Pub Date : 2007-10-24DOI: 10.1080/10519990701562360
V. Kotenko, D. V. Kurilo, Yu. F. Sergeyev, Ye. D. Volkov
Numerical studies were carried out to show that a regime implied as the standard or basic regime in the l = 2 torsatrons and heliotrons with no additional longitudinal magnetic field coils can be realized in the l = 2 torsatron with the aforementioned coils. The regime is characterized by a shift in the closed magnetic surface configuration inwards in the torus, by the transformation of the spatial magnetic axis into the planar axis and by close to minimum values of the field ripple at the magnetic surfaces. For the simulation model used, the main parameters of the magnetic surface are given at several values of the additional longitudinal magnetic field.
{"title":"The basic regime characteristics of the l = 2 torsatron with additional longitudinal magnetic field coils","authors":"V. Kotenko, D. V. Kurilo, Yu. F. Sergeyev, Ye. D. Volkov","doi":"10.1080/10519990701562360","DOIUrl":"https://doi.org/10.1080/10519990701562360","url":null,"abstract":"Numerical studies were carried out to show that a regime implied as the standard or basic regime in the l = 2 torsatrons and heliotrons with no additional longitudinal magnetic field coils can be realized in the l = 2 torsatron with the aforementioned coils. The regime is characterized by a shift in the closed magnetic surface configuration inwards in the torus, by the transformation of the spatial magnetic axis into the planar axis and by close to minimum values of the field ripple at the magnetic surfaces. For the simulation model used, the main parameters of the magnetic surface are given at several values of the additional longitudinal magnetic field.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"36 1","pages":"253 - 262"},"PeriodicalIF":0.0,"publicationDate":"2007-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81086673","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 : 2007-10-24DOI: 10.1080/10519990701669512
A. Benmoussa, H. Sisabeur, Z. Harrache, A. Belasri
In this work, we investigated the effect of gas heating near the electrode of an excimer lamp. Gas heating occurs in the cathode region as ions are accelerated by the high electric field near the cathode. The gas temperature profile was calculated by solving the one-dimensional heat transport equation in the Joule heating approximation without including the effect of reflected particles. The inter-electrode gap was filled with xenon gas and measured 0.5 cm. The model predicts correctly the waveform of the temperature and illustrates the important role of the boundary conditions. The results obtained show clearly that the increase in the temperature in the sheath is limited by the increase in the ion current density in the cathode region due to the electron multiplication.
{"title":"Study of gas heating in a xenon glow discharge","authors":"A. Benmoussa, H. Sisabeur, Z. Harrache, A. Belasri","doi":"10.1080/10519990701669512","DOIUrl":"https://doi.org/10.1080/10519990701669512","url":null,"abstract":"In this work, we investigated the effect of gas heating near the electrode of an excimer lamp. Gas heating occurs in the cathode region as ions are accelerated by the high electric field near the cathode. The gas temperature profile was calculated by solving the one-dimensional heat transport equation in the Joule heating approximation without including the effect of reflected particles. The inter-electrode gap was filled with xenon gas and measured 0.5 cm. The model predicts correctly the waveform of the temperature and illustrates the important role of the boundary conditions. The results obtained show clearly that the increase in the temperature in the sheath is limited by the increase in the ion current density in the cathode region due to the electron multiplication.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"194 1","pages":"299 - 304"},"PeriodicalIF":0.0,"publicationDate":"2007-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83495973","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 : 2007-10-24DOI: 10.1080/10519990701674074
V. Kovalenko, Yu. G. Strebkov
One of the basic purposes of the International Thermonuclear Experimental Reactor (ITER) programme is to test the elements of test blanket modules (TBMs) of the demonstration fusion reactor DEMO. In the present paper the opportunities for extrapolation of the test results of a Russian demonstration fusion reactor DEMO ceramic helium-cooled blanket TBM in the ITER to the DEMO blanket parameters are considered. A possible research and development plan for the Russian ceramic helium-cooled TBM, the types of TBM for testing in the ITER, and the TBM test plan are presented.
