Dépouillements

Dynamics simulations of a graphite block under longitudinal impact / Gyeongho Kim in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 07 p. Titre : Dynamics simulations of a graphite block under longitudinal impact Type de document : texte imprimé Auteurs : Gyeongho Kim, Auteur ; Dong-Ok Kim, Auteur ; Woo-Seok Choi, Auteur Année de publication : 2012 Article en page(s) : 07 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Earthquake  engineering  Fission  reactor  materials  Gas  cooled  reactors  Graphite  Impact  (mechanical) Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Graphite blocks are important core components of the high temperature gas-cooled reactor. As these blocks are simply stacked in array, collisions among neighboring components may occur during earthquakes or accidents. Thus, it is important to develop a reliable seismic model of the stacked graphite blocks and have them designed to maintain their structural integrity during the anticipated occurrences. Various aspects involved in modeling and calculating impact-contact dynamics can affect the resulting behavior of the graphite block. These include mesh size, time step, contact behavior, mechanical constraint formulation of impact-contact analysis, etc. This work is dedicated to perform comparative studies and the effects of these parameters will be identified. The insights obtained through these studies will help build a realistic impact-contact model of the graphite block from which a lumped or reduced dynamics model will be developed for the seismic analysis of the reactor including these graphite components. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Dynamics simulations of a graphite block under longitudinal impact [texte imprimé] / Gyeongho Kim, Auteur ; Dong-Ok Kim, Auteur ; Woo-Seok Choi, Auteur . - 2012 . - 07 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 07 p. Mots-clés : Earthquake  engineering  Fission  reactor  materials  Gas  cooled  reactors  Graphite  Impact  (mechanical) Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Graphite blocks are important core components of the high temperature gas-cooled reactor. As these blocks are simply stacked in array, collisions among neighboring components may occur during earthquakes or accidents. Thus, it is important to develop a reliable seismic model of the stacked graphite blocks and have them designed to maintain their structural integrity during the anticipated occurrences. Various aspects involved in modeling and calculating impact-contact dynamics can affect the resulting behavior of the graphite block. These include mesh size, time step, contact behavior, mechanical constraint formulation of impact-contact analysis, etc. This work is dedicated to perform comparative studies and the effects of these parameters will be identified. The insights obtained through these studies will help build a realistic impact-contact model of the graphite block from which a lumped or reduced dynamics model will be developed for the seismic analysis of the reactor including these graphite components. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Reactivity accident in a high temperature gas-cooled reactor due to inadvertent withdrawal of control rod / Zheng Yanhua in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Titre : Reactivity accident in a high temperature gas-cooled reactor due to inadvertent withdrawal of control rod Type de document : texte imprimé Auteurs : Zheng Yanhua, Auteur ; Shi Lei, Auteur Année de publication : 2012 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission  reactor  accidents  Gas  cooled  reactors  Power  engineering  computing  Rods  (structures) Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese pebble-bed modular high temperature gas-cooled reactor (HTR-PM) with single module power of 250 MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper (e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not) and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature, and xenon concentration are studied and compared in detail between these different cases. The calculating results show that the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way so that the temperature limits of the fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Reactivity accident in a high temperature gas-cooled reactor due to inadvertent withdrawal of control rod [texte imprimé] / Zheng Yanhua, Auteur ; Shi Lei, Auteur . - 2012 . - 06 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Mots-clés : Fission  reactor  accidents  Gas  cooled  reactors  Power  engineering  computing  Rods  (structures) Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Reactivity accident due to inadvertent withdrawal of the control rod is one kind of the design basis accident for high temperature gas-cooled reactors, which should be analyzed carefully in order to validate the reactor inherent safety properties. Based on the preliminary design of the Chinese pebble-bed modular high temperature gas-cooled reactor (HTR-PM) with single module power of 250 MW, several cases of reactivity accident has been studied by the help of the software TINTE in the paper (e.g., the first scram signal works or not, the absorber balls (secondary shutdown units) drop or not) and the ATWS situation is also taken into account. The dynamic processes of the important parameters including reactor power, fuel temperature, and xenon concentration are studied and compared in detail between these different cases. The calculating results show that the decay heat during the reactivity accidents can be removed from the reactor core solely by means of physical processes in a passive way so that the temperature limits of the fuel element and other components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. This will be helpful to the further detail design of the HTR-PM demonstrating power plant project. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Extraction of actinides and lanthanides by supercritical fluid / Liyang Zhu in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Titre : Extraction of actinides and lanthanides by supercritical fluid Type de document : texte imprimé Auteurs : Liyang Zhu, Auteur ; Wuhua Duan, Auteur ; Jingming Xu, Auteur Année de publication : 2012 Article en page(s) : 08 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Actinides  Fission  reactor  fuel  reprocessing  Radioactive  waste  processing  Rare  earth  metals  Sustainable  development Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Reprocessing of used nuclear fuel and nuclear waste management are important issues for the sustainable development of nuclear energy. It is necessary to develop novel nuclear waste treatment technologies to meet the goal of minimizing the secondary liquid waste. Supercritical fluids are considered green solvents in chemical engineering process. Moreover, extraction of metal ions by supercritical fluid is achieved. It gains growing interest to treat nuclear waste using supercritical fluid extraction recently because it can greatly decrease the secondary liquid waste with high radioactivity. During the past 2 decades, extraction of actinides and lanthanides by supercritical fluid has been intensively studied in many countries, and many important progresses have been made. However, the prospect of industrial application of supercritical fluid extraction technology in nuclear waste management is still unclear. In this paper, extraction of actinides and lanthanides from various matrices or from their oxides by supercritical fluid including the experimental results, extraction mechanism, and kinetic process was reviewed. The engineering demonstration projects were introduced. The trend of industrial application of supercritical fluid extraction technology in nuclear waste management was also discussed. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Extraction of actinides and lanthanides by supercritical fluid [texte imprimé] / Liyang Zhu, Auteur ; Wuhua Duan, Auteur ; Jingming Xu, Auteur . - 2012 . - 08 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Mots-clés : Actinides  Fission  reactor  fuel  reprocessing  Radioactive  waste  processing  Rare  earth  metals  Sustainable  development Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Reprocessing of used nuclear fuel and nuclear waste management are important issues for the sustainable development of nuclear energy. It is necessary to develop novel nuclear waste treatment technologies to meet the goal of minimizing the secondary liquid waste. Supercritical fluids are considered green solvents in chemical engineering process. Moreover, extraction of metal ions by supercritical fluid is achieved. It gains growing interest to treat nuclear waste using supercritical fluid extraction recently because it can greatly decrease the secondary liquid waste with high radioactivity. During the past 2 decades, extraction of actinides and lanthanides by supercritical fluid has been intensively studied in many countries, and many important progresses have been made. However, the prospect of industrial application of supercritical fluid extraction technology in nuclear waste management is still unclear. In this paper, extraction of actinides and lanthanides from various matrices or from their oxides by supercritical fluid including the experimental results, extraction mechanism, and kinetic process was reviewed. The engineering demonstration projects were introduced. The trend of industrial application of supercritical fluid extraction technology in nuclear waste management was also discussed. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Investigation of the beltline welding seam and base metal of the greifswald WWER-440 unit 1 reactor pressure vessel / Jan Schuhknecht in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Titre : Investigation of the beltline welding seam and base metal of the greifswald WWER-440 unit 1 reactor pressure vessel Type de document : texte imprimé Auteurs : Jan Schuhknecht, Auteur ; Hans-Werner Viehrig, Auteur ; Udo Rindelhardt, Auteur Année de publication : 2012 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Annealing  Dynamic  testing  Fracture  Nuclear  power  stations  Steel  Welding Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The investigation of reactor pressure vessel (RPV) materials from decommissioned nuclear power plants (NPPs) offers the unique opportunity to scrutinize the irradiation behavior under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterization. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I); irradiated and recovery annealed (IA); and irradiated, recovery annealed, and re-irradiated (IAI). The working program is focused on the characterization of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the American Society for Testing of Materials (ASTM) Test Standard E1921–08 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step, the trepans taken from the RPV Greifswald unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the master curve (MC) approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behavior. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on subsize Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy-V transition temperature TT41J estimated with results of subsize specimens after the recovery annealing was confirmed by the testing of standard Charpy-V-notch specimens. The evaluated TT41J shows a better accordance with the irradiation fluence along the wall thickness than the master curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during two campaign operations can be assumed to be low for the weld and base metal. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Investigation of the beltline welding seam and base metal of the greifswald WWER-440 unit 1 reactor pressure vessel [texte imprimé] / Jan Schuhknecht, Auteur ; Hans-Werner Viehrig, Auteur ; Udo Rindelhardt, Auteur . - 2012 . - 06 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Mots-clés : Annealing  Dynamic  testing  Fracture  Nuclear  power  stations  Steel  Welding Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The investigation of reactor pressure vessel (RPV) materials from decommissioned nuclear power plants (NPPs) offers the unique opportunity to scrutinize the irradiation behavior under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterization. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I); irradiated and recovery annealed (IA); and irradiated, recovery annealed, and re-irradiated (IAI). The working program is focused on the characterization of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the American Society for Testing of Materials (ASTM) Test Standard E1921–08 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step, the trepans taken from the RPV Greifswald unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the master curve (MC) approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behavior. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on subsize Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy-V transition temperature TT41J estimated with results of subsize specimens after the recovery annealing was confirmed by the testing of standard Charpy-V-notch specimens. The evaluated TT41J shows a better accordance with the irradiation fluence along the wall thickness than the master curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during two campaign operations can be assumed to be low for the weld and base metal. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Study of interfacial friction force for bubble flows in a 2×1 rod channel simplifying BWR / Akimaro Kawahara in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Titre : Study of interfacial friction force for bubble flows in a 2×1 rod channel simplifying BWR Type de document : texte imprimé Auteurs : Akimaro Kawahara, Auteur ; Michio Sadatomi, Auteur ; Yutaro Nakamoto, Auteur Année de publication : 2012 Article en page(s) : 08 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Bubbles  Channel  flow  Drag  Fission  reactors  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Most of the recent subchannel analysis codes are based on a multifluid model, and an accurate evaluation of the constitutive equations in the model is essential. In order to get an accurate interfacial friction force in two-phase bubble flows, experimental data on drag coefficient and interfacial area concentration have been obtained for air-water flows in a 2×1 rod channel simplifying a boiling water nuclear reactor fuel rod bundle. In order to know the effects of liquid properties on the data, the temperature of the test water was changed from 18°C to 50°C. The data are compared with the existing correlations reported in literatures. As a result, the semitheoretical correlation of Hibiki and Ishii (2001, “Interfacial Area Concentration in Steady Fully-Developed Bubbly Flow,” Int. J. Heat Mass Transfer, 44, pp. 3443–3461) was found to give the best prediction against the present interfacial area concentration data. The correlation of Delhaye and Bricard (1994, “Interfacial Area in Bubbly Flow: Experimental Data and Correlations,” Nucl. Eng. Des., 151, pp. 65–77) also gave a reasonably good prediction if their correlation was modified by incorporating liquid property effects. As for the drag coefficient, no correlation exists, which can predict the present data well. Therefore, we developed a new correlation, including three dimensionless numbers, i.e., bubble capillary number, Morton number, and Eötvös number. The correlation predicted the data of Liu et al. (2008, “Drag Coefficient in One-Dimensional Two-Group Two-Fluid Model,” Int. J. Heat Fluid Flow, 29, pp. 1402–1410) as well as the present data well. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Study of interfacial friction force for bubble flows in a 2×1 rod channel simplifying BWR [texte imprimé] / Akimaro Kawahara, Auteur ; Michio Sadatomi, Auteur ; Yutaro Nakamoto, Auteur . - 2012 . - 08 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Mots-clés : Bubbles  Channel  flow  Drag  Fission  reactors  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Most of the recent subchannel analysis codes are based on a multifluid model, and an accurate evaluation of the constitutive equations in the model is essential. In order to get an accurate interfacial friction force in two-phase bubble flows, experimental data on drag coefficient and interfacial area concentration have been obtained for air-water flows in a 2×1 rod channel simplifying a boiling water nuclear reactor fuel rod bundle. In order to know the effects of liquid properties on the data, the temperature of the test water was changed from 18°C to 50°C. The data are compared with the existing correlations reported in literatures. As a result, the semitheoretical correlation of Hibiki and Ishii (2001, “Interfacial Area Concentration in Steady Fully-Developed Bubbly Flow,” Int. J. Heat Mass Transfer, 44, pp. 3443–3461) was found to give the best prediction against the present interfacial area concentration data. The correlation of Delhaye and Bricard (1994, “Interfacial Area in Bubbly Flow: Experimental Data and Correlations,” Nucl. Eng. Des., 151, pp. 65–77) also gave a reasonably good prediction if their correlation was modified by incorporating liquid property effects. As for the drag coefficient, no correlation exists, which can predict the present data well. Therefore, we developed a new correlation, including three dimensionless numbers, i.e., bubble capillary number, Morton number, and Eötvös number. The correlation predicted the data of Liu et al. (2008, “Drag Coefficient in One-Dimensional Two-Group Two-Fluid Model,” Int. J. Heat Fluid Flow, 29, pp. 1402–1410) as well as the present data well. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Subcooled water flow boiling heat transfer in a short SUS304-tube with twisted-tape insert / Koichi Hata in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 11 p. Titre : Subcooled water flow boiling heat transfer in a short SUS304-tube with twisted-tape insert Type de document : texte imprimé Auteurs : Koichi Hata, Auteur ; Suguru Masuzaki, Auteur Année de publication : 2012 Article en page(s) : 11 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Boiling  Heat  transfer  Invertors  Pipe  flow  Pumps  Swirling  flow  Turbulence  Undercooling Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The subcooled boiling heat transfer and the steady-state critical heat fluxes (CHFs) in a short SUS304-tube with twisted-tape insert are systematically measured for mass velocities (G=4016–13,850 kg/m2 s), inlet liquid temperatures (Tin=285.82–363.96 K), outlet pressures (Pout=764.76–889.02 kPa), and exponentially increasing heat input (Q=Q0 exp(t/tau), tau=8.5 s) by the experimental water loop comprised of a multistage canned-type circulation pump controlled by an inverter. The SUS304 test tube of inner diameter (d=6 mm), heated length (L=59.5 mm), effective length (Leff=49.1 mm), L/d (=9.92), Leff/d (=8.18), and wall thickness (delta=0.5 mm) with average surface roughness (Ra=3.18 µm) is used in this work. The SUS304 twisted tape with twist ratio, y(=H/d=(pitch of 180 deg rotation)/d), of 3.39 is used. The relation between inner surface temperature and heat flux for the SUS304-tube with the twisted-tape insert are clarified from nonboiling to CHF. The subcooled boiling heat transfer for SUS304-tube with the twisted-tape insert is compared with our empty SUS304-tube data and the values calculated by our and other workers' correlations for the subcooled boiling heat transfer. The influences of the twisted-tape insert and the swirl velocity on the subcooled boiling heat transfer and the CHFs are investigated into details and the widely and precisely predictable correlations of the subcooled boiling heat transfer and the CHFs for turbulent flow of water in the SUS304-tube with twisted-tape insert are given based on the experimental data. The correlations can describe the subcooled boiling heat transfer coefficients and the CHFs obtained in this work within −25 to +15% difference. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Subcooled water flow boiling heat transfer in a short SUS304-tube with twisted-tape insert [texte imprimé] / Koichi Hata, Auteur ; Suguru Masuzaki, Auteur . - 2012 . - 11 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 11 p. Mots-clés : Boiling  Heat  transfer  Invertors  Pipe  flow  Pumps  Swirling  flow  Turbulence  Undercooling Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The subcooled boiling heat transfer and the steady-state critical heat fluxes (CHFs) in a short SUS304-tube with twisted-tape insert are systematically measured for mass velocities (G=4016–13,850 kg/m2 s), inlet liquid temperatures (Tin=285.82–363.96 K), outlet pressures (Pout=764.76–889.02 kPa), and exponentially increasing heat input (Q=Q0 exp(t/tau), tau=8.5 s) by the experimental water loop comprised of a multistage canned-type circulation pump controlled by an inverter. The SUS304 test tube of inner diameter (d=6 mm), heated length (L=59.5 mm), effective length (Leff=49.1 mm), L/d (=9.92), Leff/d (=8.18), and wall thickness (delta=0.5 mm) with average surface roughness (Ra=3.18 µm) is used in this work. The SUS304 twisted tape with twist ratio, y(=H/d=(pitch of 180 deg rotation)/d), of 3.39 is used. The relation between inner surface temperature and heat flux for the SUS304-tube with the twisted-tape insert are clarified from nonboiling to CHF. The subcooled boiling heat transfer for SUS304-tube with the twisted-tape insert is compared with our empty SUS304-tube data and the values calculated by our and other workers' correlations for the subcooled boiling heat transfer. The influences of the twisted-tape insert and the swirl velocity on the subcooled boiling heat transfer and the CHFs are investigated into details and the widely and precisely predictable correlations of the subcooled boiling heat transfer and the CHFs for turbulent flow of water in the SUS304-tube with twisted-tape insert are given based on the experimental data. The correlations can describe the subcooled boiling heat transfer coefficients and the CHFs obtained in this work within −25 to +15% difference. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Photographic study on two-phase flow patterns of water in a single-side heated narrow rectangular channel / Junfeng Wang in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 05 p. Titre : Photographic study on two-phase flow patterns of water in a single-side heated narrow rectangular channel Type de document : texte imprimé Auteurs : Junfeng Wang, Auteur ; Yanping Huang, Auteur ; Yanlin Wang, Auteur Année de publication : 2012 Article en page(s) : 05 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Boiling  Channel  flow  Heat  transfer  Photography  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Visualized experimental observation on flow patterns during flow boiling of water under single-side heated and fluid-inlet subcooled conditions in a vertical narrow rectangular channel with the cross section of 40×3 mm2 have been carried out. Four discernible flow patterns, which names dispersed bubbly, coalesced bubbly, churn flow, and annular flow are observed. Flow visualization in two dimensions of two-phase flow patterns for narrow rectangular channel, which provided clearer evidence to distinguish flow patterns, have been performed. Based on the experimental results, a flow pattern map for single-side heated narrow rectangular channel has been developed and then compared with the exiting maps and flow transition criteria. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Photographic study on two-phase flow patterns of water in a single-side heated narrow rectangular channel [texte imprimé] / Junfeng Wang, Auteur ; Yanping Huang, Auteur ; Yanlin Wang, Auteur . - 2012 . - 05 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 05 p. Mots-clés : Boiling  Channel  flow  Heat  transfer  Photography  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Visualized experimental observation on flow patterns during flow boiling of water under single-side heated and fluid-inlet subcooled conditions in a vertical narrow rectangular channel with the cross section of 40×3 mm2 have been carried out. Four discernible flow patterns, which names dispersed bubbly, coalesced bubbly, churn flow, and annular flow are observed. Flow visualization in two dimensions of two-phase flow patterns for narrow rectangular channel, which provided clearer evidence to distinguish flow patterns, have been performed. Based on the experimental results, a flow pattern map for single-side heated narrow rectangular channel has been developed and then compared with the exiting maps and flow transition criteria. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

