燃料包壳破损条件下裂变气体释放模拟实验研究

反应堆燃料包壳破损发生时需要判断其破损程度,为系统及时作出响应提供参考依据。目前使用逃脱率系数表征压水堆燃料包壳的破损程度,但对于裂变气体释放机理缺乏研究。本文采用实验方法研究燃料包壳破损时,非稳态过程中冷却剂压力和温度对裂变气体释放的影响。实验装置基于几何相似性、流动相似性以及闪蒸相变相似性设计,考察了在子通道内冷却剂压力与温度对裂变气体释放的影响,以及闪蒸对非稳态过程中裂变气体逃脱率的影响。实验结果表明:在选取的0.5 mm破口尺寸下,非稳态过程对于气体释放速率没有明显影响,实验中长期逃逸率保持稳定,释放过程符合一级动力学方程。同样冷却剂压力下,冷却剂温度从90℃增长到110℃时,长期逃逸率增长。同样冷却剂温度下,回路压力从0.3 MPa增长到0.5 MPa,长期逃逸率则下降。长期逃逸率与过冷度负相关,表明燃料包壳破口处的液膜对于裂变气体释放有影响。     

压水堆燃料元件包壳破损在线监测系统研制

针对压水堆核电厂运行工况下燃料元件包壳发生破损的情况,通过以机理性定量分析方法为基础的诊断物理模型和在线监测系统设计,给出完整的包壳破损在线监测解决方案。同时,通过理论模拟计算、原理样机带源实验以及电厂实测运行数据验证,多方面验证了系统设计的正确性。该套系统能够改进中国改进型百万千瓦级压水堆(CPR1000)机组现有燃料破损监测手段的不足,提高压水核电机组运行的安全性能。     

压水堆核电厂燃料包壳破损诊断方法研究和改进

分析了国内外压水堆核电厂燃料包壳破损诊断方法以及存在的问题，从燃料棒破损数量、破损尺寸和燃耗3个方面对压水堆核电厂燃料包壳破损的诊断方法进行了改进，并对可能影响诊断结果的因素进行了探讨。应用我国在役核电厂实际的运行数据对诊断方法进行了验证，结果表明，改进后的燃料包壳破损诊断方法可准确地诊断燃料包壳破损情况，且有更广泛的适用性。      

智能化核电厂燃料包壳破损在线监测装置研制

研制一套智能化核电厂燃料包壳破损在线监测装置,采用高纯锗反康谱顿散射探测系统在线测量一回路冷却水特征放射性核素的活度,采用多核素组耦合的分析方法实现燃料包壳破损的在线诊断。通过检定校准试验,实测⁵⁷Co、¹³⁷Cs和⁶⁰Co的相对标准偏差的绝对值小于3%;20 mL样品的可探测活度最小可达到6.5 Bq。     

基于人工神经网络的压水堆燃料破损状态监测

为提升对核反应堆燃料棒包壳破损的预测能力,建立两个串联的人工神经网络分别判断燃料棒包壳是否破损以及破损程度。通过改变沾污铀质量、增加数据扰动、改变运行功率和使用更少的特征核素进行训练,对用于判断是否破损的神经网络模型和判断破损等级的神经网络进行了性能测试和分析。在沾污铀质量小于0.5 g、数据扰动在30%以内、单棒功率在77 kW到120 kW之间的条件下,第1个人工神经网络能较好地判断出是否破损。第2个神经网络,对于考虑的5种破损程度,判断的精确性较高。与传统的碘同位素比值法相比,神经网络方法响应更快,精度更高。结果表明,人工神经网络可用于预测反应堆燃料包壳是否发生破损以及破损程度。     

智能化核电厂燃料包壳破损在线监测装置研制

An intelligent on-line monitoring device for fuel cladding defect has been developed. HPGe with anti-Compton scattering detection system is used to measure the activity of characteristic radionuclides in primary cooling water, and the multi-nuclide group coupled analysis method is used to diagnose the defect of the fuel cladding. The verification and calibration test shows that the absolute relative standard deviations of the measured typical nuclides 57Co, 137 Cs and 60Co are less than 3%, and the minimum detectable activity can reach 6.5 Bq.     

