Safety Systems Of Power Plants Research Articles

Verification on application program generation and loading for safety systems of nuclear power plants based on the reverse engineering method

The article describes an automated verification method used for application software of control safety systems based on the TPTS-SB equipment. Verification is performed by comparing two mathematical models (oriented graphs): one obtained by processing the original design data, i.e., graphical functional diagrams, and the other formed by reversing the program code loaded from the controller. The vertices in both graphs are functional blocks of mathematical and logical operations; the edges are connections between them. The constructed mathematical models undergo a comparison, covering the vertices and edges of the graphs as well as the memory cells and values of constants. The equivalence of mathematical models proves the correspondence between the program code and the initial set of design functional diagrams. The proposed automated verification method makes it possible to prove that no distortion is introduced into the program during the process of converting graphical functional diagrams into the program code with its subsequent translation and loading into the controller. It is postulated that any distortions will be detected during the verification procedure, which is performed every time after loading the code into the controller. The solution provides an acceptable speed when large volumes of vector graphics stored in a relational database are processed, and makes it possible to visualize the verification results. The proposed method is implemented in the GET-R1 instrumentation tools for TPTS-SB and is used in designing and verifying the application software of the safety systems at the Belarusian NPP.

Верификация процедур генерации и загрузки прикладных программ систем безопасности атомных электростанций методом обратного преобразования

Верификация процедур генерации и загрузки прикладных программ систем безопасности атомных электростанций методом обратного преобразования

Prediction of the moderator temperature field in a heavy water reactor based on a cellular neural network

Reactors with heavy water coolants and moderators have been used extensively in today's power industry. Monitoring of the moderator condition plays an important role in ensuring normal operation of a power plant. A cellular neural network, the architecture of which has been adapted for hardware implementation, is proposed for use in a system for prediction of the heavy water moderator temperature. A reactor model composed in accordance with the CANDU Darlington heavy water reactor design was used to form the training sample collection and to control correct operation of the neural network structure. The sample components for the adjustment and configuration of the network topology include key parameters that characterize the energy generation process in the core. The paper considers the feasibility of the temperature prediction only for the calandria's central cross-section. To solve this problem, the cellular neural network architecture has been designed, and major parts of the digital computational element and methods for their implementation based on an FPLD have also been developed. The method is described for organizing an optical coupling between individual neural modules within the network, which enables not only the restructuring of the topology in the training process, but also the assignment of priorities for the propagation of the information signals of neurons depending on the activity in a situation analysis at the neural network structure inlet. Asynchronous activation of cells was used based on an oscillating fractal network, the basis for which was a modified ring oscillator. The efficiency of training the proposed architecture using stochastic diffusion search algorithms is evaluated. A comparative analysis of the model behavior and the results of the neural network operation have shown that the use of the neural network approach is effective in safety systems of power plants.

The modelling of condensation in horizontal tubes and the comparison with experimental data

The condensation in horizontal tubes plays an important role in determining the operation mode of passive safety systems of modern nuclear power plants. In this paper, two different approaches for modelling of this phenomenon are compared and verified against experimental data. The first approach is based on the flow regime map developed by Tandon. Depending on the regime, the heat transfer coefficient is calculated according to corresponding semi-empirical correlation. The second approach uses a general, fully empirical correlation proposed by Shah. Both models are developed with utilization of the object-oriented, equation-based Modelica language and the open-source Open-Modelica environment. The results are compared with data obtained during a large scale integral test, simulating a Loss of Coolant Accident scenario performed at the dedicated Integral Test Facility Karlstein (INKA) which was built at the Components Testing Department of AREVA in Karlstein, Germany. The INKA facility was designed to test the performance of the passive safety systems of KERENA, the new AREVA boiling water reactor design. INKA represents the KERENA containment with a volume scaling of 1:24. Components heights and levels over the ground are in the full scale. The comparison of simulations results shows a good agreement.

Development of Passive In-Core Cooling System for Nuclear Safety Using Hybrid Heat Pipe

The importance of passive safety for commercial nuclear power plants has been emphasized after the nuclear accidents that occurred at Three Mile Island and Fukushima. A combination of unexpected human errors, severe natural disasters, and defects of system designs led to the accidents, thereby highlighting the vulnerability of established safety systems of commercial nuclear power plants. Various passive safety systems are under development to mitigate design-basis accidents. However, several uncertainties and problems have been pointed out. As a solution to the problems, this paper proposes various designs for a passive in-core cooling system (PINCs) based on hybrid heat pipes. The feasibility and coolability of the PINCs as a passive safety system for commercial pressurized water reactors was investigated using experimental works and numerical analyses. The PINCs showed sufficient coolability to mitigate station blackout conditions by delaying core uncovery. Additionally, several PINCs concepts for advanced nuclear power plants such as a small modular reactor and Generation IV reactors are suggested.

