Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518507
S.N. Golovato, W. Beck, P. Bonoli, M. Fridberg, M. Porkolab, Y. Takase
The Alcator C-Mod tokamak experiment relies on ICRF heating as the sole source of auxilliary power. Two different antenna designs have been produced and fabricated for the ICRF heating experiments. The first antenna has a single current strap, a two-layer Faraday shield, and is movable radially. One of these antennas was built, conditioned to high RF voltage on a test stand, and installed on C-Mod for the first experimental run. The engineering goals of the first experiments include measuring the plasma loading and testing the power handling capability of the antenna. Details of the conditioning and tuning procedures will be presented. The second antenna has two phasable current straps and a single layer Faraday shield. The predicted high disruption forces require that it be attached securely to the vacuum vessel wall at a fixed radial position. Two of these antennas have been constructed, one having a titanium carbide coated Faraday shield and the other a boron carbide coated shield. They are designed to couple up to 4 MW of source power from two FMIT transmitters. Details of the fabrication and test stand conditioning of these antennas will be presented.
{"title":"Antennas for ICRF heating in the Alcator C-Mod tokamak","authors":"S.N. Golovato, W. Beck, P. Bonoli, M. Fridberg, M. Porkolab, Y. Takase","doi":"10.1109/FUSION.1993.518507","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518507","url":null,"abstract":"The Alcator C-Mod tokamak experiment relies on ICRF heating as the sole source of auxilliary power. Two different antenna designs have been produced and fabricated for the ICRF heating experiments. The first antenna has a single current strap, a two-layer Faraday shield, and is movable radially. One of these antennas was built, conditioned to high RF voltage on a test stand, and installed on C-Mod for the first experimental run. The engineering goals of the first experiments include measuring the plasma loading and testing the power handling capability of the antenna. Details of the conditioning and tuning procedures will be presented. The second antenna has two phasable current straps and a single layer Faraday shield. The predicted high disruption forces require that it be attached securely to the vacuum vessel wall at a fixed radial position. Two of these antennas have been constructed, one having a titanium carbide coated Faraday shield and the other a boron carbide coated shield. They are designed to couple up to 4 MW of source power from two FMIT transmitters. Details of the fabrication and test stand conditioning of these antennas will be 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":"126544043","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.1993.518306
G. Miley, A. Satsangi, Y. Yamamoto, H. Nakashima, J. Javedani
Inertial Electrostatic Confinement (IEC) fusion is well-suited for the burning of advanced fuels, such as D-/sup 3/He. Small-scale experiments at the University of Illinois have produced encouraging results and are the basis for a conceptual design of a 25-MW D-/sup 3/He gridded IEC reactor. Viewed as a pilot plant, this reactor would be used to demonstrate a net power production and to study engineering problems related to high-voltage energy conversion.
{"title":"Conceptual design for a D-/sup 3/He IEC pilot plant","authors":"G. Miley, A. Satsangi, Y. Yamamoto, H. Nakashima, J. Javedani","doi":"10.1109/FUSION.1993.518306","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518306","url":null,"abstract":"Inertial Electrostatic Confinement (IEC) fusion is well-suited for the burning of advanced fuels, such as D-/sup 3/He. Small-scale experiments at the University of Illinois have produced encouraging results and are the basis for a conceptual design of a 25-MW D-/sup 3/He gridded IEC reactor. Viewed as a pilot plant, this reactor would be used to demonstrate a net power production and to study engineering problems related to high-voltage energy conversion.","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":"124428623","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.1993.518501
J.P. Smith, K. Schaubel, C. Baxi, G. Campbell, A. Hyatt, G. J. Laughon, M. Mahdavi, E. Reis, M. Schaffer, D. Sevier, R. Stambaugh, M. Menon
Phase two of a divertor cryocondensation pump, the Advanced Divertor Program, is now installed in the DIII-D tokamak at General Atomics and complements the phase one biasable ring electrode. The installation consists of a 10 m long cryocondensation pump located in the divertor baffle chamber to study plasma density control by pumping of the divertor. The design is a toroidally electrically continuous liquid helium-cooled panel with 1 m/sup 2/ of pumping surface. The helium panel is single point grounded to the nitrogen shield to minimize eddy currents. The nitrogen shield is toroidally continuous and grounded to the vacuum vessel in 24 locations to prevent voltage potentials from building up between the pump and vacuum vessel wall. A radiation/particle shield surrounds the nitrogen-cooled surface to minimize the heat load and prevent water molecules condensed on the nitrogen surface from being released by impact of energetic particles. Large currents (>5000 A) are driven in the helium and nitrogen panels during ohmic coil ramp up and during disruptions. The pump is designed to accommodate both the thermal and mechanical loads due to these currents. A feedthrough for the cryogens allows for both radial and vertical motion of the pump with respect to the vacuum vessel. Thermal performance measured on a prototype verified the analytical model and thermal design of the pump. Characterization tests of the installed pump show the pumping speed in deuterium is 42,000 l/sec for a pressure of 5 mTorr. Induction heating of the pump (at 300 W) resulted in no degradation of pumping speed. Plasma operations with the cryopump show a 60% lower density in H-mode.
