大型压水堆装载50%MOX燃料方案压力容器辐照安全计算

对含MOX燃料堆芯的压力容器(RPV)快中子注量率计算进行了初步研究,探讨了适用于MOX堆芯方案屏蔽计算的堆芯源项处理方法。采用三维离散纵标程序TORT,针对CAP1400型堆芯装载50%MOX燃料方案开展了RPV快中子注量计算,结果表明:堆芯装载50%MOX燃料可满足RPV屏蔽安全设计要求;对比分析含MOX堆芯方案和全UO2堆芯方案的RPV快中子注量率的特性差异,从RPV辐射防护最优化的角度,后续燃料管理方案优化时可重点关注关键位置处组件的布置。     

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建立了6Li D转换器中14Me V中子源强的计算模型,对转换器不带辐照样品和分别带2、3、4个辐照样品时的中子源强进行了计算,对转换器产生的中子和来自于堆芯的中子在样品内的能谱和中子注量率进行了计算。结果表明,辐照管内充水和氦气时,辐照样品内由转换器产生的能量大于13Me V的中子分别占能量在1Me V以上中子的25.7%、24.6%,辐照样品内由堆芯产生的能量大于13Me V的中子仅占能量在1Me V以上中子的10·5左右,样品内14Me V中子源强分别可达4.31×1013nT·s·1、3.34×1013 nT·s·1;中子注量率分别可达2.66×1010nT·cm·2·s·1、3.53×1010nT·cm·2·s·1。     

砷化镓探测器的镉上中子辐照改性

介绍反应堆镉上中子辐照改性的掺铬砷化镓（ＧａＡｓ：Ｃｒ）和本征砷化镓（ＧａＡｓ）光电导探测器，研究了探测器的响应时间、粒子灵敏度及输出电流随镉上中子辐照改性注量和施加偏压的变化关系。     

加速器驱动嬗变研究装置中关键材料的辐照损伤分析

反应堆中各结构部件的抗辐照性能,对整个系统的使用寿命及安全性有较大的影响。本工作通过MCNPx2.70蒙特卡罗软件建立CiADS散裂靶次临界反应堆模型,结合NJOY2016核数据截面处理软件制作的材料原子离位截面,在堆芯组件数分别为30,42,72盒的情况下分别计算和分析了316L、15-15Ti、SIMP 3种不锈钢材料和ZTA陶瓷作为候选结构材料的中子辐照损伤情况。当用作CiADS燃料包壳时,3种不锈钢材料中SIMP制成的包壳的R_dpa值最小,在燃料组件数为30,42,72盒的情况下其年辐照损伤量分别约为1.16,1.61和12.0 dpa/a。而ZTA制成的燃料包壳的R_dpa值均大于不锈钢材料的辐照损伤。在散裂靶次临界反应堆耦合区域,轴向上CiADS中心管在束靶作用面附近所受到的辐照损伤最大。燃料组件数为30盒时,由316L制成的中心管的辐照损伤率峰值约为2.7 dpa/a。     

基区表面势对栅控横向PNP晶体管中子位移损伤的影响

通过在常规横向PNP晶体管基区表面氧化层上淀积栅电极,制作了可以利用栅极偏置调制基区表面势的栅控横向PNP晶体管。对无栅极偏置电压和偏置电压分别为-10V和10V的栅控横向PNP晶体管,在西安脉冲反应堆上开展注量为2×1012,4×1012,6×1012,8×1012,1×1013 cm-2的中子辐照实验,研究基区表面势的增加和降低对栅控横向PNP晶体管中子位移损伤退化特性的影响。研究结果表明,基区表面势的增加引起栅控横向PNP晶体管共射极电流增益倒数的变化量随辐照中子注量的退化速率增加,基区表面势的降低对位移损伤退化速率无明显影响。     

位移损伤诱发CMOS图像传感器电荷转移损失退化的理论模拟研究

以4T PPD（4个晶体管的钳位光电二极管）型CMOS图像传感器为研究对象，开展注量为1×10¹¹,3×10¹¹,5×10¹¹,7×10¹¹,1×10¹² neutron/cm²的中子辐照下损伤模拟的研究，建立CMOS图像传感器的器件模型和不同注量中子辐照后位移损伤的缺陷模型；采用相关双采样技术测量由亮到暗连续两帧时序脉冲下浮置节点（FD）的输出值，建立测量电荷转移损失（CTI）的模拟方法；获得CTI随中子辐照注量的变化关系，分析CTI随中子累积注量的变化规律，结合中子辐照效应实验验证中子辐照诱发CTI退化的理论模拟计算结果的有效性。研究结果表明，CMOS图像传感器的位移损伤敏感区域为空间电荷区域，中子辐照后会在空间电荷区中引入位移损伤缺陷，这些缺陷通过不断俘获和发射载流子，使信号电荷不能快速转移到FD中，造成电荷转移损失；电荷转移损失随着中子辐照注量的增加而增大，二者在一定范围内呈线性关系。     

