破损燃料棒二次氢化行为观察与分析

燃料棒在堆内运行时,由于初次破口会导致包壳发生二次氢化现象,二次氢化是导致燃料棒发生严重破损的重要因素。针对实际工况下的破损燃料棒,在中国原子能科学研究院燃料与材料检验设施(303热室)上开展了相关辐照后检验,并采用热室金相手段,对燃料棒二次氢化行为进行了观察分析。结果表明:二次氢化破口有明显的氢化肿胀现象;氢化物分阶段从内壁扩散到外壁,并形成"日爆"现象;二次氢化部位芯块温度明显升高,并会导致芯块气孔迁移、芯块晶粒长大、柱状晶生长等现象发生;相比未破损棒,破损棒二次氢化部位水侧氧化膜厚度有增加现象,但仍处于正常范围内。     

核电厂AFA 3G钆棒破损原因分析

核电厂运行期间,发现冷却剂放射性水平增高,怀疑1根钆(Gd)棒发生破损。通过在中国原子能科学研究院进行的热室检验,确定该Gd棒下端塞环焊缝处存在破损,且堵孔焊点失效。利用氢分析仪对上、下端环焊缝附近以及破损的下端塞进行了氢分析,检验结果表明:下端塞环焊缝处的氢含量为1720μg/g,为二次氢化破口;上端塞附近氢含量为133μg/g,破损堵孔焊点为Gd棒的初次破口。堵孔焊点失效是钆棒破损的根本原因。     

压水堆燃料棒三维热-力学性能的精细模拟

压水堆燃料棒工作在复杂的辐照、热和力学环境中,对其性能进行定量评估涉及多种复杂的物理现象。目前常用的燃料性能分析程序一般对结构采用简化的轴对称假设,对辐照肿胀、辐照蠕变和高温蠕变等物理现象以及辐照-热-力等物理场之间的耦合考虑并不充分。基于ABAQUS有限元求解框架,开发了压水堆燃料棒三维热-力学性能的模拟程序,利用程序对压水堆燃料棒进行了稳态分析,以及升功率和反应性引入事故两种瞬态分析。结果表明:辐照引起燃料致密化和肿胀对燃料温度变化有重要影响;芯块应变增加主要是由裂变产物肿胀引起的;芯块几何结构导致包壳应力集中发生在芯块间的交界面处;燃料棒功率的急剧变化会加快芯块表面破裂的进程;反应性引入事故会导致芯块从内部开始破裂,并会引发芯块-包壳的接触。     

国产压水堆核电站40 GW•d/tU乏燃料棒金相检验

为评价国产燃料棒在较高燃耗水平下的辐照性能,在中国原子能科学研究院燃料与材料检验设施（303热室）对燃耗为40 GW•d/tU的国产压水堆核电站乏燃料棒进行了金相检验。检验内容包括芯块宏观与微观组织、包壳水侧腐蚀与氢化物分布、芯块-包壳相互作用状况等。金相检验结果表明：40 GW•d/tU燃耗下,芯块未发生明显的轮廓变化,气孔率为3.3%～5.8%,晶粒组织为等轴晶,平均晶粒尺寸为7.2 μm;Zr合金最大水侧氧化膜厚度为23 μm,氢化物分布和含量正常,最大氢含量约为150 μg/g,同时不同部位的包壳氢含量与水侧氧化膜厚度基本呈线性关系,水侧腐蚀处于正常水平;包壳内壁有局部轻微腐蚀,包壳与芯块之间存在间隙,未发生包壳与芯块相互作用情况。     

发生落棒事故时的PCI研究

当反应堆发生落棒事故时,燃料芯块与包壳的相互作用瞬间增强,易造成燃料棒破损,从而影响核电站的正常运行.本文介绍了反应堆Ⅱ类瞬态下燃料棒芯块与包壳相互作用的机理和定量分析方法,并针对大亚湾核电站18个月换料的燃料管理方案进行了发生落棒事故时的PCI热力学评价.初步的研究结果表明:如果在自然循环长度和延伸燃耗运行期内发生落棒事故,对于基负荷运行和基负荷一次调频运行,均有PCI的应力裕量,不会造成燃料棒破损.     