{"title":"Test strategy of a Russian ceramic helium-cooled test blanket module in the International Thermonuclear Experimental Reactor","authors":"V. Kovalenko, Yu. G. Strebkov","doi":"10.1080/10519990701674074","DOIUrl":"https://doi.org/10.1080/10519990701674074","url":null,"abstract":"One of the basic purposes of the International Thermonuclear Experimental Reactor (ITER) programme is to test the elements of test blanket modules (TBMs) of the demonstration fusion reactor DEMO. In the present paper the opportunities for extrapolation of the test results of a Russian demonstration fusion reactor DEMO ceramic helium-cooled blanket TBM in the ITER to the DEMO blanket parameters are considered. A possible research and development plan for the Russian ceramic helium-cooled TBM, the types of TBM for testing in the ITER, and the TBM test plan are presented.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"13 1","pages":"305 - 315"},"PeriodicalIF":0.0,"publicationDate":"2007-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87146443","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 : 2007-07-19DOI: 10.1080/10519990701451002
E. Azizov, G. G. Gladush, N. Rodionov, V. Rodionova
Heat removal from the structural elements of a compact tokamak with warm coils intended for the transmutation of minor actinides is an especially critical problem. This paper is concerned with the thermal processes on the first-wall elements and the toroidal field windings, ignoring the most stressed component of JUST-T, i.e. the blanket. The analysis has shown that the peak load on the first-wall elements in JUST-T does not exceed a similar load in the International Thermonuclear Experimental Reactor tokamak. The two-dimensional numerical code for calculation of the current distribution density, Ohmic heating and heat transfer due to heat conductivity and extraction of heat by the water flux has been developed.
{"title":"On heat removal from the first-wall elements and warm toroidal field coils in the JUST-T tokamak","authors":"E. Azizov, G. G. Gladush, N. Rodionov, V. Rodionova","doi":"10.1080/10519990701451002","DOIUrl":"https://doi.org/10.1080/10519990701451002","url":null,"abstract":"Heat removal from the structural elements of a compact tokamak with warm coils intended for the transmutation of minor actinides is an especially critical problem. This paper is concerned with the thermal processes on the first-wall elements and the toroidal field windings, ignoring the most stressed component of JUST-T, i.e. the blanket. The analysis has shown that the peak load on the first-wall elements in JUST-T does not exceed a similar load in the International Thermonuclear Experimental Reactor tokamak. The two-dimensional numerical code for calculation of the current distribution density, Ohmic heating and heat transfer due to heat conductivity and extraction of heat by the water flux has been developed.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"48 48 1","pages":"241 - 252"},"PeriodicalIF":0.0,"publicationDate":"2007-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79318782","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 : 2007-07-19DOI: 10.1080/10519990701450632
O. Buzhinskij, V. Otroshchenko, W. West
The application of fine-grained recrystallized RGT graphite modified by titanium is considered. Thermally stabilized fine-grained RGT graphite modified by 10 at.% Ti has a density of 2.1 g cm−3, a heat conductivity of up to 800 W m−1 K−1, a low porosity and a small retention of hydrogen isotopes. The paper presents the results of full-scale tests on RGT graphite samples carried out in the T-10 and T-11M tokamaks in Russia, in the DIII-D tokamak in the USA, and in the MK-200 and PLADIS plasma facilities. The tested samples made from RGT graphite were found to be sufficiently stable, in particular, under irradiation for 600 ms by plasma with an electron temperature of 60 eV, an electron density of 2.5×1014 cm−3 and a power of the parallel heat flux of 5 kW cm−2, which are typical for the working modes and disruptions of plasma discharges in tokamaks.
