共查询到20条相似文献,搜索用时 15 毫秒
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In this study, the method of characteristic line (MOC) was adopted to evaluate the valve-induced water hammer phenomena in a parallel pumps feedwater system (PPFS) during the alternate startup process of parallel pumps. Based on closed physical and mathematical equations supplied with reasonable boundary conditions, a code was developed to compute the transient phenomena including the pressure wave vibration, local flow velocity and slamming of the check valve disc, etc. Some interesting results were obtained and it was shown that severe slamming between the valve disc and valve seat occurred during the alternate startup of parallel pumps. The induced maximum pressure vibration amplitude is up to 5.0 MPa, which occurs under the high–high speed startup condition. The scheme of appending a damping torque with the check valve disc was also numerically performed to eliminate the water hammer for the optimum design purpose. The adoption of damping torque slows down the closing speed of the check valve and has been approved to be an effective approach. This work is expected to be instructive for the optimum design of the PPFS in NPPs so as to mitigate the potential damage caused by valve-induced water hammer. 相似文献
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压水堆核电站主泵工况变化时,由于止回阀阻止流体逆流,回路会发生程度不同的水击现象.水击严重时不但会产生瞬时超压,危害压力边界,也可能造成止回阀失效.对冲式止回阀是一种新原理止回阀,是为解决传统止回阀关闭时产生的严重水击现象而设计的.分析和实验表明对冲式止回阀能有效地解决回路的水击问题,也能可靠地阻止流体逆流. 相似文献
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对冲式止回阀局部流动特性仿真模拟 总被引:1,自引:0,他引:1
当压水堆核电站主泵工况变化时,回路会发生程度不同的水击现象,严重时不仅会产生瞬时超压危害压力边界,也可能造成止回阀失效。对冲式止回阀是一种新型止回阀,其新型的导流结构设计可很好地减轻水击现象,提高核电站运行的安全性。本工作基于FLUENT流体仿真软件,利用动网格及UDF(用户自定义函数)技术,更真实地模拟了阀门关闭过程中冲压管喉部及其联接腔体内的流场与压力分布。模拟结果表明,虽然几何形状的变化会导致关闭过程中局部流速高于其他部位的,但对阀门关闭过程的稳定性影响很小。另外,通过对流场分析发现,阀门关闭过程可分为3个阶段,每一阶段均有其独立特征,为以后阀门结构的改进及可靠性分析提供了依据。 相似文献
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基于耦合程序的流体瞬变流动水锤现象分析 总被引:1,自引:1,他引:0
水锤现象严重威胁系统的安全,而设备的启闭是产生水锤现象的重要因素之一。本文针对并联双泵系统建立耦合程序,计算研究泵启动和阀门关闭时的流体瞬变水锤现象。验证过程证明了耦合程序的正确性,并将三维稳态模型计算结果与实验结果进行了对比,二者符合良好。瞬态分析中,动网格技术成功模拟阀门关闭,并获得了闭合时阀内的重要热工水力参数。通过对比泵启动耦合计算结果与传统RELAP5计算结果可知,耦合程序能正确预测水锤压力波和水锤载荷。耦合分析较一维计算能更直观地展现系统中重要设备内的流体瞬变特性。计算获得的三维瞬态特性能对阀门的设计和优化提供重要的参考。 相似文献
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《Annals of Nuclear Energy》2007,34(1-2):1-12
Correct prediction of water hammer transients is of paramount importance for the safe operation of the plant. Therefore, verification of computer codes capability to simulate water hammer type transients is a very important issue at performing safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. In RBMK-type reactors the water hammers can occur in cases of rapid check valve operation. The performed analysis using RELAP5 code RBMK-1500 model has shown that in general the maximum values of the pressure pulses due to water hammer do not exceed the permissible loads on the pipelines. 相似文献
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This paper introduced the structure and working principle of a new designed check valve, Contra-push check valve (CPCV), which can release water hammer and valve slam in accidents and special working status of nuclear power systems. The steady and transient characteristics of CPCV are simulated by CFD codes. Based on the experimental data, it is shown that the result is highly dependent on the turbulence model. The renormalization group theory (RNG) k–ε model is proved to be more accurate to describe the flow inside the valve. Steady hydraulic characteristics computed with RNG k–ε model agreed well with the experimental data at different positions of the plug. The Sensitivity analysis of structure parameters of CPCV were carried out in this study and two key factors were revealed. 相似文献
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320 MW压水堆一回路压力边界止回阀为核Ⅰ级关键设备,严密性要求非常高,直接关系到主系统的内泄漏率.焊接式止回阀维修后常采用密封面色印检查的方式,对其密封性能进行判断.如果管道内有存水或者湿热水汽,会影响到色印检查的准确度.针对在线止回阀密封性试验的特殊性,有的核电厂采用水压压降法试验设计过在线检测装置,但存在一些缺点和使用上的限制.文章采用低压气密封试验流量测定法,设计出可靠、便携的试验装置,对压力边界止回阀检修后密封性做出准确、定量的判断. 相似文献
