scholarly journals МОДЕЛІ НАДІЙНОСТІ УГРУПОВАНЬ ФЛОТІВ БПЛА З КОВЗНИМ РЕЗЕРВУВАННЯМ ДЛЯ МОНІТОРИНГУ ПОТЕНЦІЙНО НЕБЕЗПЕЧНИХ ОБ’ЄКТІВ

2019 ◽  
pp. 147-156
Author(s):  
Герман Вікторович Фесенко ◽  
Вячеслав Сергійович Харченко
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Probability Of Failure ◽  
Reliability Models ◽  
Aerial Vehicle ◽  
Reliability Block Diagrams ◽  
The Subject ◽  
High Level ◽  
Analytical Expressions ◽  
Support Decision Making

Over the past years, unmanned aerial vehicles have been used to solve the problems of pre-and post-accident monitoring of nuclear power plants and other potentially dangerous objects. The use of a fleet (fleet grouping) of the unmanned aerial vehicle (UAV) for monitoring missions is due to the special conditions of the aggressive environment, which causes the failures of certain UAVs, and therefore needs to ensure a high level of reliability of such a fleet (the group of fleets). The most effective way to solve this problem is to use k-out-of-n redundancy. The subject of the study is a UAV fleet group with k-out-of-n redundancy. In order to take into account the reliability of the control station for various variants of the organization of UAV fleet groups, it is advisable to formulate the following tasks: to analyze the different structures of UAV fleet groups taking into account the scheme of activation of redundant UAVs; to develop and investigate models of reliability of UAV fleet groups with a centralized, decentralized and partially decentralized schemes of activation of redundant UAVs with the possibility of control station redundancy; to formulate recommendations for choosing a scheme for activation of redundant UAVs. Research results: the structure of the UAV fleet group with two-level k-out-of-n redundancy (at fleet levels and their groups using the reserve fleet) and different variants of organization of control stations were proposed. The centralized, decentralized and partially decentralized structures of activation of the redundant UAVs for a fleet group with reserve (reserve fleets) are investigated, namely: the reliability block diagrams of these systems are constructed; analytical expressions for calculating the probability of failure-free operation of the UAV fleet group based on each of these schemes are obtained; the following dependencies are obtained and investigated: probability of failure-free operation of a  UAV fleet group with different probabilities of UAV failure-free operation from the number of the main fleets in case of use of reserve fleet with three UAVs; the probability of UAV fleet group failure-free operation with different schemes of activation of redundant UAVs from the number of main fleets in the case of using a reserve fleet with three UAVs. The development and research of reliability models have made it possible to formulate tasks regarding the choice of schemes for activation of redundant UAVs and the corresponding recommendations for the organization of groups. Further research is appropriate to focus on developing software to support decision-making on choosing the options for structures and taking into account possible schemes to get areas of responsibility by UAVs.

NUCLEAR WASTE MANAGEMENT IN SPAIN: ANALYSIS OF THE CURRENT SITUATION AND ALTERNATIVE STRATEGIES

10.6036/10156 ◽  
2021 ◽  
Vol 96 (4) ◽  
pp. 355-358
Author(s):  
Pablo Fernández Arias ◽  
DIEGO VERGARA RODRIGUEZ
Keyword(s):  
Nuclear Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Fuel ◽  
Nuclear Waste Storage ◽  
Alternative Strategies ◽  
Temporary Storage ◽  
Long Term Storage ◽  
High Level

Centralized Temporary Storage Facility (CTS) is an industrial facility designed to store spent fuel (SF) and high level radioactive waste (HLW) generated at Spanish nuclear power plants (NPP) in a single location. At the end of 2011, the Spanish Government approved the installation of the CTS in the municipality of Villar de Cañas in Cuenca. This approval was the outcome of a long process of technical studies and political decisions that were always surrounded by great social rejection. After years of confrontations between the different political levels, with hardly any progress in its construction, this infrastructure of national importance seems to have been definitively postponed. The present research analyzes the management strategy of SF and HLW in Spain, as well as the alternative strategies proposed, taking into account the current schedule foreseen for the closure of the Spanish NPPs. In view of the results obtained, it is difficult to affirm that the CTS will be available in 2028, with the possibility that its implementation may be delayed to 2032, or even that it may never happen, making it necessary to adopt an alternative strategy for the management of GC and ARAR in Spain. Among the different alternatives, the permanence of the current Individualized Temporary Stores (ITS) as a long-term storage strategy stands out, and even the possibility of building several distributed temporary storage facilities (DTS) in which to store the SF and HLW from several Spanish NPP. Keywords: nuclear waste, storage, nuclear power plants.

