不同氢化物取向的Zr-4管材制备方法研究

电解渗氢制备氢含量分别为140±20?ppm（1 ppm=1 μg/g）和260±20?ppm的Zr-4管材，通过加载高压气体使管材发生氢化物应力再取向效应，最终获得不同氢化物取向的Zr-4管材。结果表明:以电解参数为105?mA/cm2×2?h和110?mA/cm2×(4?h+50?min)进行电解后可分别获得氢含量为140±20?ppm和260±20?ppm的管材。当温度循环为400~200℃，实际升温和降温速率分别约为10℃/min和0.75℃/min时，通过调节压力和保温时间，仅单次热循环即可获得氢化物取向因子高达0.7的管材。      

Zr-2包壳管氢化物分布及取向因子测量

本工作对辐照后燃料元件的Zr-2包壳管氢化物分布及取向因子进行测量与分析。通过对两根元件的样品进行微观检验，得到其中的氢含量、氢化物分布、氢化物形态和取向因子，为包壳管生产工艺、元件设计及堆内运行提供依据。     

Zr—4板中氢化物应力再取向的研究

在张应力为55—180MPa和150(?)400℃温度循环条件下,研究了应力和温度循环次数对氢含量为220μg/g的Zr-4板中氢化物再取向程度影响。随着应力增大和温度循环次数增加,氢化物再取向程度增大,但是氢化物发生再取向存在一个应力阈值,当张应力低于应力阈值时,即使增加温度循环次数,氢化物再取向也不明显。应力阈值又会随温度循环次数增加而降低。     

加工工艺对Zr-4合金薄壁管材氢化物取向的影响

采用两种轧制方法、5种退火制度与4种矫直压下量,研究了冷轧工艺、成品退火制度、矫直工艺对Zr-4合金薄壁管材氢化物取向的影响.结果表明,冷轧工艺对氢化物取向因子的影响不大;在合适的冷轧工艺条件下,消除应力退火的管材,氢化物取向因子非常小,氢化物几乎都沿周向分布;随退火温度的增加氢化物取向因子增大;矫直压下量对氢化物取向因子的影响较大,随压下量的增大氢化物取向因子增大,管材靠外壁的氢化物取向因子高于其内壁.     

Zr—4管中氢化物分布的应力再取向研究

渗200ppm 氢的 Zr-4管,在周向应力为70—180MPa 和150(?)400℃的条件下,研究了应力和温度循环次数对氢化物再取向的影响。随着应力增大或温度循环次数增加,一部份氢化物由周向分布转变成径向分布。氢化物发生再取向时,先从管壁外表面开始,逐步向内推进。在本实验条件下,当氢化物发生再取向后,并没有全部转变成径向分布,f_(45)只达到0.5,说明控制织构对控制氢化物再取向仍然有效。     

Zr-2包壳管爆破力学性能

核燃料元件是反应堆的核心部件，由燃料芯块、包壳及其构件组成。由于燃料元件的运行环境比较恶劣，中子辐照、冷却剂的腐蚀及在开堆、停堆、和运行后期燃料芯块与包壳的机械相互作用和裂变气体产物的释放，使包壳管承受双向应力，均会造成燃料元件的力学性能下降，形成安全隐患，它的安全性能直接影响反应堆的安全可靠性。为更好地模拟包壳在堆内的受力状态，一般采用内压爆破试验来获得包壳材料的断裂强度与延性数据。     

M5锆合金包壳管轴向和环向拉伸性能测试

对国产及法国产两种M5锆合金包壳管进行拉伸性能测试,包括轴向拉伸及其环向拉伸.测试温度为室温及375 ℃.测试获得了φ9.5 mm×0.57 mm M5锆合金包壳管轴向和环向在两种试验温度下的抗拉强度σb、屈服强度σ0.2、延伸率δ等性能指标.     

氢及氢化物对锆合金疲劳裂纹扩展速率的影响及估算

采用紧凑拉伸（CT）试样．研究了不同氢含量的Zr-4及Zr-Sn-Nb合金在室温下疲劳加载裂纹扩展（dα/dN）行为．用扫描电镜观察了断口形貌。结果表明，氢含量对疲劳裂纹扩展速率影响微弱，疲劳断裂受通常的裂纹萌生、稳态扩展和瞬间断裂机制控制。根据疲劳裂纹扩展机理．导出了裂纹扩展门槛值△Kth的关系式．得出了一个描述疲劳裂纹扩展速率油（dα/dN）与材料性能常数之间的关系式，该关系式可用于预测材料的疲劳裂纹扩展速率。用锆合金实验数据对（dα/dN）预测表达式进行验证．结果表明，预测值与实验值吻合较好。     

模拟裂变产物腐蚀对国产不锈钢包壳管室温爆破性能的影响

未经腐蚀的国产不锈钢包壳管与在两种不同氧势下经模拟裂变产物腐蚀后的包壳管相比，其室温爆破强度和周向延伸率有明显变化。由于氧势对包壳管腐蚀深度影响显著，从而也明显影响包壳管的爆破强度和周向延伸率。爆破口的微观形貌分析进一步显示了裂变产物腐蚀对包壳管力学性能的影响。试验方法对结果的影响也作了初步探讨。     

Effect of Pre-Hydriding on Post-Irradiation Ring-Tensile Properties of Zircaloy-2 Cladding Tubes

Abstract

Zircaloy-2 tubes were hydrided up to a nominal content of 200 ppm and irradiated as fuel claddings in HBWR. Post-irradiation ring-tensile testing revealed that hydrogen enhances the irradiation-induced decrease of elongation and wall thickness reduction at room temperature. On the other hand, no effect of hydrogen was observed on ultimate tensile strength. With testings at 300°C, the effect was negligible on elongation too. From the evaluation of the test results including metallographic observation of ring specimens after fracture, it was concluded that segregation of hydrides due to thermal diffusion of hydrogen during irradiation was at least a part responsible to the above effect of hydrogen enhancing embrittlement.     

