核电站用余热排出泵电机LOCA试验

重点介绍余热排出泵电机LOCA试验的过程及要求,并对试验出现的异常情况进行探讨。余热排出泵电机是安装在安全壳内重要的核安全级电气设备,在其一系列的质量鉴定试验过程中,LOCA试验是最为重要的试验之一,通过模拟发生冷却剂丧失事故时安全壳内的热工和化学环境,考验余热排出泵电机在LOCA事故期间以及事故后是否能够执行其原有功能。     

灰浆泵的合理选用

文中介绍了在选用灰浆泵时如何确定泵运行工况点，以及泵特性曲线形状，串联运行，净扬程，管道结垢等因素引起对泵运行工况点发生变化的影响，其结合实例，对《火力发电厂除灰设计技术规定》中有关选泵依据的具体条文提出了不同的看法，指出：按《技规》选用的泵，对实际装置扬程会出现计算和选泵的双重富裕量，将导致泵运行工况点向大流量方向偏移，作者建议，可直接采用中文中公式（4）的计算值进行选泵，同时还应尽可能地采用变速转速调节流量法来稳定泵运行工况点。     

灰浆泵的合理选用

文中介绍了在选用灰浆泵时如何确定泵运行工况点,以及泵特性由线形状、串联运行、净扬程、管道结垢等因素引起对泵运行工况点发生变化的影响,并结合实例,对《火力发电厂除灰设计技术规定》中有关选泵依据的具体条文提出了不同的看法。指出:按《技规》选用的泵,对实际装置扬程会出现计算和选泵的双重富裕量,将导致泵运行工况点向大流量方向偏移。作者建议,可直接采用文中公式(4)的计算值进行选泵,同时还应尽可能地采用变转速调节流量法来稳定泵运行工况点。     

核电设备特种电机用先进技术

核能装置和核能发电设备反应堆冷却剂循环用主泵电机机组装在安全壳内,用于驱动。由反应堆、主管道、蒸汽发生器所组成的密封系统中的冷却剂进行循环,以便将反应堆产生的热量传给2次回路。     

秦山核电厂二回路系统设计的安全考虑

鉴于核电厂反应堆堆芯中积聚着大量放射性物质,故在核电厂工程设计中,凡属与安全停堆、堆芯冷却、以及余热排出等有关的安全功能,都被认为是最重要的安全功能,在进行工程设计时,必须对有关系统进行安全分析,以便确定各工程项目的设计基准和相应的技术措施,保证该电厂的安全运行。核电厂的二回路系统,虽然与核安全功能没有直接相关,但必须能适应核电厂一回路系统的各种不同工况,包括正常运行、工况扰     

AP1000核电站额定功率运行时单台给水泵跳闸瞬态分析

AP1000核电站额定功率运行时采用3台33.3%容量电动泵提供给水,无备用泵。通过对100%和70%额定功率平台下给水流量和蒸汽流量的稳态平衡计算,验证了1台泵跳闸后,剩余2台泵能维持电厂70%额定功率运行;并通过对CENTS程序建立的单台给水泵跳闸瞬态的仿真结果数据进行分析,验证了此瞬态下无需快速降功率动作,反应堆功率自动控制系统与蒸汽旁排系统能够将一回路平均温度维持在允许范围内,期间不会发生蒸汽发生器窄量程液位低跳堆事故,上述分析结果对 AP1000核电站调试和运行具有重要参考意义。     

凝结水泵低水位运行方式的改进

1　低水位运行方式的弊端在小型机组中 ,热水井的液位控制一直采用低水位运行方式 ,或利用出口管路调节阀进行控制。在低水位运行方式中 ,泵以近 30 0 0ｒ/ｍｉｎ的转速定速运行 ,出口阀全开。泵的流量大于需要抽送的凝结水流量 ,热水井的液位大于泵入口灌注高度时 ,水位很快下降到低于泵灌注高度处的水位 ,此时 ,泵无法再抽出水 ,而继续旋转的叶轮使泵内凝结水很快汽化 ,阻止了凝结水输送。一段时间后热水井液位由于机组不断凝汽而上涨 ,很快会高于泵入口灌注高度 ,泵入口汽化汽体亦会被抽空气管吸走 ,这样泵又可以抽水了。由此 ,形成了泵…     

超临界水冷堆与WWER1000型压水堆的安全特性比较分析

超临界水冷堆(SCWR)是在高于水的临界点(374℃,22.1 MPa)的温度和压力下运行的反应堆。它的设计为一次通过循环,其中没有再循环回路。这点是与现在运行的轻水堆的最大不同。在超临界水堆电站系统中,以控制棒、汽轮机控制阀与反应堆冷却剂泵为主要的控制方式。通过对比分析超临界水冷堆与田湾核电站WWER1000型压水堆主泵卡轴事故下的安全特性,得出超临界水堆给水流量的丧失会造成反应堆冷却剂流量的丧失,而WWER1000型压水堆给水流量的丧失并不会造成反应堆冷却剂流量的丧失;WW-ER1000型压水堆的安全系统有控制棒、蒸汽发生器的主蒸汽旁排阀、应急给水泵,这些安全配置与超临界水冷堆相似;相比WWER1000型压水堆,超临界水冷堆在压力较快达到稳定状态前提下,其最高包壳温度有个剧烈变化过程,但超临界水冷堆和WWER1000型压水堆在卡轴事故发生后,都能建立稳定的自然循环。     

核电厂余热排出系统安全阀控制逻辑改进研究

M310核电机组余热排出系统安全阀采用法国WEIR公司的SEBIM先导式安全阀,是核电厂中重要的阀门.目前M310压水堆核电站核反应堆一回路压力大于2.4 MPa(一回路压力低低信号未触发)时,会触发余热排出系统安全阀关闭命令,但是主控室操纵员在执行相关试验或特殊工况下在主控通过安全阀开启按钮将安全阀开启后,无法通过关闭按钮将安全阀再次关闭.通过改进余热排出系统安全阀控制逻辑解决此问题.     

