Development of severe accident management advisory and training simulator (SAMAT)

The most operator support systems including the training simulator have been developed to assist the operator and they cover from normal operation to emergency operation. For the severe accident, the overall architecture for severe accident management is being developed in some developed countries according to the development of severe accident management guidelines which are the skeleton of severe accident management architecture. In Korea, the severe accident management guideline for KSNP was recently developed and it is expected to be a central axis of logical flow for severe accident management. There are a lot of uncertainties in the severe accident phenomena and scenarios and one of the major issues for developing a operator support system for a severe accident is the reduction of these uncertainties. In this paper, the severe accident management advisory system with training simulator, SAMAT, is developed as all available information for a severe accident are re-organized and provided to the management staff in order to reduce the uncertainties. The developed system includes the graphical display for plant and equipment status, the previous research results by knowledge-base technique, and the expected plant behavior using the severe accident training simulator. The plant model used in this paper is oriented to severe accident phenomena and thus can simulate the plant behavior for a severe accident. Therefore, the developed system may make a central role of the information source for decision-making for a severe accident management, and will be used as the training simulator for severe accident management.     

Role of the man–machine interface in accident management strategies

First, this paper gives a short general review on important safety issues in the field of man–machine interaction as expressed by important nuclear safety organisations. Then follows a summary discussion on what constitutes a modern Man–Machine Interface (MMI) and what is normally meant with accident management and accident management strategies. Furthermore, the paper focuses on three major issues in the context of accident management. First, the need for reliable information in accidents and how this can be obtained by additional computer technology. Second, the use of procedures is discussed, and basic MMI aspects of computer support for procedure presentation are identified followed by a presentation of a new approach on how to computerise procedures. Third, typical information needs for characteristic end-users in accidents, such as the control room operators, technical support staff and plant emergency teams, is discussed. Some ideas on how to apply virtual reality technology in accident management is also presented.     

Analyses for VVER-1000/320 reactor for spectrum of break sizes along with SBO

Severe accident analysis of a reactor is an important aspect for evaluation of source term. This in turn helps in emergency planning and severe accident management (SAM). Analyses have been carried out for VVER-1000 (V320) reactor following LOCA along with station blackout (SBO) to generate information on these aspects. Availability and unavailability of hydro-accumulators (HAs) are also considered for this study. Integral code ASTEC V1.3 (jointly developed by IRSN, France, and GRS, Germany) is used for analysing the transients. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for a spectrum of break sizes to provide information for probabilistic safety analysis (PSA) level-2 and severe accident management (SAM) guidelines.     

危冷系统对蒸汽发生器传热管破损事故缓解能力分析

针对蒸汽发生器传热管破损事故后果严重,危冷系统投入对事故缓解能力认识不清的问题,为提高危冷系统对事故缓解能力的认识,增强传热管破损事故处置能力,利用MELCOR程序建立了破损安全分析模型。通过计算对比分析了危冷系统投入与否对事故后果的影响,并比较了危冷系统对不同尺寸传热管破损事故的缓解能力。经仿真分析,明确了危冷系统对传热管破损事故的缓解能力,对提高运行人员事故处置能力及保证反应堆运行安全有重要意义。      

运行核电厂事故管理若干问题的探讨

本文就运行核电厂事故管理的必要性、实施原则及事故管理大纲等问题进行了探讨。本文认为,为了在超设计基准事故发展过程中进行事故管理,应制定和实施核电厂事故管理大纲,使事故管理所需要的所有物项都处在备用状态,以便需要时进行有效地事故管理。     

基于MELCOR与MCNP程序的安全壳剂量率计算方法

严重事故条件下,评估安全壳内的放射性剂量率水平对核电厂严重事故管理、应急响应等环节具有重要指导意义。本工作利用MELCOR程序模拟严重事故序列,计算不同核素组释放进入安全壳内的质量;利用ORIGEN2程序计算不同核素组的堆芯积存量及核素的γ源强;利用MCNP程序计算每组核素100%释放进入安全壳所产生的剂量率水平;最后根据拟合公式求解安全壳剂量率。中核核电运行管理有限公司30万千瓦机组安全壳剂量率的计算结果说明该方法切实可行。     

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.     

