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Journal of engineering for gas turbines and power |
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Ajouter le résultat dans votre panierEffect of oxygen potential on crack growth in alloys for advanced energy systems / Julian K. Benz in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 132 N° 10 (Octobre 2010)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Titre : Effect of oxygen potential on crack growth in alloys for advanced energy systems Type de document : texte imprimé Auteurs : Julian K. Benz, Auteur ; Ji Hyun Kim, Auteur ; Ronald G. Ballinger, Auteur Année de publication : 2011 Article en page(s) : 07 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Chromium alloys Fatigue cracks Iron alloys Nickel alloys Oxygen Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The effect of oxygen partial pressure on crack growth rates in Alloy 617 has been studied using both static and fatigue loadings at 650°C over the oxygen partial pressure range 10−19−10−3 atm. Tests were conducted at either the constant stress intensity factor K for static conditions or the constant DeltaK in fatigue. Oxygen concentration was measured on both the inlet and outlet of the test retort as well as in situ with a probe located directly at the specimen surface. For fatigue loading the crack path was observed to be transgranular but crystallographic with a decreasing growth rate as the oxygen concentration decreased. However, for static loading the crack path shifted to intergranular at the same Kmax (fatigue) and exhibited what appears to be an increasing crack growth rate with decreasing oxygen concentration. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Effect of oxygen potential on crack growth in alloys for advanced energy systems [texte imprimé] / Julian K. Benz, Auteur ; Ji Hyun Kim, Auteur ; Ronald G. Ballinger, Auteur . - 2011 . - 07 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Mots-clés : Chromium alloys Fatigue cracks Iron alloys Nickel alloys Oxygen Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The effect of oxygen partial pressure on crack growth rates in Alloy 617 has been studied using both static and fatigue loadings at 650°C over the oxygen partial pressure range 10−19−10−3 atm. Tests were conducted at either the constant stress intensity factor K for static conditions or the constant DeltaK in fatigue. Oxygen concentration was measured on both the inlet and outlet of the test retort as well as in situ with a probe located directly at the specimen surface. For fatigue loading the crack path was observed to be transgranular but crystallographic with a decreasing growth rate as the oxygen concentration decreased. However, for static loading the crack path shifted to intergranular at the same Kmax (fatigue) and exhibited what appears to be an increasing crack growth rate with decreasing oxygen concentration. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Titre : Residual life assessment of steam generators with alloy 600 TT tubing : methods and application Type de document : texte imprimé Auteurs : Ian de Curieres, Auteur ; Marie-Christine Meunier, Auteur ; Pierre Joly, Auteur Année de publication : 2011 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Boilers Corrosion Cracks Inspection Pipes Steam power stations Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The aim of the paper is to introduce methods to estimate the residual life of steam generators with alloy 600 thermally treated (TT) tubing, taking into account primary water stress corrosion cracking (PWSCC) as the main contributor damage. The methods take into account both initiation and propagation of PWSCC cracks in the expansion transition zone of steam generator tubes, as well as the current damage status (cracking and plugging) of the tube bundle, known from inspection results. A probabilistic model is used to treat initiation, while the propagation stage is treated in a deterministic way based on inspection data. After introducing the methods used to assess the residual life, a brief parametric study will be shown to illustrate the effects of initiation versus propagation. Eventually, the cases of a few actual steam generators with tubing made of alloy 600 TT showing different situations of present damage and damage evolution rates will be presented. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Residual life assessment of steam generators with alloy 600 TT tubing : methods and application [texte imprimé] / Ian de Curieres, Auteur ; Marie-Christine Meunier, Auteur ; Pierre Joly, Auteur . - 2011 . - 06 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Mots-clés : Boilers Corrosion Cracks Inspection Pipes Steam power stations Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The aim of the paper is to introduce methods to estimate the residual life of steam generators with alloy 600 thermally treated (TT) tubing, taking into account primary water stress corrosion cracking (PWSCC) as the main contributor damage. The methods take into account both initiation and propagation of PWSCC cracks in the expansion transition zone of steam generator tubes, as well as the current damage status (cracking and plugging) of the tube bundle, known from inspection results. A probabilistic model is used to treat initiation, while the propagation stage is treated in a deterministic way based on inspection data. After introducing the methods used to assess the residual life, a brief parametric study will be shown to illustrate the effects of initiation versus propagation. Eventually, the cases of a few actual steam generators with tubing made of alloy 600 TT showing different situations of present damage and damage evolution rates will be presented. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Titre : Chromatography column system with controlled flow and temperature for engineering scale application Type de document : texte imprimé Auteurs : Sou Watanabe, Auteur ; Ichiro Goto, Auteur ; Yuichi Sano, Auteur Année de publication : 2011 Article en page(s) : 07 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Actinides Chromatography Liquid metal fast breeder reactors Radioactive waste processing Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The Japan Atomic Energy Agency is conducting research and development study on the engineering scale extraction chromatography system, which uses silica-based adsorbents impregnated with an extractant for the minor actinides (Am and Cm) recovery from the high level liquid waste generated in the spent fast breeder reactor (FBR) fuel reprocessing, as a part of the fast reactor cycle technology development project. A bench scale testing system was made and provided for the first step of development. The column of the test system (inside diameter of 480 mm or 200 mm with height of 650 mm) was equipped with ports for six external sensors at its top, middle, and bottom levels for measuring the flow velocity or temperature, and for additional six heaters for simulating the decay heat of Am and Cm at the middle level of the column. The flow velocity distribution was almost constant except for the part that is very near the column wall, and it was almost uniform when the liquid flew from top to bottom direction with 4 cm/min of the velocity. The heaters scarcely influenced the temperature profile inside the column when the power applied to the heater simulated the decay heat of Am, Cm, and fission products (FPs). The decay heat generated in the column was transported to the effluents, and the temperature inside the column was kept almost constant. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Chromatography column system with controlled flow and temperature for engineering scale application [texte imprimé] / Sou Watanabe, Auteur ; Ichiro Goto, Auteur ; Yuichi Sano, Auteur . - 2011 . - 07 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Mots-clés : Actinides Chromatography Liquid metal fast breeder reactors Radioactive waste processing Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The Japan Atomic Energy Agency is conducting research and development study on the engineering scale extraction chromatography system, which uses silica-based adsorbents impregnated with an extractant for the minor actinides (Am and Cm) recovery from the high level liquid waste generated in the spent fast breeder reactor (FBR) fuel reprocessing, as a part of the fast reactor cycle technology development project. A bench scale testing system was made and provided for the first step of development. The column of the test system (inside diameter of 480 mm or 200 mm with height of 650 mm) was equipped with ports for six external sensors at its top, middle, and bottom levels for measuring the flow velocity or temperature, and for additional six heaters for simulating the decay heat of Am and Cm at the middle level of the column. The flow velocity distribution was almost constant except for the part that is very near the column wall, and it was almost uniform when the liquid flew from top to bottom direction with 4 cm/min of the velocity. The heaters scarcely influenced the temperature profile inside the column when the power applied to the heater simulated the decay heat of Am, Cm, and fission products (FPs). The decay heat generated in the column was transported to the effluents, and the temperature inside the column was kept almost constant. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Titre : Linear and nonlinear stability analysis of a supercritical natural circulation loop Type de document : texte imprimé Auteurs : Manish Sharma, Auteur ; P. K. Vijayan, Auteur ; D. S. Pilkhwal, Auteur Année de publication : 2011 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Finite volume methods Fission reactor coolants Flow instability Heat transfer Nuclear engineering computing Thermal analysis Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water cooled reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN has been developed employing supercritical water properties to carry out the steady-state and linear stability analysis of a SCW natural circulation loop (SCWNCL). The conservation equations of mass, momentum, and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure, and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been benchmarked against published results. Then the code has been extensively used for studying the effect of diameter, heater inlet temperature, and pressure on steady-state and stability behavior of a SCWNCL. A separate computer code, NOLSTA, has been developed, which investigates stability characteristics of supercritical natural circulation loop using nonlinear analysis. The conservation equations of mass, momentum, and energy in transient form were solved numerically using finite volume method. The stable, unstable, and neutrally stable points were identified by examining the amplitude of flow and temperature oscillations with time for a given set of operating conditions. The stability behavior of loop, predicted using nonlinear analysis has been compared with that obtained from linear analysis. The results show that the stability maps obtained by the two methods agree qualitatively. The present paper describes the linear and nonlinear stability analysis models and the results obtained in detail. DEWEY : 620.1 ISSN : 07472-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Linear and nonlinear stability analysis of a supercritical natural circulation loop [texte imprimé] / Manish Sharma, Auteur ; P. K. Vijayan, Auteur ; D. S. Pilkhwal, Auteur . - 2011 . - 09 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Mots-clés : Finite volume methods Fission reactor coolants Flow instability Heat transfer Nuclear engineering computing Thermal analysis Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water cooled reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN has been developed employing supercritical water properties to carry out the steady-state and linear stability analysis of a SCW natural circulation loop (SCWNCL). The conservation equations of mass, momentum, and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure, and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been benchmarked against published results. Then the code has been extensively used for studying the effect of diameter, heater inlet temperature, and pressure on steady-state and stability behavior of a SCWNCL. A separate computer code, NOLSTA, has been developed, which investigates stability characteristics of supercritical natural circulation loop using nonlinear analysis. The conservation equations of mass, momentum, and energy in transient form were solved numerically using finite volume method. The stable, unstable, and neutrally stable points were identified by examining the amplitude of flow and temperature oscillations with time for a given set of operating conditions. The stability behavior of loop, predicted using nonlinear analysis has been compared with that obtained from linear analysis. The results show that the stability maps obtained by the two methods agree qualitatively. The present paper describes the linear and nonlinear stability analysis models and the results obtained in detail. DEWEY : 620.1 ISSN : 07472-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 10 p.
