Thermal-hydraulics analysis during fuel element failure in an operating PWR

The fuel element failure in an operating pressurized water reactor (PWR), including fuel element breaks, has an effect on the operation safety of PWR. In this paper, the RELAP5 model of the fuel element failure is established for the safety analysis. The RELAP5 time step sensitivity analyses for the element pre-break steady and post-break transient simulation are carried out. And the variations of main thermal-hydraulics parameters related to the fuel element break are quantitatively studied, which include the internal gap pressure and the maximum fuel pellet temperature as well as the releasement of noncondensables in the gap. It is found that (1) the results by the RELAP5 code is very sensitive to the time step in a volume system with the noncondensables, and the time step sensitivity analysis is necessary if the effective time step range is unknown, (2) the larger the break area is, the more quickly the gap pressure increases and the maximum pellet temperature reaches to the stable value, (3) when the gap pressure increases and reaches to the coolant pressure, at the break the liquid inflow from coolant to gap will be turned to the vapor outflow from gap to coolant, (4) during the failure transient, the gap thermal conductivity experiences a sharp decrease in the break instant, which results in the decrease of heat transferred to cladding and the sharp decrease of cladding temperature as well as the sharp increase of minimum departure from nucleate boiling ratio (MDNBR). These conclusions can provide the basic for the operation safety analysis of PWR during the fuel element failure.     

压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

AP1000主回路系统热工水力瞬态计算程序RETAC的开发

针对先进压水堆AP1000的具体结构和运行特点,采用FORTRAN程序设计语言,自主开发了用于AP1000主回路系统热工水力瞬态计算的微机型程序RETAC（REactorTransientAnalysisCode）。利用程序对AP1000失流事故进行分析,得到了堆芯燃料中心最高温度、最小偏离泡核沸腾比(MDNBR)、稳压器压力、水位及蒸汽发生器二次侧压力、水位等主要系统参数的瞬态特性。分析结果表明,在失流事故初期阶段,堆芯热通道燃料中心最高温度和MDNBR不超出规定限值,满足安全准则要求。RETAC完全采用模块化编程,便于移植和二次开发,可为后续开发自主知识产权的大功率压水堆安全分析程序提供借鉴。     

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.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

AP1000自动降压系统误开启事故仿真分析

针对AP1000的具体结构和运行特点,采用FORTRAN程序设计语言,开发了AP1000瞬态热工水力计算程序RETAC。利用RETAC对AP1000自动降压系统(ADS)误开启事故进行仿真分析,得到稳压器压力、堆芯归一化热功率、堆芯归一化流量、堆芯平均温度、燃料中心最高温度和最小偏离核态沸腾比（MDNBR）等主要系统参数的响应特性。分析结果表明,在稳压器低压停堆保护的作用下,燃料中心最高温度和MDNBR未超出规定限值,满足安全准则要求。并将计算结果与美国西屋公司AP1000分析软件LOFTRAN的计算结果进行对比,对比趋势符合良好,证明了RETAC建模和自动降压系统临界流模型计算的合理性。     

Post DNB Heat Transfer Experiments for PWR Fuel Assemblies

Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis.     

基于SCWR堆芯结构的子通道程序开发与应用

为能够对超临界水堆（SCWR）堆芯进行子通道分析,开发了新的子通道分析程序SABER。该程序在COBRA程序的基础上改进了网格结构和热传导模型,加入了新的边界条件和水物性模块,以适用于SCWR慢谱燃料组件的子通道分析。为评估程序的适用性,采用该程序对SCWR堆芯概念设计中的慢谱燃料组件进行子通道建模,并进行稳态计算。结果表明,该程序能够用于SCWR堆芯的子通道计算分析,并较好地解决了慢谱组件计算中慢化通道和冷却通道间的热耦合及逆向流动的模拟问题。     

超临界水冷堆双排燃料组件子通道分析

研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.     

石墨慢化通道式熔盐堆的稳态热工水力计算模型

针对石墨慢化通道式熔盐堆的堆芯结构,基于COMSOL Multiphysics程序和MATLAB程序建立了堆芯稳态热工水力学计算模型。该模型对堆芯内固体区域的温度分布采用三维热传导方程进行模拟,对通道内熔盐温度采用一维单相流体模型进行计算。固体区域与熔盐通过熔盐通道壁面的对流换热边界建立热耦合。该模型基于平行通道压力损失相等的原则,分配堆芯内各熔盐通道的流量。通过对比RELAP5程序的计算结果,验证了模型对温度和流量分配计算的正确性。针对2 MWt 液态燃料熔盐堆的一种概念设计,分析了堆芯内三维温度分布和通道间流量分配。该模型可精确计算通道式熔盐堆堆芯内稳态温度分布和流量分配,对堆芯的热工水力学设计具有重要意义。     

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。     

FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.     

Comparisons of the RALIZA-2/02 two-phase flow model with experimental data

The RALIZA-2 computer program was designed for thermal-hydraulic analysis of flow channel and fuel element of PWR/BWR at steady-state and transient conditions. A nonhomogeneous, nonequilibrium model of a two-phase flow and a two-dimensional heat conduction model of fuel pin are used in the program. A fully implicit integration scheme for both models is used. The steady-state constitutive correlations set is used. The void fraction, pre- and post-DNB heat transfer mechanism are compared with data. Also a loss of flow experiment was calculated and compared with nuclear heated rod bundle experimental data for typical PWRs. A very good agreement was obtained.     

