Magnetic field enhanced radio frequency ion source and its application for Siincorporation diamond-like carbon film preparation

Various ion sources are key components to prepare functional coatings,such as diamond-like carbon(DLC)films.In this article,we present our trying of surface modification on basis of Si-incorporation diamond-like carbon(Si-DLC)produced by a magnetic field enhanced radio frequency ion source,which is established to get high density plasma with the help of magnetic field.Under proper deposition process,a contact angle of 111°hydrophobic surface was achieved without any surface patterning,where nanostructure SiC grains appeared within the amorphous microstructure.The surface property was influenced by ion flow parameters as well as the resultant surface microstructure.The magnetic field enhanced radio frequency ion source developed in this paper was useful for protective film applications.     

国外最新消息

美国NRG能源公司将建造更多的核电厂美国NRG能源公司执行总裁David Crane表示,公司将拥有更多的核电厂,包括计划在得克萨斯州建造的两座机组。(编译自路透社2007年11月6日报道)     

《原子能科学技术》征稿简则

须包括题目、作者姓名、作者单位、城市名、省名和邮政编码,并应写成叙述性文摘(含有研究目的、方法、结果和结论);关键词3～8个。3)文稿应采用阿拉伯数字进行分级编号。引言不编号,也不写引言字样。4)基金项目名称及项目编号、作者简介(第1作者姓名(出生年—)、性别(民族,汉族略)、籍贯、职称、学位、从事     

国外最新消息

切尔诺贝利核电机组完成最后一次卸料工作2008年4月23日,切尔诺贝利核电厂完成了其3号机组的最后一次卸料工作。此次卸料工作的完成,意味着切尔诺贝利核电厂已完成了所有的机组卸料工作。(编译自世界核新闻网站2008年4月24日报道)     

国际原子能机构成员国

1957年阿富汗、阿尔巴尼亚、阿根廷、澳大利亚、奥地利、白俄罗斯、巴西、保加利亚、加拿大、古巴、丹麦、多米尼加共和国、埃及、萨尔瓦多、埃塞俄比亚、法国、德国、希腊、危地马拉、海地、罗马教廷、匈牙利、冰岛、印度、印度尼西亚、以色列、意大利、日本、大韩民国、摩纳哥、摩洛哥、缅甸、荷兰、新西兰、挪威、巴基斯坦、巴拉圭、秘鲁、波兰、葡萄牙、罗马尼亚、俄罗斯联邦、塞尔维亚和黑山、南非、西班牙、斯里兰卡、瑞典、瑞士、泰国、突尼斯、土耳其、乌克兰、英国、美国、委内瑞拉、越南1958年比利时、厄瓜多尔、芬兰、伊朗伊斯兰共和国、卢森堡、墨西哥、菲律宾、苏丹     

茜素红的辐射降解研究

用60Co γ射线辐照茜素红水溶液,研究该染料的辐射降解特性.通过对辐照前后茜素红的紫外可见光谱、脱色率、总有机碳(Total organic carbon,TOC)去除率的研究.探讨了吸收剂量、初始浓度、溶液pH值、H2O2加入量、在不同气体饱和条件下对茜素红溶液降解和矿化效果的影响.结果表明,辐射技术能有效降解茜素红染料,在空气饱和双氧水浓度为4mmol/L的条件下,吸收剂量为7.5kGy时,茜素红脱色率可达98%,TOC去除率可达70%.     

前言

由中国文物保护技术协会释光与电子自旋共振测定年代专业委员会(挂靠在上海博物馆)和中国国家地震局地质研究所共同主办的全国第十次释光(Luminescence)与电子自旋共振(ESR)测定年代学术讨论会于2006年11月6-10日在杭州的浙江省地震局培训中心召开.来自内地和香港的大学、中国科学院及部委研究所和文博系统32个单位的专家、学者出席了讨论会.这次讨论会共有27篇论文在会上作了交流,内容涉及热释光(TL)、光释光(OSL)和电子自旋共振测定年代的理论、方法和技术;考古和地质样品年代的测定及其应用;陶瓷器古剂量的测定;释光特性与矿藏成因;热释光、光释光测量仪器研制和改进等.本选编辑录了其中13篇论文予以发表.     

