Safeguard assessment for life extension in nuclear power plants (NPPs) using a production function

Life extension is investigated as a safeguard assessment for the stability on the operation of the nuclear power plants (NPPs). The Cobb-Douglas function, one of the production functions, is modified for the nuclear safeguard in NPPs, which was developed for the life quality of the social and natural objects. Nuclear Safeguard Estimator Function (NSEF) is developed for the application in NPPs. The cases of NPPs are compared with each other in the aspect of the secure performance. The results are obtained by the standard productivity comparisons with the designed power operations. The range of secure life extension is between 1.008 and 5.353 in 2000 MW_e and the range is between 0.302 and 0.994 in 600 MW_e. So, the successfulness of the power operation increases about 5 times higher than that of the interested power in this study, which means that the safeguard assessment has been performed in the life extension of the NPPs. The technology assessment (TA) is suggested for the safe operation which is an advanced method comparing conventional probabilistic safety assessment (PSA).     

Passive system reliability in the nuclear power plants (NPPs) using statistical modeling

The probabilistic safety assessment (PSA) has been studied for the very high temperature reactor (VHTR). There is a difficulty to make the quantification of the PSA due to the deficiency of the operation and experience data. So, it is necessary to use the statistical data for the basic event. The physical data of the non-linear fuzzy set algorithm are used to quantify the designed case. The mass flow rate in natural circulation is investigated. In addition, the potential energy in the gravity, the temperature and pressure in the heat conduction, and the heat transfer rate in the internal stored energy are investigated. The values in the probability set and fuzzy set are compared for the failure explanation. The result shows how to use the newly made probability of the failure in the propagations. The failure frequencies, which are made by the GAMMA (GAs Multi-component Mixture Analysis) code, are compared with four failure frequencies by probabilistic and fuzzy methods. The results show that the artificial intelligence analysis of the fuzzy set could improve the reliability method than that of the probabilistic analysis.     

A probabilistic seismic risk assessment procedure for nuclear power plants: (I) Methodology

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology.     

Fuzzy methodology applied to Probabilistic Safety Assessment for digital system in nuclear power plants

A fuzzy inference system (FIS) modeling technique to treat a nuclear reliability engineering problem is presented. Recently, many nuclear power plants (NPPs) have performed a shift in technology to digital systems due to analog obsolescence and digital advantages. The fuzzy inference engine uses these fuzzy IF-THEN rules to determine a mapping of the input universe of discourse over the output universe of discourse based on fuzzy logic principles. The risk priority number (RPN) (typical of a traditional failure mode and effects analysis - FMEA) is calculated and compared to fuzzy risk priority number (FRPN), obtained by the use of the scores from expert opinions. It was adopted the digital feedwater control system as a practical example in the case study. The results demonstrated the potential of the inference system to this class of problem.     

Safety assessment for the ultimate heat sink (UHS) system with non-injection concept in nuclear power plants (NPPs)

Following the Fukushima accident, it is proposed to find a better safety system, which has a pool-type cooling system without coolant injections. Since the conventional piping-based injection systems have failed in treating the three major severe accidents, the artificial pool could be constructed to cover the failed reactor core systems in which the pool-like structure is constructed. Regarding this study, there were some previous studies about the ultimate heat sink (UHS). In this study, the system dynamics (SD) modeling is performed in the case of Fukushima Unit 1 accident. The basic events are obtained by the Boolean values as 0 and 1. The quantifications are obtained by the SD algorithm incorporated with the Vensim software. In the simulations work, there is a plateau region between the 25th and 45th years in the interested period. The nonlinear algorithm is applied for the UHS analysis which was not installed for the commercial use yet.     

