Indirect neutron radiography experiment on dummy nuclear fuel rods for pressurized water reactors at CMRR

Nuclear energy is a vital source of clean energy that will continue to play an essential role in global energy production for future generations.Nuclear fuel rods are core components of nuclear power plants,and their safe utilization is paramount.Due to its inherent high radioactivity,indirect neutron radiography(INR)is currently the only viable technology for irradiated nuclear fuel rods in the field of energy production.This study explores the experimental technique of indirect neutron computed tomography(INCT)for radioactive samples.This project includes the development of indium and dysprosium conversion screens of different thicknesses and conducts resolution tests to assess their performance.Moreover,pressurized water reactor(PWR)dummy nuclear fuel rods have been fabricated by self-developing substitute materials for cores and outsourcing of mechanical processing.Experimental research on the INR is performed using the developed dummy nuclear fuel rods.The sparse reconstruction technique is used to reconstruct the INR results of 120 pairs of dummy nuclear fuel rods at different angles,achieving a resolution of 0.8 mm for defect detection using INCT.

Characterization of crud deposited on fuel rods under HWC environment in Kuosheng Nuclear Power Plant

Under normal water chemistry conditions, the oxygen and hydrogen peroxide produced by water radiolysis in the coolant of boiling water reactors(BWRs) can lead to intergranular stress corrosion cracking in the constituent materials of plant components. This fact has led to the wide-scale adoption of hydrogen water chemistry(HWC) in the nuclear industry to counteract these effects.This study seeks to characterize the metallic composition and the surface properties of the constituent materials of plant components in order to determine their effects on the accumulation of chalk river unidentified deposits(crud) on fuel rods in the BWR Unit-1 of the Kuosheng Nuclear Power Plant in Taiwan. Inductively coupled plasma-atomic emission spectroscopy was used to calculate the concentrations of surface crud and gamma spectrometry was used to determine the radioactivity of the corrosion products, as well as their axial distribution across the surface of the fuel rods. X-ray diffraction analysis and scanning electron microscopy/energy-dispersive X-ray spectroscopy were used to identify the crystalline phase and morphology of the crud as irregular shapes and flakes. The amount of crud deposited during the fourth fuel cycle exceeded that of the third fuel cycle due to extended burn-up time. Our analytical results indicate that the implementation of HWC had no significant effect on the characteristics of subsequent crud.

Seismic and stress qualification of LMFR fuel rod and simple method for the determination of LBE added mass effect

In this study, two different designs of liquid metal fast reactor(LMFR) fuel rods wire-wrapped and nonwire-wrapped(bare) are compared with respect to different parameters as a means of considering the optimum fuel design. Nuclear seismic rules require that systems and components that are important for safety must be capable of bearing earthquake effects, and that their integrity and functionality should be guaranteed. Mode shapes, natural frequencies, stresses on cladding, and seismic aspects are considered for comparison using ANSYS. Modal analysis is compared in a vacuum and in lead–bismuth eutectic(LBE) using potential flow theory by considering the added mass effect. A simple and accurate approach is suggested for the determination of the LBE added mass effect and is verified by a manually calculated added mass, which further proved the usefulness of potential flow theory for the accurate estimation of the added mass effect. The verification of the hydrodynamic function(τ) over the entire frequency range further validated the finite element method(FEM) modal analysis results. Stresses obtained for fuel rods against different loading combinations revealed that they were within the allowable limits with maximum stress ratios of 0.25(bare) and 0.74(wire-wrapped). In order to verify the structural integrity of cladding tubes, stresses along the cladding length were determined during different transients and were also calculated manually for static pressure. The manual calculations could be roughly compared with the ANSYS results, and the two showed a close agreement. Contact analysis methodology was selected,and the most appropriate analysis options were suggested for establishing contact between the wire and cladding for the wire-wrapped design grid independence analysis,which proved the accuracy of the results, confirmed the selection of the appropriate procedure, and validated the use of the ANSYS mechanical APDL code for LMFR fuel rod analysis. The results provided detailed insight into the structural design of LMFR fuel rods by considering different structural configurations(i.e., bare and wire-wrapped) in the seismic loading;this not only provides a FEM procedure for LMFR fuel with complex configuration, but also guides the reference design of LMFR fuel rods.

