俄罗斯无机材料研究院关于快堆燃料包壳材料的研究

俄罗斯无机材料研究院(ВНИИНМ)是材料学研究和核燃料循环工艺、裂变核材料处理工艺等领域的著名研究机构,在快堆堆芯结构材料方面该院借助于俄罗斯丰富的钠冷快堆运行和材料学研究经验,以BOR-60和BN-600为研究试验平台,以提高BN-600和BN-800性能及开发更加先进的BN-1200为目标,开展了大量燃料棒包壳及燃料组件外套管材料的研究.本文是对ВНИИНМ近几年研究成果在俄罗斯科学杂志和研讨会上发表报告的调研、翻译和汇总,供我国有关钠冷快堆技术研究和工程设计人员参考.     

F/M钢在超临界水环境中的腐蚀性能

本文研究了F/M钢在超临界水（SCW）环境中的腐蚀性能。实验结果表明,F/M钢在SCW中的抗腐蚀性能较差,温度、溶氧浓度以及材料中的Cr含量对其腐蚀性能有较大影响。对12Cr表面进行盐浴复合处理（QPQ）、电镀Cr和磁控溅射Cr处理,以研究其对F/M钢在SCW中抗腐蚀性能的影响。研究表明,经电镀Cr和磁控溅射Cr处理的12Cr试样在SCW中具有优良的抗腐蚀性能,尤其是经磁控溅射Cr处理的试样,1 000 h后其表面氧化膜依然完整致密,而经QPQ的试样腐蚀严重。     

P92钢高温低周疲劳的实验研究

由于高的热效率和简单的系统组成,超临界水堆(SCWR)被认为是第四代核反应堆的一种选择。超临界水堆的关键问题之一是核心部件尤其是燃料组件包壳的材料。这些材料在高温下的力学性能、腐蚀和应力腐蚀开裂敏感性以及抗辐射性能等对核电厂的安全运行至关重要。本文对SCWR包壳候选材料的F/M类材料P92钢进行了高温低周疲劳实验研究。实验温度为600和650℃,控制方式为总应变控制,应变范围均为±0.2%～±0.6%。实验结果表明,在两种温度下,P92钢均为循环软化材料,但未出现循环稳定现象。由于温度升高,塑性增强,P92钢在650℃下的宏观裂纹出现周次比率随应变范围的增加,下降比较平缓,且650℃下的失效寿命显著高于600℃下的失效寿命。并得到了两种温度下的稳定循环应力-塑性应变的关系以及循环失效寿命和应变的关系。     

Effects of Fast Reactor Irradiation Conditions on Tensile and Transient Burst Properties of Ferritic/Martensitic Steel Claddings

The effects of fast neutron irradiation conditions have been investigated by focusing on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated at temperatures from 693 to 833K to 42.5 dpa (displacement per atom) in the experimental fast reactor JOYO using the PFB090 fuel test subassembly. Post-irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for the PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. On the other hand, the strength of the HT9M cladding tended to shift to lower values than those of as-received specimens. The differences in tensile and transient burst strengths between the two claddings were attributed to martensite structural stability which was related to the stable solid solution elements.     

超临界水冷堆燃料包壳管用低活性F/M钢的优化设计

应用热力学计算与实验验证,系统研究了Cr、W、C、Mn对高Cr低活性F/M（铁素体/马氏体）钢基体相及显微组织的影响规律,并在此基础上,对钢的组织和成分进行设计与优化,以适应超临界水系统对包壳材料的性能要求。研究表明：Cr是决定高Cr低活性实验钢中奥氏体Cr固溶量以及钢中是否出现铁素体的最重要影响因素;W和C对实验钢铁素体相的出现有显著影响,而Mn的影响相对较小;W对实验钢中Laves相出现的温度范围及数量具有显著影响,Laves相消失的临界温度随W量降低而降低;在不采用Co、Ni等奥氏体形成元素且不增加Mn量的情况下,通过调控W、C等含量,Cr含量≥11%的Cr-W-C-Mn系低活性F/M钢即可获得全马氏体组织。     

