Conceptual Strategy for Mitigating the Risk of Hydrogen as an Internal Hazard in Case of Severe Accidents at Nuclear Power Plant Considering Existing Risks and Uncertainties Associated with the Use of Traditional Strategies

Hydrogen challenge mitigation stands as one of the main objectives in the management of severe accidents at Nuclear Power Plants (NPPs). Key strategies for hydrogen control include atmospheric inertization and hydrogen removal with Passive Autocatalytic Recombiners (PARs) being a commonly accepted approach. However, an examination of PAR operation specificity reveals potential inefficiencies and reliability issues in certain severe accident scenarios. Moreover, during the in-vessel stage of severe accident development, in some severe accident scenarios PARs can unexpectedly become a source of hydrogen detonation. The effectiveness of hydrogen removal systems depends on various factors, including the chosen strategies, severe accident scenarios, reactor building design, and other influencing factors. Consequently, a comprehensive hydrogen mitigation strategy must effectively incorporate a combination of strategies rather than be based on one strategy, taking into consideration the probabilistic risks and uncertainties associated with the implementation of PARs or other traditional methods. In response to these considerations, within the framework of this research it has been suggested a conceptual strategy to mitigate the hydrogen challenge during the in-vessel stage of severe accident development.

Assessment of fuel-rod meltdown in a severe accident at Bushehr nuclear power plant(BNPP)

After the Fukushima disaster, interest in the evaluation of severe accidents in nuclear power plants and off-site consequences has significantly increased. Because experimental studies are difficult to conduct, computational methods play a substantial role in accident analysis. In this study, a severe accident in the Bushehr pressurized water reactor power plant caused by a station blackout with a total loss of alternating current power supply has been evaluated. This analysis presents the in-core damage of fuel rods and the release of fission products as well as the thermal hydraulic response of the station components during the loss of active emergency cooling systems. In this manner, a perfect model of the Bushehr nuclear power plant using the MELCOR code is prepared. The accident progression is simulated, and the thermal responses of the fuels and hydraulic components are presented. It is shown that, without operator intervention, steam generators will become dry in approximately 3000 s, and the heat sink of the reactor will be lost. The simulation results show that at approximately 8600 s, the upper parts of the core start melting. This model calculates the shortest available time for accident prevention and proves that the time available is sufficient for operator manual action to prevent a nuclear disaster.

SRDAAR－QNPP：a computer code system for the real－time dose assessment of an accident release for Qinshan Nuclear Power Plant

The paper presents a computer code system 'SRDAAR- QNPP' for the real-time dose as-sessment of an accident release for Qinshan Nuclear Power Plant. It includes three parts:thereal-time data acquisition system, assessment computer. and the assessment operating code system. InSRDAAR-QNPP, the wind field of the surface and the lower levels are determined hourly by using amass consistent three-dimension diasnosis model with the topographic following coordinate system.A Lagrangin Puff model under changing meteorological condition is adopted for atmosphericdispersion, the correction for dry and wet depositions. physical decay and partial plume penetrationof the top inversion and the deviation of plume axis caused by complex terrain have been taken in-to account. The calculation domain areas include three square grid areas with the sideline 10 km, 40krn and 160 km and a grid interval 0.5 km, 2.0 km, 8.0 km respectively. Three exposure pathwaysare taken into account:the external exposure from immersion cloud and passing puff, the internalexposure from inhalation and the external exposure from contaminated ground. This system is ableto provide the results of concentration and dose distributions within 10 minutes after the data havebeen inputed.

The Role of Countermeasures in Mitigating the Radiological Consequences of Nuclear Power Plant Accidents

The countermeasures are the actions that should be taken, after the occurrence of a nuclear accident to protect the public against the associated risks. These actions may be represented by sheltering, evacuation, distribution of stable iodine tablets and/or relocation. This study represents a comprehensive probabilistic study to investigate the role of the adoption of the countermeasures in case of a hypothetical accident of type LOCA for nuclear power plant of PWR （1000 Mw）. The effective doses in different organs, short and long health effects, and the associated risks are calculated with and without countermeasures. In addition, the overall costs of the accident and the costs of countermeasures are estimated which represent our first trials to know how much the proposed accident cost. The results showed that, the area around the site requires early and late countermeasures action after the accident especially in the downwind sectors. For late countermeasures, the duration time of relocation ranged from about two to 10 years. The adoption of the countermeasures increases the costs of emergency plan by 40% but reduces the risk associated the accident.

