新锆合金碘致应力腐蚀断口研究

利用扫描电镜和EDXA对新锆合金碘致应力腐蚀的断口进行了形貌观察和成分分析。在断口表面发现腐蚀产物、二次裂纹、沿晶开裂和穿晶准解理开裂等应力腐蚀断裂特征,并观察到锆合金碘致应力腐蚀的“沟槽”特征。起裂区为沿晶开裂,在裂纹扩展阶段,开裂以穿晶为主。断口上腐蚀产物的成分主要是氧和锆,局部准解理开裂区域可检测到碘。去应力退火试样上发现了平行轧面的深沟,且沿晶开裂不明显。试验温度升高,断口上的腐蚀产物增多,而且沿晶开裂更容易。碘分压不仅影响腐蚀产物层的厚度,而且碘分压较高时沿晶开裂容易发生。     

温度对Zr-Sn-Nb合金致应力碘腐蚀开裂的影响

研究了N18和N36锆合金在不同温度下的碘致应力腐蚀开裂(SCC)行为,用扫描电镜进行了断口分析.结果表明,对于再结晶状态的锆合金,随着试验温度升高,KISCC(临界应力强度因子)降低,裂纹萌生所需的应力降低,裂纹萌生所需的时间也变短;对于去应力状态的N18合金,试验温度从300℃增加到350℃,KISCC基本不变,裂纹萌生所需的应力降低,裂纹萌生所需的时间也变短;随着试验温度升高,断口上的腐蚀产物增多.     

锆合金的碘致应力腐蚀开裂研究(英文)

研究了不同热处理状态的Zr-2和Zr-4合金在不同浓度的碘介质及实验温度下的应力腐蚀开裂(SCC)行为。并对不同织构取向的试样在350℃下进行了蠕变实验,蠕变实验的载荷值选择与SCC实验相对应的一系列典型载荷。用扫描电子显微镜观察了断口特征,用透射电子显微镜和光学显微镜检查了材料的显微组织,用X-光衍射仪测定了锆合金的织构,分析讨论了材料状态、实验温度、碘浓度以及蠕变对锆合金碘致应力腐蚀行为的影响。     

新锆合金碘致应力腐蚀性能研究

研究了再结晶状态的相同成分，不同织构取向的两块新锆合金板材在不同的实验条件下的抗碘致应力腐蚀性能。用直流电位降法动态监测裂纹的长度并对试样进行了织构的测定、第二相的观察和断口分析结果表明，在实验条件下，N18-2(L-T)抗碘致应力腐蚀的能力优于N18-1(T—L)断口形貌表明，在应力腐蚀裂纹的初始扩展阶段，断口沿晶开裂；在裂纹的稳态扩展阶段，以穿晶准解理扩展为主。     

材料状态对锆合金碘致应力腐蚀的影响

在不同碘浓度(碘加热温度分别为40℃,65℃,110℃)和流动高纯氩的气氛下,经再结晶退火和消除应力退火的非辐照锆-2和锆-4合金,分别在300℃,350℃和400℃的温度下,进行了单轴拉伸试验,以便了解材料状态、微观组织对锆合金碘致应力腐蚀行为的影响。试验结果表明,在应力腐蚀过程中的初始阶段,由于晶间腐蚀行为与材料微观组织有关,晶粒度对材料应力腐蚀行为(对不同碘浓度)均很敏感;在相同试验温度下,     

锆合金耐腐蚀性能研究综述

锆合金主要用作核反应堆燃料元件的包壳材料及其他堆内构件。回顾了有关锆合金水侧腐蚀的主要研究结果及存在的问题，概括了现有的理论及面临的挑战。８０年代，关于锆合金化学成分、微观结构及辐照对耐腐蚀性能影响的研究取得了很大进展。近几年来的研究工作主要集中在探索腐蚀机理、选择最佳合金成分及控制微观结构方面，以满足提高燃耗、降低核电成本后对锆合金提出的更高要求。     