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大型非能动压水堆核电厂在发生失水事故(LOCA)后的长期堆芯冷却阶段依靠重力向堆芯注入应急冷却水,其注射管线上设置的旋启式止回阀的阻力可随流量变化,管线的阻力可能将非预期地增加。根据旋启式止回阀阻力特性,为失水事故最佳估算系统分析程序添加相应的计算功能,对压力容器直接注射(DVI)管线双端断裂事故后长期堆芯冷却工况进行了计算分析。结果表明:安全注射管线上旋启式止回阀阻力变化对大型非能动压水堆核电厂LOCA后长期冷却的影响较小;在安全裕量不足的情况下,旋启式止回阀的阻力特性将影响到非能动注射管线的安全注射功能的执行。 相似文献
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Measurements of an experiment in a pipe system with pump shutdown and valve closing have been performed in the nuclear power plant KRB II (Gundremmingen, Germany). Comparative calculations of fluid and structure including interaction show an excellent agreement with the measured results. Theory and implementation of the fluid structure interaction (FSI) and the results of the comparison are described. The following measurements have been compared with calculations: (1) experiments in Delft, Netherlands to analyse the FSI; and (2) experiment with pump shutdown and valve closing in the nuclear power plant KRB II has been performed. It turns out, that the consideration of the FSI is necessary for an exact calculation of ‘soft’ piping systems. It has significant application in current waterhammer problems. For example, water column closure, vapour collapse, check valve slamming continues to create waterhammers in the energy industry. An important consequence of the FSI is mostly a significant increase of the effective structural damping. This mitigates—so far in all KED’s calculations the FSI has taken into account—an amplification of pipe movements due to pressure waves in resonance with structural eigenvalues. To investigate the integrity of pipe systems pipe stresses are calculated. Taking FSI into account they are reduced by 10–40% in the actual case. 相似文献
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Check valves are used extensively in nuclear plant safety systems and balance-of-plant (BOP) systems. Their failures have resulted in significant maintenance efforts and, on occasion, have resulted in water hammer, overpressurization of low-pressure systems and damage to flow system components. Consequently, in recent years check valves have received considerable attention by the Nuclear Regulatory Commission (NRC) and the nuclear power industry. Oak Ridge National Laboratory (ORNL) is carrying out a comprehensive two phase aging assessment of check valves in support of the Nuclear Plant Aging Research (NPAR) program. As part of the second phase, ORNL is evaluating several developmental and/or commercially available check valve diagnostic monitoring methods; in particular, those based on measurements of acoustic emission, ultrasonics, and magnetic flux. These three methods were found to provide different (and complementary) diagnostic information. The combination of acoustic emission with either ultrasonic or magnetic flux monitoring yields a monitoring system that succeeds in providing sensitivity to detect all major check valve operating conditions. The three check valve monitoring methods described in this paper are still under development and are presently being tested as part of a program directed by the Nuclear Industry Check Valve Group (NIC) in conjunction with the Electric Power Research Institute (EPRI). Phase 1 of this program (water testing) is being carried out at the Utah Water Research Laboratory located on the Utah State University campus. 相似文献
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Werner Barten Audrius Jasiulevicius Annalisa Manera Rafael Macian-Juan Omar Zerkak 《Nuclear Engineering and Design》2008,238(4):1129-1145
