共查询到20条相似文献,搜索用时 15 毫秒
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B. Margolin B. Gurovich V. Fomenko V. Shvetsova A. Gulenko D. Zhurko M. Korshunov E. Kuleshova 《Journal of Nuclear Materials》2013,432(1-3):313-322
New fracture toughness data are represented for highly irradiated RPV materials that were obtained by testing standard compact specimens with thickness of 12.5 mm and 25 mm and pre-cracked Charpy specimens machined from the RPV decommissioned. Two advanced engineering methods, the Master Curve and the Unified Curve, are applied for treatment of the test results. Application of the dependence of fracture toughness KJC on test temperature T predicted with the Master Curve and the Unified Curve methods on the basis of surveillance specimens testing is discussed for RPV integrity assessment when the reference KJC(T) curve is recalculated to the crack front length of the postulated flaw that is considerable larger than thickness of surveillance specimens. The prediction of the KJC(T) curve transformation caused by neutron irradiation is considered. 相似文献
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In case of a postulated loss of coolant accident (LOCA) of a reactor pressure vessel (RPV), the nozzle region experiences higher stresses and lower temperatures than the remaining part of the RPV. Thus, the nozzle is to be considered in the RPV safety assessment. For a LOCA event, three-dimensional elastic–plastic finite element calculations of stresses and strains in the intact RPV were performed. Using the substructure technique, fracture mechanics analyses were then carried out for several postulated cracks in the nozzle corner and in the circumferential weld below the nozzle. For different crack geometries and locations, the J-integral and the stress intensity factor were calculated as functions of the crack tip temperature. Based on the KIC-reference curve and the JR curve, both brittle and ductile instability of the postulated cracks were excluded. In order to reduce the expenses of three-dimensional finite element analyses for various crack geometries, an analytical procedure for calculating stress intensity factors of subclad cracks in cylindrical components was extended for cracks in the nozzle corner. 相似文献
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Fracture assessment of shallow-flaw cruciform beams tested under uniaxial and biaxial loading conditions 总被引:1,自引:0,他引:1
B. R. Bass W. J. McAfee P. T. Williams W. E. Pennell 《Nuclear Engineering and Design》1999,188(3):259
A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow surface flaws. Matrices of cruciform beam tests were developed to investigate and quantify the effects of temperature, biaxial loading, and specimen size on fracture initiation toughness of two-dimensional (constant depth), shallow, surface flaws. The cruciform beam specimens were developed at Oak Ridge National Laboratory (ORNL) to introduce a far-field, out-of-plane biaxial stress component in the test section that approximates the nonlinear stresses resulting from pressurized-thermal-shock or pressure–temperature loading of an RPV. Tests were conducted under biaxial load ratios ranging from uniaxial to equibiaxial. These tests demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for an RPV material. The cruciform fracture toughness data were used to evaluate fracture methodologies for predicting the observed effects of biaxial loading on shallow-flaw fracture toughness. Initial emphasis was placed on assessment of stress-based methodologies, namely, the J–Q formulation, the Dodds–Anderson toughness scaling model, and the Weibull approach. Applications of these methodologies based on the hydrostatic stress fracture criterion indicated an effect of loading-biaxiality on fracture toughness; the conventional maximum principal stress criterion indicated no effect. A three-parameter Weibull model based on the hydrostatic stress criterion is shown to correlate with the experimentally observed biaxial effect on cleavage fracture toughness by providing a scaling mechanism between uniaxial and biaxial loading states. 相似文献
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In the frame of the multi-scale approach of the fracture toughness prediction defined in the PERFECT project, we proposed a new crystal plasticity model and applied it to the computation of stress heterogeneities within a reference polycrystalline aggregate defined in the project RPV material.The proposed crystal plasticity model is able to take into account the effects of temperature and irradiation hardening. The analysis of the results of aggregate computations shows that the distributions of the maximum values of the maximal principal stresses are found to be well described by a Gumbell function. Applying these distributions on a Griffith criterion allows settling the basis of an original fracture criterion. However the increasing resistance to fracture of the steel with temperature can be reproduced only by introducing a temperature dependence of the fracture energy. 相似文献
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带拐角裂纹的反应堆压力容器接管的安全分析 总被引:1,自引:0,他引:1
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。 相似文献
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The regulating norms for nuclear power plant component are based on the assumption that the components have no defects. Therefore it is of major interest to find what values the fracture mechanics parameters assume when a loading which produces in an unflawed elbow stresses which are acceptable according to the regulatory guides is applied to an elbow with cracks. It could be shown that with small flaws (
of arc length,
