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.     

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.     

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.     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.     

A consistent approach to severe accident management

The approach adopted for severe accident management (SAM) at the Loviisa nuclear power plant (in Finland) is presented and discussed. The approach includes a number of significant hardware changes and procedures that allow lowering of the lower head thermal insulation and neutron shield assembly, opening of the ice condenser doors, and spraying (externally) of the steel shell of the containment. It is expected that with these changes we can assure in-vessel debris coolability and retention, gradual burning of the hydrogen with good access to the ice condenser, and long term stabilization of the containment pressure, even in the absence of the residual heat removal system. Methodological aspects of demonstrating these SAM objectives, and the status of work in support of related quantifications (of key phenomena), are included in sufficient detail to provide an integrated perspective of the approach taken. The detailed quantifications, separately on each task, will follow, as respective research and quantification programs come to completion.     

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4–CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4–CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4–CANDU code predictions.     

Reactor core isolation cooling system analysis of the Fukushima Daiichi Unit 2 accident with RELAP/ScdapSIM

The reactor core isolation cooling (RCIC) system is an auxiliary system of a boiling water reactor (BWR) that provides makeup water in the case of a severe accident. During the Fukushima accident, the extended operation of the RCIC had a large influence on the accident progression and delayed the core meltdown by almost 70 h. During the Fukushima accident, the water level in the reactor pressure vessel (RPV) was assumed to rise enough to flood the main steam line (MSL), which caused the water to move through the RCIC steam turbine and reduce the overall system water injection capability. A RELAP/ScdapSIM analysis was carried out by using RCIC nodalization to reproduce the Fukushima accident and evaluate the impact of the RCIC system on the accident progression. A coefficient based on the critical flow model was included in the RELAP/ScdapSIM source code to reproduce the degradation suffered by the turbine due to the presence of water. Although highly simplified, the analysis demonstrated the RCIC system's feedback capability, which allows the RCIC to control the plant conditions for a long period of time without any human interaction.     

Development of severe accident management guidance (SAMG) for the Canadian CANDU 6 nuclear power plants

This paper discusses the severe accident management guidance (SAMG) development process undertaken for the Canadian CANDU 6 nuclear power plants (NPPs); the customization process of the generic CANDU SAMG for the Point Lepreau NPP is presented. Examples of severe accident management (SAM) guidelines related to containment pressure control are included in this paper. This paper also provides an overview summary of the severe accident analysis program at Atomic Energy of Canada Limited (AECL) that complements the SAM guidelines development process for the CANDU 6 NPPs in Canada. These analyses provided insights into the accident progression and basis to develop the SAM guidelines.     

The manipulation of time-varying dynamic variables using the rulemodification method and performance-index in NPP accident diagnosticexpert systems

When an expert system is being developed for nuclear power plant (NPP) accident diagnosis, the most difficult problem is the manipulation of the time-varying dynamic variables. To meet this need, the authors propose modification of the rules for accident diagnosis when plant parameters and conditions are varying. In addition to the rule-modification method, a pattern-matching method using the performance index is suggested. This system also uses the results of transient analysis and accident analysis codes as a database. To simulate this expert system, PROLOG was used to construct the knowledge base. The inherent backtracking inference strategy was used for the inference engine. Feedwater line piping failure was selected for system verification     

Classification method of severe accident condition for the development of severe accident instrumentation and monitoring system in nuclear power plant

Instrumentation and monitoring systems in a nuclear power plant are very important to monitor plant conditions for safe operations and a plant shutdown. The severe accident at TOKYO ELECTRIC POWER COMPANY's Fukushima Daiichi Nuclear Power Station (hereinafter called as TF1) in March 2011 caused several severe situations such as core damage, hydrogen explosion, etc. Lessons learned from the severe accident at TF1 show that an appropriate operable instrumentation and monitoring system for a severe accident should be developed so that the system will deliver an appropriate performance for mitigation of severe accident condition in a nuclear power plant.

This paper proposes the classification method of severe accident condition for the development of an appropriate operable instrumentation and monitoring system for a severe accident based on the problem analysis of monitoring variables during the severe accident at TF1. The classification is formed on the basis of the integrity of boundary for plant safety and the successful (or unsuccessful) condition of the cooling water injection, and is used for an establishment of defining severe accident environmental conditions for the instrumentation and monitoring system. Examples of the establishment method are also shown in this paper.     

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment.     

