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.

Sleep and Stress of Late Middle Age Males Who Are Forced to Live in Emergency Temporary Houses and Post-Earthquake Public Houses for a Long Period Due to the Fukushima Daiichi Nuclear Power Station Accident

It has been revealed that prolonged shelter life caused by Fukushima Daiichi Nuclear Power Station accident in conjunction with the Great East Japan Earthquake influences sleep and mental health of the residents, and therefore its influence on their physical and emotional health has become a concern. Therefore, in this study, the authors aimed at clarifying actual situations of sleep and stress of middle age males living in shelters for a long period in each of emergency temporary houses and post-earthquake public houses. For 5 males who moved from emergency temporary houses to post-earthquake public houses, their objective and subjective sleep states were measured with Actigraph and PSQI, respectively. Furthermore, their objective and subjective stresses were measured with saliva stress biomarkers and GHQ28, respectively. Their data were analyzed by paired t-test. As result, in comparison between the life in the emergency temporary houses and post-earthquake public houses, significant variation was not recognized in their objective sleep states and saliva stress biomarkers though their subjective sleep and subjective stress were significantly worsened after moving to the post-earthquake public houses.

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).

Atmospheric Dispersion and Deposition of Radionuclides (¹³⁷Cs and ¹³¹I) Released from the Fukushima Dai-ichi Nuclear Power Plant

The Lagrangian Particle Dispersion Model (LPDM) in the 594 km× 594 km model domain with the horizontal grid scale of 3 km×3 km centered at a power plant and the Eulerian Transport Model (ETM) modified from the Asian Dust Aerosol Model 2 (ADAM2) in the domain of 70° LAT × 140° LON with the horizontal grid scale of 27 km×27 km have been developed. These models have been implemented to simulate the concentration and deposition of radionuclides (137Cs and 131I) released from the accident of the Fukushima Dai-ichi nuclear power plant. It is found that both models are able to simulate quite reasonably the observed concentrations of 137Cs and 131I near the power plant. However, the LPDM model is more useful for the estimation of concentration near the power plant site in details whereas the ETM model is good for the long-range transport processes of the radionuclide plume. The estimated maximum mean surface concentration, column integrated mean concentration and the total deposition (wet+dry) by LPDM for the period from 12 March to 30 April 2011 are, respectively found to be 2.975 × 102 Bq m-3, 3.7 × 107 Bq m-2, and 1.78 × 1014 Bq m-2 for 137Cs and 1.96 × 104 Bq m-3, 2.24 × 109 Bq m-2 and 5.96 × 1014 Bq m-2 for 131I. The radionuclide plumes released from the accident power plant are found to spread wide regions not only the whole model domain of downwind regions but the upwind regions of Russia, Mongolia, Korea, eastern China, Philippines and Vietnam within the analysis period.

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.

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.

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。

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。

Measurements of Fission Products from the Fukushima Daiichi Incident in San Francisco Bay Area Air Filters,Automobile Filters,Rainwater,and Food

A variety of environmental media were analyzed for fallout radionuclides resulting from the Fukushima nuclear accident by the Low Background Facility (LBF) at the Lawrence Berkeley National Laboratory (LBNL) in Berkeley, CA. Monitoring activities in air and rainwater began soon after the onset of the March 11, 2011 tsunami and are reported here through the end of 2012. Observed fallout isotopes include 131I, 132I, 132Te, 134Cs, 136Cs, and 137Cs. Isotopes were measured on environmental air filters, automobile filters, and in rainwater. An additional analysis of rainwater in search of 90Sr is also presented. Last, a series of food measurements conducted in September of 2013 are included due to extended media concerns of 134,137Cs in fish. Similar measurements of fallout from the Chernobyl disaster at LBNL, previously unpublished publicly, are also presented here as a comparison with the Fukushima incident. All measurements presented also include natural radionuclides found in the environment to provide a basis for comparison.

喷淋去除气溶胶的模型及实验研究

针对当前的安全壳喷淋系统去除气溶胶的计算模型对于未喷淋区的处理及液滴运动计算过于简化的问题,本文开展了喷淋去除气溶胶的模型及实验研究。通过将液滴运动模型与夹带气流模型进行耦合以获取喷淋液滴的分布及夹带气流速度,然后将二者与单液滴去除气溶胶效率模型结合建立了喷淋去除气溶胶计算模型,实现对喷淋区及未喷淋区内气溶胶浓度的计算。模型的计算结果与实验数据进行了比较验证,结果表明:计算模型能够较为准确地描述喷淋条件下气溶胶的浓度变化,模型对气溶胶质量浓度的计算误差小于5%;夹带气流的搅混作用使喷淋区与未喷淋区内气溶胶的浓度趋于一致,喷嘴附近的气溶胶具有更高的去除率。

