INCONEL690镍基合金带极电渣堆焊工艺研究

通过试验研究,确定了SA-508Gr.3 CL.2管板采用INCONEL690镍基合金进行带极电渣堆焊的焊接工艺方案,各项试验数据均满足产品制造技术条件要求,说明INCONEL690镍基合金电渣堆焊工艺能满足第三代核电技术蒸汽发生器管板的堆焊要求.     

提高SA508—C13钢锻件冲击韧性研究

本文通过对SA508-C13钢锻件冲击韧性偏低的原因分析,提出采用双相区热处理可以提高SA508-C13钢锻件冲击韧性的试验研究。试验研究结果可供核电锻件材料制造和应用时参考。     

联合循环汽轮机高中压转子锻件的国产化研究

在引进、消化、吸收美国联合循环汽轮机高中压转子材料的基础上,针对国内大型转子锻件的实际生产情况进一步创新,采取3项技改措施,首次在国内研制出满足美国材料的要求并高于进口水平的联合循环汽轮机高中压转子锻件.对联合循环汽轮机高中压转子锻件成功国产化,不仅填补了国内空白,缩短转子锻件采购周期,降低机组的制造成本,同时也推动了燃气-蒸汽联合循环机组的国产化进程.     

三代核电常规岛设备制造质量控制

总结国内首台三代核电AP1000和EPR1700常规岛设备国产化制造质量控制经验,对核电常规岛产品制造质量保证体系的建立与实施进行介绍,并提出常规岛产品制造过程中的特殊质量控制要点.     

第3代核电(AP1000)关键设备工艺制造特点综述

AP1000作为一种第3代核电技术,相比于第2代压水堆,其核岛关键设备发生了许多明显的变化。然而,部分关键技术仍处于试制阶段,国内制造厂家缺乏相关的工艺技术和制造经验,给设备的国产化带来了巨大困难。通过对AP1000先进压水堆设备的制造特点进行分析,可以加深对第3代先进压水堆技术的了解,明确设备制造的难点以及国产化面临的一些问题,促进设备制造工作的顺利开展。     

世界首台套中小功率核电应急柴油机试验成功

正10月18日,由中国船舶工业集团公司研制的世界首台套中小功率核电用应急柴油机在沪东重机顺利完成了台架性能试验.这标志着我国成功突破了中小功率应急柴油机的设计、制造及试验技术,有力推动了我国核级柴油机重大设备国产化进程,同时也为我国第四代核电的市场化奠定了基础.该型柴油机为6PA6LN型柴油机,与国内外已建和在建主流堆型的额定功率6 300 kW柴油机相比有     

AP1000非能动余热排出热交换器管板堆焊技术

通过选用优良的焊接材料,制定合理的堆焊工艺,顺利完成了AP1000核岛部件非能动余热排出热交换器的管板堆焊,并获得了质量和性能优良的堆焊层。结果表明,采用两侧交替堆焊的方式,带极埋弧堆焊与焊条电弧焊结合的方法,既能最大限度的保证堆焊效率,又保证了管板堆焊后的各种性能和尺寸要求。为第三代核电AP1000技术在我国的推广应用提供了宝贵制造经验。     

秦山核电二期600 MW蒸汽发生器关键焊接技术

由美国西屋公司设计全面国产化制造的第一台秦山核电二期工程600 MW核电蒸汽发生器结构复杂,技术要求严格.简要介绍了蒸汽发生器的结构特点、功能、以及蒸汽发生器所使用的本体材料和焊接材料的成分和性能,重点介绍了蒸汽发生器关键接头的焊接及热处理的技术要点及相应的焊接工艺措施.通过对产品所涉及的焊接及热处理技术的介绍,为核电设备制造提供了一些工艺技术资料.     

核电站反应堆压力壳用钢断裂韧性要求及安全可靠性分析

本文介绍了美国ASME和法国RCC-M对核电站反应堆压力壳用SA533B-I互钢板和SA508- Ⅲ锻件断裂韧性的要求及各国生产的SA533B-I铜板、SA508-Ⅲ锻件和焊接接头的断裂韧性试验结果,并讨论了美国、法国和英国核电站反应堆压力壳安全可靠性分析方法.     

国产SA-302C材料及工艺试验研究

介绍了国产SA-302C材料性能,并通过一系列焊接性能试验和环焊缝焊接接头力学性能试验,试验结果表明国产SA-302C材料的性能与进口材料相当;可采用相同的焊接工艺进行焊接,为国产SA-302C材料替代进口打下了基础.     

