10CrMo910和316不锈钢的时间相关失效评定曲线

对10CrMo910和316不锈钢进行了总计约1万小时的蠕变拉伸试验、得到了材料的等时应力应变曲线和时间相关失效评定曲线,并给出了长时蠕变情况下的时间相关失效评定曲线方程。     

高温翅片管的长时蠕变屈曲分析及评定方法研究

由于翅片管处于高温环境和外压载荷下,需要考虑其发生蠕变屈曲失效的风险。本文对翅片管在高温环境下的蠕变屈曲分析及评定方法进行了研究,提出了一种基于塑性本构和蠕变本构的有限元长时蠕变屈曲分析方法,并通过数值拟合,获得了高温屈曲的失效评定图以及失效评定公式,提出了一种方便应用于工程的快速评定方法。针对翅片管结构,将该方法的评定结果与规范中的屈曲分析评定结果进行对比,验证了该方法的可行性。同时研究了在有压力波动的情况下,结构的临界屈曲时间与载荷历程的关系,为复杂结构和复杂载荷工况的蠕变屈曲分析奠定了基础。     

高温翅片管的长时蠕变屈曲分析及评定方法研究

由于翅片管处于高温环境和外压载荷下,需要考虑其发生蠕变屈曲失效的风险。本文对翅片管在高温环境下的蠕变屈曲分析及评定方法进行了研究,提出了一种基于塑性本构和蠕变本构的有限元长时蠕变屈曲分析方法,并通过数值拟合,获得了高温屈曲的失效评定图以及失效评定公式,提出了一种方便应用于工程的快速评定方法。针对翅片管结构,将该方法的评定结果与规范中的屈曲分析评定结果进行对比,验证了该方法的可行性。同时研究了在有压力波动的情况下,结构的临界屈曲时间与载荷历程的关系,为复杂结构和复杂载荷工况的蠕变屈曲分析奠定了基础。     

基于J-Q双参数准则的材料裂纹体失效评定曲线工程化研究

针对应力三轴性对裂纹尖端应力应变场的影响，讨论了均匀材料裂纹体的断裂准则．建立了基于J-Q双参数准则的失效评定曲线。利用有限元数值解的结果，得到了失效评定曲线中各参量的工程估算方法。通过实例，对本文所建立失效评定曲线的计算方法和现有的计算途径进行了比较。     

样品尺寸变化对T91钢蠕变力学行为的影响

T91钢是第4代反应堆的候选结构材料之一，中子辐照后的高温蠕变性能是评价其服役性能的关键指标。为充分利用辐照空间、减小辐照参数梯度和降低样品放射性，针对力学性能的研究需要使用小尺寸样品，但小样品试验数据可能与标准样品不同，导致无法准确评价材料性能。为研究样品尺寸变化对T91钢蠕变力学行为的影响，本文对T91钢小片状试样和标准棒状试样在温度675～725℃、蠕变应力80～120 MPa下的蠕变行为和断裂机理进行了对比研究。结果表明，不同尺寸样品均发生了减速蠕变、稳态蠕变和加速蠕变3个变形阶段，且断裂时间均随蠕变温度和应力的增大而减小，但小片状试样的蠕变断裂时间更长、稳态蠕变速率更小；所有试样均发生了微孔聚集型韧性断裂，但小片状试样断口的韧窝尺寸相对更小且受到剪切应力；试样尺寸变化不影响T91钢蠕变变形机制，造成差异的原因是试样应力状态差异；蠕变过程中不同的微观结构演化是蠕变试样尺寸效应随蠕变温度、应力变化规律复杂的重要原因。     

NiTiNb形状记忆合金的应力松弛研究

为了解NiTiNb形状记忆合金的高温松弛性能,进行了高温下的松弛试验。由应力松弛与蠕变的关系,推导得到松弛曲线表达式。通过实验数据回归,发现松弛曲线表达式与实验结果吻合良好,并得到了表征材料抗松弛性能的材料松弛特征系数和剩余应力比。结果表明,温度越高,初始应力越大,应力松弛越明显;当温度在300～400℃,初始应力在260～360MPa时,NiTiNb的应力松弛很小;在高温下NiTiNb的抗松弛性能优于NiTiFe,更适合于高温下使用。     