Dependence of Fast Reactor Fuel Burnup Characteristics on Nuclear Data Libraries

In the framework of the development of burnup calculation method for commercial fast reactors, a sensitivity analysis was carried out to clarify the dependence of fuel burnup characteristics on nuclear data libraries (NDLs). The following NDLs were compared: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2 and JENDL-3.3. The NDL-dependence of material balance for main heavy metal nuclides (²³⁵U, ²³⁸U, ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu) was small, since the number densities at the end of one-cycle burnup did not change over 1 or 2% among the above-mentioned libraries. Relatively large differences were found for minor actinide nuclides, especially for ²³⁶U, ²³⁷Np, ^242mAm, ²⁴³Am and curium isotopes. The number densities for these nuclides after burning up showed remarkable NDL-dependence over 5% through 50%. A burnup sensitivity analysis system based on the generalized perturbation theory enabled us to find out quantitatively the causative nuclides and reactions, as well as their energy regions.     

国产快堆燃料元件包壳材料316不锈钢的中子辐照效应

介绍国产六种不同成分与工艺的快堆燃料元件包壳材料316不锈钢(316SS)经650℃高温、积分中子注量3.1×1021n/cm2(En>0.1MeV)的辐照概况,以及辐照后在650℃与室温下的拉伸力学性能试验和金相检查的结果及评述。     

Development of an ISI Robot for the Fast Breeder Reactor MONJU Primary Heat Transfer System Piping

This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55°C; piping surface, 80°C) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB).     

Irradiation Performance of Fast Reactor MOX Fuel Assemblies Irradiated to High Burnups

Mixed oxide fuel assemblies (MFA-1 and MFA-2 assemblies) were irradiated in the fast flux test facility to evaluate the irradiation performance of fast reactor core fuels at high burnups and high fast neutron fluences. The MFA-1 and MFA-2 assemblies achieved respective peak pellet burnups of 147 and 162GWd/t, and resisted to respective peak fast neutron fluences (E > 0:1 MeV) of 21:4 _ 1026 and 23:8 _ 1026 n/m2, without any indication of fuel pin breaching. Structural components of these assemblies were made of modified type 316 stainless steel and 15Cr-20Ni base advanced austenitic stainless steel. Postirradiation examinations of these assemblies revealed dimensional changes of fuel pins and assembly ducts due to irradiation-induced void swelling and irradiation creep, and fuel cladding local oval distortions due to bundle-duct interaction (BDI). The swelling resistance of 15Cr-20Ni base advanced austenitic stainless steel fuel pin cladding was almost the same as that of the modified type 316 stainless steel cladding, while the assembly duct of the former material had a slightly higher swelling resistance than that of the latter material. Analyses of fuel pin bundle deformations indicated that these assemblies likely mitigate BDI mainly by fuel pin bowings and cladding oval distortions.     

Measurements of Reaction Rates in Zone-Type Cores of Fast Critical Assembly Simulating High Conversion Light Water Reactor

Measurements of reaction rates have been performed in three uranium-fueled zone-type cores of the FCA constructed for a series of experiments on a high conversion light water reactor (HCLWR). These cores possess central test zones of different fuel enrichments and moderator to fuel volume ratios. Radial and axial fission rates of 236U, 239Pu, 238U and 23,Np were measured in each test zone by means of the micro-fission counter traverse. A region where the fundamental mode spectrum is established in the test zone were determined by utilizing these fission rate distributions. Central reaction rate ratios relative to the 235U fission rate were obtained from the measurements by the micro-fission counters and metallic uranium foils to examine changes in the reaction rate ratios among the three cores.

The measured data were analyzed by the SRAC code system on the basis of the nuclear data file JENDL-2. The calculated fission rate distributions agree well with the experimental results for the all cases. The results of reaction rate ratios show that the calculations over- predict the experimental values of the 238U capture/235U fission and 238U fission/235U fission rate ratios in the three cores.