小型压水堆抑压水池抑压特性研究

与传统大型压水堆相比，小型压水堆安全壳自由容积小，发生冷却剂失水事故(LOCA)后安全壳压力迅速上升，需采取抑压水池、安全壳喷淋等措施保证安全壳的完整性。为探究抑压式安全壳中抑压水池的抑压特性，设计了小型压水堆抑压喷淋系统模拟装置，并使用RELAP5程序对模拟装置进行建模，模拟安全壳抑压排热过程，分析了破口质能释放对抑压水池模拟和设计的影响，对比了不同破口喷放速度和喷淋流量下抑压水池采用不同抑压管面积和气水比的抑压效果。结果表明：抑压水池的抑压管面积和气水比存在最佳值；破口面积影响喷放速度，在质能释放总量相同的前提下，破口喷放速度越快，抑压水池的最佳抑压管面积和最佳气水比越大；增设喷淋后，最佳气水比随喷淋流量增大而增大，最佳抑压管面积在小范围内变化。研究结果可为小型压水堆抑压式安全壳的设计和分析提供参考。     

抑压式安全壳的抑压特性研究

以100 MW级核电厂压水堆为对象,通过对反应堆冷却剂失水事故(LOCA)初期安全壳压力温度响应的分析,对抑压式安全壳抑压特性进行研究。由于LOCA事故喷放阶段质能释放焓值较高,安全壳喷淋难以及时有效地抑制安全壳压力的上升,而采用抑压水池对抑制事故初期的压力具有较为明显的效果。通过对抑压水池总容积、气水容积比、排放管流通面积等重要参数的分析,对抑压效果的影响表现为:其中抑压水池总容积大小对抑压效果影响程度最大;并且抑压水池气水容积存在最佳比;排放管流通面积存在最佳范围。     

小型压水堆安全壳抑压传热研究

小型压水堆(PWR)采用抑压水箱和浸没式钢制安全壳的设计,冷却剂失水事故(LOCA)后短期内利用抑压水箱抑制安全壳压力上升,长期利用钢安全壳与外部水池的传热带出热量。其抑压和传热特性与传统干式安全壳存在本质区别。本文分别对抑压水箱的抑压特性和钢制安全壳壁换热特性进行了对比研究,并针对PWR安全壳抑压水箱容量和安全壳换热面积进行了需求分析。分析结果表明,抑压水箱能有效地抑制安全壳短期升压,钢安全壳换热面积增大到350 m2时,即可满足长期供热要求。     

LOCA下小型抑压式安全壳内流动与传热现象初步研究

将沸水堆抑压技术应用到小型堆的安全设计中是当前小堆研究的热点之一.当壳内管道发生破口事故时,高温高压的蒸汽迅速喷入安全壳,导致压力、温度迅速升高,而抑压系统则可以使安全壳迅速降压.实际事故工况下的气体为蒸汽和不凝性气体混合物.按照从简到繁的研究方法,本文以没有相变的热空气作为工质进行数值模拟,研究气体在抑压系统中的流动...     

采用能动和非能动安全系统的HPLWR安全壳设计建议

基于最新的沸水堆安全壳设计概念，提出了类似带有余热排出系统的高性能轻水堆安全壳的概念。它包括1个抑压水池、4个堆芯淹没水池、4个能动余热排出系统和4个非能动余热排出系统。安全壳的内径为20m，高度约为22m。4个冗余的能动系统在低压下从抑压水池向反应堆提供补水。8个压力释放阀和上都堆芯淹没水池中的8个喷头提供快速卸压。作为一种替代方法，堆芯产生的蒸汽可通过8个非能动热移出系统排放，每个系统都有一个蒸汽喷射器和一个浸入堆芯淹没水池的冷凝器。这个方案虽然还没有进行过任何瞬态分析研究，但可以作为一个可行的安全壳设计概念的基础来讨论。     

小型压水堆抑压特性试验与数值模拟研究

为了研究小型压水堆抑压系统的抑压效果，建立了小型安全壳抑压特性试验装置，开展了定流量和变流量混合气体排放实验，以研究气-水容积比和不可凝气体对抑压效果的影响。实验结果表明，气-水容积比在2～4.55范围内，随着气-水容积比的增大，抑压效果逐渐增强；混合气体中不可凝气体含量对抑压效果影响显著。对实验进行了数值模拟，模拟结果可以反映抑压试验的现象规律，但仍需进一步优化抑压冷凝相关模型以提高模拟精度。     