{"title":"Installation and initial operation of the DIII-D advanced divertor cryocondensation pump","authors":"J.P. Smith, K. Schaubel, C. Baxi, G. Campbell, A. Hyatt, G. J. Laughon, M. Mahdavi, E. Reis, M. Schaffer, D. Sevier, R. Stambaugh, M. Menon","doi":"10.1109/FUSION.1993.518501","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518501","url":null,"abstract":"Phase two of a divertor cryocondensation pump, the Advanced Divertor Program, is now installed in the DIII-D tokamak at General Atomics and complements the phase one biasable ring electrode. The installation consists of a 10 m long cryocondensation pump located in the divertor baffle chamber to study plasma density control by pumping of the divertor. The design is a toroidally electrically continuous liquid helium-cooled panel with 1 m/sup 2/ of pumping surface. The helium panel is single point grounded to the nitrogen shield to minimize eddy currents. The nitrogen shield is toroidally continuous and grounded to the vacuum vessel in 24 locations to prevent voltage potentials from building up between the pump and vacuum vessel wall. A radiation/particle shield surrounds the nitrogen-cooled surface to minimize the heat load and prevent water molecules condensed on the nitrogen surface from being released by impact of energetic particles. Large currents (>5000 A) are driven in the helium and nitrogen panels during ohmic coil ramp up and during disruptions. The pump is designed to accommodate both the thermal and mechanical loads due to these currents. A feedthrough for the cryogens allows for both radial and vertical motion of the pump with respect to the vacuum vessel. Thermal performance measured on a prototype verified the analytical model and thermal design of the pump. Characterization tests of the installed pump show the pumping speed in deuterium is 42,000 l/sec for a pressure of 5 mTorr. Induction heating of the pump (at 300 W) resulted in no degradation of pumping speed. Plasma operations with the cryopump show a 60% lower density in H-mode.","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":"123492046","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.1993.518378
M.E. Oldaker, J. Lawson, K. Wright
The Gas Delivery System for the TFTR Neutral Beam (NB) Heating System has been replaced with a new system which permits injection of either hydrogen/deuterium or tritium into each of the twelve Long Pulse Ion Sources (LPIS). The new gas delivery system will provide throughputs up to 100 torr-l/s of deuterium or 50 torr-l/s of tritium to each of the LPISs. The desired gas can be selected and the throughput adjusted between pulses on an individual source basis. The throughput is maintained uniform to /spl plusmn/1% throughout the injection pulse by utilizing a closed loop controller which varies the voltage applied to a piezoceramic element within piezoelectric valve as a function of the measured pressure within individual local storage plenums. The capability to "condition" the LPISs with deuterium between tritium injection pulses and maximum operational flexibility is provided by installing independent deuterium and tritium subsystems on each LPIS, A programmable logic controller provides central control, interlocks and permissives for all twelve LPISs. This paper presents the final installation and the results from preoperational testing and initial operation of the Neutral Beam Deuterium-Tritium Gas Delivery System (NB D-T GDS).
