法国压水堆燃料元件新一代包壳材料的发展

概述了法国对核电站燃料元件包壳材料锆合金的开发与研究现状,着重介绍了所开发的新锆合金（Ｍ２,Ｍ３,Ｍ４,Ｍ５合金）在堆内外的性能。其中Ｍ４和Ｍ５合金包过央燃料棒燃耗达到５５ＧＷ．ｄ．ｔ＾－１的辐照考验结果表明,它们的堆内的腐蚀,蠕变和辐照伸长等性能优于改进型Ｚｒ－４合金包壳。     

燃料棒性能分析程序FUPAC V2.0的研发与验证

鉴于现有软件均缺乏CF3燃料组件N36锆合金包壳分析能力,开展了燃料棒性能分析程序FUPAC V2.0的研发工作。基于N36锆合金的堆外试验数据和N36锆合金包壳燃料棒池边检查数据,研究了N36锆合金的物理性能、腐蚀行为和辐照生长行为,初步建立了N36锆合金包壳相应模型。在现有自主化软件FUPAC V1.1的基础上,耦合入N36锆合金包壳分析模块,形成FUPAC V2.0,并进行了初步验证。验证结果表明：N36锆合金辐照生长模型和腐蚀模型与目前试验结果符合较好,FUPAC V2.0已实现计算N36锆合金包壳燃料棒性能的功能。     

燃料棒包壳氧化层溶解失效准则研究

为确定严重事故条件下燃料棒包壳温度达到金属锆的熔点后包壳氧化层的失效时间、再定位熔融物的成分以及氧化层失效对堆芯熔化进程的影响,本文基于熔融锆同时溶解UO_2和ZrO_2动力学模型及燃料棒包壳水侧氧化层的受力分析建立了氧化层在熔融锆中溶解失效的准则。以FPT-0实验结果验证后发现该失效准则可以较准确地预测包壳氧化层的溶解失效。为增加该准则在严重事故计算程序中的适用性,在燃料棒设计结构一定的条件下,进一步将该准则量化为温度的函数,分析表明包壳氧化程度和燃料棒温度上升速率是影响包壳氧化层失效温度的主要因素。利用该失效准则可以同时获得包壳氧化层失效后再定位的熔融物的质量及成分含量。     

锆合金包壳二次氢化破损模拟方法研究

压水堆燃料锆包壳管一旦出现破口，流入包壳内的水会在内外壁压差的作用下闪蒸为水蒸汽，在包壳管内壁引发锆水反应，使包壳管内壁由于大量吸氢而产生破损，称为二次氢脆。为了模拟压水堆一回路运行工况与锆包壳管的二次氢脆发生过程，通过理论强度计算与热工验证，自主设计锆合金包壳管二次氢脆实验堆外模拟装置，并针对ZIRLO合金包壳管开展双热源模拟实验。该装置实现了在一回路工况水平下的长期稳定运行，模拟结果显示ZIRLO合金管内外壁氧化并生成沿轴向自下而上浓度增加的氢化物。表明该装置解决了窄缝空间热分层现象带来的影响，可模拟包壳管二次氢脆过程中的一次破口失水、冷却水闪蒸及间隙蒸汽腐蚀行为，验证了该装置技术手段可行性。     

6 核电站用M5合金包壳管疲劳性能研究

M5锆合金是法国法马通公司开发研制的新一代燃料包壳材料，现已用作第3代改进型燃料组件AFA-3G燃料棒的包壳。     

燃料棒腐蚀和吸氢模型研究

锆合金包壳的腐蚀和吸氢性能是影响燃料棒堆内性能的重要因素。本文在锆合金包壳均匀腐蚀吸氢基本机理和现有模型的基础上,结合某特定燃料棒包壳材料的具体情况和使用特点,建立了包壳材料的均匀腐蚀和吸氢模型,并根据现有辐照数据对所建立的模型进行了验证。     