不同类型核燃料对热管冷却反应堆燃耗性能的影响

热管冷却反应堆采用固态反应堆设计理念,具有功率密度高、结构紧凑、固有安全性高等特点,在深空探索、深海勘探、偏远地区等场景中具有广阔的应用前景。核燃料作为热管冷却反应堆的重要组成部分,不同类型核燃料在堆芯燃耗分析时会呈现不同的中子学性能。基于美国爱达荷国家实验室(INL)提出的热管冷却反应堆INL Design A,利用清华大学蒙特卡罗中子输运程序RMC (Reactor Monte Carlo code)建立堆芯物理模型,选取UO₂,(U_0.9Pu_0.1)O₂,U-10Zr,U-8Pu-10Zr,UN,UC这6种核燃料开展燃耗计算,分析了不同核燃料、不同功率水平对热管冷却反应堆堆芯燃耗性能的影响。计算结果表明:在堆芯燃耗深度相同情况下(20.8 GW·d·t?1),装载U-8Pu-10Zr燃料的堆芯所需235U富集度最低(9.8%),具有较好的U-Pu增殖性能。堆芯功率处于5 MW的热管冷却反应堆,燃料中241Pu的存在不仅没起到增大堆芯燃耗深度的作用,反而导致堆芯剩余反应性和堆芯寿期末次锕系核素(MAs)的产量增大,影响反应堆的安全性与经济性。因此,对于装载含有Pu燃料的小功率长寿期热管冷却反应堆,需重点关注241Pu对堆芯燃耗性能的影响。     

高温气冷堆堆芯中子时空动力学模拟计算

研究开发三维圆柱几何堆芯多群中子时空动力学改进准静态方法模拟计算程序.对给定的模块式高温气冷堆模型进行模拟计算.初始状态下,计算结果与中子扩散程序CITATION吻合很好.动态情况下,模拟堆芯反应性、堆内各能群中子平均注量率和堆芯相对功率等物理量随时间变化,计算结果与理论分析一致.     

JMCT在CPR1000工程设计中的验证与确认

CPR1000系列反应堆是目前国内广泛应用的第二代压水堆型号之一，蒙特卡罗程序在CPR1000系列反应堆的验证与确认是该程序实现反应堆工程设计应用的关键环节。基于某CPR1000机组实际参数，使用由国内单位研发的蒙特卡罗程序JMCT在该机组开展了粒子输运建模计算，分别进行了临界计算和固定源计算，并进行了验证与确认。对于临界计算，采用JMCT建立了全堆芯pin-by-pin模型，计算了堆芯有效增殖因子和功率分布。对于固定源计算，建立适用于屏蔽分析的反应堆模型和辐照监督管精细结构模型，计算了两个核电机组多个循环的辐照监督管探测器位置累积快中子注量。通过将JMCT的计算结果与参考程序的计算结果、反应堆实际测量值进行了对比，验证了JMCT程序在CPR1000反应堆工程设计中的实际使用效果，证明了JMCT程序具备工程级的计算精度。     

堆外探测器响应函数三维空间分布计算

堆外探测器响应函数代表了堆芯活性区各组件对堆外探测器计数率的贡献,反映了堆芯功率分布与探测器计数率的关系。研究了三维离散纵标法(SN)程序TORT的共轭输运方法,并开发相应的处理程序,实现了柱坐标下的三维共轭中子注量率到压水堆各燃料组件响应函数的转换。并基于CAP1400核电厂反应堆模型,分析了其堆外探测器响应函数空间分布的特性,与采用TORT多次正向输运计算结果进行了对比分析,两者符合较好。通过本文研究,实现了压水堆核电厂堆外探测器响应函数的三维空间分布计算。     