Optimization of the fuel rod＇s arrangement cooled by turbulentnanofluids flow in pressurized water reactor (PWR)

In this paper, response surface methodology(RSM) based on central composite design(CCD) is applied to obtain an optimization design for the fuel rod's diameter and distance cooled by turbulent Al_2O_3–water nanofluid for a typical pressurized water reactor(PWR). Fuel rods and nanofluid flow between them are simulated 3D using computational fluid dynamics(CFD) by ANSYS-FLUNET package software. The RNG k–ε model is used to simulate turbulent nanofluid flow between the rods. The effect of different nanoparticles concentration is also investigated on the Nusselt number from heat transfer efficiency view point. Results reveal that when distance parameter(a) is in the minimum level and diameter parameter(r) is in the maximum possible level, cooling the rods will be better due to higher Nusselt number in this situation. Also, using the different nanoparticles on the cooling process confirms that Al_2O_3 averagely 17% and TiO_2 10% improve the Nusselt numbers.

Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor

The integrity and reliability of fuel rods under both normal and accidental operating conditions are of great importance for nuclear reactors.In this study,considering various irradiation behaviors,a fuel rod performance analysis code,named KMC-Fueltra,was developed to evaluate the thermal–mechanical performance of oxide fuel rods under both normal and transient conditions in the LMFR.The accuracy and reliability of the KMC-Fueltra were validated by analytical solutions,as well as the results obtained from codes and experiments.The results indicated that KMC-Fueltra can predict the performance of oxide fuel rods under both normal and transient conditions in the LMFR.

Revisiting Stainless Steel as PWR Fuel Rod Cladding after Fukushima Daiichi Accident

In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as cladding material mainly due to its lower neutron absorption cross section. Now, stainless steel cladding appears as a possible solution for safety problems related to hydrogen production and explosion as occurred in Fukushima Daiichi accident. The aim of this paper is to discuss the steady-state irradiation performance using stainless steel as cladding. The results show that stainless steel rods display higher fuel temperatures and wider pellet-cladding gaps than Zircaloy rods and no gap closure. The thermal performance of the two rods is very similar and the neutron absorption penalty due to stainless steel use could be compensating by combining small increase in U-235 enrichment and pitch size changes.

Multiphysics simulation of VVER-1200 fuel performance during normal operating conditions

Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions.Fuel performance is a complicated phenomenon that involves thermal,mechanical,and irradiation mechanisms and requires special multiphysics modules.In this study,a fuel performance model was developed using the COMSOL Multiphysics platform.The modeling was performed for a 2D axis-symmetric geometry of a UO2fuel pellet in the E110 clad for VVER-1200 fuel.The modeling considers all relevant phenomena,including heat generation and conduction,gap heat transfer,elastic strain,mechanical contact,thermal expansion,grain growth,densification,fission gas generation and release,fission product swelling,gap/plenum pressure,and cladding thermal and irradiation creep.The model was validated using a code-to-code evaluation of the fuel pellet centerline and surface temperatures in the case of constant power,in addition to validation of fission gas release(FGR)predictions.This prediction proved that the model could perform according to previously published VVER nuclear fuel performance parameters.A sensitivity study was also conducted to assess the effects of uncertainty on some of the model parameters.The model was then used to predict the VVER-1200 fuel performance parameters as a function of burnup,including the temperature profiles,gap width,fission gas release,and plenum pressure.A compilation of related material and thermomechanical models was conducted and included in the modeling to allow the user to investigate different material/performance models.Although the model was developed for normal operating conditions,it can be modified to include off-normal operating conditions.

