International Training Program in Support of Safety Analysis: 3D S.UN.COP—Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars

2009 ◽  
Author(s):  
Alessandro Petruzzi ◽  
Francesco D’Auria ◽  
Tomislav Bajs ◽  
Francesc Reventos
Keyword(s):  
Nuclear Power ◽  
Effective Means ◽  
Computer Code ◽  
State University ◽  
Nuclear Facilities ◽  
Joint Research ◽  
User Training ◽  
Support Organizations ◽  
Institute Of Technology ◽  
Current Code

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the ‘user effect’ and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Nine seminars have been held at University of Pisa (two in 2004), at The Pennsylvania State University (2004), at the University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006), in Buenos Aires, Argentina (October 2006), requested by Autoridad Regulatoria Nuclear (ARN), Nucleoelectrica Argentina S.A (NA-SA) and Comisio´n Nacional de Energi´a Ato´mica (CNEA), at the College Station, Texas A&M, (January-February 2007), in Hamilton and Niagara Falls, Ontario (October 2007) requested by Atomic Energy Canada Limited (AECL), Canadian Nuclear Society (CNS) and Canadian Nuclear Safety Commission (CNSC), in Petten, The Netherlands (October 2008) in cooperation with the Institute of Energy of the Joint Research Center of the European Commission (IE-JRC-EC). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed ‘permanent’ stepwise approach to user training. The 3D S.UN.COP 2008 at IE-JRC was successfully held with the attendance of 35 participants coming from more than 10 countries and 20 different institutions (universities, vendors and national laboratories). More than 30 scientists (coming from more than 10 countries and 20 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A tenth seminar will be held (October 2009) at the Royal Institute of Technology (KTH) in Amsterdam (Sweden), involving more than 30 scientists between lectures and code developers (http://dimnp.ing.unipi.it/3dsuncop/2009/index.html).

International Training Program in Support of Safety Analysis: 3D S.UN.COP—Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars

2006 ◽  
Author(s):  
Alessandro Petruzzi ◽  
Francesco D’Auria ◽  
Tomislav Bajs ◽  
Francesc Reventos
Keyword(s):  
Nuclear Power ◽  
Effective Means ◽  
Computer Code ◽  
State University ◽  
Nuclear Facilities ◽  
User Training ◽  
Support Organizations ◽  
Regulatory Bodies ◽  
International Training Program ◽  
Current Code

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the ‘user effect’ and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed ‘permanent’ stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A sixth seminar will be organized in 2007 at the Texas A&M University involving more than 30 scientists between lecturers and code developers. (http://dimnp.ing.unipi.it/3dsuncop/2007)

scholarly journals International Course to Support Nuclear Licensing by User Training in the Areas of Scaling, Uncertainty, and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes: 3D S.UN.COP Seminars

2008 ◽  
Vol 2008 ◽  
pp. 1-16 ◽  
Author(s):  
Alessandro Petruzzi ◽  
Francesco D'Auria ◽  
Tomislav Bajs ◽  
Francesc Reventos ◽  
Yassin Hassan
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Effective Means ◽  
Computer Code ◽  
Nuclear Facilities ◽  
User Training ◽  
Research Organizations ◽  
Coupled Codes ◽  
Regulatory Bodies ◽  
Computer Codes

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the “user effect” and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to the areas of the scaling, uncertainty, and 3D coupled code analysis.

scholarly journals The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

2012 ◽  
Vol 2012 ◽  
pp. 1-12 ◽  
Author(s):  
Jean-Pierre Van Dorsselaere ◽  
Ari Auvinen ◽  
David Beraha ◽  
Patrick Chatelard ◽  
Christophe Journeau ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Computer Code ◽  
Severe Accident ◽  
Research Network ◽  
Combustion Source ◽  
Joint Research ◽  
Severe Accidents ◽  
Young Researchers

Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP). After a first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…), source term issues (mainly iodine behaviour). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

scholarly journals New trends in Multihazards Probabilistic Safety Assessment for nuclear installations: the H2020-NARSIS Project

2020 ◽  
Author(s):  
Evelyne Foerster ◽  
Behrooz Bazargan-Sabet ◽  
James Daniell ◽  
Pierre Gehl ◽  
Philip J. Vardon ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Severe Accident ◽  
Management Decision ◽  
Nuclear Facilities ◽  
Probabilistic Safety Assessment ◽  
Human Aspects ◽  
Support Organizations ◽  
Probabilistic Safety

