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田湾核电站反应堆压力容器承压热冲击分析 总被引:1,自引:1,他引:0
反应堆压力容器(RPV)是核反应堆中不可替换的关键设备。田湾核电站在役前检查阶段,发现反应堆压力容器2#焊缝存在超标缺陷,2#焊缝处于堆芯筒体段,属强辐照区。为评价该缺陷的可接受性,采用有限元方法对反应堆压力容器2#焊缝进行了承压热冲击分析,在分析中考虑了小破口失水事故和安全阀误开启这两种最严酷工况。计算结果表明:有限元分析的结果与外国专家推荐方法的计算结果基本吻合,且田湾核电站反应堆压力容器2#焊缝寿期末的脆性转变温度小于最低容许脆性转变温度,能满足防脆断的设计要求。 相似文献
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反应堆压力容器(RPV)的辐照脆化问题是核安全的重中之重,影响到核电厂的安全性、经济性与公众信心。介绍了传统RPV辐照监督方案,讨论了现行技术的局限性,梳理了RPV辐照监督无损评估技术国外研究进展与存在问题,在实验与理论研究的基础上创新性地提出了中子辐照条件下RPV钢力学性能预测统一模型,并形成了基于电磁性能的RPV辐照监督无损评估技术,进一步完善后具有较好的工程应用前景。同时指出了开展RPV钢电磁性能实验研究,既有助于从一个全新的角度理解与再认识国产RPV钢长寿期服役时的辐照脆化行为,又有利于揭示RPV钢辐照脆化机理,丰富辐照脆化的基础理论。 相似文献
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结合田湾核电站2号机组反应堆压力容器(RPV)、秦山核电二期扩建工程4号机组RPV超标缺陷处理不符合项的安全审查,着重从RPV超标缺陷断裂力学安全评价的评价标准、缺陷特征化、断裂韧性、残余应力取值和承压热冲击等几个方面进行讨论,讨论结果对后续的此类不符合项的处理和安全评价有借鉴意义。 相似文献
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反应堆压力容器(RPV)是保障核电站运行安全性、经济性的核心构件。对RPV的完整性评估而言辐照脆化是必须面对的问题。我国已开发了第三代设计寿命为60 a的核电站。当达到寿期末时,辐照脆化的行为是未知的,这给国产RPV的辐照脆化预测带来了困难。为研究高注量下的辐照脆化行为,对A508-3钢的材料力学性能试样进行辐照考验,辐照温度为(288±8) ℃,中子注量水平达到反应堆压力容器60 a寿期末的辐照水平1×1020 cm-2;开展拉伸、冲击和断裂韧性试验,分析辐照脆化行为,在EONY模型基础上,提出针对国产RPV钢的改进的辐照脆化模型。模型的有效性被试验数据证实,其可准确预测国内A508-3材料的辐照脆化趋势。 相似文献
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对反应堆压力容器用Ni-Cr-Mo-V钢焊缝温度监督样品的热老化脆化行为进行了研究。焊缝属于压力容器的薄弱环节,服役时间最高达120 430 h(服役温度归一化到300 ℃)。3批次的焊缝监督样品冲击实验表明,焊缝材料在热老化过程中发生了脆化。通过研究发现,金相组织和显微维氏硬度在热老化期间未发生明显的变化,表明在热老化过程中不存在硬化脆化机制。断口分析及扫描俄歇纳米探针研究表明,晶界发生了P的偏析,弱化了晶界结合力,因此,反应堆压力容器用Ni-Cr-Mo-V钢焊缝在热老化过程中发生了由杂质元素P偏析引起的非硬化脆化。 相似文献
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The reactor pressure vessel (RPV) is the key component of pressurized water reactor. It has to comply with various rules and regulatory guides to ensure sufficient safety and operating margins during the plant lifetime. Thus, it is crucial to assure the integrity of RPV for an effective plant lifetime management program. In this paper, the status and the experiences of various integrity issues of the highly embrittled RPV are introduced. A circumferential weld in the beltline region of the Kori Unit 1 RPV was projected to be unable to satisfy the minimum upper-shelf energy requirement and the reference temperature-pressurized thermal shock requirement before the end of 40-year design lifetime. The detailed integrity assessments had been performed to resolve both issues and the results summarized. In addition several actions have been taken as aging management programs to assure the integrity of Kori Unit 1 RPV during the extended operation. Details of the activities such as, redefining initial reference temperature-nil ductility transition temperature, installing ex-vessel dosimetry, and withdrawal and testing of additional surveillance capsule are explained. Finally, the applicability of these and other activities including thermal annealing to mitigate the effects of the irradiation embrittlement are evaluated. 相似文献
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A. Ballesteros J. Bros L. Debarberis F. Sevini D. Erak S. Gezashchenko A. Kryukov Y. Shtrombakh S. Goloschapov Ionov Y. Pytkin Y. Anikeev G. Banyuk A. Plusch F. Gillemot L. Tatar V. Petrosyan 《Nuclear Engineering and Design》2005,235(2-4):411-419
The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level.The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals 2.7 × 1012 cm−2 s−1 with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%. 相似文献
