Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518408
K. Werley, C. Bathke
Start-up conditions are examined for a pulsed tokamak reactor that uses only inductively driven plasma current (and bootstrap current). A zero-dimensional (profile-averaged) model containing plasma power and particle balance equations is used to study several aspects of plasma start-up, including: (1) optimization of the start-up pathway; (2) tradeoffs of auxiliary start-up heating power versus start-up time; (3) volt-second consumption; (4) thermal stability of the operating point; (5) estimates of the divertor heat flux and temperature during the start-up transient; (6) the sensitivity of the available operating space to allowed values of the H confinement factor; and (7) partial-power operations.
{"title":"Start-up simulations of the PULSAR pulsed tokamak reactor","authors":"K. Werley, C. Bathke","doi":"10.1109/FUSION.1993.518408","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518408","url":null,"abstract":"Start-up conditions are examined for a pulsed tokamak reactor that uses only inductively driven plasma current (and bootstrap current). A zero-dimensional (profile-averaged) model containing plasma power and particle balance equations is used to study several aspects of plasma start-up, including: (1) optimization of the start-up pathway; (2) tradeoffs of auxiliary start-up heating power versus start-up time; (3) volt-second consumption; (4) thermal stability of the operating point; (5) estimates of the divertor heat flux and temperature during the start-up transient; (6) the sensitivity of the available operating space to allowed values of the H confinement factor; and (7) partial-power operations.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127872183","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518311
N. Uckan
Confinement capability of the ITER-EDA (R=7.75 m, I=25 MA) operational scenarios is evaluated and compared with the ITER CDA (R=6 m, 22 MA). The ignition capability of ITER EDA is somewhat higher than that of CDA by a factor of 1.1-1.2 with empirical power law scalings and by a factor of 1.5-2 with offset linear scalings. Simulations with the RLW /spl chi/(/spl nabla/T/sub e/)/sub crit/ model show that both the EDA and CDA scenarios operates in L-mode, however CDA ignition margin is much smaller. With empirical scalings, the required L-mode confinement enhancement factor [H=/spl tau//sub E///spl tau//sub E/(scaling)] corresponding to, for example, ITER89-P L-mode scaling would be 1.5-1.6 in ITER EDA relative to 1.8 in CDA for 10% He (plus 1% Be) concentration. At a higher concentration of He of 20-25%, the confinement capability is deteriorated and the required confinement enhancement factor (over empirical L-mode scalings) is /spl ges/2. The Ohmic confinement time is a factor of two higher in the EDA design (as compared to the CDA), yielding a strong reduction in the auxiliary power required to reach ignition. In 1.5-D simulations with L-mode enhancement factors of H/spl ges/1.2 allowed ohmic ignition with 25 MA, ignition was aided by initially peaked density profiles (and low He content) during the start-up.
{"title":"Confinement capability of ITER-EDA design","authors":"N. Uckan","doi":"10.1109/FUSION.1993.518311","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518311","url":null,"abstract":"Confinement capability of the ITER-EDA (R=7.75 m, I=25 MA) operational scenarios is evaluated and compared with the ITER CDA (R=6 m, 22 MA). The ignition capability of ITER EDA is somewhat higher than that of CDA by a factor of 1.1-1.2 with empirical power law scalings and by a factor of 1.5-2 with offset linear scalings. Simulations with the RLW /spl chi/(/spl nabla/T/sub e/)/sub crit/ model show that both the EDA and CDA scenarios operates in L-mode, however CDA ignition margin is much smaller. With empirical scalings, the required L-mode confinement enhancement factor [H=/spl tau//sub E///spl tau//sub E/(scaling)] corresponding to, for example, ITER89-P L-mode scaling would be 1.5-1.6 in ITER EDA relative to 1.8 in CDA for 10% He (plus 1% Be) concentration. At a higher concentration of He of 20-25%, the confinement capability is deteriorated and the required confinement enhancement factor (over empirical L-mode scalings) is /spl ges/2. The Ohmic confinement time is a factor of two higher in the EDA design (as compared to the CDA), yielding a strong reduction in the auxiliary power required to reach ignition. In 1.5-D simulations with L-mode enhancement factors of H/spl ges/1.2 allowed ohmic ignition with 25 MA, ignition was aided by initially peaked density profiles (and low He content) during the start-up.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117248900","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518417
B. Turck, D. Bessette, D. Ciazynski, J. Duchateau
Conductors presently retained for the coils of large fusion machines are cable-in-conduit conductors made of about one thousand of superconducting strands cabled inside a jacket and cooled by a forced flow of helium. The current density is the key parameter for die machine design, as it reacts strongly on the size and on the cost. An optimized conductor design has been pursued to maximize the current density while fulfilling several criteria regarding protection, stability and safety margin. The influence of some parameters such as: field, strain, effective diameter of filaments and heat transfer coefficient is analyzed. A Nb/sub 3/Sn conductor developed by CEA relevant for NET/ITER has been tested successfully at 6.2 K up to 50 kA and 12 T. Following the same concept a complete design of a conductor for the central solenoid of ITER is proposed.
