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Jahanfarnia, G.; Rahimi, M. H., E-mail: rezajahan@yahoo.com, E-mail: m.h.rahimi@gmx.com
Proceedings of NUCLEAR 2015 the 8th annual international conference on sustainable development through nuclear research and education. Part 1/32015
Proceedings of NUCLEAR 2015 the 8th annual international conference on sustainable development through nuclear research and education. Part 1/32015
AbstractAbstract
[en] Researches and studies on nuclear reactor core are usually subdivided into two major fields, named: Thermal-Hydraulic and Neutronic, in which, precise simulation of reactor behaviour in both fields is highly required to ensure the designers that reactor will work in a safe margin. In this study, a thermal-hydraulic analysis of pressurized water reactor core is performed using a porous media approach. Based on this approach, each fuel assembly was modelled and was divided into a network of lumped regions; each of them was characterized by a volume average parameter. In such manner, while complex geometries are easily defined and dealt with, the thermal-hydraulic parameter and phenomena like friction, shear stress, cross-flows, convective heat transfer and etc. are strictly included in simulations. To validate the applied approach, the numerical analysis and COBRA EN code results were compared for a typical PWR core and showed a good agreement. (authors)
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Constantin, Marin; Turcu, Ilie (Institute for Nuclear Research-Pitesti, 1 Campului Str., RO-115400 Mioveni, Arges (Romania)) (eds.); Institute for Nuclear Research-Pitesti, 1 Campului Str., RO-115400 Mioveni, Arges (Romania); University of Pitesti, Bd. Republicii, 71, Pitesti (Romania). Funding organisation: National Authority for Scientific Research, Bucharest (Romania); 244 p; ISSN 2066-2955;
; 2015; p. 22-30; NUCLEAR 2015: 8. annual international conference on sustainable development through nuclear research and education; Pitesti (Romania); 27-29 May 2015; Also available from author(s) or Institute for Nuclear Research-Pitesti, 1 Campului Str., RO-115400 Mioveni, Arges (RO); 9 refs., 12 figs., 1 tab.

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[en] A slab model has been widely applied in the thermal-hydraulic safety research fields for a simplified test development. However, the similarity in thermal-hydraulic behaviors is strongly dependent on the flow characteristics having a 2 D or 3 D flow field. The slab model is suitable for 2 D flow characteristics. If the slab model is applied in a strong 3 D flow field, the flow distortion grows highly. In this numerical study, the distortion of the flow field is investigated by changing the aspect ratio of the test section
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Korean Nuclear Society, Daejeon (Korea, Republic of); [1 CD-ROM]; Oct 2012; [2 p.]; 2012 autumn meeting of the KNS; Kyoungju (Korea, Republic of); 24-26 Oct 2012; Available from KNS, Daejeon (KR); 2 refs, 4 figs, 1 tab
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Nouy, E.; Crécy, A. de, E-mail: estelle.nouy@cea.fr2017
AbstractAbstract
[en] This paper describes some Uncertainty Quantification (UQ) and propagation performed within the NURESAFE European collaborative project. It especially presents the CEA results in the PREMIUM benchmark, an international benchmark organized by the Committee on the Safety of Nuclear Installations (CSNI, 2013) of the OECD/NEA, devoted to the quantification of the uncertainty of the physical models. It contains a confirmation of the estimation with a propagation of the uncertainties to two experiments including one blindly to check the capacity of the uncertainties to be extrapolated. In addition to the PREMIUM benchmark, other studies are shown, that were carried out to determine the accuracy of different quantifications using different indicators. The difficulties faced during the project are presented and analyzed, and potential high-impact improvements to the method are detected. One of them would consist in developing and assessing an evaluation matrix from SETs and IETs with two parts, one designed for the quantification and the other for the validation. The paper also highlights the need of tools to evaluate the quality of the UQ.