Some Technical Solutions for Environmental Protection System during Accidents at Nuclear Power Plants

The paper reports some technical solutions, which suggested or used for increasing of environmental protection during accidents at NPPs. For NNPs with two protective shells and pressure release system such as WWER-1000 a comprehensive, passive-mode environmental protection system of decontamination of the radioactive steam-air mixture from the containment and the intercontainment area was suggested. This system includes the “wet” stage (scrubbers, etc.), the “dry” stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For WWER-440/230 NPPs three protection levels: 1) a jet-vortex condenser;2) the spray system;3) a sorption module were suggested and installed. For modern designs of new generation NPPs, which do not provide for pressure release systems, a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable, was proposed and after modernization was installed at the KudanKulam NPP (India).

Impact Analysis of the 2011 Fukushima Nuclear Power Plant Accidents by Running Spectrum Analysis on Newspaper

Huge amount of digital data of the Great East Japan Earthquake is provided by the highly-developed digital data technology. But the method and technique for analysis of these huge digital data are not developed sufficiently. This paper proposes a running spectrum technique for text data and analyzing changes of disaster phase during the disaster management cycle. Impact analysis of the nuclear power plant accidents have been performed by using Fukushima Minpo newspaper for its verification. The result shows the dynamic characteristics of the nuclear power plant accidents. As the time interval B becomes longer, the analysis data is used from wide range period along with the smoothing effect. When observing different time intervals B, fewer keywords have been ranked in the longer time intervals of B. The proposed technique is a powerful tool to effective and efficient disaster response and management. analyze effectively the huge amount of digital data for the

Safety Implementation of Hydrogen Igniters and Recombiners for Nuclear Power Plant Severe Accident Management

Hydrogen combustion in a nuclear power plant containment building may threaten the integrity of the containment. Hydrogen recombiners and igniters are two methods to reduce hydrogen levels in containment buildings during severe accidents. The purpose of this paper is to evaluate the safety implementation of hydrogen igniters and recombiners. This paper analyzes the risk of deliberate hydrogen ignition and investigates three mitigation measures using igniters only, hydrogen recombiners only or a combination of recombiners and igniters. The results indicate that steam can effectively control the hydrogen flame acceleration and the deflagration-to-detonation transition.

Fukushima accident study using MELCOR

The accidents at the Fukushima Daiichi nuclear power station stunned the world as the sequences played out over severals days and videos of hydrogen explosions were televised as they took place. The accidents all resulted in severe damage to the reactor cores and releases of radioactivity to the environment despite heroic measures had taken by the operating personnel. The following paper provides some background into the development of these accidents and their root causes,chief among them,the prolonged station blackout conditions that isolated the reactors from their ultimate heat sink - the ocean. The interpretations given in this paper are summarized from a recently completed report funded by the United States Department of Energy (USDOE).

Possibility of Use of Noise Analysis for Identification of Reactor Conditions during Accidents

Knowledge and control of the thermo-hydraulic conditions in the reactor is required for the effective accident management. Dedicated and qualified for harsh environment instrumentation has to be in place for this purpose. Experience of the Fukushima Dai-ichi plant and the lessons learned from the European stress tests demonstrated that alternative and divers tools and methods are needed for the identification of reactor condition in extreme situations. In the paper the feasibility of development of an alternative accident monitoring via well-known noise diagnostics methods is proposed and demonstrated. The possibility of identification of reactor accident conditions using temperature and pressure fluctuations, noise of the neutron and gamma field is considered on the basis of research achievements in reactor noise. As an example the use of pressure fluctuations for accident monitoring is presented.

Investigation of the interaction of material of fuel cladding for WWER-1000 reactor with steam at a temperature of accident overheatings

The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grinding) and after additional chemical etching. The changes in the microstructure of tubes due to the interaction with steam were investigated. A comparison was made between the oxidation rate of this material (weight gain) and the data on the oxidation of other alloys for nuclear power plants. The oxidation rate of Zr1Nb(0.1% O) is close to the oxidation rate of other zirconium alloys. It is shown that after chemical treatment of the surface of the samples there is a more even growth of oxide films, and they have a smaller thickness for the same time of exposure than after mechanical grinding. Surface treatment before oxidation also affects the change of microstructure of samples when heated to high temperatures.