Sodium boiling experiments in the THORS facility

Experiments have been performed with 19- and 61-pin test assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility at the Oak Ridge National Laboratory (ORNL) since 1971. The THORS Facility is a high-temperature sodium system operated for the US Liquid-Metal Fast Breeder Reactor (LMFBR) Safety Program. The facility is used primarily for testing simulated LMFBR fuel subassemblies (pin (bundles). High-performance, electrically heated fuel pin simulators (FPSs) duplicate the heat generating capabilities and the dimensional characteristics of the nuclear fuel pins. A number of test bundles have been built and operated to obtain base thermal-hydraulic data, inlet and heated zone blockage data, and transient boiling data. Five of these bundles have been operated under two-phase conditions. Sodium boiling for periods up to twelve minutes were sustained in one bundle. (The lengths of the periods were limited only by automatic data recording capability). Clad dryout occurred in several tests. Tests were run at widely varying conditions of flow and power density. Testing with nonuniform power distribution across the bundle was also a part of the program.A 19-pin bundle with 12 peripheral guard heaters and a 6-subchannel blockage around the center pin in the heated zone was tested. The test program for this bundle was designed to determine if local boiling in the wake of the blockage propagates radially or axially during quasi-steady-state conditions. Post-test inspection revealed that significant helical distortion of the FPSs occurred in the vicinity of the blockage plate. This distortion probably influenced the boiling behavior. In the more severe tests, boiling initiated at the outlet of the heated zone and propagated radially into the unblocked subchannels after it had progressed upstream to the blockage. The subchannel analysis codes, SABRE and COBRA, accurately predict the extent of the boiling region.Experimental and analytical studies of sodium boiling behavior in unblocked 19- and 61-pin bundles indicate that cooling can be maintained for a significant period of time beyond boiling inception in a flow-power transient. Quasi-steady-state boiling occurred under natural-convection conditions.Investigations of the temperature data indicate that the thermal-hydraulic behavior during boiling transients is determined by two-dimensional effects, and that one-dimensional models cannot accurately predict the important phenomena associated with sodium boiling in test bundles. The subchannel code SABRE-2P (with a simple two-phase multiplier boiling model) and the two-region equilibrium mixture code THORAX (developed at ORNL) accurately predict the two-dimensional behavior between boiling inception and dryout.Extrapolation of the data from the smaller bundle tests to full-size fuel assemblies shows that the time between boiling inception and dryout would be lower for a 217-pin bundle than for a 61-pin bundle for a comparable transient. However, the time delay would still be significant, especially in a heterogeneous reactor core.     

Reactor cavity and ALS thermal-hydraulic evaluation in the case of fuel channels ruptures at Ignalina NPP

This paper presents the work analysis of the thermal-hydraulic parameters behavior in the RBMK-1500 reactor cavity (RC) and other connected volumes in the case of fuel channels ruptures. The analysis is performed with CONTAIN code using the models of accident localization system (ALS) and reactor cavity venting system (RCVS). The RCVS capacity is assessed and expressed as a number of ruptured fuel channels at which the integrity of RC is maintained. The uncertainty analysis of pressure behavior in RC during multiple fuel channel rupture was performed. The initial and boundary conditions and the code models were selected and their influence on the results is estimated.Calculation of coolant mass and energy release to the reactor cavity in case of fuel channels rupture performed using the main circulation circuit model of Ignalina NPP, which was developed by employing state-of-the-art code RELAP5/MOD3.2 [Fletcher et al., RELAP5/MOD3 code manual user’s guidelines, Idaho National Engineering Lab., NUREG/CR-5535 (1992)]. These results were applied further as the initial data for the analysis of the thermal-hydraulic parameters behavior in the affected compartments employing CONTAIN code.     

轻水堆燃料关联核素的特征比值与乏燃料堆型的辨别

应用燃耗分析程序MCCOOR计算压水堆和沸水堆的栅元模型满功率运行时2种不同初始燃料富集度情形下不同燃耗深度的燃料核素成分，分析轻水堆燃料的关联核素比值与燃耗深度的关系，获得了轻水堆燃料的关联核素特征比值，并探索了由乏燃料相应核素特征比值确定其堆型的可行性。      

Features and performance of a coupled three-dimensional thermal-hydraulic/kinetics TRAC-PF1/NEM pressurized water reactor (PWR) analysis code

This paper summarizes the current status of the Pennsylvania State University (PSU) version of the coupled three-dimensional (3-D) thermal-hydraulic/kinetics TRAC-PF1/NEM code for pressurized water reactor (PWR) transient and accident analysis and describes applications to reactivity insertion accident (RIA) simulations as well as recent developments. The TRAC-PF1/NEM methodology utilizes closely coupled 3-D thermal-hydraulics and 3-D core neutronics transient models to simulate the vessel and a 1-D simulation of the primary system. An efficient and flexible cross-section generation procedure was developed and implemented into TRAC-PF1/NEM. These features make the coupled code capable of modeling PWR reactivity transients, including boron dilution transients, in a reasonable amount of computer time. Three-dimensional studies on hot zero power (HZP) rod ejection and main steam line break (MSLB) transients in a PWR, as well as a large break loss-of-coolant-accident (LBLOCA) and boron dilution transients, were accomplished using TRAC-PF1/NEM. The results obtained demonstrate that this code is appropriate for analysis of the space-dependent neutronics and thermal-hydraulic coupled phenomena related to most current safety issues.     

板型燃料元件反应堆瞬态热工水力分析程序的开发与验证

针对核动力系统瞬态分析的需求,建立板型燃料反应堆的热工水力数学物理模型,开发了具有自主知识产权的核动力系统瞬态热工水力分析程序SYSTRAN,并采用中国先进研究堆（CARR堆）的设计工况和国际原子能机构（IAEA）基准题的堵流瞬态数据对程序进行了验证。计算结果表明,堆芯流量分配、出口温度等关键参数与验证数据吻合良好,初步证明了本程序适用于板型燃料反应堆系统瞬态热工水力分析。