反应堆压力容器密封面材料非正常工况下的腐蚀性能研究

针对压力容器密封面材料在核工程应用中发生的腐蚀问题,研究了反应堆压力容器密封面材料非正常工况下的腐蚀性能.利用静态高压釜研究308L不锈钢在不同Clˉ浓度条件下的腐蚀行为,采用金相显微镜和扫描电镜( SEM)对样品进行观察和分析.结果表明,在270℃、5.5 MPa条件下,Clˉ浓度低于1 mg/L时308L不锈钢没有发生点腐蚀、缝隙腐蚀和应力腐蚀;随着Clˉ浓度提高,308L不锈钢对点腐蚀、缝隙腐蚀和应力腐蚀的敏感性显著增加.     

CORA程序介绍和在腐蚀产物控制措施上的应用

为了便于国内核电厂进行较为深入的腐蚀产物产生、分布和控制措施等研究,对腐蚀产物的产生和物理特性、CORA程序的原理及使用范围、国际上采用CORA程序对核电厂腐蚀产物控制措施的情况、辐射场监测和集体剂量分析等方面的经验进行了研究,并采用CORA程序对AP1000核电厂主要的腐蚀产物控制措施进行了分析。通过研究得出如下的结论和建议:CORA程序适用于当核电厂设计条件发生变化时,对电厂中腐蚀产物的影响进行评估;国外广泛采用CORA程序应用于核电厂辐射场测量和腐蚀产物控制措施研究中;CORA程序计算可说明AP1000核电厂中采用的主要控制腐蚀产物的措施,如控制蒸汽发生器传热管钴含量、提高向反应堆冷却剂中注入的Li OH浓度等措施对于控制腐蚀产物是有效的;建议国内核电厂广泛应用CORA程序以研究电厂中辐射场的变化及腐蚀产物的控制措施。     

钚在低碳钢包装容器腐蚀产物上的吸附行为研究进展

研究低碳钢材质容器腐蚀机理及其腐蚀产物对钚的吸附行为对评价放射性废物处置场地安全性具有重要意义。通过研究低碳钢包装容器在放射性废物处置库中的腐蚀行为以及腐蚀产物对钚的吸附行为研究，可为准确预测钚的迁移行为和处置库安全评价提供理论基础。本文综述了废物地质处置条件下包装容器材料可能的腐蚀机理，系统总结了腐蚀产物对钚的吸附行为、影响吸附的主要因素以及吸附机理等方面的国内外研究进展，为今后深入研究腐蚀产物对钚的吸附机理提供参考。     

钚在低碳钢包装容器腐蚀产物上的吸附行为研究进展

研究低碳钢材质容器腐蚀机理及其腐蚀产物对钚的吸附行为对评价放射性废物处置场地安全性具有重要意义。通过研究低碳钢包装容器在放射性废物处置库中的腐蚀行为以及腐蚀产物对钚的吸附行为研究，可为准确预测钚的迁移行为和处置库安全评价提供理论基础。本文综述了废物地质处置条件下包装容器材料可能的腐蚀机理，系统总结了腐蚀产物对钚的吸附行为、影响吸附的主要因素以及吸附机理等方面的国内外研究进展，为今后深入研究腐蚀产物对钚的吸附机理提供参考。     

钛合金在硼、锂介质中的缝隙腐蚀行为分析

钛合金作为新型蒸汽发生器的主要结构材料，其耐缝隙腐蚀性能受到关注，而钛合金在硼、锂介质中的缝隙腐蚀行为研究较少。本文采用微型腐蚀回路对钛合金TA16和TA17在硼、锂介质中的5000 h缝隙腐蚀行为进行了研究，获得了2种材料的缝隙腐蚀敏感性，并对试验后钛合金氧化膜成分和结构进行了分析。结果表明：在缝隙腐蚀模拟件上未观察到缝隙腐蚀现象，TA16、TA17在硼、锂介质中对缝隙腐蚀不敏感；模拟件缝隙内、外的氧化物存在一定差异，缝隙外的颗粒状微晶钛铁矿(FeTiO₃)与钛合金缝隙腐蚀无关。     