) calculations assuming linear-elastic behavior give results almost identical as when elastic-plastic behavior is assumed. If the cracks are small and the loading is according to the regulatory guides the initiation value J1 for stable crack growth is not reached. 相似文献
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Krassimira Ilieva Sergey Belousov Antonio Ballesteros Bohumil Osmera Sergey Zaritsky 《Progress in Nuclear Energy》2009,51(1):1-13
A state of the art review of Reactor Dosimetry used for reactor pressure vessel irradiation damage assessment and lifetime evaluation of the Russian type VVER reactors is presented. The necessity of prospective studies in Reactor Dosimetry for improvements that will reduce the neutron fluence uncertainty and in this way to substantiate the extension of NPP lifetime is summarized by specialists in Reactor Dosimetry from countries operating VVER reactors such as Russia, Ukraine, Czech Republic, Finland, Hungary, and Bulgaria, together with specialists from Western European countries such as France, Spain, Germany, Belgium, The Netherlands, and UK, operating PWR and BWR type reactors. 相似文献
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反应堆压力容器承压热冲击(PTS)分析 总被引:1,自引:1,他引:0
在反应堆运行过程中发生严重的失水事故(LOCA)时,应急堆芯冷却系统启动,冷的安注水从安注接管注入反应堆压力容器(RPV)中,此时压力容器还维持较高压力,这种瞬态就称为承压热冲击,即PTS(Pressurized ThermalShock).按照10CFR50,61[2]和RCC-M规范[1],对安注接管、焊缝和堆芯筒体三个区域,进行了PTS工况评估,分析结果表明,在发生PTS时,压力容器的完整性是能够保证的. 相似文献
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A statistical theory for intergranular crack frequency as a function of crack size has been developed for polycrystalline materials. On the assumption that the cracks are penny-shaped and that brittle fracture is caused by overstressing the weakest link, the crack frequency was converted into a statistical theory of fracture. The implications of this theory were examined and compared with those of Weibull and of Fisher, Holloman and McClintock. In general, the predicted behavior is in reasonable agreement with experimental observations, although whether agreement is significantly better than Weibull's theory will require detailed examination. Among the attractive features of the new theory are: (1) a closer relationship to fracture mechanics and to material structure and properties compared with the Weibull theory and (2) a capacity for refinement so as to model the actual material structure more closely, which should result in improved agreement with experiment. 相似文献
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To investigate the crack growth and crack arrest behaviour of primary circuit materials large scale experiments were conducted on component-like specimens under pressurized thermal shock loading at MPA Stuttgart. The material characteristics varied from high tough material to low tough material with higher nil ductility transition temperature to simulate EOL or beyond EOL-state. All tests started from in-service conditions and were cooled down to room temperature. The specimens showed both stable and unstable crack growth and partly crack arrest. The crack growth behaviour was verified by post test calculations and could be explained with the help of the multiaxiality of the stress state. 相似文献
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Application of material models and assessment of strains in a surgeline under cyclic thermal loading
Inelastic constitutive models in commercial finite element (FE) programs are examined with respect to their capability of describing cyclic thermal loading. Neither isotropic nor linear kinematic hardening alone gives correct answers. Therefore, a model with combined isotropic and kinematic hardening based on Chaboche is implemented in an FE program and validated by appropriate experiments. This model is applied to assess the safety of a piping loop in a nuclear power plant subjected to cyclic loading. The numerical simulation shows a purely elastic behaviour of the surgeline after two cycles (elastic shakedown) as indicated by previous strain gauge measurements. 相似文献
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The Second Report of the Marshall Study Group [1] reviewed the design of a pressure vessel to identify those regions which were most sensitive to the expected loading conditions. The design and loading conditions, including the pressure test, were also reviewed. A stress analysis data base of sufficient detail was established to permit a full fracture analysis to be carried out. After reviewing currently available methods for fracture assessment the CEGB R6 method was adopted. Selected normal and upset, emergency and fault transients were considered. Initiation defect sizes for normal and upset transients were generally large. In the beltline and nozzle shell course, for example, the minimum initiating height for an extended line defect was approximately 70 mm. For emergency and fault conditions, providing upper shelf material properties apply, the height to initiate semi-elliptical defects exceeds 25 mm and cracks substantially larger than this will extend by only small amounts in a ductile manner and remain stable. If during the large LOCA upper shelf material properties cannot be maintained semi-elliptic defects 6 to 10 mm high may propagate transversely parallel to the inside surface of the vessel. This may be limited by warm prestressing. 相似文献
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LOCA下具有表面裂纹的反应堆压力容器承压热冲击分析 总被引:1,自引:0,他引:1
失水事故(LOCA)瞬态下,具有半椭圆形表面裂纹的反应堆压力容器(RPV)承压热冲击(PTS)问题被研究。采用有限元方法计算瞬态过程的热-应力响应;采用影响函数法计算应力强度因子,分别对母材和堆焊层内的应力进行分解,从而解决了由于堆焊层存在造成的应力拟合困难带来的计算偏差。编制了相应的断裂分析程序,对LOCA下RPV的结构完整性进行了分析。结果表明,在研究的LOCA下,整个瞬态过程中RPV应力强度因子均未超过材料断裂韧性,压力容器结构安全。本文研究为RPV在PTS下的结构完整性评估提供理论指导。 相似文献