包壳

包壳管内表面渗B，渗B量与相容时间的平方根成正比，渗B深度不随相容时间变化；

There is B penetration in cladding inner surface , the amount of B penetration is propotional to the square root of the test period .

熔化情况下Zr（-4）合金包壳管氧化膜厚度变化的计算

Calculation of the Oxide Film Thickness Changing for Zircaloy Cladding in In-pile Meltdown Case

热处理对U3Si2－Al燃料板包壳显微组织及其厚度测量的影响

The Effect of Heat-Treatment on the Microstructure and Thickness Measurement of Cladding for U_3Si_2-Al Fuel Plates

快堆包壳用ODS铁素体合金中Ti的强化作用

Effects of Ti on strengthening of ODS ferritic alloy for advanced FBR cladding application

用于超临界水堆燃料包壳的ODS铁素体钢的研究进展

Progress of Using Oxide Dispersion Strengthened Ferritic Steels as Fuel Cladding Materials in Supercritical Water Reactor

国产低锡Zr-4包壳管的电子束焊接

Electron beam welding for clad tube of lower tin Zr-4 alloy

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。

Axial and ring tensile test of M5 alloy cladding tube was carried out .

低锡Zr-4包壳管电子束焊接时发生的合金元素蒸发现象

Alloy Element Evaporation Phenomenon of Low Sn Zr 4 Cladding Tube During Electric Beam Welding

M5锆合金主要应用于核反应堆堆芯的燃料包壳和结构件。

M5 zirconium alloy is mainly applied in the nuclear reactor core as the fuel cladding tubes and structure materials .

氧化物弥散强化（ODS）铁素体合金具有强度高肿胀低等优点，是核反应堆中包壳及其它结构的极有竞争力的候选材料。

Because of their high strengthes and low swelling properties , oxide dispersion strengthened ( ODS ) ferritic steels are competitive structural materials in reactors .

用MCNP程序计算核燃料废包壳缓发裂变中子形成的热中子通量密度

Computing with MCNP the thermal flux density due to the delayed fission neutron generated in processing the clads of burned-up fuel roads

M5合金是一种适用于高燃耗组件的新型锆合金，主要用作燃料棒包壳、端塞、导向管和定位格架材料。

M5 alloy was a new zirconium alloy used in high burn up assembly such as fuel cladding , end plug , guide tube and grid spring and so on .

国内池式轻水研究反应堆中，尽管其燃料元件包壳材料、反应堆水池覆面材料和主要堆内构件材料以及运行温度不一样，其冷却剂pH值都定为偏酸性。

Although the materials of nuclear fuel cladding , internal surface and the main reactor internals , and operating temperatures of the pool type light water reactors in China are different from each other . All of pH value of water in the pools are always selected as slight acidity .

本文提出利用热处理工艺改善再结晶退火态M5锆合金的组织和性能，旨在提高核燃料包壳的寿命，为促进核燃料包壳国产化提供一定的理论参考和依据。

In order to extend the life of nuclear fuel cladding tubes , this paper proposes to improve microstructure and properties by heat treatment of M5 zirconium alloys on recrystallization annealing state .

结果表明，U3Si2-Al弥散型复合燃料板的破坏模式主要表现为芯体裂纹的萌生、扩展和失效，芯体与包壳冶金结合性能较好，拉伸及弯曲过程中未出现明显的剪切破坏。

The results indicate that the failure mechanisms of U_3Si_2-Al fuel plates during tension and flexure are crack initiation , propagation , and failure of the fuel layer , without shearing failure between the fuel-cladding sintering interfaces .

60Coγ刀治疗源采用高比活度放射性钴粒，分装焊封在双层不锈钢包壳内，经安全性能、表面污染与泄漏等质量检验，符合GB4075和GB4076的规定。

Co γ source consists of 60 high specific activity cobalt pellets and double stainless steel sealed by argon arc welding . Its safety performance , surface contamination and leakage testing meet the requirements of GB 4075 and GB 4076 .

锆合金作为轻水核反应堆燃料元件的包壳材料已得到了广泛的应用并积累了丰富的经验，因而仍被列为SCWR潜在的应用材料。

As a light water nuclear reactor fuel cladding material , Zr alloy has been wide used and its application experience accumulated , hence it is the potential materials for SCWR .