Investigation on the possibility of using the microshear test as a surveillance method to estimate the mechanical properties and fracture toughness of nuclear pressure vessel steel,A508CL3, and its joints welded by narrow-gap submerged-arc welding

The present paper has investigated the mechanical properties of nuclear pressure vessel steel, A508CL3, and its welded joints by using the microshear test method, and the fracture toughness of A508CL3 steel and its welds has also been estimated. Moreover, a comparison has been carried out between the conventional test, microshear test and fracture mechanics test. In addition, the possibility of using the microshear test on the surveillance program of nuclear pressure vessel embrittlement due to neutron irradiation has also been considered in detail, and the results indicate that the microshear test can be used successfully to estimate the degradation of mechanical properties both for A508CL3 steel and its welded joints. It has been found that the lower the microshear toughness, the smaller the Charpy V-notch (CVN) impact toughness and fracture toughness, as well as the tearing modulus. Finally, the results show that the microshear test method may be developed as a supplemental test method or standard of ASTM E185 and E636.     

Radiation embrittlement modelling for reactor pressure vessel steels: I. Brittle fracture toughness prediction

Modelling for the irradiation effect on brittle fracture toughness of reactor pressure vessel (RPV) steel is performed on the basis of the probabilistic model for fracture toughness prediction proposed by the authors earlier. The irradiation effect on parameters controlling plastic deformation and brittle fracture of RPV steels is analyzed. The physical mechanisms are considered which control the cleavage microcrack nucleation for RPV steels in the unirradiated and irradiated states and also in state after post-irradiation annealing. Prediction of the temperature dependence of brittle fracture toughness is performed as applied to irradiated 2.5Cr–Mo–V reactor pressure vessel steel. Modelling of the fluence effect and the phosphorus and copper content effect on brittle fracture toughness is carried out. It is shown that the probabilistic model based on a new formulation for brittle fracture criterion allows the adequate modelling for the irradiation effect on fracture toughness for RPV steel. Application of alternative models is discussed for fracture toughness prediction for irradiated RPV steels.     

Measurement of through-the-thickness variations of mechanical properties in SA508 Gr.3 pressure vessel steels using ball indentation test technique

The through-the-thickness variations of mechanical properties in SA508 Gr.3 pressure vessel steels were measured using the automated ball indentation (ABI) test technique. Key mechanical properties, such as the yield strength, ultimate strength, flow curve and hardness, were evaluated from indentation load-depth curves. The mechanical properties measured were location-dependent and the steepest gradients in the distributions of the mechanical properties appeared in the near-surface regions. The maximum through-the-thickness variations of the mechanical properties were in the range of 5–20% and they depended on the manufacturing process as well as the original wall thickness. It was concluded that the through-the-thickness variations in the mechanical properties were mainly caused by the location-dependent cooling rate during water quenching in the quality heat treatment which consisted of water quenching and tempering.     

The fracture mechanical significance of cracks formed during stress-relief annealing of a submerged arc weldment in pressure vessel steel of type A508 Class 2

In large weldments of type A508 Cl2 cracks can form in the heat-affected zone during stress-relief annealing. The significance of such cracks with respect to catastrophic fracture is of interest from the point of view of safety, in particular for nuclear pressure vessels. In this investigation the size of reheat cracks, as formed and after fatigue growth, has been compared with the critical size for fast fracture. The latter was assessed by determination of the fracture toughness of the heat-affected zones. The fracture toughness of the heat-affected zones did not differ much from that of the parent material. The presence of microcracks reduced the fracture toughness (of a special type of simulated specimen) at 20°C by about 20%. The fracture mechanical evaluation indicates that the cracks formed during stress-relief annealing should not impair the safety of the vessel under normal conditions, except for particular geometries and when the cracks may rapidly link together during fatigue.     

Effects of geometry and flow properties on the fracture toughness of a C–Mn reactor pressure vessel steel in the lower shelf region

To estimate the fracture toughness of thick section nuclear reactor pressure vessel (RPV) steel in the irradiated condition, it is necessary to apply a size effect correction to the test results obtained on small-scale surveillance specimens. This correction is usually derived using toughness data obtained on different sizes of fracture mechanics testpieces made of non-irradiated material, for which the flow properties are quite different from those of material in the irradiated state. This paper describes the results of a fracture toughness test programme carried out on a C–Mn steel plate for two different specimen geometries (10 mm thickness precracked Charpy and 25 mm thickness compact tension) in the lower shelf region of the temperature/fracture–toughness curve. A comparison of the fracture behaviour and failure micromechanisms has been made for the material in the ‘start of life’ condition and after the application of cold prestraining, which was used to simulate the effects of neutron irradiation on flow properties during service. Although the Master Curve methodology predicts no size effects on the lower shelf, size effects were observed.     