热管式空气预热器的变工况预测模型

针对热管空气预热器在非设计工况下运行会出现低温腐蚀、超温失效及烟风阻力过大等问题,提出了一种定量分析热管安全性的方法。通过计算热管空气预热器变工况运行时排烟温度的变化来评价换热器的运行情况,进行了变工况校核热力计算,推导出预测和计算排烟温度的模型,利用模型可以分析入口烟温、燃料量和过量空气系数对热管换热器工作性能的影响程度。该方法为热管空气预热器的合理设计和安全经济运行提供了参考。     

Some methods for achieving more efficient performance of fuel assemblies

More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.     

Verification of the computer code KORSAR taking into account the effect of nondensables on thermal-hydraulic processes

Calculations to verify the Russian computer code KORSAR were carried out for the B4.1 experimental operating conditions, in which nitrogen was supplied to the reactor coolant (primary) circuit of a reactor plant model, and which were simulated at the PKL III integral test facility. It is shown that dissolution of gases in coolant has an essential effect on the thermal-hydraulic processes during long-term passive removal of heat from the primary to secondary coolant circuit of the reactor plant model under the conditions of natural circulation.     

核主泵轴密封节流管标定技术

节流管降压的准确值直接决定核主泵机械密封的运行条件,若降压后的压力过高会使轴密封失效,将导致机械密封泄漏量增大,影响核主泵的安全运行。为了实现压降值的精确测量,本文通过对沿程损失公式进行推导得出Q/ΔP~(1/2)与Q/u×10~5的近似线性关系,并应用于节流管标定试验,标定结果完全满足规定温度和流量下的降压要求。通过与理论计算长度比较确定,由于节流管表面粗糙度不同,实际标定长度与理论长度存在长度差,表明节流管的理论设计必须经过试验验证才能应用于工程。     

Fast reactor: an experimental study of thermohydraulic processes in different operating regimes

Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.     

核主泵电动机绝缘技术研究与发展

核主泵在反应堆核岛内运行,始终受到核岛内高能射线的辐射作用,因此,电动机绝缘系统承受高能射线的水平,是核主泵电动机绝缘系统设计的必要条件。绝缘结构的快速热老化评定试验是绝缘系统的耐热等级和热寿命的科学评定方法,为核主泵电动机绝缘系统的可靠运行提供了科学的依据。     

The high-temperature sodium coolant technology in nuclear power installations for hydrogen power engineering

In the case of using high-temperature sodium-cooled nuclear power installations for obtaining hydrogen and for other innovative applications (gasification and fluidization of coal, deep petroleum refining, conversion of biomass into liquid fuel, in the chemical industry, metallurgy, food industry, etc.), the sources of hydrogen that enters from the reactor plant tertiary coolant circuit into its secondary coolant circuit have intensity two or three orders of magnitude higher than that of hydrogen sources at a nuclear power plant (NPP) equipped with a BN-600 reactor. Fundamentally new process solutions are proposed for such conditions. The main prerequisite for implementing them is that the hydrogen concentration in sodium coolant is a factor of 100–1000 higher than it is in modern NPPs taken in combination with removal of hydrogen from sodium by subjecting it to vacuum through membranes made of vanadium or niobium. Numerical investigations carried out using a diffusion model showed that, by varying such parameters as fuel rod cladding material, its thickness, and time of operation in developing the fuel rods for high-temperature nuclear power installations (HT NPIs) it is possible to exclude ingress of cesium into sodium through the sealed fuel rod cladding. However, if the fuel rod cladding loses its tightness, operation of the HT NPI with cesium in the sodium will be unavoidable. Under such conditions, measures must be taken for deeply purifying sodium from cesium in order to minimize the diffusion of cesium into the structural materials.     

The condensation of steam from steam-water mixture on water jets at high pressure

A physical model for condensation of steam in water flow at high pressure is developed, and analytical dependences for calculating heat transfer are obtained, in particular as applied to the operation of a direct-contact feedwater heater for a new-generation reactor plant with lead coolant.     

Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor

Numerical simulation of the ULOF-type accident development in a fast reactor with sodium coolant performed using the COREMELT code indicates that sodium boiling in the active core takes place. The boiling is accompanied by oscillations of the technological parameters of the reactor installation; these oscillations can go on during several tens of seconds. In this case, it is possible that a stable regime of removal of heat from residual energy release is implemented. The model of the two-phase coolant flow applied in the code has an important effect on the numerical results; that is why this model needs experimental verification. For eliminating the development of an accident resulting in destruction of the active core elements, a structural solution is proposed; the essence of it is the application of the sodium void above the reactor active core. The experimental installation was developed and the heat exchange at sodium boiling in the model fuel assembly of the fast reactor in the regimes of natural and forced circulation in the presence of the sodium void and the top end shield was studied. It was demonstrated that, in the presence of the sodium void, it is possible to provide long-term cooling of the fuel assembly for a thermal flux density on the fuel element simulator surface of up to 140 and 170 kW/m² in the natural and forced circulation modes, respectively. The obtained data are used for more precise determination of the numerical model of sodium boiling in the fuel assembly and verification of the COREMELT code.     

Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit’s purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.