非能动安全壳冷却系统空气流道自然循环的比例分析与设计

在考虑建设试验台架经济性的前提下,缩小比例的单项和整体效应试验台架对研究和开发大型先进压水堆核电站及其分析验证程序都具有重要意义。非能动安全壳冷却系统（PCS）壳外空气流道内的自然循环在安全壳非能动冷却性能中发挥着重要的作用。本文从自然循环的数学模型出发,推导出了单项和整体效应试验台架的比例设计方法。在给定壳内热流密度的条件下,通过PCCSAP-3D程序对CAP1400非能动安全壳的2/5比例单项效应试验理想比例台架（ISF）进行模拟。结果表明,本比例分析与设计方法以及在降低高度台架上模拟自然循环是可行的。     

Passive containment cooling system (PCCS) response with Drywell Gas Recirculation System (DGRS) activated during a postulated Loss of Coolant Accident (LOCA)

Passive Containment Cooling Systems (PCCS) are characterizing the design of several advanced LWR such as SBWR, ESBWR, ABWR-II, etc. These systems should ensure the mitigation of postulated accidents both under Design Basic Accident (DBA) and Beyond DBA (BDBA) conditions. Some ALWR designs integrated in the PCCS a system called Drywell Gas Recirculation System (DGRS). The DGRS works like a fan, with inlet flow lines connected to the Passive Cooling condenser (PCC) vent lines and the outflow line connected to the Drywell (DW). The present paper presents the experimental results of an integral containment test performed in the PANDA facility. The initial conditions (temperature, pressure, gas composition, decay heat, etc.) for the test represent the containment situation 1 h after a Loss of Coolant Accident (LOCA). The test consists of two phases (6 h each) for a total duration of about 12 h. In the first phase has been simulated the response of the PCCS to a LOCA, in the second phase the DGRS has been activated and has been investigated the effect of such activation on the overall PCCS response. The test shows that the activation of the DGRS has an effect on the overall PCCS characteristics, i.e. composition of gas mixture in the PCC tubes, stratification in the Wetwell (WW), DW-WW pressure differences, timing for the opening of the Vacuum Breakers (VB) and overall containment pressure.     

Development of Evaluation Code for Long-Term Cooling of Containment with Passive Safety Feature

Simplified BWRs are characterized as an adoption of a passive ECCS and a passive containment cooling system (PCCS). While a passive ECCS has a short term core cooling function, a PCCS has a long-term decay heat removal function. As a PCCS, several concepts, differing in cooling location and method employed, have been considered. From the containment thermal- hydraulic response analysis viewpoint, simplified BWRs are essentially different from the current BWRs. For evaluating and comparing the performance of several PCCSs over full break spectra, the new containment safety evaluation code TOSPAC was developed as a preliminary design tool for PCCS. This paper summarizes the thermal-hydraulic modelings of the TOSPAC code and the validity evaluation of the TOSPAC code, compared with TRAC-BF1 calculation.

From the validity evaluation concerning a main steam line break (MSLB) accident analysis for an isolation condenser (I/C) as a PCCS, it was found that the TOSPAC calculation result shows reasonable agreement with that for TRAC, even though the TOSPAC consists of simpler modelings.     

Heat Removal Performance Evaluation of Several Passive Containment Cooling Systems during Loss of Coolant Accident

As a passive containment cooling system (PCCS), which is adopted in simplified BWRs, several concepts, differing in cooling location and method, such as the suppression chamber water wall, the drywell water wall, the isolation condenser (I/C) and the drywell cooler, have been considered. This paper summarizes the characteristics of each PCCS concept, and the analysis results of the performance for several PCCSs during a main steam line break LOCA for a reference simplified BWR plant, obtained by the newly developed containment thermalhydraulic response analysis code TOSPAC.

The performance comparison suggests that I/C and drywell cooler have good heat removal capability with regard to the smallest heat transfer area among PCCS concepts evaluated in the present analysis. I/C removes decay heat efficiently, since it absorbs steam directly from the reactor pressure vessel, which is the hottest portion inside the containment. The suppression chamber water wall is ineffective, mainly due to high non-condensable gas partial pressure in the suppression chamber, and low suppression pool temperature.

Calculations of other pipe breaks were also implemented for the reference plant adopting I/C as PCCS. The results show the effectiveness of the I/C cooling over a wide range of break spectra.     