探讨了钛改性细晶再结晶RGT石墨的应用。10 at改性的热稳定细晶RGT石墨。% Ti的密度为2.1 g cm−3，导热系数高达800 W m−1 K−1，孔隙率低，氢同位素保留量小。本文介绍了在俄罗斯的T-10和T-11M托卡马克、美国的DIII-D托卡马克以及MK-200和PLADIS等离子体设备上对RGT石墨样品进行的全尺寸测试结果。用RGT石墨制成的测试样品具有足够的稳定性，特别是在电子温度为60 eV、电子密度为2.5×1014 cm−3、平行热流功率为5 kW cm−2的等离子体照射600 ms的情况下，这是托卡马克等离子体放电的典型工作模式和中断。
{"title":"Application of RGT graphite in fusion devices","authors":"O. Buzhinskij, V. Otroshchenko, W. West","doi":"10.1080/10519990701450632","DOIUrl":"https://doi.org/10.1080/10519990701450632","url":null,"abstract":"The application of fine-grained recrystallized RGT graphite modified by titanium is considered. Thermally stabilized fine-grained RGT graphite modified by 10 at.% Ti has a density of 2.1 g cm−3, a heat conductivity of up to 800 W m−1 K−1, a low porosity and a small retention of hydrogen isotopes. The paper presents the results of full-scale tests on RGT graphite samples carried out in the T-10 and T-11M tokamaks in Russia, in the DIII-D tokamak in the USA, and in the MK-200 and PLADIS plasma facilities. The tested samples made from RGT graphite were found to be sufficiently stable, in particular, under irradiation for 600 ms by plasma with an electron temperature of 60 eV, an electron density of 2.5×1014 cm−3 and a power of the parallel heat flux of 5 kW cm−2, which are typical for the working modes and disruptions of plasma discharges in tokamaks.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"37 1","pages":"201 - 208"},"PeriodicalIF":0.0,"publicationDate":"2007-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85325478","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 : 2007-07-19DOI: 10.1080/10519990701451085
A. Dnestrovskij, V. Lukash, R. Khayrutdinov
The free-boundary equilibrium time evolution code DINA is used for calculations of the current ramp-up scenarios in the Component Test Facility (CTF) machine based on the concept of low-aspect-ratio tokamaks. The current ramp-up and subsequent support of the steady phase operation in the CTF is realized by the current driven by the neutral beam (NB), the bootstrap current and the current produced owing to the variation in the poloidal field from the control coils. The problems of seeded current formation are discussed. The self-consistent scenario for current ramp-up for the CTF project is worked out. The steady-state regime parameters are analysed. NB power absorption and current drive calculations are carried out by the DINA code tested on different devices.
{"title":"Calculations of current ramp-up scenarios for the Component Test Facility tokamak using the DINA code","authors":"A. Dnestrovskij, V. Lukash, R. Khayrutdinov","doi":"10.1080/10519990701451085","DOIUrl":"https://doi.org/10.1080/10519990701451085","url":null,"abstract":"The free-boundary equilibrium time evolution code DINA is used for calculations of the current ramp-up scenarios in the Component Test Facility (CTF) machine based on the concept of low-aspect-ratio tokamaks. The current ramp-up and subsequent support of the steady phase operation in the CTF is realized by the current driven by the neutral beam (NB), the bootstrap current and the current produced owing to the variation in the poloidal field from the control coils. The problems of seeded current formation are discussed. The self-consistent scenario for current ramp-up for the CTF project is worked out. The steady-state regime parameters are analysed. NB power absorption and current drive calculations are carried out by the DINA code tested on different devices.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"10 5 1","pages":"209 - 218"},"PeriodicalIF":0.0,"publicationDate":"2007-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82877244","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 : 2007-07-19DOI: 10.1080/10519990701450616
A. Barkalov, G. G. Gladush
In this paper, the stage of plasma column formation in the JUST-T tokamak is considered on the basis of a homogeneous model. The ionization, charge exchange, equilibration of temperatures between electrons and ions, radiation losses, Ohmic heating and additional heating are taken into account. With the use of electron cyclotron resonance heating, the breakdown voltage U br decreases from 30 to 16 V for P ECRH=300 kW and p=2 mPa. The additional ion cyclotron resonance heating is less effective; U br=20 V. The optimal conditions for overcoming the ionization radiation barrier were obtained for the JUST-T tokamak.
{"title":"Numerical simulation of the initial stage of the plasma column formation in JUST-T tokamak","authors":"A. Barkalov, G. G. Gladush","doi":"10.1080/10519990701450616","DOIUrl":"https://doi.org/10.1080/10519990701450616","url":null,"abstract":"In this paper, the stage of plasma column formation in the JUST-T tokamak is considered on the basis of a homogeneous model. The ionization, charge exchange, equilibration of temperatures between electrons and ions, radiation losses, Ohmic heating and additional heating are taken into account. With the use of electron cyclotron resonance heating, the breakdown voltage U br decreases from 30 to 16 V for P ECRH=300 kW and p=2 mPa. The additional ion cyclotron resonance heating is less effective; U br=20 V. The optimal conditions for overcoming the ionization radiation barrier were obtained for the JUST-T tokamak.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"1 1","pages":"185 - 192"},"PeriodicalIF":0.0,"publicationDate":"2007-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73292515","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 : 2007-07-19DOI: 10.1080/10519990701450657
L. Begrambekov, O. Buzhinskij
The physical aspects of graphites, carbon-fibre composites (CFC) and boron–carbon coatings as plasma-facing materials are considered. The advantages and disadvantages of their applications in tokamaks are analysed. A crystalline boron carbide coating is shown to have excellent erosion characteristics, low hydrogen retention capacity and high resistance under high-energy loads; these properties of boron carbide differ significantly from those of graphite. In modern fusion devices, crystalline boron carbide deposited on graphites with a high thermal conductivity can be used successfully as a plasma-facing material. A new technique suitable for deposition of boron carbide coating in the plasmas of modern tokamaks is described. It is concluded that a thick renewable boron carbide coating can be used successfully under the conditions of the first wall of the International Thermonuclear Experimental Reactor.