两种锆合金显微组织与腐蚀机理研究

根据国产C锆合金与低锡Zr-4合金在纯水以及LiOH水溶液中的高压釜腐蚀试验的结果,采用透射电镜（TEM）观察基体和氧化膜显微组织,通过分析氧化增重数据,对C锆合金的腐蚀机理进行了研究。提出了3种腐蚀机理:即Nb元素有效抑制阴离子空位浓度提高,可减少氧元素扩散速率;缺陷阱的数量影响氧扩散带来的腐蚀,且空位阱数量与第二相颗粒总表面积成正比;第二相粒子氧化膨胀造成氧化膜压应力松弛,降低其稳定性并失去保护能力。      

材料状态对锆合金碘致应力腐蚀的影响(英文)

在不同碘浓度(碘加热温度分别为40℃,65℃,110℃)和流动高纯氩的气氛下,经再结晶退火和消除应力退火的非辐照锆-2和锆-4合金,分别在300℃,350℃和400℃的温度下,进行了单轴拉伸试验,以便了解材料状态、微观组织对锆合金碘致应力腐蚀行为的影响。试验结果表明,在应力腐蚀过程中的初始阶段,由于晶间腐蚀行为与材料微观组织有关,晶粒度对材料应力腐蚀行为(对不同碘浓度)均很敏感;在相同试验温度下,碘浓度增加使裂纹生长加快;随着试验温度提高,裂纹生长速率增加;由于织构影响,再结晶退火材料比消除应力退火的材料好。     

锆合金包壳I-SCC性能评价

对N36、Zr-4、X锆合金包壳管环形试样在350、400℃下施加周向拉伸载荷,研究N36锆合金包壳管在10~2 Pa、10~3 Pa、10~4 Pa碘分压、Zr-4及X试样在102Pa碘分压下的碘致应力腐蚀开裂行为。研究发现:在350、400℃下以最大载荷为指标时,N36、Zr-4及X试样在一定碘分压环境中均会发生不同程度的碘致应力腐蚀开裂,断裂能量迅速下降;在相同试验条件下,N36试样的最大载荷和断裂能量下降最慢。     

水化学对锆合金耐腐蚀性能影响的研究

经对包括Ｚｒ－４在内的４种不同成分锆合金耐腐蚀性能的研究发现,高湿水中添加了０．０１￣０．１ｍｏｌＬｉＯＨ后,腐蚀所折提早发生,转折后的腐蚀速度增加,这种现象随ＬｉＯＨ浓度增大变得更加显著,但在成分不同的锆合金中有明显的差异,同时添加Ｈ３ＢＯ３后又可以抑制ＬｉＯＨ的加速腐蚀作用。氧化膜显微级织的观察结果表明：转折后的加速腐蚀过程与氧化膜中孔洞簇的出现有关,添加ＬｉＯＨ以及改变合金成分后,通过影响孔     

Stress corrosion cracking of Zircaloy-4 cladding at elevated temperatures and its relevance to transient LWR fuel rod behaviour

In extensive out-of-pile experiments from 500 to 900° C it has been shown that, of all the volatile fission products in a LWR fuel rod, only iodine can cause low ductility failure of Zircaloy-4 tubing due to stress corrosion cracking up to about 800° C. The critical iodine concentration above which brittle cladding failure occurs was determined as a function of temperature in the absence and presence of UO₂ fuel. A comparison of these values with the amount expected in the fuel cladding gap during a LOCA transient shows that a clear influence of iodine on burst strain can be expected only up to 700° C. This is in agreement with the results of in-pile LOCA tests performed in the FR-2 reactor with high burnup fuel rods. Since the burst temperatures during a LOCA transient would generally be above 700° C, an influence of iodine on burst strain is not very probable in a LOCA. However, with respect to ATWS transients where the maximum cladding temperatures would be below 700° C, an influence of iodine on the mechanical properties of zircaloy can be expected.     

