Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal–hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.     

An analytical study of the dynamics and stability of a spring loaded safety valve

Spring loaded self-actuating safety valves are employed as part of the overpressure protection systems in various industrial applications. In order to design and predict their performance it is necessary to study the dynamic behavior of the valve over a range of fluid and system conditions. A one-dimensional model has been developed to study the effects of different valve parameters such as the spring-mass characteristics, geometry of internal parts, adjustment ring settings, bellows etc. which influence the dynamic behavior and stability of the valve. Analytical results for steam flow conditions are presented to demonstrate the relative effects of these parameters on the valve opening time, maximum lift, blowdown (upstream pressure differential between the valve opening and closing) and any oscillations of the valve stem. If the valve is not properly backpressure compensated, it may become unstable as the stagnation pressure at the valve inlet decreases. Lowering of the guide adjustment ring position or raising the nozzle adjustment ring generally results in improved stability, shorter valve opening time, higher lift and longer blowdown. The effect of damping on the valve stability is also demonstrated. The model can be used to evaluate the design of safety valves and damping devices to eliminate unstable valve dynamic behavior.     

离心式上充泵首级叶轮空化特性与试验研究

为研究和改善核电站离心式上充泵首级叶轮空化性能,采用数值模拟方法进行优化分析。将叶片数改为4片,研究了泵的最佳空化性能、扬程和效率。结果表明,最大流量工况点扬程模拟值与试验值的相对误差为2.9%,空化余量相对误差为3.6%,试验结果和模拟结果相吻合。将空化细分为初生空化、发展空化、临界空化、严重空化和断裂空化5个阶段,分析表明:初生空化时汽泡首先出现在叶片进口背面处,临界空化状态以后叶片工作面也开始出现汽泡;在发展空化到严重空化状态之间,空化和叶轮蜗壳动静干涉共同影响叶轮内的压力脉动规律;严重空化状态之后,空化成为主要影响因素,压力脉动变得相对稳定,叶轮进口和中部的压力脉动幅值明显减小,但叶轮出口处仍然保持较高幅值且比较规律的压力脉动。     

不同初因事件下的两相排放过程热工水力特性研究

本文针对稳压器安全阀开启后的复杂两相热工水力过程进行研究,确定不同初因事件下的稳压器安全阀两相排放特性。采用自主化系统分析程序ARSAC对稳压器安全阀的上下游进行建模分析,选取三种典型的阀门排放过程,包括稳压器安全阀误开启事故、导致一个或多个稳压器安全阀开启的主蒸汽流量完全丧失事故、以及低温超压保护条件下导致的稳压器安全阀间歇性开启的安注泵误启动事故,研究稳压器安全阀开启后水封及蒸汽（或水）排放过程中涉及的复杂两相热工水力特性,结果表明：ARSAC程序能够捕捉两相排放过程中管道内部的流型变化;水封通过下游管道会形成明显的流量峰值,且不同的上游初始条件下排放过程对于下游管道造成的流量峰值及时间特性不同。通过本文的研究可以为载荷分析、安全评价及设计优化提供指导性建议。     

An analysis of check valve performance characteristics based on valve design

Although it had been theorized by nuclear industry valve experts that the two most significant factors in assessing check valve performance were valve type (or design) and operating conditions, until recently, no data was available to support their assumptions. In co-operation with the Nuclear Industry Check Valve Group (NIC), Oak Ridge National Laboratory (ORNL) undertook a review and analysis of check valve failures recorded in the Institute of Nuclear Power Operations’ (INPO) Nuclear Plant Reliability Data System (NPRDS). This study involved the characterization of failures according to several parameters, including valve design (e.g. swing check, lift check). Since the valve design is not inherently included within the NPRDS engineering record for each component in the database, ORNL relied on input from NIC, valve manufacturers and catalogs to supply the missing information. As a result, nearly 60% of the 21 000 check valves listed in the NPRDS component database and 85% of the 838 failures occurring during 1991–1992 were identified according to valve design. This data provided the basis to perform previously unavailable cross-correlations between parameters such as valve design versus failure mode, valve design versus failure discovery method, population/failure distributions by valve design, etc. Performance assessments and predictions based on more specific sets of parameters (as opposed to generic check valve failure rates obtained from standard reference sources that generally ignore the valve design) should result in a significant impact on future nuclear plant operations, including inservice testing (IST) practices, maintenance, and probabilistic risk assessments (PRAs) by providing a means to calculate more appropriate relative (and ultimately absolute) failure rates for check valves.     