Reorientation of Hydrides and Its Effect on the Mechanical Properties of Zr-Nb-Sn-Fe Cladding Tubes

The reorientation of hydrides and its effect on the fracture of Zr—Nb—Sn—Fe cladding tubes were investigated. The reorientation of hydrides along the radial direction was most pronounced if the tube was cooled from 573 to 473 K under circumferential loading. Reorientation occurred much less frequently at either higher (from 673 to 573 K) or lower (from 473 to 373 K) temperature range. The strength decreased to 250 MPa and the ductility decreased to zero in the tube which was reorientation-treated from 573 to 473 K (RS32AC). Fracture surface of RS32AC sample exhibited flat cleavage fracture along the radial hydride platelet. The reorientation of hydrides was also found to increase with increase of loading time, suggesting time dependent stress-aided dissolution of circumferential hydrides and re-precipitation of radial hydrides.     

Zr－4合金管环向性能测试研究

采用对开式拉伸法（ＮＯＬ环法），对反应堆中常用构件Ｚｒ－４合金薄壁细管，在不同温度条件下进行了环向拉伸试验。通过对拉伸曲线的修正和炉内试样颈缩处承载面积的确定，得到了Ｚｒ－４合金管在不同温度条件下，环向拉伸的真应力一真应变关系及强度、塑性指标。     

Zr-4合金小试样高温疲劳行为研究

基于Zr-4合金漏斗薄片小试样,完成了室温和400℃高温下的等幅横向应变循环与应变疲劳试验.根据弹塑性有限元分析,建立了基于局部应变等效的应变换算方法,并结合实验结果,得到了估算Zr-4合金应变疲劳寿命的Manson-Coffin模型.结果表明:低应变幅下,Zr-4合金表现出循环软化特征;高应变幅下,Zr-4合金表现出循环强化特征.高温严重降低了低应变幅下Zr-4合金的疲劳寿命,随着应变幅增加,温度影响趋弱.分析表明,基于传统应变转换公式的M-C模型用于估算疲劳寿命偏于保守.     

Mechanical Properties of Neutron Irradiated Fuel Cladding Tubes

There was few post irradiation examination data on the mechanical properties of domestic fuel cladding tubes used for light water reactors, then those data obtained abroad have been often used in the fuel design or fuel performance codes. Although, many reports discussed the deformation mechanism of the tube, almost all the data were not obtained from irradiated specimens but unirradiated ones. In recent years, systematic post irradiation examinations on domestic fuel elements used in Japanese light water reactors and the related studies were performed.

This report first summarizes briefly the crystallographic texture which characterizes the properties of Zircaloy fuel cladding tubes, followed by an explanation of basic properties such as elasticity, plasticity, creep and fatigue. Finally, the up-to-date results are introduced.     

Effect of Iodine on Fatigue Properties of Zircaloy-2 Fuel Cladding under Cyclic Tensile Stress

Effects of iodine on the cyclic tensile properties of Zircaloy-2 have been investigated at 350°C. Notched specimens, which were grooved circumferentially, and unnotched specimens were machined from the stress-relieved fuel cladding. Cyclic tensile stress was loaded on the specimens in order to get the fatigue life and failure ductility in air and in an Ar atmosphere containing iodine.

The fatigue life and failure ductility of the unnotched specimens, which were pulled parallel to the longitudinal direction of the cladding, showed little decrease due to effects of the iodine environment, though many pittings due to iodine corrosion were observed on the inner surfaces.

On the other hand, the fatigue properties of the notched specimens indicated the effects of iodine. This difference implies that the corrosion fatigue property of Zircaloy-2 fuel cladding under the iodine environment depends on the texture and stress state of the specimens.     

锆合金薄壁细管的单调拉伸与低周疲劳试验研究

利用自行研制的高温夹具完成了Zr-1Nb合金和Zr-4合金薄壁短管试样不同温度下的单调拉伸和375℃下的等幅低周疲劳试验,获得了两种锆合金的单调和循环本构关系及Manson-Coffin寿命估算模型。研究结果表明：Zr-1Nb合金和Zr-4合金的弹性模量、屈服强度、抗拉强度以及应变硬化程度明显下降。随着温度的升高,温度对Zr-4合金的应变硬化程度的影响逐渐减弱;应变速率对Zr-4合金的拉伸性能的影响微弱。在等幅应变循环过程中,Zr-4合金表现为循环硬化,应变幅越低,硬化现象越明显;Zr-1Nb在较低应变幅下表现为循环硬化特性,而在较高应变幅下表现为循环软化。相对于单调拉伸行为,Zr-4合金在不同温度下的循环行为均表现出明显的强化特性。