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S0029-5493(16)30408-3; Available from http://dx.doi.org/10.1016/j.nucengdes.2016.10.032; Copyright (c) 2016 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
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Lakehal, Djamel, E-mail: Lakehald@ethz.ch2019
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[en] The first time I met George was in March 1998 for dinner. And the last time I had dinner with him was in March of last year, or two decades later exactly. During the first evening, we discussed and confronted our strategies for the short period of time left then until his retirement in 2004; was it sufficient time to contribute to the discipline with something risky and innovative? During the last four-hour dinner, after having evoked life and death rather joyfully and serenely, we looked back and analysed what came out of our two decades of collaboration. This paper, which was prepared with the aim to portray George’s achievements in thermal-hydraulics during the last 20 years of our partnership (1998–2018), is written in the spirit of narrating our epilogue culinary-science-talk, while trying to be faithful to his thoughts and ideas about the developments of the discipline and its perspectives. The paper introduces first the cascade of computational tools and discusses trends related to reactor safety problems and developments needed, as well as the need for new kinds of refined experimental data. Although George was not directly involved in all the examples presented here, he felt so concerned about the success of each project/case presented in this paper that he was virtually part of it; and as he wrote in our last paper (Yadigaroglu and Lakehal, 2016): this is after all his near-home work. Finally, in memory of the two other great scientists who have left us recently, Geoff Hewitt and Sam Martin, the content of the paper includes two cases in which both of them had collaborated directly or indirectly.
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S0029549319301955; Available from http://dx.doi.org/10.1016/j.nucengdes.2019.110186; © 2019 Elsevier B.V. All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
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[en] The state feedback optimal control techniques are used in designing the reactor control system. The mathematical plant model with the temperature feedback effects is established from the point kinetics equation and the singly lumped thermal-hydraulic balance equations, and is expressed in terms of state variables. The LQR control system is designed, being followed by the LQG design to determine the optimal conditions both for the rod movements and for the output responses. This approach permits the consideration of the nuclear limitations since the rod movements are included as a system constraint. Also two different servo control schemes are proposed for the purpose of input tracking. The general control characteristics such as stability margins and output responses are investigated
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Korean Nuclear Society, Taejon (Korea, Republic of); 958 p; 1994; p. 137-142; 1994 autumn meeting of the KNS; Seoul (Korea, Republic of); 29 Oct 1994; Available from KNS, Taejon (KR); 4 refs, 6 figs
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[en] This study has extensively investigated the emerging 3-D printing technologies for use of MIR-based flow field visualization methods such as PIV and LDV. As a result, mixture of Herb essential oil and light mineral oil has been evaluated to be great working fluid due to its adequate properties. Using this combination, the RIs between 1.45 and 1.55 can be accurately matched, and most of the transparent materials are found to be ranged in here. Conclusively, the proposed MIR method are expected to provide large flexibility of model materials and geometries for laser based optical measurements. Particle Image Velocimetry (PIV) and Laser Doppler Velocimetry (LDV) are the two major optical technologies used for flow field visualization in the latest fundamental thermal-hydraulics researches. Those techniques seriously require minimizing optical distortions for enabling high quality data. Therefore, matching index of refraction (MIR) between model materials and working fluids are an essential part of minimizing measurement uncertainty. This paper proposes to use 3-D Printing technology for manufacturing models for the MIR-based optical measurements. Because of the large flexibility in geometries and materials of the 3-D Printing, its application is obviously expected to provide tremendous advantages over the traditional MIR-based optical measurements. This study focuses on the 3-D printing models and investigates their optical properties, transparent printing techniques, and index-matching fluids
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Korean Nuclear Society, Daejeon (Korea, Republic of); [1 CD-ROM]; Oct 2013; [2 p.]; 2013 Fall meeting of the KNS; Kyungju (Korea, Republic of); 23-25 Oct 2013; Available from KNS, Daejeon (KR); 4 refs, 3 figs, 2 tabs
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Su, Guanghui; Fukuda, K.
