Fluorination of UF4 in a mini-tapered fluidized bed and mathematical modeling

A mini-tapered fluidized bed reactor can be used for fluorination reaction of UF₄ to produce uranium hexafluoride. By adopting the mini-tapered bed the problems associated with fluidization in a cylindrical bed such as entrainment of particles and the limitation of operating velocity can be overcome, consequently the performance of the reactor can be enhanced. Simulation of the reactor was performed employing two-phase models, bubble phase with piston flow and emulsion phase with piston (D.P-P model) or perfectly mixed flow (D.P-M model). The voidage of the emulsion and bubble phases were estimated from the distribution two-phase structure hydrodynamic model. The model predictions have been compared with the results from a pilot-scale experiment. The D.P-P model gives good agreement between computed and empirical results. The effects of various parameters on the reactor performance are discussed using the model.     

振动流化床微波脱硝研究——减压振动流化床流体力学行为的研究

根据六水硝酸铀酰 (UNH)的脱硝动力学特性 ,将微波加热减压振动流化床用于UNH的脱硝过程 ,是一个很有价值的脱硝工艺。为此 ,首先进行了减压振动流化床流化特性的研究。以床层空隙率为主要评价参数 ,讨论了减压振动流化床的假流化机理和特性 ,着重研究了真空度、床高和振动频率等参数对流化过程的影响。建立了最小流化振动强度的经验关联式。     

Dynamics modeling and stability analysis of a fluidized bed nuclear reactor

A theoretical model describing the coupling of neutronics, thermohydraulics and fluidization in a fluidized bed nuclear reactor is presented. The stability of the system is investigated by linearizing and perturbing the system around its equilibrium points and identifying the root loci of the sytem. It is found that within the operational range, the eigenvalues are located in the negative part of the phase plane, implying linear stability. Simulations of transient conditions are performed, viz. a hypothetical startup transient and a quasistatic transient related to noise resulting from stochastic movements of the fuel particles. These simulations show that although the total power of the reactor may reach high values, the fuel temperature is well below safety limits at all times.     

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.     

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。     

球床包层聚变—裂变混合堆热工安全分析

球床包层混合堆与板状元件包层混合堆相比较，前者在核燃料生产和安全方面可能具有更多的优越性。本应用ＴＨＥＲＭＩＸ程序和辅助程序对我国开发的托卡马克堆芯氮气冷却球床包层聚变－裂变合堆的包层进行了热工计算。计算中考虑了不同的燃料球材料及稳态，卸压和断流事故工况。计算结果表明，只要选用合适的燃料球材料和设置适当的控制保护系统，具有快速卸料罐的托卡马克堆芯氦气包层聚变－裂变混合堆的概念设计在安全上的可行的。     

Thermofluid-neutronic stability of the rotating, fluidized bed, space-power reactor

A rotating fluidized bed nuclear reactor has the potential of being a vary attractive option for ultra-high power space systems, especially for propulsion. Research has already examined fuel bed expansion due to variations in state variables, propellant flow rate, and rotational speed, and has also considered problems related to thermal stress. This paper describes the results of a coupled thermofluid-neutronic analysis where perturbations in fuel bed height caused by maneuvering changes in operating conditions alter power levels due to varying absorption of neutrons which would otherwise leak from the system, mainly through the nozzle. This first analysis was not a detailed stability analysis. Rather, it utilized simplified neutronic methods, and was intended to provide an order-of-magnitude assessment of the stability of the reactor with the intention to determine whether or not stability might be a “concept killer”.Stability was compared with a fixed-fuel-bed reactor of identical geometry for three different cases comprising a set of small, medium and large sizes/powers from 250 MW to 5 GW. It was found that power fluctuations in the fluidized bed reactor were larger by 100 db or more than expected in a packed bed reactor of the same geometry, but never resulted in power excursions. Margins to unit gain in some cases, however, were sufficiently small that the approximations in this quasi-2-dimensional model may not be sufficiently accurate to preclude significant excursions.     