Contribution to the Design of Concrete Columns Reinforced With Externally Fiber Polymers

2009 ◽  
Author(s):  
Ferran Prats Bella ◽  
Ramo´n Gonza´lez-Drigo ◽  
Adrina Bachiller San˜a
Keyword(s):  
Structural Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Pushover Analysis ◽  
Concrete Columns ◽  
Seismic Evaluation ◽  
Modal Pushover Analysis ◽  
Nuclear Domain ◽  
Concrete Building ◽  
The Subject

In the design of seismic category 1 buildings in nuclear power plants (NPP) or, outside the nuclear domain, in the conventional structural design of buildings, the seismic evaluation of these buildings may be done. In the occurrence of an earthquake in a NPP or in the case of changing the use of a conventional building, the seismic levels are modified. Then a new analysis need to be performed. This paper focusses on the situation where reinforcing the concrete building is needed and it also analyses how an extern reinforcement performed using polymers can be carried out to fulfill the new seismic requeriments. We present two main results: a) the resulting momentum-curvature diagrams obtained reinforcing standard segments embraced with polymers; b) the evaluation of the structure capacity on the basis of the modified diagrams. Finally, a modal pushover analysis is selected to perform the seismic evaluation of two types of concrete columns, those having a polymer reinforcement and those without it. This paper presents the basis of the subject in a theoretical form.

Underground Siting of Nuclear Power Plants: Insights From Fukushima

2012 ◽  
Author(s):  
Jay F. Kunze ◽  
James M. Mahar ◽  
Kellen M. Giraud ◽  
C. W. Myers
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Light Water ◽  
Waste Storage ◽  
Small Modular Reactor ◽  
Fukushima Daiichi ◽  
Water Reactors ◽  
High Level

Siting of nuclear power plants in an underground nuclear park has been proposed by the authors in many previous publications, first focusing on how the present 1200 to 1600 MW-electric light water reactors could be sited underground, then including reprocessing and fuel manufacturing facilities, as well as high level permanent waste storage. Recently the focus has been on siting multiple small modular reactor systems. The recent incident at the Fukushima Daiichi site has prompted the authors to consider what the effects of a natural disaster such as the Japan earthquake and subsequent tsunami would have had if these reactors had been located underground. This paper addresses how the reactors might have remained operable — assuming the designs we previously proposed — and what lessons from the Fukushima incident can be learned for underground nuclear power plant designs.

Centralization of Canada’s Spent Nuclear Fuel

2013 ◽  
Author(s):  
Krista Nicholson ◽  
John McDonald ◽  
Shona Draper ◽  
Brian M. Ikeda ◽  
Igor Pioro
Keyword(s):  
Nuclear Fuel ◽  
Site Selection ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Selection Process ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Candu Reactors ◽  
High Level

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel. The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method. The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

Instrument Health Monitoring for Extending Calibration Intervals of Safety Related Instruments

2016 ◽  
Author(s):  
R. Steven Black ◽  
Aaron J. Hussey ◽  
Randall L. Bickford
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Path ◽  
Maintenance Cost ◽  
Calibration Error ◽  
Or Efficiency ◽  
Active Monitoring ◽  
Plant Safety ◽  
Calibration Program ◽  
High Level

The ability to extend calibration intervals for nuclear plant instrumentation has multiple benefits for improving productivity and reducing operating costs at nuclear plants. Benefits include fewer calibrations inside containment during an outage and associated reduced critical path time and ALARA exposure, reduced risk of calibration error or instrument damage during removal and replacement, and reduced operations and maintenance cost for instrument removal, calibration and replacement. A good instrument calibration program ensures instruments are checked frequently enough to provide a high level of confidence that they are performing within acceptable limits, but no more frequently. Over-testing of plant instruments and equipment should be avoided for two reasons: valuable resources are expended on maintenance that might not measurably improve plant safety, reliability, or efficiency; and the potential exists for adjustment errors or equipment damage each time an instrument is removed from service for testing. Over-testing increases the risk of errors or damage being introduced without a justifiable improvement in reliability. This paper discusses the regulatory framework for extending calibration intervals of safety related instruments for U.S. based nuclear power plants. Necessary changes to licensing, plant processes and procedures, training, and configuration management are summarized. An example application of pattern recognition modeling is provided to highlight the analytical support for the processes provided by active monitoring to confirm on-going instrument heath. The paper concludes with a listing of recommended steps to implement a practical program for extending calibration intervals of safety related instruments within the U.S. nuclear regulatory environment.

Development Trends in Nuclear Technology and Related Safety Aspects

2002 ◽  
Author(s):  
B. Kuczera ◽  
P. E. Juhn ◽  
K. Fukuda
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Nuclear Reactors ◽  
Nuclear Technology ◽  
User Requirements ◽  
Safety Standards ◽  
Safety Requirements ◽  
Fuel Cycles ◽  
High Level

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO subtask “User Requirements and Nuclear Energy Development Criteria in the Area of Safety” have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R&D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle.

