Thermohydraulic characteristics of coolant gas flow through a control rod channel in a very high temperature gas cooled reactor

A calculation code was developed to evaluate the thermohydraulic performance of a coolant flow through a control rod channel in a very high temperature gas cooled reactor (VHTR) and a high temperature engineering test reactor (HTTR). A one-dimensional flow network model was employed in the present calculation code. The calculated results agreed well with the experimental ones on the flow rate distribution and the total pressure loss in an isothermal coolant flow. The thermohydraulic characteristics of the HTTR control rod channel were evaluated by the code under various conditions, including the normal operating conditions of a HTTR.     

Comprehensive Analysis of Mechanical and Thermohydraulic Processes in the Graphite Masonry of Channel Rectors during an Accidental Rupture of a Fuel Channel

A methodology and preliminary results of a computational analysis of the processes occurring in the graphite masonry in RBMK channel reactors during the rupture of a fuel channel as a result of accidental overheating are presented. The methodology for the computational analysis is implemented using the U_STACK code, simulating the thermohydraulic and mechanical processes occurring in masonry with varying geometry in continuous coupling with the processes in the circulation loop and accident containment system. The deformation and rupture of a pipe in the damaged channel and coolant efflux are calculated using the prescribed accident scenario. After channel rupture, the parameters of the medium, the displacement of graphite blocks, and the sagging of channel pipes in the entire volume of the core are calculated. As a result, the additional force loads on neighboring fuel channels in the rupture zone, the reactor case, and the top and bottom plates are estimated. The objective of the calculations performed with the U_STACK code are assessment of the integrity of the components of the reactor core and substantiation of the impossibility of multiple ruptures occurring in the reactor fuel channels.     

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。     

某游泳池式反应堆全厂断电事故分析

采用RELAP5/MOD3.3程序对某游泳池式反应堆的全厂断电事故工况进行计算,对堆内冷却剂流动逆转过程进行了模拟计算,并对全厂断电事故下堆芯漏流和组件间流动等相关参数对流动逆转的影响进行了深入分析。结果表明,该反应堆在失去全部强迫流动的情况下,能形成足够的自然循环流量,以导出堆芯余热,燃料组件不会发生破损。     

Thermal-hydraulic modeling of flow inversion in a research reactor

The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic–thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge–Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor.     

Evolutionary design of reactor vessel assembly for liquid metal cooled battery

The Battery Omnibus Reactor Integral System (BORIS) is being developed as a multipurpose integral fast reactor at the Seoul National University. This paper focuses on developing design methodology for optimizing geometry of the liquid metal cooled reactor vessel assembly. The key design parameters and constraints are chosen considering technical specifications such as thermal limits and manufacturing difficulties. The evolution strategy is adopted in optimizing the geometry. Two objective functions are selected based upon economic and thermohydraulic reasons. Optimization is carried out in the following steps. First, selected design values are supplied to the momentum integral model code to evaluate steady-state mass flow rate and coolant temperature distribution of the reactor vessel assembly utilizing the thermodynamic boundary condition on heat exchanger calculated by the thermodynamics code. Second, the objective function values are calculated and compared against the previous results. The steps are repeated until an optimum value is obtained. Results of the improved design of the reactor vessel assembly are presented and their characteristics are discussed.     

Reactivity insertion limits in a typical pool-type research reactor cooled by natural circulation

The course of reactivity insertion in a pool type research reactor, with scram disabled under natural circulation condition is numerically investigated. The analyses were performed by a coupled kinetic–thermal–hydraulic computer code developed specifically for this task. The 10-MW IAEA MTR research reactor was subjected to unprotected reactivity insertion (step and ramp) for both low and high-enriched fuel with continuous reactivity feedback due to coolant and fuel temperature effects. In general, it was found that the power, core mass flow rate and clad temperature under fully established natural circulation are higher for high-enriched fuel than for low enriched fuel. This is unlike the case of decay heat removal, where equal clad temperatures are reported for both fuels. The analysis of reactivity represented by the maximum insertion of positive reactivity ($0.73) demonstrated the high inherent safety features of MTR-type research reactor. Even in the case of total excess reactivity without scram, the high reactivity feedbacks of fuel and moderator temperatures limit the power excursion and avoid consequently escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation. The code can also be modified to provide an accurate capability for the analyses of research reactor transients under forced convection.     