破损燃料组件中裂变产物向冷却剂释放量计算

分析包壳破损情况下裂变产物从燃料芯块向冷却剂的释放机理,建立裂变产物从燃料芯块向冷却剂的释放量的计算模型;采用CPR1000机型的设计参数对燃料包壳破损率、包壳破损尺寸和燃耗开展敏感性分析,计算等效逃脱率系数并与AP1000设计控制文件中给出的逃脱率系数进行比较。结果表明,包壳破损尺寸对裂变产物释放的影响较大,燃耗和包壳破损率对裂变产物释放影响较小。在包壳破口尺寸为34μm时,采用建立的计算模型计算所得部分核素的等效逃脱率系数与AP1000设计控制文件中给出的逃脱率系数极为接近。     

烧结二氧化铀燃料裂变气体肿胀的数值模拟研究

建立低温条件下烧结二氧化铀燃料(简称UO2燃料)中裂变气体的肿胀计算模型,采用有限差分方法编写计算程序,定量计算不同燃耗和温度条件下UO2燃料中固溶态的裂变气体份额、裂变气体气泡的密度与平均半径以及它们对燃料肿胀的贡献.计算表明,该模型能用于预测低温条件下UO2燃料中裂变气体所导致的肿胀随燃耗的变化规律.     

基于弥散燃料颗粒开裂的裂变气体释放模型

根据弥散燃料颗粒开裂后裂变气体的3种释放途径,分别建立了裂纹连通释放模型、气泡连通释放模型以及原子扩散释放模型,综合得到了基于弥散燃料颗粒开裂的裂变气体释放模型,并采用该模型对裂变气体释放量进行了计算。结果表明:裂变气体释放量主要由裂纹连通释放途径贡献;燃耗深度越高,裂变气体释放量的增加速率会越大;随着退火温度的增加,裂变气体释放量迅速增加,而退火时间越长,裂变气体释放量的增加速率越低。通过裂变气体释放量模型计算得到的裂纹宽度与实验观察到的裂纹宽度符合较好,对比结果验证了基于弥散燃料颗粒开裂的裂变气体释放模型的合理性。      

燃料元件破损监测裂变气体法的理论计算

及时发现燃料元件包壳破损，与核安全有着密切的关系。结合裂变气体分离沉降装置和片型静电沉降器进行理论公式的推导和计算，并将理论计算与试验结果进行的比较。试验是在４９－２反应堆元件破损监测小回路上进行的。由于裂变气体子代产物β衰变的计数率不但与试验小元件、反应堆运行参数、探测仪表的特性有关，而且与试验回路的特性和运行参数有着密切的关系。因此，该计算对建立新的试验回路，对设计新型的静电沉降器和对从事于反应堆元件破损监测的科技人员都有一定的参考价值。     

Release Behavior of Fission Gas from Coated Fuel Particles under Irradiation

TRISO coated fuel particles for HTGR were irradiated by two sweep gas capsules in order to study the release behavior of the fission gas and try to predict the failure fraction of the particles on the basis of the measurement. For verification of the predicted failure fraction, post irradiation examination was conducted, and failure fraction in a visual inspection and acid leaching fraction were measured. Agreement between the predicted failure fraction and the acid leaching fraction was good for these samples except one. From the release behavior from the intact particles, in-pile diffusion coefficients of Kr in LTI-PyC were estimated and expressed as D=(2.9–6.0)×10⁴exp(-2.55×10°/RT) (cm²/s), where R ids the gas constant (=8.314 J/K) and T the absolute temperature. It was recognized that the release from failed particles was controlled by diffusion at 1,600°C and that from intact particles, predominantly by recoil at 1,400°C.     