铍组件堆内长期服役后的尺寸测量

铍是核反应堆内的重要反射层材料,其辐照后的尺寸变化对反应堆的安全性具有重要的影响。为获得铍组件堆内长期服役后的尺寸变化,以对其堆内的服役性能评价提供基础数据,设计并加工了一套高放样品远程转运平台,使用三坐标测量机完成了绵阳SPRR-300堆内铍组件的尺寸变化测量实验。实验测量结果表明,SPRR-300堆的铍组件在服役29 a后,在最高中子通量高达6.78×1021 cm?2的辐照环境下,铍组件外形尺寸总体上保持良好,截面有微量的收缩变形,最大形变约0.13 mm,这表明在长期中子辐照环境下,辐照蠕变是导致铍组件尺寸变化的主要原因。     

6061-Al合金的自离子辐照损伤效应

铝合金是国内外研究堆的主要结构材料,在前期300#研究堆主要结构材料铝合金辐照性能研究的基础上,通过离子辐照研究6061-Al合金的微观结构损伤和引起的硬度变化,以开展较高辐照剂量下6061-Al合金损伤效应的前期探索。结果表明,经过自离子辐照后,6061-Al合金中产生了夹角为72°的位错环等缺陷,随着辐照剂量从0.218×1016 cm?2增加到4.367×1016 cm?2,缺陷密度明显增加,但选区电子衍射表明合金保持了很好的晶体结构,并没有发生非晶化。纳米压痕测试表明,不同辐照剂量下,样品中产生了不同程度的硬化,且微观硬度随着辐照剂量的增加而增加,当剂量增加到2.183×1016和4.367×1016 cm?2时,辐照硬化达到饱和,约为11%。研究结果可为初步预测较高中子辐照剂量下6061-Al合金结构和性能的变化提供数据支撑。     

Investigation of stability of ITER candidate Hall sensors under neutron irradiation

Hall sensors offer an attractive true non-inductive method of magnetic field measurements for fusion reactors. Their use for steady state magnetic diagnostics of ITER is presently limited by their questionable radiation and thermal stability. Issues of stable and reliable operation in ITER like radiation and thermal environment are addressed by the contribution. Recently, novel Hall sensors, compatible with temperatures up to 200°C, were developed and their radiation stability was tested at LVR-15 experimental fission reactor. Overview of the experimental set-up on LVR-15 reactor is given. Degradation of the sensor’s sensitivity by several tens of percents was observed after neutron irradiation by the total neutron fluence of 2 × 10¹⁷ n/cm² in LVR-15. This level of neutron fluence is comparable to that expected to occur over the whole ITER life time for a sensor location just outside the ITER vessel. The in-situ recalibration techniques are expected to handle the observed degree of Hall sensors performance degradation in ITER environment.     

Kinetics of radiation damage and martensitic transformations in TiNi alloys irradiated with neutrons

The variation of the temperatures of martensitic transformations and the rate of radiation damage in TiNi alloys were studied upon irradiation with reactor neutrons. The irradiation was performed at temperatures of 120 and 335 K. In the process of irradiation, electrical resistance of the alloys was measured continuously and thermal cycling through the temperature range of martensitic transformations was carried out. The transformation temperatures were shown to decrease at different rates with increasing irradiation fluence. The electrical resistance increases linearly with increasing neutron fluence to 6.7×10¹⁸ cm^?2 irrespective of the irradiation temperature. Deviation from a linear dependence is only observed when the irradiation leads to a change in the phase state of the alloy. The rate of the resistance increase only slightly depends on the irradiation temperature. In martensite, it is greater by a factor of 2–4 than that in austenite. Mechanisms of irradiation-induced modification of the structure of TiNi alloys that explain the experimental data obtained are discussed.     

Martensitic transformation in a copper-aluminum-nickel alloy single crystal after short-term neutron irradiation

The influence of neutron irradiation on the temperature kinetics of thermoelastic martensitic transformation in a Cu-Al(13.4%)-Ni(5%) alloy single crystal is investigated by measuring the electrical resistivity directly under irradiation of the sample in a nuclear reactor channel. It is revealed that, after irradiation of the crystal in a martensitic or two-phase state, the temperature of the phase transition upon heating becomes 25–30 K higher than that prior to irradiation. This shift in the transition temperature is observed only upon the first heating, and the kinetics of martensitic transformation is restored in subsequent thermocycles. The shift in the transformation temperatures after irradiation increases with an increase in the fluence. The experimental results are explained by a disturbance of coherence at the interfaces in the irradiated crystals.     