基于差分电感原理的稳压器安全阀阀位测量系统研究

为实现稳压器安全阀阀位的精准测量,核电厂通常采用安全级阀位测量系统进行实时信号监测。对核电厂阀门位移量测量技术进行了深入研究。设计了一种基于差分电感原理的稳压器安全阀阀位测量系统。通过采用分体式的设计方案,有效提高了测量设备的耐事故性能,并依托研制的试验样机开展了鉴定试验。试验结果表明,基于差分电感原理的稳压器安全阀阀位测量系统具有较高的测量精度,能够在正常和事故条件下实现安全阀阀位的准确测量。该系统后续可推广应用于其他恶劣环境条件下的位移测量领域。

核电厂严重事故人员可靠性分析研究现状与展望

福岛事故缓解过程中暴露出的较多人因问题,使得核电厂严重事故场景下的人员可靠性逐渐成为研究热点。为分析国内外核电厂严重事故人员可靠性的研究现状,探究后续的研究方向,本文基于文献调研、核电厂走访、专家访谈等方法,阐述了开展核电厂严重事故下人员可靠性分析研究的必要性,从国内外核电厂严重事故管理、HRA方法、严重事故人员可靠性研究等3个方面对核电厂严重事故人员可靠性分析的研究现状进行评述,指出国内外针对核电厂严重事故场景的HRA方法研究还处于初步阶段,并题出未来的研究工作应从应急人员行为模型、人员失误因果模型、人因数据的收集和规范化等方面开展。

PISAA软件中含不凝性气体的蒸汽冷凝模型研究与验证

针对核电厂事故状态下含不凝性气体的蒸汽在安全壳内壁面上的冷凝传热问题,本文以严重事故一体化分析软件的开发为背景,建立了基于热质比拟方法的安全壳内含不凝性气体蒸汽冷凝计算模型,用于求解安全壳内壁面由于冷凝造成的传热量以及产生的凝结水的量,并通过Wisconsin实验以及ISP47-TOSQAN实验对该模型进行了验证。结果表明:严重事故一体化分析软件中的冷凝模型针对Wisconsin实验各工况下的冷凝传热系数计算结果与实验测量值基本一致,冷凝传热系数平均偏差在10%左右,最大偏差不超过30%,整体与实验测量值吻合较好,均在工程可接受偏差范围内;严重事故一体化分析软件针对ISP47-TOSQAN实验中的阶段A和阶段B的计算结果与实验测量值吻合,稳态1/2/3/4保持一致,准确地模拟了实验过程。

移动式放射性废水应急处理装置的设计与试验研究

为了应对核电站突发严重事故时产生的高含盐放射性废水,设计制造了一套处理量为5 m~3/h的移动式放射性废水应急处理装置。该装置采用了“过滤+反渗透+离子交换”的组合工艺。配制Co^(2+)、Sr^(2+)、Cs^(+)质量浓度均为100 mg/L,TDS=7 000 mg/L的模拟放射性废水,用于评价该处理装置的性能。结果表明,该移动式放射性废水应急处理装置对Co^(2+)、Sr^(2+)、Cs^(+)的总去污因子(DF)分别为1.19×10^(5)、1.02×10^(5)、7.30×10^(3),对3种核素的综合去污因子为1.74×10^(4),3种核素的总去除率大于99.99%;同时,该装置对模拟放射性废水中的盐类也具有良好的去除效果。基于该移动式放射性废水应急处理装置对废水中核素的优异去除效果,其可以作为核电站放射性废水的应急处理方案。

典型事故工况下压水堆核电厂内工作人员辐射风险分析方法研究

基于概率论和确定论分析方法建立了事故工况下场内工作人员辐射剂量控制的体系。针对典型三代压水堆核电厂,建立了事故工况下场内工作人员辐射风险分析的方法论,并采用典型事故进行验证。验证结果表明,对于选取的堆外放射性系统相关典型事故,建立的辐射风险控制体系和分析方法可很好地评估并控制事故工况下场内工作人员的辐射风险。该方法可进一步扩展至堆芯相关事故以及其他堆外放射性系统相关事故,从而提升压水堆核电厂辐射防护最优化水平。

混合神经网络的核电站故障程度评估方法

为了准确处理大量复杂数据,避免由于压力上升导致的操纵员判断失误,本文提出了一种基于改进的粒子群算法和留一交叉验证的混合神经网络超参数优化方法,辅助操纵员评估核电站故障程度。该方法通过改进的粒子群算法优化超参数组合,利用留一交叉验证评估深度学习模型泛化性能,最终构建高精度故障程度评估模型。本文以核电站失水事故为对象,对所提出方法进行测试验证。结果表明:本文提出的混合神经网络超参数优化方法能够搜索最优超参数组合,构建绝对精度为97%的神经网络模型,能有效评估核电站故障程度,辅助操纵员维修决策。