Radiation embrittlement modelling for reactor pressure vessel steels: II. Ductile fracture toughness prediction

Modelling for the irradiation effect on ductile fracture toughness of reactor pressure vessel steels (RPV) is performed on the basis of ductile fracture criterion proposed earlier by the authors. The irradiation effect on mechanisms controlling ductile fracture is considered from a physical viewpoint. Modelling of the irradiation effect is carried out on the critical strain for smooth cylindrical specimens and on the local critical strain for cracked specimens. On the basis of the performed studies a scheme that allows an evaluation of the upper shelf level of the K_IC(T) curve for irradiated RPV steels is proposed.     

A comparison of Western and Eastern nuclear reactor pressure vessel steels

The task was essentially to compare the irradiation response of `East' and `West' steels. Since the plates and forgings of pressure vessels must be welded together, it is obvious that the strength requirements of the welds and heat affected zones (HAZ) can be no less demanding than those of the plates and the forgings themselves, particularly as experience has shown that the most likely location for flaws is in the welds or their HAZs. These and the highly stressed regions of the reactor pressure vessel (RPV) are important because neutron irradiation degrades the mechanical properties of steels.After comparing the various designs, manufacture and materials of the various RPVs, a comparison was made of the irradiation response of these different steels. The role of mitigating the change in mechanical properties on irradiation by thermal annealing was also considered.Particular codes/guides could only be used for the predicting results underpinning their own database because a major difference between these national codes/guides is that the elements conferring irradiation sensitivity are different for the two cases considered, i.e. Russian codes [1] (PNAE G-7-002-86) and the USNRC guide [2] (RG 1.99 Rev. 2). In the former, copper and phosphorus are significant, while copper and nickel are identified as significant in the latter case.Predictions were compared for `real' materials used in NPPPVs whose compositions were known. The irradiation response of these steels is coincidentally similar. The essential difference in behaviour is in the lifetime fluence. Eastern steels are irradiated to a much higher fluence than Western steels. Differences in the predictions of the Eastern–Western codes/guides are a reflection of differences in the concentration of deleterious elements and pessimisms of the various codes/guides, particularly at low concentrations of deleterious elements where they are most conservative. Thirdly, and on a `fitness for purpose' basis, the shift in transition temperature produces a limitation to the lifetime of the earlier Eastern RPVs. However, by thermally annealing the RPV to mitigate the effect of neutron irradiation, where the conditions to recover the mechanical properties of both Eastern and Western steels are nearly the same, the operational life of these older Eastern plants has been extended. Life assurance of these plants has, therefore, become practicable.This aspect of RPV technology, which is currently being considered in the US, could extend the operational life of nuclear power plants and thereby reduce the cost of the electricity generated.     

Tensile and fracture properties evaluation of PHT system piping material of PHWR

The aim of this paper is to report the tensile and fracture properties of SA333 Gr.6 carbon steel material which is used for the primary heat transport (PHT) system piping of the Indian pressurized heavy water reactor (PHWR). Tensile and J integral tests have been carried out on specimens machined from the base material as well as weldments of actual PHT pipes. The effects of test temperature and specimen orientation on the material properties have been discussed.     

Development of a statistical lower bound fracture toughness curve

A collection of mechanical property data has been used to develop reference lower bound fracture toughness curves. Many heats of nuclear pressure vessel steel (A533B-1, A508-1, A508-2, A540, A302B, A537-1 and A537-2), heat affected zone material, manual arc and submerged arc weld material were used in the study. Several referencing techniques were used to remove heat to heat variation for a particular material type. The currently employed technique, using RT_NDT, was found to inflate variance in some situations instead of reducing it. Alternative and more promising techniques were developed, the best one using the precracked instrumented Charpy impact test. First, mean curves relating the reference test to fracture toughness were developed. Second, the distribution of actual data about the curves was characterised in terms of variance and the form of distribution. Both variance and form of distribution changed with temperature. The results were used to generate a lower bound reference curve (a local tolerance bound), using the function:

$\text{f = A + B tanh}\text{t ? T}{}_0\text{C}$

($\text{f}$ is normalised fracture toughness and $\text{t}$ is normalised temperature).