核级石墨辐照变形对辐照应力和寿命的影响

石墨是高温气冷堆的堆芯关键结构材料,其机械性能,尤其是辐照后特性,对反应堆的运行安全至关重要.不同牌号的石墨在制备工艺上有较大差异,导致内部微观结构的不同,从而影响石墨的辐照变形.本工作通过对高温气冷堆堆芯侧反射层石墨砖的辐照行为进行数值仿真,分析不同石墨材料的辐照变形对石墨结构的辐照应力和辐照寿命的影响.结果表明,石墨结构的辐照应力和辐照寿命对石墨材料的辐照变形高度敏感.相关结论将为高温气冷堆堆芯石墨砖的结构设计提供重要的数值依据.     

隐藏爆炸物检测装置快中子辐照安全分析

为推广隐藏爆炸物检测装置在反恐领域的应用,对快中子辐照炸药、食品及药品的安全性进行了分析。通过蒙特卡罗方法建立了装置的快中子辐照炸药安全性评估模型。通过能量沉积计算及炸药起爆机理分析可知,快中子辐照炸药不会产生爆炸的危险。采用剂量分析法对快中子辐照食品及药品的安全性进行了分析,结果表明,在隐藏爆炸物检测装置的快中子辐照条件下,食品及药品的辐照剂量在国家和国际限定的标准内,快中子辐照食品及药品的安全性是可接受的。     

非均匀辐照环境下燃料组件变形的初步数值模拟

以典型17×17燃料组件为研究对象,考虑非均匀快中子环境、材料辐照效应等因素,基于一定的假设与简化,初步实现了燃料组件整体辐照变形的三维数值模拟。模拟结果表明,在所设定的较大快中子注量和快中子注量梯度作用下,燃料组件仍具有较低的整体应力水平,但其引起的燃料组件辐照变形会对控制棒落棒和燃料组件装卸造成一定影响。     

核石墨耐辐照性能的评价方法

在高温气冷堆运行过程中,作为堆内构件的石墨经受高温和快中子的辐照,会经历先收缩后膨胀的宏观尺寸形变,并在膨胀至原始尺寸时到达使用寿命。在石墨尺寸形变的过程中,石墨内部气孔的结构和数目均有明显变化。当辐照剂量接近使用寿命时,石墨内部气孔数目明显增加,导致其力学性能急剧下降而退出服役。He⁺、C⁺、Xe⁺离子辐照实验表明,在200keV10¹⁴cm^-2Xe⁺离子辐照下,石墨气孔形貌变化明显。这一结果可作为石墨辐照性能的评价方法。     

CADDS5与ADINA的接口程序开发

ＣＡＤＤＳ５与ＡＤＩＮＡ的接口程序开发旨在使国际上先进的机械设计软件包ＣＡＤＤＳ５与通用有限元分析ＡＤＩＮＡ相结合，利用ＣＡＤＤＳ５的前处理功能形成有限元分析所必需的相关信息，编制前处理接口程序将春转移为ＡＤＩＮＡ的输入格式，编制后处理接口程序交ＡＤＩＮＡ的计算结果转换为ＣＡＤＤＳ５可接受的格式，返回ＣＡＤＤＳ５进行后处理工作。     

西安脉冲反应堆辐照腔参数理论计算分析

介绍了西安脉冲反应堆辐照腔参数的理论计算模型和计算程序。计算了辐照腔中的快中子（E≥0.1MeV)注量和辐照腔屏蔽外表面处的当量剂量率，理论计算与实测值比较表明，二者符合很好。     

辐照装置中锆-4合金圆管高温变形的堆外验证试验

新型核燃料入堆辐照考验时，需要设计专用的辐照装置。将锆-4合金圆管用作辐照装置的高温承压部件时，需要考虑其长期在高温高压下运行时的变形，以防止其变形过大影响辐照装置的热工设计，进而影响辐照参数的稳定性和堆芯安全，为此在堆外对锆-4合金圆管开展验证试验。将锆-4合金圆管制成的试验件充至不同的内压，并通电加热至不同的温度，稳定考验一段时间，测量其外径的变化，重复5次，最终得到其在不同温度和环向应力下考验后的变形量。将试验结果中由氧化和蠕变导致的变形，分别与Leistikow-Schanz公式和Rosinger公式的计算结果进行对比，发现Leistikow-Schanz公式计算结果比试验氧化增重结果要大15%左右，而Rosinger公式计算结果与试验蠕变变形结果符合较好，验证了试验结果的合理性。对试验结果进行分析，认为锆-4合金圆管在外壁环向应力29.10 MPa和460～470℃的工况下可以较长时间的运行。     