Titre : Recent moisture separator reheater design technologies Type de document : texte imprimé Auteurs : Jun Manabe, Auteur ; Jiro Kasahara, Auteur ; Issaku Fujita, Auteur Année de publication : 2011 Article en page(s) : 10 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Heat exchangers Nuclear power stations Steam turbines Thermal stress cracking Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The moisture separator reheater (MSR) is a key piece of equipment in reheat systems in nuclear steam turbines that use saturated main steam, where it helps improve turbine efficiency and suppress flow-accelerated corrosion. Fundamental to achieving a compact, reliable MSR design are methods for predicting mist separator vane performance and suppressing tube drainage instability. First, we devised a method for predicting separator performance based on the observation of mist separation behavior under an air-water test. We then developed a method for predicting performance under steam conditions from air-water test data and verified it by means of a comparison with the actual results of a steam condition test. The instability of tube drainage associated with both subcooling and temperature oscillation at turbine partial load, which might adversely affect the seal welding of the tubes to the tube sheet due to thermal fatigue, was measured on an existing unit to clarify the behavior. We then developed a technique for increasing venting steam, which had been operating at a constant flow rate, to suppress instability and verified its effectiveness. Both methods were applied to current MSR models, which were adopted for nuclear power plant turbines commercially placed in service from 1984 to 2009, and the effectiveness of the methods was demonstrated. The separator vane mist carryover rate was less than 0.1%, and tube drainage instability was suppressed, demonstrating the effectiveness of the simple design concept of a two-flow U-tube instead of the prevailing four-flow U-tube design. We put forth a new concept in the design of MSRs for 1700 MW class advanced pressurized water reactor (APWR) units based on associated technologies, along with advanced technology for the compact design of pressure vessels and multidisciplinary optimum design for evaluating heat exchanger tube bundles. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Recent moisture separator reheater design technologies [texte imprimé] / Jun Manabe, Auteur ; Jiro Kasahara, Auteur ; Issaku Fujita, Auteur . - 2011 . - 10 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 10 p.
Mots-clés : Heat exchangers Nuclear power stations Steam turbines Thermal stress cracking Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The moisture separator reheater (MSR) is a key piece of equipment in reheat systems in nuclear steam turbines that use saturated main steam, where it helps improve turbine efficiency and suppress flow-accelerated corrosion. Fundamental to achieving a compact, reliable MSR design are methods for predicting mist separator vane performance and suppressing tube drainage instability. First, we devised a method for predicting separator performance based on the observation of mist separation behavior under an air-water test. We then developed a method for predicting performance under steam conditions from air-water test data and verified it by means of a comparison with the actual results of a steam condition test. The instability of tube drainage associated with both subcooling and temperature oscillation at turbine partial load, which might adversely affect the seal welding of the tubes to the tube sheet due to thermal fatigue, was measured on an existing unit to clarify the behavior. We then developed a technique for increasing venting steam, which had been operating at a constant flow rate, to suppress instability and verified its effectiveness. Both methods were applied to current MSR models, which were adopted for nuclear power plant turbines commercially placed in service from 1984 to 2009, and the effectiveness of the methods was demonstrated. The separator vane mist carryover rate was less than 0.1%, and tube drainage instability was suppressed, demonstrating the effectiveness of the simple design concept of a two-flow U-tube instead of the prevailing four-flow U-tube design. We put forth a new concept in the design of MSRs for 1700 MW class advanced pressurized water reactor (APWR) units based on associated technologies, along with advanced technology for the compact design of pressure vessels and multidisciplinary optimum design for evaluating heat exchanger tube bundles. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Titre : Competitiveness of small-medium, new generation reactors : a comparative study on decommissioning Type de document : texte imprimé Auteurs : Giorgio Locatelli, Auteur ; Mauro Mancini, Auteur Année de publication : 2011 Article en page(s) : 07 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Costing Fission reactor decommissioning Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Small-medium reactors (SMRs) are going to be important players in the worldwide nuclear renaissance. The economy of scale plays against the development of this kind of reactors, even if sometimes its influence is overestimated so that SMRs appear to have a levelized unit electricity cost significantly higher than large reactors (LRs). However, the economy of scale applies only if SMR designs are similar to that of LRs, but this is not the case since the small size allows for original design solutions not accessible to large sized reactors. The literature already presents studies showing how, under certain assumptions, the capital cost and the operation and maintenance cost of a site provided by one large reactor is quite similar to another site composed of four SMRs providing the same power output. However, the literature still lacks this kind of analysis on the decommissioning cost. The paper fills this gap, investigating the cost breakdown of a decommissioning project and providing a literature review about its cost estimate techniques and managerial approach. This paper identifies and briefly discusses the different cost drivers related to the decommissioning phase of a nuclear plant focusing the attention on those critical ones in the comparison between SMR and LR (economy of scale, multiple units in a single site, technical savings, and decommissioning strategy—“immediate decommissioning” or “deferred decommissioning”). The International Reactor Innovative and Secure reactor is used as the example of a SMR to quantify the effect of these drivers, but the analysis and conclusions are applicable to the whole spectrum of new small nuclear plants. The results show that when all of these factors are accounted for in a set of realistic and comparable configurations, and with the same power installed on the site, the decommissioning costs of a SMR with respect to a LR drop from three times higher to two times. If more than one large reactor is considered, the gap increases since the large reactor investment also reaps advantages from site sharing. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Competitiveness of small-medium, new generation reactors : a comparative study on decommissioning [texte imprimé] / Giorgio Locatelli, Auteur ; Mauro Mancini, Auteur . - 2011 . - 07 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Mots-clés : Costing Fission reactor decommissioning Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Small-medium reactors (SMRs) are going to be important players in the worldwide nuclear renaissance. The economy of scale plays against the development of this kind of reactors, even if sometimes its influence is overestimated so that SMRs appear to have a levelized unit electricity cost significantly higher than large reactors (LRs). However, the economy of scale applies only if SMR designs are similar to that of LRs, but this is not the case since the small size allows for original design solutions not accessible to large sized reactors. The literature already presents studies showing how, under certain assumptions, the capital cost and the operation and maintenance cost of a site provided by one large reactor is quite similar to another site composed of four SMRs providing the same power output. However, the literature still lacks this kind of analysis on the decommissioning cost. The paper fills this gap, investigating the cost breakdown of a decommissioning project and providing a literature review about its cost estimate techniques and managerial approach. This paper identifies and briefly discusses the different cost drivers related to the decommissioning phase of a nuclear plant focusing the attention on those critical ones in the comparison between SMR and LR (economy of scale, multiple units in a single site, technical savings, and decommissioning strategy—“immediate decommissioning” or “deferred decommissioning”). The International Reactor Innovative and Secure reactor is used as the example of a SMR to quantify the effect of these drivers, but the analysis and conclusions are applicable to the whole spectrum of new small nuclear plants. The results show that when all of these factors are accounted for in a set of realistic and comparable configurations, and with the same power installed on the site, the decommissioning costs of a SMR with respect to a LR drop from three times higher to two times. If more than one large reactor is considered, the gap increases since the large reactor investment also reaps advantages from site sharing. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 05 p.