压水堆释放源项快速估算程序开发

事故时向环境释放的源项是确定核电厂(NPP)应急响应水平和防护行动决策的重要依据。基于电厂工况估算源项是核电厂严重事故应急响应期间重要的应急评价内容之一。在国际原子能机构(IAEA)和美国核管会(NRC)的有关技术文档基础上,本文介绍了基于压水反应堆(PWR)工况进行事故释放源项估算的步骤和基础数据,并归纳了7种实用的事故释放源项估算方法。基于这些方法,开发了PWR事故时环境释放源项快速估算程序。该程序为不同估算方法提供4种释放途径:安全壳泄漏、安全壳旁通、蒸汽发生器传热管破裂(SGTR)和直接环境释放,除直接环境释放途径外,其他释放途径都估算了核素释放过程中的衰变、滞留、喷淋和过滤等减弱过程。对比发现,软件计算结果与美国核管会的RASCAL软件释放源项计算结果接近。     

SAMEX: A severe accident management support expert

A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R&D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.     

间接氦气透平循环高温堆紧急停堆动态仿真

通过对10 MW高温气冷堆氦气透平发电装置(HTR-10GT)的堆芯、热交换器和透平压气机组等主要设备的数学建模和程序编制,初步建立起了一套模拟该装置瞬态特性的仿真程序.通过对该装置于5s时刻堆内引入0.1$阶跃正反应性引发的紧急停堆事故的瞬态模拟,初步验证了该装置紧急停堆预案设置的安全性和合理性,证明了旁路快开阀的设...     

核电厂事故后应急环境监测谱仪国产化研制

事故后应急环境监测谱仪是福岛事故后开始在我国核电厂环境监测系统使用的一种新型探测器,但目前国内使用的此类设备基本都为进口设备,综合使用成本高昂,严重制约了核电厂应急监测能力的提高。本工作成功研制出事故后应急环境监测谱仪,经过一系列测试,性能指标已达到国际先进产品水平,且软件系统完全开放,批量生产成本仅约为进口设备采购价格的二分之一。研制的事故后应急环境监测谱仪能够在核电厂正常运行和事故情况下发挥有效监测作用,未来具有在其他核设施推广使用的前景。     

Thermal-hydraulic analysis of accidents leading to local coolant flow decrease in the main circulation circuit of RBMK-1500

There are a few transient and loss-of-coolant accident conditions in RBMK-1500 reactors that lead to a local flow decrease in fuel channels. Because the coolant flow decreases in fuel channels (FC) leads to overheating of fuel claddings and pressure tube walls, mitigation measures are necessary. The accident analysis enabled the suggestion of the new early reactor scram actuation and emergency core cooling system (ECCS) initiation signal, which ensures the safe shutdown of the reactor and compensates the stagnation flow. Analysis of such conditions is presented in this paper. Thermal-hydraulic analysis was conducted using the state-of-the-art RELAP5 code. Results of the analysis demonstrated that, after implementation of the developed management strategy for destruction of local flow stagnation, the Ignalina nuclear power plant (NPP) would be adequately protected following accidents, leading to local coolant flow decrease in the primary circuit.     

核事故早期可防止剂量的计算

本文介绍了一种事故早期可防止剂量的计算方法，分析讨论了时间，空间，环境和不同防护行动之间相互作用等对防护行动实施效果有主要影响的因素；通过计算不采取任何防护行为时所受到的剂量与采取防护行动后的剂量之差得到可防止剂量，将其与国际原子能机构建议采用的通用干预水平进行比较，可得到应采取紧急防护行为的区域，同时也可以为进一步的防护决策最优化提供剂量数据。该方法已经应用在广东省核事故场外后果预测评价系统（GNARD2．0）和秦山地区环境核事故后果评价系统（QS－NUCAS1．0）之中。     

Development of a short-term emergency assessment system of the marine environmental radioactivity around Japan*