Investigation of the interaction of material of fuel cladding for WWER-1000 reactor with steam at a temperature of accident overheatings

The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grinding) and after additional chemical etching. The changes in the microstructure of tubes due to the interaction with steam were investigated. A comparison was made between the oxidation rate of this material (weight gain) and the data on the oxidation of other alloys for nuclear power plants. The oxidation rate of Zr1Nb(0.1% O) is close to the oxidation rate of other zirconium alloys. It is shown that after chemical treatment of the surface of the samples there is a more even growth of oxide films, and they have a smaller thickness for the same time of exposure than after mechanical grinding. Surface treatment before oxidation also affects the change of microstructure of samples when heated to high temperatures.

燃料棒性能分析程序KMC-fuel的力学模块开发

KMC-fuel是针对液态金属冷却快堆棒状燃料元件开发的性能分析程序,具有综合评估燃料棒在服役期间的热、力、辐照性能的功能。本文围绕该程序的力学分析模块展开详细论述。KMC-fuel的力学分析模块采用基于应力的数值算法求解了结构的力学方程组。采用横观各向同性假设模拟了芯块开裂现象,引入重定位模型模拟开裂碎片对间隙的吸收,在滑移接触条件下得到了分析芯块-包壳机械相互作用的显示迭代算法,分析中考虑了燃料棒热膨胀、密实化、肿胀、蠕变等材料行为。经过与理论及实验数据对比,结果显示KMC-fuel的力学分析模块能够准确预测燃料棒的应力应变场分布,表明该力学分析模块可用于燃料棒性能分析。

Irradiation Testing of Coated Particle Fuel at HANARO

TRISO （Tri-structural iso-tropic）-coated particle fuel is being developed to support the development of a VHTR （very high temperature reactor） in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in the VHTR in HANARO （high-flux advanced neutron application reactor） at KAERI （Korea Atomic Energy Research Institute）. This experiment is currently undergoing under an atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak bum-up of about 4% and a peak fast neutron fluence of about 1.7 × 1021 n/cm2, PIE （post irradiation examination） will be carried out at KAERI＇s irradiated material examination facility. This paper describes the characteristics of coated particle fuels, and the design of the test rod and irradiation device for the coated particle fuels, and discusses the technical results of irradiation testing at HANARO.

乏燃料棒M5锆合金包壳的透射电镜分析

压水堆燃料元件的锆合金包壳,在服役期间会经受高中子注量辐照,其微观组织将发生很大变化,从而影响其宏观性能,因此锆合金包壳的中子辐照行为研究一直是核领域的研究重点。但由于材料经中子辐照后具有较强的放射性,相关的实验必须在热室内进行,因此针对辐照后燃料包壳微观组织的研究也一直是工作的难点。本文在中国原子能科学研究院热室设施上,通过透射电镜分析手段,研究了M5锆合金包壳材料中子辐照后的微观组织。样品来源于国内商业压水堆AFA3G型乏燃料棒,其燃耗分别为14 GW·d/tU和41 GW·d/tU。从燃料棒上截取长度约10 mm的包壳样品,在热室内完成去芯块与化学清洗,获得空包壳样品,然后通过机械制样方法,制备出?3 mm薄片状包壳基体样品,最后采用电解双喷减薄方法,制备出包壳透射电镜观察分析样品。另外,为对比锆包壳辐照后的组织变化,采用同样方法制备了相同材料的冷态观察分析样品。冷态样品与辐照样品的观察分析结果表明:冷态Zr合金包壳基体组织内部存在原生的第二相粒子,基体内部整体较为干净,纳米析出相稀少,未观察到明显的位错结构;辐照后,基体内原生的第二相粒子尺寸和分布与冷态样品差异不明显,但出现了明显的纳米析出相和高密度位错组织;随着燃耗的增加,纳米析出相尺寸有增加的现象;低燃耗与高燃耗样品位错组织具有相似性,表明在14 GW·d/tU燃耗下,锆合金包壳内由辐照产生的位错组织已基本趋于饱和状态;电子选取衍射结果表明,辐照后,基体内原生的第二相粒子虽存在一些非晶组织,但仍以bcc晶体结构为主,表明在41 GW·d/tU燃耗下,第二相粒子保持了一定的辐照稳定性;另外,第二相的EDS结果表明,随着燃耗的增加,Nb元素的含量有贫化趋势;分析认为,Zr合金经中子辐照,第二相粒子中的Nb原子扩展至Zr基体内,将促进Nb元素以纳米富Nb相形式在Zr基体中析出。