Development of diagnosis algorithm for the check valve with spectral estimations and neural network models using acoustic signals

We have developed a method for detecting and diagnosing a disk wear failure and a foreign object failure among the various failure modes of check valves. The method is based on the acoustic emission sensors which can detect the sound wave of the leakage flow and the estimation of the power spectral densities with an auto-regressive model. For validating the method, we implemented a hydraulic test loop with an artificially failed check valve. We have found that the frequency spectrums from the acoustic signals are strongly dependent on the failure modes of the check valve and that they are nearly independent of the failure size and operating pressure through an estimation of the power spectral density with an auto-regressive signal processing model. In addition, the root mean square values of the acoustic signal and the amplitudes of the power spectral density as well as the loop pressure have a strong dependency on the failure size in each failure mode of the check valve. We developed a diagnosis algorithm by using neural network models in order to identify the type and size of the failure in the check valve. The diagnosis algorithm consists of a hierarchical model composed of three back-propagation neural networks. The results of our research and the experiments show that the diagnosis algorithm is proven to be a good solution for identifying the failures of the check valves without any disassembling work.     

Mixture loss coefficient of safety valves used in nuclear plants

Theoretical and experimental investigations on the loss coefficient of gas–liquid mixture across safety relief valves have been carried out. Experiments were performed for three different types of safety valves and under different flow conditions. Using the Darcy equation and based on the presented experimental results, a new empirical correlation has been developed to calculate the loss coefficient and hence pressure loss. By consideration of flow contraction, high viscous fluids, Reynolds number and safety valve geometry, the model includes therefore the relevant primary influencing parameters. The reproductive accuracy of the proposed model and the statistical comparison, based on about 2000 measured data in the literature, demonstrated that the proposed model is the best overall agreement with the data. The standard deviation of the data is less than 27%. The model fits the data well and is sufficiently accurate for engineering purposes. The reported results of the tested safety relief valve are very important to improve the practical and safety design of the nuclear plants.     

核电厂一回路压力边界止回阀在线密封性能测试

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.     

Wear in check valves due to flow induced motion

Experiments showing the frequency and amplitude of the flow induced motion of the gate for a 2- and a 4-in. swing check valve have been performed. The gate motion is due to turbulence in approach flow. We have found the dominant turbulent frequency of the approach flow is about half the natural frequency of the valves. The valves appear to be almost critically damped. Because of this, the valves respond almost as they would to a static force of the magnitude characteristic of the turbulent fluctuation in the flow. Both the dimensionless exciting force and the damping ratio have been found to be independent of valve size so the above statements are true for larger valves also. The recommended valve oscillation amplitudes and frequencies are used to calculate the wear at the shaft and at the stop. For an unpegged check valve, such as one of the 10-in. valves which was used at the San Onofre Nuclear Generation Station, it was found that shaft bearing wear would amount to 0.27 in.³/year and stop wear to 0.03 in.³/year.     

Best Estimated and Conservative Analyses for a Feedwater Pipe Break Accident of an Integral Type Reactor

An integral type reactor, which is an innovative design to achieve a high degree of safety, is currently being developed at the Korea Atomic Energy Research Institute. A feedwater pipe break accident is one of the most important accidents regarding the safety of an integral type reactor. A best estimated calculation, a conservative calculation, and a parameter study for a feedwater pipe break have been carried out. The sensitivity analysis in this paper performed is to establish the parameters which greatly affect the feedwater pipe break accident. A power level, an initial system pressure, a moderator reactivity coefficient and a break size are the major parameters which maximize a system pressure. The important function that must operate following a feedwater pipe break accident is an opening of the pilot operated safety relief valves, and an initiation of the passive residual heat removal system. The integral reactor safety systems function properly and thus secure the reactor to a safe condition with respect to the safety parameters.     