A Simulation Experiment of the Results of Fission Product Gas Release due to Fuel Cladding Failure using Water Circulation System

To examine the effects of fission product gas release into the coolant channel due to fuel cladding failure, two preliminary simulation experiments were undertaken using an electrical heater pin set in a system for water circulation. The first experiment represented a continuous release of gas from a small hole in the cladding, and the second a sudden burst of gas into the system due to plenum rupture. The first experiment revealed that the cladding surface temperature registered a sharp dip at the stagnation point of the water flow created immediately upstream of the gas releasing hole and that the temperature at the bottom of the dip was very sensitive to the ratio between the gas release rate and water flow rate. In the second experiment, representing plenum rupture, a pressure pulse of about 8 kg/cm² was registered on the wrapper tube, and the cladding surface temperature was found to rise by about 10°C within 0.1 sec (with heating at 100 W/cm), due to gas blanketing. The magnitude of the deformation and the transient strain on the cladding surface were also measured.     

回堆考验元件裂变气体测量分析

反应堆燃料元件的裂变气体释放率测量是辐照后检验的一项重要内容,它对于评价燃料元件的性能起着重要作用.回堆考验组件采用3×3-2再组装小组件方式,由一期考验的3根老棒、4根新棒和2根控制棒导向管组成.3×3-2小组件在中国原子能科学研究院重水研究堆辐照到燃耗(以金属铀计,全文同)30.9 GW*d/t(老棒)时,堆内出现破损信号.随后将其运至热室,非破坏性检验未发现元件棒破损.为此,采用激光刺孔方法将7根元件棒刺穿,测量元件棒气腔内压和裂变气体释放率.结果表明,元件棒内压均不低于再回堆考验前的压力值,从而进一步证实元件棒未发生破损,与一期考验元件相比,回堆后的燃料棒裂变气体释放率无明显增加.     

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.     

Fuel Microstructural Analysis of Incubation Burn-Up for Fission Gas Release

A study was made on an incubation burn-up for fission gas release using fuel swelling microstructural analysis. Conclusions of the study are: (1) The fuel microstructural analysis successfully determined the incubation burn-up. The analyzed values agreed with those estimated by the Halden empirical gas release model. (2) The incubation burn-up obtained from the Halden model was correlated with the fuel center temperature, but the micro-structural analysis was more dependent on the local fuel swelling temperature. (3) The incubation burn-up was attributed to the grain boundary diffusion process and the fuel local gaseous swelling.     

弥散型燃料裂变产物释放行为分析研究

为了获得弥散型燃料裂变产物向一回路冷却剂的释放特性,开展了弥散型燃料裂变产物释放行为研究,开发了适用于弥散型燃料的裂变产物源项计算程序,并对裂变产物源项进行了影响分析。结果表明:沾污铀和起泡破损后裂变产物的核素谱存在一定差异;裂变产物的释放与起泡当量直径的平方成正比;对于弥散型燃料而言,起泡破损中通过反冲释放的占比较低;相同破口条件下的弥散型和陶瓷型燃料中裂变产物的释放存在量级的差别。本文开发的程序能够用于分析弥散型燃料的裂变产物源项,为后续相关研究工程设计奠定基础。      

LOCA工况下环形燃料元件外包壳鼓胀爆破试验研究

为获得环形燃料元件外包壳在压水堆冷却剂丧失事故（LOCA）工况下鼓胀爆破温度和应变的经验关系式,为设计计算提供必要的输入,并初步评价其LOCA工况下的鼓胀爆破性能,在堆外对其开展了LOCA工况下的鼓胀爆破试验研究。在不同的升温速率和内压下,蒸汽环境中,以外表面红外加热的方式对环形燃料元件外包壳进行了鼓胀爆破试验。总结了试验得到的经验关系式,分析了试验中爆破温度和应变的影响因素,并将试验结果与美国核管理委员会出版的NUREG0630中的结果进行对比,验证了试验结果的合理性。获得的试验数据可用于环形燃料的设计、计算和改进。     

Effect of Fuel Pellet Eccentricity with Cladding on Fuel Rod Thermal Behavior under Reactivity Initiated Accident Condition

Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g?UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum temperature difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g?UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA conditions.