{"title":"Design methods and actual performances of conductors for the superconducting coils of tokamaks","authors":"B. Turck, D. Bessette, D. Ciazynski, J. Duchateau","doi":"10.1109/FUSION.1993.518417","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518417","url":null,"abstract":"Conductors presently retained for the coils of large fusion machines are cable-in-conduit conductors made of about one thousand of superconducting strands cabled inside a jacket and cooled by a forced flow of helium. The current density is the key parameter for die machine design, as it reacts strongly on the size and on the cost. An optimized conductor design has been pursued to maximize the current density while fulfilling several criteria regarding protection, stability and safety margin. The influence of some parameters such as: field, strain, effective diameter of filaments and heat transfer coefficient is analyzed. A Nb/sub 3/Sn conductor developed by CEA relevant for NET/ITER has been tested successfully at 6.2 K up to 50 kA and 12 T. Following the same concept a complete design of a conductor for the central solenoid of ITER is proposed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115437775","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518275
A. Reggiori, R. Carlevaro, G. Riva, G. Daminelli
Among the different methods of pellet injection for Tokamak refueling, the two-stage gun is presently the most suitable for pellet speeds over 2000 m/s, but for continuous operation two main problems are to be solved: wear of the pistons and plasma contamination by the propulsion gas. An experimental program is under way with the CNPM two-stage gas gun aimed at the solution of these problems. Different piston geometries and materials have been tested. Moreover, a fast valve has been installed at the exit of the launch barrel, in order to minimize the propellant gas flow. Results of the tests are presented.
{"title":"Experiments with the CNPM gas gun for minimizing piston wear and propellant gas flow","authors":"A. Reggiori, R. Carlevaro, G. Riva, G. Daminelli","doi":"10.1109/FUSION.1993.518275","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518275","url":null,"abstract":"Among the different methods of pellet injection for Tokamak refueling, the two-stage gun is presently the most suitable for pellet speeds over 2000 m/s, but for continuous operation two main problems are to be solved: wear of the pistons and plasma contamination by the propulsion gas. An experimental program is under way with the CNPM two-stage gas gun aimed at the solution of these problems. Different piston geometries and materials have been tested. Moreover, a fast valve has been installed at the exit of the launch barrel, in order to minimize the propellant gas flow. Results of the tests are presented.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114257631","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518532
J. Galambos, D. Conner, P. Goranson, D. Lousteau, D. Williamson, B. Nelson, F. Davis
An analytic parametric scoping tool has been developed for application to first wall (FW) design problems. Both thermal and disruption force effects are considered. For the high heat flux and high disruption load conditions expected in the International Thermonuclear Experimental Reactor (ITER) device, vanadium alloy and dispersion-strengthened copper offer the best stress margins using a somewhat flattened plasma-facing configuration. Ferritic steels also appear to have an acceptable stress margin, whereas the conventional stainless steel 316 does not appear feasible. If a full semicircle shape FW is required, only the vanadium and ferritic steel alloy have acceptable solutions.