Pressure suppression pool mixing in passive advanced BWR plants

In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data.     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

Containments design for the advanced and simplified boiling water reactor standard plants

The Advanced Boiling Water Reactor (ABWR) design is based on construction and operating experience of nuclear power plants in Japan, United States, and Europe. To optimize the plant arrangement of the Advanced Boiling Water Reactor (ABWR) and to verify the structural feasibility to carry design loads a study was conducted. To arrive at an optimized plant arrangement with a minimum size reactor building (RB), a circular cylindrical reinforced concrete containment vessel (RCCV) with a flat top slab and a monolithically connected diaphragm slab has been selected.The Simplified Boiling Water Reactor (SBWR) is being developed as a standardized 600 MWe Advanced Light Water Reactor. The design concept of the SBWR is based on simplicity and passive features to enhance safety and reliability, improve performance and increase economic viability. Due to the use of passive containment cooling, SBWR has features that are different from those of existing designs.The objectives of the study for the ABWR containment and RB are to perform a structural analysis of the containment and RB and to evaluate the structure for conformance to the U.S. NRC requirements. The main objective of the studies for the SBWR is to demonstrate the structural design feasibility of the containment for the pressure and the temperature response associated with the passive systems adopted for the SBWR.     

Thermal-hydraulic and stress analysis of AP1000 reactor containment during LOCA in dry cooling mode

Some kinds of break in the reactor coolant system may cause the coolant to exit rapidly from the failure site,which leads to the loss of coolant accident (LOCA).In this paper,a stress analysis of an AP1000 reactor containment is performed in an LOCA,with the passive containment cooling system (PCCS) being available and not available for cooling the wall's containment.The variations in the mechanical properties of the wall's containment,including elastic modulus,strength,and stress,are analyzed using the ABAQUS code.A general two-phase model is applied for modeling thermal-hydraulic behavior inside the containment.Obtained pressure and temperature from thermal-hydraulic models are considered as boundary conditions of the ABAQUS code to obtain distributions of temperature and stress across steel shell of the containment in the accident.The results indicate that if the PCCS fails,the peak pressure inside the containment exceeds the design value.However,the stress would still be lower than the yield stress value,and no risk would threaten the integrity of the containment.     

Measurement of heat transfer coefficients for steam condensation on a vertical 21.5-mm-O.D. tube in the presence of air

ABSTRACT

Governing the rate of heat transport by condenser tubes in the passive containment cooling system (PCCS), the steam condensation over a vertical cylinder in the presence of air was investigated experimentally. The main objective of this study was to explore if the condensation heat transfer coefficient relies on the tube dimension, which has been a variable missed in most condensation models or has been embraced without experimental demonstration under phase change environments. The mean heat transfer coefficient was measured in the condensation test facility named JERICHO (JNU Experimental Rig for Investigation of Condensation Heat transfer On tube). The outer diameter of the condenser tube used in this study was set to 21.5 mm. The measured heat transfer coefficients were compared to those obtained from the 40-mm-O.D. tube, and a multiplier to correct the variation of the heat transfer coefficient with the tube diameter was proposed for its application to Lee correlation. The proposed correlation was further validated against another set of experimental data obtained from a separate test facility housing the 31.8-mm-O.D. tube.     

Use of an ideal scaled model for scaling evaluation

In this paper a method is described for using RELAP5 models to corroborate the scaling methodology that has been used for design of the Purdue University multidimensional Test Apparatus. This facility was built for the U.S. NRC to obtain data on the performance of the passive safety systems of the General Electric Company simplified boiling water reactor. Similarity between the prototype system and the scaled test facility is investigated for a main steam line break accident.     

基于威斯康辛冷凝试验本体的结构改进及分析

先进压水堆(APWR)是第三代核电技术的代表堆型之一,它采用了非能动安全系统,提高了安全性能。非能动安全壳冷却系统(PCCS)主要利用蒸汽的冷凝来带走安全壳内的热量。本文主要介绍了威斯康辛大学进行的冷凝试验的试验本体结构,应用ANSYS软件对其结构进行了应力分析,并在现有结构的基础上对外部加强筋布置进行了一定的改进和优化。通过计算和比较可以看出,经过改进后的加强筋布置,不仅满足原有的试验要求,结构布置合理,更提高了试验本体的承压能力,使其能够满足更高试验压力的需要。     

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.

For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems.     