反应堆压力容器钢SA533B的修正θ-projection蠕变模型研究

反应堆严重事故下,压力容器存在因冷却不充分而蠕变失效的风险。基于修正θ-projection蠕变模型提出了一种针对反应堆压力容器钢SA533B的蠕变模型,该模型能够完整描述3段蠕变过程,模拟结果与实验结果蠕变曲线符合较好,同时可通过插值方法预测任意载荷下的蠕变行为。该模型可进一步用于压力容器失效的相关算例分析。      

NiTiNb形状记忆合金的应力松驰研究

为了解NiTiNb形状记忆合金的高温松驰性能，进行了高温下的松驰试验，由应力松驰与蠕变的关系。推导得到松驰曲线表达式，通过实验数据回归，发现松驰曲线表达式与实验结果吻合良好，并得到了表征材料抗松驰性能的材料松驰特征系数和剩余应力比，结果表明，温度越高，初始应力越大，应力松驰越明显：当温度在300－400℃，初始应力在260－360MPa时，NiTiNb的应力松驰很小，在高温下NiTiNb的抗松驰性能优于NiTiFe，更适合于高温下使用。     

基于PepS的高温核一级管道蠕变疲劳分析方法研究

钍基熔盐堆(Thorium Molten Salt Reactor-Liquid Fuel,TMSR-LF1)回路管道最高运行温度达650℃,高温服役下的管道蠕变-疲劳损伤分析及评定至关重要。目前仅ASME-BPVC-III-5-HBB规范中有适用于高温核一级管道的蠕变-疲劳损伤暂行评定方法,但该方法对于复杂管道系统使用起来过于繁琐。本文旨在使用管道分析软件PepS软件实现高温核一级复杂管系的分析与结构完整性评估。首先结合管道结构在多种载荷组合作用下的截面应力状态解析解,进行管道截面应力分析及应力线性化,并将结果与有限元数值解进行对比分析,两者的误差结果基本一致。随后,利用PepS软件对TMSR-LF1回路管道进行了力学分析和结构完整性评估,结其蠕变疲劳损伤结果位于包络线以内,满足蠕变疲劳极限的要求。该研究将管道分析软件与ASME评定规范进行了有效衔接,明确了评定方法,实现了高温核一级复杂管系的蠕变疲劳评估。     

钍基熔盐堆回路管道系统应力分析与评定

钍基熔盐堆(TMSR)管道设计温度可达700℃,设计标准采用美国机械工程师协会ASME-NH分卷。高温管道评定时除需要进行应力评定外,还需进行应变变形限值和蠕变疲劳限值等评定。利用通用有限元分析软件(ANSYS)对整体回路系统进行计算,并通过优化计算,使得管道应力达到ASME规范中限值要求。     

基于孔洞长大理论的多轴蠕变设计模型及其工程应用

多轴蠕变是高温构件失效的重要原因,工程设计中必须予以考虑.本文介绍了以孔洞长大理论为基础的多轴蠕变设计方法;探讨了孔洞长大的基本原理;讨论了基于该理论的5个常用模型;分析了这些模型在R5、ASME III、RCC-MR 及VGB-R 509L 中的应用,指出了高温构件多轴蠕变设计的困难,并对进一步的工作提出了建议.     

Assessing the defect-free structure

There has been a strong incentive within Nuclear Electric to develop a comprehensive assessment procedure to evaluate the high temperature response of structures; the so-called R5 procedure is extensive and Vols. 2 and 3 deal with the analytical approach and the assessment of creep fatigue damage for defect-free structures. In this paper we describe the approach which has been adopted and identify the relationship between the analytical requirement and material damage assessment under cyclic loading.     

Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall.     

