核设施烟囱气态流出物取样系统气溶胶穿透效率试验研究

核设施烟囱气态流出物的取样是否具有代表性,将直接影响流出物测量的准确性。气溶胶穿透效率是取样代表性的关键指标之一。本文介绍了取样系统气溶胶穿透效率的试验方法和试验要求,针对国内某在建核电站,开展了Da=10μm粒径下的穿透效率验证试验,其结果为48. 42%;三级取样管线取消弯头,采用直管连接,通过此改进后取样管路的穿透效率提升至53. 21%,满足标准中大于50%的要求。     

核电厂气态流出物监测样品分析中的核素衰变校正应用探讨

在放射性样品分析工作中,样品测量阶段、闲置阶段、采样阶段都涉及到核素衰变校正。根据放射性核素的衰变规律,阐明了这三种类型的衰变校正方法原理及其系数的计算。针对核电厂气态流出物中短半衰期的放射性惰性气体和碘核素的监测分析工作,研究了对监测样品分析结果进行衰变校正的必要性。基于实验测量结果,探讨了衰变校正对γ能谱分析法中核素的活度浓度探测下限的影响。介绍并分析了核电厂气态流出物监测样品分析工作中的衰变校正方案,对于短半衰期放射性惰性气体和碘核素,采用合理保守的衰变校正方案,可减少监测分析中的偏差,同时又可保证统计排放量的偏保守性。     

秦山核电厂气态流出物气溶胶粒径分布测量

对秦山核电基地烟囱流出物的粒径分布谱进行了实验测量,结果表明:气态流出物中的气溶胶粒径大多在1 μm以下,但总粒子数浓度随机组类型不同而不同。测量结果为评估核电厂气态流出物监测系统采样的有效性提供了参考。     

气态流出物取样不具代表性的案例浅析

本文通过对核燃料循环设施局部排风系统取样的工艺描述、取样现状调查研究、概率模型计算等对气态流出物局排取样不具代表性的问题进行了分析,对其中反映出来的短板弱项问题,如管理不够完善、核安全文化缺失、麻痹松懈等对营运单位落实核安全全面责任的不利影响进行了进一步分析,提出了一些强化风险防控、排查管理隐患的思考.     

秦山核电厂气态流出物监测系统的研究及其改进

秦山核电厂气态流出物取样监测系统采用了较长的水平取样管和垂直取样管．使核电站气态流出物排放监测存在设计上的缺陷由于取样管道过长（特别是水平取样管道）．会造成气溶胶和碘取样过程的管道沉积损失改进后,将气溶胶和碘的取样监测系统移至烟囱底部的校正室．取消了长度为85．4m水平取样管线．取样管改为只有70．6m的垂直管．试验测试及评价结果表明：放射碘的取样校正因子可取作1．65对于放射性气溶胶．按最保守的估计．事故工况下．其校正因子最大也不可能大于1．65;在正常运行期间,气溶胶粒子均以小粒径为主,可将校正因子取作 1.1.     

我国核电厂气态流出物中惰性气体监测现状

核电厂流出物尤其是气态流出物中的放射性惰性气体监测多为低水平放射性核素,我国运行核电厂的环境监测结果均低于探测限,无法计算照射剂量.探测能力决定了放射性惰性气体排放评价的结果.本文分析了我国各运行核电厂流出物放射性惰性气体监测和排放评价的现状,比较欧盟的相关建议,研究我国核电厂流出物放射性惰性气体监测能力存在的问题,并提出了建议.     

核电厂流出物放射性核素监测项目探讨

针对核电厂流出物放射性核素监测项目,从国际原子能机构通用要求、我国各运行核电厂放射性核素常规测量项目以及流出物低水平放射性核素实验研究成果等三方面进行了对比与分析,总结了核电厂流出物放射性核素监测中的必要性项目和推荐性项目,为各核电厂流出物放射性监测活动的开展以及审管部门的监管提供参考。     

核设施烟囱气态流出物取样代表性验证的技术要求研究

核设施的排放烟囱都设有连续取样和在线监测系统,对流出物中排放的气载放射性活度进行测量,以判明气态流出物排放是否满足管理限值要求,并及时发现污染异常,提供报警。但是,取样和监测系统中气态流出物的取样是否具有代表性,将直接影响流出物测量的准确性。ISO标准(ISO 2889—2010)和美国标准(ANSI/HPS N13.1—2011)都对核设施气态流出物取样监测的有关性能,提出了详细的量化指标和判定方法。本文在对上述标准编写的技术背景和国内外所开展的相关工作进行调研的基础上,研究分析了这些标准推荐的各项性能指标和验证方法。     

核电厂流出物低水平放射性核素监测问题的探讨

核电厂流出物低水平放射性核素监测能力是评价核电厂放射性流出物排放量的保证。为了探讨我国核电厂流出物监测管理存在的问题,通过改变运行核电厂常规采样、测量方式,实验研究了核电厂流出物低水平放射性核素活度水平及其探测限,并探讨了核电厂现行数据处理的计算结果与实际排放之间的差距。研究结果表明,流出物中低水平放射性核素的实际水平远低于其现阶段排放量统计水平,确实成为了我国核电厂统计排放量偏高的因素之一。在此基础上给出了建立核电厂流出物监测探测限指标监管要求和低水平监测结果数据处理的建议。     

Calculation of risk-based detection limits for radionuclides in the liquid effluents from Korean nuclear power plants