Determination of elemental impurities in polymer materials of electrical cables of safety systems of nuclear power plants by k 0-INAA

Instrumental neutron activation analysis with k0 standardization (k0-INAA) was employed for elemental characterization of individual components of electrical cables currently used in safety systems of nuclear power plants (NPP). The analysis yielded information about the content of 35 elements. It has been found that the presently used cables would not satisfy requirements of qualification set for the new generation of NPP by “The European Utility Requirements for LWR Nuclear Power Plants”, because content of several elements exceeded the maximal admissible levels. This demonstrates that cable polymer materials for safety systems of the new generation NPP must be synthesized and controlled regarding the content of additives, which are undesired from the hygienic and radiological points of view. k0-INAA proved to be a method of choice for elemental characterization of polymer materials of electrical cables as part of their qualification for use in NPP.

Particle Swarm Optimization of safety components and systems of nuclear power plants under uncertain maintenance planning

Maintenance planning is a subject of concern to many industrial sectors as plant safety and business depend on it, and can be formulated in terms of a multi-objective optimization problem where reliability, availability, maintainability and cost act as decision criteria and surveillance test and maintenance strategies act as decision variables. Usually, the frequency of performing a maintenance task is considered, in the optimization process, as a constant value but a certain range of variation from such value is observed in real practice. Thus, to obtain a more realistic approach, a certain degree of uncertainty should be considered in the decision variables. This paper presents two examples of maintenance optimization using Particle Swarm as optimization technique and a tolerance interval based approach to address uncertainty, one is focused on a safety component and the other considers a nuclear power plant safety system.

A level-1 PSA based systemic approach to optimal technical specifications of safety systems in nuclear power plants

Level-1 PSA provides an integrated picture of the plant safety, incorporating all the design features, operational failure data, operational practices and maintenance information. The nuclear power plant operations are governed by the Technical Specifications, which specify Surveillance test frequencies, amongst other limiting condition of operations. With the detailed plant safety assessment having been conducted using the state of the art methodologies of PSA, there is a potential for evaluating the impact of varying the Testing frequencies on the plant risk and cost metrics. However, the complexity of the study is so vast that any optimization within the PSA framework, with change of input variables parametrically, is not feasible. In this paper, a multi-objective approach to optimization of surveillance test frequencies of two systems is studied using genetic algorithms, with inputs taken from the Level-1 PSA study of the 540 MWe Indian Pressurized Heavy Water Reactor.

Validation of the periodicity of checks of nuclear power plant safety systems using the bars system of programs

The application of the BARS program system for choosing the optimal parameters for checking the serviceability of the safety systems of nuclear power plants is described. The types of checks as well as the mathematical models determining the reliability of the components of the system for each type of check are described. Calculations of the failure probability of the safety systems of nuclear power plants which incorporate components with different levels of reliability are presented. The values of the optimal period between serviceability checks for each type of system are determined. An evaluation of the effect of failures of a general type on the reliability of safety systems is presented for two forms of the α-factor model.

An application of risk-informed evaluation on MOVs and AOVs for Taiwan BWR-type nuclear power plants

Implementing a risk-informed inservice testing (RI-IST) program provides a good aspect to the nuclear power plant licensee as an alternating program in the current ASME Section XI and 10 CFR 50.55a relevant testing programs. RI-IST concentrates testing resources on highly significant components, reduces excess testing burden, increases plant's availability, decreases dose rate on the plant's staff and also reduces cost on plant's operation and maintenance under nuclear safety expectations. Furthermore, RI-IST also gives a feature on prospective licensing change basis to a nuclear power plant's licensee. This study will focus on safety-related and PRA-molded motor-operated valves (MOVs) and air-operated valves (AOVs) under the inservice testing program in boiling water reactor (BWR)-type nuclear power plant. As MOVs and AOVs have crucial safety functions throughout the nuclear power plant's safety systems, the steady operation and performance of MOVs and AOVs will definitely ensure that the nuclear power plant operates under safety expectations; therefore, this is the key reason to implement risk-informed evaluation for MOVs and AOVs in this study and being able to provide the safety significance classification for MOVs and AOVs under the current IST program to the plant's management. As a pilot study of RI-IST, the methodology of qualitative assessment will incorporate with probabilistic risk assessment (PRA) analyzing MOVs’ and AOVs’ safety significance within the current PRA model. The evaluating result will then classify its safety significance into a high-safety significant component (HSSC) and a low-safety significant component (LSSC) for MOVs and AOVs based on relevant regulatory criteria. With this initiating achievement, it can provide a cornerstone for further studies on the other types of valves and pumps in RI-IST program and also provide a valuable reference as proposing license change to the licensee.

