Deformable-volume-based thermal-hydraulic model for a tanks-in-series water-gas system in a vertically stratified flow. Part II: Under non-adiabatic conditions

In this paper, heat and mass transfers in a high-pressure water-gas system with a clear water-gas interface are modeled, and the concept of a pseudo-deformable heat structure is described. The model of the pseudo-deformable heat structure and water vapor bulk condensation under nearly saturated conditions has two distinguishing features. First, the heat structure contacting a deformable fluid volume is modeled as if it deforms along with the deformable fluid; due to this feature, the exact average temperature of the heat structure can be predicted. Second, the rate of change of water vapor mass in the gas dome under nearly saturated conditions is determined by a differential equation derived from the Gibbs free energy function. Then, the rate of bulk condensation is determined from the water vapor mass conservation in the gas dome. The proposed model was partially validated by comparison with the results of the MARS3.1 accumulator model at low temperature conditions.     

Experimental validation of the TASS/SMR code for an integral type pressurized water reactor

The transient and setpoint simulation small and medium reactor (TASS/SMR) code has been applied to perform the safety analysis and performance evaluation of an integral type pressurized water reactor. Till now, the code has only been verified by using simplified and analytical problems as well as a reliable system code due to the lack of available experimental data. Recently, several kinds of experiments have been performed by focusing on an identification of the heat transfer characteristics at a heat sink and source, and the thermal hydraulic characteristics and the natural circulation performance in an integral effect test facility. In this paper, the TASS/SMR code has been validated by using the experimental data obtained from a separate effect test facility by focusing on the heat transfer characteristics and an integral effect test facility by focusing on the thermal hydraulic characteristics and the natural circulation performance. According to the validation results of the TASS/SMR code against the separate effect test and the integral effect test, the code predicts the overall variation of the thermal hydraulic parameters well, including the system pressure, fluid temperature, mass flow rate, etc., and it is applicable for the safety analysis and performance evaluation of an integral type pressurized water reactor.     

Experimental validation of the helical steam generator model in the TASS/SMR code

The transients and setpoint simulation/system-integrated modular reactor (TASS/SMR) code has been used to identify the safety margin of a 65-MWt advanced integral reactor and to evaluate its design performance. Although, the code has been verified by using simplified and analytical problems as well as a reliable system code, its validation has not been fully established. This paper deals with a validation of the TASS/SMR code by using two kinds of separate effect tests related to heat transfer at a helically coiled steam generator. The heat transfer experiments were performed by using a full-scale prototype of the steam generator cassette of the advanced integral reactor and a scaled-down steam generator cassette. Analytical results show that the TASS/SMR code predicts the thermal hydraulic parameters, including the system pressure and fluid temperature at the primary and secondary sides of the steam generator cassette, and the heat transfer rate through the steam generator cassette well. The validation results in this study show that the TASS/SMR code is applicable for heat transfer calculations related to the helically coiled steam generator of the advanced integral reactor.     

Development and steady state level experimental validation of TASS/SMR core heat transfer model for the integral reactor SMART

An advanced integral pressurized water reactor (PWR) of a small size (330 MWt) is being developed by the Korea Atomic Energy Research Institute (KAERI). The purposes of the reactor are a sea water desalination and an electricity generation. To enhance its safety, many advanced design concepts are introduced such as a passive residual removal system and a low power density core. For the safety validation of the designed reactor, a system analysis code named TASS/SMR, was developed. TASS/SMR code uses a one dimensional node/path modeling for the thermal hydraulic calculation and point kinetics for the core power calculation. The code also has specific models for the developed integral reactor, such as a helical tube heat transfer model and a passive residual heat transfer model. One of the important models for the safety or performance calculation is the core heat transfer model. The core heat transfer model of TASS/SMR was developed to meet the requirements of the 10 CFR 50 appendix K EM model as well as the realistic models. The developed model was validated with experimental data. The results show that the model predicts the heat transfer phenomena in the reactor core with a reasonable conservatism.     

