Research of Screening and Grading Method of Ageing Sensitive Objectives in Nuclear Power Plant

2014 ◽  
Vol 543-547 ◽  
pp. 858-861
Author(s):  
Xiao Tian Liu ◽  
Yong Wang ◽  
Shao Rui Niu ◽  
Yan Zhao Zhang ◽  
Zhen Hao Shi ◽  
...  
Keyword(s):  
Fuzzy Logic ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Nonlinear Coupling ◽  
Highly Nonlinear ◽  
Ageing Management ◽  
First Time

This first step of ageing management in nuclear power plant is to determine the objectives and their priorities. The characteristics of the objectives are complex and highly nonlinear coupling. A fuzzy logic based screening and grading method have been developed in this research for the first time which combined the genetic ageing lessons learned and field expert experience to resolve the problem. The method have been approved of highly applicability and applied to ageing management in multiple nuclear power plants.

scholarly journals Overview of NRC NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves, and Inservice Examination and Testing of Dynamic Restrains (Snubbers) at Nuclear Power Plants

2014 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Regulatory Guidance ◽  
American Society ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) staff issued Revision 2 to NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plant,” to assist the nuclear power plant licensees in establishing a basic understanding of the regulatory basis for pump and valve inservice testing (IST) programs and dynamic restraints (snubbers) inservice examination and testing programs. Since the Revision 1 issuance of NUREG-1482, certain tests and measurements required by earlier editions and addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) have been clarified, updated, revised or eliminated. The revision to NUREG-1482 incorporates and addresses those changes, and includes the IST programs guidelines related to new reactors. The revised guidance incorporates lessons learned and experience gained since the last issue. This paper provides an overview of the contents of the NUREG-1482 and those changes and discusses how they affect NRC guidance on implementing pump and valve inservice testing (IST) programs. For the first time, this revision added dynamic restraint (snubber) inservice examination and testing program guidelines along with pump and valve IST programs. This paper highlights important changes to NUREG-1482, but is not intended to provide a complete record of all changes to the document. The NRC intends to continue to develop and improve its guidance on IST methods through active participation in the ASME OM Code consensus process, interactions with various technical organizations, user groups, and through periodic updates of NRC-published guidance and issuance of generic communications as the need arises. Revision 2 to NUREG-1482 incorporates regulatory guidance applicable to the 2004 Edition including 2005 and 2006 Addenda to the ASME OM Code. Revision 0 and Revision 1 to NUREG-1482 are still valid and may continue to be used by those licensees who have not been required to update their IST program to the 2004 Edition including the 2005 and 2006 Addenda (or later Edition) of the ASME OM Code. The guidance provided in many sections herein may be used for requesting relief from or alternatives to ASME OM Code requirements. However, licensees may also request relief or authorization of an alternative that is not in conformance with the guidance. In evaluating such requested relief or alternatives, the NRC uses the guidelines/recommendations of the NUREG, where applicable. The guidelines and recommendations provided in this NUREG and its Appendix A do not supersede the regulatory requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 50.55a, “Codes and standards”. Further, this NUREG does not authorize the use of alternatives to, grant relief from, the ASME OM Code requirements for inservice testing of pumps and valves, or inservice examination and testing of dynamic restraints (snubbers), incorporated by reference in 10 CFR 50.55a. Paper published with permission.

The IMS Development Plan in Construction Phase for Nuclear Power Plants

2010 ◽  
Author(s):  
Kyung Hwa Chung ◽  
Jae Hee Han
Keyword(s):  
Information Technology ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Construction Projects ◽  
Nuclear Power ◽  
Power Plants ◽  
Information Management System ◽  
Lessons Learned ◽  
3D Cad ◽  
Plant Construction

The implementation of advanced information technology for nuclear power plant construction industry in the Republic of Korea is a challenging task to improve the competitive edge for APR1400 across the global nuclear business market. This paper will briefly describe the phases of development of an information management system (hereinafter IMS) in nuclear power construction during the last 40 years in Korea. Additional development task which strengthen the IMS capability will be described from our experience. It will analyze the IMS development and implementation stages for Korean nuclear power plant construction projects such as YGN 3&4, UCN 3&4, YGN 5&6 and UCN 5&6, and the successful application case of the UCN 5&6 Radwaste Building in which 3D CAD technology was implemented for the first time. By reviewing lessons learned, this paper will define the Information Technology advancements resulting the reduction of project costs and construction schedule both by project execution procedures and IT systems including 3D CAD application. The future plan will include integrating project management systems based on data-centric approach and handover strategy for better O&M phase through configuration management technology. In this report, we address the functions to be developed and added in the new IMS.