Environmental assessment of nuclear power generation in Korea

Nuclear energy is the major source of the electricity generation due to the lack of domestic energy in Korea. In recent years, environmental preservation concern has been increased for more efficient management of nuclear energy. Therefore it is necessary to apply Life Cycle Assessment (LCA) methodology including environmental management system, waste reduction schemes and environmental analysis methods. In this study, LCA methodology is introduced to the nuclear power generation system and environmental burden caused by this system is assessed. This study suggests new methodology for environmental assessment and establishes the extensive infra-database related with nuclear power generation system. Also, it is possible to improve the scientific basis of LCA with the emphasis on the nuclear power generation system. Therefore, this study consists of general framework of LCA such as goal and scope, inventory analysis, impact assessment, and valuation of the nuclear power generation system. The major objectives are to identify the environmental impact and the environmentally most dominant stage in life cycle of nuclear power generation system and to suggest the new methodology to solve the problem when LCA is applied to facility releasing the radioactive wastes. Thus this study is useful to improve the environmental impact assessment of nuclear power generation system and to promote the methodology of LCA.     

Research-reactor neutron fields for calibrating in-reactor detectors in nuclear power plants

The neutron spectra of research and power reactors are compared. The spectra were measured by the neutron-activation method and calculated using the KASKAD computer code. The a priori spectrum in the calculation was constructed as a superposition of physically validated spectra. A method of calibrating in-reactor detectors in nuclear power plants on the basis of the sensitivity to the ²³⁵U fission rate in 1 g of uranium using the neutron fields of research reactors is proposed. __________ Translated from Atomnaya énergiya, Vol. 100, No. 2, pp. 97–107, February, 2006.     

Probabilistic assessment of seismic risk for nuclear power plants

Probabilistic approaches to the design, siting, and safety analysis of nuclear power plants have been proposed by Farmer, Wall, and Garrick. Farmer and Wall established a limit line which delineates between acceptable and unacceptable risks. To implement the method, all accidental chains are systematically analyzed to determine their probability and associated fission product release magnitude; the combination is compared to the limit line. For proper implementation, the seismic risk should be evaluated in a quantified manner. Conceptually, this evaluation is made in two stages: the probability of an earthquake occurrence as a function of its intensity and, given a seismic intensity, the conditional probability of damage. This paper reports on an initial study into the latter aspect.The effect of uncertainty in several parameters which determine the response of a nuclear reactor building to earthquake forces is assessed. Probability distributions for material properties were determined from site measurements and these distributions were utilized for determining the building response and the damage criterion. A subjective probability density function for damping was assigned from the available information and the judgment of experienced engineers. Four accelerograms, El Centro N---S 1940, and three artificial earthquakes were used to represent the variability in the forcing functions. The uncertainty in the model idealization was assessed by evaluating three alternate models. A versatile computer program was developed to compute the response of the mathematical model to the forcing functions using matrix formulation and modal method of analysis. An exact solution, rather than numerical integration, was used to obtain the dynamic response of the system in generalized coordinates.The stresses within the reactor building are similar for different earthquakes considered in this study when they are normalized to ground acceleration, indicating that the shape of the accelerogram and its frequency content are less significant than the magnitude of the maximum ground acceleration for the reactor building. The variation in modulus of elasticity for concrete had a significant effect on the building response. Damping, in general, reduced the response, but in cases where the duration of an earthquake is short the effect was not very significant.A simple failure criteria for ultimate shear stress in shear walls, τ_ult = 4.75 √ƒ′_c, where ƒ′_c is the ultimate compressive strength of concrete, is used to estimate the initiation of cracking in the walls. The normal design of the reactor building is deterministic and is based on a 0.2 g design basis earthquake. Using the results obtained by the parametric study on the variation of response, the probability of damage was estimated by a Monte Carlo analysis. It was estimated that, given the occurrence of a design basis earthquake, there is less than approximately 10⁻³ probability of cracking in the shear walls of the reactor building. The initiation of cracking in the concrete should not lead to a significant release of contained fission products.     