螺旋十字燃料棒束热力耦合特性研究

螺旋十字燃料（Helical-Cruciform Fuel,HCF）是一种革新形式的燃料组件,具有比传热面大、导热距离短、旋流交混强、无须格架支撑的特点,可提高堆芯功率密度。然而,HCF组件自支撑位置可能发生应力集中,导致包壳产生塑性变形甚至破裂。本文研究不同工况下HCF棒束热力耦合响应,以包含中心目标棒及周围的3×3棒束单元为计算域进行热力耦合分析,获得HCF组件的应力应变响应,评价包壳完整性。结果表明：HCF包壳外表面von Mises应力和塑性变形最大值总是出现在相邻燃料的接触位置;翼片顶部包壳应力受接触约束和包壳内外表面温差影响,肋部凹槽区域的应力与包壳径向温度梯度有关;相对于单相工况,饱和沸腾工况下HCF包壳塑性变形大但von Mises应力小;反应性引入事故和破口失水事故下,包壳的von Mises应力和塑性应变分别低于350 MPa和0.04,且包壳温度低于锆水反应温度。     

核电站用M5合金包壳管疲劳性能研究

M5锆合金是法国法马通公司开发研制的新一代燃料包壳材料,现已用作第3代改进型燃料组件AFA-3G燃料棒的包壳。核燃料包壳管在正常服役工况下,经受强烈的中子辐照,同时管内外均承受交变应力和温度作用,以及电厂定期开停堆,使得包壳经常产生周期性塑性变形,因此,锆合金包壳的疲劳行为研究成为核安全防护的重要课题之一。特别是当前核电站追求高燃耗、低燃料循环成本,换料周期更长,这样就对燃料包壳材料提出了更高的要求。核电站用包壳管疲劳失效是导致反应堆燃料元件发生破坏的主要原因之一。本工作对国产及法国产两种M5锆合金包壳管的疲劳…     

燃料棒破损热力学模块开发及在严重事故分析程序中的应用

包壳肿胀和破损是严重事故早期阶段的重要现象。包壳形变不仅会造成局部流动堵塞,同时,水蒸气会从破裂处进入包壳气隙,增加包壳被蒸汽氧化的表面积。广泛使用的一体化严重事故分析程序不能分析早期事故过程中燃料棒的热力学行为,判断包壳破裂也只是基于简单的参数模型。本文开发了分析燃料棒热力学行为的FRTMB模块,并集成在严重事故分析程序ISAA中。使用开发的耦合系统ISAA FRTMB分析了CAP1400反应堆直接注射（DVI）管线小破口事故过程中燃料棒的热力学行为,并预计了包壳破裂时间及相应的失效温度。计算结果整体验证了ISAA FRTMB分析瞬态事故过程中燃料棒热力学行为以及判断包壳破裂的适用性和可靠性。     

燃料棒破损热力学模块开发及在严重事故分析程序中的应用

包壳肿胀和破损是严重事故早期阶段的重要现象。包壳形变不仅会造成局部流动堵塞,同时,水蒸气会从破裂处进入包壳气隙,增加包壳被蒸汽氧化的表面积。广泛使用的一体化严重事故分析程序不能分析早期事故过程中燃料棒的热力学行为,判断包壳破裂也只是基于简单的参数模型。本文开发了分析燃料棒热力学行为的FRTMB模块,并集成在严重事故分析程序ISAA中。使用开发的耦合系统ISAA-FRTMB分析了CAP1400反应堆直接注射(DVI)管线小破口事故过程中燃料棒的热力学行为,并预计了包壳破裂时间及相应的失效温度。计算结果整体验证了ISAA-FRTMB分析瞬态事故过程中燃料棒热力学行为以及判断包壳破裂的适用性和可靠性。     

Simulation Test on Interactions of Zircaloy Claddings Deforming in Fuel Assembly under LOCA Conditions

In a hypothetical LOCA of LWR, it is assumed that Zircaloy claddings of fuel assemblies occur ballooning and cause thermal and mechanical interactions among themselves. To investigate the phenomena, burst tests were conducted with a single rod fuel simulator placed in the center of eight non-pressurized external heater pipes. It was found that the deformation and rupture behavior of ballooning cladding which made contact with the pipes depends on the following conditions: (1) temperature difference between cladding and heater pipes; higher temperature in the cladding than in the pipes produces an extended deformation and contact area, (2) internal pressure of cladding, and (3) temperature level at which ballooning takes place.