Alloying effects on the corrosion behavior of binary Nb-based and Mo-based alloys in liquid Li

Alloying effects on the corrosion in liquid Li at 1473 K have been investigated for both binary Nb-based and Mo-based alloys. For the binary Nb-based alloys, the weight change due to the corrosion varied largely with alloying elements. Also, there was a clear difference in the surface morphology between them after corrosion tests. Either one or two corrosion products and large cracks were observed on the surface of all the Nb-based alloys. The weight change of a Nb–Hf alloy was the smallest among a variety of Nb-based alloys, indicating that the Hf addition was very effective in improving the corrosion resistance. On the other hand, for the binary Mo-based alloys the weight change was about a factor of 10 smaller than that for the binary Nb-based alloys. There were no cracks and a little corrosion products on the surface of them, indicating that the Mo-based alloys have much superior corrosion resistance in liquid Li than the Nb-based alloys.     

Influence of Nb concentration in the α-matrix on the corrosion behavior of Zr-xNb binary alloys

The influence of the Nb concentration in the α-matrix on the corrosion behavior of Zr-xNb (x=0-0.6 wt%) binary alloys was evaluated using a static autoclave in the temperature range from 300 to 500 °C. Corrosion tests and precipitate analysis of Zr-xNb binary alloys showed that corrosion resistance increased with the increase of the Nb concentration in the α-matrix, and the best corrosion resistance was obtained when the Nb concentration was nearly at its equilibrium solubility limit at all test temperatures. The alloys containing a higher Nb concentration than their equilibrium solubility also showed good corrosion resistance, which could be attributed mainly to the formation of Nb-precipitates, resulting in an equilibrium Nb concentration in the α-matrix. These results imply that the corrosion resistance of Nb-containing Zr-alloys can be controlled by the Nb concentration in the α-matrix rather than the Nb-precipitates.     

Nb元素和Fe元素对锆合金耐腐蚀性能的影响

将几种成分不同的锆合金样品放入高压釜中,在350℃、16.8MPa、70μg/gLi+LiOH水溶液中腐蚀。结果显示:第二相几乎全是Zr-Nb-Fe粒子的3#样品耐腐蚀性能最好,而不含Zr-Nb-Fe粒子的1#和5#样品耐腐蚀性能很差,这说明Zr-Nb-Fe第二相粒子对改善锆合金耐腐蚀性能起着关键作用。只有合金中Nb元素和Fe元素配比合理,才可使合金中第二相主要是Zr-Nb-Fe粒子。     

A review of radionuclide behaviour in the primary system of a very-high-temperature reactor

During normal operation of (V)HTRs radiologically-significant contamination of the primary system will occur this being of prime importance in relation to depressurization accidents. This paper reviews information relevant to radiocontaminant transport in (V)HTR primary systems paying particular attention to chemical forms, interactions with dust and overall distribution. The primary-system environment comprises nuclear graphites, alloys, dust and high-purity helium into which low releases of the isotopes ¹³⁴Cs, ¹³⁷Cs, ⁹⁰Sr, ^110mAg, ¹³¹I, ¹³⁵Xe and ⁸⁵Kr can be anticipated. Additionally, proper treatment of radiological risk requires accounting for tritium.A likely gas-phase speciation of the chemically-active fission products is proposed:

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for caesium and strontium, hydroxides would be dominant with iodides as minor species if a relatively low iodine concentration can be assumed;

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for iodine, a split between CsI and HI are likely to dominate with atomic iodine as a minor species.

Strong sorption of radionuclides onto carbonaceous dust can be expected. Such dust is likely to cover all surfaces in a pebble-bed (V)HTR so radionuclides will principally associate with this dust rather than underlying alloys. This is unlikely in prismatic (V)HTRs with lower and uneven dust deposits. Where caesium interacts with alloys strong implanting of a large fraction can occur via adsorption and reaction with low-concentration silicon. Silver shows no special affinity for carbonaceous dust but may interact preferentially with nickel-rich alloys, i.e., in the IHX and/or the gas turbine. Quantitative evaluations of radionuclide distribution are hampered by a lack of data regarding sorption onto the graphites, alloys and carbonaceous dust of modern (V)HTR systems; a long time will elapse before sufficient data are forthcoming. In the meantime, some form of best-estimate distribution and upper-bound concentration for contamination is needed if deterministic safety evaluations are to begin. This distribution will be different for pebble-bed and prismatic designs.     