Design and modification of steam generator safety system of FBR MONJU

MONJU is a prototype fast breeder reactor (FBR) in Japan. The sodium–water reaction in the steam generator (SG) is one of the important safety assessment items for a sodium cooled reactor like MONJU. MONJU is equipped with hydrogen gas detectors for the small water leak detection, gas pressure gauges for the medium leak and sensors of rupture discs for the large leak. As a design basis accident, one tube failure then failure propagation of neighboring three tubes is assessed to verify the structural integrity of the secondary components. A latest evaluation method on the design margin against the overheating tube rupture showed that the present SG system had not an enough margin in the worst case. For improving the margin, it needs to shorten the time of the sodium–water reaction by earlier water leak detection in the SG and sooner water ejection from the SG tubes. Therefore, MONJU is now carrying out the following modification works: (1) addition of steam relief valves, (2) addition of a gas pressure gauge with changing the interlock logic and lowering the trigger level, (3) reducing the opening of the valves on the SG gas flow line to the dump tank because of earlier detection for the pressure rise. After this modification, the design margin of the SG system will be sufficiently improved.     

对改善弹簧直接作用式安全阀密封特性的探讨

由于弹簧直接作用式安全阀的密封比压力是系统工作压力的递减函数，所以安全阀在正常工作压力下的前期泄漏－直未能克服。本文提出了一种新的密封原理。根据该原理，弹簧直接作用式安全阀的密封比压力变成了系统工作压力的阶跃函数，可改善安全阀的密封特性。     

Cavitation model of choked flow of underheated water

A necessary condition for cavitation to appear in fluid flow is a local drop of the pressure below the saturation pressure at the flow temperature. Such a pressure drop can be due to centrifugal acceleration or more accurately the centrifugal force engendered by this acceleration and directed along the radius of curvature of a convex surface bounding the flow, across the flow direction. A mathematical model of cavitation is constructed on this basis in the form of a dependence of the critical flow rate on the pressure, density, and underheating of water at the entrance into a hydraulic apparatus and concretized for nozzles and Venturi tubes with a smooth entry as well as for a stop valve in a RBMK process channel. This model is used to evaluate the state of the flow in the valve in different RBMK-1000 operating regimes.     

大压降管路节流特性分析及孔板优化设计

大压降孔板节流管路面临汽蚀导致的高频振动和孔板间流速过大导致的低频振动这两方面的危害。针对核电厂容积和硼控制系统典型大压降节流管路的振动现象,基于CFD方法分析了单级孔板节流管路中压降、速度、流线、涡流等关键水力特性,发现单级孔板下游产生负压区而发生汽蚀,且因孔板射流导致局部速度过大而形成涡流。采用阻塞压差评估了多级同心孔板的节流性能。相比于单级孔板,多级同心孔板的汽蚀危害得到了较大改善,但最后一级孔板仍存在过度节流的风险。按多级孔板节流压降几何级数递降的原则设计的渐扩型五级孔板可消除汽蚀的发生,但一级孔板压降过大导致其下游流速过大。综合考虑汽蚀特性和流速分布而设计的多级偏心孔板结构既能规避汽蚀危害,又能最大程度降低流速过大引发的管路低频振动,且增大孔板间距可提高上游孔板的节流能力,增加下游孔板的汽蚀裕度,可作为大压降孔板节流管路振动综合治理的优化设计方案。     

百万千瓦级压水堆严重事故卸压阀高温瞬态分析

由于核电厂严重事故的恶劣工况,在卸压过程中严重事故卸压阀门可能会经历阀门无法承受的高温瞬态而导致不可用。本文在可能导致高压熔堆的事故序列中筛选出具有一定的包络性并包含各种典型严重事故现象的典型严重事故序列。针对该事故序列考虑严重事故管理中的开阀时间范围开展了高温瞬态计算,并针对重要的影响因素阀门开启时刻的稳压器水位开展分析。最终确定了百万千瓦级核电厂具备典型性及一定包络性的严重事故卸压阀工作条件,并得到了阀门开启前后阀门可能经历的最高流体温度及流体温度变化曲线,为严重事故卸压阀门的设备鉴定及功能应用提供了重要基础。      

核一级低压差旋启式止回阀研制

为了实现核电站关键阀门的国产化,研制了核一级低压差旋启式止回阀。本文介绍了该阀门的技术参数及其研制过程和型式试验的情况。经过各种测试,各种性能指标均达到设计要求。     