The capabilities of the nuclear system transient codes TRACE and RELAP5 to model coupled two-phase flow and pressure wave propagations in a pipe are assessed by analyzing the UMSICHT PPP cavitation water hammer experiments 329 and 135 after valve closure. Time-dependent pressure, flow behaviour, and the generation and collapse of vapor bubbles at the valve and the first bridge are discussed. We show that both codes are able to model the flow behaviour of the water hammer for the high pressure and high temperature case 329 (initially 10–13 bar and 420 K), however condensation heat transfer for the base case needed to be increased in order to accurately model the magnitude of the first pressure excursion. The experimental broadening and damping of the subsequent pressure peaks by Fluid-Structure Interaction (FSI) phenomena arising from the interaction of the flow with the vibrations of the piping structure are not considered in the modeling results. For the lower pressure and temperature case 135 (initially 1–4 bar and 294 K), the TRACE code provides a good approximation of the propagation of the pressure wave and the void fraction behaviour, already with base case conditions, while RELAP5 overpredicts the vapor generation along the pipe and, as a result, considerably underpredicts the pressure amplitudes and overpredicts the water hammer frequency. 相似文献
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Shiro Takahashi Akinori Tamura Shunichi Sato Toshitaka Goto Michiaki Kurosaki Noriyuki Takamura 《Journal of Nuclear Science and Technology》2016,53(8):1164-1177
Some problems due to flow-induced vibrations related to closed side branch pipes have been observed in thermal and nuclear power plants. Fluctuating pressure generated in the main pipes was unusually, acoustically excited in closed side branch pipes, and intense vibrations were caused at pipes and components. For example, flow-excited acoustic resonance in closed side branches of stub pipes of safety relief valves caused the failure of steam dryers in the United States Quad City Unit 2 nuclear power plant. Furthermore, there was a possibility that residual air or gas in a closed side branch pipe unexpectedly caused severe vibrations of low frequency in the feed water piping system. We have investigated the root cause and influence of air on severe vibrations. Intense fluctuating pressure was often caused by water hammer due to valve closure and it became larger in the closed side branch pipes. We showed that an additional side branch with an orifice was very effective to suppress the flow-induced acoustic resonance. Design methods of the orifice to attenuate fluctuating pressure generated by water hammer were presented considering Mach number, the pressure loss coefficient of orifice and the intensity of particle velocity. Moreover, suitability of the characteristic curve method was confirmed for evaluation of the attenuation effect of an orifice on fluctuating pressure generated by water hammer. Finally, we considered some flow-induced vibration problems related to closed side branch pipes and their attenuation methods. 相似文献
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In the current design of the simplified boiling water reactor, the vacuum breaker check valve is an important safety component. The vacuum breaker check valve is the only key safety components which is not passive in nature. Failure of this mechanical valve drastically reduces the passive containment cooling system cooling capability and hence containment pressure may exceed the design pressure. To eliminate this problem novel vacuum breaker check valve was developed to replace the mechanical valve. This new design is based on a passive hydraulic head, which is fail-safe and is truly passive in operation. Moreover this new design needs only one additional tank and one set of piping each to the wetwell and drywell. This system is simple in design and hence is easy to maintain and to qualify for operation. The passive vacuum breaker check valve performance was first evaluated using RELAP5. Then the passive vacuum breaker check valve was constructed and implemented in the PUMA integral test facility. Its performance was studied in a large break loss of coolant accident simulation test performed in PUMA facility. 相似文献