Titre : Steady propagation of the vaporization front in metastable liquid Type de document : texte imprimé Auteurs : S. P. Aktershev, Auteur ; V. V. Ovchinnikov, Auteur Année de publication : 2011 Article en page(s) : 05 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Bubbles Convection Film boiling Metastable states Nuclear power stations Vaporisation Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The boiling up of a metastable liquid when the vaporization fronts appear is considered theoretically and experimentally. Boiling up occurs usually on the surface of a heater. At the first stage, the growth of a spherical vapor bubble is observed. If the temperature of liquid exceeds the threshold value, the vaporization fronts develop near to the line of contact of a vapor bubble and heater. Fronts of vaporization extend along a heater with constant speed. It is a direct transition from one phase convection to film boiling. Such scenario of crisis of a convective heat transfer is also possible in the nuclear reactor equipment. The model of steady propagation of the vaporization front is developed. The temperature and velocity of propagation of the interface are determined from the balance equations for the mass, momentum, and energy in the neighborhood of the vaporization front and the condition of stability of motion of the interface. It is shown that a solution of these equations exists only if the liquid is heated above a threshold value. The velocity of propagation of the vaporization front also has a threshold value. The calculated velocity of the interface motion and the threshold value of temperature are in reasonable agreement with available experimental data for various liquids within wide ranges of saturation pressures and temperatures of the overheated liquid. The developed model adequately describes the experimental data for various substances in a wide range of temperature of an overheated fluid. In this model, the steady propagation of the vaporization front is possible only if the temperature of a metastable liquid exceeds some threshold value. The velocity of the vaporization front also has a threshold value. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Steady propagation of the vaporization front in metastable liquid [texte imprimé] / S. P. Aktershev, Auteur ; V. V. Ovchinnikov, Auteur . - 2011 . - 05 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 05 p.
Mots-clés : Bubbles Convection Film boiling Metastable states Nuclear power stations Vaporisation Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The boiling up of a metastable liquid when the vaporization fronts appear is considered theoretically and experimentally. Boiling up occurs usually on the surface of a heater. At the first stage, the growth of a spherical vapor bubble is observed. If the temperature of liquid exceeds the threshold value, the vaporization fronts develop near to the line of contact of a vapor bubble and heater. Fronts of vaporization extend along a heater with constant speed. It is a direct transition from one phase convection to film boiling. Such scenario of crisis of a convective heat transfer is also possible in the nuclear reactor equipment. The model of steady propagation of the vaporization front is developed. The temperature and velocity of propagation of the interface are determined from the balance equations for the mass, momentum, and energy in the neighborhood of the vaporization front and the condition of stability of motion of the interface. It is shown that a solution of these equations exists only if the liquid is heated above a threshold value. The velocity of propagation of the vaporization front also has a threshold value. The calculated velocity of the interface motion and the threshold value of temperature are in reasonable agreement with available experimental data for various liquids within wide ranges of saturation pressures and temperatures of the overheated liquid. The developed model adequately describes the experimental data for various substances in a wide range of temperature of an overheated fluid. In this model, the steady propagation of the vaporization front is possible only if the temperature of a metastable liquid exceeds some threshold value. The velocity of the vaporization front also has a threshold value. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Titre : Experimental study on gas entrainment due to nonstationary vortex in a sodium cooled fast reactor : comparison of onset conditions between sodium and water Type de document : texte imprimé Auteurs : Nobuyuki Kimura, Auteur ; Toshiki Ezure, Auteur ; Hiroyuki Miyakoshi, Auteur Année de publication : 2011 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Gas cooled reactors Sodium Stratified flow Vortices Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development project. In the reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designed to reduce the construction cost. One of the thermal hydraulic problems in this design is gas entrainment at the free surface in the R/V. In most of past studies, water experiments were performed to investigate the gas entrainment in the reactor. It is necessary to evaluate an influence of fluid physical property on the gas entrainment phenomena. In this study, sodium experiments were carried out to clarify the onset criteria of the gas entrainment due to a free surface vortex. Water experiments using a test section in which geometry is the same as that in the sodium tests were also performed. The gas entrainment in water slightly tended to take place in comparison with that in sodium under low velocity conditions. Overall, the onset condition map on the lateral and downward flow velocities in the sodium and water experiments were in good agreement. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Experimental study on gas entrainment due to nonstationary vortex in a sodium cooled fast reactor : comparison of onset conditions between sodium and water [texte imprimé] / Nobuyuki Kimura, Auteur ; Toshiki Ezure, Auteur ; Hiroyuki Miyakoshi, Auteur . - 2011 . - 06 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Mots-clés : Gas cooled reactors Sodium Stratified flow Vortices Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development project. In the reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designed to reduce the construction cost. One of the thermal hydraulic problems in this design is gas entrainment at the free surface in the R/V. In most of past studies, water experiments were performed to investigate the gas entrainment in the reactor. It is necessary to evaluate an influence of fluid physical property on the gas entrainment phenomena. In this study, sodium experiments were carried out to clarify the onset criteria of the gas entrainment due to a free surface vortex. Water experiments using a test section in which geometry is the same as that in the sodium tests were also performed. The gas entrainment in water slightly tended to take place in comparison with that in sodium under low velocity conditions. Overall, the onset condition map on the lateral and downward flow velocities in the sodium and water experiments were in good agreement. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 08 p.
Titre : Measurement and analysis for rewetting velocity under post-BT conditions during anticipated operational occurrence of BWR Type de document : texte imprimé Auteurs : Sibamoto Yasuteru, Auteur ; Maruyama Yu, Auteur ; Nakamura Hideo, Auteur Année de publication : 2011 Article en page(s) : 08 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Boiling Fission reactor accidents Heat transfer Light water reactors Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : A series of experiments was performed for rewetting phenomena on dried-out fuel rod surfaces under post-boiling transition (post-BT) conditions with high-pressure and high-water flow rate simulating anticipated operational occurrences of a BWR. An analytical model for rewetting velocity, defined by a propagation velocity of a quench front, has been developed on the basis of the experimental results. The rewetting for the post-BT condition is characterized by the faster propagation of the quench front than that for reflood phase conditions during a postulated large-break loss-of-coolant accident. In order to provide an explanation of this characteristic, the present analytical model took an effect of a precursory cooling into account by modifying the existing correlation by Sun et al. (1975, “Effects of Precursory Cooling on Falling-Film Rewetting,” ASME J. Heat Transfer, 97, pp. 360–365), which is based on a one-dimensional analysis in a flow direction during the reflood phase. The present model demonstrates that the precursory cooling can significantly increase the rewetting velocity by more than an order of magnitude. Applying the experimental correlation developed in the separately conducted experiment into the heat transfer coefficient in the present model at a wet and a dry region with precursory cooling, our data of the rewetting velocity as well as the wall temperature profiles for the variable flow rates are successfully predicted. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Measurement and analysis for rewetting velocity under post-BT conditions during anticipated operational occurrence of BWR [texte imprimé] / Sibamoto Yasuteru, Auteur ; Maruyama Yu, Auteur ; Nakamura Hideo, Auteur . - 2011 . - 08 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 08 p.