近年来的研究工作表明：低活性Fe-Cr-Mn系合金最有希望替代高洁性Fe-Cr-Ni系合金，使用在核反应堆中，用作第一壁结构材料和包壳材料的。

Lately studies showed that the low activation Fe-Cr-Mn type alloys were very likely to substitute the high activation Fe-Cr-Ni type alloys and can be used as first wall and cladding materials in fusion reactors .

PMC核燃料装卸贮存系统直接吊装核燃料组件，即直接面对核电站三道屏障中的第一道屏障&燃料包壳，其安全性是核电站核安全的重要保障。

Nuclear fuel handling and storage system ( PMC ) is a system which directly handles nuclear fuel assemblies , namely , it directly faces to nuclear fuel cladding which is the first of the three barriers of nuclear safety for nuclear power station .

研究了Al-Li合金为芯体、Al-Mg-Cu-Si合金为包壳的扩散偶，经共挤压及520℃/2h固溶处理后，芯体与包壳间的互扩散现象。

Tth interdiffusion between core and cladding was investigated . Core material is Al-Li alloy and cladding material is Al-Mg-Cu-Si alloy . The diffusion couple have coextruded and solution-heat-treated ( 520C , 2h ) .

在堆用不锈钢包壳管内分别填装不同B/C比的B4C芯块及核级钠，以模拟快堆控制棒内的B4C/Na/S.S。

B 4C pellets with different B / C ratio and nuclear purity grade sodium were put into a stainless steel cladding tubes and the out of pile tests were carried out at 550,650 and 750 ℃ .

基于LOFTL2-5试验的一维与三维计算结果比较，初步分析了有关三维效应及其对ECC旁通等应急堆芯冷却过程以及燃料包壳峰值温度（PCT）的影响。

By comparing the 1-D and 3-D calculation results of LOFT L2-5 test progress , related 3-D effects and its influence on such emergency core cooling progress as ECC bypass etc and that on the fuel peaking cladding temperature ( PCT ) is preliminarily investigated .

其中M4和M5合金包壳在燃料棒燃耗达到55GW·d·t-1的辐照考验结果表明，它们在堆内的腐蚀、蠕变和辐照伸长等性能优于改进型Zr4合金包壳。

The results obtained from irradiation test show that the corrosion resistance , creep resistance and irradiation growth for M4 and M5 are superior to the optimized Zircaloy 4 , especially the performances of M5 in and out of pile are much better .

通过堆外模拟试验，研究了国产316－Ti不锈钢包壳管在4种氧势下对FCCI（燃料包壳化学相互作用）和FPLME（裂变产物液态金属脆化效应）的敏感性。

The susceptibility of domestic 316-Ti stainless steel cladding tube with respect to Fuel Cladding Chemical interaction ( FCCI ) and Fission Products - induced Liquid Metal Embrittlement ( FPLME ) under four oxygen POtentials has been investigated by out-of-pile simulation test .

采用对核颗粒进行还原掺杂和将其包壳的方法制备了一系列不同DMAB用量、内部有还原敏化中心（Ag2）的AgBr核壳乳剂。

A series of cubic AgBr core-shell emulsions were prepared from the core emulsions reduced by DMAB and shelled by AgBr . These core-shell emulsions doped with Ag_2 in the interior of grains exhibited a significant increase in sensitivity .

特大型钢包烘烤过程包壳表面温度场研究

Thermal Field on Shell Surface of Large Ladle During Baking

用矢量电位有限元法改进新型感应钢包炉包壳结构的设计

Structure Improvement for Induction Ladle Furnace Shell with Vector Electric Potential FEM

对堆芯熔融过程中包壳和燃料栅元的径向和轴向分段失效模式进行了模拟；

Radial and axial failure mode of fuel and cladding are simulated ;

一种用于退役核燃料元件包壳的破裂检测技术

A safety check-up method for decommissioned reactor fuel element cladding

锆-4合金包壳管抗疖状腐蚀性能研究

Research on Nodular Corrosion Resistance Performance of Zircaloy-4 Cladding Tubes