Validation of a Computational Simulation Method for Evaluating Thermal Stratification in the Reactor Vessel Upper Plenum of Fast Reactors

Validation of a numerical simulation method is carried out for thermal stratification phenomena in the reactor vessel upper plenum of advanced sodium-cooled fast reactors. The study mainly focuses on the fundamental applicability of commercial computational fluid dynamics (CFD) codes as well as an inhouse code to the evaluation of thermal stratification behavior including the simulation methods such as spatial mesh distribution and RANS-type turbulence models in the analyses. Two kinds of thermal stratification tests are used in the validation, which is done for relatively simple- and conventional-type upper plenum geometries with water and sodium as working fluids. Quantitative comparison between the simulation and test results clarifies that when used with a high-order discretization scheme of the convection term, the investigated CFD codes are applicable to evaluations of the basic behaviors of thermal stratification and especially the vertical temperature gradient of the stratification interface, which is important from the viewpoint of structural integrity. No remarkable difference is seen in the simulation results obtained using different RANS turbulence models, namely, the standard kε model, the RNG k-ε model, and the Reynolds stress model. It is further confirmed in a numerical experiment that the distribution of two or more meshes within the stratification interface will lead to accurate simulation of the interface temperature gradient with less than 10% error.     

基于ECC安注热混合试验的LOCUST 1.2分析与验证

在失水事故（LOCA）工况下安注系统投入使用时,蒸汽与安注冷却剂会发生流体热力学混合,热混合过程中冷腿段的冷却是直接影响堆芯再淹没与否的重要因素。中国广核集团有限公司自主研发了一款两相流热工水力系统分析软件LOCUST,可用于压水堆核电厂事故工况的分析计算。基于西安交通大学堆芯应急冷却系统（ECCS-XJTU）试验台架进行的堆芯应急冷却（ECC）安注热混合试验,本文使用LOCUST软件对ECC热混合试验进行了几何建模及计算分析。ECC热混合试验工况主要为不同流量下主管纯蒸汽与安注管过冷水的混合,蒸汽流量为25~125 kg/h,过冷水流量为100~500 kg/h。模拟计算结果和试验结果的对比分析表明：试验段出口质量流量计算值的最大相对误差在13.8%以内,混合后温度计算值的最大相对误差在8%以内,LOCUST在计算高温蒸汽和过冷水混合时的计算结果相对保守,总体上验证了LOCUST在LOCA下两相热混合安注计算的可靠性和准确性。     

Code validation on a passive safety system test with the SMART-ITL facility

A thermal-hydraulic integral effect test facility, SMART-ITL, was constructed to examine the system performance of SMART, a 330 MWt integral type reactor, and to provide data for validation of related thermal-hydraulic models in the system analysis codes. SMART is equipped with various passive systems such as a passive residual heat removal system (PRHRS), a passive safety injection system (PSIS), and an automatic depressurization system (ADS). The PSIS of SMART is made up of four core makeup tanks (CMTs), four safety injection tanks (SITs), and related piping. Over 10 tests have been performed to investigate the behavior of a single train of a PSIS (a CMT and a SIT) in connection with PRHRSs and an ADS. Using a system analysis code, MARS-KS, we validated the experimental results for a representative test. All geometrical and thermal-hydraulic conditions of SMART-ITL were reflected in the code input construction. Through the validation process, several models, including a break flow model, heat transfer models, and pressure drop models, were examined. Overall, the major system parameters were well reproduced.     

Analysis of pressurizer level control system using MAAP4 code

The code initialization effort has been troubling code users for decades for system transient and severe accident analyses using codes such as RETRAN, MAAP4, MAAP5 and MELCOR. The purpose of this work is to demonstrate an approach that could be considered a generic method to address the code initialization problem. This was demonstrated by developing a pressurizer level control model and temperature dependent level control logic in MAAP4 without re-compiling with the source code. The method would enhance the simulation capability and accuracy of a severe accident analysis by transient and severe accident analyses codes. The demonstration case used MAAP4 to show that the adopted proportional-integral controller with the temperature dependent level control logic would reduce its code steady state errors to zero. The subsequent transient response would become more realistic. The proposed method provides a convenient and exemplified approach for code initialization which is applicable to the next generation of codes that couple with the balance of plant models. These codes include the MAAP5 code and others future codes that could simulate the whole plant by a single and elaborate plant model with exhausting component and phenomenological models.     