非能动安全壳冷却系统性能提高的初步分析

为分析先进非能动型压水堆(AP1000)非能动安全壳冷却系统(PCCS)的运行特征,建立单体积模型来计算安全壳内部气体混合物的传热,并结合前期已开发的安全壳外部水膜蒸发冷却的多体积程序,对冷却剂丧失事故(LOCA)下的安全壳运行压力进行计算,计算结果与WGOTHIC以及相关文献的结果吻合良好。通过讨论安全壳的运行特性,分析安全壳的排热能力。最后提出抑压方案来提高PCCS的性能,新的方案能够抑制大约10%的安全壳压力,降低PCCS的工程投资和难度。     

浮动核电站抑压水池液舱晃荡研究

浮动核电站运行环境与陆地核电站有着显著不同，其专设安全设施的设计需考虑海洋环境适应性，尤其是涉及液体流动的设施更应考察船体运动激励的影响。本文以抑压水池为分析对象，采用有限体积法对液舱晃荡过程进行仿真分析，研究在极限海洋环境下水池内水位的变化，以及有/无制流板情况下水位的区别。研究表明，船体纵/横摇引起的抑压水池最低水位相比于初始水位显著降低，由于抑压水池共振周期与船体晃荡周期错开，因此水位变化主要由晃荡幅值和内部结构件共同作用决定。      

某核电厂LOCA下预应力混凝土安全壳响应规律初探

核电厂LOCA发生后,预应力混凝土安全壳结构内温度场分布具有明显的非线性特征,但现行的混凝土安全壳设计规范未对LOCA下温度和应力的组合作用提出具体的计算方法。基于用ANSYS程序建立的包含预应力钢束的混凝土安全壳结构的有限元模型,本文计算了LOCA下不同时刻安全壳壳壁内的温度场分布,并与理论值进行了比较,验证了计算模型的正确性。初步分析了高温、高压作用下安全壳结构变形的规律,总结了混凝土温度效应和预应力系统的作用,可为安全壳结构设计提供参考。     

ACP100+失水事故应对策略研究

对先进模块化小型堆(ACP100+)失水事故后的应急堆芯冷却、安全壳压力控制、余热长期导出提出了相应的应对手段和策略。初步计算分析表明:通过限制ACP100+反应堆冷却剂破口尺寸可取消安注箱,使安注系统得到简化;对于小型钢安全壳带来的事故后压力控制问题,可采用抑压水池和顶部非能动水池设计,使事故后安全壳压力可长期控制在设计限值以下;由于小型堆余热量较小,可利用钢安全壳体作为导热媒介,通过浸没安全壳顶盖,以自然对流的方式长期导出余热,其长期冷却的固有安全性得到进一步提高。     

Optimization Analysis on Pressure Suppression System of Small Reactor Containment

The containment pressure rises rapidly after LOCA, especially for the small reactors containment with very small free capacity, in order to avoid the rapid rise of containment pressure in the short term after LOCA, a pressure suppression system should be arranged in the containment. In this paper, the GOTHIC program was used to model the containment with pressure suppression system, and sensitivity analysis was carried out on the thermal response of containment after LOCA under different configuration schemes of pressure suppression system, the demonstration method of containment capacity with pressure suppression pool system and the optimal scheme were obtained. The analysis results show that the pressure suppression pool can significantly reduce the pressure in the containment, the pressure in the containment varies greatly under different configurations of pressure suppression pool modules, and the optimal configuration should be carried out for the containment design scheme in the design process.     

Vent clearing analysis of a Mark III pressure suppression containment

An analysis of the vent clearing transient in a Mark III pressure suppression containment after a hypothetical LOCA is carried out. A two-dimensional numerical model solving the transient fluid dynamic equations is used. The geometry of the pressure suppression pool is represented and the pressure and velocity fields in the pool are obtained from the moment the LOCA occurs until the first vent in the drywell wall clears. The results are compared to those obtained with the one-dimensional model used for containment design, with special interest on two-dimensional effects. Some conclusions concerning the effect of the water discharged into the suppression pool through the vents on submerged structures are obtained. Future improvements to the model are suggested.     

华龙一号非能动安全壳冷却系统热工水力分析

本文采用不可压缩流体均匀流模型对华龙一号(HPR1000)的非能动安全壳冷却系统（PCS）进行数值模拟,在反应堆冷却剂系统（RCS）大破口丧失冷却剂事故（LOCA）工况下对PCS进行热工水力分析,并对PCS设计工况进行性能分析计算。结果表明：PCS的非能动运行特性与事故进程具有很好的匹配能力,能在事故早期极快启动,并在24 h内将安全壳的温度和压力稳定在安全范围内。通过PCS设计工况的换热性能分析,PCS在运行5 h后进入两相流传热阶段,当换热水箱介质达到饱和温度后仍能长期稳定运行,导出安全壳内热量。     

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition.     