燃料组件破损棒自动超声检查系统关键技术研究

燃料组件长期服役过程中,会因为振动、材料老化、强辐照等原因发生破损,这为核电机组安全运行带来隐患,而燃料组件破损自动超声检测系统尚未达到完全国产化。为此,该文从检测理论、探头和采集系统方面讨论了自动超声检测系统中的关键技术,同时对声束传播过程和响应进行了仿真和试验验证。试验结果表明,实际测量结果与理论模型吻合,自主研制的自动超声检查系统可有效分辨燃料组件破损棒。

基于POD方法的EPR燃料棒流致振动特性分析

EPR(European pressurized reactor)燃料棒相比M310燃料棒的棒长更长,导致其频率降低、幅值增大,在冷却剂的作用下,会加剧格架与棒束之间微动磨损(grid-to-rod fretting,GTRF),进而造成放射性物质的泄漏.将EPR燃料棒简化为3D梁模型,将刚凸和弹簧对燃料棒的约束等效为弹性约束,将带格架的燃料棒简化为多跨连续简支梁模型,然后基于ANSYS-APDL建立了EPR燃料棒的有限元模型.阐述了湿模态分析和湍流激振响应分析的基本原理,整理了12个格架失效工况,系统地研究了格架失效对湿模态和湍流激振响应的影响.针对EPR燃料棒流致振动问题,提出了采用本征正交分解(proper orthogonal decomposition,POD)原理分析EPR燃料棒流致振动特性的方法,通过对快照矩阵进行POD分解生成投影子空间,将湍流激振响应投影到子空间进行模型降阶,最后在物理空间快速地重构湍流激振响应.结果表明:格架失效会导致频率降低,且湍流激振响应的幅值会在格架失效处增大;当格架失效使得EPR燃料棒模型成为悬臂梁结构时,湍流激振响应最大;前2阶POD降阶模型基本能快速重构燃料棒的湍流激振响应,且误差非常小.文章的研究将有助于核反应堆工程的优化和设计.

压水堆棒状燃料氧化与吸氢模型开发及验证

核反应堆长期运行过程中,锆合金包壳与冷却水接触会发生氧化反应、吸氢脆化行为,使包壳的导热性能和力学性能恶化,威胁燃料元件的安全特性。因此,开展棒状燃料氧化与吸氢行为研究具有重要意义。本文实现了MOOSE-BEEs燃料性能分析程序中压水堆棒状燃料堆内腐蚀计算模型的开发,主要包括氧化腐蚀模型和吸氢腐蚀模型。基于相关的实验数据和BISON程序计算结果,对氧化腐蚀、氢扩散、氢析出等单个模型及耦合模型开展了验证。验证结果表明BEEs的模拟结果与实验数据和BISON程序吻合较好,说明BEEs能够准确模拟燃料棒的氧化与吸氢行为。