破损燃料组件热室检查技术研究

燃料组件破损直接影响了反应堆的安全运行,分析燃料组件破损原因是燃料组件研发改进的重要环节。通过破损燃料组件水下解体、破口位置定位、破口试样取样等关键技术的研究,建立了破损燃料组件热室检查方法。研究结果表明,该技术路线合理,检查方法可行,为热室条件下开展燃料元件破损检查提供了技术途径。?      

3×3-2回堆考验组件辐照后检验

文章涉及3×3-2可拆卸组件和相应热室检验装置的设计及热室内组件的远距离再组装和拆卸;组件检验及平均燃耗(以金属铀计,全文同)为30.9 GW*d/t和6.5 GW*d/t燃料棒的非破坏性检验和破坏性检验;室温和高温下的燃料棒检漏.主要结果是可拆卸组件及相应热室检验装置的设计合理,热室内远距离操作可行;经堆内考验后组件结构完整,但格架辐照弹簧松弛量较大;燃料棒性能完好,环脊明显,棒径减小显著;采用多种方法检漏,未发现破损的燃料棒.     

压水堆核电站乏燃料元件γ扫描测量

在压水堆核电站乏燃料元件检验中,完成了4根完整元件棒、4根破损元件棒的γ扫描测量,元件燃耗分布在9600～45000 MW•d/t（U）之间,获得了完整元件轴向相对燃耗分布、破损元件¹³⁷Cs分布及迁移流失情况。结果显示,破损元件均存在不同程度的Cs迁移流失,破口处存在¹³⁷Cs计数突变（降低）。破损元件¹³⁴Cs/¹³⁷Cs原子比分布与相邻完整元件基本一致,表明¹³⁴Cs、¹³⁷Cs流失比例近似相等,可用¹³⁴Cs/¹³⁷Cs原子比表征其相对燃耗分布;破口处可通过低挥发性核素¹⁵⁴Eu计数水平判断燃料芯块是否缺失。检验结果可为燃料元件破损原因分析及堆内行为分析提供重要依据。     

芯块偏置对核反应堆燃料棒温度场和热通量分布的影响

本文分析芯块偏置对核反应堆燃料棒稳态温度分布和热通量分布的影响,推导出芯块偏置的燃料棒无量纲温度及热通量分布关系式。计算结果表明:芯块偏置不仅使燃料最高温度位置发生偏移,而且造成燃料棒表面温度和热通量沿圆周方向不均匀分布。     

压水堆核电站乏燃料元件γ扫描测量

在压水堆核电站乏燃料元件检验中,完成了4根完整元件棒、4根破损元件棒的γ扫描测量,元件燃耗分布在9 600～45 000 MW·d/t(U)之间,获得了完整元件轴向相对燃耗分布、破损元件~(137)Cs分布及迁移流失情况。结果显示,破损元件均存在不同程度的Cs迁移流失,破口处存在~(137)Cs计数突变(降低)。破损元件~(134)Cs/~(137)Cs原子比分布与相邻完整元件基本一致,表明~(134)Cs、~(137)Cs流失比例近似相等,可用~(134)Cs/~(137)Cs原子比表征其相对燃耗分布;破口处可通过低挥发性核素~(154)Eu计数水平判断燃料芯块是否缺失。检验结果可为燃料元件破损原因分析及堆内行为分析提供重要依据。     

Rupture of spent fuel Zircaloy cladding in dry storage due to delayed hydride cracking

Delayed hydride cracking in the Zircaloy alloy has been considered as a possible degradation mechanism of spent nuclear fuel cladding in interim dry storage. Some recent in-core fuel failures indicated that a long axial crack developed in the cladding was a secondary failure by delayed hydride cracking. The aim of this study is to define the effects of hydride reorientation on the failure of Zircaloy cladding. Different hydride orientations, the amount of zirconium hydride and various cracking types, all have been considered for their effects on the crack growth and stability of the cladding, and have been thoroughly discussed in this paper. A finite element computer code, ANSYS, has been used in conjunction with the strain energy density theory. In summary, the crack propagation will be aggravated if the hydride orientation is shifted from the circumferential to the radial direction. For a larger crack length, the zirconium hydride plays an important role in affecting the crack growth because the strain energy density factor increases as the hydride approaches the crack tip. Furthermore, when thermal effects are considered, a compressive stress exists at the inner side of the cladding, while a tensile stress is found at the outer side of cladding, thus resulting in crack propagation from the outer side to the inner side of the cladding. These findings are in accordance with other experimental results in related literature.     

