IPWRs非能动余热排出系统运行特性分析

由于结构紧凑和采用模块化及非能动安全技术,一体化压水堆(IPWRs)特别适合于舰船核动力装置的应用。本文研究对象为基于固有安全一体化动力堆UZrHx和俄罗斯一体化压水堆ABV-6M的运行特点而概念设计的一体化压水堆。堆芯采用弧形板状燃料元件,直流蒸汽发生器形式为套管式,利用3个回路的自然循环排出堆芯余热的非能动余热排出系统以及一套能动的停堆冷却系统。运用RE-LAP5/MOD3.4程序对该反应堆在全船断电事故工况下反应堆停堆,非能动余热排出系统和能动停堆冷却系统分别投入运行进行仿真计算,分析其热工水力动态特性,保证堆芯安全。     

The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR

1 Introduction The technology of passive safety is the trend of safety systems in nuclear power plant, and various novel reactor concepts, including AP600, EPP1000, SPWR, WWER1000, and MS600, have adopted pas- sive safety systems [1]. Passive safety system is one of the main features of Chinese advanced PWR, which is different from other conventional PWR [2]. Passive residual heat removal system (PRHRS), which ac- counts for the majority of passive safety systems of Chinese advanced…     

Development of the analysis tool for the water cooling type passive residual heat removal system of Chinese pressurized reactor

An innovative design for Chinese pressurized reactor is the steam generator (SG) secondary side water cooling passive residual heat removal system (PRHRS). The new design is expected to improve reliability and safety of the Chinese pressurized reactor during the event of feed line break or station blackout (SBO) accident. The new system is comprised of a SG, a cooling water pool, a heat exchanger (HX), an emergency makeup tank (EMT) and corresponding valves and pipes. In order to evaluate the reliability of the water cooling PRHRS, an analysis tool was developed based on the drift flux mixture flow model. The preliminary validation of the analysis tool was made by comparing to the experimental data of ESPRIT facility. Calculation results under both high pressure condition and low pressure condition fitted the experimental data remarkably well. A hypothetical SBO accident was studied by taking the residual power table under SBO accident as the input condition of the analysis tool. The calculation results showed that the EMT could supply the water to the SG shell side successfully during SBO accident. The residual power could be taken away successfully by the two-phase natural circulation established in the water cooling PRHRS loop. Results indicate the analysis tool can be used to study the steady and transient operating characteristics of the water cooling PRHRS during some accidents of the Chinese pressurized reactor. The present work has very important realistic significance to the engineering design and assessment of the water cooling PRHRS for Chinese NPPs.     

非能动余热排出系统数学模型研究与运行特性分析

利用某型核动力装置非能动余热排出系统1:10原理性试验的8个稳态工况、6个启动工况的试验数据,验证RELAP5/MOD3.2程序对本类型非能动余热排出系统的适用性。结果表明:垂直管内蒸汽凝结换热系数对两相流自然循环的流动与传热影响大;RELAP5/MOD3.2程序过低估算了垂直管内蒸汽流速对蒸汽凝结换热系数的影响,计算结果与试验结果偏差大。对RELAP5/MOD3.2程序垂直管内的蒸汽凝结换热模型进行修正,修正后的计算结果与试验值基本吻合;采用RELAP5程序对垂直管内两相流自然循环传热进行计算,须选择热前沿跟踪模型。对非能动余热排出系统的稳态与瞬态运行特性进行分析,理论计算与试验结果均表明:稳态工况下,系统可以实现稳定的两相流自然循环,系统排热能力受蒸汽发生器水位的影响大,冷却水入口温度与系统压力的影响相对较小;系统的启动特性良好,可快速地建立环路的自然循环,带走反应堆的衰变热。     

Research on operational characteristics of passive residual heat removal system under rolling motion

The operational characteristics of passive residual heat removal system under rolling motion were investigated experimentally. The passive residual heat removal system under rolling motion was simulated with the advanced RELAP5 code. The results are consistent with experiments. The relative discrepancy between calculating and experimental results is less than 10%. The modified condensation heat transfer model can also be used to calculate the condensation heat transfer coefficient with droplet carryover precisely. The fluctuation of condensate temperature and steam pressure is not noticeable. As the power becomes larger for the same rolling motion, the oscillation amplitude of condensate flow rate becomes larger. The effect of rolling motion upon heat transfer coefficient and flow resistance was investigated with experimental results. Rolling motion can increase the flow resistance in a great extent. The more serious the rolling is, the more the flow resistance is. Additional pressure drop does not effect on average flow velocity. The decreasing of average flow velocity is due to the decreasing average gravity pressure drop and the increasing of flow resistance. The contribution of gravity pressure drop on the decrement of average flow velocity is less than 20%. The other is due to the increasing flow resistance. In the present paper, the experimental results are listed first, and then the simulation results comparing with the experimental results are listed in the second part. At last, the effect of rolling motion is investigated theoretically.     

