石墨粉尘通过高温气冷堆堆芯球床结构的运动行为研究

高温气冷堆在运行过程中产生带有放射性的石墨粉尘,对反应堆的运行安全和环境安全造成一定影响。本文选取二维球床流场,采用离散相模型分析了堆芯球床结构对石墨粉尘颗粒的扩散和沉积的影响。计算结果表明:球床结构能有效阻碍石墨粉尘颗粒的扩散;沉积在球床结构上的石墨粉尘颗粒数目随堆芯内氦气流速的增加而增大,而由于受到颗粒惯性及热泳力的作用其增长趋势逐渐放缓;石墨粉尘颗粒在球床结构上的沉积效率随粒径的逐渐增加呈现"几乎不变-快速增长-缓速增长"的态势。     

HTR-PM高温气冷示范堆堆芯石墨粉尘产生量估算

在球床式高温气冷堆堆芯内,影响石墨球摩擦磨损率的关键条件为载荷与温度。此前,中国辐射防护研究院研究了载荷对石墨球摩擦磨损性能的影响,得到了石墨球磨损率与载荷的关系。本文在此基础上进一步研究了温度对石墨球磨损率的影响,通过拟合得到了石墨球磨损率与石墨球所受载荷、温度之间的关系式,结合HTR-PM高温气冷示范堆内燃料元件所受载荷和温度的分布情况,计算得出石墨球之间摩擦产生的石墨粉尘量约为14.01 g/d(5.1 kg/a)。     

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF₂ (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.     

Nuclear graphite friction properties and the influence of friction properties on the pebble bed

Recently, the role of friction and wear in the safety of pebble bed reactors was reconsidered. The friction properties of graphite were analyzed in this paper. In the first part of this paper, we investigated the past research of graphite and analyzed the influence of environment gas, temperature and radiation on friction coefficient. The variation laws of friction coefficient with the effect factors were given, especially in helium environment (including pure helium and HTGR helium). In the second part, the mechanical behavior of pebble bed is discussed while considering the influence of the friction properties of graphite. The results indicated that the fuel elements cannot be crushed failure even though the fuel elements have a large friction coefficient.     

Air ingress benchmarking with computational fluid dynamics analysis

The air ingress accident is a complicated accident scenario that may limit the deployment of high-temperature gas reactors. The complexity of this accident scenario is compounded by multiple physical phenomena that are involved in the air ingress event. These include diffusion, natural circulation, and complex chemical reactions with graphite and oxygen. In an attempt to better understand the phenomenon, the FLUENT-6 computational fluid dynamics code was used to assess two air ingress experiments. The first was the Japanese series of tests performed in the early 1990s by Takeda and Hishida. These separate effects tests were conducted to understand and model a multi-component experiment in which all three processes were included with the introduction of air in a heated graphite column. MIT used the FLUENT code to benchmark these series of tests with quite good results. These tests are generically applicable to prismatic reactors and the lower reflector regions of pebble-bed reactors. The second series of tests were performed at the NACOK facility for pebble bed reactors as reported by Kuhlmann [Kuhlmann, M.B., 1999. Experiments to investigate flow transfer and graphite corrosion in case of air ingress accidents in a high-temperature reactor]. These tests were aimed at understanding natural circulation of pebble bed reactors by simulating hot and cold legs of these reactors. The FLUENT code was also successfully used to simulate these tests. The results of these benchmarks and the findings will be presented.     

HTR-10 full core first criticality analysis with MCNP

Experimental facilities like HTR-10, HTTR, and ASTRA serve as the source of information for the currently designed high temperature gas-cooled nuclear reactors. It is also desired to verify the existing codes against the data obtained in such facilities. In this study, first criticality calculations of a pebble bed gas-cooled reactor, HTR-10, is performed with MCNP-4B, a code system for Monte Carlo particle transport simulation. HTR-10 has rather unique characteristics in terms of the randomness in geometry as in the case of all pebble bed reactors. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP. Modeling details are discussed with necessary simplifications. Results obtained by Monte Carlo simulations are compared with available data. It is observed that Monte Carlo simulations yield sufficiently accurate results in terms of initial criticality of the HTR-10 reactor.     

