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控制棒驱动机构在地震情况下保证及时落棒插入堆芯安全停堆,是中国实验快堆(CEFR)抗震安全的重要保证,是核安全局特别关注和重点审评的重要问题之一,同时也是CEFR取得装料许可证的必要条件之一。为了验证控制棒(包括安全棒、补偿一调节棒)驱动机构在地震情况下的落棒功能及落棒时间,中国原子能科学研究院中国实验快堆工程部委托核动力研究设计院完成了CEFR控制棒驱动机构抗震试验,快堆工程部做了技术上的论证和配合。 相似文献
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磁悬浮控制棒驱动线抗震试验研究 总被引:2,自引:1,他引:1
为验证设备的稳定性、可靠性以及在极端条件下的安全功能,在地震模拟振动台上,采用一组控制棒驱动线实物作为足尺模型,进行了控制棒驱动线的抗震试验研究。得到了不同的地震输入对控制棒驱动线落棒时间的影响;测量了运行安全地震(SL-1)、极限安全地震(SL-2)水平下,控制棒驱动线的加速度响应值和应变值;验证了不同工况下控制棒驱动线的安全功能。试验数据表明,该驱动线在运行基准地震(OBE)、安全停堆地震(SSE)工况下,均能保持结构的完整性,并能实现运行功能。 相似文献
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高温气冷堆控制棒驱动机构冷态落棒试验研究 总被引:5,自引:0,他引:5
高温气冷堆控制棒驱动机构是执行反应堆功率调节、紧急停堆的重要核安全设备,具有固有安全特性,当断电后,控制棒能够靠重力快速下落实现停堆。为验证控制棒驱动机构的可靠性,必须对其进行设计分析和试验验证。本文建立了全尺寸的冷态试验台架,并采用1:1的控制棒驱动机构样机,对高温气冷堆控制棒驱动机构的落棒功能进行了验证和分析。进行多次全行程及不同高度的落棒实验,验证落棒过程的稳定性,测定控制棒在冷态条件下的落棒时间、落棒速度,试验结果满足规定的限值。对落棒过程进行分析,建立落棒运动方程式,进而得到控制棒运行速度的解析解。理论及试验的结果符合较好,均表明本文研究的控制棒驱动机构落棒可靠具有固有安全特性,为商用高温气冷堆中的实践应用提供了理论依据。 相似文献
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《核科学与工程》2015,(3)
高温气冷堆控制棒驱动机构是执行反应堆功率调节、紧急停堆的重要核安全设备,具有固有安全特性,当断电后,控制棒能够靠重力快速下落实现停堆。为验证控制棒驱动机构的可靠性,必须对其进行设计分析和试验验证。本文建立了全尺寸的冷态试验台架,并采用1:1的控制棒驱动机构样机,对高温气冷堆控制棒驱动机构的落棒功能进行了验证和分析。进行了多次全行程及不同高度的落棒实验,验证了落棒过程的稳定性,测定了控制棒在冷态条件下的落棒时间、落棒速度,试验结果满足规定的限值。对落棒过程进行了分析,建立了落棒运动方程式,进而得到了控制棒运行速度的解析解。理论及试验的结果符合较好,均表明本文研究的控制棒驱动机构落棒可靠具有固有安全特性,为商用高温气冷堆中的实践应用提供了理论依据。 相似文献
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控制棒组件是快堆控制系统和安全保护系统的重要组成部分,快堆控制棒价值的准确求解至关重要。基于PASC?5程序的快堆少群均匀化群常数计算中使用直接体积均匀化方式,这会导致控制棒价值严重高估,必须对控制棒组件的非均匀效应进行修正。基于群常数修正的思路,本论研究了体积?通量权重、反应率之比守恒和反应性守恒3种方法在快堆控制棒组件非均匀效应修正中的应用;基于二维特征线程序开发了群常数修正因子计算程序FRHP。通过中国实验快堆算例进行测试验证,修正后的控制棒价值计算结果与MCNP计算的参考结果符合较好,表明3种方法均能对控制棒组件的非均匀效应实现有效修正,其中反应性守恒方法修正效果最好。 相似文献
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Toshihiko Hirama Masashi Goto Minoru Kanechika Tsutomu Mieda Katsuki Takiguchi 《Nuclear Engineering and Design》2005,235(13):1335-1348
In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses. 相似文献
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Toshihiko Hirama Masashi Goto Toshiyasu Hasegawa Minoru Kanechika Takahiro Kei Tsutomu Mieda Hiroshi Abe Katsuki Takiguchi Hiroshi Akiyama 《Nuclear Engineering and Design》2005,235(13):1128
In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses. 相似文献
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Toshihiko Hirama Masashi Goto Hitoshi Kumagai Yukio Naito Atsushi Suzuki Hiroshi Abe Katsuki Takiguchi Hiroshi Akiyama 《Nuclear Engineering and Design》2007,237(11):1128-1139
In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. Test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses. 相似文献
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论文以国内某新建核电站控制室盘台抗震鉴定为例,阐述了基于有限元模型验证的盘台抗震鉴定方法。通过样机试验和模型验证分析,将盘台结构设计与有限元分析进行了有机结合,同时在盘台整体有限元分析验证过程中引入了修正因子,保证了盘台的抗震性能,并使其具有一定的安全裕度。 相似文献
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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