Life Time Management: A Practical Approach of Nuclear Power Plants

2002 ◽  
Author(s):  
Jaroslav Bartonicek ◽  
Klaus-Juergen Metzner ◽  
Friedrich Schoeckle
Keyword(s):  
Time Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mechanical Systems ◽  
Life Time ◽  
Practical Approach ◽  
Degradation Mechanisms ◽  
High Level ◽  
Quality Status

A comprehensive life time management has to take care of all safety and availability relevant components in nuclear power plants, with different intensity, of course. For instance, mechanical systems and components can be basically classified/ranked into three different groups: (1): The quality status of the components in this group has to be guaranteed on a pre-defined (high) level. (2): The quality status of the components in this group has to be maintained on its actual level. (3): Other components with no specific quality demands. Regarding the first group, integrity has to be guaranteed. Therefore it is necessary to monitor the possible root causes of degradation mechanisms during plant operation; thus the degradation effects can be assessed and — more important — controlled to maintain the safety standard on the demanded high level without any compromise. The monitoring of consequences of degradation mechanisms is being performed as an additional redundant measure. The requirements to maintain the quality status of the second group of components can be fulfilled by monitoring of the consequences of operational degradation mechanisms to be performed by preventive maintenance activities, in terms of tests, inspections and repairs, using either time dependant procedures or component condition orientated methods. For the third group of components, no preventive action is necessary. However, failures and malfunctions have to be assessed statistically to avoid a reduction of the required basic component quality. In the first two groups all safety relevant components and systems are included. Generally, aging management programs cover these two groups of components; life time management covers all of above groups. This paper concentrates on mechanical systems and components; it summarizes the practical approach to life time management as it is realized in German nuclear power plants. The application is discussed using dedicated examples.

Design, Loading, Transport and Storage Experience of CASTOR® Casks for Vitrified High Level Waste

2003 ◽  
Author(s):  
Andre´ Voßnacke ◽  
Wilhelm Graf ◽  
Roland Hu¨ggenberg ◽  
Astrid Gisbertz
Keyword(s):  
Nuclear Power ◽  
High Performance ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Level Waste ◽  
Spent Fuel ◽  
Highly Active ◽  
Dry Conditions ◽  
Intermediate Storage ◽  
High Level

The revised German Atomic Act together with the Agreement between the German Government and the German Utilities of June 11, 2001 form new boundary conditions that considerably influence spent fuel strategies by stipulation of lifetime limitations to nuclear power plants and termination of reprocessing. The contractually agreed return of reprocessing residues comprises some 156 casks containing vitrified highly active waste, the so-called HAW or glass canisters, coming form irradiated nuclear fuel assemblies to be shipped from COGEMA, France and BNFL, UK to Germany presumably until 2011. Several hundred casks with compacted residues and other waste will follow. The transports are scheduled presumably beyond 2020. The central interim storage facilities in Ahaus and Gorleben, formerly intended to accumulate up to 8,000 t of heavy metal (HM) of spent fuel from German nuclear power plants, offer sufficient capacity to receive the totality of residues to be returned from reprocessing abroad. GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and is one of the world leaders for delivering spent fuel and high level waste casks. Long-term intermediate storage of spent fuel is carried out under dry conditions using these casks that are licensed for transport as well as for storage. Standardized high performance casks such as the types CASTOR® HAW 20/28 CG, CASTOR® V/19 and CASTOR® V/52 meet the needs of most nuclear power plants in Germany. Up to now GNS has co-ordinated the loading and transport of 27 casks loaded with 28 canisters each from COGEMA back to Germany for storage in Gorleben for up to 40 years. In all but one case the cask type CASTOR® HAW 20/28 CG has been used.

Building the UK ABWR: Developments in Structural Integrity Safety Cases for Highest Reliability Components in the UK

2018 ◽  
Author(s):  
Tim Jelfs ◽  
James O’Neill ◽  
Angus Beveridge
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Advisory Group ◽  
Manufacturing Processes ◽  
Design And Manufacturing ◽  
Safety Case ◽  
Safety Cases ◽  
The Uk ◽  
High Level

Nuclear power plants contain certain components whose gross failure would lead to intolerable radiological consequences. In the UK, a common terminology used for such components is Very High Integrity (VHI). If it is not possible to engineer lines of protection for these components, a safety case must demonstrate to UK regulators that the probability of gross failure is demonstrably so low that it can be discounted. A previous paper [Ref. 1] has described, at a high level, how the structural integrity safety case for a nuclear new build project in the UK — the UK Advanced Boiling Water Reactor (UK ABWR) is being structured. As described in [Ref. 1], the structural integrity safety case for the UK ABWR is based on the guidance provided by the UK Technical Advisory Group on Structural Integrity (TAGSI) and aims to demonstrate a multi-legged safety case with robust and independent legs giving confidence of defense in depth. Design to the internationally recognized ASME code [Refs. 2, 3, 4] is supplemented by a significant number of beyond code requirements such as supplementary inspection and inspection qualification, augmented material testing requirements, defect tolerance assessment to the well-established R6 procedure [Ref. 5], and demonstration that design and manufacturing processes have reduced risks to As Low as Reasonably Practicable (ALARP). This paper provides an updated position of the progress made on the UK ABWR project. It also provides more specific details on the activities the future licensee, Horizon Nuclear Power, has performed in support of the demonstration that design and manufacturing processes have reduced risks to ALARP. This kind of additional work is vital to providing the UK regulator with confidence that the risk of failure of VHI components has been reduced to ALARP.

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