Failure Propensities of Pressurized Zircaloy Tube Containing Iodine and Other Chemical Species

Abstract

Failure propensities of Zircaloy-4 cladding tube internally pressurized with Ar gas containing iodine and iodine plus each of other chemical species were examined at 360°C, to study the effect of corrosive fission products (FPs) on the integrity of spent nuclear fuel rods during dry storage, and also to assess the capability of preventing the spent fuel rod degradation.

The iodine stress corrosion cracking (SCC) of Zircaloy tube occurred in the long time/low stress exposure tests at stresses much lower than the conventional “threshold stress”, with considerably large strains at failure. The addition of cesium to iodine perfectly suppressed the SCC. It is inferred from these results that the degradation of spent fuel rods induced by corrosive fission products is unlikely during dry storage. Even if iodine alone should take effect, a proper strain limit could prevent spent fuel rods from incurring iodine induced effects because of considerably large strains necessary for iodine SCC of Zircaloy tube at low stresses.     

热处理工艺对冷旋U-6.5Nb合金电化学腐蚀行为的影响

采用冷旋方法制备了管状U-6.5Nb合金零件,分别在400、600和700℃下对合金零件进行1h退火处理,考察了不同状态的U-6.5Nb合金在含50μg/gCl-的氯化钾水溶液中的电化学腐蚀行为,采用扫描电镜表征了腐蚀前后的形貌特征。结果表明：所有状态的合金均未发生钝化,冷旋态和700℃退火态合金为单相组织,具有较高的腐蚀电位和较小的腐蚀电流,400℃退火和600℃退火态合金为双相组织,具有较低的腐蚀电位和较大的腐蚀电流。单相合金比双相合金具有更好的抗腐蚀能力,但更易发生点蚀。双相合金表面Nb成分不均是其抗腐蚀性不佳的主要原因。     

CAP1400燃料组件用新锆合金研究

在Zr-Sn-Nb系合金的基础上添加微量合金元素Ge和Si等,采用真空电弧熔炼,制备了多种新锆合金。使用透射电子显微镜(Transmission electron microscope,TEM)对合金基体进行显微组织分析,分别通过堆外高压釜腐蚀试验、定氢分析仪和万能材料试验机对合金的腐蚀、吸氢和拉伸性能进行评估。结果表明,常规工艺处理后,SZA-4和SZA-6合金均发生了完全再结晶,第二相细小、均匀弥散分布在晶粒内和晶界上;SZA-4和SZA-6合金在三种水化学条件下均具有优良的耐腐蚀性能,SZA-6合金的耐腐蚀性能优于参考合金,SZA-4合金的耐腐蚀性能略优于SZA-6合金;SZA-6合金的吸氢性能略优于SZA-4合金;两种合金的拉伸性能满足设计要求。基于SZA-4和SZA-6合金优良的耐腐蚀、吸氢和力学性能,未来将有望用于CAP1400自主化燃料组件。     

Tarnish Testing of Copper-Based Alloys Coated with SiO_2-Like Films by PECVD

The tarnishing test in the presence of hydrogen sulfide（H2S） vapors has been used to investigate the tarnish resistance capability of copper-based alloys coated with Si02-like films by means of plasma-enhanced chemical vapor deposition（PECVD） fed with a tetraethoxysilane/oxygen mixture.The chemical and morphological properties of the films have been characterized by using infrared absorption spectroscopy（IR） and scanning electron microscopy（SEM）with energy disperse spectroscopy（EDS）.The corrosion products of the samples after the tarnishing test have been identified by X-ray diffraction analysis（XRD）.It has been found that SiO2-like films formed via PECVD with a high O2 flow rate could protect copper-based alloys from H2S vapor tarnishing.The alloys coated at the O2 flow rate of 20 sccm remain uncorroded after 54days of H2S vapor tarnish testing.The corrosion products for the alloys deposited at a low O2flow rate after 54 days of tarnish testing are mainly composed of brochantite.