压水堆一回路颗粒性腐蚀产物沉积机理的实验研究

一回路腐蚀产物在传热管内表面的沉积对压水堆核电厂安全运行和放射性产物屏蔽有非常不利的影响。为研究一回路腐蚀产物在传热管中的沉积机理,搭建了一回路腐蚀产物沉积机理研究实验系统。实验研究了颗粒性腐蚀产物Fe_3O_4在传热管内的沉积分布,并对比分析了不同弯管半径、pH值和粒径对颗粒沉积的影响。实验得到了颗粒状腐蚀产物在管内分布的规律,给出了最大沉积量所对应的pH值和粒径范围,可为压水堆化学控制提供参考。     

钛合金在高温高压过热蒸汽中的腐蚀机理研究

研究了T225NG和T42NG合金在高温、高压、过热蒸汽中长时间的氧化腐蚀规律。对试样表面氧化膜元素、相组成及表面形貌进行了测试与分析，给出了两种合金的腐蚀动力学模型．并对两种合金的腐蚀机理进行了初步分析。结果表明，两种钛合金中主元素Ti优先产生氧化腐蚀，合金元素Al和Fe及N、H、C等杂质元素对腐蚀起到促进作用，并在合金表面呈现出典型的结晶学特征。两种钛合金的腐蚀动力学规律均服从Wagner理论中的近抛物线规律。     

基于分形理论的核动力管道腐蚀坑深度预测模型研究

针对核动力管道腐蚀环境的复杂性和腐蚀过程的随机性，利用分形理论，建立了腐蚀坑深度预测模型，通过模型对核动力管道的最大腐蚀坑深度和不大于允许腐蚀深度rd的概率进行了预测研究。为进一步提高预测精度，对非线性方程进行了线性化处理。计算结果证明，利用分形理论对腐蚀坑深度进行预测是可行的，同时通过对核动力管道的最大腐蚀坑深度的预测，对分析腐蚀管道的可靠性和安全性及了解核动力管道的性能具有一定的意义。     

纯水中微量铜、氯离子对铝的腐蚀的影响

本文重点研究了纯水中微量铜、氯离子对铝的点腐蚀的影响,以及γ辐射对点腐蚀的作用. 试验表明,铜、氯离子浓度在<0.05毫克/升时,АО铝、АД1铝和铝合金可完全避免点腐蚀的发生.对于АД1铝和铝合金,单一氯离子浓度在0.1—0.2毫克/升时,亦不会产生点腐蚀. 在低浓度的铜、氯离子介质中(浓度<1毫克/升),γ辐射可以促使点腐蚀转化为均匀腐蚀,但其腐蚀量却大大增加. 试验还证明,铝经过阳极氧化后,可以有效地防止点腐蚀发生,而且均匀腐蚀量也大为减少. 本文还叙述了铁、铅、镍和硝酸根等离子对铝的点腐蚀影响,以及重铬酸钾在铜、氯离子溶液中对铅的缓蚀作用.     

熔盐堆中结构材料的腐蚀研究

本文针对第四代先进核能系统熔盐堆用结构材料的腐蚀问题背景,从材料的发展,氟盐腐蚀的特点、驱动力和控制机制等几方面综述了合金结构材料在高温氟化物熔盐中的腐蚀研究进展。镍基合金在氟盐中的主要腐蚀机制是合金中的活性元素Cr的选择性溶解。根据驱动力,可以将腐蚀分为四种类型:本征腐蚀,杂质腐蚀,温差腐蚀和异质材料腐蚀。结合钍基熔盐核能系统的运行环境和特点,对钍基熔盐核能系统特有的七大腐蚀问题进行了论述,介绍了我国在熔盐堆结构材料腐蚀研究领域取得的进展。     

Effect of gamma-ray irradiation on the deoxygenation of salt-containing water using hydrazine