The ASME Foundation, Inc., Three Park Avenue, New York, NY 10016-5990 (United States)2002
The ASME Foundation, Inc., Three Park Avenue, New York, NY 10016-5990 (United States)2002
AbstractAbstract
[en] New methods of studying reactor thermohydraulic problematic such as CHF, DNBR, boiling and instability etc. were introduced briefly in this paper. Some of these methods have been successfully applied in the reactor thermohydraulic analysis. And the others have not yet been applied in this field. The further study work must be done to show that these new methods can be successfully applied in this field with high accuracy because they are interdisciplinary fields. (authors)
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2002; 5 p; American Society of Mechanical Engineers - ASME; New York (United States); ICONE-10: 10. international conference on nuclear engineering; Arlington - Virginia (United States); 14-18 Apr 2002; Country of input: France
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[en] This summary discusses a new methodology to support the construction of surrogate models for computationally expensive models with many input parameters and output responses. In realistic engineering calculations, such as thermal-hydraulics simulation, the cost of the simulation is typically expensive, especially when detailed models are employed to describe system behavior, thereby justifying the need for surrogate models. The cost of surrogate model construction is typically a function of the number of model parameters and the degree of model nonlinearity. For sufficiently complex models, the cost of constructing the surrogate model could be by itself overwhelmingly expensive, thereby diminishing its value. One key requirement of surrogate model construction is the ability to preselect the best parametric form to describe the model behavior over the expected range of parameter variations. This summary presents a new algorithm to help select the best approximation out of a template of functions using recent advances in reduced order modeling (ROM) techniques. The idea of selecting functions out of a template is typically referred to as subset selection in the statistical and applied mathematics community. Different from existing subset selection algorithms, we rely on ROM to help identify the most influential functions for the surrogate model. Previous work has shown that ROM can help reduce the effective dimensionality of the parameter space, which can help reduce the cost of surrogate model construction. In this summary, we adopt a different philosophy, where the fitting functions are treated as pseudo parameters and ROM is applied on the pseudo parameter space, with the result now being a ranked list of the pseudo parameters along with their contribution to the quality of the surrogate model fit. This approach allows the analyst to discard all terms that have negligible impact on the quantity of interest. A relap5 model is employed to demonstrate the application of the proposed methodology. (authors)
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2016 ANS Winter Meeting and Nuclear Technology Expo; Las Vegas, NV (United States); 6-10 Nov 2016; Country of input: France; 7 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)
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Transactions of the American Nuclear Society; ISSN 0003-018X;
; v. 115; p. 1740-1743

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AbstractAbstract
[en] Highlights: • Overview on state of the art of the numerical modeling in SYS-TH codes is presented. • Effect of virtual mass force and interfacial pressure on hyperbolicity is discussed. • Role of physical and numerical regularizations in ill-posed equations is discussed. - Abstract: The paper provides a brief overview on the state of the art of the numerical modeling in present system thermal-hydraulic codes in the framework of the Forum & Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). FONESYS is a network among code developers who share the common objective to strengthen current technology and to bring technical argumentation against possible disbelief in SYS-TH codes. The aim of the FONESYS network is to highlight the capabilities and the robustness as well as the limitations of current SYS-TH codes to predict the main phenomena during transient scenarios in nuclear reactors for safety issues. This objective will be achieved primarily through the development of a strategy that aims to identify the key aspects for the improvement of SYS-TH codes and then establish a work program regarding actions to be taken to provide an answer to the needs listed in the Road Map of FONESYS network. In this regard, key topics, like the hyperbolicity, the numerical methods and some closure laws with particular reference to the modeling of some interfacial forces in the current SYS-TH codes, have been treated by the FONESYS network and are summarized in this paper.
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S0029-5493(17)30304-7; Available from http://dx.doi.org/10.1016/j.nucengdes.2017.06.033; Copyright (c) 2017 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
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[en] Investigation of resaturation and thermal-hydro-mechanical behavior for the buffer of a repository requires measuring the water content of compacted bentonite. This study investigated the relative humidity of compacted bentonites using a humidity sensor (Vaisala HMT 334) applicable under high temperature and pressure, and then conducted a multi-regression analysis based on the measured results to determine relationships among the water content, relative humidity, and temperature. The relationships for the compacted bentonites with the dry densities of 1,500 kg/m3 and 1,600 kg/m3 were expressed as ω = 0.196RH - 0.029T + 1.391 (r2 = 0.96) and ω = 0.199RH - 0.029T + 2.596 (r2 = 0.96), respectively. These were then used to interpret the resaturation of bentonite blocks in the KENTEX test.
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6 refs, 12 figs, 1 tab
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Journal of the Korean Radioactive Waste Society; ISSN 1738-1894;
; v. 7(2); p. 101-107

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