共查询到20条相似文献,搜索用时 140 毫秒
1.
2.
3.
供热堆控制棒驱动线试验模型动力分析 总被引:1,自引:1,他引:0
对200MW低温核供热堆的控制棒及其相关部件(包括燃料组件等)试验模型进行抗震分析,包括有限元模型的简化、建立模型、模态分析、响应谱分析及时程响应分析等。通过分析得到结构的基频以及不同因素对基频的影响;根据频谱响应分析和时程分析的结果,得到结构受到地震载荷冲击时的最大位移响应和加速度响应。 相似文献
4.
叙述了在压水堆核电站主冷却剂泵的研制中,对安全一级部件的主泵泵壳进行“光弹”和“电测”实验应力分析,旨在保证泵壳的结构设计合理,并确保压力边界的完整和安全。实验分别用几何相似的模型泵壳,在内压、自重、地震以及接管系统载荷作用下进 相似文献
5.
6.
7.
8.
建立三维非线性有限元模型,对反应堆结构进行抗冲击动力响应分析,克服了结构的间隙、接触、摩擦、阻尼、预紧、碰撞、流固耦合、连接刚度等非线性因素;对于吊篮与压力容器间的流固耦合作用,建立水动力质量矩阵,并采用ANSYS声单元验证其准确性;设置了三维的堆芯上下板,并建立了多组燃料组件模型,考虑其预紧、跳起、与围板的间隙及碰撞,并考虑水平与竖直方向的耦合,更加精确地模拟了反应堆结构动力响应;以3个方向的冲击加速度时程作为计算输入,得到了各部件的响应,为各部件应力分析及控制棒驱动线抗冲击试验提供输入。该方法为反应堆结构的三维动力响应分析提供借鉴。 相似文献
9.
10.
11.
压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值(PCT)最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1204 ℃的限值。 相似文献
12.
Shinobu Yoshimura Kei Kobayashi Hiroshi Akiba Satoru Suzuki Masao Ogino 《Journal of Nuclear Science and Technology》2013,50(4):546-567
This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value. 相似文献
13.
Reactor Coolant Pumps (RCPs) are very important to the safe operation of Nuclear Power Plants (NPPs), especially during the earthquake, which needs detailed seismic analysis of individual RCPs and the boundary conditions, for example, at the nozzles. In this paper, three-dimensional finite element model of Reactor Coolant System (RCS) is constructed from a systematic perspective to perform dynamic evaluation, in which the boundary conditions could be given. The seismic spectrum analysis with three orthotropic directions is performed to obtain the stress and displacement response, which shows that the maximum Tresca stress locates in the connection part of SG with RCP and the maximum displacement occurs at the surge line. Sensitivity analysis of spectrum input angle and stiffness of supports is performed, which may be useful to further design and analysis. Furthermore, direct integration method is used to perform time-history analysis, and the boundary conditions of RCP, the loads, acceleration and displacement at nozzles are obtained, which could support the detailed analysis of RCP components. Besides, the lumped mass model of RCS is also constructed to compare with three-dimensional finite element model, which means that for the complicated geometry the 3-D model is better than the lumped mass model. 相似文献
14.
A level 1 probabilistic risk assessment of the Experimental Breeder Reactor 11 has recently been completed. Seismic events are among the external initiating events included in the assessment. The analysis indicated that the reactor shutdown system had a high reliability of operation in response to internally initiated events. One of the major tasks within the seismic assessment concentrated on the ability to shut down the reactor under seismic conditions. A comprehensive analysis of the shutdown system, including the development of a finite element model of the reactor control rod drive system, has been used to estimate the system response when subjected to input seismic accelerations. The results indicate the control rod driven system has a high seismic capacity and that the overall reactor shutdown system is capable of maintaining its high reliability under seismic conditions. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure. 相似文献
15.
核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统(CVR1.0)的重要组成。 相似文献
16.
An application of the finite element method to a three-dimensional perforated plate structure is presented using a nested modeling technique. Stresses calculated by a general three-dimensional finite element computer program were compared with those obtained from a model test. The structure considered in this analysis is a plenum cover of a pressurized water reactor internal which is a circular perforated plate stiffened with welded cross ribs. This type of structure is common in reactor internals. The nested model analyses consist of two finite element models; one is for the overall structure model and the other for an isolated portion of the structure with refined grid system for more accurate stress calculation. The first model was analyzed to obtain the nodal displacements under the applied loads. Then the second model was run using the displacement boundary conditions obtained from the first model analysis. A fully instrumented Plexiglas model test was conducted to verify this method. Comparison between the test results and the calculated stresses from the second model analysis showed good agreement. 相似文献
17.
Yi-Hsiang Cheng Jong-Rong Wang Hao-Tzu Lin Chunkuan Shih 《Nuclear Engineering and Design》2009,239(11):2343-2348
The pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) nuclear power plants. An accurate modeling of the pressurizer is needed to determine the pressure response of the primary coolant system, and thus to successfully simulate overall PWR nuclear power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: (1) turbine trip test from 100% power (Test PAT-50); (2) large-load reduction at 100% power (Test PAT-49); (3) net-load trip at 100% power (Test PAT-51); and (4) net-load trip at 50% power (Test PAT-21). The simulation results show that the predictions of the pressure response are in reasonable agreement with the power plant's start-up tests, and thus the pressurizer model built in this study is successfully verified and validated. 相似文献
18.
19.
20.
燃料组件属I类抗震物项,其抗震问题直接关系核电厂运行安全,通常需通过抗震试验验证反应堆燃料组件抗震分析方法的合理性。本文模拟反应堆实际堆芯燃料组件安装方式,设计压水堆燃料组件抗震试验件与试验装置,针对不同组件数量布置方案,在高性能地震模拟振动台上开展试验研究。结果表明,水介质中燃料组件的第一阶频率为2.96 Hz,最大冲击力出现在燃料组件偏中间位置处,试验获取了地震作用下燃料组件的格架冲击力、格架相对位移、模拟堆芯板与围板的加速度等响应。试验结果可用于设计基准事故工况中燃料组件抗震分析模型的建立与分析软件的验证。 相似文献