{"title":"Installation, preoperational testing and initial operation of the TFTR neutral beam deuterium-tritium gas delivery system","authors":"M.E. Oldaker, J. Lawson, K. Wright","doi":"10.1109/FUSION.1993.518378","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518378","url":null,"abstract":"The Gas Delivery System for the TFTR Neutral Beam (NB) Heating System has been replaced with a new system which permits injection of either hydrogen/deuterium or tritium into each of the twelve Long Pulse Ion Sources (LPIS). The new gas delivery system will provide throughputs up to 100 torr-l/s of deuterium or 50 torr-l/s of tritium to each of the LPISs. The desired gas can be selected and the throughput adjusted between pulses on an individual source basis. The throughput is maintained uniform to /spl plusmn/1% throughout the injection pulse by utilizing a closed loop controller which varies the voltage applied to a piezoceramic element within piezoelectric valve as a function of the measured pressure within individual local storage plenums. The capability to \"condition\" the LPISs with deuterium between tritium injection pulses and maximum operational flexibility is provided by installing independent deuterium and tritium subsystems on each LPIS, A programmable logic controller provides central control, interlocks and permissives for all twelve LPISs. This paper presents the final installation and the results from preoperational testing and initial operation of the Neutral Beam Deuterium-Tritium Gas Delivery System (NB D-T GDS).","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":"121668110","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.1993.518310
H. Shimotohno, S. Kondo
The engineering feasibility of a D/sup 3/He fueled tokamak fusion power reactor is studied using a system code to obtain a consistent set of design variables and to optimize the reactor system. The study shows that the desirable ion temperature of the D/sup 3/He core plasma is in the range of 50-65 keV to obtain a system with reasonable power density. The fusion energy of a core plasma is lost mainly by transport, bremsstrahlung and synchrotron radiation, of which shares are strongly dependent on the confinement characteristics, namely, on Lawson parameter. Considering physics and engineering constraints, we propose a reactor system with the double null divertor in which we expect to recover a large fraction of fusion power transported. Economical consideration suggests that the desirable minimum plasma beta is around 0.21 and thermal heat flux on the first wall and the divertor wall is larger than /spl sim/3 MW/m/sup 2/.
{"title":"A system study on the engineering feasibility of D/sup 3/He tokamak fusion power reactor","authors":"H. Shimotohno, S. Kondo","doi":"10.1109/FUSION.1993.518310","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518310","url":null,"abstract":"The engineering feasibility of a D/sup 3/He fueled tokamak fusion power reactor is studied using a system code to obtain a consistent set of design variables and to optimize the reactor system. The study shows that the desirable ion temperature of the D/sup 3/He core plasma is in the range of 50-65 keV to obtain a system with reasonable power density. The fusion energy of a core plasma is lost mainly by transport, bremsstrahlung and synchrotron radiation, of which shares are strongly dependent on the confinement characteristics, namely, on Lawson parameter. Considering physics and engineering constraints, we propose a reactor system with the double null divertor in which we expect to recover a large fraction of fusion power transported. Economical consideration suggests that the desirable minimum plasma beta is around 0.21 and thermal heat flux on the first wall and the divertor wall is larger than /spl sim/3 MW/m/sup 2/.","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":"125516526","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.1993.518313
C. Annese, E. F. Watkins, E. Greenspan, W. F. Miller, J. Latkowski, J.D. Lee, P. Soran, M. Tobin
Minimum cost design concepts of the primary shield for the National Ignition laser fusion experimental Facility (NIF) are searched with the help of the optimization code SWAN. The computational method developed for this search involves incorporating the time dependence of the delayed photon field within effective delayed photon production cross sections. This method enables one to address the time-dependent problem using relatively simple, time-independent transport calculations, thus significantly simplifying the design process. A novel approach was used for the identification of the optimal combination of constituents that will minimize the shield cost; it involves the generation, with SWAN, of effectiveness functions for replacing materials on an equal cost basis. The minimum cost shield design concept was found to consist of a mixture of polyethylene and low cost, low activation materials such as SiC, with boron added near the shield boundaries.
{"title":"Optimization of the National Ignition Facility primary shield design","authors":"C. Annese, E. F. Watkins, E. Greenspan, W. F. Miller, J. Latkowski, J.D. Lee, P. Soran, M. Tobin","doi":"10.1109/FUSION.1993.518313","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518313","url":null,"abstract":"Minimum cost design concepts of the primary shield for the National Ignition laser fusion experimental Facility (NIF) are searched with the help of the optimization code SWAN. The computational method developed for this search involves incorporating the time dependence of the delayed photon field within effective delayed photon production cross sections. This method enables one to address the time-dependent problem using relatively simple, time-independent transport calculations, thus significantly simplifying the design process. A novel approach was used for the identification of the optimal combination of constituents that will minimize the shield cost; it involves the generation, with SWAN, of effectiveness functions for replacing materials on an equal cost basis. The minimum cost shield design concept was found to consist of a mixture of polyethylene and low cost, low activation materials such as SiC, with boron added near the shield boundaries.","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":"130424956","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.1993.518540
J. Rosenfeld, J. Lindemuth
Several types of porous media heat exchangers are being evaluated for use in cooled plasma-facing components. Monel/water thermosyphon heat pipes with a porous metal wick are being evaluated for use in Faraday shields. A subscale prototype has been fabricated, and tests at Oak Ridge National Laboratory are being planned. An advanced gyrotron microwave cavity is being developed which uses water cooling in a copper porous metal heat exchanger. Tests on a single-cell prototype demonstrated absorbed heat flux capability in excess of 60 MW/m/sup 2/. Porous metal heat exchangers with helium, water, or liquid metal coolants are being evaluated for divertor cooling. Tests on helium/copper porous metal heat exchanger demonstrated absorbed heat flux capability in excess of 15 MW/m/sup 2/. Applications, conceptual designs, fabricated hardware and test results are summarized.