In order to review if present detection limits of radionuclides in liquid effluent from nuclear power plants are effective enough to warrant compliance with regulatory discharge limits, a risk-based approach is developed to derive a new detection limit for each radionuclide based on radiological criteria. Equations and adjustment factors are also proposed to discriminate the validity of the detection limits for multiple radionuclides in the liquid effluent with or without consideration of the nuclide composition. From case studies to three nuclear power plants in Korea with actual operation data from 2006 to 2015, the present detection limits have turned out to be effective for Hanul Unit 1 but may not be sensitive enough for Kori Unit 1 (8 out of 14 radionuclides) and Wolsong Unit 1 (9 out of 42 radionuclides). However, it is shown that the present detection limits for the latter two nuclear power plants can be justified, if credit is given to the radionuclide composition. Otherwise, consideration should be given to adjustment of the present detection limits. The risk-based approach of this study can be used to determine the validity of established detection limits of a specific nuclear power plant.     

An efficient Neuro-Fuzzy approach to nuclear power plant transient identification

Transient identification in nuclear power plants (NPP) is often a computational very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in NPPs. The bases for the transient identification relay on the evidence that different system faults and anomalies lead to different pattern evolution in the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments that represents a specific type of event. Recently, several works have been developed for transient identification. These works frequently present a non reliable response, using the “don´t know” as the system output. In this work, we investigate the possibility of using a Neuro-Fuzzy modeling tool for efficient transient identification, aiming to helping the operator crew to take decisions relative to the procedure to be followed in situations of accidents/transients at NPPs. The proposed system uses artificial neural networks (ANN) as first level transient diagnostic. After the ANN has done the preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. A validation of this identification system was made at the three loops Pressurized Water Reactor (PWR) simulator of the Human-System Interface Laboratory (LABIHS) of the Nuclear Engineering Institute (IEN/CNEN/Brazil). The obtained results show the potential of this new transient identification system to be used in an operational NPP in order to assist the operators to take decisions during transients/accidents.     

确定核电厂气载放射性流出物排放限值的设计方法研究

为了使核电厂的气载放射性流出物排放所致环境公众辐射剂量满足法规标准要求,本文从辐射防护剂量限值、大气污染物扩散规律及核电厂源项特点等方面考虑,通过实例分析,探讨了确定核电厂气载放射性流出物排放限值(排放率、排放浓度、地面浓度限值)的设计方法,提出了限值的建议值,并与已有核电厂的气载流出物排放报警阈值进行了验证,确认了方法的合理性。     

基于CFD方法的核设施气载流出物取样代表性研究

核设施气载流出物取样代表性的优劣,决定着对环境监测和辐射安全评价的准确性。应用计算流体力学(Computational Fluid Dynamics,CFD)方法,根据新版ISO 2889标准《核设施的烟道和通风管道中的放射性物质取样》中对取样位置的要求,对某核设施排气系统的流场进行了数值模拟并分析了不同取样位置的取样代表性。指出现有取样位置选取不尽合理的结论,通过对模拟结果进行分析,确定了合理取样位置,从而可以有效地监测放射性气载流出物的核素活度浓度,为辐射环境影响评价提供科学数据支持。     

Investigation of loss factors of aerosol particles in the sampling systems of a nuclear power plant

The loss factor of particles with different sizes in a sampling system of a nuclear power plant has been experimentally determined. As shown by optical particle size measurements, the particle size distribution of the test aerosol titanium dioxide (aerodynamic diameter between 0.6 and 8 μm) agrees well with that of the ambient aerosol in the nuclear power plant. The total loss factor for titanium dioxide lies between 1.3 and 1.6. This means a loss of particles of about 23–38% and shows that the transfer properties of the sampling system are good. The experimental results for small particles (aerodynamic diameter 0.6–8 μm) are in good agreement with those of theoretical estimations. The transfer properties of the sampling system for larger particles (aerodynamic diameters up to 3 mm) have been investigated, too. The results of the measured loss factors show that the sampling system is suitable even in this case.     

双通道人工放射性气溶胶监测系统设计

人工放射性气溶胶的监测环境较为复杂,有时甚至要在地下洞库、坑道或地下核设施测量,其环境中氡及其子体产物的浓度高达10~3-10~4 Bq·m~(-3)。在高氡钍环境下,氡钍能峰拖尾导致其放射性计数影响了人工气溶胶道址计数,进而影响人工放射性气溶胶的探测下限。本文研制了一种适合在高氡环境下工作的常压、真空双通道人工气溶胶监测仪。虽然常压测量通道测量准确度较差,但响应速度快,可以弥补真空测量通道测量准确但响应速度慢的缺点。本仪器对大气环境中钚、铀气溶胶检测的响应时间快,在有人工核素泄露的情况下最快30 min内可以得到测量结果,理论探测下限可达到10~(-4)-10~(-3) Bq·m~(-3),可验证探测下限为铀0.1 Bq·m-3、钚0.02 Bq·m~(-3),适用于各类需要对人工核素进行无人监测的高氡环境场所。     

核电站实物保护系统集成模式浅析

实物保护系统是对核电站及核设施的重要设备和核材料实施保卫,使其在任何情况下,重要设备遭受破坏和核材料丢失的危险减至最小。早期实物保护系统涉及的各子系统相互独立,无法形成整个系统的真正集成,导致资源浪费和投资成本的增加。本文研究了实物保护系统的不同集成模式,建立了基于对象的子系统相互平等互连的模式进行集成的模型,并着重研究了该集成模型的体系结构。