一株耐辐射菌的分离与初步鉴定

对新疆古尔班通古特沙漠土样中分离到的一株耐辐射菌RL2进行了多相分类鉴定.结果发现,此菌为革兰氏阳性,球形,菌落为淡红色;菌株的(G C)mol%含量为71.62%;16S rDNA序列分析表明,菌株RL2的16S rDNA基因序列与D.radiodurans DSM20539T同源性最高(97.2%).通过表型及系统进化树分析,确定RL2菌种分类应归于Deinococcus菌属,并可能是该菌属中的-个新种.     

Activated persulfate by DBD plasma and activated carbon for the degradation of acid orange Ⅱ

In this study, the effect of activated peroxydisulfate(PDS) by dielectric barrier discharge(DBD) plasma and activated carbon(HGAC) on the removal of acid orange Ⅱ(AOⅡ) was investigated. The effects of applied voltage, PDS dosage, HGAC dosage, initial pH value, and inorganic anions on the removal rate of AOⅡ were discussed. The main free radicals degrading azo dyes during the experiment were also studied. Experimental results show that the removal rate of AOⅡ in DBD/HGAC/PDS synergistic system is much higher than that in the single system. With the applied voltage of 16 kV, HGAC dosage of 1 g l-1, PDS and AOⅡ molar ratio of 200:1, initial pH value of 5.4 and concentration of AOⅡ solution of 20 mg l-1, the removal rate of AOⅡ reached 97.6% in DBD/HGAC/PDS process after 28 min of reaction.Acidic and neutral conditions are beneficial for AOⅡ removal. Sulfate and hydroxyl radicals play an important role in the removal of AOⅡ. Inorganic anions are not conducive to the removal of AOⅡ.     

二级PSA在严重事故管理中的应用研究

二级概率安全分析（PSA）可用来定量评估严重事故风险,是评价严重事故管理的良好工具。通过研究二级PSA应用于严重事故管理的一般方法与流程,以某二代改进型核电厂二级PSA模型为例,对严重事故管理导则中“一回路卸压”和“一回路应急注水”两个关键操作进行了定量评价。评价表明进入严重事故管理导则后立即执行“一回路卸压操作”可大幅度降低大量放射性释放风险,执行“一回路应急注水操作”对于降低进程较慢的事故序列大量放射性释放风险贡献较大。研究表明国内核电厂针对严重事故的管理还有进一步提升空间。      

核电站严重事故后果概率安全评价方法研究

核电站严重事故后果概率安全评价(PSA)是采用概率论的方法对核电站放射性后果进行分析,并定量给出放射性物质对核电站周围公众的健康效应影响。以国内某压水堆核电站为参考厂址,建立合适的场外后果分析模型。采用分层抽样方法对参考厂址1a的气象数据进行抽样,源项和释放特征等数据取自二级PSA的研究结果。利用事故后果评价程序对核电站严重事故后果进行计算,并用概率论方法对结果进行评估。通过计算将各事故和事故谱的场外个人剂量表示为CCDF曲线和总频率-剂量曲线,再用概率论方法得到不同距离处个人剂量超过指定剂量的条件概率;也可用此方法对确定烟羽应急计划区的安全准则中所描述的"大多数严重事故序列"进行量化。     

Effects of accident management strategy on the severe accident environmental conditions

This paper presents a methodology utilizing an accident management strategy in order to determine accident environmental conditions to be used as inputs to equipment survivability assessments. In the case that there is a well-established accident management strategy for a specific nuclear power plant (NPP), an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for the accident management strategy or appropriate actions. For this work, three different tools are introduced; probabilistic safety assessment (PSA) outcomes, major accident management strategy actions, and accident environmental stages (AESs). In order to quantitatively investigate an applicability of accident management strategy on equipment survivability, the accident simulation for most likely scenario in Korean standard nuclear power plants (KSNPs) is performed with the MAAP4 code. The accident management guideline (AMG) actions such as the reactor coolant system (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparison to actions from previous normal accident simulation, especially focusing on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages. This implies that plant-specific AMG actions need to be considered in order to determine accident environmental conditions in equipment survivability assessments.     