核电站应急系统EmInfoSys简介

核安全是核能与核技术利用事业发展的生命线，核事故应急响应是核安全纵深防御的最后一道屏障，是确保核能事业安全健康发展的重要措施之一。建立功能齐全、反应灵敏、运转高效的核与辐射事故应急管理体系，是我国核安全中重要任务之一。北京广利核系统工程有限公司从2001年开始参与核应急指挥系统的建设，先后参与了秦山、田湾、环境保护部、海阳、台山、防城港、三门等大量的核应急项目，形成了拥有自主知识产权的核应急平台———EmInfoSys （Emergency Management Information System），本文对EmInfoSys系统做了简要介绍。

BINELOCA 程序与广东岭澳核电站大破口失水事故分析

ＢＩＮＥＬＯＣＡ程序是在吸收国外先进大破口失水事故分析计算机程序的基础上，针对我国现有大破口失水事故分析程序的不足和工程应用方面的问题，采用了一些成熟的、新型的和当今国际公认的瞬态热工水力数学物理模型以及水和蒸汽性质参数计算公式改编而成的大破口失水事故分析程序。通过对广东岭澳核电站大破口失水事故的计算表明，ＢＩＮＥＬＯＣＡ程序计算的结果与法国计算的结果是一致的。

Text Mining Analysis of Efficiency of the Continuously Implemented Gathering Type Action Plan for Male Elderly People Obtained

Aim: To clarify transformation of the participants’ consciousness for rebuilding the community and its factors from the discussion contents by actions for male elderly people in Town A in Fukushima prefecture. Design: This study was an action research. Method: The author verbalized discussion contents of the action conducted in 2018-2019 and analyzed them for each year by the text mining method. Results: The word appearance frequency was high in the order of “Person” and “Town A” in both years. One large word network was formed in 2018 and its topic was about what the participants feel in their life in Town A. Two large word networks were formed in 2019 and their topic was about the community participation including difficulty in motivating others such as how people who do not participate can feel like joining it.

模拟事故工况下非能动核电厂安全相关涂层的可靠性测试及评估方法研究

安全相关涂层在非能动核电厂中起着重要的作用,涂层的失效会影响核电厂安全系统的功能执行,影响核安全。国内外核监管机构对于在设计基准事故(DBA)工况下涂层系统的可靠性及评估方法非常重视。文章结合非能动核电厂涂层系统的工程应用,针对其在DBA下的可靠性及评估方法进行了研究。研究表明:在DBA下非能动核电厂安全相关涂层的可靠性要综合考虑涂层的模拟DBA性能、干膜密度、导热性能等。而非能动核电厂安全相关涂层工程应用,则需从涂层的模拟DBA性能、干膜密度、导热性能、涂层碎片(数量、大小、位置和性能等)以及包络涂层碎片后的碎片裕量等角度进行综合评估,以确定在事故工况下涂层的可靠性,不对系统安全产生影响,保证核电厂更安全、高效和经济性运行。

基于差分电感的分体式压力/差压测量系统研究

随着安全级压力/差压变送器在核工业领域的广泛应用,提高其中传感器的可靠性已成为发展重点。对国内外核安全级压力/差压测量技术进行了深入研究。采用易远传的电感式传感器结构,设计了一种基于差分电感原理的安全级分体式压力/差压测量系统。阐述了测量系统中传感器和信号处理装置的详细设计方案,并对测量电路的设计进行了分析。通过分体式的设计方案,可有效提高测量设备的耐事故性能。该方案可为国内高可靠核级压力变送器产品的研发奠定基础,适用于核级压力、液位和流量信号的测量。该方案也适用于其他恶劣环境条件下非核测量领域的变送器产品研发。

核电厂安全级电气连接器的设计与试验

为实现安全级仪表信号的可靠传输,核电厂通常采用可拆卸的电气连接器连接仪表与电缆以及电缆与电缆。对电气连接器技术进行调研,设计了一种结构简单、性能可靠、操作安装方便、在地震以及严重事故下能有效吸收振动载荷、能够承受更长时间的辐照老化和热老化的安全级电气连接器。依托研制样机开展了功能性能试验和鉴定试验,试验结果表明:安全级电气连接器具有极高的可靠性,能够满足核电厂事故环境下的需求。该连接器可推广于其他恶劣环境条件下的应用领域。