氟盐冷却球床堆与HTR-10和MSRE的定量相似性分析

氟盐冷却球床堆是当前国际上一种新的研究堆型,尚无已经建造完成的反应堆,因此,选择相似且具有运行经验的反应堆作为基准题有助于堆芯核设计软件适用性分析。利用国际上常采用的相似性分析软件,可对熔盐实验堆(Molten Salt Reactor Experiment,MSRE)及10 MW高温气冷堆(10 MW high-temperature gas-cooled test reactor,HTR-10)与氟盐冷却球床堆的相似性进行分析,定量判断它们作为基准题的合理性。分析结果表明,MSRE和氟盐冷却球床堆的能谱峰位能量接近且堆内元素种类相近,二者相似程度较高;常温临界HTR-10和氟盐冷却球床堆冷却剂不同,且能谱峰位能量差异较大,二者相似程度较低。因此,MSRE是氟盐冷却球床堆中子物理设计软件较理想的基准题。     

Development and testing of analytical models for the pebble bed type HTRs

The pebble bed type gas cooled high temperature reactor (HTR) appears to be a good candidate for the next generation nuclear reactor technology. These reactors have unique characteristics in terms of the randomness in geometry, and require special techniques to analyze their systems. This study includes activities concerning the testing of computational tools and the qualification of models. Indeed, it is essential that the validated analytical tools be available to the research community. From this viewpoint codes like MCNP, ORIGEN and RELAP5, which have been used in nuclear industry for many years, are selected to identify and develop new capabilities needed to support HTR analysis. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP. The coupled MCNP-ORIGEN code is used to estimate the burnup and the refuelling scheme. Results obtained from Monte Carlo analysis are interfaced with RELAP5 to analyze the thermal hydraulics and safety characteristics of the reactor. New models and methodologies are developed for several past and present experimental and prototypical facilities that were based on HTR pebble bed concepts. The calculated results are compared with available experimental data and theoretical evaluations showing very good agreement. The ultimate goal of the validation of the computer codes for pebble bed HTR applications is to acquire and reinforce the capability of these general purpose computer codes for performing HTR core design and optimization studies.     

高温气冷堆侧反射层纵向窄缝中的旁流研究

高温气冷堆的堆内构件由大量石墨块与碳砖构成,石墨块之间的窄缝会造成堆芯旁流,影响堆芯的流量与温度分布,需细致研究。石墨侧反射层有垂直方向的窄缝,是主要的旁流通道之一,氦气可能从冷氦联箱通过这些窄缝直接流入热氦联箱,也会与球床中的氦气发生横向交混。通过对球床流动及垂直窄缝中的旁流建立流体网络模型,分析了横向交混对窄缝旁流的影响,并讨论在不同窄缝大小及窄缝分布情况下旁流的变化规律。研究结果表明,球床边缘的氦气横向交混对旁流量影响较为明显,需在旁流分析中考虑,尺寸较大的窄缝对整个旁流的影响较为明显,窄缝尺寸较大时,堆芯的旁流量也更大。     

基于折射率匹配方法的球床三维重构

固态钍基熔盐堆(Thorium-based Molten Salt Reactor with Solid Fuel,TMSR-SF)是第四代核反应堆堆型之一,它融合了高温气冷堆的石墨基质包覆颗粒燃料球技术和熔盐堆的高温熔盐冷却剂技术。堆芯的物理设计和几何设计依赖于燃料球在堆芯中的堆积因子,为研究球床堆堆芯模型内燃料球的堆积三维结构,本文提出基于折射率匹配的方法对球床进行三维重构的方案,并通过初步的模拟实验对程序进行验证,旨在探索该方法在球床三维重构中的可行性。针对三维重构中的一系列关键问题进行阐释,并提出相应的解决方案;同时给出了三维重构方案的完整流程,并计算出了衡量三维重构精确度的度量值:直径重叠量。最后,搭建了一个小型规则排布的球床实验装置,通过折射率匹配技术开展球床可视化实验以探索该方案在球床三维重构中的精确度,并说明该方法的可行性。试验结果表明,颗粒间平均重叠量为1.43 mm,重构精度有待提高,重构方法有待改进。     

Development of an innovative detection system for pebble bed kinematic studies

Pebble bed reactors enable the circulation of pebble fuel elements when the reactors are in operation.This unique design helps to optimize the burnup and power distribution, reduces the excessive reactivity of the reactor,and provides a mean to identify and segregate damaged fuel elements during operation. The movement of the pebbles in the core, or the kinematics of the pebble bed,significantly affect the above features and is not fully understood. We designed and built a detection system that can measure 3-axis acceleration, 3-axis angular velocity,3-axis rotation angles, and vibration and temperature of multiple pebbles anywhere in the pebble bed. This system uses pebble-shaped detectors that can flow with other pebbles and does not disturb the pebble movement. We used new technologies to enable instant response, precise measurement, and simultaneous collection of data from a large number of detectors. Our tests show that the detection system has a negligible zero drift and the accuracy is better than the designed value. The residence time of the pebbles in a moving pebble bed was also measured using the system.     