In spent fuel pools at the Fukushima Daiichi nuclear power plant, hydrazine was added to salt-containing water in order to reduce dissolved oxygen. Hydrazine is known to reduce dissolved oxygen in high-temperature pure water, but its deoxygenation behavior in salt-containing water at ambient temperature in the presence of radiation is unknown. Deoxygenation using hydrazine in salt-containing water was thus investigated using a ⁶⁰Co gamma-ray source and artificial seawater at room temperature. Water samples containing a small amount of hydrazine were irradiated at dose rates of 100–10,000 Gy/h. The concentration of dissolved oxygen in the water samples was measured before and after irradiation. Notably, a decrease in the dissolved oxygen was only observed after irradiation, and the dissolved oxygen concentration decreased with increasing dose rate and irradiation time. The rate of decrease in the amount of dissolved oxygen using hydrazine was slow in the presence of salts. Kinetic considerations suggested that the deoxygenation of the salt-containing water exposed to gamma-ray irradiation using hydrazine was suppressed by chloride ions.     

冷变形316不锈钢在高温水中的应力腐蚀开裂行为

对不同冷变形量的核级316和316L不锈钢在高温水中的应力腐蚀开裂(SCC)行为进行了研究。通过试验,对溶解氧、氯离子和温度对裂纹扩展速率的影响进行了深入探讨和分析。试验结果显示,溶解氧和氯离子能明显加快材料的应力腐蚀开裂速率。当水化学条件一致时,325℃时的裂纹扩展速率较288℃时的裂纹扩展速率高。     

Evaluation of the stability of precipitated uranyl peroxide and its storage characteristics in solution

This work studied the stability of uranyl peroxide, which can be obtained as the final product of several processes to treat uranium mixture waste and uranium ore, in solution using various temperature, pH, and ionic strength conditions. The change in concentration of dissolved uranium and hydrogen peroxide from uranyl peroxide, the form of the dissolved uranium species, and the change in morphology of dissolved uranyl peroxide were investigated for 100 or more days. Uranyl peroxide was stable in distilled water at elevated temperatures, but dissolved in other solutions at temperatures higher than 40?°C; a greater amount of uranyl peroxide dissolved in more acidic conditions at elevated temperatures. Uranyl ions that dissolved from uranyl peroxide were able to be recovered as uranyl peroxide in the solution where the dissolution occurs by adding hydrogen peroxide. After the precipitation of uranyl peroxide, the uranyl concentration in the supernatant is low enough for the supernatant to be recycled or released into the environment. The morphologies of the partially dissolved uranyl peroxide and the re-precipitated uranyl peroxide from dissolved uranyl ions were different from that of the initial uranyl peroxide.     

铝包壳,镁-二氧化铀元件的溶解

一、前言一种以金属铝为包壳,由镁和二氧化铀粉末组成的弥散体铀芯的核燃料元件(其中Mg占15%,UO_2占85%),经过反应堆辐照、冷却后,需要溶解成溶液,以便回收铀和钚。     

Dissolution of spent nuclear fuel in carbonate–peroxide solution

This study shows that spent UO₂ fuel can be completely dissolved in a room temperature carbonate–peroxide solution apparently without attacking the metallic Mo–Tc–Ru–Rh–Pd fission product phase. In parallel tests, identical samples of spent nuclear fuel were dissolved in nitric acid and in an ammonium carbonate, hydrogen peroxide solution. The resulting solutions were analyzed for strontium-90, technetium-99, cesium-137, europium-154, plutonium, and americium-241. The results were identical for all analytes except technetium, where the carbonate–peroxide dissolution had only about 25% of the technetium that the nitric acid dissolution had.     

高温气冷堆二回路加药系统工艺优化及自动控制方法研究

目前高温气冷堆（HTGR）的二回路水质指标无相关的成熟标准可以参考，加药控制指标也没有标准。本文通过对HTGR二回路的材质和运行工况的研究，参考压水堆（PWR）和火电直流炉的运行经验，确定二回路给水pH值控制标准在9.5~9.8之间，联氨的控制标准在80~120 μg/L之间可以使二回路取得较好的防腐效果；针对HTGR二回路加药系统设计现状，对加药系统设计工艺提出了优化和变更方案，用联氨表代替溶氧表，用电导率计算pH，避免了溶氧表和pH表的滞后性和不稳定性；最后，通过控制方法的改进，实现HTGR二回路加氨和加联氨系统的全自动精准控制。      

