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.

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)

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).

Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes

2012 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Computer Code ◽  
Safety Analysis ◽  
Fuel Rod ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fuel Pin ◽  
Computer Codes ◽  
Flow Blockage

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis. This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.

scholarly journals A Procedure to Address the Fuel Rod Failures During LB-LOCA Transient in Atucha-2 NPP

2010 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria ◽  
Oscar Mazzantini
Keyword(s):  
Hydraulic System ◽  
Initial Conditions ◽  
Probabilistic Method ◽  
Three Dimensional ◽  
Computer Code ◽  
Calculation Methods ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pin ◽  
Computer Codes

Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes. This may imply the use of a dedicated fuel rod thermo-mechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermo-mechanical code. The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions (e.g. pin power axial profiles) are provided by core physics and three dimensional neutron kinetic coupled thermal-hydraulic system codes (RELAP5-3D©) calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, with the aim of a better understanding of the uncertainties involved and their technological consequences on the behavior of the fuel rods, not addressed in the current paper.

scholarly journals A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

2011 ◽  
Vol 2011 ◽  
pp. 1-11 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D'Auria ◽  
Oscar Mazzantini
Keyword(s):  
Hydraulic System ◽  
Initial Conditions ◽  
Probabilistic Method ◽  
Three Dimensional ◽  
Computer Code ◽  
Calculation Methods ◽  
Neutron Physics ◽  
Fuel Rod ◽  
Fuel Pin ◽  
Computer Codes

Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

Developing Procedure for Nuclear Regulatory Vulnerability Assessment of Nuclear Power Plants

2004 ◽  
Author(s):  
Venceslav Kostadinov ◽  
Stojan Petelin ◽  
Andrej Stritar
Keyword(s):  
Vulnerability Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Long Term Effects ◽  
Nuclear Facilities ◽  
Computer Tools ◽  
Nuclear Operators ◽  
Regulatory Bodies

First attempts of developing the procedure and adopting the methodology for nuclear regulatory vulnerability assessment of nuclear power plants are presented. The development of the philosophy and computer tools could be new and important insight for management of nuclear operators and nuclear regulatory bodies who face difficult questions about how to assess the vulnerability of nuclear power plants and other nuclear facilities on external and internal threats. In the situation where different and hidden threat sources throughout the world are dispersed, the assessments of security and safe operation of nuclear power plants are very important. Most risk assessments of nuclear power plants emphasize probability more than consequences. In addition, the probabilities omit terrorists. As of 09/11/2001, the emphasis has changed and this paper tries to show how to do this. Probabilities are enhanced by showing they are dynamic, interactive, and include countermeasures. Consequences are also enhanced by their dynamic nature dividing them on short, intermediate and long term effects. Risk is enhanced by creatively distinguishing between perceived and estimated.

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.

scholarly journals Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

2005 ◽  
Vol 20 (1) ◽  
pp. 3-15 ◽  
Author(s):  
Francesco d’Auria ◽  
Marco Cherubini ◽  
Maria Galassi ◽  
Nikolaus Muellner
Keyword(s):  
Nuclear Power ◽  
Computer Code ◽  
Scaling Analysis ◽  
Behavior Prediction ◽  
Plant Behavior ◽  
Test Methodology ◽  
Qualitative Level ◽  
Reactor Type ◽  
New Phenomena ◽  
Selection Of

This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

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