核动力装置运行支持系统的设计与开发

利用实时嵌入式操作系统VxWorks开发了核动力装置运行支持系统(NPPOSS),该系统可以利用计算机技术的优势给操纵员提供装置运行状态监测信息、警报分析信息、故障诊断信息以及应急操作规程等。本文对核动力装置运行支持系统进行了设计和开发,并以蒸汽发生器传热管破裂为例,在全范围仿真机上对该系统的可行性进行实验验证。结果表明,采用NPPOSS可以减轻操纵员的工作负担和精神压力,提高核动力装置的安全性。     

Safeguard management for operation security in nuclear power plants (NPPs)

Safeguard modeling is conducted for the successful operations in the nuclear power plants (NPPs). The characteristics of the secure operation in NPPs are investigated using the network effect method which is quantified by the Monte-Carlo algorithm. Fundamentally, it is impossible to predict the exact time of a terror incident. So, the random sampling for the event frequency is a reasonable method, including the characteristics of network effect method such as the zero-sum quantification. The performance of operation with safeguard is the major concern of this study. There are three kinds of considerations as the neutronics, thermo-hydraulics, and safeguard properties which are organized as an aspect of safeguard considerations. The result, therefore, can give the stability of the operations when the power is decided. The maximum value of secure operation is 12.0 in the third month and the minimum value is 1.0 in the 18th and 54th months, in a 10 years period. Thus, the stability of the secure power operation increases 12 times higher than the lowest value according to this study. This means that the secure operation is changeable in the designed NPPs and the dynamical situation of the secure operation can be shown to the operator.     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.     

A study on physicochemical properties of epoxy coating system for nuclear power plants

In this work, the thermal properties of epoxy coating system on the liner plate in the containment structure have been investigated by irradiation dose rate and design basis accident (DBA) conditions. Also, the effect of immersion in hot water on adhesion strengths of epoxy coating system has been studied. The glass transition temperature (T_g) and thermal stability of the epoxy coating system after DBA tests were measured by differential scanning calorimeter (DSC) and thermogravimetric analyser (TGA) analyses, respectively. Contact angle measurements were used to determine the effect of immersion on the surface energetics of epoxy coating system, including surface free energy and work of adhesion. Adhesion tests were also executed to evaluate the adhesion strength at interfaces between carbon steel plate and epoxy resins. As a result, the DBA test led to the improvement of the internal structure in cured epoxy systems, resulting in significantly increasing the thermal stability, as well as the T_g. Also, the immersion in hot water had a role in the post curing of epoxy resins and increased the mechanical interlocking of the network system, resulting in increasing the adhesion strengths of the epoxy coating system.     

软件仿真核电站堆芯核测量系统的设计与实现

针对堆芯核测量系统设计制造周期长、关键机械设备采购难、控制系统调试困难等问题,设计了仿真堆芯核测量系统。基于RSView32的仿真测量系统仿真电气设备和机械设备的功能,并配备了测量系统的状态监控界面。本文介绍了仿真测量系统的原理和实现方法。通过仿真测量系统与真实控制软件的联调试验,证实在系统设计过程中,仿真测量系统完全可以代替电气和机械设备,辅助测量系统的控制程序进行调试运行。该仿真测量系统的使用可显著减少机械磨损,缩短控制系统的调试周期。     

A dose constraint proposal for members of the public taking into account multi-unit nuclear power plants in Korea

In the 2007 recommendations, the International Commission on Radiological Protection (ICRP) changed from a process-based system of practices and intervention to a system based on the characteristics of the radiation exposure situation. In addition, the ICRP now recommends the application of source-related dose constraints under a planned exposure situation as a tool for the optimization of measures to protect the workers and members of the public. In this study, an analysis of radioactive effluents from Korean nuclear power plants and a public dose assessment were conducted using these source-related dose constraints. As a result, this analysis suggests appropriate dose constraints for members of the public taking into account the operation of multi-unit nuclear reactors at a single site in Korea.     

An earthquake isolation system for steam generators in nuclear power plants

This paper describes an experimental/analytical study of the effectiveness of base isolation and damped interaction between a model of a steam generator and its primary housing structure in a nuclear power plant subjected to earthquake ground motion. The design of the test generator model, its connection to the primary structure by yielding elements and the influence of such yielding restrainers on the response of the generator are included. Details of an optimal design problem for selection of the “best” combination of isolation and energy absorption devices are presented and their effectiveness demonstrated.