Post-irradiation examinations focused on fuel integrity of spent BWR-MOX and PWR-UO2 fuels stored for 20 years

Post-irradiation examinations (PIEs) of spent BWR-MOX and PWR-UO₂ fuel rods irradiated in commercial LWRs and stored for 20 years were carried out to evaluate fuel integrity during storage. Average burn-up of five fuel rods of the BWR-MOX fuel was about 20 MWd/kgHM and that of the PWR-UO₂ fuel was 58 MWd/kgHM. The PIE items included: (a) visual inspection of the cladding surface, (b) puncture test, (c) ceramographic observation on the pellet and cladding, (d) pellet density, (e) electron probe microanalysis of the pellet, (f) cladding tensile test, (g) hydrogen content and hydride orientation in the cladding, and (h) hydrogen redistribution in the cladding under temperature gradient. The PIE results showed no marked difference in the visual inspection, fission gas release, oxide layer thickness, pellet microstructure, and cladding mechanical properties or hydride orientation after storage. The result of the hydrogen redistribution experiment showed that hydrogen migration had little effect on the fuel integrity during dry storage. Hydrogen migration on the fuel rod for 40 years of storage was evaluated using the heat of transport obtained in the hydrogen redistribution experiment and calculated result showed that hydrogen migration had little effect on the fuel integrity during dry storage.     

Simulation of the Hydriding Failure of Fuel Elements in Water-Cooled Reactors

A physical model is proposed for local massive hydriding of the cladding of an initially sealed fuel element with fuel containing excess moisture. The growth of massive hydride in the cladding is described in a two-dimensional geometry taking account of hydrogen diffusion in zirconium and thermal diffusion. Incorporating the model developed into the RTOP code makes it possible to perform self-consistent calculations of various scenarios of unsealing of the fuel elements as a result of massive hydriding of the cladding under prescribed operating conditions. The results of test calculations are compared with existing data. It is shown that there exists a critical content of moisture inside a sealed fuel element for which through growth of hydride and unsealing of the fuel element occur. This threshold value depends on the state of the oxide film on the inner surface of the cladding, the power density distribution per unit length over the height of a fuel element, the geometry of the fuel element, the temperature distrubution in the cladding, and the coefficient of hydrogen transfer in the cladding and hydride.     

Failure of high burnup fuels under reactivity-initiated accident conditions

Pulse irradiation experiments of high burnup light-water-reactor fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO₂ rod at a burnup of 69 GW d/t failed due to pellet-cladding mechanical interaction (PCMI) in the test LS-1. The fuel enthalpy at which fuel failure occurred was comparable to those for PWR-UO₂ rods of 71 to 77 GW d/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuel rods suggested that the morphology of hydride precipitation, which depends on the cladding texture, affects the fuel failure limit. The tests BZ-1 and BZ-2 with PWR-MOX rods of 48 and 59 GW d/t, respectively, also resulted in PCMI failure. The fuel enthalpies at failure were consistent with a tendency formed by the previous test results with UO₂ fuel rods, if the failure enthalpy is plotted as a function of the cladding outer oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states including oxidation and hydride precipitation, and the same failure limit is applicable to UO₂ and MOX fuels below 59 GW d/t.     

Influence of Cladding-Peripheral Hydride on Mechanical Fuel Failure under Reactivity-Initiated Accident Conditions

In order to investigate the influence of hydrogen embrittlement on fuel failure under reactivity-initiated accident (RIA) conditions, pulse irradiation experiments were performed with unirradiated fuel rods at the Nuclear Safety Research Reactor (NSRR). Fresh cladding was pre-hydrided so that the other factors of cladding degradation, such as irradiation damage and oxidation, were excluded. Hydride clusters are circumferentially oriented and localized in the cladding periphery in order to simulate ‘hydride rim’ which is formed in high burnup PWR cladding. The present study demonstrated hydride-assisted pellet-cladding mechanical interaction (PCMI) failure which has been observed in high burnup fuel experiments. The fuel enthalpy at failure was lower when the cladding had a thicker hydride rim where surface cracks were easily generated. It indicates that the failure limit is highly correlated with the stress intensity factor assuming that the crack depth is equivalent to the hydride rim thickness. Hence, we conclude that hydride rim formation is the primary factor of decreasing the failure limit for high burnup fuels. Based on the experimental results together with an analysis on cladding mechanical state during PCMI, the present study suggests a process of through-wall crack generation which is originated with brittle cracking within the hydride rim.     

