核反应堆一回路系统水锤数值模拟

采用特征线法,针对核反应堆一回路系统特点建立完整的数学物理方程和边界条件。自主开发WAHAP水锤计算程序,并对含止回阀的4泵并联系统在启动及切换工况下的水锤特性进行模拟计算。计算结果表明,所研究工况中有2个不同阶段存在阀瓣多次剧烈撞击阀座的过程和2种形式的压差震荡。     

双泵并联给水系统切换方式模拟研究

建立双泵并联给水系统模型,对双泵切换过程中三种不同切换方式进行模拟研究,第一种切换方式为先启动备用泵后关闭运行泵、第二种方式为先关闭运行泵后启动备用泵、第三种方式为同时进行操作.通过试验验证了模型的准确性.模拟结果表明:第一种切换方式条件下系统流量失水量最小,第三种切换方式条件下系统流量波动最小;相同切换方式条件下,双泵切换时间间隔越短,系统流量波动越小,系统恢复稳定时间越短.在实际情况中,要尽量缩短双泵切换时间间隔.     

双泵切换过程模拟及灵敏度分析

对双泵并联给水系统中泵切换过程进行了试验与模拟研究,通过试验结果验证了模型的正确性。利用模型对试验过程进行灵敏度分析,得出以下结论：双泵并联给水系统在泵切换过程中系统流量越小、离心泵转动惯量越大,则系统失水量越大,最低流量越小,系统恢复稳定时间越长;止回阀特征时间越小,主回路流量波动越趋于稳定,但对切换过程系统流量波动影响十分有限。     

海洋条件下自然循环蒸汽发生器U型管内倒流特性研究

当压水堆处于自然循环工况时,蒸汽发生器U型管内可能发生倒流现象,导致一回路流动阻力增大、自然循环流量降低,为反应堆安全运行带来不利影响。基于RELAP5程序建立了海洋条件下的附加力模型及控制体空间坐标求解模型,对蒸汽发生器所有U型管进行建模和节点划分,计算了海洋条件下蒸汽发生器内U型管的倒流临界质量流量及进出口压差,最后分析了3种海洋条件对U型管内流体倒流的影响。结果表明,倾斜条件下有可能会改变倒流现象;而在航行过程中可能遇到的起伏条件都无法改变倒流现象;当摇摆条件比较剧烈时有可能改变倒流现象。      

自然循环过渡过程UTSG一次侧倒流特性研究

自然循环条件下,立式倒U型管型蒸汽发生器（UTSG）并联倒U型传热管内存在非均匀流动,部分传热管出现倒流,倒流的发生对反应堆自然循环能力产生显著的影响。按管长对并联倒U型传热管进行分类,建立分布式的结构模型。采用最佳评估程序RELAP5/MOD3.3,对主泵不同转动惯量下的自然循环过渡过程进行研究,得到了转动惯量对UTSG倒U型管内非均匀流动的影响特性。研究结果表明,转动惯量的增加可以延缓倒流的发生,但对倒流的空间分布和倒流流量无影响。     

基于耦合程序的流体瞬变流动水锤现象分析

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。     

水锤实验装置和压水堆用新型止回阀的水锤试验研究

本文介绍了我们自行设计的以轻水(去离子水)为介质的的常温不锈钢回路的水锤冲击实验装置。采用专门设计的四通转换阀及其气动操作机构,可以对核反应堆冷却剂回路在“关、停、并、转”过程中产生的倒流现象进行水锤冲击模似试验。同时,对 φ150和 φ200两种压水堆用新型无冲击止回阀进行了水锤的试验研究,得出了它们的瞬态冲击波的波形、峰值大小和波动的持续时间等动态特性。对现有水锤计算方法作了一些分析。     

管长对UTSG倒流管空间分布的影响分析

自然循环条件下,自然循环U型管蒸汽发生器(UTSG)并联倒U型传热管内存在非均匀流动,部分传热管出现倒流;现有研究对倒流管的空间分布存在不同认识。通过建立一维、稳态条件下的传热管内流动换热模型,发现传热管管长对并联倒U型管间的流量压降对比关系产生影响,进而对倒流管的空间分布产生影响。传热管管长较小的UTSG,短管将先于长管发生流量漂移;反之,长管将先于短管发生流量漂移。采用最佳评估程序RELAP5/MOD3.3对研究结果进行了验证。     

多泵并联给水系统给水泵切换运行规律仿真研究

多泵并联给水系统作为核动力系统的主要子系统之一,其给水泵的切换运行规律对系统运行经济性以及系统运行特性至关重要。本研究利用系统仿真支撑软件APROS建立了多泵并联给水系统仿真模型,并依据额定设计值验证了模型的准确性。基于此,通过进行不同切换条件下的线性升、降负荷仿真,对给水泵切换运行规律和系统动态特性进行了研究。研究结果表明,针对本研究对象,其高负荷工况切换点选取为70%额定流量,低负荷工况切换点选取为30%额定流量时,既能获得良好的系统动态响应,还能保持给水泵运行经济性较高。此外,低负荷工况对给水泵切换引入的扰动更为敏感。低负荷工况下,若切换条件选取不当,则会导致降负荷过程中系统触发超压排放。     

旋启式止回阀关阀水锤优化分析

建立了双泵并联给水系统数学物理模型,针对旋启式止回阀关闭过程中形成的水锤进行编程计算,分析作用在止回阀阀板上的力矩以及阻尼扭簧力矩对关阀水锤的影响。结果表明：在双泵并联给水系统中,旋启式止回阀在关闭时会产生明显水锤,作用在阀板上的力矩对水锤的作用效果有一定影响,选择合适的止回阀阀板材料以及加装合适的弹簧可有效缓解水锤危害。     

Numerical simulation and optimization on valve-induced water hammer characteristics for parallel pump feedwater system

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.     