Mots-clés : Boiling Fission reactor accidents Heat transfer Light water reactors Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : A series of experiments was performed for rewetting phenomena on dried-out fuel rod surfaces under post-boiling transition (post-BT) conditions with high-pressure and high-water flow rate simulating anticipated operational occurrences of a BWR. An analytical model for rewetting velocity, defined by a propagation velocity of a quench front, has been developed on the basis of the experimental results. The rewetting for the post-BT condition is characterized by the faster propagation of the quench front than that for reflood phase conditions during a postulated large-break loss-of-coolant accident. In order to provide an explanation of this characteristic, the present analytical model took an effect of a precursory cooling into account by modifying the existing correlation by Sun et al. (1975, “Effects of Precursory Cooling on Falling-Film Rewetting,” ASME J. Heat Transfer, 97, pp. 360–365), which is based on a one-dimensional analysis in a flow direction during the reflood phase. The present model demonstrates that the precursory cooling can significantly increase the rewetting velocity by more than an order of magnitude. Applying the experimental correlation developed in the separately conducted experiment into the heat transfer coefficient in the present model at a wet and a dry region with precursory cooling, our data of the rewetting velocity as well as the wall temperature profiles for the variable flow rates are successfully predicted. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Titre : Risk-informed approach for the regulation of decommissioning of nuclear facilities Type de document : texte imprimé Auteurs : Yukihiro Iguchi, Auteur ; Masami Kato, Auteur Année de publication : 2011 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission reactor decommissioning Fission reactor safety Risk management Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : During decommissioning of nuclear facilities, it is generally thought that the risk is relatively low after high activity inventory such as the spent fuel is removed. However, dismantlement works may be carried out with nonmultiple barriers with a nonregular process depending mainly on human activities. Moreover, fire or gas incidents caused by conventional industry methods may lead to accidents with radioactivity release. This means more attention is necessary for safer dismantlement, especially for nuclear reactors with high activity. Therefore, utilization of risk information based on risk assessment of the decommissioning was proposed. A method of risk assessment for decommissioning was developed and applied for the dismantlement of typical reactor facilities and nuclear fuel facilities (a uranium enrichment plant and a reprocessing plant). The results show that the consequences of such troubles are low but their occurrences are still not negligible. This fact is supported by past trouble cases. Taking into account the risk assessment results, a methodology to secure the safety of decommissioning was proposed. It consists of four steps, i.e., (1) risk-informed approach, (2) graded approach, (3) phased approach, and (4) layered approach and the results can be reflected to the management and regulation. The regulation means, for example, the review of decommissioning plan or the operational safety program, the periodic safety inspections and usual monitoring. The methodology can evaluate the risk level of decommissioning more objectively and enable reasonable regulation based on the risk level. This leads to the appropriate distribution of resources for safety enhancement. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Risk-informed approach for the regulation of decommissioning of nuclear facilities [texte imprimé] / Yukihiro Iguchi, Auteur ; Masami Kato, Auteur . - 2011 . - 09 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Mots-clés : Fission reactor decommissioning Fission reactor safety Risk management Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : During decommissioning of nuclear facilities, it is generally thought that the risk is relatively low after high activity inventory such as the spent fuel is removed. However, dismantlement works may be carried out with nonmultiple barriers with a nonregular process depending mainly on human activities. Moreover, fire or gas incidents caused by conventional industry methods may lead to accidents with radioactivity release. This means more attention is necessary for safer dismantlement, especially for nuclear reactors with high activity. Therefore, utilization of risk information based on risk assessment of the decommissioning was proposed. A method of risk assessment for decommissioning was developed and applied for the dismantlement of typical reactor facilities and nuclear fuel facilities (a uranium enrichment plant and a reprocessing plant). The results show that the consequences of such troubles are low but their occurrences are still not negligible. This fact is supported by past trouble cases. Taking into account the risk assessment results, a methodology to secure the safety of decommissioning was proposed. It consists of four steps, i.e., (1) risk-informed approach, (2) graded approach, (3) phased approach, and (4) layered approach and the results can be reflected to the management and regulation. The regulation means, for example, the review of decommissioning plan or the operational safety program, the periodic safety inspections and usual monitoring. The methodology can evaluate the risk level of decommissioning more objectively and enable reasonable regulation based on the risk level. This leads to the appropriate distribution of resources for safety enhancement. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Titre : Upgrading of waste heat for combined power and hydrogen production with nuclear reactors Type de document : texte imprimé Auteurs : C. Zamfirescu, Auteur ; G. F. Naterer, Auteur ; I. Dincer, Auteur Année de publication : 2011 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Chlorine Copper Fission reactor cooling Heat pumps Heat recovery Hydrogen production Waste heat Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : This paper presents a new heat upgrading method that utilizes waste heat from nuclear reactors for thermochemical water splitting with a copper-chlorine (Cu–Cl) cycle. Through combined power, hydrogen, and oxygen generation, the exergy efficiency of a power plant can be significantly augmented. The heat rejected to the environment for moderator cooling, a relatively small amount of low pressure superheated steam and a small fraction of generated power, are extracted from the nuclear reactor and used to drive a Cu–Cl hydrogen plant. More specifically, the moderator heat transfer at ~80°C is used as a source to a newly proposed vapor compression heat pump with a cascaded cycle, operating with retrograde fluids of cyclohexane (bottoming cycle) and biphenyl (topping supercritical cycle). Additionally, the heat pump uses as input the heat recovered from within the Cu–Cl cycle itself. This heat is recovered at two levels: ~80–130°C and ~250–485°C. This heat input is upgraded up to 600°C by work-to-heat conversion and then used to supply the endothermic water splitting process. The extracted steam is fed into the Cu–Cl cycle and split into hydrogen and oxygen as overall products. Electricity is partly used for an electrochemical process within the Cu–Cl cycle, and also partly for the heat pump compressors. This paper analyses the performance of the proposed heat pump and reports the exergy efficiency of the overall system. The proposed system is about 4% more efficient than generating electricity alone from the nuclear reactor. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Upgrading of waste heat for combined power and hydrogen production with nuclear reactors [texte imprimé] / C. Zamfirescu, Auteur ; G. F. Naterer, Auteur ; I. Dincer, Auteur . - 2011 . - 09 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Mots-clés : Chlorine Copper Fission reactor cooling Heat pumps Heat recovery Hydrogen production Waste heat Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : This paper presents a new heat upgrading method that utilizes waste heat from nuclear reactors for thermochemical water splitting with a copper-chlorine (Cu–Cl) cycle. Through combined power, hydrogen, and oxygen generation, the exergy efficiency of a power plant can be significantly augmented. The heat rejected to the environment for moderator cooling, a relatively small amount of low pressure superheated steam and a small fraction of generated power, are extracted from the nuclear reactor and used to drive a Cu–Cl hydrogen plant. More specifically, the moderator heat transfer at ~80°C is used as a source to a newly proposed vapor compression heat pump with a cascaded cycle, operating with retrograde fluids of cyclohexane (bottoming cycle) and biphenyl (topping supercritical cycle). Additionally, the heat pump uses as input the heat recovered from within the Cu–Cl cycle itself. This heat is recovered at two levels: ~80–130°C and ~250–485°C. This heat input is upgraded up to 600°C by work-to-heat conversion and then used to supply the endothermic water splitting process. The extracted steam is fed into the Cu–Cl cycle and split into hydrogen and oxygen as overall products. Electricity is partly used for an electrochemical process within the Cu–Cl cycle, and also partly for the heat pump compressors. This paper analyses the performance of the proposed heat pump and reports the exergy efficiency of the overall system. The proposed system is about 4% more efficient than generating electricity alone from the nuclear reactor. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Titre : Modeling of Fe–Cr martensitic steels corrosion in liquid lead alloys Type de document : texte imprimé Auteurs : F. Balbaud-Célérier, Auteur ; L. Martinelli, Auteur Année de publication : 2011 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Bismuth alloys Chromium alloys Corrosion Dissolving Fission reactor coolants Fission reactor cooling Iron alloys Lead alloys Liquid alloys Martensitic steel Oxidation Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Among the Generation IV systems, sodium fast reactors (SFRs) are promising and benefits of considerable technological experience. However, the availability and acceptability of the SFR are affected by the problems linked with the sodium-water reaction. One innovative solution to this problem is the replacement of the sodium in the secondary loops by an alternative liquid fluid. Among the fluids considered, lead-bismuth is at the moment being evaluated. Liquid lead-bismuth has been considerably studied in the frame of the research program on accelerator driven systems for transmutation applications. However, lead alloys are corrosive toward structural materials. The main parameters impacting the corrosion rate of Fe–Cr martensitic steels (considered as structural materials) are the nature of the steel (material side), temperature, liquid alloy velocity, and dissolved oxygen concentration (liquid alloy side). In this study, attention is focused on the behavior of Fe-9Cr steels, and more particularly, T91 martensitic steel. It has been shown that in the case of Fe–Cr martensitic steels, the corrosion process depends on the concentration of oxygen dissolved in Pb–Bi. For an oxygen concentration lower than the one necessary for magnetite formation (approximately <10−8 wt % at T[approximate]500°C for Fe-9Cr steels), corrosion proceeds by dissolution of the steel. For a higher oxygen content dissolved in Pb–Bi, corrosion proceeds by oxidation of the steel. These two corrosion processes have been experimentally and theoretically studied in CEA Saclay and also by other partners, leading to some corrosion modeling in order to predict the life duration of these materials as well as their limits of utilization. This study takes into account the two kinds of corrosion processes: dissolution and oxidation. In these two different processes, the lead alloy physico-chemical parameters are considered: the temperature and the liquid alloy velocity for both processes and the oxygen concentration for oxidation. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Modeling of Fe–Cr martensitic steels corrosion in liquid lead alloys [texte imprimé] / F. Balbaud-Célérier, Auteur ; L. Martinelli, Auteur . - 2011 . - 09 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Mots-clés : Bismuth alloys Chromium alloys Corrosion Dissolving Fission reactor coolants Fission reactor cooling Iron alloys Lead alloys Liquid alloys Martensitic steel Oxidation Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Among the Generation IV systems, sodium fast reactors (SFRs) are promising and benefits of considerable technological experience. However, the availability and acceptability of the SFR are affected by the problems linked with the sodium-water reaction. One innovative solution to this problem is the replacement of the sodium in the secondary loops by an alternative liquid fluid. Among the fluids considered, lead-bismuth is at the moment being evaluated. Liquid lead-bismuth has been considerably studied in the frame of the research program on accelerator driven systems for transmutation applications. However, lead alloys are corrosive toward structural materials. The main parameters impacting the corrosion rate of Fe–Cr martensitic steels (considered as structural materials) are the nature of the steel (material side), temperature, liquid alloy velocity, and dissolved oxygen concentration (liquid alloy side). In this study, attention is focused on the behavior of Fe-9Cr steels, and more particularly, T91 martensitic steel. It has been shown that in the case of Fe–Cr martensitic steels, the corrosion process depends on the concentration of oxygen dissolved in Pb–Bi. For an oxygen concentration lower than the one necessary for magnetite formation (approximately <10−8 wt % at T[approximate]500°C for Fe-9Cr steels), corrosion proceeds by dissolution of the steel. For a higher oxygen content dissolved in Pb–Bi, corrosion proceeds by oxidation of the steel. These two corrosion processes have been experimentally and theoretically studied in CEA Saclay and also by other partners, leading to some corrosion modeling in order to predict the life duration of these materials as well as their limits of utilization. This study takes into account the two kinds of corrosion processes: dissolution and oxidation. In these two different processes, the lead alloy physico-chemical parameters are considered: the temperature and the liquid alloy velocity for both processes and the oxygen concentration for oxidation. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Titre : A comparative study of single-phase, two-phase, and supercritical natural circulation in a rectangular loop Type de document : texte imprimé Auteurs : P. K. Vijayan, Auteur ; M. Sharma, Auteur ; D. S. Pilkhwal, Auteur Année de publication : 2011 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission reactor cooling Heat transfer Two-phase flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : A one-dimensional theoretical model has been used to analyze the steady state and stability performance of a single-phase, two-phase, and supercritical natural circulation in a uniform diameter rectangular loop. Parametric influences of diameter, inlet temperature, and system pressure on the steady state and stability performance have been studied. In the single-phase liquid filled region, the flow rate is found to increase monotonically with power. On the other hand, the flow rate in two-phase natural circulation systems is found to initially increase, reach a peak, and then decrease with power. For the supercritical region also, the steady state behavior is found to be similar to that of the two-phase region. However, if the heater inlet temperature is beyond the pseudo critical value, then the performance is similar to single-phase loops. Also, the supercritical natural circulation flow rate decreases drastically during this condition. With an increase in loop diameter, the flow rate is found to enhance for all the three regions of operation. Pressure has a significant influence on the flow rate in the two-phase region, marginal effect in the supercritical region, and practically no effect in the single-phase region. With the increase in loop diameter, operation in the single-phase and supercritical regions is found to destabilize, whereas the two-phase loops are found to stabilize. Again, pressure has a significant influence on stability in the two-phase region. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] A comparative study of single-phase, two-phase, and supercritical natural circulation in a rectangular loop [texte imprimé] / P. K. Vijayan, Auteur ; M. Sharma, Auteur ; D. S. Pilkhwal, Auteur . - 2011 . - 06 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Mots-clés : Fission reactor cooling Heat transfer Two-phase flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : A one-dimensional theoretical model has been used to analyze the steady state and stability performance of a single-phase, two-phase, and supercritical natural circulation in a uniform diameter rectangular loop. Parametric influences of diameter, inlet temperature, and system pressure on the steady state and stability performance have been studied. In the single-phase liquid filled region, the flow rate is found to increase monotonically with power. On the other hand, the flow rate in two-phase natural circulation systems is found to initially increase, reach a peak, and then decrease with power. For the supercritical region also, the steady state behavior is found to be similar to that of the two-phase region. However, if the heater inlet temperature is beyond the pseudo critical value, then the performance is similar to single-phase loops. Also, the supercritical natural circulation flow rate decreases drastically during this condition. With an increase in loop diameter, the flow rate is found to enhance for all the three regions of operation. Pressure has a significant influence on the flow rate in the two-phase region, marginal effect in the supercritical region, and practically no effect in the single-phase region. With the increase in loop diameter, operation in the single-phase and supercritical regions is found to destabilize, whereas the two-phase loops are found to stabilize. Again, pressure has a significant influence on stability in the two-phase region. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Titre : Three dimensional modeling of the hydrodynamics of oblique droplet-hot wall interactions during the reflood phase after a LOCA Type de document : texte imprimé Auteurs : D. Chatzikyriakou, Auteur ; S. P. Walker, Auteur ; B. Belhouachi, Auteur Année de publication : 2011 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Computational fluid dynamics Convection Hydrodynamics Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : During the reflood phase, following a loss-of-coolant-accident (LOCA), the main mechanism for the precursory cooling of the fuel is by convective heat transfer to the vapor, with the vapor being cooled by the evaporation of the entrained saturated droplets. However, it is believed that the droplets that reach the rod could have an effect on this cooling process. Despite the fact that those droplets do not actually wet the fuel rod due to the formation of a vapor film that sustains them and prevents them from touching the wall, the temperature drop caused by the impingement of such water droplets on a very hot solid surface (whose temperature is beyond the Leidenfrost temperature (1966, “A Track About Some Qualities of Common Water,” Int. J. Heat Mass Transfer, 9, pp. 1153–1166)) is of the order of 30–150°C (2008, The Role of Entrained Droplets in Precursory Cooling During PWR Post-LOCA Reflood, TOPSAFE, Dubrovnik, Croatia, 1995, “Heat Transfer During Liquid Contact on Superheated Surfaces,” ASME J. Heat Transfer, 117, pp. 693–697). The associated heat flux is of the order of 105–107 W/m2 and the heat extracted is in the range of 0.05 J over the time period of the interaction (a few ms) (2008, The Role of Entrained Droplets in Precursory Cooling During PWR Post-LOCA Reflood, TOPSAFE, Dubrovnik, Croatia, 1995, “Heat Transfer During Liquid Contact on Superheated Surfaces,” ASME J. Heat Transfer, 117, pp. 693–697). The hydrodynamic behavior of the droplets upon impingement is reported to affect the heat transfer effectiveness of the droplets. In the dispersed flow regime the droplets are more likely to impinge on the hot surface at a very small angle sliding along the solid wall, still without actually touching it, and remaining in a close proximity for a much larger time period. This changes the heat transfer behavior of the droplet. Here, we investigate numerically the hydrodynamics of the impingement of such droplets on a hot solid surface at various incident angles and various velocities of approach. For our simulations, we use a computational fluid dynamics (CFD), finite-volume computational algorithm (TransAT©). The level set method is used for the tracking of the interface. We present three-dimensional results of those impinging droplets. The validation of our simulation is done against experimental data already available in the literature. Then, we compare the findings of those results with previous correlations. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Three dimensional modeling of the hydrodynamics of oblique droplet-hot wall interactions during the reflood phase after a LOCA [texte imprimé] / D. Chatzikyriakou, Auteur ; S. P. Walker, Auteur ; B. Belhouachi, Auteur . - 2011 . - 06 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Mots-clés : Computational fluid dynamics Convection Hydrodynamics Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : During the reflood phase, following a loss-of-coolant-accident (LOCA), the main mechanism for the precursory cooling of the fuel is by convective heat transfer to the vapor, with the vapor being cooled by the evaporation of the entrained saturated droplets. However, it is believed that the droplets that reach the rod could have an effect on this cooling process. Despite the fact that those droplets do not actually wet the fuel rod due to the formation of a vapor film that sustains them and prevents them from touching the wall, the temperature drop caused by the impingement of such water droplets on a very hot solid surface (whose temperature is beyond the Leidenfrost temperature (1966, “A Track About Some Qualities of Common Water,” Int. J. Heat Mass Transfer, 9, pp. 1153–1166)) is of the order of 30–150°C (2008, The Role of Entrained Droplets in Precursory Cooling During PWR Post-LOCA Reflood, TOPSAFE, Dubrovnik, Croatia, 1995, “Heat Transfer During Liquid Contact on Superheated Surfaces,” ASME J. Heat Transfer, 117, pp. 693–697). The associated heat flux is of the order of 105–107 W/m2 and the heat extracted is in the range of 0.05 J over the time period of the interaction (a few ms) (2008, The Role of Entrained Droplets in Precursory Cooling During PWR Post-LOCA Reflood, TOPSAFE, Dubrovnik, Croatia, 1995, “Heat Transfer During Liquid Contact on Superheated Surfaces,” ASME J. Heat Transfer, 117, pp. 693–697). The hydrodynamic behavior of the droplets upon impingement is reported to affect the heat transfer effectiveness of the droplets. In the dispersed flow regime the droplets are more likely to impinge on the hot surface at a very small angle sliding along the solid wall, still without actually touching it, and remaining in a close proximity for a much larger time period. This changes the heat transfer behavior of the droplet. Here, we investigate numerically the hydrodynamics of the impingement of such droplets on a hot solid surface at various incident angles and various velocities of approach. For our simulations, we use a computational fluid dynamics (CFD), finite-volume computational algorithm (TransAT©). The level set method is used for the tracking of the interface. We present three-dimensional results of those impinging droplets. The validation of our simulation is done against experimental data already available in the literature. Then, we compare the findings of those results with previous correlations. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Titre : FAST code system : review of recent developments and near-future plans Type de document : texte imprimé Auteurs : Konstantin Mikityuk, Auteur ; Jiri Krepel, Auteur ; Sandro Pelloni, Auteur Année de publication : 2011 Article en page(s) : 07 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Gas cooled reactors Liquid metal fast breeder reactors Nuclear engineering computing Transient analysis Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The FAST code system is currently being developed and used at the Paul Scherrer Institut for static and transient analysis of the main Generation 4 fast-spectrum reactor concepts: sodium-, helium-, and gas-cooled fast reactors. The code system includes the ERANOS code system for static neutronics calculations, as well as coupled TRACE/PARCS/FRED for neutron kinetics, thermal hydraulic, and fuel transient analysis. The paper presents the status of the recent developments in neutronics (new 3D procedure for equilibrium cycle simulation and new transient cross section generation procedure), in thermal hydraulics and chemistry (equations-of-state for new coolants, two-phase flow models for sodium, and new model for oxide layer buildup in heavy-metal flow), and in fuel behavior (new model for the dispersed gas-cooled fast reactor fuel). Near-future plans for the further development of FAST are outlined. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] FAST code system : review of recent developments and near-future plans [texte imprimé] / Konstantin Mikityuk, Auteur ; Jiri Krepel, Auteur ; Sandro Pelloni, Auteur . - 2011 . - 07 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Mots-clés : Gas cooled reactors Liquid metal fast breeder reactors Nuclear engineering computing Transient analysis Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The FAST code system is currently being developed and used at the Paul Scherrer Institut for static and transient analysis of the main Generation 4 fast-spectrum reactor concepts: sodium-, helium-, and gas-cooled fast reactors. The code system includes the ERANOS code system for static neutronics calculations, as well as coupled TRACE/PARCS/FRED for neutron kinetics, thermal hydraulic, and fuel transient analysis. The paper presents the status of the recent developments in neutronics (new 3D procedure for equilibrium cycle simulation and new transient cross section generation procedure), in thermal hydraulics and chemistry (equations-of-state for new coolants, two-phase flow models for sodium, and new model for oxide layer buildup in heavy-metal flow), and in fuel behavior (new model for the dispersed gas-cooled fast reactor fuel). Near-future plans for the further development of FAST are outlined. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 08 p.
Titre : Integration of degradation models into generation risk assessment : challenges and modeling approaches Type de document : texte imprimé Auteurs : Mikko I. Jyrkama, Auteur ; Mahesh D. Pandey, Auteur ; Stephen M. Hess, Auteur Année de publication : 2011 Article en page(s) : 08 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fault trees Nuclear power stations Power generation faults Power generation reliability Risk management Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The main objective of generation risk assessment (GRA) is to assess the impact of equipment unavailability and failures on the ability of the plant to produce power over time. The system reliability models employed for this purpose are based on the standard fault tree/event tree approach, which assumes failure rates to be constant. However, this ignores the impact of aging degradation and results in static estimates of expected generation loss. Component and equipment degradation not only increases the probability of failure over time, but also contributes to generation risk through increased unavailability and costs arising from greater requirement for inspection and replacement of degraded components. This paper discusses some of the key challenges associated with integrating the results of component degradation models into GRA. Because many analytical and simulation methods are subject to limitations, the methodology and modeling approach proposed in this work builds on the current GRA practice using the fault tree approach. The modeling of component degradation can be done separately at the fault tree cut set level, assuming the cut sets are independent and the component unavailabilities are relatively small. In order to capture the joint contribution of equipment failure and unavailability to generation risk, new risk-based importance measures are also developed using the concept of net present value. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Integration of degradation models into generation risk assessment : challenges and modeling approaches [texte imprimé] / Mikko I. Jyrkama, Auteur ; Mahesh D. Pandey, Auteur ; Stephen M. Hess, Auteur . - 2011 . - 08 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 08 p.
Mots-clés : Fault trees Nuclear power stations Power generation faults Power generation reliability Risk management Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The main objective of generation risk assessment (GRA) is to assess the impact of equipment unavailability and failures on the ability of the plant to produce power over time. The system reliability models employed for this purpose are based on the standard fault tree/event tree approach, which assumes failure rates to be constant. However, this ignores the impact of aging degradation and results in static estimates of expected generation loss. Component and equipment degradation not only increases the probability of failure over time, but also contributes to generation risk through increased unavailability and costs arising from greater requirement for inspection and replacement of degraded components. This paper discusses some of the key challenges associated with integrating the results of component degradation models into GRA. Because many analytical and simulation methods are subject to limitations, the methodology and modeling approach proposed in this work builds on the current GRA practice using the fault tree approach. The modeling of component degradation can be done separately at the fault tree cut set level, assuming the cut sets are independent and the component unavailabilities are relatively small. In order to capture the joint contribution of equipment failure and unavailability to generation risk, new risk-based importance measures are also developed using the concept of net present value. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Titre : A once-through fuel cycle for fast reactors Type de document : texte imprimé Auteurs : Kevan D. Weaver, Auteur ; John Gilleland, Auteur ; Charles Ahlfeld, Auteur Année de publication : 2011 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission reactor fuel Nuclear power Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : A paradigm shift has altered the design targets for advanced nuclear energy systems that use a fast neutron spectrum. Whereas designers previously emphasized the ability of fast reactors to extend global reserves of fissile fuels, the overriding desire now is for reactor technologies that are “cleaner, more efficient, less waste-intensive, and more proliferation-resistant.” (Cheney, 2001, “U.S. National Energy Policy,” National Energy Policy Development Group, Washington, DC) This shift in priorities, along with recent design advances that enable high fuel burnup even when using fuels that have been minimally enriched, creates an opportunity to use fast reactors in an open nuclear fuel cycle. One promising route to this goal exploits a phenomenon known as a traveling wave, which can attain high burnups without reprocessing. A traveling-wave reactor (TWR) breeds and uses its own fuel in place as it operates. Recent design work has demonstrated that TWRs could be fueled almost entirely by depleted or natural uranium, thus reducing the need for initial enrichment. The calculations described here show that a gigawatt-scale electric TWR can achieve a burnup of 20%, which is four to five times that realized in current light water reactors. Burnups as high as 50% appear feasible. The factors that contribute to these high burnups and the implications for materials design are discussed. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] A once-through fuel cycle for fast reactors [texte imprimé] / Kevan D. Weaver, Auteur ; John Gilleland, Auteur ; Charles Ahlfeld, Auteur . - 2011 . - 06 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Mots-clés : Fission reactor fuel Nuclear power Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : A paradigm shift has altered the design targets for advanced nuclear energy systems that use a fast neutron spectrum. Whereas designers previously emphasized the ability of fast reactors to extend global reserves of fissile fuels, the overriding desire now is for reactor technologies that are “cleaner, more efficient, less waste-intensive, and more proliferation-resistant.” (Cheney, 2001, “U.S. National Energy Policy,” National Energy Policy Development Group, Washington, DC) This shift in priorities, along with recent design advances that enable high fuel burnup even when using fuels that have been minimally enriched, creates an opportunity to use fast reactors in an open nuclear fuel cycle. One promising route to this goal exploits a phenomenon known as a traveling wave, which can attain high burnups without reprocessing. A traveling-wave reactor (TWR) breeds and uses its own fuel in place as it operates. Recent design work has demonstrated that TWRs could be fueled almost entirely by depleted or natural uranium, thus reducing the need for initial enrichment. The calculations described here show that a gigawatt-scale electric TWR can achieve a burnup of 20%, which is four to five times that realized in current light water reactors. Burnups as high as 50% appear feasible. The factors that contribute to these high burnups and the implications for materials design are discussed. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010)
Titre : Characteristics of the new embrittlement correlation method for the japanese reactor pressure vessel steels Type de document : texte imprimé Auteurs : Naoki Soneda, Auteur ; Akiyoshi Nomoto, Auteur Année de publication : 2011 Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Correlation methods Embrittlement Pressure vessels Steel Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Neutron irradiation embrittlement of reactor pressure vessel steels is an important aging issue for the long-term operation of light water reactors. A new embrittlement correlation method was developed by Central Research Institute of Electric Power Industry and the Japanese electric utilities in 2007. This method is primarily based on the fundamental understandings on the embrittlement mechanisms, i.e., microstructural changes were modeled by the mathematical form of rate equations, and the predicted microstructural changes were further correlated with the mechanical property changes in transition temperature region. The coefficients of the rate equations were optimized using the Japanese surveillance data of RPV embrittlement. This method was adopted as the revision of the Japanese code, JEAC4201-2007, in 2007. In this paper, after a brief explanation on the new correlation method, the predictions of the new method will be investigated through comparisons with the previous correlation, JEAC4201-2004, and the U.S. surveillance data in order to identify the characteristics of the new method. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Characteristics of the new embrittlement correlation method for the japanese reactor pressure vessel steels [texte imprimé] / Naoki Soneda, Auteur ; Akiyoshi Nomoto, Auteur . - 2011.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010)
Mots-clés : Correlation methods Embrittlement Pressure vessels Steel Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Neutron irradiation embrittlement of reactor pressure vessel steels is an important aging issue for the long-term operation of light water reactors. A new embrittlement correlation method was developed by Central Research Institute of Electric Power Industry and the Japanese electric utilities in 2007. This method is primarily based on the fundamental understandings on the embrittlement mechanisms, i.e., microstructural changes were modeled by the mathematical form of rate equations, and the predicted microstructural changes were further correlated with the mechanical property changes in transition temperature region. The coefficients of the rate equations were optimized using the Japanese surveillance data of RPV embrittlement. This method was adopted as the revision of the Japanese code, JEAC4201-2007, in 2007. In this paper, after a brief explanation on the new correlation method, the predictions of the new method will be investigated through comparisons with the previous correlation, JEAC4201-2004, and the U.S. surveillance data in order to identify the characteristics of the new method. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 08 p.