{"title":"Methodology for first wall design","authors":"J. Galambos, D. Conner, P. Goranson, D. Lousteau, D. Williamson, B. Nelson, F. Davis","doi":"10.1109/FUSION.1993.518532","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518532","url":null,"abstract":"An analytic parametric scoping tool has been developed for application to first wall (FW) design problems. Both thermal and disruption force effects are considered. For the high heat flux and high disruption load conditions expected in the International Thermonuclear Experimental Reactor (ITER) device, vanadium alloy and dispersion-strengthened copper offer the best stress margins using a somewhat flattened plasma-facing configuration. Ferritic steels also appear to have an acceptable stress margin, whereas the conventional stainless steel 316 does not appear feasible. If a full semicircle shape FW is required, only the vanadium and ferritic steel alloy have acceptable solutions.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116225541","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518536
D. Sevier, E. Chin, T.R. Hodapp, R. Junge, K. Redler, H. Mantz
The TPX limiter and armor systems are designed for steady-state operation from day one operation, at 18 MW plasma input power, to a possible upgrade to 45 MW. All components are designed for remote handling. Carbon-carbon (C-C) composites are the baseline plasma facing material for all limiter and armor systems. Where applicable, all components are made from low activation materials. The TPX limiter system consists of the inboard toroidal limiter, the outboard toroidal limiter, and three discrete poloidal limiters. These limiters are used for plasma startup and to protect the vessel, passive plates, and equipment in the ports from the energetic particle fluxes during steady-state operation. In addition, the inboard limiter protects the vacuum vessel from steady-state neutral beam shine-though and from neutral beam faults. The TPX armor components consist of two major systems: the neutral beam armor that protects the outer vessel wall and equipment in the ports, and the ripple armor that intercepts the trapped energetic particles that are drifting vertically in the ripple region. Different design concepts are employed for these plasma facing components (PFC) depending on their expected heat loads. Inboard and outboard limiters are designed with mechanically restrained C-C composite tiles mounted on cooled support plates. Components which must withstand higher heat loads, such as neutral beam and ripple armor, are made of C-C composite tiles brazed to actively-cooled copper.
{"title":"Design of the TPX limiter and armor components","authors":"D. Sevier, E. Chin, T.R. Hodapp, R. Junge, K. Redler, H. Mantz","doi":"10.1109/FUSION.1993.518536","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518536","url":null,"abstract":"The TPX limiter and armor systems are designed for steady-state operation from day one operation, at 18 MW plasma input power, to a possible upgrade to 45 MW. All components are designed for remote handling. Carbon-carbon (C-C) composites are the baseline plasma facing material for all limiter and armor systems. Where applicable, all components are made from low activation materials. The TPX limiter system consists of the inboard toroidal limiter, the outboard toroidal limiter, and three discrete poloidal limiters. These limiters are used for plasma startup and to protect the vessel, passive plates, and equipment in the ports from the energetic particle fluxes during steady-state operation. In addition, the inboard limiter protects the vacuum vessel from steady-state neutral beam shine-though and from neutral beam faults. The TPX armor components consist of two major systems: the neutral beam armor that protects the outer vessel wall and equipment in the ports, and the ripple armor that intercepts the trapped energetic particles that are drifting vertically in the ripple region. Different design concepts are employed for these plasma facing components (PFC) depending on their expected heat loads. Inboard and outboard limiters are designed with mechanically restrained C-C composite tiles mounted on cooled support plates. Components which must withstand higher heat loads, such as neutral beam and ripple armor, are made of C-C composite tiles brazed to actively-cooled copper.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114091637","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518500
C. A. Foster, H. C. McCurdy
A cryopumping system based on the snail continuous cryopump concept is being developed for fusion applications under a DOE SBIR grant. The primary pump is a liquid helium cooled compound pump designed to continuously pump and fractionate deuterium/tritium and helium. The D/T pumping stage is a 500 mm bore cryocondensation pump with a nominal pumping speed of 45,000 L/s. It will be continuously regenerated by a snail regeneration head every 12 minutes. Continuous regeneration will dramatically reduce the vulnerable tritium inventory in a fusion reactor. Operating at an inlet pressure of 1 millitorr, eight of these pumps could pump the projected D/T flow in the ITER CDA design while reducing the inventory of tritium in the pumping system from 630 to 43 grams. The helium fraction will be pumped in a compound argon frost stage. This stage will also operate continuously with a snail regeneration head. In addition the argon spray head will be enclosed inside the snail, thereby removing gaseous argon from the process chamber. Since the cryocondensation stage will intercept over 90% of the D/T/H stream, a purified stream from this stage could be directly reinjected into the plasma as gas or pellets, thereby bypassing the isotope separation system and further simplifying the fuel cycle. Experiments were undertaken in Phase I which demonstrated continuous cryosorption pumping of hydrogen on CO/sub 2/ and argon frosts. The pumping system and its relevance to fusion reactor pumping will be discussed.