Mitigating check valve slamming and subsequentwater hammer events for PPFS using MOC

The method of characteristic (MOC) was adopted to analyze the check valve-induced water hammer behaviors for a Parallel Pumps Feedwater System (PPFS) during the alternate startup process. The motion of check valve disc was simulated using inertial valve model. Transient parameters including the pressure oscillation, local flow velocity and slamming of the check valve disc etc. have been obtained. The results showed 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. The scheme of appending a damping torque to slow down the check valve closing speed was also performed to mitigate of water hammer. It has been numerically approved to be an effective approach.     

加热功率对低压低高差自然循环系统两相流动特性影响研究

对浮动式核电站中一类具有倾斜热管段的低压低高差自然循环系统的两相流动特性进行了实验研究,分析了加热功率对两相流动特性的影响。结果表明,不同功率条件下系统存在两相稳定冷凝和伴随蒸汽冷凝诱发水锤两相振荡2种流动模式,热管段内过冷水倒流和蒸汽与低温过冷水直接接触冷凝是导致2种流动模式的内在机制。此外,蒸汽冷凝诱发水锤的发生会产生较大压力脉冲,并导致过冷水倒流长度显著增加,进而加剧系统流动不稳定。进一步研究表明,加热段出口含气率可以作为流动不稳定判断依据。      

Startup Thermal Analysis of a High-Temperature Supercritical-Pressure Light Water Reactor

The startup systems of a high-temperature supercritical-pressure light-water-cooled thermal reactor (SCLWR-H), in which the core outlet temperature is 500°C and downward-flowing water rods are used as moderators, are studied by thermal-hydraulic analysis. The thermal analyses are carried out for various startup phases and detailed procedures for these phases are investigated. In constant pressure startup system, the reactor starts at supercritical pressure. A flash tank and pressure-reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at subcritical pressure. A steam-water separator and a drain tank are required. The separator is designed by referring to those of supercritical fossil-fired power plants (FPPs). The maximum cladding surface temperature is restricted not to exceed the rated value of 620°C. The minimum flow rate is 25% for constant pressure startup and 35% for sliding pressure startup. Both constant pressure and sliding pressure startup systems are found feasible from thermal analysis. Because of lower flow rate than SCFR, of which the core outlet temperature is about 430°C, the component weight required is reduced in SCLWR-H. The sliding pressure startup system should be used to reduce the component weight and to simplify the plant system.     

Thermal and Stability Considerations of Super LWR during Sliding Pressure Startup

The feasibility of the sliding pressure startup of a high-temperature supercritical-pressure light water reactor (super LWR, SCLWR-H) is assessed from both thermal and stability considerations. In the sliding pressure startup, nuclear heating starts at subcritical pressure and the reactor is pressurized to supercritical pressure at a low power and high enough flow rate. The reactor power and flow rate are then raised gradually to the rated normal values at constant supercritical operating pressure. During startup, the maximum cladding surface temperature must not exceed 620°C. For two-phase flow at subcritical pressures, the homogeneous equilibrium model is used. The thermal-hydraulic and coupled neutronic thermal-hydraulic stabilities during pressurization and power-raising are investigated by a frequency-domain linear analysis for both supercritical-pressure and subcritical-pressure operating conditions. The same stability criteria as those of BWRs are used. From the analysis results, a sliding pressure startup procedure is proposed for super LWR. The thermal criteria are satisfied by keeping the core power between the maximum allowable limit and minimum limit required for turbine startup and operation. The thermal-hydraulic stability and coupled neutronic thermal-hydraulic stability can be maintained by applying an orifice pressure drop coefficient at the inlet of fuel assembly and by controlling the power and flow rate during startup.     

Flow-induced vibrations in closed side branch pipes and their attenuation methods

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.     

超临界水堆全系统启动特性研究

为研究超临界水堆（SCWR）全系统启动特性，以SCTRAN程序为计算工具，基于中国超临界水堆（CSR1000）堆芯参数、高性能轻水反应堆（HPLWR）热力循环回路和日本SCWR再循环启动回路，建立了SCWR完整再循环启动系统模型。通过与HPLWR热力循环回路的稳态参数对比，验证了完整回路模型的正确性。分析在控制系统控制下的CSR1000再循环启动过程，得到了启动过程中堆芯、汽鼓、汽轮机、各级抽汽、再热器、各级回热器的瞬态响应曲线。计算结果表明，启动序列和启动过程各热工参数的变化符合预期，系统稳定启动；堆芯始终处于单相状态；汽轮机入口为超临界蒸汽；经过高压和低压回热器后堆芯入口温度能够达到280℃；高压缸入口压力维持恒定；在启动的过程中最大燃料包壳表面温度低于限值温度650℃，整个启动过程安全可靠。      

CSR1000启动过程控制特性研究

启动系统和启动特性分析是超临界水堆（SCWR）设计的重要组成部分，为了实现全系统启动分析，以SCWR瞬态分析程序SCTRAN为基础，提出了新的宽参数范围的壁面换热模型，在此基础上设计了启动过程的控制系统，包括冷却剂流量、堆芯入口温度、系统压力、堆芯功率、汽鼓水位控制。根据启动各阶段的不同控制目标建立不同的控制方案，并以中国百万千瓦SCWR（CSR1000）为研究对象，建立了包括再循环回路和直流冷却回路的分析模型，提出了采用控制系统的SCWR的4个启动过程。计算结果表明，再循环回路和直流冷却回路在各个启动过程中，各热工参数变化符合预期，最高包壳表面温度不超过限值温度650℃，验证了启动方案的可行性和启动过程的安全性。