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.     

A novel approach for radionuclide diffusion in the enclosed environment of a marine nuclear reactor during a severe accident

A severe accident in a marine nuclear reactor leads to radionuclide leakage, which causes hidden dangers to workers and has adverse effects of environmental pollution. It is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear reactor. Therefore, this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe...     

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.     

Comparison of stress-based and strain-based creep failure criteria for severe accident analysis

We conducted a parametic analysis of stress-based and strain-based creep failure criteria to determine if there is a significant difference between the two criteria for SA533B vessel steel under severe accident conditions. Parametric variables include debris composition, system pressure, and creep strain histories derived from different testing programs and mathematically fit, with and without tertiary creep. Results indicate significant differences between the two criteria. Stress gradient plays an important role in determining which criterion will predict failure first. Creep failure was not very sensitive to different creep strain histories, except near the transition temperature of the vessel steel (900 K to 1000 K). Statistical analyses of creep failure data of four independent sources indicate that these data may be pooled, with a spline point at 1000 K. We found the Manson-Haferd parameter to have better failure predictive capability than the Larson-Miller parameter for the data studied.     

Structural analysis of electrosleeved tubes under severe accident transients

A flow stress model was developed for predicting failure of electrosleeved PWR steam generator tubing under severe accident transients. The electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400°C during severe accidents because of grain growth. A grain growth model and the Hall–Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data, as well as high-temperature failure tests, on notched electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of electrosleeved tubes with throughwall and part-throughwall axial cracks in the parent tube during a postulated severe accident transient.     

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.     

Investigation of ALPHA experiment by severe accident analysis code SAMPSON

The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al₂O₃) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the code to simulate the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined where the results show its necessity in order to correctly reproduce the experimental trends.     

A generic approach for steel containment vessel success criteria for severe accident loads

Safety has been defined as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR) by the U.S. Department of Energy (DOE), Office of New Production Reactors (NP). The DOE-NP issued the Deterministic Severe Accident Criteria (DSAC) concept to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for containment vessel success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements and (2) Success Criteria. This paper is limited to a discussion of a generic approach for steel containment vessel success criteria. These criteria define acceptable containment response measures and limits for each problem statement. The criteria are based on the “best estimate” of failure with no conservatism. Rather, conservatism, if required, is to be provided in the problem statements prepared by the designer and/or the regulatory authorities. The success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles that may impact the containment. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements. The rationale upon which these criteria are based is contained in referenced documents. Rigorous validation of the criteria by comparison with results from analytical or experimental programs and application of the criteria to a containment design remain as future tasks.     

A simpler approach for predicting LWR vessel failure during severe accident conditions

Using closed-form solution techniques, models were developed for assessing the thermal and structural response of light water reactor (LWR) vessels and penetrations during severe accident conditions. Results from models are displayed as failure maps, generally developed in terms of non-dimensional groups, so that a broader range of reactor design parameters and severe accident conditions can be considered. In this paper, failure maps are used to compare LWR vessel response to three accident conditions. Results discussed within this paper illustrate the importance of vessel and tube geometrical parameters and material properties for predicting which vessel failure mode occurs first.     

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.     

A use of prototypic material for the investigation of severe accident progression

An experimental research platform using corium melts is established for the understanding of safety related important phenomena during a severe accident progression. The research platform includes TROI facility for corium water interaction experiments and VESTA facility for corium-structural material interaction experiments. A cold crucible technology is adapted and improved for a generation of 5–100 kg of corium melts at various compositions. TROI facility is used for experiments to investigate premixing and explosion behaviors during a fuel coolant interaction process. More than 70 experiments using corium at various compositions were performed to simulate steam explosion phenomena in a reactor situation. The results indicate that the conversion efficiency of steam explosion for corium is less than 1%. VESTA facility is used to investigate molten corium-structural material interaction phenomena. VESTA facility consists of two cold crucibles. One crucible is used for the melting of charged material and pouring of corium melt. The other crucible is used for the corium-structural material interaction while providing an induction heating to simulate the decay heat. The results of an experiment on the interaction between corium melt and a specimen made of Inconel performed in the VESTA facility is reported.     

The EPR overall approach for severe accident mitigation

The EPR design includes provisions to cope with severe accidents including core melt situations:

- Situations that would lead to large early releases such as containment bypass, high reactivity accidents, high-pressure core melt or global hydrogen detonation have to be prevented.

- All other situations, including low pressure core melt have to be mitigated in such a way that the corresponding radiological consequences would necessitate only very limited protective countermeasures in a relatively small area and for a limited time for the population living in the neighborhood of the power plant. This means that there would be no need for permanent relocation, no need for evacuation outside the immediate vicinity of the plant, limited sheltering and no long-term restrictions in food consumption.

