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.     

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。     

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.     

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...     

Preliminary study on the application of the heat pipe to the passive decay heat removal system of the modular HTR

To investigate the applicability of heat pipe to the decay hat removal (DHR) system of the modular HTRs, preliminary study of the Heat Pipe DHR System was performed. The results show that the Heat Pipe DHR System is applicable to the modular HTRs and its heat removal capability is sufficient. Especially, by applying the variable conductance heat pipe, possibility of a fully passive DHR system with lower heat loss during normal operation is suggested.

The experiments to obtain the fundamental characteristics data of the variable conductance heat pipe were carried out and the experimental results are also shown.     

Above-ground HTGR design for passive decay heat removal

The design of a small high-temperature gas-cooled reactor (HTGR) for passive decay heat removal which could be located deeply underground was proposed previously. In the present work, analogue design analyses of passive decay heat removal for an above-ground long-life small prismatic HTGR was carried out to obtain the conditions for successful decay heat removal by radiation and conduction inside the reactor building, and by radiation and natural cooling by air at the outer surface of the reactor building. Sensitivity analysis of the peak temperatures of both the core and the reactor building after reactor shutdown was performed by changing the physical characteristics of the reactor regions. Enlarging the reactor building was found to be an effective way to reduce the peak reactor building temperature to within its design limit. By using the obtained condition for design parameters, the appropriate sizes of reactor core and reactor building were evaluated for some reactors. Consequently, criticality and burnup analyses for the proposed reactors were performed to confirm the possibility of designing a long-life core for the core size and reactor power which meet the condition of removing decay heat successfully. Using our design, all the reactors with 20 wt% uranium enrichment could be critical for over nine years.     

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。     

Possible design of PBR for passive decay heat removal

Conditions for design parameters of above-ground and underground, prismatic high-temperature gas-cooled reactor (HTGR)s for passive decay heat removal based on fundamental heat transfer mechanisms were obtained in the previous works. In the present study, analogous conditions were obtained for pebble bed reactors by performing the same procedure using the model for heat transfer in porous media of COMSOL 4.3a software, and the results were compared. For the power density profile, several approximated distributions together with original one throughout the 10-MWt high-temperature gas-cooled reactor-test module (HTR-10) were used, and it was found that an HTR-10 with a uniform power density profile has the higher safety margin than those with other profiles. In other words, the safety features of a PBR can be enhanced by flattening the power density profile. We also found that a prismatic HTGR with a uniform power density profile throughout the core has a greater safety margin than a PBR with the same design characteristics. However, when the power density profile is not flattened during the operation, the PBR with the linear power density profile has more safety margin than the prismatic HTGR with the same design parameters and with the power density profile by cosine and Bessel functions.     

Survey of experiments and code development for the passive residual heat removal system of PWR in China

Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of PWR performed in Nuclear Power Institute of China (NPIC) recent ten years. The typical results of MISAP. a special code for PWR passive residual heat removal system developed and assessed by NPIC,are also described briefly in this paper.     

Sensitivity analysis of a passive decay heat removal system under a post-loss of coolant accident condition

Passive safety features are now of interest to the design of future generation reactors because of their peculiar characteristics of simplicity, reduction of human interaction, and avoidance of failures of active components. However, the large uncertainty associated with the responses of passive systems might not be ignored. Therefore, it is necessary to identify the uncertain inputs that have the important impact on the uncertainty of the system performance. In this study, two global sensitivity measures, the first-order sensitivity index and the total-order sensitivity index, are applied to a natural circulation decay heat removal system of a gas-cooled fast reactor for identifying the important system inputs. It is found that the uncertainty in the system pressure contributes the most to the uncertainty in the system outputs. In addition, the cooler (the heat exchanger of the emergency cooling system) wall temperature, the Nusselt number in the mixed convection regime, and the friction factor in the mixed convection flow regime also have small impact on the uncertainty of the system outputs.     