核电厂压力容器材料在49-2反应堆的辐照技术研究

应用MCNP程序对堆芯建模,计算得出辐照孔道内距堆心底部高25 cm处的中子能谱,结合多箔活化法测量结果,通过SANDII程序解谱得出该位置的快中子注量率;通过相对快中子注量率测量,获得孔道内轴向快中子注量率分布,从而确定辐照时长和辐照方案,使样品辐照达到快中子(E≥1 Me V)注量~6×1019cm-2的技术指标。为完成辐照样品解体,应用ORIGEN2程序计算,获得待解体样品源项;使用MCNP程序对解体时的操作环境进行建模,计算得出不同屏蔽层厚度的γ剂量率数据;与实测结果进行对比,计算结果与实测结果符合较好,证明屏蔽设计有效。本次辐照考验完全满足技术指标。。     

计入辐照效应的快堆燃料组件外套管截面变形有限元分析

快堆燃料组件外套管截面的辐照变形计算对快堆堆芯设计非常重要。本文研究考虑材料辐照蠕变和辐照肿胀效应,利用有限单元法计算外套管截面变形的方法。首先介绍了采用的辐照蠕变和辐照肿胀材料模型,其次给出了通过力学简化模型研究截面变形的理论方法,最后提出一种本构关系应力更新方案,通过将其编入ABAQUS子程序接口UMAT对外套管在压差作用下的截面变形进行了有限元分析计算,并比较讨论分析结果。结果表明有限元方法成功计算出了截面的变形,并在小变形时与理论解吻合较好。研究表明本文提出的本构关系应力更新方案是有效的;变形较大时理论解的偏差增大;内壁角点处应力水平最高,并伴随应力松弛效应。     

影响反应堆压力容器安全运行的根本原因及应对措施

从实现核反应堆安全目标和运行工况的角度,分析了反应堆压力容器在承受压力、温度和快中子辐照条件下的失效形式及根本原因。针对能量≥1MeV快中子辐照损伤,给出了预测和监督方法;对承压热冲击下可能引发脆性断裂进行了分析,并提出了分析方法。分析和介绍了各运行工况下RPV安全运行的压力一温度限值计算方法。     

Development of Thermal/Irradiation Stress Analytical Code “VIENUS” for HTTR Graphite Block

A new thermal/irradiation stress analysis code “VIENUS” has been developed for the graphite block in the High-Temperature Engineering Test Reactor (HTTR). The VIENUS is a two- dimensional finite element visco-elastic analysis code to take account of graphite behavior under irradiation in detail. In the analysis, the effects of both fast neutron fluence and temperature on material properties are considered.

The code has been evaluated by the irradiation test results of the Peach Bottom fuel elements to confirm the thermal/irradiation stresses in the graphite block. It is clarified that the calculated results are able to estimate a tendency of the test results, and that both the irradiation- induced creep and dimensional change are the most important parameters in the thermal/irradiation stress analysis. From the present study, it is suggested that the VIENUS code is a useful tool to evaluate the thermal/irradiation stresses in the HTTR graphite blocks.     

MC simulation of thermal neutron flux of large samples irradiated by 14 MeV neutrons

The response of a 14 MeV neutron-based prompt gamma neutron activation analysis (PGNAA) system, i.e.the prompt gamma-rays count rate and the average thermal neutron flux, is studied with a large concrete sample and with a homogeneous large sample, which is made of polyethylene and metal with various concentrations of hydrogen and cadmium using the MCNP-5 (Monte Carlo N-Particle) code. The average thermal neutron flux is determined by the analysis of the prompt gamma-rays using the thermal neutron activation of hydrogen in the sample, and the thermal and fast neutron activation of carbon graphite irradiation chamber of the PGNAA-system. Our results demonstrated that the graphite irradiation chamber of the PGNAA-system fairly operates, and is useful to estimate the average thermal neutron flux of large samples with various compositions irradiated by 14 MeV neutrons.     

Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

Evaluation of the influence of bypass flow gap distribution on the core hot spot in a prismatic VHTR core

Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage/swelling and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of fast neutron fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass gap and flow distributions are closely related to the local hot spot and its location and the core restraint mechanism preventing outward movement of the graphite block by a fastening device reduces the bypass gap size, which results in the decrease of maximum fuel temperature not less than 100 °C, when compared to the case without it.     

锆锡合金的辐照生长及在燃料组件结构设计中的考虑

锆锡合金由于中子辐照而引起生长。生长量是锆锡合金材料的织构、辐照温度、快积分通量和冷作量的函数。在燃料组件结构设计中应给予足够的重视。     

Analyses of a restrained growth graphite irradiation creep experiment

Graphite is used as a moderator, reflector and structural component in pebble bed and prism High Temperature Reactors (HTRs). It is fortunate to reactor designers that irradiated graphite shows remarkably high creep behaviour under the influence of fast neutron irradiation at temperatures far below those required for significant creep strains to be generated in unirradiated graphite. This creep behaviour is important in the design of nuclear graphite reactor cores because the self-induced shrinkage stresses generated in typical core components during irradiation can be relieved. However, there are no reliable data on high fluence irradiation creep and the mechanistic understanding of the irradiation creep is insufficiently developed to reliably extrapolate to the high fluences expected of graphite in future HTR designs. The understanding of irradiation creep is further complicated because it has been experimentally observed that irradiation creep strain in graphite modifies other properties in particular the coefficient of thermal expansion. In addition modified changes in Young's modulus in crept specimens have been reported and it has also been postulated that irradiation creep may also modify dimensional change. The assessment of irradiation creep in graphite components is based on empirical laws derived from data obtained from small samples irradiated in a materials test reactor. However, due to the complicated irradiation rigs required and the amount of dimensional and property measurements needed to be taken, constant stress irradiation creep experiments are difficult and very expensive to carry out successfully. However, restrained creep experiments are simple to implement, less expensive and can be easily included as part of other, more conventional irradiation graphite experimental programmes. However, in the past, the disadvantage of these experiments has been that the results have been difficult to interpret using the then available analytical methods. In this paper the restrained creep experiment is revisited and analysed numerically and the possible benefit of using a restrained creep experiment in future graphite irradiation experiments is investigated. It is shown that a numerical simulation of the restrained creep experiment behaviour would be an essential tool to ensure that the stress within the specimen remains within defined limits so that specimen failure can be avoided.     

Relationship between nuclear graphite moderator brick bore profile measurement and irradiation-induced dimensional change

A stress analysis for a hypothetical nuclear graphite moderator brick is presented, considering dimensional and other property changes due to fast neutron irradiation, to illustrate the relationship between the change in moderator brick bore profile and dimensional change of the material. The results give the stresses and deformations of the brick during operation and at shutdown, with the effect of irradiation creep on the deformation of the brick also considered. The analyses provide information useful to reactor designers and operators for planning graphite monitoring campaigns.     

The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.     

熔盐/辐照环境中细结构等静压石墨的结构和化学稳定性研究

熔盐堆以石墨作为反射体和慢化体,熔盐与石墨直接接触,石墨在熔盐中的腐蚀反应和辐照损伤是值得研究的问题。本文采用自主研发的细结构石墨,阻隔熔盐浸渗,采用30 MeV He+模拟中子辐照,研究不同温度及熔盐环境下石墨微观形貌、微结构和化学结构的变化。研究结果表明,高温环境下,由于高温的退火效应,石墨缺陷密度的增加及形貌的变化都远小于室温环境。辐照后的石墨与熔盐接触,其缺陷密度略微降低。这种微结构的改善与高温熔盐环境中的退火效应及熔盐固化引起内部微裂纹的闭合有关。辐照后的熔盐浸泡可在石墨C—C键结构中引入C—F键,且C—F键的形成与缺陷密度及缺陷类型密切相关。稳定的空位簇及间隙原子的迁移均会影响层间化合物的形成,从而产生限制C—F键形成的环境,进而降低由层间化合物的形成对石墨表面结构的破坏。     

反应堆压力容器中子辐照损伤的分析和评价

对于材料已经确定的反应堆压力容器,其辐照脆化效应的主要因素是快中子积分通量。本文应用中子输运格林函数法验算了秦山核电站压力容器1/4厚度处最大快中子通量。分析和评价结果表明,该压力容器的设计对中子辐照是安全的。