国产反应堆压力容器用16MND5钢的蠕变损伤本构模型研究

为了获得反应堆压力容器（RPV）材料在高温下的蠕变行为,保证RPV在严重事故工况下的完整性,本研究对国产RPV用16MND5钢的高温蠕变性能进行了测试,获得了600~900℃下材料的蠕变性能,并基于应变强化的基本蠕变本构模型与基于延性耗竭理论的蠕变损伤模型,建立了适用于16MND5钢的蠕变损伤本构模型,给出了材料的蠕变损伤模型参数。结果表明,本文提出的蠕变损伤本构模型的有限元模拟数据与试验数据符合性较好,验证了此蠕变损伤模型的正确性。该方法可用于严重事故情况下RPV的蠕变损伤分析,为RPV的完整性分析提供支持。      

带拐角裂纹的反应堆压力容器接管的安全分析

从疲劳断裂的角度对反应堆压力容器(RPV)在两种停堆过程中,带拐角裂纹的RPV接管的安全性进行了分析评定。针对压力和温度随时间变化的复杂情况,采用“半解析法”和弹塑性断裂力学观点进行了大量的分析计算。结果表明,ASME规范第Ⅺ卷的观点(以LEFM为基础)似偏保守。     

Evidence for chromium stabilisation of irradiation damage in RPV weld metals

A study using some available high-quality data from German surveillance and research programmes has been carried out in order to provide evidence on the prediction capabilities of a semi-mechanistic model for changes in ductile to brittle transition temperature. The model was developed and tuned originally on model alloys and later proved to be suitable also for WWER and other materials. This study demonstrates that the model can predict very well all the results obtained on low nickel–manganese weld metals and that the corrections required for high nickel–manganese weld metals are consistent with those found for model alloys.     

Grain size effects in the quasi-static fracture resistance of a thermally embrittled RPV steel (annealed microstructures)

The elastic-plastic fracture toughness and crack extension behavior under quasi-static loading regimes of the as-received and several thermally embrittled states of a reactor pressure vessel (RPV) steel were assessed on the basis of microstructural parameters. Through a simple rule of mixture that is typically applied for composite materials, it was found that the equivalent grain size (EGS) of dual-phase annealed microstructures is the controlling parameter of the fracture properties. It was concluded that a Hall-Petch type relationship correlates the J-fracture mechanics criteria to the EGS.     

秦山核电厂反应堆压力容器水压试验应变测量与分析

对秦山核电厂反应堆压力容器出厂水压试验测点布置作了说明,与一回路系统水压试验位移测量的主要结果作了分析对比,给出了实测载荷与主螺栓测试数据,讨论了表征密封性能的法兰转动,认为出厂水压试验此项结果有错.     

核电厂压力容器支座“流场-温度场”耦合分析

反应堆压力容器(RPV)支座在核电厂设计中属ASME核安全1级支承,是关系到RPV安全的关键设备。通过建立支座的"流-固"耦合CFD模型和基于响应面法的实验设计,分析了支座入口空气流速和温度对支座温度场和空气流场的影响规律;进一步面向支座底面最高温度限制和减小空气流速的双目标要求,获得了优化的入口空气参数组合;最后通过对比"流-固"耦合模型结果和响应面法拟合结果,证明了模型和分析方法的正确性。     

反应堆压力容器钢辐照脆化的影响因素分析

压力容器（RPV）在工况条件下经中子辐照后将导致力学性能退化,发生辐照脆化现象。相关研究表明,引起RPV钢辐照脆化的因素包括自身材料因素和服役环境因素。材料因素是RPV钢自身性质对辐照脆化的影响,包括材料化学成分、微观组织特性与晶粒尺寸,环境因素是RPV服役环境对辐照脆化的影响,包括中子注量、中子注量率、中子能谱和辐照温度。     

Performance of a Monte-Carlo solution for the H.B. Robinson-2 Pressure Vessel Dosimetry Benchmark

At the Paul Scherrer Institut, a methodology for PWR fast neutron fluence estimations, based on the Monte-Carlo particle transport code MCNPX with general-purpose neutron data libraries and using neutron source data from deterministic 3-D core-follow calculations, has been developed. The methodology has been validated on the basis of experimental data related to the fluence at the inner surface of a Swiss PWR Reactor Pressure Vessel. In this technical note, a first objective is to enlarge the validation basis of the methodology as well as to extend it for applications to RPV outer-surfaces. To that aim, a preliminary analysis with the MCNPX-2.4.0 code, along with the JEFF-3.1 continuous-energy neutron data library, of the “H.B. Robinson-2 Pressure Vessel Benchmark”, providing in-vessel and out-vessel experimental dosimetry data, is presented. In addition, considering that the available original solutions of the benchmark employed deterministic transport methods with associated libraries, a second objective of this technical note is to assess the progress achieved for this type of problems when applying modern Monte-Carlo based methods. The results show that for the ²³⁷Np and ²³⁸U fission dosimeters, which were of primary interest in the given study, a non-negligible improvement is seen in the MCNPX solution, indicating thus a rather good performance of the employed Monte-Carlo method and providing thereby, additional confidence for the overall PSI fluence methodology. For other high-energy dosimeters, the presented new results do not show yet any significant accuracy improvement versus previously reported results. This can however not be confirmed before additional studies, e.g., with focus on improvements of the involved modelling approximations and the statistical precision of the MCNPX calculations, be carried out. Similarly, investigations of neutron cross-section and dosimetry libraries effects remain to be addressed. These further studies are however not included here since at this stage, the principal aim was mainly to model and analyse this benchmark at most consistent manner with the previous solutions using a continuous-energy Monte-Carlo based method.     

反应堆压力容器模拟钢中富Cu原子团簇对材料力学性能的影响

采用淬火和时效处理方法诱发压力容器模拟钢中的微量杂质元素Cu以富铜原子团簇析出。力学性能试验结果表明,富铜原子团簇的出现导致压力容器模拟钢韧脆转变温度出现明显变化,而屈服强度和抗拉强度增长较小,塑性也只有较小程度下降。三维原子探针分析结果表明,富铜原子团簇的数密度为1023 m-3数量级,富铜原子团簇直径为1～3 nm。