AC600非能动安全壳冷却系统长期效应分析

利用自主开发的用于先进压水堆AC600非能动安全壳冷却系统的专用三维热工水力分析程序PCCSAC－3D，对AC600安全壳在大破口失水事故情况下进行了长期效应分析，该程序把钢安全壳内部的工质分为水蒸汽，不可凝干空气，连续相水和非连续相水，对气相引入k-ε湍流计算模型并考虑由于气体浓度差引起的扩散效应。PCCSAC－3D程序充分考虑了各种空间非均匀的物理因素的影响，能够较精细描述在发生核电厂设计基准情况下出现与安全壳非能动冷却系统有关的各种物理现象，本文对安全壳进行长期效应的分析结果表明，AC600非能动安全壳冷却系统能够保证安全壳的完整性。     

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.     

采用分离式热管的非能动安全壳冷却系统研究

非能动安全壳冷却系统（PCCS）是第3代先进压水堆核电厂重要的专设安全系统。本文提出一套采用分离式热管技术的PCCS,通过原理性试验和系统性热工水力分析程序研究系统启动和稳态运行的流动和传热特性,研究影响系统运行及热传输能力的关键因素,验证系统设计的可行性。研究结果表明,该系统的传热性能随安全壳的状态变化有极强的自适应能力,在事故工况下利用该系统作为非能动的安全壳热量移出措施是可行、有效的。程序分析结果与试验结果及国际上已有研究成果的对比分析表明,RELAP5程序对于该系统热工水力分析是适用的。蒸发段传热管内流型、传热模式、空泡份额等关键流动、传热参数的变化表明,系统初始充液率对系统传热性能有重要影响。较小的冷热芯位差即能提供足够的自然循环驱动力,冷热芯位差不是系统布置的主要制约因素。     

一维自然循环比例分析的理论模型

整体性能试验研究是验证先进非能动压水堆核电站堆芯冷却系统设计有效性的核心技术,一回路系统两相自然循环热工水力特性比例分析是确定整体性能试验装置尺度的主要理论依据。以一维漂移流模型为基础,对整个一回路两相自然循环系统控制方程积分,并求得稳态解,由此获得了系统的流动条件。应用初始流动条件与边界条件,对两相自然循环系统控制方程直接无量纲化,最终得到了整体性能试验装置与实际非能动电站热工水力特性的相似准则。     

Adequacy of Power-to-volume Scaling Philosophy to Simulate Natural Circulation in Integral Test Facilities

Theoretical and experimental investigations were carried out to study the adequacy of power-to-volume scaling philosophy for the simulation of natural circulation and to establish the scaling philosophy applicable for the design of the Integral Test Facility (ITF-AHWR) for the Indian Advanced Heavy Water Reactor (AHWR). The results indicate that a reduction in the flow channel diameter of the scaled facility as required by the power-to-volume scaling philosophy may affect the simulation of natural circulation behaviour of the prototype plants. This is caused by the distortions due to the inability to simulate the frictional resistance of the scaled facility. Hence, it is recommended that the flow channel diameter of the scaled facility should be as close as possible to the prototype. This was verified by comparing the natural circulation behaviour of a prototype 220MWe Indian PHWR and its scaled facility (FISBE-1) designed based on power-to-volume scaling philosophy. It is suggested from examinations using a mathematical model and a computer code that the FISBE-1 simulates the steady state and the general trend of transient natural circulation behaviour of the prototype reactor adequately. Finally the proposed scaling method was applied for the design of the ITF-AHWR.     

非能动安全壳冷却系统膜状冷凝强化换热设计

非能动安全壳冷却系统(PCCS)能在反应堆发生事故时将安全壳内部的热量及时导出,避免安全壳因超温、超压而失效。为强化换热,本文设想在安全壳内部安装阻隔带和液滴收集装置,通过降低层流区液膜厚度、扰动不可凝气体隔离层并充分利用湍流的换热强化作用,降低总的换热热阻,提高换热效率。以AP1000为例,依托GDLM模型对改进前后安全壳的换热情况进行分析,结果表明,通过安装阻隔带和液滴收集装置,能降低安全壳壁面的液膜厚度,提高壁面热流量,从而实现强化换热。     

Design and test of hydraulic vacuum breaker check valve for simplified boiling water reactor

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.     

IIST small break LOCA experiments with passive core cooling injection

The purpose of this study is to evaluate the performance of a passive core cooling system (PCCS) with passive injection during the cold-leg small break loss-of-coolant accidents (SBLOCAs) experiments conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. Four tests were performed simulating break sizes of 0.2–2% (approximately corresponding to 1.25–4″ breaks for a referenced nuclear power plant) at cold-leg for assessing the PCCS capability in accident management. The key thermal–hydraulic phenomena to core heat removal for PCCS are observed and discussed. The experimental results show that the PCCS has successfully provided a continuous removal of core heat and a long term core cooling can be reached for all cases of SBLOCA.