固态燃料熔盐堆自适应功率控制器设计及分析

钍基熔盐堆核能系统(Thorium-based Molten Salt Reactor,TMSR)是中国科学院首批启动实施的战略性先导科技专项,旨在研发第四代反应堆核能系统。固态燃料钍基熔盐实验堆(The Solid Fuel Thorium-based Molten Salt Experimental Reactor,TMSR-SF1)是一个10 MW热功率的氟盐冷却球床堆,目前已经完成方案设计和初步工程设计。功率控制系统是反应堆一个关键控制系统,实现反应堆正常启动、功率运行和正常停堆功能,对保证反应堆安全和稳定运行起着极其重要的作用。根据TMSR-SF1运行控制要求,结合自适应控制理论,基于Lyapunov稳定性理论设计了一种TMSR-SF1模型参考自适应功率控制器。基于TMSR仿真平台,使用MATLAB/Simulink建立了自适应功率控制系统模型,并开展了控制器特性分析。结果表明,自适应功率控制器具备良好的负荷跟随能力,抗干扰能力强、稳定性好、可靠性高,能够满足TMSR-SF1功率控制的要求,确保堆芯的输出功率与功率设定值相匹配。     

先进的规则床模块式高温气冷堆概念

规则床模块堆是燃料球呈规则堆积的一种先进的模块式高温气冷堆设计。燃料球在平面上成正方形排列,四个球的中心是次一层球的位置,形成正四棱锥堆积。当燃料球落入被做成一定几何形状的堆芯空腔时,就自动形成规则堆积。燃料球可以从反应堆顶部装入和卸出,能够在较短的停堆时间内完成换料操作。规则床堆芯是一种密实体,具有很强的结构适应性和稳定性。在模块化设计中,保持非能动冷却和限制最高燃料温度的条件下,它能够提高输出功率和降低堆芯压降,同时还兼有球形燃料堆和柱状燃料堆的主要优点。本文介绍规则堆积床特性和预测规则床模块堆的设计性能。     

高温气冷堆不确定性分析的新进展

球床高温气冷堆由于采用流动球床堆芯和燃料多次通过的运行方式,不能直接套用轻水堆中一般采用的“系统分解,逐级传递”的分析思路,其不确定性的传播和分析具有特殊性。清华大学核能与新能源技术研究院基于高温气冷堆的设计分析经验,开展了高温堆的不确定性研究,并取得了一些进展。目前高温气冷堆已建立起完整的不确定性分析计算框架。在此框架内,基于VSOP程序,开发能反映球床高温气冷堆实际运行特点的不确定性分析程序VSOP-UAM,实现了核数据不确定性隐式效应和显式效应的完整分析。然后使用SCALE/TSUNAMI-3D和VSOP-UAM程序,建立燃料球、堆芯单元、初装堆芯和平衡堆芯的分析模型,量化了核数据的不确定性对各种模型关键参数的影响。此外,还量化了球流混流效应、燃料富集度、燃料孔隙率这些球床堆芯参数的不确定性对堆芯有效增殖因数k_eff和功率分布的影响。从计算结果可看出,高温气冷堆的不确定性分析显示出了有别于传统轻水堆的结果。     

冲击载荷下核动力船只控制棒的动态响应研究

对于核动力船只,当受到冲击载荷作用时,尤其是沿竖直方向的冲击载荷作用时,反应堆的控制棒会产生一定的位移,改变反应堆的反应性,影响反应堆的正常运行,进而会影响到船只的正常航行。在此,通过建立合理的计算模型,研究了核动力船只仅受竖直方向冲击载荷作用时控制棒的位移与加速度的响应情况,并给出了反应堆基频与载荷作用时间对控制棒位移响应的影响情况。     

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.     

Dryout heat flux experiments with deep heterogeneous particle bed

A test facility has been constructed at Technical Research Centre of Finland (VTT) to simulate as accurately as possible the ex-vessel core particle bed in the conditions of Olkiluoto nuclear power plant. The STYX particle bed reproduces the anticipated depth of the bed and the size range of particles having irregular shape. The bed is immersed in water, creating top flooding conditions, and internally heated by an array of electrical resistance heating elements. Dryout tests have been successfully conducted at 0.1–0.7 MPa pressure for both uniformly mixed and stratified bed geometries. In all tests, including the stratified ones, the dry zone first formed near the bottom of the bed. The measured dryout heat fluxes increased with increasing pressure, from 232 kW/m² at near atmospheric pressure to 451 kW/m² at 0.7 MPa pressure. The data show some scatter even for the uniform bed. The tests with the stratified bed indicate a clear reduction of critical power due to the presence of a layer of small particles on top of the uniform bed. Comparison of data with various critical power (dryout heat flux) correlations for porous media shows that the most important parameter in the models is the effective particle diameter. Adiabatic debris bed flow resistance measurements were conducted to determine the most representative particle diameter. This diameter is close, but not equal, to the particle number-weighted average diameter of the bed material. With it, uniform bed data can be calculated to within an accuracy of 3–28% using Lipinski's 0-D model. In the stratified bed experiments, it appears that the top layer was partially fluidized, hence the measured critical power was significantly higher than calculated. Future experiments are being planned with denser top layer material to eliminate non-prototypic fluidization.     