Fuzzy multi-objective decision making approach for nuclear power plant installation

2020 ◽  
Vol 39 (5) ◽  
pp. 6339-6350
Author(s):  
Esra Çakır ◽  
Ziya Ulukan
Keyword(s):  
Linear Programming ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Energy Supply ◽  
Energy Demand ◽  
Nuclear Power ◽  
Power Plants ◽  
Total Duration ◽  
Objective Functions ◽  
Multi Objective

Due to the increase in energy demand, many countries suffer from energy poverty because of insufficient and expensive energy supply. Plans to use alternative power like nuclear power for electricity generation are being revived among developing countries. Decisions for installation of power plants need to be based on careful assessment of future energy supply and demand, economic and financial implications and requirements for technology transfer. Since the problem involves many vague parameters, a fuzzy model should be an appropriate approach for dealing with this problem. This study develops a Fuzzy Multi-Objective Linear Programming (FMOLP) model for solving the nuclear power plant installation problem in fuzzy environment. FMOLP approach is recommended for cases where the objective functions are imprecise and can only be stated within a certain threshold level. The proposed model attempts to minimize total duration time, total cost and maximize the total crash time of the installation project. By using FMOLP, the weighted additive technique can also be applied in order to transform the model into Fuzzy Multiple Weighted-Objective Linear Programming (FMWOLP) to control the objective values such that all decision makers target on each criterion can be met. The optimum solution with the achievement level for both of the models (FMOLP and FMWOLP) are compared with each other. FMWOLP results in better performance as the overall degree of satisfaction depends on the weight given to the objective functions. A numerical example demonstrates the feasibility of applying the proposed models to nuclear power plant installation problem.

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Vanderley Vasconcelos ◽  
Wellington Antonio Soares ◽  
Raissa Oliveira Marques ◽  
Silvério Ferreira Silva Jr ◽  
Amanda Laureano Raso
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Human Factors ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Cooling System ◽  
Human Reliability ◽  
Non Destructive ◽  
Plant Components

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. This inspection is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI is reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components, such as FMEA (Failure Modes and Effects Analysis) and THERP (Technique for Human Error Rate Prediction). An example by using qualitative and quantitative assessesments with these two techniques to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues, is presented.

scholarly journals Python TensorFlow Big Data Analysis for the Security of Korean Nuclear Power Plants

Electronics ◽  
2020 ◽  
Vol 9 (9) ◽  
pp. 1467
Author(s):  
Sangdo Lee ◽  
Jun-Ho Huh ◽  
Yonghoon Kim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Network Traffic ◽  
Nuclear Power ◽  
Power Plants ◽  
Security Requirements ◽  
Republic Of Korea ◽  
Continuous Data ◽  
Cyber Attack ◽  
The Republic

The Republic of Korea also suffered direct and indirect damages from the Fukushima nuclear accident in Japan and realized the significance of security due to the cyber-threat to the Republic of Korea Hydro and Nuclear Power Co., Ltd. With such matters in mind, this study sought to suggest a measure for improving security in the nuclear power plant. Based on overseas cyber-attack cases and attacking scenario on the control facility of the nuclear power plant, the study designed and proposed a nuclear power plant control network traffic analysis system that satisfies the security requirements and in-depth defense strategy. To enhance the security of the nuclear power plant, the study collected data such as internet provided to the control facilities, network traffic of intranet, and security equipment events and compared and verified them with machine learning analysis. After measuring the accuracy and time, the study proposed the most suitable analysis algorithm for the power plant in order to realize power plant security that facilitates real-time detection and response in the event of a cyber-attack. In this paper, we learned how to apply data for multiple servers and apply various security information as data in the security application using logs, and match with regard to application of character data such as file names. We improved by applying gender, and we converted to continuous data by resetting based on the risk of non-continuous data, and two optimization algorithms were applied to solve the problem of overfitting. Therefore, we think that there will be a contribution in the connection experiment of the data decision part and the optimization algorithm to learn the security data.

Transition to Online Cable Insulation Condition Monitoring

2021 ◽  
Author(s):  
S. W. Glass ◽  
Leonard S. Fifield ◽  
Mychal P. Spencer
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Management Program ◽  
Capacitance Measurement ◽  
Test Bed ◽  
Test Environment ◽  
Specific Test ◽  
Motor Test

Abstract Nuclear power plant cables were originally qualified for 40 year life and generally have not required specific test verification to assure service availability through the initial plant qualification period. However, license renewals to 60 and 80 years of operation require a cable aging management program that depends on some form of test and verification to assure fitness for service. Environmental stress (temperature, radiation, chemicals, water, and mechanical) varies dramatically within a nuclear power plant and, in some cases, cables have degraded and required repair or replacement before their qualified end-of-life period. In other cases, cable conditions have been mild and dependable cable performance confirmed to extend well beyond the initial qualified life. Most offline performance-based testing requires cables to be decoupled and de-energized for specially trained technicians to perform testing. These offline tests constitute an expensive operational burden that limits the economic viability of nuclear power plants. Although initial investment may be higher, new online test practices are emerging as options or complements to offline testing that avoid or minimize the regularly scheduled offline test burden. These online methods include electrical and fiber-optic partial discharge measurement, spread spectrum time or frequency domain reflectometry, distributed temperature profile measurements, and local interdigital capacitance measurement of insulation characteristics. Introduction of these methods must be supported by research to confirm efficacy plus either publicly financed or market driven investment to support the start-up expense of cost-effective instrumentation to monitor cable condition and assure reliable operation. This work summarizes various online cable assessment technologies plus introduces a new cable motor test bed to assess some of these technologies in a controlled test environment.

Research on Earthquake Acceleration Alarm of Nuclear Power Plant

2021 ◽  
Author(s):  
Li Liang ◽  
Pan Rong ◽  
Ren Guopeng ◽  
Zhu Xiuyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
False Alarm ◽  
Ground Motion ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Time History ◽  
Seismic Acceleration ◽  
Trigger Signal

Abstract Almost all nuclear power plants in the world are equipped with seismic instrument system, especially the third generation nuclear power plants in China. When the ground motion measured by four time history accelerometers of containment foundation exceeds the preset threshold, the automatic shutdown trigger signal will be generated. However, from the seismic acceleration characteristics, isolated and prominent single high frequency will be generated the acceleration peak, which has no decisive effect on the seismic response, may cause false alarm, which has a certain impact on the smooth operation of nuclear power plant. According to the principle of three elements of ground motion, this paper puts forward a method that first selects the filtering frequency band which accords with the structural characteristics of nuclear power plants, then synthesizes the three axial acceleration time history, and finally selects the appropriate acceleration peak value for threshold alarm. The results show that the seismic acceleration results obtained by this method can well represent the actual magnitude of acceleration, and can solve the problem of false alarm due to the randomness of single seismic wave, and can be used for automatic reactor shutdown trigger signal of seismic acceleration.

A Quantitative Approach for Reliability Evaluation of Safety I&C Systems in Nuclear Power Plants

2017 ◽  
Author(s):  
Liu Dongxu ◽  
Xu Dongling ◽  
Zhang Shuhui ◽  
Hu Xiaoying
Keyword(s):  
Probability Theory ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Boolean Algebra ◽  
Nuclear Power ◽  
Power Plants ◽  
Block Diagram ◽  
Reliability Assessment ◽  
Simple Approach ◽  
Assessment Approach

The probability that the safety I&C system fails to actuate or advertently actuates RT or ESF functions, in part, essentially determines whether a nuclear power plant could operate safely and efficiently. Since more conservative assumptions and simplifications are introduced during the analysis, this paper achieves solid results by performing the modeling and calculation based on a relatively simple approach, the reliability block diagram (RBD) method. A typical safety I&C platform structure is involved in the model presented in this paper. From the perspective of conservation and simplicity, some assumptions are adopted in this paper. A group of formulas is derived in this paper based on Boolean algebra, probability theory, basic reliability concepts and equations, to facilitate the calculations of probabilities that the safety I&C system fails to actuate or advertently actuates RT or ESF functions. All the inputs of the analysis and calculation in this paper, which includes the I&C platform structure, the constitution of the hardware modules, and reliability data, are referenced to the nuclear power plant universal database where applicable. Although the conclusion drawn in the paper doesn’t apply to the I&C platform assessment for a specific plant, the method of modeling and process of analysis provides an illustration of an alternative quantitative reliability assessment approach for a typical safety I&C system installed in the nuclear power plant.

scholarly journals Efficiency of Using the Auxiliary Boiler for Peak Power Generation by a Steam Turbine of NPP

2021 ◽  
Vol 64 (1) ◽  
pp. 78-90
Author(s):  
V. A. Khrustalev ◽  
M. V. Garievskii
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Normal Operation ◽  
Heat Output ◽  
Generation Process ◽  
System Effect ◽  
Industrial Enterprises

The article presents the technique of an estimation of efficiency of use of potential heat output of an auxiliary boiler (AB) to improve electric capacity and manoeuvrability of a steam turbine unit of a power unit of a nuclear power plant (NPP) equipped with a water-cooled water-moderated power reactor (WWER). An analysis of the technical characteristics of the AB of Balakovo NPP (of Saratov oblast) was carried out and hydrocarbon deposits near the NPP were determined. It is shown that in WWER nuclear power plants in Russia, auxiliary boilers are mainly used only until the normal operation after start-up whereas auxiliary boiler equipment is maintained in cold standby mode and does not participate in the generation process at power plants. The results of research aimed to improve the systems of regulation and power management of power units; general principles of increasing the efficiency of production, transmission and distribution of electric energy, as well as the issues of attracting the potential of energy technology sources of industrial enterprises to provide load schedules have been analyzed. The possibility of using the power complex NPP and the AB as a single object of regulation is substantiated. The authors’ priority scheme-parametric developments on the possibility of using the thermal power of the auxiliary boilers to increase the power of the steam turbine of a nuclear power plant unit equipped with WWER reactors unit during peak periods, as well as the enthalpy balance method for calculating heat flows, were applied. The surface area of the additional heater of the regeneration “deaerator – high pressure heaters” system and its cost were calculated. On the basis of calculations, it was shown that the additional power that can be obtained in the steam turbine of the NPP with a capacity of 1200 MW due to the use of heat of the modernized auxiliary boiler in the additional heat exchanger is 40.5 MW. The additional costs for the implementation of the heat recovery scheme of the auxiliary boiler at different prices for gas fuel and the resulting system effect were estimated in an enlarged way. Calculations have shown the acceptability of the payback period of the proposed modernization.

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