{"title":"Evaluation of porous media heat exchangers for plasma facing components","authors":"J. Rosenfeld, J. Lindemuth","doi":"10.1109/FUSION.1993.518540","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518540","url":null,"abstract":"Several types of porous media heat exchangers are being evaluated for use in cooled plasma-facing components. Monel/water thermosyphon heat pipes with a porous metal wick are being evaluated for use in Faraday shields. A subscale prototype has been fabricated, and tests at Oak Ridge National Laboratory are being planned. An advanced gyrotron microwave cavity is being developed which uses water cooling in a copper porous metal heat exchanger. Tests on a single-cell prototype demonstrated absorbed heat flux capability in excess of 60 MW/m/sup 2/. Porous metal heat exchangers with helium, water, or liquid metal coolants are being evaluated for divertor cooling. Tests on helium/copper porous metal heat exchanger demonstrated absorbed heat flux capability in excess of 15 MW/m/sup 2/. Applications, conceptual designs, fabricated hardware and test results are summarized.","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":"116581018","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.1993.518389
D. Cepraga, A. Panini, D. Diamanti, G. Cambi, G. Cavallone, M. Costa, S. Boeriu
The reduction of hafnium content in blanket materials containing zirconium is costly. Therefore, a sensitivity analysis was performed to estimate the impact of hafnium content on nuclear properties of lithium metazirconate Li/sub 2/ZrO/sub 3/ and of Zr/sub 5/Pb/sub 3/, promising breeder-multiplier system candidate for fusion reactors. This paper summarises the results of extensive shielding neutronic analysis and transmutation-activation calculations aiming to evaluate the tritium breeding ratio, the inventories of various radio nuclides, and the surface /spl gamma/-dose rate, both for breeder and multiplier materials containing hafnium impurities. 1-D neutron transport fixed source analysis has been performed with the XSDRNPM code, using the 171-group VITAMIN-C cross section library, based on ENDF/B-V and JEF2.2 basic data. The activation calculations were performed with the ANITA code, using updated cross section and decay data libraries based on EAF-3 evaluation files. A comparison between Zr/sub 5/Pb/sub 3/ and beryllium multipliers in different design configurations including homogenised solid breeder mixtures of Li/sub 2/ZrO/sub 3/ and structural materials has been performed.
{"title":"Impact of hafnium content on nuclear performance of breeder and multiplier materials containing zirconium in fusion assemblies","authors":"D. Cepraga, A. Panini, D. Diamanti, G. Cambi, G. Cavallone, M. Costa, S. Boeriu","doi":"10.1109/FUSION.1993.518389","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518389","url":null,"abstract":"The reduction of hafnium content in blanket materials containing zirconium is costly. Therefore, a sensitivity analysis was performed to estimate the impact of hafnium content on nuclear properties of lithium metazirconate Li/sub 2/ZrO/sub 3/ and of Zr/sub 5/Pb/sub 3/, promising breeder-multiplier system candidate for fusion reactors. This paper summarises the results of extensive shielding neutronic analysis and transmutation-activation calculations aiming to evaluate the tritium breeding ratio, the inventories of various radio nuclides, and the surface /spl gamma/-dose rate, both for breeder and multiplier materials containing hafnium impurities. 1-D neutron transport fixed source analysis has been performed with the XSDRNPM code, using the 171-group VITAMIN-C cross section library, based on ENDF/B-V and JEF2.2 basic data. The activation calculations were performed with the ANITA code, using updated cross section and decay data libraries based on EAF-3 evaluation files. A comparison between Zr/sub 5/Pb/sub 3/ and beryllium multipliers in different design configurations including homogenised solid breeder mixtures of Li/sub 2/ZrO/sub 3/ and structural materials has been performed.","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":"120814284","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.1993.518318
J. Morrow-Jones, M. Firestone, S. Jardin, C. Kessel, T. Mau
Dynamical models of a tokamak can be used to augment on-line measurements, giving real-time estimates of the state of the plasma. We outline modeling of tokamak dynamics, and implementing the dynamics into a real-time filter. The real-time estimation of a current profile is constructed from simulated time-dependent measurements of PF coil currents and total plasma current.