The design features of the HTR-10

The 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is a modular pebble bed type reactor. This paper briefly introduces the main design features and safety concept of the HTR-10. The design features of the pebble bed reactor core, the pressure boundary of the primary circuit, the decay heat removal system and the two independent reactor shutdown systems and the barrier of confinement are described in this paper.     

基于CITATION-ORIGEN2球床堆平衡态计算程序的实现

基于确定论中子扩散软件CITATION和点燃耗软件ORIGEN2,编写了球床堆分析程序COBBLE,以实现指定燃料球加载策略下的球床堆平衡态燃耗计算。COBBLE程序采用谱区能谱修正方法,通过迭代求解得到球床堆堆芯平衡态参数。本文选取简化的球床模块高温气冷堆(PBMR)堆芯进行建模,计算其功率分布及燃耗分布,并使用基于蒙特卡罗方法的球床堆燃耗计算程序PBRE进行了验证与分析。结果表明,COBBLE程序适用于球床堆的平衡态燃耗计算。     

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.     

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.     

Simulations of flow behavior of fuel particles in a conceptual helium-cooled spout fluidized bed nuclear reactor

Hydrodynamics of helium and fuel particles are simulated in a conceptual helium-cooled spout fluidized bed nuclear reactor. The conceptual reactor consists of an axis-symmetric column with a sharp cone inside which the fuel particles are fluidized by helium. An isothermal gas-solid two-fluid flow model is presented. The kinetic-frictional constitutive model for dense assemblies of solids is incorporated. The kinetic stress is modeled using the kinetic theory of granular flow, while the friction stress is from the normal frictional stress model proposed by (Johnson, P.C., Nott, P., Jackson, R., 1990. Frictional-collisional equations of motion for particulate flows and their application to chutes. Journal of Fluid Mechanics 210, 501-535). Detailed spatial/temporal concentration and velocity profiles have been obtained in a conceptual spout fluidized bed nuclear reactor. The influence of inlet spouting jet velocity and conical angles on flow behavior of fluid and fuel particles is analyzed. The numerical simulations show that the unique mixing ability of the spout fluidized bed nuclear reactor gives rise, as expected, to uniform particle distributions. This uniformity enhances the heat transfer and therefore the power produced by the reactor.     

Control rod worth calculation for VVER-1000 nuclear reactor using WIMS and CITATION codes

This paper summarizes efforts related to developing a technically justifiable approach for investigating the control rod worth of VVER-1000 reactor. For this assessment, computer simulation of nuclear reactor was needed. In this study nuclear reactor behavior was modeled by WIMS code, which solves transport equation for fuel assemblies' modeling at first step, and CITATION code that solves diffusion equation for core modeling. From these two codes, neutronic calculation of reactor was performed and control rod worth was calculated. Results of this study are comparable with the plant's FSAR.On comparing results of this study and reference some unacceptable errors were discerned. To find out the cause of these errors, some efforts had been performed and finally was discerned that the method of cell calculation, i.e., DSN method, was the important cause of errors. Therefore, some analysis had been performed by WIMS in PIJ + PERSEUS method and was shown that the results were improved.     

10MWt固态燃料熔盐堆控制棒失控抽出事故分析

钍基熔盐堆核能系统项目是中科院先导科技专项之一,其战略性目标是研发第四代熔盐冷却裂变反应堆核能系统。基于10 MWt固态燃料熔盐堆的系统设计,开发了适用于球床式反应堆系统的安全分析软件,并以高温气冷堆为对象对程序计算结果的准确性进行了验证。基于该软件程序,对固态燃料球床堆(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF)控制棒失控抽出事故进行了分析计算,研究了不同停堆限值及各停堆信号对事故的影响。计算结果表明,超功率停堆限值越高,出口温度限值越大,信号延迟时间越长,反应堆停堆越晚,堆芯功率和燃料最高温度越高。在TMSR-SF控制棒失控抽出事故下,燃料最高温度不超过860°C,远低于1 600°C的熔化温度限值。