杂质硼在高温气冷堆中的燃耗特性

在球床式高温气冷堆的堆芯和石墨反射层中,不可避免地含有少量杂质硼。硼杂质的存在及其燃耗会对反应堆的反应性产生影响。对于多次通过的球床堆芯,根据燃料元件的运行历史计算所有元件的硼燃耗,对于中子注量率差别较大的反射层,分区计算了硼燃耗。再采用微扰理论,计算燃耗过程中硼反应性价值的变化。计算结果表明,硼杂质燃耗很快,因此,硼杂质对反应性的影响降低很快。     

Individual pebble temperature peaking factor due to local pebble arrangement in a pebble bed reactor core

Scientists at the German AVR pebble bed nuclear reactor discovered that the surface temperature of some of the pebbles in the AVR core were at least 200 K higher than previously predicted by reactor core analysis calculations. The goal of this research paper is to determine whether a similar unexpected fuel temperature increase of 200 K can be attributed solely or mostly to elevated power production resulting from exceptional configurations of pebbles. If it were caused by excessive pebble-to-pebble local power peaking, there could be implications for the need for core physics monitoring which is not now being considered for pebble bed reactors. The PBMR-400 core design was used as the basis for evaluating pebble bed reactor safety. Through exhaustive Monte Carlo modeling of a PBMR-400 pebble environment, no simple pebble-to-pebble burn-up conditions were found to cause a sufficiently high local power peaking to lead to a 200 K temperature increase. Simple thermal hydraulics analysis was performed which showed that a significant core coolant flow anomalies such as higher than expected core bypass flows, local pebble flow variation or even local flow blockage would be needed to account for such an increase in fuel temperature. The identified worst case scenarios are presented and discussed in detail. The conclusion of this work is that the stochastic nature of the pebble bed cannot lead to highly elevated fuel temperatures but rather local or core-wide coolant flow reductions are the likely cause.     

Optimization of a pebble bed configuration for quasi-direct numerical simulation

A high temperature reactor (HTR) is envisaged to be one of the renewed reactor designs to play a role in nuclear power generation including process heat applications. The HTR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in the event of an accident such as loss of coolant. However, the possible appearance of hot spots in the pebble bed cores of HTR may affect the integrity of the pebbles. This has drawn the attention of several scientists to understand this highly three-dimensional complex phenomenon. A good prediction of the flow and heat transport in such a pebble bed core is a challenge for CFD based on the available turbulence models and computational power. Such models need to be validated in order to gain trust in the simulation of these types of flow configurations. Direct numerical simulation (DNS), while imposing some restrictions in terms of flow parameters and numerical tools corresponding to the available computational resources, can serve as a reference for model development and validation. In the present article, a wide range of numerical simulations has been performed in order to optimize a pebble bed configuration for quasi-DNS which may serve as reference for validation.     

Burnup characteristics and fuel cycle economics of mixed uranium–thorium fuel in a simplified small pebble bed reactor

The limitation of natural uranium resources and the improvement of economic values of nuclear reactors are important issues to be solved in the future development of these reactors. In our previous study, we presented an innovative design for simplifying a pebble bed reactor, and the optimization of this design showed that burnup values could be increased and natural uranium uses could be reduced. The purposes of the current study were to design a simplified pebble bed reactor by removing the unloading device from the reactor system and to further optimize the burnup characteristics of this reactor with a peu à peu fuel-loading scheme by introducing thorium in the fuel configuration as a fertile material. Another goal was to optimize the fuel composition so that the system could achieve even better burnup characteristics and use scarce uranium resources more efficiently. Using a specially developed computer code, we analyzed and optimized the performance of a 110-MWt simplified pebble bed reactor using a peu à peu fuel-loading scheme. An optimized design using 30% of fertile thorium mixed with uranium fuel with 15% ²³⁵U enrichment and a 7% packing fraction calculated to achieve a high burnup of 140 GWD/T for more than 21 years' operation time that could save 13 to 33% of natural uranium use compared with the savings noted in our previous study. Neutronic, burnup and fuel economic analysis for this optimized design are discussed in this study.     