Transient analyses of the passive residual heat removal system

The steam generator secondary residual heat removal system is a new design with three-fold natural circulation loop. It is the candidate of the safeguard systems of next generation pressurized water reactor design in China. This paper presents the development of a program, which would be used in the system design. One-dimensional drift flux mixture flow model is used for analysis of the steam water loop, whereas single-phase fluid model is applied to single flow in primary system and air loop. Based on the mathematical model, the code SGSPRHR has been developed. The transient analyses, when station black out accident happens, have been made. Sensitivity study results have also been given. The calculated parameter variation trends are reasonable.     

Screening of critical parameters influencing thermal-hydraulic reliability of simple passive system

Robust safety nature of passive safety systems (PSSs) accounts for their increasing applications. Critical parameters (CPs) which influence reliability of thermal-hydraulic (t-h) PSSs are considered independent in most cases while considering their effects purposely for simplicity which may not be realistic. Findings affirmed reliability of t-h PSSs to be influenced by CPs that are dependent in most scenarios and thus, effects of CPs dependency which can directly/indirectly influence t-h reliability need to be considered. Reliability assessment methodologies (RAM) can thus be improved upon by considering the dependency of CPs in reliability analysis. In this regard, this paper considers the screening of CPs required to justify their dependency consideration in evaluating t-h reliability. The Pearson’s product moment correlation coefficient and covariance method were applied as illustration for the screening of the possible realistic CPs, which affect natural circulation of a passively water-cooled steam generator. The approach was used to determine the combinations of the CPs that are dependent and screens out those adjudged independent. Based on the results obtained, appropriate considerations (dependency/independency) can be made and further analysis of interest (failure/reliability) can be conducted for the system. Incorporation of this screening approach into the existing t-h RAMs will improve their efficiency.     

Numerical investigation of natural convection characteristics of a heat pipe-cooled passive residual heat removal system for molten salt reactors

The limited availability of studies on the natural convection heat transfer characteristics of fluoride salt has hindered progress in the design of passive residual heat removal systems(PRHRS) for molten salt reactors. This paper presents results from a numerical investigation of natural convection heat transfer characteristics of fluoride salt and heat pipes in the drain tank of a PRHRS. Simulation results are compared with experimental data,demonstrating the accuracy of the calculation methodo...     

Development of a thermal-hydraulic subchannel analysis code for motion conditions

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable.     

Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR

Thermal-hydraulic characteristic investigation on passive residual heat removal system (PRHRS) of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features. A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified. On the basis of theory analysis, a correlation of two-phase natural circulation was obtained, and relative errors of 95% test data were less than ±16%. There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink, and its correlation of two-phase natural circulation system has been obtained. The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.     

压力管式反应堆非能动余热排出系统方案研究

本文针对压力管式钍基先进核能系统(TANES)提出了一种非能动余热排出(PRHR)系统方案。该方案利用两个回路的自然循环，将事故工况下的堆芯余热排出到最终热阱。利用RETRAN02程序，以全厂断电事故为设计基准事故，对TANES非能动余热排出系统的余热排出能力进行了计算。计算表明，TANES的PRHR系统能够将余热导向最终热阱并且保持冷却剂回路和慢化剂回路的压力低于设计限值。另外对诸如设备间高度差等因素进行了敏感性分析。     

用AC-600非能动余热排出系统实验评估RELAP5程序

利用RELAP5程序对先进堆二次侧非能动堆芯余热排出系统实验的瞬态过程进行数值模拟。在微循环启动,有注水的工况下,比较了RELAP5程序的计算结果和实验数据,计算结果与实验基本一致。由此可见,利用RELAP5程序分析此类问题是可行的。瞬态计算结果还为先进压水堆非能动余热排出系统的设计提供参考。     

Thermal hydraulic calculation in a passive residual heat removal system of the SMART-P plant for forced and natural convection conditions

An investigation of the thermal hydraulic characteristics in the passive residual heat removal system of the System integrated Modular Advanced ReacTor-P (SMART-P) has been carried out using the MARS code, which is a best estimate system analysis code. The SMART-P is designed to cool the system during accidental conditions by a natural convection. The dominant heat transfer in the steam generator is a boiling mode under a forced convection condition, and it is a single-phase liquid and a boiling heat transfer under a natural convection condition. Most of the heat is removed in the heat exchanger of the passive residual heat removal system by a condensation heat transfer. The passive residual heat removal system can remove the energy from the primary side as long as the heat exchanger is submerged in the refueling water tank. The mass flow is stable under a natural circulation condition though it oscillates periodically with a small amplitude. The parameter study is performed by considering the effects of an effective height between the steam generator and the heat exchanger, a hydraulic resistance, an initial pressure, a non-condensable gas fraction in the compensating tank, and a valve actuation time, which are useful for the design of the passive residual heat removal system. The mass flow in the passive residual heat removal system has been affected by the height between the steam generator and the heat exchanger, and the hydraulic resistance of the loop.     