高温水中氯离子对316L不锈钢应力腐蚀裂纹扩展速率的影响

本文采用直流电压降(DCPD)方法,使用恒K(K=27.5 MPa·m1/2)加载方式,在核电厂高温高压水环境中研究了氯离子对316L不锈钢的应力腐蚀裂纹扩展速率的影响。实验结果表明:在高温除氧水中,氯离子会加快316L不锈钢的应力腐蚀裂纹扩展速率,且当水中存在溶解氧时,氯离子对应力腐蚀裂纹扩展速率的影响更明显。     

氯和氧对304N在高温水中应力腐蚀开裂的影响

采用慢应变速率拉伸试验（SSRT）与断口形貌分析技术研究304N不锈钢（固溶退火态）在300 ℃高温水中的应力腐蚀行为与机理。结果表明：304N在300 ℃高温水中的最大抗拉强度、延伸率以及断裂吸收能随Cl^-浓度的增大显著降低;随氧浓度的急剧降低而显著增大;304N在高温水中发生应力腐蚀开裂（SCC）主要为穿晶型;随Cl^-浓度增加,304N的应力腐蚀敏感性也迅速增加,在含50 mg/L Cl^-的空气饱和高温水环境中,试样断口形貌表现为完全脆断;在溶解氧浓度急剧降低时,氯致应力腐蚀开裂的敏感性大幅降低,表明溶解氧对304N在高温水中的氯致应力腐蚀开裂具有明显的促进作用。     

Investigation of the corrosion behavior of 304L and 316L stainless steels at high-temperature borated and lithiated water

The effects of temperature and dissolved oxygen on the electrochemical behavior and the oxide film formation of grades 304L and 316L stainless steels at high-temperature borated and lithiated water were investigated by means of potentiodynamic polarization, scanning electron microscopy and X-ray photoelectron spectroscopy. The results revealed that increasing the solution temperature degrades the passivity of the oxide films formed on both grades of steel and shifts their corrosion potential toward more negative potentials. The oxide films formed on the steel samples immersed into the solution containing 20 ppb dissolved oxygen (DO) showed a duplex structure, in which the inner layer was mostly a composition of Cr oxides and the outer layer mainly was a Fe oxide and Ni–Fe spinel. Only a single layer of Cr-rich oxide was observed in the oxide films formed in the solutions with the DO concentrations higher than 20 ppb. Higher amount of Cr in the oxide films formed on the type 316L compared to 304L improves the passivity of the oxide film of this grade of steel and results in a wider passive region in its potentiodynamic polarization curves.     

The Influence of Dissolved Hydrogen on the Corrosion of Type 304 Stainless Steels Treated with Inhibitive Chemicals in High Temperature Pure Water

In order to assess the influence of dissolved hydrogen on the intergranular stress corrosion cracking (IGSCC) characteristics of Type 304 stainless steels treated with inhibitive chemicals, electrochemical corrosion potential (ECP) measurements and slow strain rate tensile (SSRT) tests were conducted in high temperature pure water. A number of thermally sensitized specimens were prepared and then pre-oxidized in a 288°C pure water environment with the presence of 300ppb dissolved oxygen for 360h. Most of the specimens were then separately treated with various inhibitive chemicals including powdered zirconium oxide (ZrO₂), powdered titanium oxide (TiO₂), and zirconyl nitrate [ZrO(NO₃)₂] via hydrothermal deposition at 150°C. Test environments with a dissolved oxygen concentration of 300ppb and various dissolved hydrogen concentrations at 288°C were created. Test results showed that the ECPs of the treated specimens were lower than that of the untreated one no matter what the dissolved hydrogen concentration was. In addition, IGSCC was observed on all specimens (treated or untreated) in all tested environments. However, the untreated specimen exhibited lower elongation, shorter failure time, and more secondary cracks on the lateral surfaces. It was therefore suggested that inhibitive chemicals such as ZrO₂, TiO₂, and ZrO(NO₂)₂ did provide a certain degree of enhancement in improving the mechanical behavior of the treated specimens and in prolonging IGSCC initiation times.