{"title":"Use of tokamak dynamics models for digital filtering and control","authors":"J. Morrow-Jones, M. Firestone, S. Jardin, C. Kessel, T. Mau","doi":"10.1109/FUSION.1993.518318","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518318","url":null,"abstract":"Dynamical models of a tokamak can be used to augment on-line measurements, giving real-time estimates of the state of the plasma. We outline modeling of tokamak dynamics, and implementing the dynamics into a real-time filter. The real-time estimation of a current profile is constructed from simulated time-dependent measurements of PF coil currents and total plasma current.","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":"121043010","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.1993.518478
H. Khater, L. Wittenberg, M. Sawan
Environmental and safety analyses have been performed for SIRIUS-P, and its target factory and fuel reprocessing facilities. Both the C/C composite chamber and steel-reinforced concrete shield would easily qualify as Class A low level waste (LLW). Due to the high /sup 14/C activity, the Li/sub 2/O solid breeder and TiO/sub 2/ coolant would only qualify for Class C LLW. The radiological dose to the population in the vicinity of the reactor site due to the routine release of tritium is 0.56 mrem/yr. During a loss of coolant accident (LOCA) or loss of flow accident (LOFA), the whole body (WB) early dose at the site boundary (1 km) only amounts to 1.55 and 58.2 mrem for the chamber and shield, respectively. The WB early dose at the site boundary due to the Li/sub 2/O and TiO/sub 2/ are 93.5 /spl mu/rem and 93 mrem, respectively. A 100% release of the 156.3 g of tritium contained inside the reactor containment at any moment would produce a WB early dose on the order of 1.4 rem. Even though the target factory processes a total of 580,000 targets/day, the total tritium inventory along the production line is limited to only 285 g. The maximum WB early dose projected as a result of a severe accident involving the target factory of SIRIUS-P would be limited to 2.57 rem. In addition, a 100% release of the tritium contained in the fuel reprocessing facility would only result in a WB early dose of 640 mrem. The use of N-stamp nuclear grade components in SIRIUS-P can be avoided due to the low off-site doses.
{"title":"Environmental and safety assessment of the inertially confined direct drive laser fusion power reactor SIRIUS-P","authors":"H. Khater, L. Wittenberg, M. Sawan","doi":"10.1109/FUSION.1993.518478","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518478","url":null,"abstract":"Environmental and safety analyses have been performed for SIRIUS-P, and its target factory and fuel reprocessing facilities. Both the C/C composite chamber and steel-reinforced concrete shield would easily qualify as Class A low level waste (LLW). Due to the high /sup 14/C activity, the Li/sub 2/O solid breeder and TiO/sub 2/ coolant would only qualify for Class C LLW. The radiological dose to the population in the vicinity of the reactor site due to the routine release of tritium is 0.56 mrem/yr. During a loss of coolant accident (LOCA) or loss of flow accident (LOFA), the whole body (WB) early dose at the site boundary (1 km) only amounts to 1.55 and 58.2 mrem for the chamber and shield, respectively. The WB early dose at the site boundary due to the Li/sub 2/O and TiO/sub 2/ are 93.5 /spl mu/rem and 93 mrem, respectively. A 100% release of the 156.3 g of tritium contained inside the reactor containment at any moment would produce a WB early dose on the order of 1.4 rem. Even though the target factory processes a total of 580,000 targets/day, the total tritium inventory along the production line is limited to only 285 g. The maximum WB early dose projected as a result of a severe accident involving the target factory of SIRIUS-P would be limited to 2.57 rem. In addition, a 100% release of the tritium contained in the fuel reprocessing facility would only result in a WB early dose of 640 mrem. The use of N-stamp nuclear grade components in SIRIUS-P can be avoided due to the low off-site doses.","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":"129979929","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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