Numerical design of a 20 MW lead–bismuth spallation target for an accelerator-driven system

A spallation target system is a key component to be developed for an accelerator-driven system (ADS). It is known that a 15–25 MW spallation target is required for the practical size of an ADS. Although there have been some design studies for small power spallation targets, that is, less than 10 MW, designs of high power target systems for ADS are relatively rare. The design of a 20 MW spallation target is very challenging because more than 60% of the beam power is deposited as heat in a small volume of the target system. In the present work, a numerical design study was performed to get optimal design parameters for a 20 MW spallation target for a 1000 MW ADS. The cylindrical beam tube and the hemispherical beam window were adopted in the basic target design concept with 1 GeV proton energy, and the thermal-hydraulic and the structural analyses were performed with the CFX and ANSYS codes. The beam window diameter and thickness were varied to find the optimal parameter set based on the design criteria: maximum lead–bismuth eutectic (LBE) temperature <500 °C, maximum beam window temperature <600 °C, maximum LBE velocity <2 m/s, and the maximum beam window stress <160 MPa. The results of the present study show that a 40 cm wide proton beam with a uniform beam profile should be adopted for the spallation target of 20 MW power. It was found that a 2.5 mm thick beam window is needed to sustain the mechanical load.     

采用典型模糊控制器实现压水堆稳压器的综合控制

对压水堆稳压器的压力和水位控制．提出了一种模糊综合控制方案。采用3个典型模糊控制器分别对电加热器、喷淋卸压阀和上充阀进行控制;在稳压器压力典型模糊控制器中采用了积分分离方法。本文对汽轮机负荷阶跃变化、线性变化、甩负荷3种工况进行了控制系统的仿真实验。结果表明,稳压器的压力以及水位的瞬态和稳态控制性能都得到了较大改善,明显优于GA-FC和PID控制方案。     

An investigation of the degradation characteristics for casting stainless steel, CF8M, under high temperatures

Stainless steel castings are used in pipes and valves subjected to high pressure and temperatures. The primary coolant system of a nuclear power plant is made of a stainless steel casting and the operating temperatures are in the range of 290–330°C. If the coolant system is exposed to these temperature ranges for a long period, it may be possible to experience degradation of the material. The present investigation is concerned with the degradation characteristics of CF8M (cast duplex stainless steel), exposed to the thermal and σ-phase degradation temperatures, 430 and 700°C, respectively. After the CF8M specimens are held 100–3600 h at 430°C for the thermally degraded specimens and maintained 20 min to 150 h at 700°C for the σ-phase degraded specimens, respectively, all specimens are water quenched. Each specimen of the thermally and σ-phase degraded materials is classified into five classes depending on the holding time at the given temperatures. In order to investigate the characteristics of degradation, microstructure, micro Vickers hardness, tensile, impact tests, and fatigue crack growth tests are performed for each class of the specimens. From the present investigation the following results were obtained: (1) the difference between the thermally and σ-phase degraded specimens can be distinguished through their microstructures, (2) hardness and tensile strength are increased with degradation, while elongation, reduction area, and impact energy are decreased by increasing the degradation, (3) the fatigue crack growth rate (FCG) of the σ-phase degradation at 700°C is larger than that of the thermally degraded specimens, and (4) the FCG for both thermally and σ-phase degraded specimens are larger than those of the virgin (nondegraded) specimens.     

Application of signature analysis for determining the operational readiness of motor-operated valves under blowdown test conditions

In support of the NRC-funded Nuclear Plant Aging Research (NPAR) program, Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of motor-operated valves (MOVs).As part of this work, ORNL participated in the gate valve flow interruption blowdown (GVFJB) tests carried out in Huntsville, Alabama, The tests provided an excellent opportunity to evaluate signature analysis methods for determining the operability of MOVs under accident conditions.ORNL acquired motor current and torque switch shaft angular position signauresnon two test MOVs during several GVFIB tests. The reduction in operating “margin” of both MOVs due to the presence of additional value running loads imposed by high flow was clearly observed in motor current and troque switch angular signatures. In addition, the effects of differential pressure, fluid temperature, and line voltage on MOV operations were observed and more clearly understood as a result of utilizing the signature analysis techniques.