Neutron-irradiation-induced shape memory effect in a TiNi alloy

The recovery of inelastic strains in Ti-Ni alloy samples irradiated in a nuclear reactor under isothermal conditions was studied. Before irradiation, the cylindrical samples were compressed to a residual strain of 3–6% in the martenstici state at room temperature. The samples were irradiated at a temperature of 45°C, which does not exceed the temperature of the onset of the reverse martensitic transformation A _S . Irradiation with a fastneutron fluence of 5 × 10²⁰ cm^?2 is established to result in the recovery of the residual strain. The value of the recoverable strain is comparable to that observed under the conditions of the shape memory effect on heating of the deformed alloy and even somewhat exceeds it. The obtained data show that neutron irradiation can induce the shape-memory effect in the TiNi alloy. This is due to a decrease in the temperatures of the martensitic transformations under irradiation.     

Magnetic hysteretic characterization of the irradiation-induced embrittlement of Fe,Fe–Cu model alloys,and reactor pressure vessel steel

Neutron irradiation is known to cause embrittlement of iron-based materials; in the nuclear industry, this effect can be detrimental for reactor pressure vessel steels. In this paper, we investigate the variations of the magnetic hysteretic behavior due to neutron irradiation, for four materials, i.e. nominally pure Fe, Fe-0.1 wt% Cu and Fe-0.3 wt%Cu model alloys, and a reactor pressure vessel steel, JRQ A533-B. Two parameters related to the magnetization loop shape, i.e. maximum relative differential permeability and peak intensity of local interaction field distribution, are measured as a function of neutron fluence. For all materials both parameters decrease with increasing fluence, due to the irradiation-induced formation of nano-size defects. This decreasing trend in magnetic parameters during embrittlement is noticeable regardless the origin of the embrittlement, which can be only Cu-precipitation (thermal aging of Fe–Cu), only matrix damage (irradiation of pure Fe), or both mechanisms (irradiation of Fe–Cu or steel). The magnetic parameters relatively change up to 40%, which indicates the potential of magnetic characterization to assess irradiation-induced material hardening and embrittlement.     

SPRR-300研究堆随堆辐照石墨的中子辐照效应北大核心CSCD

随SPRR-300研究堆约30 a的长时间运行,位于活性区附近的石墨箱体经历了长期的中子辐照。在长期服役的石墨箱体上取样,研究了其热学、力学以及微观结构变化,并与商用IG110,NG-CT-10石墨进行了对比。研究结果表明,经长时间低剂量率的中子辐照后,SPRR-300堆内随堆辐照石墨的晶格中出现了明显的辐照损伤缺陷,这些缺陷主要为位错环、层错、孔洞和微裂纹等,并出现了一定程度的非晶化。这些辐照损伤缺陷直接或间接地引起了石墨热学、力学性能的变化,主要表现为热膨胀系数、热扩散系数、抗压强度和抗弯强度的下降以及弯曲弹性模量的上升。     

Magnetic hysteresis properties of neutron-irradiated VVER440-type nuclear reactor pressure vessel steels

The development of non-destructive evaluation methods for irradiation embrittlement in nuclear reactor pressure vessel steels has a key role for safe and long-term operation of nuclear power plants. In this study, we have investigated the effect of neutron irradiation on base and weld metals of Russian VVER440-type reactor pressure vessel steels by measurements of magnetic minor hysteresis loops. A minor-loop coefficient, which is obtained from a scaling power-law relation of minor-loop parameters and is a sensitive indicator of internal stress, is found to change with neutron fluence for both metals. While the coefficient for base metal exhibits a local maximum at low fluence and a subsequent slow decrease, that for weld metal monotonically decreases with fluence. The observed results are explained by competing mechanisms of nanoscale defect formation and recovery, among which the latter process plays a dominant role for magnetic property changes in weld metal due to its ferritic microstructure.     

Electron irradiation-enhanced core/shell organization of Al(Cr,Fe, Mn)Si dispersoids in Al–Mg–Si alloys

The age hardening 6061-T6 aluminium alloy has been chosen as structural material for the core vessel of the material testing Jules Horowitz nuclear reactor. The alloy contains incoherent Al(Cr, Fe, Mn)Si dispersoids whose characterization by energy-filtered transmission electron microscopy (EFTEM) analysis shows a core/shell organization tendency where the core is (Mn, Fe) rich, and the shell is Cr rich. The present work studies the stability of this organization under irradiation. TEM characterization on the same particles, before and after 1 MeV electron irradiation, reveals that the core/shell organization is enhanced after irradiation. It is proposed that the high level of point defects, created by irradiation, ensures a radiation-enhanced diffusion process favourable to the unmixing forces between (Fe, Mn) and Cr. Shell formation may result in the low-energy interface segregation of Cr atoms within the (Fe, Mn) system combined with the unmixing of Cr, Fe and Mn components.