Thermohydraulic studies of sodium-cooled reactor facilities

Thermohydraulic studies of reactor facilities with fast reactors are complex experimentally and computationally. Extensive experimental data are obtained on the velocity and temperature profiles, hydrodynamic resistance and heat emission, initial heat section, and interchannel mixing of the coolant in the fuel assemblies. These are used to develop engineering methods of performing thermohydraulic calculations of fuel assemblies as well as computational compute codes. The particulars of the hydrodynamics and heat transfer in intermediate heat exchangers and steam generators of reactor facilities with fast reactors are studied. This has made it possible to validate their thermohydraulic characteristics.     

Calculating decay heat removal rates via natural circulation along variable heat flux reactor fuel plates

A recently developed integral technique is applied to natural convection cooling along test reactor fuel plates. The technique is demonstrated for water and air flow. In the case of air flow, the process is characterized by a large temperature rise along the fuel channel, thereby rendering the commonly applied Boussinesq approximation invalid. This case is a heat transfer problem of particular interest in accident analyses such as determining the level of decay heat dissipation possible, without exceeding the melting temperature of the fuel, subsequent to a hypothetical loss of primary coolant.     

Natural circulation in parallel vertical channels with different heat inputs

When forced circulation of the coolant is lost in a gas cooled reactor, natural circulation takes place in the reactor core. Natural circulation in the many parallel vertical channels with different heat inputs is quite complicated because the flow rate and direction depend on the time history of the heat input of each channel. Since only a few previous studies were concerned with this problem, an experimental and analytical study was performed to investigate the natural circulation in an array of parallel channels with different heat inputs.The experimental apparatus consisted of four vertical channels filled with water. Each channel was furnished with a heater pin and heated by various combinations of time-varying heat inputs. Flow velocity of the water was measured, and an abrupt change of the flow direction was observed. An analytical code predicted well the experimental results.     

Steady-state thermal-hydraulic analysis of the Moroccan TRIGA MARK II reactor by using PARET/ANL and COOLOD-N2 codes

The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maâmora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor.     

Visual Study of Flow Behavior in Upper Plenum during Simulated Reflood Phase of PWR-LOCA

Fuel temperature is an important parameter in reactor safety. However, temperatures of fuel sub-assemblies in power reactors usually cannot be directly measured. A means therefore has been devised to use heat removed by coolant from fuel sub-assemblies following a reactor trip to estimate the pre-trip average fuel temperatures and fuel-to-coolant heat transfer coefficients. Sub-assembly coolant mass flow and outlet temperature measurements are used to determine the heat removed. After accounting for the contribution from release of stored heat from non-fuel components and from decay heat, the contribution due to release of fuel-stored heat alone during the trip transient can be inferred. Average fuel temperatures and fuel-to-coolant heat transfer coefficients of the FBR “MONJU” at 45% thermal power were estimated using this energy balance approach. The trip-test data derived estimates compared favorably with plant simulation code calculated values.     

超临界水冷堆堆芯子通道稳态热工分析

超临界水冷堆(SCWR)作为6种第四代未来堆型中唯一的水冷堆,冷却剂出口温度可达500℃,具有良好的经济性.本文采用改进的COBRA-IV程序对超临界水冷堆方形组件子通道进行稳态热工分析.对计算结果进行分析可知:减小慢化剂通道中给水质量流量份额和加大慢化剂通道与相邻子通道之间的热阻,可以降低热管焓升,后者还可以得到较好的慢化效果.通过热通道的传热恶化分析发现,超临界水冷堆的设计不能避免传热恶化,必须精确计算传热恶化条件下的包壳温度才能确定包壳能否保证其完整性.     

海洋条件下反应堆热工水力参数的子通道计算

针对海洋条件下反应堆的子通道热工水力分析,建立了海洋运动附加力模型和瞬态入口边界,将起伏、摇摆及复合运动的附加力关系式用于子通道模型的轴向和横向动量方程,并应用到COBRAⅢC程序将其改造为适应海洋条件的反应堆子通道分析程序。作为验证,计算了加热实验通道和"奥陆"堆在起伏运动情况下热通道的临界热流密度比(CHFR)、出口空泡份额和冷却剂流量,并与文献结果对比。还详细计算了"奥陆"堆在起伏、不同摇摆中心和复合运动情况下,热通道的CHFR和不同位置子通道出口的热工水力参数。研究表明:海洋条件下反应堆的子通道热工水力参数随运动呈周期性变化;起伏运动对子通道的压降影响较大,摇摆运动对子通道冷却剂的流量和温度影响较大。     

Numerical research on natural convection in molten salt reactor with non-uniformly distributed volumetric heat generation

Molten salt reactor is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The present work presents a numerical investigation on natural convection with non-uniform heat generation through which the heat generated by the fluid fuel is removed out of the core region when the reactor is under post-accident condition or zero-power condition. The two-group neutron diffusion equation is applied to calculated neutron flux distribution, which leads to non-uniform heat generation. The SIMPLER algorithm is used to calculate natural convective heat transfer rate with isothermal or adiabatic rigid walls. These two models are coupled through the temperature field and heat sources. The peculiarities of natural convection with non-uniform heat generation are investigated in a range of Ra numbers (10³ ∼ 10⁷) for the laminar regime of fluid motion. In addition, the numerical results are also compared with those containing uniform heat generation.     