Modeling pressure suppression pool hydrodynamics in the ABWR containment

This paper focuses on the assessment of pressure suppression pool hydrodynamics in the advanced boiling water reactor (ABWR) containment under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a mechanistic model for predicting various suppression pool hydrodynamics parameters. A phenomena identification and ranking table (PIRT) applicable to the ABWR containment pool hydrodynamics analysis is used as a basis for the development of the model. The highly ranked phenomena are represented by analytic equations or empirical correlations. The best estimate and several sensitivity calculations are performed for the ABWR containment using this model. Results of the sensitivity calculations are also presented that demonstrate the influence of key model parameters and assumptions on the pool hydrodynamics parameters. A comparison of model predictions to the results of the licensing analyses shows reasonable agreement. Comparison of the results of the proposed model to experimental data shows that the model predicted top vent clearance time, the pool swell height, and the bubble breakthrough elevation are within 10% of the data. The predicted pool surface velocity and the liquid slug thickness are within 30% of the measurements, which is considered adequate given the large uncertainties in the experimental measurements.     

小破口失水事故非能动系统瞬态特性研究

为了解先进压水堆小破口失水事故下非能动安全壳冷却系统、非能动堆芯冷却系统、非能动余热排出系统的瞬态响应特性,需开展小破口失水事故下反应堆冷却剂系统和安全壳的耦合响应特性研究。分析结果表明,小破口失水事故下,耦合分析中非能动余热排出系统、非能动堆芯冷却系统、自动卸压系统和非能动安全壳冷却系统的特性与独立计算有较大差异,小破口失水事故下耦合分析得到的安全壳压力峰值小于独立计算。      

1000MW级压水堆安全壳压力温度计算分析

使用计算机程序PAREO9对157XL 1000MW级压水堆核电厂安全壳的压力温度进行计算分析．论证安全壳的自由容积．并计算分析了安全壳自由容积为53163m^3时157XL 1000MW级压水堆核电厂安全壳的压力和温度.     

LOCA steam condensation loads in BWR Mark II pressure suppression containment system

Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates ( 30 kg/m²·s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.     

Some aspects of the reliability-based design of reactor containment structures

A rational procedure for the design of reactor containment structures is carried out within a probabilistic framework. Various risk concepts such as the return period, non-encounter probability and the reliability function are discussed. Internal load conditions caused by system failure such as LOCA pressure loads, and external load conditions caused, for instance, by impact due to aircraft crashes, external pressure waves and natural hazards such as earthquakes and severe storms, are described by extreme value distributions of the largest values of the Fisher-Tippett types. Statistical and physical arguments are given to support their application. The occurrence of these rare events with respect to time is modeled by a Poisson process. The ultimate strength of a PWR containment structure for the steel (liner) shell is also modeled by an extreme value distribution (of the smallest values). As a good approximation the load action of the shell structure is determined by linear elastic analysis. The failure criterion considered here is that of reaching the ultimate tensile strength at one point of the structure. A numerical example of the reliability analysis of a steel shell structure under internal overpressure is carried out.     

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods.An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m³ and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1:64 compared to the 1200 MW PWR plant Biblis A.Up to now the test facility has been used for four trial runs and nine PWR LOCA experiments with single- and double-ended pipe ruptures of 100 mm dia. in a steam generator compartment and in the nozzle compartment. The initial conditions of the pressurized water in the coolant circuit before rupture were 140 bar and 290°C. About 0.1 sec after the rupture the flow rate at the site of rupture reaches its maximum of about 400 kg/sec (single-ended rupture) and 800 kg/sec (double-ended rupture). From the compartment where the rupture takes place a water-steam-air mixture streams through openings into the other compartments of the containment. Differential pressures between the compartments were measured with maximums of up to a few bar 0.15–0.5 sec after rupture, depending on the positions of rooms and transducers.Approximately 30–40 sec after rupture the blowdown has finished and the pressure in the containment has reached about 4–5 bar. The maximum pressure in a model containment is lower and the decrease of the pressure by condensation is faster than in a full-scale containment, due to the greater ratio of inner surface area to volume of a model containment. During blowdown the temperature of the containment atmosphere rises to about 150°C. Several minutes later the temperature of the concrete walls has increased non-uniformly causing considerable stress in the walls. Approximately 30 min after rupture measurements on the outside of the outer containment wall show a temperature-caused strain of about 30–60% of the maximum pressure-caused strain. A comparison between experiments and calculations shows discrepancies indicating the need for further development of calculational methods.