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

During a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.     

Nuclear Magnetic Resonance Study of Ligand Exchange Reaction in Lanthanum Nitrate Complex with n-Octyl(phenyl)-N,N-Diisohutylcarbamoylmethylphophine Oxide

The Molten Salt Reactor (MSR) concept has recently been considered as one of the candidates for the generation IV nuclear power systems. MSRs have many advantages such as improved safety, proliferation resistance, resource sustainability and waste reduction. But MSR developmental activities have lagged and there are few data available to support detailed analyses. However, the authors believe that additional investigations are merited for future study. From this point of view, the authors have analyzed the depressurization accident of the MSR “Fuji-12” using a newly developed MSR transient analysis code. In Fuji-12, a small amount of helium gas bubbles are circulated in the primary loop for stripping out gaseous fission products. A depressurization of the primary system would cause these bubbles to expand, resulting in a positive reactivity insertion. We have attempted to determine the severity of such an accident. Although the analysis is still preliminary and the assumptions are crude, it can be expected that the depressurization would not result in a severe accident in Fuji-12.     

二级PSA中人员可靠性分析方法研究

在概率安全分析(PSA)中,人员可靠性分析(HRA)是必不可少的组成部分。国内在一级PSA中的HRA做了大量的研究工作,已有良好的基础和工程实践,但由于核电厂严重事故下人员响应的复杂性,有关二级PSA的HRA还处于摸索阶段。通过研究二级PSA中人员响应特点,调研国内外在二级PSA中采用的HRA方法,最后以我国某三代压水堆核电厂严重事故下一回路快速卸压为例,采用THERP、HCR+THERP以及SPAR-H三种方法,分别进行了HRA,并给出相应的结论和建议。     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.     

漳州核电厂严重事故场外后果评价准则探讨

为了对“在技术上实现减轻放射性后果的场外防护行动是有限的甚至是可以取消的”这一基本目标进行量化评价,本文从简化事故后场外应急的角度,提出了严重事故后“3 km外不需要撤离、5 km外不需要隐蔽及服碘”的设计目标。结合漳州核电厂的厂址条件,推导出了一套用于漳州核电厂的严重事故后放射性后果评价准则。通过对“华龙一号”典型严重事故过程及放射性释放过程进行分析,结果表明,漳州核电厂“华龙一号”堆型满足本文提出的放射性后果评价准则,能够实现在严重事故后“3 km外不需要撤离、5 km外不需要隐蔽及服碘”的目标。     

Development of accumulator computational aid for determining RCS injection volume

Reactor coolant system (RCS) injection using accumulator is an important strategy for both emergency operating procedure (EOP) and severe accident management guideline (SAMG) of pressurized water reactor (PWR) nuclear power plant. Once accumulator injection starts, the operator is requested to close the accumulator isolation valve to avoid nitrogen gas flow into RCS as the water level is low. Current accumulator water level indication system is not designed for this purpose. In emergency operating procedure, it relies on the steam generator pressure to close the accumulator isolation valve.The purpose of this paper is to develop a computational aid for estimating RCS injection volume of accumulator. First of all, simple accumulator model is verified using the plant data during a station blackout incident of Maanshan nuclear power plant. An isentropic expansion model is found better than adiabatic expansion model. Then, a computational aid is developed based on this model. Using this computational aid, the accumulator water level can be judged directly from the accumulator pressure. This computational aid can be applied for typical PWR nuclear power plants in both emergency operating procedure and severe accident management guideline.     

池式钠冷快堆放射性释放风险概率安全评价事件树分析

池式钠冷快堆的安全特性和放射性释放机制与压水堆有着显著不同,在核安全新要求下,亟待开展放射性释放风险概率安全评价（PSA）研究。本文以池式钠冷快堆为研究对象,通过分析放射性来源、包容边界及破坏包容边界完整性的严重事故现象,确定了池式钠冷快堆大量放射性释放的主要位置和释放模式,构建分析了放射性释放事件树。本文分析结果可为进一步开展池式钠冷快堆放射性释放风险PSA提供参考。