High temperature aging and corrosion study on alloy 617 and alloy 230 / Kun Mo in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Titre : High temperature aging and corrosion study on alloy 617 and alloy 230 Type de document : texte imprimé Auteurs : Kun Mo, Auteur ; Gianfranco Lovicu, Auteur ; Hsiao-Ming Tung, Auteur Année de publication : 2012 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Ageing  Alloys  Corrosion  Crystal  microstructure  Fission  reactor  materials  Hardness  Scanning  electron  microscopy  Tensile  strength  Transmission  electron  microscopy  X-ray  spectroscopy Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The very high temperature gas-cooled reactor (VHTR), with dual capacities of highly efficient electricity generation and thermochemical production of hydrogen, is considered as one of the most promising Gen-IV nuclear systems. The primary candidate materials for construction of the intermediate heat exchanger (IHX) for the VHTR are alloy 617 and alloy 230. To have a better understanding of the degradation process during high temperature long-term service and to provide practical data for the engineering design of the IHX, aging experiments were performed on alloy 617 and alloy 230 at 900°C and 1000°C. Mechanical properties (hardness and tensile strength) and microstructure were analyzed on post-aging samples after different aging periods (up to 3000 h). Both alloys attained increased hardness during the early stages of aging and dramatically soften after extended aging times. Microstructural analysis including transmission electron microscopy, scanning electron microscopy, energy dispersive X-ray spectroscopy, and electron backscatter diffraction was carried out to investigate the microstructure evolution during aging. A carbide particle precipitation, growth, and maturing process was observed for both alloys, which corresponds to the changes of the materials' mechanical properties. Few changes in grain boundary character distribution and grain size distribution were observed after aging. In addition, high temperature corrosion studies were performed at 900°C and 1000°C for both alloys. Alloy 230 exhibits much better corrosion resistance at elevated temperature compared with alloy 617. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] High temperature aging and corrosion study on alloy 617 and alloy 230 [texte imprimé] / Kun Mo, Auteur ; Gianfranco Lovicu, Auteur ; Hsiao-Ming Tung, Auteur . - 2012 . - 09 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Mots-clés : Ageing  Alloys  Corrosion  Crystal  microstructure  Fission  reactor  materials  Hardness  Scanning  electron  microscopy  Tensile  strength  Transmission  electron  microscopy  X-ray  spectroscopy Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The very high temperature gas-cooled reactor (VHTR), with dual capacities of highly efficient electricity generation and thermochemical production of hydrogen, is considered as one of the most promising Gen-IV nuclear systems. The primary candidate materials for construction of the intermediate heat exchanger (IHX) for the VHTR are alloy 617 and alloy 230. To have a better understanding of the degradation process during high temperature long-term service and to provide practical data for the engineering design of the IHX, aging experiments were performed on alloy 617 and alloy 230 at 900°C and 1000°C. Mechanical properties (hardness and tensile strength) and microstructure were analyzed on post-aging samples after different aging periods (up to 3000 h). Both alloys attained increased hardness during the early stages of aging and dramatically soften after extended aging times. Microstructural analysis including transmission electron microscopy, scanning electron microscopy, energy dispersive X-ray spectroscopy, and electron backscatter diffraction was carried out to investigate the microstructure evolution during aging. A carbide particle precipitation, growth, and maturing process was observed for both alloys, which corresponds to the changes of the materials' mechanical properties. Few changes in grain boundary character distribution and grain size distribution were observed after aging. In addition, high temperature corrosion studies were performed at 900°C and 1000°C for both alloys. Alloy 230 exhibits much better corrosion resistance at elevated temperature compared with alloy 617. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Experimental Comparisons of 3D Reconstructed Pin-Power Distributions in Full-Scale BWR Fuel Assemblies / Flavio D. Giust in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 13 p. Titre : Experimental Comparisons of 3D Reconstructed Pin-Power Distributions in Full-Scale BWR Fuel Assemblies Type de document : texte imprimé Auteurs : Flavio D. Giust, Auteur ; Peter Grimm, Auteur ; Rakesh Chawla, Auteur Année de publication : 2012 Article en page(s) : 13 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Gas  turbine  power  stations Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In this paper, reconstructed local fission rates obtained with the two-group nodal diffusion program PRESTO-2, used at the Leibstadt Nuclear Power Plant (KKL) in Switzerland, are compared with experimental results and MCNPX calculations. The experimental facility consists of a test zone, where the measurements are made, surrounded by a buffer zone and two driver zones that render the system critical and also contain the control rods. The test zone consists of a tank that contains a 3×3 array of BWR fuel assemblies of type SVEA-96+. Four cases are considered, all corresponding to a 1.23 m high active zone moderated with light water at room temperature: (1) axially uniform enrichment and gadolinium content, (2) like case 1 but with an L-shaped control blade completely inserted, (3) enrichment and gadolinium content change at the core midplane, and (4) like case 2 but with the control blade partially inserted. The comparisons give insight into the accuracy of the pin-power reconstruction methodology. The axially uniform case without control blade shows a good radial agreement and a well predicted axial curvature of the flux. On the other hand, systematic deviations are observed in the radial direction for the controlled cases, with the axial heterogeneities causing deviations around the discontinuity and also in the axial curvature of the flux. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Experimental Comparisons of 3D Reconstructed Pin-Power Distributions in Full-Scale BWR Fuel Assemblies [texte imprimé] / Flavio D. Giust, Auteur ; Peter Grimm, Auteur ; Rakesh Chawla, Auteur . - 2012 . - 13 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 13 p. Mots-clés : Gas  turbine  power  stations Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In this paper, reconstructed local fission rates obtained with the two-group nodal diffusion program PRESTO-2, used at the Leibstadt Nuclear Power Plant (KKL) in Switzerland, are compared with experimental results and MCNPX calculations. The experimental facility consists of a test zone, where the measurements are made, surrounded by a buffer zone and two driver zones that render the system critical and also contain the control rods. The test zone consists of a tank that contains a 3×3 array of BWR fuel assemblies of type SVEA-96+. Four cases are considered, all corresponding to a 1.23 m high active zone moderated with light water at room temperature: (1) axially uniform enrichment and gadolinium content, (2) like case 1 but with an L-shaped control blade completely inserted, (3) enrichment and gadolinium content change at the core midplane, and (4) like case 2 but with the control blade partially inserted. The comparisons give insight into the accuracy of the pin-power reconstruction methodology. The axially uniform case without control blade shows a good radial agreement and a well predicted axial curvature of the flux. On the other hand, systematic deviations are observed in the radial direction for the controlled cases, with the axial heterogeneities causing deviations around the discontinuity and also in the axial curvature of the flux. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Design and construction of IFMIF/EVEDA lithium test loop / H. Kondo in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Titre : Design and construction of IFMIF/EVEDA lithium test loop Type de document : texte imprimé Auteurs : H. Kondo, Auteur ; Furukawa, T., Auteur ; Y. Hirakawa, Auteur Année de publication : 2012 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fusion  reactor  materials  Lithium  Neutron  flux Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The International Fusion Materials Irradiation Facility (IFMIF) is a D+–Li neutron source aimed at producing an intense high energy neutron flux (2 MW/m2) for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDAs) of IFMIF started on July 2007. Regarding the lithium (Li) target facility, design, construction, and tests of EVEDA Li test loop (ELTL) is a major Japanese activity. The detail design of the loop was started since early 2009. Construction of the loop was started at the middle of 2009, and completion is scheduled at the end of February 2011. This paper focuses on the design of the loop configuration and the major components. ELTL was designed to consist of two major Li loops, which are a main loop and a purification loop including an impurity monitoring loop. The main loop equips a target assembly which produces a high-speed free-surface Li flow to test the flow stability as the D+ beam target. The maximum flow rate of an electromagnetic pump in the main loop was set to 3000 l/min, so that flow velocity in the target assembly is 20 m/s at the maximum. Regarding the purification loop, a cold trap and two hot traps and impurity monitors are installed in order to purify and monitor impurities in Li. The configuration of these components in addition to the specification and configuration of the whole loop is presented. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Design and construction of IFMIF/EVEDA lithium test loop [texte imprimé] / H. Kondo, Auteur ; Furukawa, T., Auteur ; Y. Hirakawa, Auteur . - 2012 . - 06 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Mots-clés : Fusion  reactor  materials  Lithium  Neutron  flux Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The International Fusion Materials Irradiation Facility (IFMIF) is a D+–Li neutron source aimed at producing an intense high energy neutron flux (2 MW/m2) for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDAs) of IFMIF started on July 2007. Regarding the lithium (Li) target facility, design, construction, and tests of EVEDA Li test loop (ELTL) is a major Japanese activity. The detail design of the loop was started since early 2009. Construction of the loop was started at the middle of 2009, and completion is scheduled at the end of February 2011. This paper focuses on the design of the loop configuration and the major components. ELTL was designed to consist of two major Li loops, which are a main loop and a purification loop including an impurity monitoring loop. The main loop equips a target assembly which produces a high-speed free-surface Li flow to test the flow stability as the D+ beam target. The maximum flow rate of an electromagnetic pump in the main loop was set to 3000 l/min, so that flow velocity in the target assembly is 20 m/s at the maximum. Regarding the purification loop, a cold trap and two hot traps and impurity monitors are installed in order to purify and monitor impurities in Li. The configuration of these components in addition to the specification and configuration of the whole loop is presented. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Measurement of wavy surface oscillations on liquid metal lithium jet for IFMIF target / Hirokazu Sugiura in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Titre : Measurement of wavy surface oscillations on liquid metal lithium jet for IFMIF target Type de document : texte imprimé Auteurs : Hirokazu Sugiura, Auteur ; Takuji Kanemura, Auteur ; Sachiko Yoshihashi-Suzuki, Auteur Année de publication : 2012 Article en page(s) : 06 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fusion  reactor  materials  Jets  Liquid  metal  ion  sources  Neutron  sources  Oscillations  Particle  accelerators Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The International Fusion Materials Irradiation Facility (IFMIF) has been conceived as a high-flux 14 MeV neutron source for testing candidate fusion reactor materials. In the current design, neutrons are generated by irradiating a target with a deuteron beam and high-speed free-surface flow of liquid metal lithium (Li) is adopted as the target. To reveal the stability of the Li flow, we have examined characteristics of surface waves at a location 175 nm downstream from a nozzle exit, which corresponds to the center of the beam irradiated region. In this study, the characteristics of surface waves just downstream of the nozzle exit were measured experimentally, since the initial growth of surface waves exerts a definite influence on the surface behavior of the Li flow in the downstream region. Experiments were carried out with a focus on surface oscillations of the Li flow using the lithium circulation loop at Osaka University. These oscillations are measured using an electro-contact probe apparatus, which can detect electrically a contact between the probe tip and the Li surface and provide local height data of surface waves. The apparatus was installed at a location 15 mm downstream from the nozzle exit and scanned the Li surface by moving along the liquid-depth direction. The experiments were performed for the velocity range of 3-15 m/s under argon gas atmosphere at a pressure of 0.13 MPa. The contact signal recorded in the experiment was used to analyze the characteristics of surface waves, and then the root-mean-square wave amplitude and the frequency of surface waves were calculated. It was found that the root-mean-square wave amplitudes of surface waves increased with a rise in the flow velocity, and reached approximately 0.18 mm at 14-15 m/s. And also, obtained frequencies were analyzed using a linear stability theory, and the variation of frequencies was examined with the mean flow velocity. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Measurement of wavy surface oscillations on liquid metal lithium jet for IFMIF target [texte imprimé] / Hirokazu Sugiura, Auteur ; Takuji Kanemura, Auteur ; Sachiko Yoshihashi-Suzuki, Auteur . - 2012 . - 06 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 06 p. Mots-clés : Fusion  reactor  materials  Jets  Liquid  metal  ion  sources  Neutron  sources  Oscillations  Particle  accelerators Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The International Fusion Materials Irradiation Facility (IFMIF) has been conceived as a high-flux 14 MeV neutron source for testing candidate fusion reactor materials. In the current design, neutrons are generated by irradiating a target with a deuteron beam and high-speed free-surface flow of liquid metal lithium (Li) is adopted as the target. To reveal the stability of the Li flow, we have examined characteristics of surface waves at a location 175 nm downstream from a nozzle exit, which corresponds to the center of the beam irradiated region. In this study, the characteristics of surface waves just downstream of the nozzle exit were measured experimentally, since the initial growth of surface waves exerts a definite influence on the surface behavior of the Li flow in the downstream region. Experiments were carried out with a focus on surface oscillations of the Li flow using the lithium circulation loop at Osaka University. These oscillations are measured using an electro-contact probe apparatus, which can detect electrically a contact between the probe tip and the Li surface and provide local height data of surface waves. The apparatus was installed at a location 15 mm downstream from the nozzle exit and scanned the Li surface by moving along the liquid-depth direction. The experiments were performed for the velocity range of 3-15 m/s under argon gas atmosphere at a pressure of 0.13 MPa. The contact signal recorded in the experiment was used to analyze the characteristics of surface waves, and then the root-mean-square wave amplitude and the frequency of surface waves were calculated. It was found that the root-mean-square wave amplitudes of surface waves increased with a rise in the flow velocity, and reached approximately 0.18 mm at 14-15 m/s. And also, obtained frequencies were analyzed using a linear stability theory, and the variation of frequencies was examined with the mean flow velocity. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Gamma-ray spectrum analysis of chang'E-1 for lunar detection / Xu HongKun in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 04 p. Titre : Gamma-ray spectrum analysis of chang'E-1 for lunar detection Type de document : texte imprimé Auteurs : Xu HongKun, Auteur ; Fang Fang, Auteur ; Ni Shijun, Auteur Année de publication : 2012 Article en page(s) : 04 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Gamma-ray  applications  Radioactivity  measuring  apparatus Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Gamma-ray spectrum analysis was essential for radioactive environmental monitoring, and it had been widely used in many areas of nuclear engineering. However, for the low-energy region of gamma-ray spectrum, weak peaks were contained in the fast-decreasing background, so it was difficult to extract characteristic information from original spectra. In order to get a better analytic result based on wavelet methods in frequency domain, we had processed the gamma-ray spectrometer data of Chang'E-1 and well extracted some useful information of spectral characteristic peaks. Then, we preliminarily mapped the distribution of net peak counts for potassium on lunar surface, which indirectly reflected the distribution of elemental abundance. At last, we compared our analytic result with that of Apollo and Lunar Prospector and found some consistencies and differences. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Gamma-ray spectrum analysis of chang'E-1 for lunar detection [texte imprimé] / Xu HongKun, Auteur ; Fang Fang, Auteur ; Ni Shijun, Auteur . - 2012 . - 04 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 04 p. Mots-clés : Gamma-ray  applications  Radioactivity  measuring  apparatus Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : Gamma-ray spectrum analysis was essential for radioactive environmental monitoring, and it had been widely used in many areas of nuclear engineering. However, for the low-energy region of gamma-ray spectrum, weak peaks were contained in the fast-decreasing background, so it was difficult to extract characteristic information from original spectra. In order to get a better analytic result based on wavelet methods in frequency domain, we had processed the gamma-ray spectrometer data of Chang'E-1 and well extracted some useful information of spectral characteristic peaks. Then, we preliminarily mapped the distribution of net peak counts for potassium on lunar surface, which indirectly reflected the distribution of elemental abundance. At last, we compared our analytic result with that of Apollo and Lunar Prospector and found some consistencies and differences. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Gas flow simulations in randomly distributed pebbles / Xiang Zhao in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Titre : Gas flow simulations in randomly distributed pebbles Type de document : texte imprimé Auteurs : Xiang Zhao, Auteur ; Trent Montgomery, Auteur ; Sijun Zhang, Auteur Année de publication : 2012 Article en page(s) : 08 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Computational  fluid  dynamics  Gas  turbine  power  stations  Heat  transfer  Navier-Stokes  equations Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In this paper, computational fluid dynamics (CFD) gas flow simulations are carried out for the pebble bed reactor. In CFD calculations, geometry modeling and physical modeling are crucial to CFD results. The effects of the treatments of the interpebble contacts on gas flow fields and heat transfer are examined. A sensitivity analysis for the gap size is conducted with two spherical pebbles, in which the interpebble region is modeled by means of two types of interpebble gap and two kinds of direct contact. Both large eddy simulation and Reynolds-averaged Navier–Stokes models are employed to investigate the turbulent effects. It is found that the flow fields and relevant heat transfer are significantly dependent on the modeling of the interpebble region. The calculations indicate the complex flow structures present within the voids between the fuel pebbles. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Gas flow simulations in randomly distributed pebbles [texte imprimé] / Xiang Zhao, Auteur ; Trent Montgomery, Auteur ; Sijun Zhang, Auteur . - 2012 . - 08 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Mots-clés : Computational  fluid  dynamics  Gas  turbine  power  stations  Heat  transfer  Navier-Stokes  equations Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In this paper, computational fluid dynamics (CFD) gas flow simulations are carried out for the pebble bed reactor. In CFD calculations, geometry modeling and physical modeling are crucial to CFD results. The effects of the treatments of the interpebble contacts on gas flow fields and heat transfer are examined. A sensitivity analysis for the gap size is conducted with two spherical pebbles, in which the interpebble region is modeled by means of two types of interpebble gap and two kinds of direct contact. Both large eddy simulation and Reynolds-averaged Navier–Stokes models are employed to investigate the turbulent effects. It is found that the flow fields and relevant heat transfer are significantly dependent on the modeling of the interpebble region. The calculations indicate the complex flow structures present within the voids between the fuel pebbles. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Computational analysis of downcomer boiling phenomena using a component thermal hydraulic analysis code, CUPID / Hyoung Kyu Cho in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Titre : Computational analysis of downcomer boiling phenomena using a component thermal hydraulic analysis code, CUPID Type de document : texte imprimé Auteurs : Hyoung Kyu Cho, Auteur ; Byong-Jo Yun, Auteur ; Ik Kyu Park, Auteur Année de publication : 2012 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Boilers  Fission  reactor  theory  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : For the analysis of transient two-phase flows in nuclear reactor components such as a reactor vessel, a steam generator, and a containment, KAERI has developed a three-dimensional thermal hydraulic code, CUPID. It adopts a three-dimensional, transient, two-phase and three-field model and includes various physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, the CUPID code and its two-phase flow models were assessed against the downcomer boiling experiment, which was performed to simulate the downcomer boiling phenomena. They may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. The computational analysis using the CUPID code showed that it can appropriately predict the multidimensional boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size model. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Computational analysis of downcomer boiling phenomena using a component thermal hydraulic analysis code, CUPID [texte imprimé] / Hyoung Kyu Cho, Auteur ; Byong-Jo Yun, Auteur ; Ik Kyu Park, Auteur . - 2012 . - 09 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Mots-clés : Boilers  Fission  reactor  theory  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : For the analysis of transient two-phase flows in nuclear reactor components such as a reactor vessel, a steam generator, and a containment, KAERI has developed a three-dimensional thermal hydraulic code, CUPID. It adopts a three-dimensional, transient, two-phase and three-field model and includes various physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, the CUPID code and its two-phase flow models were assessed against the downcomer boiling experiment, which was performed to simulate the downcomer boiling phenomena. They may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. The computational analysis using the CUPID code showed that it can appropriately predict the multidimensional boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size model. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

CANDU reactor space-time kinetic model for load following studies / Lingzhi Xia in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Titre : CANDU reactor space-time kinetic model for load following studies Type de document : texte imprimé Auteurs : Lingzhi Xia, Auteur ; Jin Jiang, Auteur Année de publication : 2012 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission  reactor  kinetics  Neutron  flux Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : This paper presents the development of a three-dimensional space-time neutronic kinetic model of a Canadian deuterium uranium (CANDU) reactor using a modal method. In this method, the reactor space-time neutron flux is synthesized by a time-weighted series of precalculated neutron flux modes. The modes are eigenfunctions of the governing neutron diffusion equation during reference steady-state operation. The xenon effect has also been considered. The reactor model is then implemented within a simulation platform of a CANDU6 reactor regulating system in MATLAB/SIMULINK. A nondimensionalized SIMULINK representation of the reactor kinetic model is established. The behavior of the reactor during load following transients has been simulated using the developed reactor-modeling module. The simulation results prove the efficiency of the model. A three-dimensional neutron flux distribution during transients is represented. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] CANDU reactor space-time kinetic model for load following studies [texte imprimé] / Lingzhi Xia, Auteur ; Jin Jiang, Auteur . - 2012 . - 09 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Mots-clés : Fission  reactor  kinetics  Neutron  flux Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : This paper presents the development of a three-dimensional space-time neutronic kinetic model of a Canadian deuterium uranium (CANDU) reactor using a modal method. In this method, the reactor space-time neutron flux is synthesized by a time-weighted series of precalculated neutron flux modes. The modes are eigenfunctions of the governing neutron diffusion equation during reference steady-state operation. The xenon effect has also been considered. The reactor model is then implemented within a simulation platform of a CANDU6 reactor regulating system in MATLAB/SIMULINK. A nondimensionalized SIMULINK representation of the reactor kinetic model is established. The behavior of the reactor during load following transients has been simulated using the developed reactor-modeling module. The simulation results prove the efficiency of the model. A three-dimensional neutron flux distribution during transients is represented. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Transfer function modeling of zero-power dynamics of circulating fuel reactors / A. Cammi in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Titre : Transfer function modeling of zero-power dynamics of circulating fuel reactors Type de document : texte imprimé Auteurs : A. Cammi, Auteur ; V. Di Marcello, Auteur ; C. Guerrieri, Auteur Année de publication : 2012 Article en page(s) : 08 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Fission  reactors  Hydrodynamics Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In this paper, the zero-power behavior of circulating fuel reactors (CFRs) has been investigated by means of a zero-dimensional neutron kinetics model that provides a simplified but useful approach to the simulation of the dynamics of this class of nuclear reactors. Among CFRs, the most promising is the molten salt reactor (MSR), which is one of the six innovative concepts of reactor proposed by the “Generation IV International Forum” for future nuclear energy supply. One of the key features of CFRs is represented by the fission material, which is dissolved in a liquid mixture that serves both as fuel and coolant. This causes a relevant coupling between neutronics and thermo-hydrodynamics, so that fuel velocity plays a relevant role in determining the dynamic performance of such systems. In the present study, a preliminary model has been developed that is based on the zero-power kinetics equations (i.e., reactivity feedbacks due to temperature change are neglected), modified in order to take into account the effects of the molten salt circulation on the drift of delayed neutron precursors. The system dynamic behavior has been analyzed using the theory of linear systems, and the transfer functions of the neutron density with respect to both reactivity and fuel velocity have been calculated. The developed model has been assessed on the basis of the available experimental data from the molten salt reactor experiment (MSRE) provided by the Oak Ridge National Laboratory. The results of the present work show that the developed simplified theoretical model is well descriptive of the MSRE zero-power dynamics, allowing a preliminary evaluation of the effects due to the circulation of the fuel salt on the neutronics of the system. Moreover, the model is of general validity for any kind of CFRs, and hence is applicable to study other MSR concepts in order to have some indications on the control strategy to be adopted in the MSR development envisaged by Generation IV. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Transfer function modeling of zero-power dynamics of circulating fuel reactors [texte imprimé] / A. Cammi, Auteur ; V. Di Marcello, Auteur ; C. Guerrieri, Auteur . - 2012 . - 08 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 08 p. Mots-clés : Fission  reactors  Hydrodynamics Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In this paper, the zero-power behavior of circulating fuel reactors (CFRs) has been investigated by means of a zero-dimensional neutron kinetics model that provides a simplified but useful approach to the simulation of the dynamics of this class of nuclear reactors. Among CFRs, the most promising is the molten salt reactor (MSR), which is one of the six innovative concepts of reactor proposed by the “Generation IV International Forum” for future nuclear energy supply. One of the key features of CFRs is represented by the fission material, which is dissolved in a liquid mixture that serves both as fuel and coolant. This causes a relevant coupling between neutronics and thermo-hydrodynamics, so that fuel velocity plays a relevant role in determining the dynamic performance of such systems. In the present study, a preliminary model has been developed that is based on the zero-power kinetics equations (i.e., reactivity feedbacks due to temperature change are neglected), modified in order to take into account the effects of the molten salt circulation on the drift of delayed neutron precursors. The system dynamic behavior has been analyzed using the theory of linear systems, and the transfer functions of the neutron density with respect to both reactivity and fuel velocity have been calculated. The developed model has been assessed on the basis of the available experimental data from the molten salt reactor experiment (MSRE) provided by the Oak Ridge National Laboratory. The results of the present work show that the developed simplified theoretical model is well descriptive of the MSRE zero-power dynamics, allowing a preliminary evaluation of the effects due to the circulation of the fuel salt on the neutronics of the system. Moreover, the model is of general validity for any kind of CFRs, and hence is applicable to study other MSR concepts in order to have some indications on the control strategy to be adopted in the MSR development envisaged by Generation IV. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Comparison of countercurrent flow limitation experiments performed in two different models of the hot leg of a pressurized water reactor with rectangular cross section / Christophe Vallée in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Titre : Comparison of countercurrent flow limitation experiments performed in two different models of the hot leg of a pressurized water reactor with rectangular cross section Type de document : texte imprimé Auteurs : Christophe Vallée, Auteur ; Tobias Seidel, Auteur ; Dirk Lucas, Auteur Année de publication : 2012 Article en page(s) : 09 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Boilers  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In order to investigate the two-phase flow behavior during countercurrent flow limitation in the hot leg of a pressurized water reactor, two test models were built: one at the Kobe University and the other at the TOPFLOW test facility of Forschungszentrum Dresden-Rossendorf (FZD). Both test facilities are devoted to optical measurement techniques; therefore, a flat hot leg test section design was chosen. Countercurrent flow limitation (CCFL) experiments were performed, simulating the reflux condenser cooling mode appearing in some accident scenarios. The fluids used were air and water, both at room temperature. The pressure conditions were varied from atmospheric at Kobe to 3.0 bars absolute at TOPFLOW. According to the presented review of literature, very few data are available on flooding in channels with a rectangular cross section, and no experiments were performed in the past in such flat models of a hot leg. Commonly, the macroscopic effects of CCFL are represented in a flooding diagram, where the gas flow rate is plotted versus the discharge water flow rate, using the nondimensional superficial velocity (also known as Wallis parameter) as coordinates. However, the classical definition of the Wallis parameter contains the pipe diameter as characteristic length. In order to be able to perform comparisons with pipe experiments and to extrapolate to the power plant scale, the appropriate characteristic length should be determined. A detailed comparison of the test facilities operated at the Kobe University and at FZD is presented. With respect to the CCFL behavior, it is shown that the essential parts of the two hot leg test sections are very similar. This geometrical analogy allows us to perform meaningful comparisons. However, clear differences in the dimensions of the cross section (H×W=150×10 mm2 in Kobe, 250×50 mm2 at FZD) make it possible to point out the right characteristic length for hot leg models with rectangular cross sections. The hydraulic diameter, the channel height, and the Laplace critical wavelength (leading to the Kutateladze number) were tested. A comparison of our own results with similar experimental data and empirical correlations for pipes available in literature shows that the channel height is the characteristic length to be used in the Wallis parameter for channels with rectangular cross sections. However, some limitations were noticed for narrow channels, where CCFL is reached at lower gas fluxes, as already observed in small scale hot legs with pipe cross sections. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Comparison of countercurrent flow limitation experiments performed in two different models of the hot leg of a pressurized water reactor with rectangular cross section [texte imprimé] / Christophe Vallée, Auteur ; Tobias Seidel, Auteur ; Dirk Lucas, Auteur . - 2012 . - 09 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 09 p. Mots-clés : Boilers  Two-phase  flow Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : In order to investigate the two-phase flow behavior during countercurrent flow limitation in the hot leg of a pressurized water reactor, two test models were built: one at the Kobe University and the other at the TOPFLOW test facility of Forschungszentrum Dresden-Rossendorf (FZD). Both test facilities are devoted to optical measurement techniques; therefore, a flat hot leg test section design was chosen. Countercurrent flow limitation (CCFL) experiments were performed, simulating the reflux condenser cooling mode appearing in some accident scenarios. The fluids used were air and water, both at room temperature. The pressure conditions were varied from atmospheric at Kobe to 3.0 bars absolute at TOPFLOW. According to the presented review of literature, very few data are available on flooding in channels with a rectangular cross section, and no experiments were performed in the past in such flat models of a hot leg. Commonly, the macroscopic effects of CCFL are represented in a flooding diagram, where the gas flow rate is plotted versus the discharge water flow rate, using the nondimensional superficial velocity (also known as Wallis parameter) as coordinates. However, the classical definition of the Wallis parameter contains the pipe diameter as characteristic length. In order to be able to perform comparisons with pipe experiments and to extrapolate to the power plant scale, the appropriate characteristic length should be determined. A detailed comparison of the test facilities operated at the Kobe University and at FZD is presented. With respect to the CCFL behavior, it is shown that the essential parts of the two hot leg test sections are very similar. This geometrical analogy allows us to perform meaningful comparisons. However, clear differences in the dimensions of the cross section (H×W=150×10 mm2 in Kobe, 250×50 mm2 at FZD) make it possible to point out the right characteristic length for hot leg models with rectangular cross sections. The hydraulic diameter, the channel height, and the Laplace critical wavelength (leading to the Kutateladze number) were tested. A comparison of our own results with similar experimental data and empirical correlations for pipes available in literature shows that the channel height is the characteristic length to be used in the Wallis parameter for channels with rectangular cross sections. However, some limitations were noticed for narrow channels, where CCFL is reached at lower gas fluxes, as already observed in small scale hot legs with pipe cross sections. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

Investigations and countermeasures for deactivation of hydrogen recombination catalyst at hamaoka units 4 and 5 / Toru Kawasaki in Transactions of the ASME . Journal of engineering for gas turbines and power, Vol. 133 N° 5 (Mai 2011) 

Public ISBD [article]  in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 05 p. Titre : Investigations and countermeasures for deactivation of hydrogen recombination catalyst at hamaoka units 4 and 5 Type de document : texte imprimé Auteurs : Toru Kawasaki, Auteur ; Motohiro Aizawa, Auteur ; Hidehiro Iizuka, Auteur Année de publication : 2012 Article en page(s) : 05 p. Note générale : Génie Mécanique Langues : Anglais (eng) Mots-clés : Catalysts  Gas  turbines  Heat  treatment  Hydrogen  production  Organic  compounds  Sealing  materials  Silicon Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The hydrogen concentration in the outlets of off-gas recombiners increased at Hamaoka Units 4 and 5, and their reactors could not continue the startup operations. Therefore, we investigated why the recombination reactions were deactivated and we selected appropriate countermeasures for both plants. Two types of deactivation mechanisms were found from our investigations. The first cause was the decrease in the active surface area of alumina as support material due to dehydrative condensation. The other cause was the catalyst being poisoned by organic silicon compounds. Organic silicon was introduced from the organosilicon sealant used at the junctions of low-pressure turbines. We also found that a boehmite rich catalyst was deactivated more easily by organic silicon than gamma alumina because boehmite had numerous hydroxyl groups. Finally, we estimated that the deactivation of hydrogen recombination catalysts was caused by two combined factors; these were the characteristics of boehmite as the ingredient of catalysts support and the organic silicon poisoning the catalyst surface. As countermeasures, the boehmite was changed into more stable gamma alumina by adding heat treatment in a hydrogen atmosphere at 500°C for 1 h, and the source of organic silicon, organosilicon sealant, was removed. The improved catalysts were applied at Hamaoka Units 4 and 5. Moreover, the linseed oil that used to be used at the plants was applied again as sealant in the low-pressure turbine casing instead of organosilicon sealant. As a result of the application of these countermeasures, the reactors could be started without increasing the hydrogen concentration at these plants. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...] [article] Investigations and countermeasures for deactivation of hydrogen recombination catalyst at hamaoka units 4 and 5 [texte imprimé] / Toru Kawasaki, Auteur ; Motohiro Aizawa, Auteur ; Hidehiro Iizuka, Auteur . - 2012 . - 05 p. Génie Mécanique Langues : Anglais (eng) in Transactions of the ASME . Journal of engineering for gas turbines and power > Vol. 133 N° 5 (Mai 2011) . - 05 p. Mots-clés : Catalysts  Gas  turbines  Heat  treatment  Hydrogen  production  Organic  compounds  Sealing  materials  Silicon Index. décimale : 620.1 Essais des matériaux. Défauts des matériaux. Protection des matériaux Résumé : The hydrogen concentration in the outlets of off-gas recombiners increased at Hamaoka Units 4 and 5, and their reactors could not continue the startup operations. Therefore, we investigated why the recombination reactions were deactivated and we selected appropriate countermeasures for both plants. Two types of deactivation mechanisms were found from our investigations. The first cause was the decrease in the active surface area of alumina as support material due to dehydrative condensation. The other cause was the catalyst being poisoned by organic silicon compounds. Organic silicon was introduced from the organosilicon sealant used at the junctions of low-pressure turbines. We also found that a boehmite rich catalyst was deactivated more easily by organic silicon than gamma alumina because boehmite had numerous hydroxyl groups. Finally, we estimated that the deactivation of hydrogen recombination catalysts was caused by two combined factors; these were the characteristics of boehmite as the ingredient of catalysts support and the organic silicon poisoning the catalyst surface. As countermeasures, the boehmite was changed into more stable gamma alumina by adding heat treatment in a hydrogen atmosphere at 500°C for 1 h, and the source of organic silicon, organosilicon sealant, was removed. The improved catalysts were applied at Hamaoka Units 4 and 5. Moreover, the linseed oil that used to be used at the plants was applied again as sealant in the low-pressure turbine casing instead of organosilicon sealant. As a result of the application of these countermeasures, the reactors could be started without increasing the hydrogen concentration at these plants. DEWEY : 620.1 ISSN : 0742-4795 En ligne : http://scitation.aip.org/getabs/servlet/GetabsServlet?prog=normal&id=JETPEZ00013 [...]

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