<p>The methodology for Probabilistic Safety Assessment (PSA) of Nuclear Power Plants (NPPs) has been used for decades by practitioners to better understand the most probable initiators of nuclear accidents by identifying potential accident scenarios, their consequences, and their probabilities. However, despite the remarkable reliability of the methodology, the Fukushima Dai-ichi nuclear accident in Japan, which occurred in March 2011, highlighted a number of challenging issues (e.g. cascading event - cliff edge - scenarios) with respect to the application of PSA questioning the relevance of PSA practice, for such low-probability but high-consequences external events. Following the Fukushima Dai-ichi accident, several initiatives at the international level, have been launched in order to review current practices and identify shortcomings in scientific and technical approaches for the characterization of external natural extreme events and the evaluation of their consequences on the safety of nuclear facilities.</p><p>The H2020 project “New Approach to Reactor Safety ImprovementS” (NARSIS, 2017-2021) aims at proposing some improvements to be integrated in existing PSA procedures for NPPs, considering single, cascade and combined external natural hazards (earthquakes, flooding, extreme weather, tsunamis). It coordinates the research efforts of eighteen partners encompassing leading universities, research institutes, technical support organizations (TSO), nuclear power producers and suppliers, reactor designers and operators from ten countries.</p><p>The project will lead to the release of various tools together with recommendations and guidelines for use in nuclear safety assessment, including a Bayesian-based multi-risk framework able to account for causes and consequences of technical, social/organizational and human aspects and as well as a supporting Severe Accident Management decision-making tool for demonstration purposes.</p><p>The NARSIS project has now been running for two years and a half, and the first set of deliverables and tools have been produced as part of the effort of the consortium. Datasets have been collected, methodologies tested, states of the art have been produced, and various criteria and plans developed. First results have started to emerge and will be presented here.</p>

Second International scientific conference «Vegetation of Eastern Europe and Northern Asia» (Bryansk, 12­–13 October 2020)

2020 ◽  
pp. 149-153
Author(s):  
V. V. Neshataev ◽  
D. D. Karsonova ◽  
A. A. Kurka
Keyword(s):  
Eastern Europe ◽  
Vegetation Classification ◽  
Scientific Conference ◽  
State University ◽  
Research Projects ◽  
Joint Research ◽  
Northern Asia ◽  
Oral Presentations ◽  
Made In ◽  
Second International

On October 12th and 13th, 2020, Bryansk State University held an international scientific online conference "Vegetation of Eastern Europe and Northern Asia". The Proceedings of abstracts includes 66 reports by 118 authors and co-authors from 5 countries, 34 localities and 51 organizations. During the meeting, 41 oral presentations were made. In conclusion, it was noted that it is necessary to promote an integration of geobotanists and florists from different regions in order to implement joint research projects. In particular, this concerns a project of making a vegetation classification in Russia.

Sobol′ Sensitivity Analysis Using a Neural Network Model of a LB-LOCA in the ZION Nuclear Power Plant With CATHARE-2 V2.5 Code

2013 ◽  
Author(s):  
Fabrice Fouet ◽  
Pierre Probst
Keyword(s):  
Neural Network ◽  
Sensitivity Analysis ◽  
Nuclear Power ◽  
Uncertainty Propagation ◽  
Correlation Coefficients ◽  
Computer Code ◽  
Numerical Application ◽  
Loss Of Coolant Accident ◽  
Input Parameters ◽  
Artificial Neural Network Ann

In nuclear safety, the Best-Estimate (BE) codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, developed a method of uncertainty propagation. This method has been tested with the BE code used is CATHARE-2 V2.5 in order to evaluate the Peak Cladding Temperature (PCT) of the fuel during a Large Break Loss Of Coolant Accident (LB-LOCA) event, starting from a large number of input parameters. A sensitivity analysis is needed in order to limit the number of input parameters and to quantify the influence of each one on the response variability of the numerical model. Generally, the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the Sobol′ methodology. The GSA requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. We have tested Artificial Neural Network (ANN) methodology for its construction and the Sobol′ methodology for the GSA. The paper presents a numerical application of the previously described methodology on the ZION reactor, a Westinghouse 4-loop PWR, which has been retained for the BEMUSE international problem [8]. The output is the first maximum PCT of the fuel which depends on 54 input parameters. This application outlined that the methodology could be applied to high-dimensional complex problems.