It was observed and analyzed that cladding hoop tensile stress is largest at the inflection points generated in the contact surface of the expanding cladding.     

The effect of anisotropy on the ballooning of Zircaloy cladding

The effect of creep anisotropy on the ballooning of Zircaloy LWR fuel rod cladding tubes is investigated. A perturbation method for calculating the effect of temperature inhomogenities is developed further. The results are compared with a simple method that is not restricted to small deviations from axisymmetry. The perturbation method is shown to have only limited applicability to the Zircaloy ballooning problem. The other method which assumes that the cladding tubes retain a circular cross-section provides a more useful technique for fuel rod behaviour analysis. Studies of the bending of cladding tubes and the effect of restraint on deformation and failure are presented. Apart from cladding tube bending the effects of creep anisotropy on clad deformation and failure are not large.     

One and two dimension analysis of the clad ballooning in pressurized fuel rods

After a literature survey on the Zircaloy clad burst of PWR pressurized fuel rods, a burst model is proposed which is coupled to a viscoplastic deformation law for the Zircaloy at high temperatures. It is then possible to take into account experimental data by fitting only one parameter of the constitution law. The two-dimensional analysis gives a good approach of the ballooning shape but does not seem really necessary to the fuel modelling in a LOCA code.     

Effects of Rod Pre-Pressurization on Light Water Reactor Fuel Behavior during Reactivity Initiated Accident Conditions

Previously pressurized (pre-pressurized) fuel rod tests recently performed in the Nuclear Safety Research Reactor (NSRR) investigate the effects of initial internal pressure on fuel rod behavior during reactivity initiated accident (RIA) conditions. A single PWR type fuel rod was contained within a waterfilled, ambient temperature and ambient pressure capsule. The fuel rod was then heated by the pulsing operation of the NSRR.

Results from the tests show that the effect of pre-pressurization was significant for the fuel rods with initial internal pressure of 0.8 MPa and above, and fuel rod failure occurred from rupture of the cladding with lower threshold energy deposition for failure as the initial internal pressure was increased. The cladding rupture was governed mainly by the cladding temperature rise, not by the rod internal pressure rise during the transient. The relationships between cladding burst pressure and cladding burst temperature and between cladding strain and cladding temperature at cladding rupture obtained in the present study under an RIA condition agree with the results obtained from various in- and ex-reactor experiments under a LOCA condition, although the obtained time-averaged strain rate of the Zircaloy cladding was much greater than that in a LOCA condition.     

Fracture of Zircaloy cladding under simulated power ramp conditions

Behaviour of Zircaloy cladding under stresses imposed by fuel dimensional changes is discussed and results presented which suggest that the correct choice of cladding properties, in particular cold-work level and texture, is important if the probability of low ductility failures in fuel pins is to be reduced. Evidence is presented, based on laboratory tests and post-irradiation examination of cladding, which leads to the conclusion that failure of Zircaloy can take place by stress-corrosion cracking in the presence of fission product iodine and it is further demonstrated that the process is strongly stress-dependent. Consideration is given to the influence of irradiation exposure, strain rate and adverse stress systems, for instance at inter-pellet ridges or fuel cracks, on the cracking susceptibility of Zircaloy cladding. The significance of the results is discussed in relation to a proposed stress-corrosion cracking mechanism.     

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code.The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from the plenum - resulting in a considerable delay of the predicted moment of cladding rupture.     