Considerations of Irradiation-Induced Transient Creep in Fuel Particle Modeling*

An analytical assessment is made of the potential effects of irradiation-induced transient creep on the behavior of the TRISO-coated fuel particles of the New Production Modular High Temperature Gas-Cooled Reactor (NP-MHTGR). An analytical solution is presented for the three-layer particles, which includes transient creep in addition to steady-state creep behavior. The solution allows for evaluating the effects that transient creep has on individual particle stresses and for determining failure probabilities for particle batches using the Monte Carlo approach. Because experimental data needed to determine parameters for a transient component in a creep model for the pyrocarbons is not available, a range of possible parameter values were considered in the assessments. It was shown that transient creep measurably affects particle stresses early in the irradiation life of the particle. At that time, the hoop stress in the primary load bearing layer of the particle is in compression and the article is not vulnerable to pressure vessel failure. Later in irradiation, the effects of transient creep were typically shown to be less significant. Thus, transient creep had less than an order of magnitude effect on batch failure probabilities for prototypical NP-MHTGR fuel particles and was much less significant than steady-state creep. Whether the presence of transient creep increased or decreased the particle failure probability was dependent on the specific values used for the transient creep material properties.     

Probability assessment of graphite brick in the HTR-10

A probability finite element assessment program was developed to evaluate the security of graphite components in the HTR-10 (10 MW high temperature gas-cooled reactor-test module), based on the MARC non-linear finite element code and the strength uncertainty of the graphite material. Using user-defined subroutines (UDS), the irradiation thermal analysis subroutine, irradiation static analysis subroutine and probability assessment subroutine are embedded into the MARC program. The recompiled MARC program take into account irradiation-induced changes in graphite components such as the thermal conductivity coefficient, the thermal expansion coefficient, the creep coefficient, the elastic modulus, and the strength. The failure probabilities of the graphite components in the HTR-10, either under normal operating conditions or cold shutdown conditions, were evaluated. Additional analyses were done with the irradiation deformation increasing 20% and the creep coefficient decreasing 20%, to see the influence of irradiation deformation and the creep effect on the failure probability. The study showed that the probability finite element assessment method is an effective tool to assess the probability of structure failure.     

基于耦合CFD-FEM方法的严重事故下RPV蠕变失效风险评估

核电站严重事故发生后,反应堆压力容器（RPV）的剩余固壁在高温差、内压、熔池重量等的作用下可能发生蠕变失效。本文以CPR1000 RPV为研究对象,基于FLUENT软件二次开发求解反应堆压力容器下封头烧蚀温度场,然后基于ANSYS Workbench开展耦合CFD-FEM力学分析,求解严重事故下RPV烧蚀温度场稳定后72 h内的等效应力、等效塑性应变和等效蠕变应变,并评估了RPV的蠕变失效风险。结果表明：当堆坑注水等措施投运后,RPV剩余固壁在72 h内不会发生蠕变失效和塑性变形失效,有效卸压可明显提升RPV结构完整性的安全裕度。     

Risk Assessment of Creep Failure for RPV under Severe Accident Condition Based on Coupled CFD-FEM Method

After a reactor core melt accident, creep failure may occur in the residual solid wall of the reactor pressure vessel (RPV) under the influence of high temperature difference, internal pressure and the weight of the molten pool. In this work, the CPR1000 RPV was used as a research object. The ablation temperature field of the lower head of RPV was solved through the secondary development of the FLUENT software. And then, a CFD-FEM coupling analysis was carried out based on ANSYS Workbench software. The equivalent stress, the equivalent plastic strain and the equivalent creep strain of the RPV within 72 h under severe accident after the wall ablation and temperature field distribution formed stably were calculated. The risk of creep failure of the RPV was evaluated. The results show that when the reactor pit water injection measure puts into operation, the residual solid wall of the RPV will not experience creep failure and plastic deformation failure within 72 h, and besides, the pressure relief can significantly increase the safety margin of the structural integrity of the RPV.