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。     

The influence of ocean conditions on thermal-hydraulic characteristics of a passive residual heat removal system

There are many differences between the flow and heat transfer characteristics of nuclear reactors under ocean and land-based conditions for the effects of ocean waves. In this paper, thermal hydraulic characteristics of a passive residual heat removal system (PRHRS) for an integrated pressurized water reactor (IPWR) in ocean environment were investigated theoretically. A series of reasonable theoretical models for a PRHRS in an IPWR were established. These models mainly include the core, once-through steam generator, nitrogen pressurizer, main coolant pump, flow and heat transfer and ocean motion models. The flow and heat transfer models are suitable for the core with plate-type fuel element and the once-through steam generator with annular channel, respectively. A transient analysis code in FORTRAN 90 format has been developed to analyze the thermal–hydraulic characteristics of the PRHRS under ocean conditions. The code was implemented to analyze the effects of different ocean motions on the transient thermal-hydraulic characteristics of PRHRS. It is found that the oscillating amplitudes and periods of the system parameters are determined by those of the ocean motions. The effect of rolling motion is more obvious than that of pitching motion when the amplitudes and periods of rolling and pitching motions are the same. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the IPWR.     

Development of a new decay heat removal system for a high temperature gas-cooled reactor

The heat removal capacity of a RCCS is one of the major parameters limiting the capacity of a HTGR based on a passive safety system. To improve the plant economy of a HTGR, the decay heat removal capacity needs to be improved. For this, a new analysis system of an algebraic method for the performance of various RCCS designs was set up and the heat transfer characteristics and performance of the designs were analyzed. Based on the analysis results, a new passive decay heat removal system with a substantially improved performance, LFDRS was developed. With the new system, one can have an expectation that the heat removal capacity of a HTGR could be doubled.     

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.     

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.     

中国铅基研究堆非能动余热排出系统可靠性分析

铅冷快堆是第四代核能系统推荐堆型之一,世界上多个铅冷快堆采用非能动余热排出系统。非能动系统中作为驱动的自然力与阻力在数量级上接近,由周边环境、材料参数的变化引起的波动不可忽略,因此需要研究非能动系统可靠性。改进了常用的响应面分析法,并应用于中国铅基研究堆反应堆容器空气冷却系统(Reactor Vessel Air Cooling System,RVACS)中。分析中使用流体计算软件Fluent模拟中国铅基研究堆RVACS系统的余热排出过程,研究了输入参数的不确定性对系统可靠性及反应堆安全产生的影响。在大量模拟数据的基础上结合神经网络法建立了输入参数不确定性和结果不确定性之间的映射关系,并以此分析RVACS非能动失效概率。分析结果表明在全厂断电的情况下,RVACS四组并联排热管中的两组也能够可靠地导出反应堆余热。     

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.     

Two phase natural circulation and the heat transfer in the passive residual heat removal system of an integral type reactor

An investigation of the thermal hydraulic characteristics and the natural circulation performance in the passive residual heat removal system (PRHRS) for an integral type reactor have been carried out using the VISTA facility and the calculated results using the MARS code, which is a best estimate system analysis code have been compared with the experimental results. The VISTA facility consists of the primary, secondary, and the PRHRS circuits, to simulate the SMART design verification program. The experimental results show that the fluid is well stabilized in the PRHRS loop and the PRHRS heat exchanger accomplishes well its functions in removing the transferred heat from the primary side in the steam generator as long as the heat exchanger is submerged in the water in the emergency cooldown tank (ECT). The decay heat and the sensible heat can be sufficiently removed from the primary loop with the operation of the PRHRS. The MARS code predicts reasonably well the characteristics of the natural circulation in the PRHRS. From the calculation results, most of the heat transferred from the primary system is removed at the PRHRS heat exchanger by a condensation heat transfer.     

Conceptual design and analysis of a passive residual heat removal system for a 10 MW molten salt reactor experiment

A conceptual design of a passive residual heat removal system was developed for a 10 MW molten salt reactor experiment (MSRE) designed by Oak Ridge National Laboratory (ORNL). The principle, main components and design parameters of the system were presented, and thermal-hydraulic behaviors, such as natural circulation and heat removal ability, were numerically analyzed in the code of C++, especially for the bayonet cooling thimbles. The results show that the system can effectively remove decay heat in the molten salt in an MSRE and has a heat removal rate that approximates to the decay heat generation rate, thus causing the temperature of the molten salt to decrease steadily. The width of the gas gap in the bayonet cooling thimbles has little effect on either the heat exchange or the natural circulation inside the thimbles, while the width of the steam riser, in spite of its slight effect on the heat transfer of the system, greatly influences the natural circulation. With the width of the steam riser increase from 3.6 to 5.1 mm, the mass flow rate increases from 1.9 kg/s to 4.79 kg/s. Finally, three operational schemes were proposed for the passive residual heat removal system, among which that of reducing the bayonet cooling thimbles by three-quarters had the best comprehensive performance.     

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.