A numerical study of radiation heat transfer in sodium pool combustion and response surface modeling of luminous flame emissivity

A response surface model of the luminous flame emissivity of sodium pool fire has been proposed for use in safety analysis computer codes of a liquid metal fast reactor. The liquid sodium burns in air resulting in not only heat generation but also release of sodium oxide aerosols of sub-micron diameters. Aerosols levitating in air are radiative and they influence the allocation of combustion heat from the flame to atmospheric gas or sodium pool. The emissivity of the flame needs to be quantified, as it is one of user-specified parameters of the computer codes for the sodium fire analysis. The response surface model of the flame emissivity is developed based on numerical experiments on the physics of mass and heat transfer and behavior of the aerosol. Thermal-hydraulic equations have been solved coupled with aerosol dynamics and chemical reaction. Three influential variables on the emissivity are identified as pool temperature, gas temperature and oxygen molar fraction in the air. It has been found that the emissivity is calculated reasonably as a function of the three variables. The proposed response surface model can be easily employed in the sodium fire analysis codes because it is a simple quadratic expression. For the safety evaluation of the sodium fire, combined use is recommended of the proposed model and the lumped-mass zone model code.     

The development of a fusion specific depletion interface code—FATI

Knowledge of nuclide burn-up within lithium blankets has a crucial part to play in the safety, reliability and feasibility of a fusion reactor. A new depletion interface code is presented called FATI (Fusion Activation and Transport Interface) which interfaces MCNP with FISPACT. The intended primary application of FATI is the simulation of nuclide burn-up within fusion reactor blankets. This paper describes some of the functionality of FATI and presents a comparison of percentage variation of the nuclide atomic densities, for a simple spherical blanket model, calculated by FATI and VESTA. The inventories of the two depletion interface codes differ by less than 1% for lithium and lead isotopes, while H and He isotopes differ by larger amounts due to variations in the methods used to model gas production in FISPACT and the PHOENIX burn-up codes.     

TRAC-PF1程序简介及其移植

ＴＲＡＣ－ＰＦ１程序是压水堆系统安全分析的最佳估算程序［１］。它采用两流体模型处理两相流动，是目前核反应堆系统分析软件中模型比较完善、简化较少的少数软件之一。为了充分利用国外这一先进的系统软件和国内现有的计算设备，将从美国引进的ＩＢＭ版ＴＲＡＣ－ＰＦ１程序经修改移植到了ＣＤＣＮＯＳ／ＶＥ系统及ＳＵＮ工作站上，对不同机器在ＦＯＲＴＲＡＮ语言及汇编语言上的差别进行了修改。对随程序带来的所有标准例题进行了校核计算，结果表明移植是成功的。     

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。     

COSINE子通道均相流分析程序的研发

COSINE一体化软件包的子通道安全分析程序cosSubc基于子通道控制体三维网格模型,采用轴向及横向的热工水力控制方程,包括两流体和均相流两种求解算法。本文介绍了子通道均相流程序的物理模型和数值算法,并用cosSubc均相流程序和参考程序COBRA-TF分别对典型1 000MW核电厂稳态算例进行计算分析,结果表明:cosSubc均相流程序与COBRA-TF吻合较好,具备堆芯子通道的热工水力计算能力。     

A droplet entrainment model based on the force balance of an interfacial wave in two-phase annular flow

Droplets are generated at the interface of annular flow due to an interaction between a liquid film and gas core flow. Therefore, knowledge of the interfacial wave structure is essential for making an accurate prediction of the amount of entrained droplets. A new droplet entrainment model was proposed based on the force balance of interfacial waves in vertical annular flow. An analytic wave shape function was developed reflecting the detailed experimental findings, and was used in the development of a new model. The model was validated using the experimental data reported by Hewitt and Pulling at low pressures and by Keeys et al. at high pressures, which had been performed in adiabatic vertical tubes. The root-mean-square error of the prediction of the amount of entrainment was approximately 27% when the model was implemented into COBRA-TF code, which is approximately 23% less than that determined by the Würtz model. The models proposed by Okawa et al. and Stevanovic et al. were also implemented into COBRA-TF and compared with the proposed model.     

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.     