Nuclear Power Plant Compliance with the Principle of Power-System Indepedence from Safety Systems Operation

The purpose of this paper is to show that during the operation of safety systems at nuclear power plants the principle of independence from the power system, which is one of the basic principles inocrporated in the design of safety systems, is not satisfied and the power system, especially if it is deficient, cannot guarantee the required electricity and protection for safety systems from general failures. To satisfy the independence principle, guarantee the required quality of electricity, and protect the safety systems in nuclear power plants from general failures, it is proposed that the presently operative algorithm for starting up diesel generators be reexamined. When the safety systems at nuclear power plants perform their required functions, they should operate from autonomous diesel generators at the nuclear power plant, which are equipped with electricity quality regulators (frequency and voltage), and not from the power system. It is also suggested that the variant of the algorithm where diesel generators are started up as a preventative measure when the quality of the electricity in the power system drops below admissable limits be reexamined.

Modernizing and Maintaining Instrumentation and Control Systems in Nuclear Power Plants

Deregulation of the electric utilities has made a major impact on nuclear power plants. To be competitive, more emphasis is being put on cost-effective production of electricity with a more critical look at whether a system should be modernized due to obsolescence, reliability, or productivity concerns. Instrumentation and control (I&C) systems play an important role in reducing the cost of producing electricity while maintaining or enhancing safety. Systems that are well designed, reliable, enhance productivity, and are cost-effective to operate and maintain can reduce the overall costs. Modern technology with its ability to better provide and use real-time information offers an effective platform for modernizing systems. At the same time, new technology brings new challenges and issues, especially for safety systems in nuclear power plants. To increase competitiveness, it is important to take advantage of the opportunities offered by modern technology and to address the new challenges and issues in a cost-effective manner. The Electric Power Research Institute (EPRI) and its member utilities have been working together with other members of the nuclear industry since 1990 to address I&C modernization and maintenance issues. The EPRI I&C Program has developed a life-cycle management approach for I&C systems that involves the optimization of maintenance, monitoring, and capital resources to sustain safety and performance throughout the plant life. Strategic planning methodologies and implementation guidelines addressing digital I&C issues in nuclear power plants have been developed. Work is ongoing in diverse areas to support the design, implementation, and operation of new digital systems. Technology transfer is an integral part of this I&C program.

APPLICATION OF DETECTORS BASED ON THE WIDE-GAP SEMICONDUCTORS CdTe, CdZnTe, AND GaAs FOR NUCLEAR POWER PLANT SAFETY MONITORING

Units for detecting ionizing radiation which are based on ionization counters, fission chambers, and scintillation sensors are used in nuclear and radiation safety systems in nuclear power plants. In comparison, semiconductor detectors are smaller and have a small mass, a wider dynamical range, a higher quantum yield, better radiation resistance, and better energy resolution and signal/noise ratio. Investigations confirming these qualities are presented in the paper.

99/00435 A verification and validation method and its application to digital safety systems in ABWR nuclear power plants

99/00435 A verification and validation method and its application to digital safety systems in ABWR nuclear power plants

A verification and validation method and its application to digital safety systems in ABWR nuclear power plants

A verification and validation (V&V) method has been developed and applied to newly developed digital safety systems for the first ABWR plant, Kashiwazaki–Kariwa unit No. 6 of the Tokyo Electric Power Company. This paper describes the method and experience gained from its application. The method was developed on the basis of domestic and foreign standards and guidelines, and covers the more concrete procedures required for actual V&V. The application of problem oriented language (POL) helps make the V&V feasible and reliable. A personal-computer-based automatic test tool for the validation test has been developed and utilized. This tool is used to carry out the pre-defined validation test procedure automatically and produce a test report, and it was found to be effective in reducing the time and manpower required for the validation test. The validation test covers dynamic transient tests in which the response of the digital safety system against the simulated design based transients are tested. The tool enables automatic execution of the dynamic test.

Revised standard addresses design requirements for computers used in nuclear power-plant safety systems

The IEEE Nuclear Engineering Committee has standards that address hardware issues. The IEEE Computer Society has standards addressing software. The P-7-4.3.2, IEEE standard criteria for digital computers in safety systems of nuclear power generating stations, which brings together standards from both groups to address the integration of computer hardware and software with noncomputer hardware, is discussed. P-7-4.3.2 is being developed to show the interrelationship of software and hardware requirements as they apply throughout the development process. The influence of system design development on system quality, system qualification, and system integrity, are described. >