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The most important effect of the degradation by radiation is the decrease in the ductility of the pressure vessel of the reactor (RPV) ferritic steels. The main way to determine the mechanical behaviour of the RPV steels is tensile and impact tests, from which the ductile to brittle transition temperature (DBTT) and its increase due to neutron irradiation can be calculated. These tests are destructive and regularly applied to surveillance specimens to assess the integrity of RPV. The possibility of applying validated non-destructive ageing monitoring techniques would however facilitate the surveillance of the materials that form the reactor vessel.The JRC-IE has developed two devices, focused on the measurement of the electrical properties to assess non-destructively the embrittlement state of materials. The first technique, called Seebeck and Thomson Effects on Aged Material (STEAM), is based on the measurement of the Seebeck coefficient, characteristic of the material and related to the microstructural changes induced by irradiation embrittlement. With the same aim the second technique, named Resistivity Effects on Aged Material (REAM), measures instead the resistivity of the material.The purpose of this research is to correlate the results of the impact tests, STEAM and REAM measurements with the change in the mechanical properties due to neutron irradiation. These results will make possible the improvement of such techniques based on the measurement of material electrical properties for their application to the irradiation embrittlement assessment. 相似文献
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T. Oku T. Kodaira K. Fukaya M. Suzuki Y. Tanaka T. Iwadate K. Suzuki I. Sato 《Nuclear Engineering and Design》1990,119(2-3)
The Japan Atomic Energy Research Institute (JAERI) has carried out a series of research and development work related to the high temperature gas-cooled reactor (HTGR) and, accordingly the high temperature engineering test reactor (HTTR) will be constructed in the near future. As the reactor pressure vessel (RPV) material,
Mo steel will be used. Material characterization tests have been carried out to evaluate the applicability of the
Mo steel for the RPV and to prepare for the licensing. The present paper summarizes the fracture toughness behavior including KId and KIa, irradiation embrittlement susceptibility and degradation of steel due to the long term aging at high temperature of the forged low
Mo steel. These tests reveal good fracture toughness which well meets the requirements of the ASME Code, low neutron irradiation embrittlement susceptibility, little embrittlement by long term aging and so on. The present test results demonstrate good applicability of forged low
Mo steel to the RPV of HTGR. 相似文献
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E. A. Kuleshova B. A. Gurovich Ya. I. Shtrombakh D. Yu. Erak O. V. Lavrenchuk 《Journal of Nuclear Materials》2002,300(2-3):127-140
Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel. 相似文献
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The surveillance test results of the reactor pressure vessels (RPV) of three Russian WWER-1000 units designated unit-1, -2 and -3 are given and the embrittlement rates compared to those predicted by the Russian Regulatory Guide. Dependence of the radiation behavior of WWER-1000 type RPV steels on metallurgical variables and the damage dose is considered. The trend curves for the steels under investigation are proposed. 相似文献
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反应堆压力容器老化敏感性分析方法 总被引:1,自引:0,他引:1
结合近期开展的大亚湾反应堆压力容器老化分析及大纲编写工作,归纳总结了反应堆压力容器老化敏感性分析方法,提出了较为明确的表单化的老化分析流程,可以为相关的老化分析与评价活动提供借鉴. 相似文献
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The reactor pressure vessel (RPV) of the HTTR is 5.5 m (inside diameter), 13.2 m (inside height), and 122 mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr–1Mo steel is chosen as the material for RPV. The temperature reaches about 400 °C at normal operation. The fluence of the RPV is estimated to be less than 1 × 1017 n/cm2 (E > 1 MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr–1Mo steel for the RPV; embrittlement parameters, J-factor and are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr–1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and , and nil-ductility transition temperatures, TNDT, by drop weight tests, are shown. In-service inspection and results of R&Ds are also described. 相似文献