Comparison of coolant flow predictions with those measured on a full-scale mockup of a pebble bed reactor core

In a companion paper, a mathematical model for the analysis of coupled thermal-hydraulic problems in steady-state, axisymmetric pebble bed nuclear reactor cores was presented. In this paper, predictions by the computer code PEBBLE, which is based on the previously reported model, are compared with flow measurements from the full-scale mockup of the Oak Ridge National Laboratory Pebble Bed Reactor Experiment. The code PEBBLE is shown to adequately predict distributions and magnitudes of velocity and pressure for high Reynolds number flows in packed sphere beds. Limitations of the code and its ability to adequately calculate flow distributions in large pebble bed power reactors are discussed.     

高温气冷堆石墨构件断裂性能数值研究

为探究球床模块式高温气冷堆（HTR-PM）石墨堆内构件抗断裂破坏特性,提供石墨堆内构件设计和完整性评估的依据,利用经实验验证的基于内聚力模型的扩展有限元方法（XFEM）对球床模块式高温气冷堆侧反射层石墨砖的燕尾键 键槽结构进行了断裂性能的模拟分析,并对石墨断裂参数及几何尺寸等参数进行了敏感性分析。模拟结果显示：该石墨砖燕尾键 键槽结构的最大失效载荷P_max为50.7 kN,且随圆角半径而增大;P_max对石墨材料抗拉强度敏感,圆角越大越敏感,对材料断裂功、杨氏模量敏感度较小,但随着结构圆角变小变得相对更敏感,对泊松比几乎不敏感。分析结果与文献预测及实验结论具有较好的一致性。本文研究能对其他类型反应堆（如熔盐堆和快堆）的石墨构件断裂性能分析评价提供参考。     

On the few-group cross-section generation methodology for PBR analysis

The generation of accurate few-group nodal parameters, which strongly influences the accuracy of a reactor core criticality analysis, constitutes an important aspect of the work in developing advanced neutronics methodologies for pebble bed reactors (PBRs). As compared to other methodologies currently used for light water reactors (LWRs), the calculation at the core sub-region level for PBRs needs to account for the third spatial dimension because of the double heterogeneity of the configuration. This paper considers two modeling approaches for few-group cross-section generation: Monte Carlo (the MCNP code) and deterministic transport (the MICROX-2 code). Results from these codes are used to discuss the effect of the modeling assumptions (parameters) on the accuracy of the solution. Few-group cross-sections and spectral indices are calculated and compared for different packing fractions of the pebbles in the pebble lattice cell and different values of the moderator-to-fuel pebble ratio, temperature, and fuel burnup.     

Characteristics of a different fuel cycle in a PBMR-400 for burning reactor grade plutonium

An outstanding feature of the high temperature, gas-cooled, version of a Generation IV type reactor is its versatility in application. Apart from the capacity in high temperature gas-cooled reactors to generate electricity it could desalinate (multi-effect distillation [MED] by deploying excess heat or reverse osmosis by deploying excess electricity), produce hydrogen (by deploying excess electricity), whereas this article showcases the on-line fuelling characteristics of a pebble bed reactor concept for the incineration of reactor grade plutonium, whilst producing electricity.The VSOP-A system of codes is employed to demonstrate by calculation how the standard PBMR-400 commercial reactor design offers similar inherent safety characteristics with a Pu-Th/U advanced fuel cycle. This implies that no significant design changes are necessary to implement such a fuel cycle. Furthermore, the flexibility of the pebble fuel concept is deployed to house the fertile material in one type of pebble, whilst a second type will contain the fissile or driver material.     

Experimental measurement of effective thermal conductivity of packed lithium-titanate pebble bed

Lithium titanate is a promising solid breeder material for the fusion reactor blanket. Packed lithium titanate pebble bed is considered for the blanket. The thermal energy; that will be produced in the bed during breeding and the radiated heat from the reactor core absorbed must be removed. So, the experimental thermal property data are important for the blanket design. In past, a significant amount of works were conducted to determine the effective thermal conductivity of packed solid breeder pebble bed, in helium atmosphere, but no flow of gas was considered. With increase in gas flow rate, effective thermal conductivity of pebble bed increases. Particle size and void fraction also affect the thermal properties of the bed significantly. An experimental facility with external heat source was designed and installed. Experiments were carried out with lithium-titanate pebbles of different sizes at variable gas flow rates and at different bed wall temperature. It was observed that effective thermal conductivity of pebble bed is a function of particle Reynolds number and temperature. From the experimental data two correlations have been developed to estimate the effective thermal conductivity of packed lithium-titanate pebble bed for different particle Reynolds number and at different temperatures. The experimental details and results are discussed in this paper.