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.

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.

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.

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.

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。

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。

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。

核燃料棒电阻焊设备研制关键技术研究

为充分掌握影响核燃料棒电阻焊成形质量的关键因素,利用自研电阻焊专用设备研究了焊接电流、顶锻压力以及焊接时间等因素对电阻焊成形质量的影响规律,并与进口设备比较。结果表明,自研设备在顶锻结构和电极夹头等核心部件上采用新设计,使得焊缝成形与进口设备有显著差异,典型顶锻结合长度超过2 mm,该长顶锻变形可以采用超声检测直观表征焊缝缺陷,焊缝外观和组织结构检测均表明自研设备具有良好的焊接成形能力。

绕丝燃料棒自动外观缺陷检测系统设计

绕丝燃料棒是核反应堆中燃料组件的核心部件,燃料棒外径绕丝是燃料棒制造的关键工序,其外观质量是燃料组件十分关键的质量特性。本文以视频图像为切入点,机器视觉为技术支撑,借助深度学习建立绕丝燃料棒外观缺陷智能判别及自动定位方法。针对燃料棒表面缺陷特征,搭建高精度的图像采集系统,获取便于识别的燃料棒外观图像信息,研究检测算法能准确识别焊缝氧化色、划伤、异物、焊缝成形不良等缺陷,实现绕丝燃料棒外观缺陷的自动化检测。

复杂燃料组件的CFD优化设计研究

对螺旋十字型燃料棒的复杂燃料组件开展全组件CFD模拟分析,保留燃料组件内复杂结构和微间隙特征,基于全域高精度网格划分结果,使用Realizable k-ε湍流模型开展数值模拟计算。基于CFD数值仿真数据,以组件结构参数为变量建立多参数优化方法,采用均匀设计方案、二阶响应面模型与遗传算法开展燃料组件系统的水力优化研究。优化结果指出分流量圆盘的中心部分直径D较小,导向头的长度H较大时优化效果更好;形成的6通流孔的优化方案组件压降减小8.61%,4通流孔的优化方案组件压降减小22.23%,并且4个通流孔的优化方案的速度均匀性更好。

压水堆核燃料棒环焊缝及堵孔点的数字射线检测

压水堆核燃料棒是核电站使用最多的一种燃料元件,需求数量逐年增加,为提高压水堆核燃料棒环焊缝及堵孔点的检测效率和质量,对其进行了数字射线检测试验。试验结果声明,其检测图片灵敏度最高可达0.25 mm,图像分辨率达D11,归一化信噪比为122,燃料棒缺陷图片与胶片试验、金相试验结果一致;该检测方法能满足燃料棒的质量评价要求。

AFA3G燃料组件拉棒机爪头收爪工艺分析与改进

分析AFA3G燃料组件拉棒机拉杆爪头收爪工艺的原理及存在的问题,提出有针对性的改进措施,包括爪头和卡管的材质、尺寸、表面处理等方面的优化。通过定性和定量对比分析,改进后的收爪工艺的可靠性明显提升,有效避免了燃料组件产品的质量隐患。

压水堆完整和破损燃料棒燃料包壳化学相互作用层拉曼特征分析

为研究压水堆完整和破损燃料棒燃料包壳化学相互作用(FCCI)层物相结构组成及影响因素,通过拉曼光谱对燃耗为45 GW·d/tU和41 GW·d/tU的完整和破损燃料棒FCCI层进行了研究分析。结果表明:完整燃料棒形成了周向厚度为14~19μm的FCCI层,主要由两个不同相结构区域组成,分别为靠近包壳界面的单斜和四方相氧化锆混合区域及靠近燃料芯块的四方相区域,在包壳界面附近约7μm范围内,观察到明显的705 cm^(-1)特征峰,该峰的形成源于界面压应力和辐照缺陷的共同作用;破损燃料棒形成了周向厚度为37~61μm的FCCI层,主要由两个不同形貌和相结构区域组成,即靠近包壳界面附近具有多孔、裂纹特征的单斜相氧化锆区域以及靠近燃料芯块的非晶结构区域。对FCCI层相结构的分布及转变影响因素进行了分析讨论,完整燃料棒FCCI层中四方相氧化锆的稳定与界面压应力、中子辐照缺陷和裂变产物作用有关,破损燃料棒FCCI层中单斜相氧化锆的存在则主要来源于应力的弛豫和氧的正常化学计量比。