{"title":"Argon frost continuous cryopump for fusion applications","authors":"C. A. Foster, H. C. McCurdy","doi":"10.1109/FUSION.1993.518500","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518500","url":null,"abstract":"A cryopumping system based on the snail continuous cryopump concept is being developed for fusion applications under a DOE SBIR grant. The primary pump is a liquid helium cooled compound pump designed to continuously pump and fractionate deuterium/tritium and helium. The D/T pumping stage is a 500 mm bore cryocondensation pump with a nominal pumping speed of 45,000 L/s. It will be continuously regenerated by a snail regeneration head every 12 minutes. Continuous regeneration will dramatically reduce the vulnerable tritium inventory in a fusion reactor. Operating at an inlet pressure of 1 millitorr, eight of these pumps could pump the projected D/T flow in the ITER CDA design while reducing the inventory of tritium in the pumping system from 630 to 43 grams. The helium fraction will be pumped in a compound argon frost stage. This stage will also operate continuously with a snail regeneration head. In addition the argon spray head will be enclosed inside the snail, thereby removing gaseous argon from the process chamber. Since the cryocondensation stage will intercept over 90% of the D/T/H stream, a purified stream from this stage could be directly reinjected into the plasma as gas or pellets, thereby bypassing the isotope separation system and further simplifying the fuel cycle. Experiments were undertaken in Phase I which demonstrated continuous cryosorption pumping of hydrogen on CO/sub 2/ and argon frosts. The pumping system and its relevance to fusion reactor pumping will be discussed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114232649","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518530
A. Torossian, W. Specking, J. Duchateau, P. Decool
The strain sensitivity of Nb/sub 3/Sn cable is well known. However the practical process to compensate for this effect when 316 LN is used for the jacket has never been considered. In this paper different proposals are analysed in order to prevent the 316 LN jacket contracting more than the Nb/sub 3/Sn cable. A first experiment performed in the FBI test facility of KfK has shown that a prestrain of 0.3% carried out at 275 K on a short straight sample of cable in conduit conductor (3/spl times/3/spl times/4 Nb/sub 3/Sn strands of 0.73 mm in a 316 L conduit) produced an improvement of the critical current. The improvement in this condition is about 80%. Different designs of tooling usable for the CS and TF coils of ITER are described.