Titre : Industry practice for the neutron irradiation embrittlement of reactor pressure vessels in Japan Type de document : texte imprimé Auteurs : Norimichi Yamashita, Auteur ; Masanobu Iwasaki, Auteur ; Koji Dozaki, Auteur Année de publication : 2011 Article en page(s) : 08 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Ageing Embrittlement Failure analysis Fission reactors Fracture Nuclear power stations Steel Surveillance Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material aging issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation, based on the appropriate recognition of the neutron irradiation embrittlement, is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complemental use of the surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old boiling water reactor (BWR) RPV steels with high impurity content. Some discrepancies of pressurized water reactor (PWR) RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201-2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long term surveillance program,” which is to be applied for the operation beyond 40 years. The use of the reloading, reirradiation, and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201-2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Industry practice for the neutron irradiation embrittlement of reactor pressure vessels in Japan [texte imprimé] / Norimichi Yamashita, Auteur ; Masanobu Iwasaki, Auteur ; Koji Dozaki, Auteur . - 2011 . - 08 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 08 p.
Mots-clés : Ageing Embrittlement Failure analysis Fission reactors Fracture Nuclear power stations Steel Surveillance Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material aging issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation, based on the appropriate recognition of the neutron irradiation embrittlement, is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complemental use of the surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old boiling water reactor (BWR) RPV steels with high impurity content. Some discrepancies of pressurized water reactor (PWR) RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201-2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long term surveillance program,” which is to be applied for the operation beyond 40 years. The use of the reloading, reirradiation, and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201-2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 05 p.
Titre : Numerical solution on spherical vacuum bubble collapse using MPS method Type de document : texte imprimé Auteurs : Wen xi Tian, Auteur ; Sui-zheng Qiu, Auteur ; Guang-hui Su, Auteur Année de publication : 2011 Article en page(s) : 05 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Bubbles Finite difference methods Finite element analysis Finite volume methods Hydrodynamics Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Single vacuum bubble collapse in subcooled water has been simulated using the moving particle semi-implicit (MPS) method in the present study. The liquid is described using moving particles, and the bubble-liquid interface was set to be the vacuum pressure boundary without interfacial heat mass transfer. The topological shape of the vacuum bubble is determined according to the location of interfacial particles. The time dependent bubble diameter, interfacial velocity, and bubble collapse time were obtained within a wide parametric range. Comparison with Rayleigh's prediction indicates a good consistency, which validates the applicability and accuracy of the MPS method. The potential void-induced water hammer pressure pulse was also evaluated, which is instructive for the cavitation erosion study. The present paper discovers fundamental characteristics of vacuum bubble hydrodynamics, and it is also instructive for further applications of the MPS method to complicated bubble dynamics. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Numerical solution on spherical vacuum bubble collapse using MPS method [texte imprimé] / Wen xi Tian, Auteur ; Sui-zheng Qiu, Auteur ; Guang-hui Su, Auteur . - 2011 . - 05 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 05 p.
Mots-clés : Bubbles Finite difference methods Finite element analysis Finite volume methods Hydrodynamics Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Single vacuum bubble collapse in subcooled water has been simulated using the moving particle semi-implicit (MPS) method in the present study. The liquid is described using moving particles, and the bubble-liquid interface was set to be the vacuum pressure boundary without interfacial heat mass transfer. The topological shape of the vacuum bubble is determined according to the location of interfacial particles. The time dependent bubble diameter, interfacial velocity, and bubble collapse time were obtained within a wide parametric range. Comparison with Rayleigh's prediction indicates a good consistency, which validates the applicability and accuracy of the MPS method. The potential void-induced water hammer pressure pulse was also evaluated, which is instructive for the cavitation erosion study. The present paper discovers fundamental characteristics of vacuum bubble hydrodynamics, and it is also instructive for further applications of the MPS method to complicated bubble dynamics. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Titre : Design, development, testing and ualification of diverse safety rod and its drive mechanism for a prototype fast breeder reactor Type de document : texte imprimé Auteurs : R. Vijayashree, Auteur ; R. Veerasamy, Auteur ; Sudheer Patri, Auteur Année de publication : 2011 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Liquid metal fast breeder reactors Safety systems Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Prototype fast breeder reactor is U–PuO2 fueled sodium cooled pool type fast reactor and it is currently under construction at Kalpakkam, India. Prototype fast breeder reactor is equipped with two independent fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms, and neutron absorbing rods. The two shutdown systems of prototype fast breeder reactor are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of prototype fast breeder reactor are called as diverse safety rods (DSRs) and their drive mechanisms are called as diverse safety rod drive mechanisms (DSRDMs). DSRs are normally parked above active core by DSRDMs. On receiving scram signal, the electromagnet of DSRDM is de-energized and it facilitates fast shutdown of the reactor by dropping the DSR into the active core. For the development of prototypes of DSR and DSRDM, three phases of testing, namely, individual component testing, integrated functional testing in room temperature, and endurance testing at high temperature sodium, were done. The electromagnet of DSRDM has been separately tested at room temperature, in furnace, and in sodium. Specimens simulating the contact conditions between electromagnet and armature of DSR have been tested to rule out self-welding possibility. The prototype of DSR has been tested in flowing water to determine the pressure drop and drop time. The functional testing of the integrated prototype DSRDM and DSR in aligned and misaligned conditions in air/water has been completed. The performance testing of the integrated system in sodium has been done in three campaigns. During the third campaign of sodium testing, the performance of the system has been verified with 30 mm misalignment at various temperatures. The third campaign has qualified the system for 10 years of operation in reactor. This paper presents design, development, testing, and qualification of the prototype DSR and DSRDM. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual and detailed design features are explained with the help of figures. Details on material of construction are given at appropriate places. Test plans and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are briefed. Brief explanation of test setups and typical test results are also given. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Design, development, testing and ualification of diverse safety rod and its drive mechanism for a prototype fast breeder reactor [texte imprimé] / R. Vijayashree, Auteur ; R. Veerasamy, Auteur ; Sudheer Patri, Auteur . - 2011 . - 09 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 09 p.
Mots-clés : Liquid metal fast breeder reactors Safety systems Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Prototype fast breeder reactor is U–PuO2 fueled sodium cooled pool type fast reactor and it is currently under construction at Kalpakkam, India. Prototype fast breeder reactor is equipped with two independent fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms, and neutron absorbing rods. The two shutdown systems of prototype fast breeder reactor are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of prototype fast breeder reactor are called as diverse safety rods (DSRs) and their drive mechanisms are called as diverse safety rod drive mechanisms (DSRDMs). DSRs are normally parked above active core by DSRDMs. On receiving scram signal, the electromagnet of DSRDM is de-energized and it facilitates fast shutdown of the reactor by dropping the DSR into the active core. For the development of prototypes of DSR and DSRDM, three phases of testing, namely, individual component testing, integrated functional testing in room temperature, and endurance testing at high temperature sodium, were done. The electromagnet of DSRDM has been separately tested at room temperature, in furnace, and in sodium. Specimens simulating the contact conditions between electromagnet and armature of DSR have been tested to rule out self-welding possibility. The prototype of DSR has been tested in flowing water to determine the pressure drop and drop time. The functional testing of the integrated prototype DSRDM and DSR in aligned and misaligned conditions in air/water has been completed. The performance testing of the integrated system in sodium has been done in three campaigns. During the third campaign of sodium testing, the performance of the system has been verified with 30 mm misalignment at various temperatures. The third campaign has qualified the system for 10 years of operation in reactor. This paper presents design, development, testing, and qualification of the prototype DSR and DSRDM. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual and detailed design features are explained with the help of figures. Details on material of construction are given at appropriate places. Test plans and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are briefed. Brief explanation of test setups and typical test results are also given. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Titre : Photographic study on bubble motion in subcooled pool boiling Type de document : texte imprimé Auteurs : Tomio Okawa, Auteur ; Takahiro Harada, Auteur ; Yuta Kotsusa, Auteur Année de publication : 2011 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Boiling Bubbles Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Using a static contact angle of a vertical heated wall as a main experimental parameter, a photographic study was carried out to elucidate the mechanisms to determine the vapor bubble dynamics during subcooled pool boiling. The test fluid was distilled water and the experiments were performed under the atmospheric pressure; liquid subcooling was set to around 5 K. To enable clear observation of bubble behavior with a high speed camera, the experiments were conducted in an isolated bubble regime near the onset of nucleate boiling. Distinctly different bubble behaviors were observed on hydrophobic and hydrophilic surfaces: the bubbles were adhered to the surface for a long period of time when the contact angle was large while lifted-off the surface within a short period of time after the nucleation when the contact angle was small. Since buoyancy does not remove the bubble from the vertical surface, the mechanisms of bubble lift-off were investigated. It was indicated that the change in bubble shape induced by the surface tension force, unsteady growth force, and local liquid flow induced by heterogeneous condensation around the bubble are considered to promote the bubble lift-off while the surface tension force acting on the three-phase common line prevented the lift-off. Effects of the surface wettability on the lift-off bubble diameter, the elapsed time from the nucleation at the lift-off, and the condensation rate after the lift-off were also investigated. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Photographic study on bubble motion in subcooled pool boiling [texte imprimé] / Tomio Okawa, Auteur ; Takahiro Harada, Auteur ; Yuta Kotsusa, Auteur . - 2011 . - 06 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Mots-clés : Boiling Bubbles Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Using a static contact angle of a vertical heated wall as a main experimental parameter, a photographic study was carried out to elucidate the mechanisms to determine the vapor bubble dynamics during subcooled pool boiling. The test fluid was distilled water and the experiments were performed under the atmospheric pressure; liquid subcooling was set to around 5 K. To enable clear observation of bubble behavior with a high speed camera, the experiments were conducted in an isolated bubble regime near the onset of nucleate boiling. Distinctly different bubble behaviors were observed on hydrophobic and hydrophilic surfaces: the bubbles were adhered to the surface for a long period of time when the contact angle was large while lifted-off the surface within a short period of time after the nucleation when the contact angle was small. Since buoyancy does not remove the bubble from the vertical surface, the mechanisms of bubble lift-off were investigated. It was indicated that the change in bubble shape induced by the surface tension force, unsteady growth force, and local liquid flow induced by heterogeneous condensation around the bubble are considered to promote the bubble lift-off while the surface tension force acting on the three-phase common line prevented the lift-off. Effects of the surface wettability on the lift-off bubble diameter, the elapsed time from the nucleation at the lift-off, and the condensation rate after the lift-off were also investigated. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Titre : Study on the coupled neutronic and thermal-hydraulic characteristics of the new concept molten salt reactor Type de document : texte imprimé Auteurs : Peng Wang, Auteur ; Libo Qian, Auteur ; Dalin Zhang, Auteur Année de publication : 2011 Article en page(s) : 07 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission reactor coolants Fission reactor design Fission reactor fuel Neutrons Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The new concept molten salt reactor is the only liquid-fuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutron precursors are drifted by the fuel flow, which leads the spread of delayed neutrons' distribution to noncore parts of the primary circuit, and it also results in reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the molten salt reactor are quite different from the conventional nuclear reactors using solid fissile materials. Besides, there is no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particles considering the flow effect of the fuel salt in the molten salt reactor, while the thermal-hydraulic model applies the fundamental conservation laws: the mass, momentum, and energy conservation equations. Then, the neutronic and thermal-hydraulic calculations are coupled and the influences of inflow temperature and flow velocity on the reactor physical properties are obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor keff, while the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron fluxes and delayed neutron precursors slightly, and makes a significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady-state, but affects the delayed neutron precursors' distribution significantly. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Study on the coupled neutronic and thermal-hydraulic characteristics of the new concept molten salt reactor [texte imprimé] / Peng Wang, Auteur ; Libo Qian, Auteur ; Dalin Zhang, Auteur . - 2011 . - 07 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 07 p.
Mots-clés : Fission reactor coolants Fission reactor design Fission reactor fuel Neutrons Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The new concept molten salt reactor is the only liquid-fuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutron precursors are drifted by the fuel flow, which leads the spread of delayed neutrons' distribution to noncore parts of the primary circuit, and it also results in reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the molten salt reactor are quite different from the conventional nuclear reactors using solid fissile materials. Besides, there is no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particles considering the flow effect of the fuel salt in the molten salt reactor, while the thermal-hydraulic model applies the fundamental conservation laws: the mass, momentum, and energy conservation equations. Then, the neutronic and thermal-hydraulic calculations are coupled and the influences of inflow temperature and flow velocity on the reactor physical properties are obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor keff, while the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron fluxes and delayed neutron precursors slightly, and makes a significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady-state, but affects the delayed neutron precursors' distribution significantly. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Titre : Core melt solidification characteristics in PRV lower head-experimental results from LIVE tests Type de document : texte imprimé Auteurs : Xiaoyang Gaus-Liu, Auteur ; Alexei Miassoedov, Auteur ; Thomas Cron, Auteur Année de publication : 2011 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission reactor cooling Light water reactors Solidification Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Core melt solidification phenomena in the lower plenum of pressurized reactor vessel during external reactor vessel cooling is investigated in late in-vessel phase experiment tests under different external cooling conditions and melt pouring positions. The melt solidification behavior, which has not yet been given sufficient attention, is an important issue since it influences not only the transient but also the steady state of melt pool thermal hydraulics. A noneutectic melt (80 mol % KNO3–20 mol % NaNO3) was used to simulate the core melt. It has been found out that when the vessel is cooled with water during the whole test period (water cooling), the cooling is more effective than the case that the vessel lower head is first cooled with air and flooded by water (air/water cooling). Water cooling at the beginning leads to faster buildup of crust layer on the vessel inner wall and lower crust thermal conductivity compared with air/water cooling. In comparison with the air/water cooling, the water cooling also achieves shorter time period of crust growth. During the solidification period in all tests, the constitutional supercooling condition is fulfilled. Pouring position near the vessel wall results in considerable asymmetry in the heat flux distribution through the vessel wall. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Core melt solidification characteristics in PRV lower head-experimental results from LIVE tests [texte imprimé] / Xiaoyang Gaus-Liu, Auteur ; Alexei Miassoedov, Auteur ; Thomas Cron, Auteur . - 2011 . - 06 p.
Génie Mécanique
Langues : Anglais (eng)
in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 132 N° 10 (Octobre 2010) . - 06 p.
Mots-clés : Fission reactor cooling Light water reactors Solidification Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Core melt solidification phenomena in the lower plenum of pressurized reactor vessel during external reactor vessel cooling is investigated in late in-vessel phase experiment tests under different external cooling conditions and melt pouring positions. The melt solidification behavior, which has not yet been given sufficient attention, is an important issue since it influences not only the transient but also the steady state of melt pool thermal hydraulics. A noneutectic melt (80 mol % KNO3–20 mol % NaNO3) was used to simulate the core melt. It has been found out that when the vessel is cooled with water during the whole test period (water cooling), the cooling is more effective than the case that the vessel lower head is first cooled with air and flooded by water (air/water cooling). Water cooling at the beginning leads to faster buildup of crust layer on the vessel inner wall and lower crust thermal conductivity compared with air/water cooling. In comparison with the air/water cooling, the water cooling also achieves shorter time period of crust growth. During the solidification period in all tests, the constitutional supercooling condition is fulfilled. Pouring position near the vessel wall results in considerable asymmetry in the heat flux distribution through the vessel wall. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]
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