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 °C which is substantially lower than ∼627 °C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a crucial factor for consideration in safety design. This study provides future researchers with a guideline on designing safety measures for the fourth generation of the fast reactors with no particular reference to any specific manufacturer.     

DEMO堆包层第一壁热工水力优化分析研究

聚变堆包层第一壁是影响包层换热效率与运行安全性最重要的部件,为了研究第DEMO堆包层第一壁的热工水力性能,对第一壁流道内氦气冷却剂的流动及其与结构材料的换热进行了数值模拟研究及优化分析。结果表明,通过增大氦气进口质量流量可以有效地降低第一壁结构材料的最高温度,但是由此带来的压力损失很大,不能作为强化换热的主要途径。此外,增加每组流道的盘绕次数能起到强化换热的效果,目前每组流道盘绕五次的方案是合理的。流道中存在的圆角包层第一壁的流动换热影响不大,但圆角的存在会使第一壁最高温度有一定的升高。铍涂层的导热系数与第一壁最高温度成反相关关系,但是对第一壁流道的对流换热影响不大。结构材料的导热系数的增大能显著降低第一壁结构材料的最高温度。     

Design and transient analyses of emergency passive residual heat removal system of CPR1000

The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the steam impact on the EPRHRs, and improve the stability of the system.     

Application of grey model on analyzing the passive natural circulation residual heat removal system of HTR-10

Using the grey correlation analysis, it can be concluded that the reactor pressure vessel wall temperature has the strongest effect on the passive residual heat removal system in HTR （High Temperature gas-cooled Reactor）, the chimney height takes the second place, and the influence of inlet air temperature of the chimney is the least. This conclusion is the same as that analyzed by the traditional method. According to the grey model theory, the GM（1,1） and GM（1, 3） model are built based on the inlet air temperature of chinmey, pressure vessel temperature and the chimney height. Then the effect of three factors on the heat removal power is studied in this paper. The model plays an import,ant role on data prediction, and is a new method for studying the heat removal power. The method can provide a new theoretical analysis to the passive residual heat removal system of HTR.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

非能动余热排出热交换器流动和传热数值模拟

非能动余热排除系统(Passive Residual Heat Removal system,PRHR)是非能动核电厂的重要安全设施,在全厂断电事故下,大部分的堆芯衰变热是通过PRHR热交换器传递至内置换料水箱(In-containment Refueling Water Storage Tank,IRWST)。但PRHR热交换器属于大型非稳态换热器,其传热机理十分复杂。基于PRHR系统的重要性和复杂性,有必要研究PRHR系统的流动和传热特性。利用计算流体动力学(Computational Fluid Dynamics,CFD)软件针对非能动堆芯冷却系统试验装置中的PRHR系统进行建模计算,分析了PRHR热交换器及IRWST的流动和传热特性,发现IRWST内部沿垂直高度上呈现明显的温度分层现象,温度沿水平方向的分布趋于均匀;IRWST内部的流动主要是沿着C型传热管竖直段向上流动,流速逐渐增大,但在两相阶段,水箱上部区域流动明显增强;C型传热管上部水平段和竖直段上部区域的换热系数要明显高于其它区域,且在上部水平段与竖直段连接弯管处换热系数最大,在两相阶段,上部区域的换热系数明显增大。     

The influence of ocean conditions on two-phase flow instability in a parallel multi-channel system

In this paper, the two-phase flow instability between multi-channels (FIBM) operating under ocean conditions is studied theoretically. The physical and mathematical model of the multi-channel system which is based on Lee and Pan [Lee, J.D., Pan, C., 1999. Dynamics of multiple parallel boiling channel systems with forced flows. Nuclear Engineering and Design 192, 31–44], is extended to analyze the influence of ocean conditions. The influence of ocean conditions on the FIBM is analyzed, particularly with respect to the periodic total mass flow rate and rolling motion. Furthermore, the instability oscillation trajectories of the multi-channel system are obtained on the phase plane of the inlet velocity and boiling boundary. Some of the trajectories show chaotic characteristic. The instability zone of a nine-channel system under rolling motion is then obtained.