LWR fuel rod behavior during severe accidents

Chemical interactions between UO₂ fuel and Zircaloy cladding up to 2350°C are described. UO₂/Zircaloy single effects tests have been performed with short LWR fuel rod segments in inert gas and under oxidizing conditions. The reaction kinetics of molten Zircaloy cladding with solid UO₂ fuel has been investigated with UO₂ crucibles containing molten Zircaloy. The UO₂/Zircaloy reactions obey parabolic rate laws. The oxygen uptake by solid Zircaloy due to chemical interaction with UO₂ occurs nearly as quickly as that from the reaction with steam or oxygen.To study the competing effects of the external and internal cladding oxidation under realistic boundary conditions and the influence of the uncontrolled temperature escalation due to the exothermic steam/Zircaloy reaction on the maximum cladding temperature, single rod and bundle experiments have been performed. Electrically heated fuel rod simulators, including absorber rod material (Ag, In, Cd alloy), guide tubes and grid spacers are used. The maximum measured cladding temperature during the temperature escalation was about 2200°C. The failure temperature of the absorber rods and the extent of bundle damage depends on the guide tube material (Zircaloy or stainless steel) and varies between 1200 and 1350°C. The molten materials and liquid reaction products can relocate and form large coherent lumps on solidification, which may result in complete blockage of the fuel rod bundle cross section. In the future, 7 × 7 bundle experiments of 2 m overall length will be performed in the new CORA facility to study, in addition, the influence of quenching on fuel rod integrity.     

Zircaloy-4 Cladding Embrittlement due to Inner Surface Oxidation under Simulated Loss-of-Coolant Condition

Embrittlement of Zircaloy-4 cladding by oxidation of the inner surface occurring in an LWR loss-of-coolant accident was studied using simulated fuel containing of A1₂O₃ pellets sheathed in Zircaloy-4 specimen cladding, filled with Ar gas, and sealed. This simulated fuel rod was heated from outside until the isothermal oxidation temperature between 880 and 1,167°C was obtained after the cladding burst. This exposed the inner surface of the cladding to the environmental atmosphere, provided by steam flowing at a constant rate in the range of 0.13–1.6 g/cm²-min.

The embrittlement of the specimen due to inner surface oxidation is influenced primarily by the amount of hydrogen absorbed by the Zircaloy-4. Ring compression tests conducted at 100°C on test pieces constituted of sliced sections of oxidized specimen showed that Zircaloy containing more than 200–300 wt.ppm of absorbed hydrogen became brittle when oxidized at temperatures above 1,000°C. In the range of oxidation temperature 932 to 972°C, brittleness did not appear below 500–750 wt.ppm absorbed hydrogen.

Hydrogen absorbed by the Zircaloy precipitated in the form of fine hydride crystals formed along previous β-phase grain boundaries. Peaks were found in the distribution of hydrogen absorbed on the inner surface, at a distance of 15–45 mm upward and downward of the rupture opening. Within this range, the distance was influenced by the oxidation temperature and steam flow rate.     

Calculational procedure for determining creep collapse of LWR fuel rods

Observed collapses in pressurized water reactor fuel rods have been attributed to the radiation enhanced creep of Zircaloy cladding into regions where separations in the fuel pellet stack have occurred. A computer code, COLAPX, has been written to determine the growth of ovality and the ultimate collapse of fuel rod cladding under reactor operating conditions. This paper describes the theoretical bases of this code, the finite element formulation used, the constitutive relations between the displacement fields and the element forces, and the radiation, temperature and stress dependent material model for creep of Zircaloy tubing. Comparisons of the creep rate predictions and of the ovality predictions with data from irradiated tubes and fuel cladding are presented.     

Measurement of Irradiation Creep Coefficient of Fine–Grained Isotropic Graphite

In order to clarify the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes, the stress-strain distribution of the cladding was estimated by finite element method (FEM) analysis. The mechanical properties of α-phase of zirconium and zirconium hydride required for the analysis were examined by means of an ultrasonic pulse-echo method and a tensile test. It was found from the analysis that the non-hydrided cladding has the highest equivalent plastic strain at the inner surface of the cladding, suggesting that the fracture initiated at the inner surface of the cladding. Since the hydride accumulated layer located in the outer surface of the hydrided cladding fails at a lower internal pressure, the crack appears to initiate at the outer surface of the cladding. The fracture behavior estimated from the stress states of the hydrided cladding was in good agreement with the experimental results obtained by pulse irradiation tests of the Nuclear Safety Research Reactor (NSRR) and high-pressurization-rate burst tests in the Japan Atomic Energy Research Institute (JAERI).