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The seismic qualification of equipment in operating nuclear plants has been identified as a potential safety concern in U.S. Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment in Operating Nuclear Power Plants”. In response to this concern, the Seismic Qualification Utility Group (SQUG), with support from the Electric Power Research Institute (EPRI), has undertaken a program to demonstrate the seismic adequacy of essential equipment by the use of actual experience with such equipment in plants which have undergone significant earthquakes and by the use of available test data for similar equipment. An important part of this program is the development of the methodology and test data for verifying the functionality of electrical relays used in essential circuits needed for plant shutdown during a seismic event. This paper describes the EPRI supported relay testing program to supplement existing relay test data. Many old relays which are used in safe shutdown systems of SQUG plants and for which seismic test data do not exist have been shake-table tested. The testing performed on these relays and the test results for two groups of relays are summarized in this paper. 相似文献
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利用试验和修正后的集中质量有限元模型预测安装在管道中阀门在不同频率成分地震激励下的响应,研究高频地震激励对管道中质量较大核级阀门的危害性。研究结果表明:高频地震激励对核级阀门的危害在于使阀门以其自身固有振型发生共振,此时阀门顶部取代阀门与管道连接位置成为阀门中响应最大的位置,这会导致安装于阀门顶端的驱动机构遭受苛刻的地震工况。增加管道阻尼和阀门刚度能有效降低高频激励对阀门的危害,但增加阀门刚度会导致管道响应增大。利用等效静力法对阀门进行抗震鉴定时,分析结果对阀体水平部位内力估计不足,对阀体垂直部分、阀盖等阀门上部构件的内力估计结果具有较大裕度。 相似文献
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The test and the updated lumped mass finite element model were used to predict the response of the valve installed in the pipeline under the seismic excitation of different frequency components, and the hazard of high frequency seismic excitation to large-mass nuclear safety class valves in the pipeline was studied. The results show that the high frequency seismic excitation causes the nuclear safety class valve to resonate with its own mode of vibration. At this moment, the top of the valve replaces the position where the valve is connected to the pipe to become the position with the largest response amplitude in the valve, which causes the drive mechanism installed on the top of the valve to suffer severe seismic conditions. Increasing the pipe damping and valve stiffness can effectively reduce the hazard of high frequency excitation to the valves, but increasing the valve stiffness will lead to the increase of the pipe response amplitude. When the equivalent static method is used for seismic identification of the valve, the analysis result is insufficient to estimate the internal force of the horizontal part of the valve body, and has a large margin to estimate the internal force of the vertical part of the valve body, the valve cover and other upper parts of the valve. 相似文献