Optimum RCS depressurization strategy for effective severe accident management of station black out accident

Plant specific severe accident management guidelines (SAMG) for operating plants are developed and implemented in Korea as was required by government policy on severe accident. Korea Institute of Nuclear Safety (KINS) has recently reviewed feasibility of the developed SAMG for Ulchin unit 1 plant. Among the strategies referred in SAMG, we have intensively analyzed the reactor coolant system (RCS) depressurization strategy during station black out (SBO) accident scenario, which has a high probability of occurrence according to Ulchin unit 1 Probabilistic Safety Analysis (PSA). In depressurization strategy of the current SAMG, operators need to depressurize rapidly RCS pressure below 2.75 MPa using pressurizer (PZR) pilot operated safety relief valves (POSRVs) for high pressure accident like SBO. The rapid depressurization is effective in allowing the water of safety injection tank (SIT) to be injected into the core, but an excessive discharge of the SIT water is not desirable for an economical use of SIT inventory. Lack of SIT water accelerates the core damage in case the failed electric power do not recover in due to time. The SIT inventory economy means here that we should not waste the water inventory of SIT and use it in the most efficient way to cool the core. In case we do not use it in an economical way, the SIT might be depleted too rapidly, thus leaving an insufficient reservoir for post-depressurization cooling. The quantification of this SIT inventory economy for plant specific situation is of interest to develop an optimum depressurization strategy. In this study we have analyzed an effectiveness of current depressurization strategy for SBO accident with the severe accident analysis code MELCOR 1.8.5 which has been used for regulatory purpose in KINS. The entry time of severe accident management, a grace time gained by the current strategy, and the economy of the discharge mass flow rate for Ulchin plant were evaluated. Moreover, through a simple energy balance equation we could find an optimum strategy for RCS depressurization. The proposed strategy is based on finding an optimum discharge rate for an efficient use of the SIT inventory and it allows us to handle an SBO accident with higher confidence. The proposed strategy is yet a theoretical one, but possibilities of how to incorporate this strategy into engineered safety features are also discussed.     

Study on severe accident mitigation measures for the development of PWR SAMG

In the development of the Severe Accident Management Guidelines （SAMG）, it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident （LOCA）, Steam Generator Tube Rupture （SGTR）, Station Blackout （SBO）, and Anticipated Transients without Scram （ATWS） in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management （SAM） such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.     

小破口引发的严重事故工况及事故缓解的研究

利用MAAP4程序对方家山核电站进行建模,针对事故后果较为严重的小破口事件进行了计算分析,得到了假设事故下电厂系统的反应以及相应的严重事故现象.对事故中发生的DCH(安全壳直接加热)现象和安全壳失效以及裂变产物向环境的释放进行了分析.随后,本文根据相关的严重事故管理导则和该事故的特点,对缓解该事故的策略进行了研究和计算...     

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.     

The IAEA accident management programme

Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff.     

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

During a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.     

Applicability of Phebus FP results to severe accident safety evaluations and management measures

The international Phebus FP (Fission Product) programme is the largest research programme in the world investigating core degradation and radioactive product release should a core meltdown accident occur in a light water reactor plant. Three integral experiments have already been performed. The experimental database obtained so far contains a wealth of information to validate the computer codes used for safety and accident management assessment.     

A dynamic neural network based accident diagnosis advisory system for nuclear power plants

In this work, an accident diagnosis advisory system (ADAS) using neural networks is developed. In order to help the plant operators quickly identify the problem, perform diagnosis and initiate recovery actions ensuring the safety of the plant, many operator support systems and accident diagnosis systems have been developed. The ADAS is a kind of such accident diagnosis system, which makes the task of accident diagnosis easier, reduces errors, and eases the workload of operators by quickly suggesting likely accidents based on the highest probability of their occurrence. In order to perform better than other accident diagnosis systems, the ADAS has three main objectives. To satisfy these three objectives, two kinds of neural networks that consider time factors are used in this work. A simple accident diagnosis system is implemented in order to validate the ADAS. After training the prototype, several accident diagnoses were performed. The results show that the prototype can detect the accidents correctly with good performance.     

Simulation of the chernobyl accident

An analysis of the April 26, 1986 accident at the Chernobyl-4 nuclear power plant in the Soviet Union is presented. The peak calculated core power during the accident was 550 000 MW_t. The analysis provides insights that further understanding of the plant behavior during the accident. The plant was modeled with the RELAP5/MOD2 computer code using information available in the open literature. RELAP5/MOD2 is an advanced computer code designed for best-estimate thermal-hydraulic analysis of transients in light water reactors. The Chernobyl-4 model included the reactor kinetics effects of fuel temperature, graphite temperature, core average void fraction, and automatic regulator control rod position. Preliminary calculations indicated the effects of recirculation pump coast down during performance of a test at the plant were not sufficient to initiate a reactor kinetics-driven power excursion. Another mechanism, or “trigger” is required. The accident simulation assumed the trigger was recirculation pump performance degradation caused by the onset of pump cavitation. Fuel disintegration caused by the power excursion probably led to rupture of pressure tubes. To further characterize the response of the Chernobyl-4 plant during severe accidents, simulations of an extended station blackout sequence with failure of all feedwater are also presented. For those simulations, RELAP5/MOD2 and SCDAP/MOD1 (an advanced best-estimate computer code for the prediction of reactor core behavior during a severe accident) were used. The simulations indicated that fuel rod melting was delayed significantly because the graphite acted as a heat sink.