{"title":"Drastic improvement of I/sub c/ of Nb/sub 3/Sn CIC conductor by prestraining at room temperature","authors":"A. Torossian, W. Specking, J. Duchateau, P. Decool","doi":"10.1109/FUSION.1993.518530","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518530","url":null,"abstract":"The strain sensitivity of Nb/sub 3/Sn cable is well known. However the practical process to compensate for this effect when 316 LN is used for the jacket has never been considered. In this paper different proposals are analysed in order to prevent the 316 LN jacket contracting more than the Nb/sub 3/Sn cable. A first experiment performed in the FBI test facility of KfK has shown that a prestrain of 0.3% carried out at 275 K on a short straight sample of cable in conduit conductor (3/spl times/3/spl times/4 Nb/sub 3/Sn strands of 0.73 mm in a 316 L conduit) produced an improvement of the critical current. The improvement in this condition is about 80%. Different designs of tooling usable for the CS and TF coils of ITER are described.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114249502","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518286
A. G. Heics, T. Shmayda, D. Muller
A special 150 L/s turbomolecular pump was used to pump pure tritium in closed loop mode over the course of six separate tritium runs. About 5.3 kCi of tritium was pumped during a cumulative period of 237 days. The pump, supplied by Leybold Heraeus, Germany for the purpose of tritium testing, features heavy duty bearings and motor, metal body seals, and radiation resistant coatings on electrical components. No tritium related degradation of the pump has been observed after 12,000 hours of operation. A liquid nitrogen cooled trap, used to trap the outgassing products from the system, required periodic regeneration. The loop was vacuum baked and recharged with tritium every 14 to 63 days of tritium testing. A bearing failure at 7176 hours is believed to have resulted from excessive bearing loads imposed by high system pressures. Testing was resumed after the original oil lubricated steel bearings were replaced with grease lubricated ceramic bearings. Vibration monitoring of turbomolecular pump bearings can be effectively used to predict remaining bearing life and to schedule maintenance. The study is intended to continue until the pump has undergone one year of cumulative tritium service.
{"title":"Tritium effects on the performance of turbomolecular pumps","authors":"A. G. Heics, T. Shmayda, D. Muller","doi":"10.1109/FUSION.1993.518286","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518286","url":null,"abstract":"A special 150 L/s turbomolecular pump was used to pump pure tritium in closed loop mode over the course of six separate tritium runs. About 5.3 kCi of tritium was pumped during a cumulative period of 237 days. The pump, supplied by Leybold Heraeus, Germany for the purpose of tritium testing, features heavy duty bearings and motor, metal body seals, and radiation resistant coatings on electrical components. No tritium related degradation of the pump has been observed after 12,000 hours of operation. A liquid nitrogen cooled trap, used to trap the outgassing products from the system, required periodic regeneration. The loop was vacuum baked and recharged with tritium every 14 to 63 days of tritium testing. A bearing failure at 7176 hours is believed to have resulted from excessive bearing loads imposed by high system pressures. Testing was resumed after the original oil lubricated steel bearings were replaced with grease lubricated ceramic bearings. Vibration monitoring of turbomolecular pump bearings can be effectively used to predict remaining bearing life and to schedule maintenance. The study is intended to continue until the pump has undergone one year of cumulative tritium service.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114751101","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518524
J. Freidberg, A. Shajii, E. Chaniotakis, J. Mccarrick
This paper presents a theoretical model describing quench propagation in cable in conduit conductors (CICC) with an additional central flow channel. The central channel is used to enhance the flow capabilities in the conduit during steady state operation as well as during quench events. Such a system is the proposed design for certain conductors in the International Thermonuclear Experimental Reactor (ITER). Here, the additional channel is formed by a metal spring located at the center of the conduit. We describe the separate thermal evolution in both the cable bundle and the central channel; in particular, the mass, momentum and heat transfer due to flow between the cable bundle and the central channel are included in the model. Several simplifications are introduced which greatly reduce the complexity of the model without sacrificing accuracy. The resulting reduced model is solved both numerically and approximately analytically for ITER parameters.
{"title":"Modelling of quench in CICC with a central channel in the conduit","authors":"J. Freidberg, A. Shajii, E. Chaniotakis, J. Mccarrick","doi":"10.1109/FUSION.1993.518524","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518524","url":null,"abstract":"This paper presents a theoretical model describing quench propagation in cable in conduit conductors (CICC) with an additional central flow channel. The central channel is used to enhance the flow capabilities in the conduit during steady state operation as well as during quench events. Such a system is the proposed design for certain conductors in the International Thermonuclear Experimental Reactor (ITER). Here, the additional channel is formed by a metal spring located at the center of the conduit. We describe the separate thermal evolution in both the cable bundle and the central channel; in particular, the mass, momentum and heat transfer due to flow between the cable bundle and the central channel are included in the model. Several simplifications are introduced which greatly reduce the complexity of the model without sacrificing accuracy. The resulting reduced model is solved both numerically and approximately analytically for ITER parameters.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125897568","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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