核电站事故环境对环氧涂层理化性能的影响

以双酚A型环氧树脂和聚酰胺固化剂为主原料制备了环氧涂层,分别研究了核电站设计基准事故(DBA)和严重事故环境对其理化性能的影响。利用红外光谱(FTIR)对试验后的样品结构进行了对比分析,样品在辐照试验后检测出了苯醌结构的特征吸收峰。采用热重分析仪(TGA)测定了样品试验后的热稳定性参数。通过涂层附着力测试,评价了各试验对碳钢板与环氧涂层黏结强度的影响。结果表明:辐照试验产生苯醌是导致涂层黄变的主要原因;辐照和高温高压处理过程会导致环氧涂层的网络结构重排和断裂,对不同热稳定性参数的影响方向并不完全一致;湿热试验和DBA试验对涂层黏结强度影响较小,而辐照试验和严重事故则会显著降低涂层附着力。

东部区域核事故核素大气扩散特征研究

针对核事故核素大气环境影响,建立一种基于区域尺度天气预报模式WRF(weather research forecasting model)的核素不同尺度大气扩散特征分析方法。以东部多个核电厂为研究对象,基于再分析气象数据(final reanalysis data,FNL)驱动WRF,并结合高斯大气扩散模式研究了东部区域尺度以及核电厂周边尺度的核事故核素大气扩散分布特征。结果表明:区域尺度大气弥散因子最大值主要分布在计算区域的西部地区,并且其随距离增加呈现下降趋势,距离500 m和5000 m处二者相差约一个量级。核电厂周边尺度大气弥散因子最大值比区域尺度大气弥散因子最大值偏低一个量级;除秦山核电厂外,其他核电厂距离500 m和5000 m处的大气弥散因子最大值相差约一个量级。不同核电厂大气弥散因子最大值出现的方位受厂址气象条件影响存在显著差异;东部沿海地区比内陆大气扩散能力较强,有利于放射性核素的大气传输。

Experience gained in analyzing severe accidents for WWER RP using CC SOCRAT

The current Russian regulatory documents on the safety of nuclear power plant(NPP)specify the requirements regarding design basis accidents(DBAs)and beyond design basis accidents(BDBAs),including severe accidents(SAs)with core meltdown,in NPP design(NP-001-15,NP-082-07,and others).For a rigorous calculational justification of BDBAs and SAs,it is necessary to develop an integral CC that will be in line with the requirements of regulatory documents on verification and certification(RD-03-33-2008,RD-03-34-2000)and will allow for determining the amount of data required to provide information within the scope stipulated by the requirements for the structure of the safety analysis report(SAR)(NP-006-16).The system of codes for realistic analysis of severe accidents(SOCRAT)(formerly,thermohydraulics(RATEG)/coupled physical and chemical processes(SVECHA)/behavior of core materials relocated into the reactor lower plenum(HEFEST))was developed in Russia to analyze a wide range of SAs at NPP with water-cooled water-moderated power-generating reactor(WWER)at all stages of the accident.Enhancements to the code and broadening of its applicability are continually being pursued by the code developers(Nuclear Safety Institute of the Russian Academy of Sciences(IBRAE RAN))with OKB Gidropress JSC and other organizations.Currently,the SOCRAT/В1 code can be used as a base tool to obtain realistic estimates for all parameters important for computational justification of the reactor plant(RP)safety at the in-vessel stage of SAs with fuel melting.To perform analyses using CC SOCRAT/В1,the experience gained during execution of thermohydraulic codes is applied,which allows for minimizing the uncertainties in the results at the early stage of an accident scenario.This study presents the results of the work performed in 2010–2020 in OKB Gidropress JSC using the CC SOCRAT/В1.Approaches have been considered to develop calculational models and analyze SAs using CC SOCRAT.This process,which is clearly structured in OKB Gidropress JSC,provides a noticeable reduction in human involvement,and reduces the probability of erroneous results.This study represents the principal results of the work performed in 2010–2020 in OKB Gidropress JSC using the CC SOCRAT,as well as a list of the tasks planned for 2021–2023.CC SOCRAT/B1 is used as the base thermohydraulic SAs code.

日本福岛核事故后的海洋放射性监测进展

全面总结日本福岛核事故后不同核素通过海洋途径排放的放射性总量;介绍多个国家开展的海洋放射性监测工作;基于海洋放射性本底与各国制定的限值标准,对福岛核事故后海洋放射性监测结果进行评价;最后对国内外的海洋放射性监测工作进行展望,从海洋环境安全角度出发,提出建立我国海洋放射性数据库的必要性.