To reach this objective, which is one of the Safety Authority's requirements, the EPR relies on a very robust containment and on various design measures intended to withstand extreme loads caused by a large range of internal events and external hazards. The deterministic method is used for the EPR safety demonstration, supplemented by probabilistic methods and appropriate R&D work.This paper outlines the major mitigating design features, summarizes the results of the Level 2 PSA and gives the main results of the evaluation of the radiological consequences of core melt on the environment.     

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.     

Deterministic accident analysis for RBMK

Within the framework of an European Commission sponsored activity, an assessment of the deterministic safety technology of the ‘post-Chernobyl modernized’ Reactor Bolshoy Moshchnosty Kipyashiy (RBMK) has been completed. The accident analysis, limited to the area of Design Basis Accident, constituted the key subject for the study; events not including the primary circuit were not considered, as well as events originated from plant status different from the nominal operating conditions. Therefore, the notorious Chernobyl Unit 4 event was outside the scope of the investigation.Following the evaluation of the current state of the art in the area including the identification of critical issues, targets for the analysis were established together with suitable chains of computational tools. The outcomes from this part of the study are (a) the list of transient scenarios whose parameter values are assumed to constitute the boundaries for the evolution of any relevant safety transient and; (b) a set of computational tools with characteristics consistent with current technological achievements, suitable for performing safety analyses.The availability of computational tools, including codes, nodalisations and boundary and initial conditions for the Smolensk 3 NPP, brought to their application to the prediction of the selected transient evolutions that, however, are not classified as licensing studies. The results demonstrated proper safety margins and relatively long time constants associated with the huge values for the ratios between mass of moderator and mass of coolant and unit generated power.The results at the item above, suggested a qualitative, though non rigorous, comparison between accident analysis aspects in LWR and RBMK having the main purpose to show strengths in RBMK safety features heavily criticized not always in a consistent way following the Chernobyl event.The results of supporting analyses for the present paper are discussed in five companions papers in this Journal volume. The second (over six) and the third paper deal with the RBMK Main Coolant Circuit and Confinement thermal-hydraulic performance, respectively. Key specific issues in the RBMK safety technology, constituted by addressing of the “Multiple Pressure Tube Rupture (MPTR)” and by the application of coupled three-dimensional neutron-kinetics thermal-hydraulics, are discussed in the fourth and fifth papers. The proposal to instrument the core channels (ICM = Individual Channel Monitoring) has been formulated in this context and is discussed in the sixth companion paper.     

Design considerations for concrete containments under severe accident loads

Many existing containments in the United States have been shown to accommodate credible severe accident loads. Future containments should be explicitly designed for severe accident loads to reduce the uncertainty associated with the response of containments to these low-probability events. This paper examines the experiences from the application of current structural design codes for concrete containments, ultimate pressure capacity evaluation of existing containments, and pressure fragility testing of scale model concrete containments to arrive at the directions for modification of national codes. Recommendations are provided to consider the severe accidents directly in the concrete containment design.     

Consideration of decontamination model for severe accident consequence assessment

Cost per severe accident, which can cover various kinds of consequences, namely health effects, economic, social and environmental impacts, has been used as an index for severe accident consequence assessment in the authors' previous studies. Decontamination of the contaminated area was concluded influential to the cost per severe accident, though it employed simple and conservative assumptions, which made further discussion difficult. A new decontamination model is formed to address this issue. A sensitivity analysis using the elementary effects method is performed to identify parameters which have large influence on the cost per severe accident. We identify 25 important parameters, and fix most negligible parameters to their medians to form a simplified decontamination model. Calculations of the cost per severe accident with the simplified model and the full model are performed and compared. The differences of the results of the two model are not significant, which ensure the validity of the simplified model. The cost per severe accident calculated by the simplified model is compared with the previous study. The decontamination cost increases its importance significantly. It is therefore necessary to pay attention to decontamination-related issues, e.g., determination of decontamination target area and decontamination waste management scheme.     

A fast and simple iteration scheme for the temperature calculation in a fuel rod

A new and simple iteration scheme for solving the nonlinear heat conduction in a fuel rod is developed. The method is based on a Newton-Raphson iteration scheme and takes two distinctive characteristics of the problem into account. The result is a scalar Regula Falsi technique which is superior to a relaxation technique. The new iteration scheme has been applied successfully over 2 years in the TRANSURANUS code, analysing several hundred complicated data cases under normal, off-normal and accident conditions. Convergence problems never occurred and the computational costs were significantly reduced.     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。