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.     

Reactor core thermohydraulic transients with thermohydrodynamic coupling

A study of the reactor core thermohydraulics in an LMFBR has been performed for the strongly coupled thermo-hydrodynamic transients. A numerical method to solve the coupled energy-momentum equations among multichannels in a core is presented and the computer code ORIFS-TRANSIENT has been developed.The results of sample calculations for a flow coastdown transient to natural circulation following a reactor scram in a typical loop-type LMFBR are as follows: (1) the inter-subassembly coolant flow redistribution due to buoyancy forces is significant under the low flow condition, such as natural circulation; (2) the maximum coolant temperature was decreased by about 80°C (corresponding to about 22% in terms of hot channel factor) due to the flow redistribution; (3) due to thermohydrodynamic coupling between upper plenum and other regions, the maximum coolant temperature was decreased by about 9°C; (4) due to inter-subassembly heat redistribution, the maximum coolant temperature was increased by about 7°C.     

Natural circulation capability of Pb-Bi cooled fast reactor: PEACER

First-principle calculations were performed to analyze the natural circulation heat removal from the core of a liquid metal reactor (LMR). The lead-bismuth (Pb-Bi) was chosen as the primary coolant for the LMR system. From the single channel analysis the temperature and the pressure drop are calculated along the fuel assembly. The total pressure drop of the core is less than 100kPa due to the large pitch-to-diameter ratio and the small height of the fuel pin. The natural circulation potential is a key characteristics of the LMR design. The steady-state momentum and energy equations are solved along the primary coolant path. The calculations are divided into two parts: an analytical model and a one-dimensional lumped parameter flow loop model. Results of the analytical model indicate that the elevation difference of 4.5m between thermal centers of the core and the steam generators could remove as much as 10% of the nominal operating reactor power. The flow loop model yielded the total pressure drop and the natural circulation heat removal capacity.     

A subchannel analysis of DUPIC fuel bundle for the CANDU reactor

Thermal characteristics of the reference DUPIC fuel has been studied for its feasibility of loading in the CANDU reactor. Half of the DUPIC fuel bundle has been modeled for a subchannel analysis of the ASSERT-IV Code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions in subchannels of the fuel bundle, it is found that the gravity effect may be pronounced in the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. The asymmetric distribution of the coolant in the fuel bundle is known to be undesirable since the minimum critical heat flux ratio can be reduced for a given value of the channel flow rate. On the other hand, the central region of the DUPIC fuel bundle has been found to be cooled more efficiently than that of the standard fuel bundle in the subcooled and the local boiling regimes due to the fuel geometry and the fuel element power changes. Based upon the subchannel modeling used in this study, the location of minimum critical heat flux ratio in the DUPIC fuel bundle turned out to be very similar to that of the standard fuel when the equivalent values of channel power and channel flow rate are used. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the subchannel-wise mixture enthalpy and void fraction peaks are located in the peripheral region of the DUPIC fuel bundle while those are located in the central region of the standard CANDU fuel bundle. Reduced values of the channel flow rates were used to study the effect of channel flow rate variation. The effect of the channel flow reduction on different thermal-hydraulic parameters have been discussed. This study shows that the subchannel analysis for the horizontal flow is very informative in developing new fuel for the CANDU reactor.     

三环路压水堆压力容器上腔室交混矩阵数值研究

使用STAR-CCM+软件对三环路压水堆压力容器上腔室流场进行了大规模、精细化三维数值模拟,并采用组分跟踪方法分别对157个燃料组件出口冷却剂流动进行计算,构造了一个具有3×157个元素的“上腔室交混矩阵”,用该矩阵即可定量、精确地描述冷却剂从堆芯流出后,经上腔室内交混并再分配到各热管道的复杂流动过程。研究发现堆芯流出的冷却剂在压力容器上腔室内的交混是并不充分的,径向上不同位置燃料组件流出的冷却剂会在上腔室同热管道的接口区域存在明显的对应关系,而燃料组件径向功率分布的差异必然导致热管道中冷却剂热分层现象的产生。