非能动余热排出换热器换热能力数值分析

基于多孔介质模型,对AP1000非能动余热排出换热器(PRHR-HX)运行初始阶段进行了数值模拟。一回路的入口温度及流量采用RELAP5的计算结果,并以此作为CFD计算的边界条件。采用多孔介质模型处理C型管束区,添加管束区分布阻力。通过商业CFD软件FLUENT计算得到安全壳内置换料水箱(IRWST)侧冷却剂的三维温度及速度分布,通过用户自定义函数UDF完成一回路侧与IRWST侧的耦合换热计算,获得一回路温度分布及换热量。计算结果表明,随着IRWST内冷却剂温度升高,换热器热负荷降低,并出现明显的热分层现象,同时证明采用多孔介质模型与耦合换热计算是分析PRHR/IRWST系统瞬态热工水力特性的有效方法。     

Control rod drop analysis by finite element method using fluid–structure interaction for a pressurized water reactor power plant

The control rod drop analysis is very important for safety analysis. For seismic and loss of coolant accident event, the control rod assemblies shall be capable of traveling from a fully withdrawn position to 90% insertion without any blockage and within specified time and displacement limits. The analysis has been executed by analytical method using in-house code. In this method, several field data are needed. These data are obtained from nuclear, thermal–hydraulic and mechanical design groups, peculiar codes, those work groups need to cooperate together.Following the enhancement of a computer and development of the multi-physics analysis code, a new method for the control rod drop analysis is proposed by finite element method. This analysis model incorporates the structure and fluid parts, termed as a fluid and structure interaction (FSI). Because a control rod is submerged inside a guide tube of a fuel assembly, the FSI boundary condition is applied. In this model, it is assumed that the fluid is incompressible laminar flow. The structures are modeled with the solid elements because there is no deformation due to the fluid flow. The analysis two-dimensional plane model is created in the analysis with considering an axi-symmetric geometry. Therefore, the proposed analysis model will be very simple and the design data from other fields will be unnecessary.The analysis results are compared with those of the in-house code, which have been used for a commercial design. After validation, it is found that the present analysis gives a useful tool in the design of the control rod and fuel assembly.     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

New gas concentration measurement system for the PANDA containment test facility

The increased use of computational fluid dynamics code for analysis and design purposes demands high quality experimental data to validate the simulation codes. Experimental data of fluid stratification and stratification break-up phenomena are generated in the frame of OECD/SETH-2 project at the PANDA facility. A new gas concentration measurement system is presented that is based on speed of sound measurements. Speed of sound in gas mixtures is a unique function of the temperature and the fractions of the components and therefore can be used to compute the fractions in varying compositions. The speed of sound is measured indirectly measuring the time of flight of an ultrasound pulse between two ultrasound transducers. The 30 transducers employed proved to be able to withstand the unfavorable conditions inside the facility with temperatures of up to 110 °C and steam that may condense. A frame rate (1 frame = each transducer has been excited) of 1.6 Hz and a helium fraction resolution of 1.5% in steam are achieved.     

System Code Calculation and Experimental Verification for Double-hole Steam Jet Condensation Heat Transfer Process Based on ADS1 3 Scaled Model

The steam direct contact condensation of high-temperature steam in sub-cooled water is an important way to reduce the temperature and pressure in the primary circuit in the third generation of advanced pressurized water reactors such as AP1000 and CAP1400 in the event of accidental overpressure. Based on the system codes of RELAP5 and COSINE, the process of saturated steam injecting into large volume sub-cooled water through a double-hole nozzle was modeled, calculated and analyzed. The temperature distributions along the axial direction of the high-temperature steam ejected from the nozzle were obtained. At the same time, the visual experiments of steam jet condensation were performed. The thermocouple matrix and high-speed camera were used to measure the key thermal-hydraulic parameters to obtain the temperature distributions along the steam plume and the flow patterns of the steam jet, which were used to verify the accuracy of the system code to simulate the process of steam spraying and condensation. The results show that the system code RELAP5 can basically simulate the general trend of ADS steam condensation process under the simplified model. The average error of the simulation results is 2.97% compared with the experimental results. In addition, the COSINE code was used to further modify and improve the model of the spraying condensation process. Considering the influence of the overall flow in the water tank on the condensation characteristics, the simulation results fit well with the experimental results, with an average error of 1.89%. However, the actual double-hole spraying process is complex and has obvious three-dimensional characteristics, so the relevant condensation heat transfer model in the system code still needs to be further improved to simulate its local condensation characteristics more accurately.