{"title":"Features and advantages of boron carbide as a protective coating of the tokamak first wall","authors":"L. Begrambekov, O. Buzhinskij","doi":"10.1080/10519990701450657","DOIUrl":"https://doi.org/10.1080/10519990701450657","url":null,"abstract":"The physical aspects of graphites, carbon-fibre composites (CFC) and boron–carbon coatings as plasma-facing materials are considered. The advantages and disadvantages of their applications in tokamaks are analysed. A crystalline boron carbide coating is shown to have excellent erosion characteristics, low hydrogen retention capacity and high resistance under high-energy loads; these properties of boron carbide differ significantly from those of graphite. In modern fusion devices, crystalline boron carbide deposited on graphites with a high thermal conductivity can be used successfully as a plasma-facing material. A new technique suitable for deposition of boron carbide coating in the plasmas of modern tokamaks is described. It is concluded that a thick renewable boron carbide coating can be used successfully under the conditions of the first wall of the International Thermonuclear Experimental Reactor.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"21 1","pages":"193 - 199"},"PeriodicalIF":0.0,"publicationDate":"2007-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72950861","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 : 2007-07-19DOI: 10.1080/10519990701451119
V. Kotov
The determinant role of raw uranium usage for atomic power production in the twenty-first century is shown. The limitations of modern reactors are the low fission materials breeding in thermal reactors and the large amount of uranium needed for start-up of fast reactors. Thermal reactor breeding can be increased under the following conditions: a decrease in the neutron losses in structural materials, leakage and operative control, and the use of a closed balanced fuel cycle. In this case, some of the raw 235U in the fuel at the start of the campaign decreases. Replacement of the raw 235U by 233U produced from thorium increases the raw usage up to 100%. The yield of 233U in a fission reactor decreases the burning-out. Production of 233U by a tokamak-based volumetric neutron source (VNS) removes the burning-out limitation. In this case, the energy consumed by the VNS is much less than the gain in the fission reactor energy production.
{"title":"Application of a volumetric neutron source to enhance the use of raw materials in nuclear power from thermal reactors","authors":"V. Kotov","doi":"10.1080/10519990701451119","DOIUrl":"https://doi.org/10.1080/10519990701451119","url":null,"abstract":"The determinant role of raw uranium usage for atomic power production in the twenty-first century is shown. The limitations of modern reactors are the low fission materials breeding in thermal reactors and the large amount of uranium needed for start-up of fast reactors. Thermal reactor breeding can be increased under the following conditions: a decrease in the neutron losses in structural materials, leakage and operative control, and the use of a closed balanced fuel cycle. In this case, some of the raw 235U in the fuel at the start of the campaign decreases. Replacement of the raw 235U by 233U produced from thorium increases the raw usage up to 100%. The yield of 233U in a fission reactor decreases the burning-out. Production of 233U by a tokamak-based volumetric neutron source (VNS) removes the burning-out limitation. In this case, the energy consumed by the VNS is much less than the gain in the fission reactor energy production.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"18 1","pages":"219 - 224"},"PeriodicalIF":0.0,"publicationDate":"2007-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74709121","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 : 2007-07-19DOI: 10.1080/10519990701467024
A. Mineev
Problems of the plasma scenario realization in a volumetric neutron source of the tokamak type are discussed in the paper, namely the generation of the fore-current followed by the non-inductive plasma current ramp-up up to its basic value.
{"title":"Non-inductive plasma operation scenario in tokamaks—possibilities for realization","authors":"A. Mineev","doi":"10.1080/10519990701467024","DOIUrl":"https://doi.org/10.1080/10519990701467024","url":null,"abstract":"Problems of the plasma scenario realization in a volumetric neutron source of the tokamak type are discussed in the paper, namely the generation of the fore-current followed by the non-inductive plasma current ramp-up up to its basic value.","PeriodicalId":54600,"journal":{"name":"Plasma Devices and Operations","volume":"183 1","pages":"225 - 240"},"PeriodicalIF":0.0,"publicationDate":"2007-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85077044","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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