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.     

简化源项计算方法

利用程序包STCP估算核电厂在发生严重事故情况下释放到环境的放射性源项需要耗费大量计算机时间和费用。所以,对所有感兴趣的事故情景完成源项程序包的计算实际上是不可能的。为此,发展了一个简化源项计算方法。它的基本思想如下:首先利用STCP计算几个选择的序列,然后根据放射性释放和事故进程的特点处理和分析由STCP计算得到的结果,得出一些特定参数,其它事故序列的源项能够通过这些参数的重新组合得到。     

Quantification of severe accident source terms of a Westinghouse 3-loop plant

Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment.     

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。     

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.     

Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

In order to improve LWR source term under severe accident conditions, the first version of a fission product chemistry database named ‘ECUME’ was developed. The ECUME is intended to include several datasets of major chemical reactions and their effective kinetic constants for representative severe accident sequences. It is expected that the ECUME can serve as a fundamental basis from which fission product chemical models can be elaborated for use in the severe accident analysis codes. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system from 300 to 3000 K. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate fission product chemistry in Cs-I-B-Mo-O-H system under LWR severe accident conditions, where kinetics effects should be considered.     

Development of a risk-informed accident diagnosis and prognosis system to support severe accident management

RISARD, risk-informed severe accident risk diagnosis system, is a computerized tool developed to improve a severe accident management (SAM) for a nuclear power plant and to effectively support the MCR and the TSC in executing the relevant SAM activities. In order to provide a diagnostic capability to a state of the plant and a prognostic capability for an anticipated accident progression, the system examines (a) a symptom-based diagnosis of a plant damage state (PDS) sequence in a risk-informing way and (b) a PDS sequence-based prognosis of key plant parameter behavior, through a prepared database (DB) containing plant-specific severe accident risk (SAR)-related information. For a given accident, the replicated use of these two processes makes it possible to obtain information about the functional states of the plant and containment safety systems expected at the time of a severe accident as well as future trend of the key plant parameters that are essentially required for taking the relevant SAM action, eventually leading to an answer about the best strategy for SAM. The foregoing concept for an accident diagnosis and prognosis can give the SAM practitioners more time to take action for mitigating the consequences of the potential accident scenarios since they are made in a simple, fast, and efficient way through a prepared SAR database and it is useful especially when the plant information available for SAM is incomplete and limited. The main purpose of this paper is to (a) introduce the concept of the RISARD system proposed to support SAM and its implementation through a prepared OPR1000 plant- and MAAP code-specific SAR database and (b) assess prediction capabilities of major events expected during the evolution of a severe accident through the system.     

秦山一厂严重事故管理导则典型事故缓解对策有效性验证

在日本福岛核事故后,国家核安全局要求核电运营单位提升应对严重事故的能力。按照国家核安全局要求,秦山一厂开发了严重事故管理导则。应用MELCOR程序建立了秦山一厂严重事故分析模型,模拟典型严重事故序列,根据严重事故管理导则的缓解对策,分析实施事故缓解对策对核电厂主要参数的影响,从而验证事故缓解对策的有效性。分析结果表明:在严重事故情况下,按照严重事故管理导则实施缓解对策,可有效地延缓或终止堆芯损坏的过程。     

CANDU6核电厂无过滤安全壳通风模式的研究

CANDU6核电厂早期设计未考虑严重事故对策,在严重事故下,CANDU6核电厂的安全壳容易失效。为了解决这一问题,本文研究了无过滤安全壳通风模式对CANDU6核电厂安全壳的影响。本文选取典型的全厂断电严重事故,利用重水蒸气回收系统作为无过滤安全壳通风的路径,初步研究了该通风模式下对安全壳完整性的保持和对裂变产物源项的滞留能力。研究表明:该通风模式可以有效保持安全壳的完整性,同时,对裂变产物源项也有一定的滞留能力。     

压水堆核电站严重事故紧凑型仿真机开发

为了缓解压水堆核电站可能发生的严重事故的后果,也为了满足安全分析工程师和概率风险评价人员的需求,并在与国际原子能机构合作框架协议内,研制开发了紧凑型的严重事故仿真分析机MELSIM－PC。该仿真系统主要由仿真核心程序、同步通讯程序、人机界面程序等几个部分组成,可以工作在一台普通的微型计算机上,成功地实现MELCOR程序变量的运行数据库管理、电站动态图形显示、仿真计算控制、再启动和仿真重演等重要功能。     

严重事故下正常余热排出系统向堆芯注水策略分析

选取导致堆芯熔化频率最高的始发严重事故--直接注入（DVI）管线断裂事故,以及典型高压熔堆事故--丧失主给水始发事故(LOFW),利用MAAP4程序,分析反应堆堆芯热工水力行为,并对正常余热排出系统（RNS）堆芯注水策略的有效性与负面效应进行评估。分析结果表明,在DVI管线断裂事故和LOFW严重事故序列中,利用RNS进行堆芯注水可有效终止堆芯熔化进程,维持堆芯长期冷却。但堆芯再淹没会产生更多的氢气,存在增加安全壳氢气燃烧风险的可能性。此外通过分析利用严重事故管理导则中辅助计算文件给出的堆芯最小流量实施堆芯注水策略,讨论注水流量对堆芯冷却的影响,结果表明,在实施堆芯注水策略时,建议在系统允许的情况下采用更高的流速进行堆芯冷却。     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

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.     

Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic studies as a major contributor to core melt frequency and source term risk during shutdown conditions. The wide range of plant states encountered and the unavailability of certain safety features make it difficult to guarantee that safety systems operation will always be sufficient to terminate the accident evolution. In this context analyses are performed using MELCOR 1.8.5 for loss of residual heat removal (RHR) at various times during mid-loop operation of a Westinghouse two-loop PWR. In the absence of recovery of RHR or other accident management (AM) measures, the sequences necessarily lead to a long term core uncovery, heat-up and degradation, loss of geometry and eventual failure of the reactor pressure vessel (RPV). The results show an extensive time window before uncovery and additionally before core damage, which increase progressively with increasing time after shutdown at which loss of RHR occurs. Significant oxidation of the cladding may result in concentrations of hydrogen sufficient for deflagration. The slow evolution implies an opportunity for the plant operators to initiate AM measures even after core uncovery has started. The analyses indicate a substantial time window during the uncovery within which the injection can recover the core without damage. The upper end of the window is determined by the temperature at which heat from cladding oxidation becomes a dominant factor, marking a critical point for the effectiveness of this recovery mode. The results provide confidence in the inherent robustness of the plant with respect to accident sequences of this type.     

Accident management strategy in Sweden — implementation and verification

A comprehensive program for severe accident mitigation was completed in Sweden by the end of 1988. As described in this paper, this program included plant modifications such as the introduction of filtered containment venting, and an accident management system comprising emergency operating strategies and procedures, training and emergency drills. The accident management system at Vattenfall has been further developed since 1988 and some results and experience from this development are reported in this paper. The main aspects covered concern the emergency organization and the supporting tools developed for use by the emergency response teams, the radiological implications such as accessibility to various locations and the long-term aspects of accident management.     

核电厂二级PSA释放类划分及代表性事故序列选取研究

国内AP1000、EPR、华龙一号等核电工程项目已将二级概率安全分析（PSA）源项用于应急输入,但二级PSA释放类的划分以及各释放类代表性事故序列的选取尚无明确可操作的方法,需要进一步开展研究。对比研究国内先进核电厂二级PSA释放类划分和代表性事故序列选取情况,以国内某三代先进压水堆核电厂为例,在同一释放类中根据频率和后果选取4个不同的严重事故序列开展源项计算。结果表明,同一释放类4个不同事故序列的源项结果差别较大,建议释放类划分以应用为导向,根据分析目的进行迭代,对同一释放类应选取多个事故序列进行对比分析,以论证释放类划分的合理性和事故序列的代表性。      

Preliminary steps toward assessing aerosol retention in the break stage of a dry steam generator during severe accident SGTR sequences

Severe accidents SGTR sequences are identified as major contributors to risk of PWRs. Their relevance lies in the potential radioactive release from reactor coolant system to the environment. Lack of knowledge on the source term attenuation capability of the steam generator has avoided its consideration in probabilistic safety studies and severe accident management guidelines. This paper describes a research program presently under way on the aerosol retention on the tubes surrounding the breach within the secondary side of the steam generator in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST program. Experimental activities are focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development.     

高压安全注射系统对压水堆全厂断电事故的缓解能力分析

选择一个典型的3环路压水堆作为参考对象,采用最佳估算程序RELAP/SCDAPSIM/MOD3.2建立了一个典型的3环路压水堆严重事故计算模型。分析了全厂断电(SBO)事故引发的堆芯熔化基准事故后,高压安全注射系统对该事故的缓解能力。敏感性分析表明,堆芯出口温度达到920 K时,采用卸压充水缓解措施可以有效地阻止堆芯熔化,维持堆芯长期处于稳定、安全状态。     

Use of a fuzzy decision-making method in evaluating severe accident management strategies

In developing severe accident management strategies, an engineering decision would be made based on the available data and information that are vague, imprecise and uncertain by nature. These sorts of vagueness and uncertainty are due to lack of knowledge for the severe accident sequences of interest. The fuzzy set theory offers a possibility of handling these sorts of data and information. In this paper, the possibility to apply the decision-making method based on fuzzy set theory to the evaluation of the accident management strategies at a nuclear power plant is scrutinized. The fuzzy decision-making method uses linguistic variables and fuzzy numbers to represent the decision-maker's subjective assessments for the decision alternatives according to the decision criteria. The fuzzy mean operator is used to aggregate the decision-maker's subjective assessments, while the total integral value method is used to rank the decision alternatives. As a case study, the proposed method is applied to evaluating the accident management strategies at a nuclear power plant.     

Steam generator tube rupture (SGTR) scenarios

The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000–2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER).