Qualification of Pumps and Valves for the Safety Injection Path of Nuclear Power Plants

2014 ◽  
Author(s):  
Ingo Ganzmann ◽  
Holger Schmidt
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Factor ◽  
International Organization ◽  
Insulation Material ◽  
Hydraulic Test ◽  
Iso 17025 ◽  
Operation And Maintenance ◽  
Related Plant ◽  
Construction Operation

The reliability of a nuclear power plant depends on the safe functioning of its components during its lifetime: from design through construction, operation and maintenance. This is valid for new build projects as well as for the current fleet. As plants undergo modifications for increased performance or extended lifetimes, component integrity becomes a critical factor in those efforts, particularly for safety-related plant functions. This paper focuses on the qualification of pumps and valves of the safety-injection path, considering new requirements. Going back to the Barsebäck event in the year 1992, it is known that insulation material may cause clogging. Consequently, the presence of debris material in the water may have an impact on the functioning of pumps and valves. For this purpose, AREVA has built new thermo-hydraulic test loops in its accredited test and inspection body (according to International Organization for Standardization (ISO) 17025 and 17020) to consider this effect as it relates to components qualification (Ref. 1). The main relevant aspects of these tests will be discussed together with corresponding thermal shock tests. Paper published with permission.

The application efficiency of intumescent coatings for power cables of safety systems of nuclear power plants under fire conditions

2021 ◽  
Vol 30 (4) ◽  
pp. 36-47
Author(s):  
O. S. Lebedchenko ◽  
S. V. Puzach ◽  
V. I. Zykov
Keyword(s):  
Outer Surface ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fire Retardant ◽  
Insulation Material ◽  
Power Cables ◽  
Fire Retardants ◽  
Safety Systems ◽  
Fire Conditions

Introduction. The reliable operation of safety systems, that allows for the failure of no more than one safety system component, entails the safe shutdown and cool-down of an NPP reactor in the event of fire. However, the co-authors have not assessed the loss of performance by an insulating material, treated by intumescent compositions and used in the power cables of the above safety systems exposed to the simultaneous effect of various modes of fire and current loads.Goals and objectives. The purpose of the article is the theoretical assessment of the application efficiency of intumescent fire-retardant coatings in power cables used in the safety systems of nuclear power plants having water-cooled and water-moderated reactors under fire conditions. To achieve this goal, the temperature of the outer surface of the insulation and the intumescent fire-retardant coating was analyzed depending on the mode of fire. Theoretical foundations. A non-stationary one-dimensional heat transfer equation is solved to identify the temperature distribution inside the multilayered insulation and the fire-protection layer of a conductive core.Results and their discussion. The co-authors have identified dependences between the temperature of the outer surface of the insulation and the fire retarding composition of the three-core cable VVGng (A)-LS 3x2.5-0.66, on the one hand, and the temperature of the indoor gas environment for three standard modes of fire and one real fire mode. It is found that before the initiation of the process of destruction of the insulation material, the intumescence of the fire-retardant coating occurs only in case of a hydrocarbon fire. Under real fire conditions, the maximal insulation melting time before the initiation of intumescence of the fire-retardant coating at the minimal temperature of intumescence is 4.75 minutes, while the maximal time period from the initiation of destruction of the insulation material to the moment of the insulation melting is 6.0 minutes.Conclusions. An experimental or theoretical substantiation of parameters of intumescent fire retardants, performed using standard modes of fire, has proven the potential loss of operational properties by insulating materials of power cables, used in the safety systems of nuclear power plants, in case of a real fire. Therefore, it is necessary to establish a scientific rationale for the efficient use of fire retardants in the above cables with regard for the conditions of a real fire.

The Use of the ASME PRA Standard in Pressure Vessel and Piping Work

2002 ◽  
Author(s):  
Richard A. Hill
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Risk Information ◽  
Technical Requirements ◽  
Operation And Maintenance ◽  
Construction Operation ◽  
Standard Sets

After several years of intense labor by many industry people, ASME is about to issue its newly approved PRA standard. This standard is for probabilistic risk assessment (PRA) for nuclear power plant applications. It is not a standard on how to build a PRA model; although, that could be inferred from the standard’s technical requirements. This Standard sets forth requirements for PRAs used to support risk-informed decisions related to design, licensing, procurement, construction, operation, and maintenance. It also prescribes a method for applying these requirements depending the degree to which risk information is needed and credited.

UNCERTAINTY OF HUMAN ERROR AND FUZZY APPROACH TO HUMAN RELIABILITY ANALYSIS

2006 ◽  
Vol 14 (01) ◽  
pp. 111-129 ◽  
Author(s):  
B. J. KIM ◽  
RAM R. BISHU
Keyword(s):  
Reliability Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Critical Factor ◽  
Fuzzy Approach ◽  
Human Reliability ◽  
Human Reliability Analysis ◽  
Human Errors ◽  
Fuzzy Logic Approach

Human error is regarded as a critical factor in catastrophic accidents such as disasters at nuclear power plants, air plane crashes, or derailed trains. Several taxonomies for human errors and methodologies for human reliability analysis (HRA) have been proposed in the literature. Generally, human errors have been modeled on the basis of probabilistic concepts with or without the consideration of cognitive aspects of human behaviors. Modeling of human errors through probabilistic approaches has shown a limitation on quantification of qualitative aspects of human errors and complexity of attributes from circumstances involved. The purpose of this paper is to investigate the methodologies for human reliability analysis and introduce a fuzzy logic approach to the evaluation of human interacting system's reliability. Fuzzy approach could be used to estimate human error effects under ambiguous interacting environments and assist in the design of error free work environments.

scholarly journals Hydrometeorological monitoring in the Nizhny Novgorod NPP siting area

2021 ◽  
Vol 263 ◽  
pp. 05012
Author(s):  
Fedor Bryukhan ◽  
Aleksey Vinogradov ◽  
Ivan Vinogradov
Keyword(s):  
Surface Waters ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Climate Changes ◽  
Environmental Safety ◽  
Data Series ◽  
Observation Data ◽  
Industrial Association ◽  
Construction Operation

Ensuring the technological and environmental safety of nuclear power plants (NPP) involves the collection and analysis of data on the state of the natural environment near nuclear power plants, including the atmosphere and surface waters. To obtain and organize such data, as well as for their subsequent processing and engineering calculations, appropriate monitoring observations are provided. The latter begin to be carried out long before the start of NPP construction and continue at all stages of the NPP life cycle, including the periods of construction, operation and decommissioning of the plant. The purpose of this research is to summarize the results of hydrometeorological monitoring at the Nizhny Novgorod NPP site and its vicinity, which was launched by the Scientific & Industrial Association Gidrotekhproekt in 2011. The description of stationary observation points and examples of calculation of regime hydrological and meteorological characteristics are given. It is noted that the accumulation of observation data series over a long period of time, which make it possible to identify potential climate changes in the study area, is of great importance.

ASME Operation and Maintenance (OM) Code Overview

2016 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Power ◽  
Regulatory Requirements ◽  
General Overview ◽  
Operation And Maintenance ◽  
American Society ◽  
General Organization

The purpose of this paper is to provide a general overview of the organization and content of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code. This will involve a brief description of the regulatory requirements associated with Inservice Testing (IST) as well as a brief overview of the OM Code scope and requirements. This paper will discuss, in general, the regulations requiring IST as well as a brief discussion on when Preservice Testing (PST) and IST become required. A general organization of the ASME OM Code will be provided as well as general topics associated with how to determine when testing and examination intervals are established; what documentation is required; and general discussion regarding the various subsections of the OM Code and the components associated with the OM Code. Alternatives to the OM Code requirements and how to obtain these alternatives will also be provided as well as how the edition applicability of the ASME OM Code is determined. There is also discussion regarding a few general issues associated with the OM Code regarding existing reactor power plants as well as the “new builds” and advanced reactor plants and designs.

The Role of Human Information Processing Models in System Development

1989 ◽  
Vol 33 (16) ◽  
pp. 1059-1063 ◽  
Author(s):  
Barry H. Kantowitz
Keyword(s):  
Information Processing ◽  
Nuclear Power ◽  
Power Plants ◽  
System Development ◽  
Theoretical Models ◽  
Cognitive Engineering ◽  
Human Cognition ◽  
Human Information Processing ◽  
Operation And Maintenance

Humans are complicated devices. Thus, systems in which people are embedded necessarily are complex. In order to better develop such systems, a means to organize and understand human complexity is required. Theoretical models of human information processing are one cognitive-engineering tool to help system development. This paper discusses the kinds of models that might be effective in solving practical problems. Suggestions are given for selecting a useful model from the plethora of available theoretical models. These issues are illustrated in the context of current research aimed at providing a general model of human cognition and action for application to the development, operation, and maintenance of nuclear power plants in Japan.

scholarly journals Analysis of Systems Thinking Safety by Artificial Intelligence (AI) Algorithm in South Korean Nuclear Power Plants (NPPs)

2021 ◽  
Vol 11 (10) ◽  
pp. 56-62
Author(s):  
Chang Hyun Baek ◽  
◽  
Kyung Bae Jang ◽  
Tae Ho Woo
Keyword(s):  
Artificial Intelligence ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Critical Factor ◽  
Safety Analysis ◽  
The Other ◽  
South Korean ◽  
Initial Stage ◽  
Relative Value

The artificial intelligence (AI) is applied to the safety analysis in the South Korean nuclear power plants (NPPs). The reinforcement learning (RL) is one of promising skills in the wise manipulations for the nuclear safety analysis where the reward is a critical factor to make the modelling. In the simulations, Y-axis means the relative value which shows the quantity of the accident possibility. The highest value is 4.0 in 46.25th year in which the values are increasing gradually. Otherwise, the values in the case with Agent gradually decrease. The highest value is near initial stage, which means the operation in NPPs is comparatively unstable. In the result, the values in the AI based controller graphs are higher than those of the other one. The RL algorithm is expressed by the Agent in this modelling, which is the most important factor in the AI-based operation in NPPs.

Benchmark Accelerated Aging of Harvested Hypalon/EPR and CSPE/XLPE Power and I&C Cable in Nuclear Power Plants

2016 ◽  
Author(s):  
Robert C. Duckworth ◽  
Emily Frame ◽  
Leonard S. Fifield ◽  
Samuel W. Glass
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Accelerated Aging ◽  
Functional Difference ◽  
Insulation Material ◽  
Cable Insulation ◽  
Control Rod ◽  
Oak Ridge

As part of the Light Water Reactor and Sustainability (LWRS) program in the U.S. Department of Energy (DOE) Office of Nuclear Energy, material aging and degradation research is currently geared to support the long-term operation of existing nuclear power plants (NPPs) as they move beyond their initial 40 year licenses. The goal of this research is to provide information so that NPPs can develop aging management programs (AMPs) to address replacement and monitoring needs as they look to operate for 20 years, and in some cases 40 years, beyond their initial, licensed operating lifetimes. For cable insulation and jacket materials that support instrument, control, and safety systems, accelerated aging data are needed to determine priorities in cable aging management programs. Before accelerated thermal and radiation aging of harvested, representative cable insulation and jacket materials, the benchmark performance of a new test capability at Oak Ridge National Laboratory (ORNL) was evaluated for temperatures between 70 and 135°C, dose rates between 100 and 500 Gy/h, and accumulated doses up to 200 kGy. Samples that were characterized and are representative of current materials in use were harvested from the Callaway NPP near Fulton, Missouri, and the San Onofre NPP north of San Diego, California. From the Callaway NPP, a multiconductor control rod cable manufactured by Boston Insulated Wire (BIW), with a Hypalon/ chlorosulfonated polyethylene (CSPE) jacket and ethylene-propylene rubber (EPR) insulation, was harvested from the auxiliary space during a planned outage in 2013. This cable was placed into service when the plant was started in 1984. From the San Onofre NPP, a Rockbestos Firewall III (FRIII) cable with a Hypalon/ CSPE jacket with cross-linked polyethylene (XLPE) insulation was harvested from an on-site, climate-controlled storage area. This conductor, which was never placed into service, was procured around 2007 in anticipation of future operation that did not occur. Benchmark aging for both jacket and insulation material was carried out in air at a temperature of 125°C or in a uniform 140 Gy/h gamma field over a period of 60 days. Their mechanical properties over the course of their exposures were compared with reference data from comparable cable jacket/insulation compositions and aging conditions. For both accelerated thermal and radiation aging, it was observed that the mechanical properties for the Callaway BIW control rod cable were consistent with those previously measured. However, for the San Onofre Rockbestos FRIII, there was an observable functional difference for accelerated thermal aging at 125°C. Details on possible sources for this difference and plans for resolving each source are given in this paper.

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