Transient Elongation of a Fresh Fuel Rod under Reactivity Initiated Accident Conditions

In-pile experiments of fresh fuel rods under reactivity initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor at the Japan Atomic Energy Research Institute in order to understand the basic pellet cladding mechanical interaction (PCMI) behavior. Rapid fuel pellet expansion due to a power excursion would cause radial and longitudinal deformation of the cladding. This PCMI could be one of the possible incipient failure modes of an embrittled cladding of a high burnup fuel under the RIA conditions.

Basic PCMI behavior was studied by measuring cladding deformation of a fresh fuel rod without complicated irradiation effects. The transient elongation measurements of the fuel with two kinds of gap width indicated not only PCMI-induced cladding elongation, but also reduction of the pellet stack displacement by the cladding constraint. In the tests under a high-pressure and high-temperature condition simulating an operation condition of BWRs, additional ridge-type cladding deformation was generated due to the axial collapse of the cladding. A preliminary analysis for interpretation of the tests was made using a computer code for the transient analysis of fuel rods, FRAP-T6.     

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.     

Theoretical evaluation of rifled cladding for LWR fuel: A study based on the computer codes INTERPIN and FREY-01

Rifled Cladding deviates from regular fuel designs mainly by the inner surface of the cladding tube being prismatic, with a large number of faces. The design has been tested in the R2 test reactor at Studsvik, and is scheduled for testing in a Swedish BWR. Its observed technical merit is to reduce the fission gas release and increase the failure threshold during overpowers. The present study aims to find an explanation of the observed increase in failure threshold, i.e., the heat rating at which cladding failure occurs. Failure is strongly related to the maximum mechanical stress. The INTERPIN code was used to calculate the starting conditions for an overpower event, and the FREY-01 code was used for finite element calculations of cladding stress distributions during the overpower. FREY-01 accounts for the kinetics of fuel pellet cracking and its impact on cladding stresses. The study focuses on the conditions that produce the highest cladding stress. It was found that the worst case is associated with cracks opening close to the fuel-cladding contact loci, i.e., the midface positions of the prismatic surface, with no additional cracks at non-contact positions. This situation is similar to what holds for regular fuel. If, however, cracks also open at corner positions, then the calculated maximum stresses are considerably less. This latter situation is, in fact, observed in hot cell examinations of ramped rifled cladding rods. Therefore, the improved failure resistance of rifled cladding rods can be explained in terms of a reduction of the strong azimuthal mechanical interaction between fuel and cladding that may occur in fuel rods of regular design.     

Pellet compliance model based on out-of-pile simulation

The thermal and mechanical behavior of fuel rods is significantly influenced by the extent of their relocation and by compliance of the cracked pellets. Movement of the cracked pellet pieces towards the cladding results in softer pellets with crack voids which accommodate some fraction of the thermoelastic pellet deformation and make the pellet more compliant under the restraint of the cladding. It is difficult to model such a pellet compliance independently of experimental observations because the cracked pellet behavior is uncertain by nature.Electrically heated simulation of pellet-cladding mechanical interaction (PCMI) facilitates much quicker and more flexible experimentation than actual in-pile tests. Testing apparatus consists of the simulated fuel rod with hollow UO₂ pellets and a tungsten rod in the center, and a diameter measuring device including three pairs of diameter sensors. Test parameters include the pellet-cladding gap and the cladding thickness. Results show that rods with a smaller gap have a larger increasing rate of cladding diameter. This suggests that a group of cracked pellet pieces induced by thermal stress has an apparent compliance which increases with pellet-cladding gap. Results also show more sensitivity to cladding thickness than those calculated assuming pellets having intrinsic stiffness. This also suggests the compliant nature of cracked pellets.Such a compliant nature can almost be described by reducing the elasticity of the pellet. A simple pellet compliance model was obtained by fitting calculations with measurements to describe a cracked pellet as a uniform axisymmetric body with apparent elasticity.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.