的理论及面临的挑战。８０年代，关于锆合金化学成分、微观结构及辐照对耐腐蚀性能影响的研究取得了很大进展。近几年来的研究工作主要集中在探索腐蚀机理、选择最佳合金成分及控制微观结构方面，以满足提高燃耗、降低核电成本后对锆合金提出的更高要求。     

Nuclear power technologies at the stage of sustainable nuclear power development

It is not simple to solve the problem of competitiveness of nuclear power technologies in evolutionary upgrading the conventional nuclear power plants (NPP) such as light water reactors (LWR), which requires high expenditure for safety. Moreover, the existing LWRs cannot provide nuclear power (NP) for a long time (hundreds of years) because the efficiency of use of natural uranium is low and closing the nuclear fuel cycle (NFC) for those reactors is not expedient.The highlighted problem can be solved in the way of use of innovative nuclear power technology in which natural uranium power potential is used effectively and the intrinsic conflict between economic and safety requirements has been essentially mitigated.The technology that is most available and practically demonstrated is the use of reactors SVBR-100 — small power multi-purpose modular fast reactors (100 MWe) cooled by lead-bismuth coolant (LBC). This technology has been mastered for nuclear submarines’ reactors in Russia.High technical and economical parameters of the NPP based on RF SVBR-100 are determined from the fact that the potential energy stored in LBC per a volume unit is the lowest.The compactness of the reactor facility SVBR-100 that results from integral arrangement of the primary circuit equipment allows realizing renovation of power-units LWRs, the vessels’ lifetime of which has been expired. So due to this fact, high economical efficiency can be obtained.The paper also validates the economical advantage of launching the uranium-fueled fast reactors with further changeover to the closed NFC with use of plutonium extracted from the own spent nuclear fuel in comparison with launching fast reactors directly with on uranium-plutonium fuel on the basis of plutonium extraction from spent nuclear fuel of LWRs.     

Mössbauer spectroscopy of tin in unirradiated and neutron irradiated Zircaloys

Mössbauer spectroscopy of the 23.9 keV γ-rays in ¹¹⁹Sn nuclei was applied to study Zircaloy-2, Zircaloy-4, and other tin-bearing zirconium-based alloys of interest to nuclear power technology. Zircaloys are extensively used in nuclear reactors as fuel cladding. In CANDU reactors, Zircaloys are also used as major structural components such as calandria tubes, and were used until the late 1970's as pressure tubes (now replaced by Zr–2.5Nb alloy). Unirradiated specimens of these alloys, as well as radioactive specimens, both neutron-irradiated in high-flux test reactors and extracted from nuclear power-reactor components after many years of service, were examined. The obtained spectra consistently showed tin in substitutional solid solution in α-Zr, whereas no evidence was found of metallic Sn or intermetallic Zr₄Sn precipitates. In oxide scrapes removed from Zircaloy-2 pressure tube of one of CANDU reactors, where the alloy was exposed for about 10 years to pressurized heavy water coolant at temperatures of ∼280°C, a considerable fraction of tin was found in the Sn(IV) state, in the form that coincides with the state of tin in stannic oxide, SnO₂. The same form of tin was identified in filterable deposits in the primary heavy water coolant of CANDU reactors. For comparison, in Zircaloy heated in air, SnO₂ was formed only at temperatures above 500°C.     

Properties of DU-10 wt% Mo alloys subjected to various post-rolling heat treatments

Studies were completed to obtain mechanical properties of depleted uranium-molybdenum (U-Mo) alloys subjected to different post-processing treatments using microhardness, quasi-static tensile tests, and scanning electron microscopy failure analysis. U-Mo alloy foils are currently under investigation for potential fuel conversion of high power research reactors to low enriched uranium fuel. Although mechanical properties take on a secondary effect during irradiation, an understanding of the alloy behavior during fabrication and the effects of irradiation on the integrity of the fuel are essential. In general, the microhardness was insensitive to annealing temperature but decreased with annealing duration. Yield strength, Young's modulus, and ultimate tensile strength were affected in varying manners with both increasing annealing temperature and duration, and subjecting the alloy to rolling. The failure mode was insensitive to annealing conditions, but was significantly controlled by the impurity concentration of the alloy, especially carbon. Values obtained from literature are also provided with reasonable agreement based on extrapolation of annealing duration, even though processing conditions and applications were quite different in some instances.     

Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

The results of radiation tests are discussed and the character of the failure of fuel compositions and the operability of fuel elements is analyzed as a function of the type of fuel and the irradiation conditions. The intense interaction of the fuel with the matrix material is considered as the main factor limiting the operability of fuel elements in power-dense high-flux nuclear reasearch reactors. It is concluded that low-enrichment high-density uranium—molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium—molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 35–44, January, 2005.     

Trends in computational intelligence in nuclear engineering

Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.