热管堆固态堆芯燃料辐照-热-力耦合性能分析

热管冷却反应堆(简称:热管堆)具有高可靠性和固有安全性、体积小、模块化和全固态堆芯等特点。固态堆芯燃料服役过程在高温、强辐照、固态约束多因素作用下堆芯的传热和力学性能受到严重影响,基体接触导致应力与间隙换热都随燃耗加深而发生较大非线性改变,且两者相互影响,因此基体在服役过程中的多物理场耦合的辐照-热-力行为复杂。本文基于有限元多物理场分析软件针对固态堆芯燃料开展辐照-热-力耦合分析,考虑UO2芯块与316不锈钢基体的辐照变形效应以及蠕变效应,并在固态堆芯间隙中引入间隙传热模型,探究固态堆芯寿期内间隙变化特点以及传热和力学耦合作用特性。结果显示:基体与燃料包壳的完全接触会导致芯块温度上升以及基体与包壳蠕变现象加强,燃料棒周围平均热管数量较少会导致附近区域较高的温度和应力分布,且寿期中该区域包壳因燃料棒内压和基体-包壳接触压力具有蠕变失效风险。分析结果表明间隙接触会对热管堆固态堆芯的传热和力学性能造成影响,甚至提高包壳的失效风险。

基于深度学习的X射线燃料棒端塞缺陷自动检测方法研究

为了提高深度学习在X射线燃料棒端塞缺陷检测中的准确性,实现更高精度的无损检测,本文基于YOLOX的目标检测模型,针对该场景下目标缺陷尺寸极小的特点,对网络结构和损失函数进行了相应的改进,并在工业数据集上进行了验证。结果表明,该算法方案在保持较高识别速度的同时,识别精度获得了明显的提升,达到生产检测要求。该研究方法为今后燃料棒端塞焊缝X射线数字检测图像的高精度自动分析评价打下了坚实的基础。

氦氙气冷小堆燃料棒辐射散热特性分析

反应堆堆芯热工安全特性是反应堆系统安全特性最重要的组成部分。氦氙气冷小堆中燃料棒间温差大,辐射散热不可忽略。因此针对氦氙气冷小堆堆芯,不止需要考虑对流换热和热传导两种换热形式,也需要考虑辐射散热对堆芯换热的影响。本文通过建立氦氙气冷小堆堆芯1∶1精细化模型,开展反应堆堆芯内部三维辐射散热特性分析,同时获得辐射角系数。将本文模型计算结果与蒙特卡罗方法结果进行对比验证,两者相对误差小于1%,证明该辐射角系数计算方法的准确性。开展了燃料棒径距比、燃料棒表面温度、燃料棒长度等几何参数敏感性分析,关注其对燃料棒辐射散热特性的影响,并开发出一套具有普适性的堆芯内燃料棒辐射角系数经验关系式。本文通过研究燃料棒间辐射散热特性了解反应堆内辐射散热规律,为后续反应堆堆芯热工安全特性研究提供了技术支撑。

用于燃料棒氦质谱检漏的多级差分真空系统设计方案研究

用于燃料棒氦检漏的装置是燃料棒生产过程中的核心部件之一,关系到燃料棒的生产质量和核电站运行稳定。现有的检漏装置采用卧式检漏装置,将燃料棒成批放入其中进行检测,检漏效率有待提高。文章开发一种多级差分真空系统,用于燃料棒的在线通过式检漏,根据系统特点,提出了系统的设计思路和流程,通过迭代计算的方法,进行极限差分比、抽速和稳态压力计算,并对系统的参数进行分析。研究表明,该多级差分真空系统的研制是可行的,检漏室压力能维持在较低的水平,检漏响应时间较快,检漏效率优于现有设备,且系统冗余度较高,有较大的降低成本的空间。该系统能实现单根燃料棒的氦质谱检漏,为燃料棒的质量管控和溯源提供了基础。

聚丙烯酸酯基耐刮擦涂层的制备与性能研究

核燃料棒作为核电堆芯的第二道放射性保护屏障,在安装过程中极易被固定格架划伤。而现有的核燃料棒防护涂层存在防护效果不佳、清洗不达标等难题。基于乳液聚合技术制备硬度和柔韧性均衡的丙烯酸乳胶,并制备水性耐刮擦涂料。结果显示:涂料粒径分布较窄;涂层在200℃以下具有热稳定性,划痕形貌为韧性断裂,水接触角降低至48.66°;并且烘烤温度90℃烘烤时间30 min所得涂层耐刮擦性能和水洗性能达到最优。该涂层有望减少核燃料棒表面刮伤,助力我国核电事业安全发展。