The Japan Atomic Energy Agency (JAEA) has, for many years, been developing a radionuclide dispersion model for the ocean, and has validated the model through application in many sea areas using oceanic flow fields calculated by the oceanic circulation model. The Fukushima Dai-ichi Nuclear Power Station accident caused marine pollution by artificial radioactive materials to the North Pacific, especially to coastal waters northeast of mainland Japan. In order to investigate the migration of radionuclides in the ocean caused by this severe accident, studies using marine dispersion simulations have been carried out by JAEA. Based on these as well as the previous studies, JAEA has developed the Short-Term Emergency Assessment system of Marine Environmental Radioactivity (STEAMER) to immediately predict the radionuclide concentration around Japan in case of a nuclear accident. Coupling the STEAMER with the emergency atmospheric dispersion prediction system, such as Worldwide version of System for Prediction of Environmental Emergency Dose Information version II (WSPEEDI-II), enables comprehensive environmental pollution prediction both in air and the ocean.     

关于通用干预水平及其应用的讨论

本文着重介绍了自前苏联的切尔诺贝利事故以来，IAEA、ICRP等国际组织在发展与完善用于核事故应急决策中的干预体系所进行的一系列活动与成果，阐述了在干预的概念、干预原则以及干预水平的表示、取值等 问题上发生的变化。鉴于目前国际上推荐的干预水平，从建立的方法采用的剂量学表示量及取值方法等均比过去更趋合理和便于应用，建议加强这方面的研究，并结合国情与核电站实际，对已有的法规、安全导则及核电站场内、外应急计划中有关应急干预的部分加以修改。     

Development and analysis of LOCA sequences for severe accident risk database

Although a Level 2 PSA has been performed for the Korean Standard Power Plants (KSNPs), and that it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access of the results. At present, KAERI is intensively calculating the severe accident sequences for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviors. The developed database (DB) system includes a graphical display for a plant and equipment status, previous research results by a knowledge-based technique, and the expected plant behavior. The plant model used in this paper is oriented to the cases of LOCAs related to severe accident phenomena and thus can simulate the plant behaviors of a severe accident. Therefore, the developed system may play a central role as an information's source during the decision-making for a severe accident management, and be used as a training simulator for a severe accident management.     

The need for further nuclear safety research: A present French view

The Chernobyl accident strengthens once more the need for a high level of safety for nuclear plants. We can still improve this safety both by operational feedback and by research which must be considered as an important and continuous way of progress.A review of the French evolution in safety research is presented, with special attention paid to the actions accelerated after Chernobyl (venting system for containment building, Phebus Fission Products, Ressac) and to those being considered recently after this accident.Finally, if the Chernobyl accident did not modify in depth the safety research priorities settled after TMI, it stresses the need to maintain strong support to research and to prepare the means capable of dealing with an emergency situation and efficient crisis management.     

Chernobyl experience of emergency data management

The use of the Chernobyl experience in emergency data management is presented. Information technologies for the generalization of practical experience in the protection of the population after the Chernobyl accident are described. The two main components of this work are the development of the administrative information system (AIS) and the creation of the central data bank. The current state of the AIS, the data bank and the bank of models is described. Data accumulated and models are used to estimate the consequences of radiation accidents and to provide different types of prognosis. Experience of accumulated analysis data allows special software to be developed for large-scale simulation of radiation consequences of major radiation accidents and to organize practical exercises. Some examples of such activity are presented.     

乏燃料水池严重事故现象及管理策略研究

针对我国二代改进型三环路核电厂乏燃料水池冷却管线破口事故(LOCA)引发的严重事故,使用MECLOR1.8.6程序进行了建模计算,分析研究了严重事故进程和乏燃料组件加热、熔化以及氢气的产生等主要现象。结果表明,乏燃料水池严重事故进程相对缓慢,但乏燃料组件的熔化及产生的氢气风险还是可能最终造成放射性向环境的大量释放。此外,本文还对乏燃料水池严重事故管理导则中的应急注水策略和氢气风险管理策略的有效性进行了计算分析,得到了严重事故下执行相关策略的时间窗口,从而为同类型核电厂严重事故管理导则的开发和有效执行提供支持。     

Approach to accident management in RBMK-1500

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK.Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed.This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account.