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

Development of New Constraction Method for LSBWR

2002 ◽  
Author(s):  
H. Heki ◽  
M. Nakamaru ◽  
T. Maruyama ◽  
H. Hirai ◽  
M. Aritomi
Keyword(s):  
Nuclear Power ◽  
Seismic Isolation ◽  
Three Dimensional ◽  
Operating Cycle ◽  
Construction Period ◽  
Hull Structure ◽  
Plant Equipment ◽  
Tokyo Institute ◽  
Institute Of Technology ◽  
Economic Improvement

LSBWR (Long operating cycle Simplified BWR) is a modular, direct cycle, light water cooled, and small power (100–300MWe) reactor. The design considers requirements from foreign utilities as well as from Japanese. LSBWR is currently being developed by Toshiba Corporation and Tokyo Institute of Technology. Major characteristics of the LSBWR are: 1) Long operating cycle (target: over 15 years), 2) Simplified systems and building, 3) Factory fabrication in module. From the perspective of economic improvement of nuclear power plant, it is needed to shorten the plant construction period and to reduce building volume. In designing LSBWR building, a new building structure, where the hull structure of a ship is applied to floors and walls of LSBWR has been studied. Since the hull structure is manufactured at a shipyard, building module that includes plant equipment becomes possible. The application of the hull structure, which can make large modules at a shipyard, is an effective solution to the lack of laborer and economic improvement. LSBWR is a small size BWR, turbine is smaller size and lighter weight than medium or larger size plant. Then, it has been studied to install a reactor and a turbine in the same building for decreasing building volume. From the view point of standardization, whole building is supported by three dimensional seismic isolation mechanism.

scholarly journals Reviewer Acknowledgements for Journal of Mathematics Research, Vol. 9, No. 2

2017 ◽  
Vol 9 (2) ◽  
pp. 155
Author(s):  
Sophia Wang
Keyword(s):  
United Arab Emirates ◽  
Medical Center ◽  
State University ◽  
National Academy Of Sciences ◽  
Bulgarian Academy ◽  
University Of Houston ◽  
Institute Of Technology ◽  
Mathematics Research ◽  
Academy Of Sciences

Journal of Mathematics Research wishes to acknowledge the following individuals for their assistance with peer review of manuscripts for this issue. Their help and contributions in maintaining the quality of the journal is greatly appreciated.Many authors, regardless of whether Journal of Mathematics Research publishes their work, appreciate the helpful feedback provided by the reviewers.Reviewers for Volume 9, Number 2  Alberto Simoes, University of Beira Interior, PortugalAli Berkol, Space and Defense Technologies & Baskent University, TurkeyArman Aghili, University of Guilan, IranCecilia Maria Fernandes Fonseca, Polytechnic of Guarda, PortugalGane Sam Lo, Universite Gaston Berger de Saint-Louis, SenegalMarek Brabec, Academy of Sciences of the Czech Republic, Czech RepublicMaria Alessandra Ragusa, University of Catania, ItalyMohammad Sajid, Qassim University, Saudi ArabiaMohd Hafiz, Universiti Sains Malaysia, , MalaysiaN. V. Ramana Murty, Andhra Loyola College, IndiaOlivier Heubo-Kwegna, Saginaw Valley State University, USAOmur Deveci, Kafkas University, TurkeyÖzgür Ege, Celal Bayar University, TurkeyPeng Zhang, State University of New York at Stony Brook, USAPhilip Philipoff, Bulgarian Academy of Sciences, BulgariaRovshan Bandaliyev, National Academy of Sciences of Azerbaijan, AzerbaijanSanjib Kumar Datta, University of Kalyani, IndiaSelcuk Koyuncu, University of North Georgia, USASergiy Koshkin, University of Houston Downtown, USAShenghua Ni, Vanderbilt University Medical Center, USAVishnu Narayan Mishra, Sardar Vallabhbhai National Institute of Technology, IndiaWaleed Al-Rawashdeh, Montana Tech, USAYifan Wang, University of Houston, USAYoussef Ei Foutayeni, Modeling and Simulation Laboratory Lams Hassan II University, MoroccoYoussef El-Khatib, United Arab Emirates University, United Arab EmiratesZoubir Dahmani, University of Mostaganem, Algeria Sophia WangOn behalf of,The Editorial Board of Journal of Mathematics ResearchCanadian Center of Science and Education

Sign in / Sign up

Export Citation Format

Share Document